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{{#Wiki_filter:ATTACHMENT 1 VOLUME 6 KEWAUNEE POWER STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.1 REACTIVITY CONTROL SYSTEMS  
{{#Wiki_filter:Attachment 1, Volume 6, Rev. 0, Page 1 of 213 ATTACHMENT 1 VOLUME 6 KEWAUNEE POWER STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.1 REACTIVITY CONTROL SYSTEMS Revision 0 Attachment 1, Volume 6, Rev. 0, Page 1 of 213


Revision 0  
Attachment 1, Volume 6, Rev. 0, Page 2 of 213 LIST OF ATTACHMENTS
: 1. ITS 3.1.1
: 2. ITS 3.1.2
: 3. ITS 3.1.3
: 4. ITS 3.1.4
: 5. ITS 3.1.5
: 6. ITS 3.1.6
: 7. ITS 3.1.7
: 8. ITS 3.1.8
: 9. Relocated/Deleted Current Technical Specifications (CTS)
Attachment 1, Volume 6, Rev. 0, Page 2 of 213
, Volume 6, Rev. 0, Page 3 of 213 ATTACHMENT 1 ITS 3.1.1, SHUTDOWN MARGIN , Volume 6, Rev. 0, Page 3 of 213


LIST OF ATTACHMENTS
Attachment 1, Volume 6, Rev. 0, Page 4 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
: 1. ITS 3.1.1
Attachment 1, Volume 6, Rev. 0, Page 4 of 213
: 2. ITS 3.1.2
: 3. ITS 3.1.3
: 4. ITS 3.1.4
: 5. ITS 3.1.5
: 6. ITS 3.1.6
: 7. ITS 3.1.7
: 8. ITS 3.1.8
: 9. Relocated/Deleted Current Technical Specifications (CTS)  


ATTACHMENT 1 ITS 3.1.1, SHUTDOWN MARGIN Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
Attachment 1, Volume 6, Rev. 0, Page 5 of 213 ITS                                                              A01                                     ITS 3.1.1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS APPLICABILITY Applies to the limits on core fission power distributions and to the limits on control rod operations.
A01 ITS 3.1.1 ITS 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS APPLICABILITY Applies to the limits on core fission power distributions and to the limits on control rod operations.
OBJECTIVE To ensure: 1) core subcriticality after reactor trip, 2) acceptable core power distribution during power operation in order to maintain fuel integrity in normal operation transients associated with faults of moderate frequency, supplemented by automatic protection and by administrative procedures, and to maintain the design basis initial conditions for limiting faults, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.
OBJECTIVE To ensure: 1) core subcriticality after reactor trip, 2) acceptable core power distribution during power operation in order to maintain fuel integrity in normal operation transients  
 
associated with faults of moderate frequency, supplemented by automatic protection and by administrative procedures, and to maintain the design basis initial conditions for limiting faults, and 3) limited potential reactivity insertions caused by hypothetical control rod  
 
ejection.
SPECIFICATION
SPECIFICATION
: a. Shutdown Reactivity  
: a. Shutdown Reactivity                                                                                 A02 Applicability      When the reactor is subcritical prior to reactor startup, the SHUTDOWN MARGIN shall be at LCO 3.1.1          least that as specified in the COLR Add proposed ACTION A                    M01
 
: b. Power Distribution Limits See ITS 3.2.1 and
When the reactor is subcritical prior to reactor startup, the SHUTDOWN MARGIN shall be at  
: 1. At all times, except during Low Power Physics Tests, the hot channel factors defined in           3.2.2 the basis must meet the following limits:
 
See ITS A. FQN(Z) Limits shall be as specified in the COLR.                                               3.2.1 B. FHN Limits shall be as specified in the COLR.
least that as specified in the COLR  
: 2. If FHN not within limits:
: b. Power Distribution Limits  
A. Perform the following:
: 1. At all times, except during Low Power Physics Tests, the hot channel factors defined in the basis must meet the following limits:
: i. Within 4 hours either, restore FHN to within its limit or reduce thermal power to less than 50% of RATED POWER.                                                           See ITS 3.2.2 ii. Reduce the Power Range Neutron Flux-High Trip Setpoint to  55% of RATED POWER within 72 hours.
A. F Q N (Z) Limits shall be as specified in the COLR.
iii. Verify FHN within limits within 24 hours.
B. FH N Limits shall be as specified in the COLR.  
: 2. If FH N not within limits:
A. Perform the following:  
: i. Within 4 hours either, restore FH N to within its limit or reduce thermal power to less than 50% of RATED POWER.
ii. Reduce the Power Range Neutron Flux-High Trip Setpoint to  55% of RATED POWER within 72 hours.
iii. Verify FH N within limits within 24 hours.
B. If the actions of TS 3.10.b.2.A are not completed within the specified time, then reduce thermal power to  5% of rated power within the next 6 hours.
B. If the actions of TS 3.10.b.2.A are not completed within the specified time, then reduce thermal power to  5% of rated power within the next 6 hours.
A pplicabilit y A02M01Add proposed ACTION A LCO 3.1.1 See ITS 3.2.1 and 3.2.2 See ITS 3.2.2 See ITS 3.2.1  Amendment No. 167 TS 3.10-1 04/04/2003 Page 1 of 3 A01ITS ITS 3.1.1 4.1 OPERATIONAL SAFETY REVIEW APPLICABILITY Applies to items directly related to safety limits and LIMITING CONDITIONS FOR OPERATION.
Amendment No. 167 TS 3.10-1                                     04/04/2003 Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 5 of 213
OBJECTIVE  To assure that instrumentation shall be checked, tested, and calibrated, and that equipment and sampling tests shall be conducted at sufficiently frequent intervals to


ensure safe operation.
Attachment 1, Volume 6, Rev. 0, Page 6 of 213 ITS                                                      A01                                    ITS 3.1.1 4.1 OPERATIONAL SAFETY REVIEW APPLICABILITY Applies to items directly related to safety limits and LIMITING CONDITIONS FOR OPERATION.
OBJECTIVE To assure that instrumentation shall be checked, tested, and calibrated, and that equipment and sampling tests shall be conducted at sufficiently frequent intervals to ensure safe operation.
SPECIFICATION
SPECIFICATION
: a. Calibration, testing, and checking of protective instrumentation channels and testing of logic channels shall be performed as specified in Table TS 4.1-1.  
: a. Calibration, testing, and checking of protective instrumentation channels and testing of   See other ITS logic channels shall be performed as specified in Table TS 4.1-1.
: b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and TS 4.1-3.  
SR 3.1.1.1    b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and TS 4.1-3.
: c. Deleted  
: c. Deleted
: d. Deleted  
: d. Deleted
: e. Deleted See otherITS SR 3.1.1.1 Amendment No. 119 TS 4.1-1 04/18/95 Page 2 of 3 A01 ITS 3.1.1 ITS  TABLE TS 4.1-2 MINIMUM FREQUENCIES FOR SAMPLING TESTS Amendment No. 119 Page 1 of 2 04/18/95  SAMPLING TESTS TEST FREQUENCY  1. Reactor Coolant Samples a. Gross Radioactivity Determination (excluding tritium) 5/week (1) b. DOSE EQUIVALENT I-131 Concentration 1/14 days (2) c. Tritium activity Monthly d. Chemistry (Cl, F, O 2)(3)  3/week (4)e. Determination 1/6 months (5) f. RCS isotopic analysis for Iodine Once per 4 hours in accordance with TS 3.1.c.2.C. 2. Reactor Coolant Boron (6) Boron Concentration (3) (3) 2/week 2/week (1) Maximum time between tests is 3 days.
: e. Deleted Amendment No. 119 TS 4.1-1                                       04/18/95 Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 6 of 213
(2) Sample required only when in the OPERATING MODE.
(3) Test required in all plant modes.
(4) Maximum time between tests is 4 days.
(5) Sample after a minimum of 2 EFPD and 20 days of OPERATING MODE operation have elapsed since the reactor was last subcritical f or    48 hours.
(6) A reactor coolant boron concentration sample does not have to be taken when the core is completely unloaded. Add proposed SR 3.1.1.1 See ITS 3.4.16 See ITS 3.4.16 M02 See CTS 3.1.e A02 L01 See CTS 3.1.e See ITS 3.9.1 M02 LA01 Page 3 of 3 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM)
ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
A02 CTS 3.10.a states that when the reactor is subcritical prior to startup, the SHUTDOWN MARGIN (SDM) shall be at least that as specified in the COLR. CTS Table 4.1-2 Sampling Test 2 footnote (3) states that the boron concentration test is required in all plant modes. ITS 3.1.1 is applicable in MODE 2 with k eff < 1.0 and in MODES 3, 4, and 5. This changes the CTS by specifically stating that the applicability is in MODE 2 with k eff < 1.0 and in MODES 3, 4, and 5. The change in Applicability for MODE 1 and MODE 2 with k eff  1.0 is discussed in DOC L01. In MODE 6, the reactor head is detensioned or removed, so a reactor startup cannot occur. In addition, ITS 3.9.1 provides the MODE 6 boron concentration limits.
This change is acceptable because the applicability has not changed. This change results in a format change only to comply with the ISTS presentation of the Applicability. The purpose of CTS Table 4.1-2 Sampling Test 2 is to verify that the boron concentration is within limits to satisfy the SDM. In MODE 2 with


k eff < 1.0 and in MODES 3, 4, and 5 the boron concentration is taken into consideration as part of the reactivity balance calculation required in ITS SR 3.1.1.1. The deletion of the MODE 1 and MODE 2 with k eff  > 1.0 is discussed in DOC L01 and the MODE 6 requirement is covered in ITS 3.9.1.
ITS                                                                                  A01                                                                  ITS 3.1.1 TABLE TS 4.1-2 See CTS                                                                                                                              See ITS 3.1.e                                  MINIMUM FREQUENCIES FOR SAMPLING TESTS                                                      3.4.16 SAMPLING TESTS                                        TEST                                              FREQUENCY (1)
Therefore, the change is acceptable because the boron concentration  
: 1. Reactor Coolant            a. Gross Radioactivity Determination (excluding tritium)  5/week Samples                    b. DOSE EQUIVALENT I-131 Concentration                    1/14 days(2)
: c. Tritium activity                                        Monthly
: d. Chemistry (Cl, F, O2)(3)                                3/week(4)
: e. Determination                                        1/6 months(5)
: f. RCS isotopic analysis for Iodine                        Once per 4 hours in accordance with TS 3.1.c.2.C.
: 2. Reactor Coolant Boron(6)  Boron Concentration(3)                                    2/week M02 LA01 Add proposed SR 3.1.1.1     M02 , Volume 6, Rev. 0, Page 7 of 213                                                                                                                                                                Attachment 1, Volume 6, Rev. 0, Page 7 of 213 (1)                                                                                                                                          See ITS Maximum time between tests is 3 days.                                                                                        3.4.16 (2)                                                                                                        L01 Sample required only when in the OPERATING MODE.                                                                        A02 (3)                                                                                                                            See CTS Test required in all plant modes.                                                                                            3.1.e (4)
Maximum time between tests is 4 days.
(5)
Sample after a minimum of 2 EFPD and 20 days of OPERATING MODE operation have elapsed since the reactor was last subcritical for 48 hours.
(6)
A reactor coolant boron concentration sample does not have to be taken when the core is completely unloaded. See ITS 3.9.1  Amendment No. 119 Page 1 of 2                                                  04/18/95 Page 3 of 3


requirements have not changed. This change is designated as an administrative change since it does not result in any technical changes to the CTS.
Attachment 1, Volume 6, Rev. 0, Page 8 of 213 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM)
 
ADMINISTRATIVE CHANGES A01  In the conversion of the Kewaunee Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
MORE RESTRICTIVE CHANGES
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
 
A02  CTS 3.10.a states that when the reactor is subcritical prior to startup, the SHUTDOWN MARGIN (SDM) shall be at least that as specified in the COLR.
M01 CTS 3.10.a does not supply any explicit time to restore the SDM to within limits when the SDM is not within its limit. As a result, LCO 3.0.c would be entered, which requires action to be initiated within 1 hour and to place the unit in HOT  
CTS Table 4.1-2 Sampling Test 2 footnote (3) states that the boron concentration test is required in all plant modes. ITS 3.1.1 is applicable in MODE 2 with keff
 
    < 1.0 and in MODES 3, 4, and 5. This changes the CTS by specifically stating that the applicability is in MODE 2 with keff < 1.0 and in MODES 3, 4, and 5. The change in Applicability for MODE 1 and MODE 2 with keff  1.0 is discussed in DOC L01. In MODE 6, the reactor head is detensioned or removed, so a reactor startup cannot occur. In addition, ITS 3.9.1 provides the MODE 6 boron concentration limits.
STANDBY (equivalent to ITS MODE 2) within the next 6 hours, in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours, and in COLD SHUTDOWN (equivalent to ITS MODE 5) within the subsequent 36 hours.
This change is acceptable because the applicability has not changed. This change results in a format change only to comply with the ISTS presentation of the Applicability. The purpose of CTS Table 4.1-2 Sampling Test 2 is to verify that the boron concentration is within limits to satisfy the SDM. In MODE 2 with keff < 1.0 and in MODES 3, 4, and 5 the boron concentration is taken into consideration as part of the reactivity balance calculation required in ITS SR 3.1.1.1. The deletion of the MODE 1 and MODE 2 with keff > 1.0 is discussed in DOC L01 and the MODE 6 requirement is covered in ITS 3.9.1.
Under similar conditions, ITS 3.1.1 provides 15 minutes to initiate boration to restore SDM to within limits. This changes the CTS by providing a maximum time limit of 15 minutes to initiate boration to restore SDM to within limits prior to  
Therefore, the change is acceptable because the boron concentration requirements have not changed. This change is designated as an administrative change since it does not result in any technical changes to the CTS.
 
MORE RESTRICTIVE CHANGES M01 CTS 3.10.a does not supply any explicit time to restore the SDM to within limits when the SDM is not within its limit. As a result, LCO 3.0.c would be entered, which requires action to be initiated within 1 hour and to place the unit in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours, in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours, and in COLD SHUTDOWN (equivalent to ITS MODE 5) within the subsequent 36 hours.
entering ITS LCO 3.0.3.  
Under similar conditions, ITS 3.1.1 provides 15 minutes to initiate boration to restore SDM to within limits. This changes the CTS by providing a maximum time limit of 15 minutes to initiate boration to restore SDM to within limits prior to entering ITS LCO 3.0.3.
 
The purpose of CTS 3.10.a SDM requirements is to provide sufficient reactivity margin to ensure that fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). This change is acceptable Kewaunee Power Station                 Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 8 of 213
The purpose of CTS 3.10.a SDM requirements is to provide sufficient reactivity margin to ensure that fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). This change is acceptable Kewaunee Power Station Page 1 of 3 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) because the Completion Time is consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the allowed Completion Time. The ITS 3.1.1 Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. In addition, the ITS Bases for the ACTION states that boration must be initiated promptly. The current action (LCO 3.0.c) allows up to 1 hour to start adding negative reactivity. The proposed action requires negative reactivity addition to begin in 15 minutes, and to continue until SDM is restored. This change is designated as more restrictive because a shorter time is allowed in ITS to restore parameters to within the LCO limits than was allowed in the CTS.
 
M02 CTS 3.10.a does not provide any Surveillance Requirements for verifying that the SDM is within the limits specified in the COLR. CTS Table 4.1-2 provides one parameter of the SDM in the boron concentration test which is performed two times per week. ITS SR 3.1.1.1 requires verifying that the SDM is within the limits specified in the COLR every 24 hours. This changes the CTS by adding a new Surveillance Requirement to verify the SDM.
 
This change is acceptable because the added Surveillance Requirement proves that the SDM requirements are being met. Furthermore, the boron concentration test is a portion of the reactivity effects that are necessary to perform a reactivity balance calculation but is not included in the reactivity balance calculation. In addition to the RCS boron concentration, the control bank position, RCS average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, samarium concentration and isothermal temperature must be taken into account in the performance of the reactivity balance calculation. This change is designated as more restrictive because a new Surveillance Requirement is being added that was not specified in the CTS.


RELOCATED SPECIFICATIONS None  
Attachment 1, Volume 6, Rev. 0, Page 9 of 213 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) because the Completion Time is consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the allowed Completion Time. The ITS 3.1.1 Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. In addition, the ITS Bases for the ACTION states that boration must be initiated promptly. The current action (LCO 3.0.c) allows up to 1 hour to start adding negative reactivity. The proposed action requires negative reactivity addition to begin in 15 minutes, and to continue until SDM is restored. This change is designated as more restrictive because a shorter time is allowed in ITS to restore parameters to within the LCO limits than was allowed in the CTS.
M02  CTS 3.10.a does not provide any Surveillance Requirements for verifying that the SDM is within the limits specified in the COLR. CTS Table 4.1-2 provides one parameter of the SDM in the boron concentration test which is performed two times per week. ITS SR 3.1.1.1 requires verifying that the SDM is within the limits specified in the COLR every 24 hours. This changes the CTS by adding a new Surveillance Requirement to verify the SDM.
This change is acceptable because the added Surveillance Requirement proves that the SDM requirements are being met. Furthermore, the boron concentration test is a portion of the reactivity effects that are necessary to perform a reactivity balance calculation but is not included in the reactivity balance calculation. In addition to the RCS boron concentration, the control bank position, RCS average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, samarium concentration and isothermal temperature must be taken into account in the performance of the reactivity balance calculation. This change is designated as more restrictive because a new Surveillance Requirement is being added that was not specified in the CTS.
RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS Table 4.1-2 Sampling Test 2 requires sampling tests of RCS boron concentration. ITS 3.1.1 does not have a specific requirement to test RCS boron concentration. This changes the CTS by removing the details of the RCS boron concentration test to the Bases.
The removal of these details for performing Surveillance Requirement from the Technical Specification is acceptable since this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. In ITS 3.1.1, RCS boron concentration is a parameter, in addition to control bank position, RCS average temperature, fuel Kewaunee Power Station                  Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 9 of 213


REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)  CTS Table 4.1-2 Sampling Test 2 requires sampling tests of RCS boron concentration. ITS 3.1.1 does not have a specific requirement to test RCS boron concentration. This changes the CTS by removing the details of the RCS boron concentration test to the Bases.
Attachment 1, Volume 6, Rev. 0, Page 10 of 213 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) burnup based on gross thermal energy generation, xenon concentration, samarium concentration and isothermal temperature, used in performing the reactivity balance calculation. Therefore, the details of how SDM is calculated do not need to appear in the Specification in order for the requirement to apply.
The removal of these details for performing Surveillance Requirement from the Technical Specification is acceptable since this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. In ITS 3.1.1, RCS boron concentration is a parameter, in addition to control bank position, RCS average temperature, fuel Kewaunee Power Station Page 2 of 3 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM)
Kewaunee Power Station Page 3 of 3 burnup based on gross thermal energy generation, xenon concentration, samarium concentration and isothermal temperature, used in performing the reactivity balance calculation. Therefore, the details of how SDM is calculated do not need to appear in the Specification in order for the requirement to apply.
Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specifications Bases Control Program in Chapter 5. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.
Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specifications Bases Control Program in Chapter 5. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.
LESS RESTRICTIVE CHANGES
LESS RESTRICTIVE CHANGES L01   (Category 5 - Deletion of Surveillance Requirement) CTS Table TS 4.1-2 Sampling Test 2, footnote (3) states that the boron concentration test is required in all plant modes. ITS 3.1.1 is applicable in MODE 2 with keff < 1.0 and in MODES 3, 4, and 5. This changes CTS by not requiring the boron concentration test in MODE 1 and MODE 2 with keff 1.0. ITS 3.9.1 provides the MODE 6 boron concentration limits.
 
The purpose of CTS Table TS 4.1-2, Sampling Test 2 is to verify the boron concentration. When the reactor is critical (MODE 1 and MODE 2 with keff 1.0),
L01 (Category 5 - Deletion of Surveillance Requirement)
SDM is verified by ensuring that the control rods are within the insertion limits (see ITS 3.1.5 and ITS 3.1.6). If the control banks are not within their insertion limits in MODE 1 and MODE 2 with keff 1.0, ITS 3.1.6, Required Action A.1.1 and B.1.1 require the SDM to be verified within limits. This requires performing a reactivity balance calculation considering Reactor Coolant System (RCS) boron concentration, control bank position, RCS average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, samarium concentration and isothermal temperature. Therefore, the boron concentration sampling test is not needed in MODE 1 and MODE 2 with keff 1.0. This change has been designated as less restrictive because a Surveillance that was required in CTS is no longer required in ITS.
CTS Table TS 4.1-2 Sampling Test 2, footnote (3) states that the boron concentration test is required in all plant modes. ITS 3.1.1 is applicable in MODE 2 with k eff < 1.0 and in MODES 3, 4, and 5. This changes CTS by not requiring the boron concentration test in MODE 1 and MODE 2 with k eff 1.0. ITS 3.9.1 provides the MODE 6 boron concentration limits.  
Kewaunee Power Station                Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 10 of 213
 
The purpose of CTS Table TS 4.1-2, Sampling Test 2 is to verify the boron concentration. When the reactor is critical (MODE 1 and MODE 2 with k eff 1.0), SDM is verified by ensuring that the control rods are within the insertion limits (see ITS 3.1.5 and ITS 3.1.6). If the control banks are not within their insertion limits in MODE 1 and MODE 2 with k eff 1.0, ITS 3.1.6, Required Action A.1.1 and B.1.1 require the SDM to be verified within limits. This requires performing a reactivity balance calculation considering Reactor Coolant System (RCS) boron concentration, control bank position, RCS average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, samarium concentration and isothermal temperature. Therefore, the boron concentration sampling test is not needed in MODE 1 and MODE 2 with k eff 1.0. This change has been designated as less restrictive because a Surveillance that was required in CTS is no longer required in ITS.
Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
SDM 3.1.1  WOG STS 3.1.1-1 Rev. 3.0, 03/31/04  CTS 3.1  REACTIVITY CONTROL SYSTEMS
 
====3.1.1 SHUTDOWN====
MARGIN (SDM)


LCO  3.1.1 SDM shall be within the limits specified in the COLR.
Attachment 1, Volume 6, Rev. 0, Page 11 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
3.10.a 
Attachment 1, Volume 6, Rev. 0, Page 11 of 213


APPLICABILITY: MODE 2 with k eff < 1.0, 3.10.a MODES 3, 4, and 5.  
Attachment 1, Volume 6, Rev. 0, Page 12 of 213 CTS                                                                                              SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1    SHUTDOWN MARGIN (SDM) 3.10.a  LCO 3.1.1            SDM shall be within the limits specified in the COLR.
3.10.a  APPLICABILITY:       MODE 2 with keff < 1.0, MODES 3, 4, and 5.
ACTIONS CONDITION                      REQUIRED ACTION                  COMPLETION TIME DOC M01  A. SDM not within limits. A.1      Initiate boration to restore    15 minutes SDM to within limits.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY DOC M02  SR 3.1.1.1      Verify SDM to be within the limits specified in the      24 hours COLR.
WOG STS                                    3.1.1-1                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 12 of 213


ACTIONS  CONDITION REQUIRED ACTION COMPLETION TIME
Attachment 1, Volume 6, Rev. 0, Page 13 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1, SHUTDOWN MARGIN (SDM)
None Kewaunee Power Station            Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 13 of 213


A. SDM not within limits.  
Attachment 1, Volume 6, Rev. 0, Page 14 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 14 of 213


A.1 Initiate boration to restore SDM to within limits.  
Attachment 1, Volume 6, Rev. 0, Page 15 of 213 SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)
 
BASES BACKGROUND                   According to GDC 26 (Ref. 1), the reactivity control systems must be INSERT 1        redundant and capable of holding the reactor core subcritical when shut       8 down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.
15 minutes DOC M01 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth 1
 
is fully withdrawn.           control            (RCCA)
SR  3.1.1.1 Verify SDM to be within the limits specified in the COLR.
The system design requires that two independent reactivity control           8 INSERT 2      systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor Coolant System (RCS). The Control Rod System can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the Control Rod System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn. The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions.
24 hours DOC M02 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1, SHUTDOWN MARGIN (SDM)
During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.6, "Control Bank Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.
None Kewaunee Power Station Page 1 of 1 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
APPLICABLE                   The minimum required SDM is assumed as an initial condition in safety SAFETY                       analyses. The safety analysis (Ref. 2) establishes an SDM that ensures       9 ANALYSES                     specified acceptable fuel design limits are not exceeded for normal (which includes reactor  operation and AOOs, with the assumption of the highest worth rod stuck safety analyses not in  out on scram. For MODE 5, the primary safety analysis that relies on the     10 USAR Chapter 14)
SDM B 3.1.1   WOG STS B 3.1.1-1 Rev. 3.0, 03/31/04 B 3.1 REACTIVITY CONTROL SYSTEMS  
SDM limits is the boron dilution analysis.
 
WOG STS                                              B 3.1.1-1                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 15 of 213
B 3.1.1 SHUTDOWN MARGIN (SDM)  
 
BASES BACKGROUND According to GDC 26 (Ref. 1), the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.
8INSERT 1  SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth is fully withdrawn.
1control (RCCA)
The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor Coolant System (RCS). The Control Rod System can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the Control Rod System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn. The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions.
8INSERT 2 During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.6, "Control Bank Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.
APPLICABLE The minimum required SDM is assumed as an initial condition in safety SAFETY analyses. The safety analysis (Ref. 2) establishes an SDM that ensures ANALYSES specified acceptable fuel design limits are not exceeded for normal operation and AOOs, with the assumption of the highest worth rod stuck out on scram. For MODE 5, the primary safety analysis that relies on the SDM limits is the boron dilution analysis.
9(which includes reactor safety analyses not in USAR Chapter 14) 10 B 3.1.1 8 INSERT 1   Two independent reactivity control systems, preferably from different principles, shall be provided (Ref. 1). The reactivity control systems shall be capable of making and holding the core sub-critical from any hot standby or hot operating condition (Ref. 2).


Attachment 1, Volume 6, Rev. 0, Page 16 of 213 B 3.1.1 8
INSERT 1 Two independent reactivity control systems, preferably from different principles, shall be provided (Ref. 1). The reactivity control systems shall be capable of making and holding the core sub-critical from any hot standby or hot operating condition (Ref. 2).
8 INSERT 2 Furthermore, USAR GDC 30, "Reactivity Holddown Capability," requires that the reactivity control system provided shall be capable of making the core subcritical under credible accident conditions with appropriate margins for contingencies and limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public (Ref. 3).
8 INSERT 2 Furthermore, USAR GDC 30, "Reactivity Holddown Capability," requires that the reactivity control system provided shall be capable of making the core subcritical under credible accident conditions with appropriate margins for contingencies and limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public (Ref. 3).
Insert Page B 3.1.1-1 SDM  B 3.1.1    WOG STS B 3.1.1-2 Rev. 3.0, 03/31/04  BASES APPLICABLE SAFETY ANALYSES  (continued)
Insert Page B 3.1.1-1 Attachment 1, Volume 6, Rev. 0, Page 16 of 213


The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:  
Attachment 1, Volume 6, Rev. 0, Page 17 of 213 SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)
: a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events, 2  
The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:
: b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature limits for  
: a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events,                                         2
: b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature limits for AOOs, and  280 cal/gm energy deposition for the rod ejection              1 accident), and      200                                                        2
: c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
The most limiting accident for the SDM requirements is based on a main steam line break (MSLB), as described in the accident analysis (Ref. 2).            1 4
The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. As RCS temperature decreases, the severity of an MSLB decreases until the MODE 5 value is reached. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs, is a guillotine break of a main steam line inside containment initiated at the end of core life. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, no fuel damage occurs as a result of the post trip return to power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1.
In addition to the limiting MSLB transient, the SDM requirement must also protect against:
: a. Inadvertent boron dilution,                                                  2
: b. An uncontrolled rod withdrawal from subcritical or low power condition,                                                                    2
                                    ; and WOG STS                                  B 3.1.1-2                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 17 of 213


AOOs, and  280 cal/gm energy deposition for the rod ejection accident), and  
Attachment 1, Volume 6, Rev. 0, Page 18 of 213 SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)
: c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.  
: c. Startup of an inactive reactor coolant pump (RCP), and 3
c d. Rod ejection.
Each of these events is discussed below.
In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest.
Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high an overtemperature T                                                                                  1 power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits.
The startup of an inactive RCP will not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur 3
due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition.
The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure.
The ejection of a rod also produces a time dependent redistribution of core power.
SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed from the control room, SDM is considered an initial 4
condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.
LCO                      SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.
WOG STS                                          B 3.1.1-3                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 18 of 213


The most limiting accident for the SDM requirements is based on a main steam line break (MSLB), as described in the accident analysis (Ref. 2). The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. As RCS temperature decreases, the severity of an MSLB decreases until the MODE 5 value is reached. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs, is a guillotine break of a main steam line inside containment initiated at the end of core life. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, no fuel damage occurs as a result of the post trip return to power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1.
Attachment 1, Volume 6, Rev. 0, Page 19 of 213 SDM B 3.1.1 BASES LCO (continued) 4                                          5 The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most                       1 limiting analyses that establish the SDM value of the LCO. For MSLB 50.67 accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4).                       1 Accident Source Term For the boron dilution accident, if the LCO is violated, the minimum                       6 required time assumed for operator action to terminate dilution may no longer be applicable.
 
APPLICABILITY             In MODE 2 with keff < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, in MODE "Boron Concentration." In MODES 1 and 2, SDM is ensured by with keff > 1.0                      1 complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6.
In addition to the limiting MSLB transient, the SDM requirement must also protect against:
ACTIONS                   A.1 If the SDM requirements are not met, boration must be initiated promptly.
: a. Inadvertent boron dilution,  b. An uncontrolled rod withdrawal from subcritical or low power condition, ;; 200; 2 1 1 4 2 2; and SDM  B 3.1.1    WOG STS B 3.1.1-3 Rev. 3.0, 03/31/04  BASES APPLICABLE SAFETY ANALYSES  (continued)
: c. Startup of an inactive reactor coolant pump (RCP), and 3 d. Rod ejection.
c Each of these events is discussed below.
 
In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest.
 
Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits.
The startup of an inactive RCP will not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition. 
 
The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The ejection of a rod also produces a time dependent redistribution of
 
core power.
 
SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.
 
LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.
3 4an overtemperature T 1 SDM  B 3.1.1   WOG STS B 3.1.1-4 Rev. 3.0, 03/31/04  BASES LCO (continued)
The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most limiting analyses that establish the SDM value of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4).
For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable.
APPLICABILITY In MODE 2 with k eff < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration." In MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6.  
 
ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly.
A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met.
A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met.
In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly concentrated solution, such as that normally found in the boric acid storage tank, or the borated water storage tank. The operator should borate with the best source available for the plant conditions.  
In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly refueling concentrated solution, such as that normally found in the boric acid storage tank, or the borated water storage tank. The operator should                           1 borate with the best source available for the plant conditions.
 
In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm. Assuming that a value of 1% k/k must be recovered and a boration flow rate of [ ] gpm, it is 40                      5 possible to increase the boron concentration of the RCS by 100 ppm in 12                                                                                                1 approximately 35 minutes. If a boron worth of 10 pcm/ppm is assumed, 40 this combination of parameters will increase the SDM by 1% k/k. These 13,000 boration parameters of [ ] gpm and [ ] ppm represent typical values and                          5 1
In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm. Assuming that a value  
are provided for the purpose of offering a specific example.
are the typical values when borating from the boric acid storage tank WOG STS                                                    B 3.1.1-4                                    Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 19 of 213


of 1% k/k must be recovered and a boration flow rate of [  ] gpm, it is possible to increase the boron concentration of the RCS by 100 ppm in approximately 35 minutes. If a boron worth of 10 pcm/ppm is assumed, this combination of parameters will increase the SDM by 1% k/k. These boration parameters of [  ] gpm and [  ] ppm represent typical values and
Attachment 1, Volume 6, Rev. 0, Page 20 of 213 SDM B 3.1.1 BASES SURVEILLANCE           SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2 with Keff 1.0, SDM is verified by observing that the                         7 requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum MODE 2 with worth.
 
keff <1.0 and in In MODES 3, 4, and 5, the SDM is verified by performing a reactivity                           6 balance calculation, considering the listed reactivity effects:
are provided for the purpose of offering a specific example.
: a. RCS boron concentration,                                                                   2
 
: b. Control bank position,                                                                     2
1 1 1 5 4 6 in MODE with k eff > 1.0 40 13,000 40 5 550.67 Accident Source Term refueling 1 12 1 1are the typical values when borating from the boric acid storage tank SDM B 3.1.1    WOG STS B 3.1.1-5 Rev. 3.0, 03/31/04  BASES SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2 with K eff 1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum  
: c. RCS average temperature,                                                                   2
 
: d. Fuel burnup based on gross thermal energy generation,                                     2
worth. 7 In MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation, considering the listed reactivity effects: MODE 2 with k eff <1.0 and in    a. RCS boron concentration,  
: e. Xenon concentration,                                                                       2
: b. Control bank position, c. RCS average temperature, d. Fuel burnup based on gross thermal energy generation, e. Xenon concentration, f. Samarium concentration, and  
: f. Samarium concentration, and                                                               2
: g. Isothermal temperature coefficient (ITC).
: g. Isothermal temperature coefficient (ITC).
Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the  
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.
 
same rate as the RCS.  
 
The Frequency of 24 hours is based on the generally slow change in required boron concentration and the low probability of an accident occurring without the required SDM. This allows time for the operator to collect the required data, which includes performing a boron concentration analysis, and complete the calculation.
The Frequency of 24 hours is based on the generally slow change in required boron concentration and the low probability of an accident occurring without the required SDM. This allows time for the operator to collect the required data, which includes performing a boron concentration analysis, and complete the calculation.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.  
USAR, Section 3.1.2.3, GDC 27, 8
: 2. FSAR, Chapter [15].
REFERENCES             1. 10 CFR 50, Appendix A, GDC 26.           "Redundancy of Reactivity Control."
: 3. FSAR, Chapter [15].  
4                                                   2. USAR, Section 3.1.2.4, GDC 28,                 8
: 4. 10 CFR 100.
: 2. FSAR, Chapter [15].      Section 14.2.5 "Reactivity Hot Shutdown Capability."           5 U
; ; ; ; ; ; Section 14.2.5 Section 14.1.450.67 USAR, Section 3.1.2.3, GDC 27, "Redundancy of Reactivity Control." 6 2 2 2 2 2 2 8 4 82. USAR, Section 3.1.2.4, GDC 28, "Reactivity Hot Shutdown Capability
: 3. USAR, Section 3.1.2.6, GDC 30,
." 5 U 3. USAR, Section 3.1.2.6, GDC 30, "Reactivity Holddown Capability." 5 8 5 6 1 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1 BASES, SHUTDOWN MARGIN (SDM)
: 3. FSAR, Chapter [15].      Section 14.1.4 "Reactivity Holddown Capability."       8    5 5
: 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.  
6
: 2. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
: 4. 10 CFR 100. 50.67                                                                      1 WOG STS                                        B 3.1.1-5                                  Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 20 of 213
: 3. ISTS 3.1.1 Applicable Safety Analyses states, in part, that in addition to the limiting Main Steam Line Break (MSLB), the SDM requirements must also protect against inadvertent boron dilution, an uncontrolled rod withdrawal from subcritical or low power condition, startup of an inactive reactor coolant pump (RCP), and rod ejection. For the description of the event, the ISTS Bases states that the "startup of an inactive RCP will not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition. USAR, Section 14.1.5, the Startup of an Inactive Reactor Coolant Loop, states that if the plant were to operate with one


reactor coolant pump (RCP) out-of-service, there would be reverse flow through the inactive loop due to the pressure difference across the reactor vessel and because there are no isolation valves or check valv es in the reactor coolant loops. The cold leg temperature in the inactive loop is identical to the cold leg temperature of the active loop (the reactor core inlet temperature). Therefore, a startup of an inactive RCP would not cause a discernable power excursion, since water colder than the operating loop is not being added to the reactor vessel. Thus, the ITS 3.1.1 Bases has been modified to delete the startup of an inactive RCP requirements.  
Attachment 1, Volume 6, Rev. 0, Page 21 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1 BASES, SHUTDOWN MARGIN (SDM)
: 4. The Applicable Safety Analyses discussion states SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). It also says that even though SDM is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of the accident analysis assumptions. The additional sentence has been  
: 1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 2. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
: 3. ISTS 3.1.1 Applicable Safety Analyses states, in part, that in addition to the limiting Main Steam Line Break (MSLB), the SDM requirements must also protect against inadvertent boron dilution, an uncontrolled rod withdrawal from subcritical or low power condition, startup of an inactive reactor coolant pump (RCP), and rod ejection.
For the description of the event, the ISTS Bases states that the startup of an inactive RCP will not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition. USAR, Section 14.1.5, the Startup of an Inactive Reactor Coolant Loop, states that if the plant were to operate with one reactor coolant pump (RCP) out-of-service, there would be reverse flow through the inactive loop due to the pressure difference across the reactor vessel and because there are no isolation valves or check valves in the reactor coolant loops. The cold leg temperature in the inactive loop is identical to the cold leg temperature of the active loop (the reactor core inlet temperature). Therefore, a startup of an inactive RCP would not cause a discernable power excursion, since water colder than the operating loop is not being added to the reactor vessel. Thus, the ITS 3.1.1 Bases has been modified to delete the startup of an inactive RCP requirements.
: 4. The Applicable Safety Analyses discussion states SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). It also says that even though SDM is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of the accident analysis assumptions. The additional sentence has been deleted. The NRC Final Policy Statement on Technical Improvements of July 22, 1993 (58 FR 39132) states that process variable captured by Criterion 2 are not limited to only those directly monitored and controlled from the control room. It also states that Criterion 2 includes other features or characteristics that are specifically assumed in Design Basis Accident and Transient analyses even if they cannot be directly observed in the control room (e.g., moderator temperature coefficient and hot channel factors). Since the Final Policy Statement provides guidance on which types of parameters satisfy Criterion 2, there is no reason to duplicate these words in the KPS ITS.
: 5. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 6. ISTS SR 3.1.1.1 Bases states, in part, that in MODES 1 and 2 with keff  1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. It continues by stating, in part, that in MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation. ITS SR 3.1.1.1 Bases states, in part, that Kewaunee Power Station                  Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 21 of 213


deleted. The NRC Final Policy Statement on Technical Improvements of July 22, 1993 (58 FR 39132) states that process variable captured by Criterion 2 are not limited to only those directly monitored and controlled from the control room. It also states that Criterion 2 includes other features or characteristics that are specifically assumed in Design Basis Accident and Transient analyses even if they
Attachment 1, Volume 6, Rev. 0, Page 22 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1 BASES, SHUTDOWN MARGIN (SDM) in MODES 1 and 2 with keff 1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. It continues by stating, in part, that in MODE 2 with keff < 1.0 and in MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation. This changes the ISTS by stating that in MODE 2 with keff < 1.0 the SDM is verified by performing a reactivity balance calculation. This is acceptable because the ISTS did not state that a reactivity balance calculation is performed when the keff < 1.0 in MODE 2.
 
: 7. Typographical error corrected.
cannot be directly observed in the cont rol room (e.g., moderator temperature coefficient and hot channel factors). Since the Final Policy Statement provides guidance on which types of parameters satisfy Criterion 2, there is no reason to
: 8. The ISTS lists GDC 26 of Appendix A to 10 CFR 50 as the reference document for the requirement that there be two independent reactivity control systems of different design principles. Per the information contained in USAR Section 1.8, Kewaunee Power Station (KPS) was designed, constructed, and is being operated to comply with the Atomic Energy Commission (AEC) General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, as proposed on July 10, 1967. Since the plant was approximately 50% complete prior to the February 20, 1971 issuance of 10 CFR 50 Appendix A General Design Criteria, KPS was not required to be reanalyzed and the Final Safety Analysis Report (FSAR) was not required to be revised to reflect these later criteria. However, the AEC Safety Evaluation Report (SER), issued July 24, 1972, acknowledged that the AEC staff assessed the plant, as described in the FSAR (Amendment No. 7), against the Appendix A design criteria and determined that the plant design generally conforms to the intent of the Appendix A criteria. As a result, KPS utilizes AEC GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot Shutdown Capability," and GDC 30, "Reactivity Holddown Capability", as the licensing reference documents for the design principles.
 
: 9. Clarification has been added since all the reactor safety analyses are not in USAR Chapter 14.
duplicate these words in the KPS ITS.
: 10. KPS does not analyze a boron dilution event in MODE 5. As stated in KPS USAR, Section 14.1.4.2, boron dilutions during cold shutdown are not part of the KPS licensing basis.
: 5. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
Kewaunee Power Station                  Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 22 of 213
: 6. ISTS SR 3.1.1.1 Bases states, in part, that in MODES 1 and 2 with k eff 1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. It continues by stating, in part, that in MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation. ITS SR 3.1.1.1 Bases states, in part, that Kewaunee Power Station Page 1 of 2 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1 BASES, SHUTDOWN MARGIN (SDM)
Kewaunee Power Station Page 2 of 2 in MODES 1 and 2 with k eff  1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. It continues by stating, in part, that in MODE 2 with k eff < 1.0 and in MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation. This changes the ISTS by stating that in MODE 2 with


k eff < 1.0 the SDM is verified by performing a reactivity balance calculation. This is acceptable because the ISTS did not state that a reactivity balance calculation is performed when the k eff  < 1.0 in MODE 2.
Attachment 1, Volume 6, Rev. 0, Page 23 of 213 Specific No Significant Hazards Considerations (NSHCs)
: 7. Typographical error corrected.
Attachment 1, Volume 6, Rev. 0, Page 23 of 213
: 8. The ISTS lists GDC 26 of Appendix A to 10 CFR 50 as the reference document for the requirement that there be two independent reactivity control systems of different design principles. Per the information contained in USAR Section 1.8, Kewaunee


Power Station (KPS) was designed, constr ucted, and is being operated to comply with the Atomic Energy Commission (AEC) General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, as proposed on July 10, 1967. Since the plant was approximately 50% complete prior to the February 20, 1971 issuance of 10 CFR 50 Appendix A General Design Criteria, KPS was not required to be reanalyzed and the Final Safety Analysis Report (FSAR) was not required to be revised to reflect these later criteria. However, the AEC Safety Evaluation Report (SER), issued July 24, 1972, acknowledged that the AEC staff assessed the plant, as described in the FSAR (Amendment No. 7), against the Appendix A design criteria and determined that the plant design generally conforms to the intent of the Appendix A criteria. As a result, KPS utilizes AEC GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot Shutdown Capability," and GDC 30, "Reactivity Holddown Capability", as the licensing reference documents for the design principles.
Attachment 1, Volume 6, Rev. 0, Page 24 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.1, SHUTDOWN MARGIN (SDM)
: 9. Clarification has been added since all the reactor safety analyses are not in USAR Chapter 14.
There are no specific NSHC discussions for this Specification.
: 10. KPS does not analyze a boron dilution event in MODE 5. As stated in KPS USAR, Section 14.1.4.2, boron dilutions during cold shutdown are not part of the KPS licensing basis.
Kewaunee Power Station               Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 24 of 213
Specific No Significant Haza rds Considerations (NSHCs)
, Volume 6, Rev. 0, Page 25 of 213 ATTACHMENT 2 ITS 3.1.2, CORE REACTIVITY , Volume 6, Rev. 0, Page 25 of 213
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.1, SHUTDOWN MARGIN (SDM) There are no specific NSHC discussions for this Specification.
, Volume 6, Rev. 0, Page 26 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 26 of 213
Kewaunee Power Station Page 1 of 1 ATTACHMENT 2 ITS 3.1.2, CORE REACTIVITY Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
A01 ITS 3.1.2 ITS 4.9 REACTIVITY ANOMALIES APPLICABILITY Applies to potential reactivity anomalies.  


OBJECTIVE   To require evaluation of reactivity anomalies within the reactor.
Attachment 1, Volume 6, Rev. 0, Page 27 of 213 ITS                                                        A01                                          ITS 3.1.2 4.9 REACTIVITY ANOMALIES APPLICABILITY Applies to potential reactivity anomalies.
SPECIFICATION Add proposed Applicability A02  Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be periodically compared with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the equivalent of 1% in reactivity, an evaluation as to the cause of the discrepancy  
OBJECTIVE To require evaluation of reactivity anomalies within the reactor.
 
SPECIFICATION                                                                                             A02 Add proposed Applicability Following a normalization of the computed boron concentration as a function of burnup, the M01 SR 3.1.2.1    actual boron concentration of the coolant shall be periodically compared with the predicted value. If the difference between the observed and predicted steady-state concentrations LCO 3.1.2      reaches the equivalent of 1% in reactivity, an evaluation as to the cause of the discrepancy shall be made and reported to the Commission within 30 days.
shall be made and reported to the Commission within 30 days.
Add proposed ACTION A and B           M02 Amendment No. 122 TS 4.9-1                                           12/21/95 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 27 of 213
M01 SR 3.1.2.1 LCO 3.1.2 M02 Add proposed ACTION A and B Amendment No. 122 TS 4.9-1 12/21/95 Page 1 of 1 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).


Attachment 1, Volume 6, Rev. 0, Page 28 of 213 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY ADMINISTRATIVE CHANGES A01  In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.
3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
A02 CTS 4.9 requires a comparison of the actual boron concentration with the predicted (computer) value, and requires the difference between the computed boron concentration, as a function of burnup, and the actual boron concentration  
A02 CTS 4.9 requires a comparison of the actual boron concentration with the predicted (computer) value, and requires the difference between the computed boron concentration, as a function of burnup, and the actual boron concentration to be +/- 1% k/k. ITS 3.1.2 also requires the comparison of the actual boron concentration with the predicted value and requires the difference to be within
 
    +/- 1% k/k, but specifically states that the Applicability is in MODES 1 and 2.
to be +/- 1% k/k. ITS 3.1.2 also requires the comparison of the actual boron concentration with the predicted value and requires the difference to be within  
This changes the CTS by stating the specific MODES.
+/- 1% k/k, but specifically states that the Applicability is in MODES 1 and 2. This changes the CTS by stating the specific MODES.
This change is acceptable because the required MODES have not changed.
This change is acceptable because the required MODES have not changed. When full power is initially reached, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the actual curve relating burn-up and reactivity is compared with that of the predicted curve. Therefore, since the comparison of the actual boron concentration to the predicted concentration and the computed boron concentration is a function of burnup (which can only occur when the reactor is critical) it can only be performed in MODES 1 and 2. This change is considered administrative because the technical requirements have not changed.
When full power is initially reached, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the actual curve relating burn-up and reactivity is compared with that of the predicted curve. Therefore, since the comparison of the actual boron concentration to the predicted concentration and the computed boron concentration is a function of burnup (which can only occur when the reactor is critical) it can only be performed in MODES 1 and 2. This change is considered administrative because the technical requirements have not changed.
 
MORE RESTRICTIVE CHANGES M01 CTS 4.9 states, in part, that after normalization, the actual boron concentration of the coolant shall be periodically compared with the predicted value. ITS SR 3.1.2.1 requires verification of the measured core reactivity is within
MORE RESTRICTIVE CHANGES M01 CTS 4.9 states, in part, that after normalization, the actual boron concentration of the coolant shall be periodically compared with the predicted value. ITS SR 3.1.2.1 requires verification of the measured core reactivity is within  
    +/- 1% k/k of predicted values at least once prior to entering MODE 1 after each refueling and after 60 effective full power days (EFPD) it should be verified every 31 EFPD thereafter. This changes the CTS by specifying Frequencies for verifying measured core reactivity is within +/- 1% k/k of predicted values.
 
+/- 1% k/k of predicted values at least once prior to entering MODE 1 after each refueling and after 60 effective full power days (EFPD) it should be verified every 31 EFPD thereafter. This changes the CTS by specifying Frequencies for verifying measured core reactivity is within +/- 1% k/k of predicted values.  
 
This change is acceptable because it requires a verification that core reactivity is within +/- 1% k/k of predicted values thus assuring agreement between the actual core design and the core design predictions prior to entering MODE 1 after each refueling, and every 31 EFPD after 60 EFPD. This verification provides additional confidence that the core design is acceptable for operation at full power. This change is designated as more restrictive because it adds Surveillance Requirement frequencies that do not appear in the CTS.
This change is acceptable because it requires a verification that core reactivity is within +/- 1% k/k of predicted values thus assuring agreement between the actual core design and the core design predictions prior to entering MODE 1 after each refueling, and every 31 EFPD after 60 EFPD. This verification provides additional confidence that the core design is acceptable for operation at full power. This change is designated as more restrictive because it adds Surveillance Requirement frequencies that do not appear in the CTS.
M02 CTS 4.9 requires that if the difference between the observed and predicted steady-state concentrations reach the equivalent of 1% in reactivity, that an Kewaunee Power Station Page 1 of 3 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY evaluation as to the cause of the discrepancy shall be made and reported to the Commission within 30 days. ITS 3.1.2 requires that if the measured core reactivity is not within limits, within 7 days a re-evaluation of the core design and safety analysis, a determination that the reactor core is acceptable for continued operation, and establishment of appropriate operating restrictions and SRs are required. If the re-evaluation of the core design and safety analysis to determine that the reactor core is acceptable for continued operation, and the establishment of appropriate operating restrictions and SRs can not be accomplished within 7 days, then the reactor is required to be in MODE 3 within 6 hours. This changes the CTS by requiring a re-evaluation of the core design and safety analysis, a determination that the reactor core is acceptable for continued operation, and establishment of appropriate operating restrictions and SRs instead of requiring an evaluation of the cause of the discrepancy shall be made and reported to the Commission. This also changes the Completion Time from
M02 CTS 4.9 requires that if the difference between the observed and predicted steady-state concentrations reach the equivalent of 1% in reactivity, that an Kewaunee Power Station                 Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 28 of 213


30 days to 7 days.
Attachment 1, Volume 6, Rev. 0, Page 29 of 213 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY evaluation as to the cause of the discrepancy shall be made and reported to the Commission within 30 days. ITS 3.1.2 requires that if the measured core reactivity is not within limits, within 7 days a re-evaluation of the core design and safety analysis, a determination that the reactor core is acceptable for continued operation, and establishment of appropriate operating restrictions and SRs are required. If the re-evaluation of the core design and safety analysis to determine that the reactor core is acceptable for continued operation, and the establishment of appropriate operating restrictions and SRs can not be accomplished within 7 days, then the reactor is required to be in MODE 3 within 6 hours. This changes the CTS by requiring a re-evaluation of the core design and safety analysis, a determination that the reactor core is acceptable for continued operation, and establishment of appropriate operating restrictions and SRs instead of requiring an evaluation of the cause of the discrepancy shall be made and reported to the Commission. This also changes the Completion Time from 30 days to 7 days.
This change is acceptable because the proposed Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing an appropriate amount of time to ensure continued safe operation and re-evaluate the core design and safety analyses. The Required Actions are consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. Should the core reactivity balance requirement not be met, time is required to determine the cause of the disagreement and what, if any, adjustments are needed to the operating conditions of the core. The startup physics testing program is used to verify most of the critical core design parameters, such as control rod worth, boron worth, and moderator temperature coefficient. In addition, there is considerable conservatism in the application of these values in the accident analysis. Therefore, allowing a time to evaluate the difference and make any adjustments to the operational controls is acceptable.
This change is acceptable because the proposed Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing an appropriate amount of time to ensure continued safe operation and re-evaluate the core design and safety analyses. The Required Actions are consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. Should the core reactivity balance requirement not be met, time is required to determine the cause of the disagreement and what, if any, adjustments are needed to the operating conditions of the core. The startup physics testing program is used to verify most of the critical core design parameters, such as control rod worth, boron worth, and moderator temperature coefficient. In addition, there is considerable conservatism in the application of these values in the accident analysis. Therefore, allowing a time to evaluate the difference and make any adjustments to the operational controls is acceptable.
The 7 day Completion time is reasonable considering the complexity of the evaluations and the time to meet administrative requirements, such as 10 CFR 50.59 safety evaluation preparation and approval. If it cannot be determined within 7 days that the core is acceptable for continued operation, the unit must be shutdown. This change is designated as more restrictive because more stringent Required Actions are being applied in the ITS than were applied in the CTS.  
The 7 day Completion time is reasonable considering the complexity of the evaluations and the time to meet administrative requirements, such as 10 CFR 50.59 safety evaluation preparation and approval. If it cannot be determined within 7 days that the core is acceptable for continued operation, the unit must be shutdown. This change is designated as more restrictive because more stringent Required Actions are being applied in the ITS than were applied in the CTS.
 
RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None Kewaunee Power Station                   Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 29 of 213
RELOCATED SPECIFICATIONS None
 
REMOVED DETAIL CHANGES None   Kewaunee Power Station Page 2 of 3 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY Kewaunee Power Station Page 3 of 3 LESS RESTRICTIVE CHANGES None Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Core Reactivity 3.1.2  WOG STS 3.1.2-1 Rev. 3.0, 03/31/04  CTS 3.1  REACTIVITY CONTROL SYSTEMS
 
3.1.2 Core Reactivity
 
LCO  3.1.2 The measured core reactivity shall be within +/- 1% k/k of predicted values. 4.9 APPLICABILITY: MODES 1 and 2. DOC A02 ACTIONS  CONDITION REQUIRED ACTION COMPLETION TIME
 
A. Measured core reactivity not within limit.
A.1 Re-evaluate core design and safety analysis, and determine that the reactor core is acceptable for continued operation.
 
AND  A.2 Establish appropriate operating restrictions and SRs. 7 days


7 days B. Required Action and associated Completion Time not met.  
Attachment 1, Volume 6, Rev. 0, Page 30 of 213 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY LESS RESTRICTIVE CHANGES None Kewaunee Power Station          Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 30 of 213


B.1 Be in MODE 3.  
Attachment 1, Volume 6, Rev. 0, Page 31 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 31 of 213


6 hours 4.9 4.9
Attachment 1, Volume 6, Rev. 0, Page 32 of 213 CTS                                                                                        Core Reactivity 3.1.2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2      Core Reactivity 4.9    LCO 3.1.2              The measured core reactivity shall be within +/- 1% k/k of predicted values.
DOC A02 APPLICABILITY:          MODES 1 and 2.
ACTIONS CONDITION                    REQUIRED ACTION                  COMPLETION TIME 4.9     A. Measured core reactivity    A.1    Re-evaluate core design          7 days not within limit.                  and safety analysis, and determine that the reactor core is acceptable for continued operation.
AND A.2    Establish appropriate            7 days operating restrictions and SRs.
4.9     B. Required Action and          B.1    Be in MODE 3.                    6 hours associated Completion Time not met.
WOG STS                                    3.1.2-1                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 32 of 213


Core Reactivity 3.1.2   WOG STS 3.1.2-2 Rev. 3.0, 03/31/04  CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 ---------------------------NOTE----------------------------------
Attachment 1, Volume 6, Rev. 0, Page 33 of 213 CTS                                                                                            Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                 FREQUENCY 4.9 SR 3.1.2.1   ---------------------------NOTE----------------------------------
The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading. ---------------------------------------------------------------------
The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.
Verify measured core reactivity is within +/- 1% k/k of predicted values.
Verify measured core reactivity is within +/- 1% k/k                  Once prior to of predicted values.                                                  entering MODE 1 after each refueling AND
                                                                                        --------NOTE--------
Only required after 60 EFPD 31 EFPD thereafter WOG STS                                        3.1.2-2                                Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 33 of 213


Once prior to
Attachment 1, Volume 6, Rev. 0, Page 34 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.2, CORE REACTIVITY None Kewaunee Power Station            Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 34 of 213


entering MODE 1  
Attachment 1, Volume 6, Rev. 0, Page 35 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 35 of 213


after each refueling AND 
Attachment 1, Volume 6, Rev. 0, Page 36 of 213 Core Reactivity B 3.1.2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Core Reactivity BASES BACKGROUND                  According to GDC 26, GDC 28, and GDC 29 (Ref. 1), reactivity shall be INSERT 1 controllable, such that subcriticality is maintained under cold conditions,    1 and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity balance is used as a measure of the predicted versus measured core reactivity during power operation. The periodic confirmation of core reactivity is necessary to ensure that Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity difference could be the result of unanticipated changes in fuel, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus (which includes reactor measured core reactivity validates the nuclear methods used in the safety      6 safety analyses not in analysis and supports the SDM demonstrations (LCO 3.1.1, USAR Chapter 14)
--------NOTE--------
                            "SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, subcritical conditions.
Only required  
When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers producing zero net reactivity.
Excess reactivity can be inferred from the boron letdown curve (or critical boron curve), which provides an indication of the soluble boron concentration in the Reactor Coolant System (RCS) versus cycle burnup.
Periodic measurement of the RCS boron concentration for comparison with the predicted value with other variables fixed (such as rod height, temperature, pressure, and power), provides a convenient method of ensuring that core reactivity is within design expectations and that the calculational models used to generate the safety analysis are adequate.
In order to achieve the required fuel cycle energy output, the uranium enrichment, in the new fuel loading and in the fuel remaining from the previous cycle, provides excess positive reactivity beyond that required to sustain steady state operation throughout the cycle. When the reactor is WOG                                                B 3.1.2-1                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 36 of 213


after 60 EFPD ------------------------
Attachment 1, Volume 6, Rev. 0, Page 37 of 213 B 3.1.2 1
 
INSERT 1 Two independent reactivity control systems, preferably from different principles, shall be provided (Ref. 1). The reactivity control systems provided shall be capable of making and holding the core sub-critical from any hot standby or hot operation condition (Ref. 2). One of the reactivity control systems provided shall be capable of making the core sub-critical under any anticipated operating condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margin should assure sub-criticality with the most reactive control rod fully withdrawn (Ref. 3). The reactivity control system provided shall be capable of making the core sub-critical under credible accident conditions with appropriate margins for contingencies and limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public (Ref. 4).
31 EFPD thereafter
Insert Page B 3.1.2-1 Attachment 1, Volume 6, Rev. 0, Page 37 of 213
 
===4.9 JUSTIFICATION===
FOR DEVIATIONS ITS 3.1.2, CORE REACTIVITY None Kewaunee Power Station Page 1 of 1 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Core Reactivity B 3.1.2    WOG B 3.1.2-1 Rev. 3.0, 03/31/04 B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.2 Core Reactivity
 
BASES  BACKGROUND According to GDC 26, GDC 28, and GDC 29 (Ref. 1), reactivity shall be controllable, such that subcriticality is maintained under cold conditions, and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity balance is used as a measure of the predicted versus measured core reactivity during power operation. The periodic confirmation of core reactivity is necessary to ensure that Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity difference could be the result of unanticipated changes in fuel, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, subcritical conditions. INSERT 1 1 When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers producing zero net reactivity. Excess reactivity can be inferred from the boron letdown curve (or critical boron curve), which provides an indication of the soluble boron concentration in the Reactor Coolant System (RCS) versus cycle burnup.
Periodic measurement of the RCS boron concentration for comparison with the predicted value with other variables fixed (such as rod height, temperature, pressure, and power), provides a convenient method of ensuring that core reactivity is within design expectations and that the calculational models used to generate the safety analysis are adequate.
 
In order to achieve the required fuel cycle energy output, the uranium enrichment, in the new fuel loading and in the fuel remaining from the previous cycle, provides excess posit ive reactivity beyond that required to sustain steady state operation throughout the cycle. When the reactor is  (which includes reactor safety analyses not in USAR Chapter 14) 6 B 3.1.2 1 INSERT 1   Two independent reactivity control systems, preferably from different principles, shall be provided (Ref. 1). The reactivity control systems provided shall be capable of making and holding the core sub-critical from any hot standby or hot operation condition (Ref. 2). One of the reactivity control systems provided shall be capable of making the core sub-critical under any anticipated operating condition (including anticipated operational tr ansients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margin should assure sub-criticality with the most reactive control rod fully withdrawn (Ref. 3). The reactivity control system provided shall be capable of making the core sub-critical under credible accident conditions with appropriate margins for contingencies and limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public (Ref. 4).
Insert Page B 3.1.2-1 Core Reactivity B 3.1.2    WOG B 3.1.2-2 Rev. 3.0, 03/31/04  BASES BACKGROUND  (continued)
 
critical at RTP and moderator temperature, the excess positive reactivity is compensated by burnable absorbers (if any), control rods, whatever neutron poisons (mainly xenon and samarium) are present in the fuel, and the RCS boron concentration.


Attachment 1, Volume 6, Rev. 0, Page 38 of 213 Core Reactivity B 3.1.2 BASES BACKGROUND (continued) critical at RTP and moderator temperature, the excess positive reactivity is compensated by burnable absorbers (if any), control rods, whatever neutron poisons (mainly xenon and samarium) are present in the fuel, and the RCS boron concentration.
When the core is producing THERMAL POWER, the fuel is being depleted and excess reactivity is decreasing. As the fuel depletes, the RCS boron concentration is reduced to decrease negative reactivity and maintain constant THERMAL POWER. The boron letdown curve is based on steady state operation at RTP. Therefore, deviations from the predicted boron letdown curve may indicate deficiencies in the design analysis, deficiencies in the calculational models, or abnormal core conditions, and must be evaluated.
When the core is producing THERMAL POWER, the fuel is being depleted and excess reactivity is decreasing. As the fuel depletes, the RCS boron concentration is reduced to decrease negative reactivity and maintain constant THERMAL POWER. The boron letdown curve is based on steady state operation at RTP. Therefore, deviations from the predicted boron letdown curve may indicate deficiencies in the design analysis, deficiencies in the calculational models, or abnormal core conditions, and must be evaluated.
APPLICABLE The acceptance criteria for core reactivity are that the reactivity balance SAFETY limit ensures plant operation is maintained within the assumptions of ANALYSES the safety analyses.  
APPLICABLE     The acceptance criteria for core reactivity are that the reactivity balance SAFETY         limit ensures plant operation is maintained within the assumptions of ANALYSES       the safety analyses.
 
Accurate prediction of core reactivity is either an explicit or implicit 5
Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations. Every accident evaluation (Ref. 2) is, therefore, dependent upon accurate evaluation of core reactivity. In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod ejection accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity balance additionally ensures that the nuclear methods provide an accurate representation of the core reactivity.
assumption in the accident analysis evaluations. Every accident evaluation (Ref. 2) is, therefore, dependent upon accurate evaluation of         1 core reactivity. In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod ejection accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity balance additionally ensures that the nuclear methods provide an accurate representation of the core reactivity.
5 1 Design calculations and safety analyses are performed for each fuel cycle for the purpose of predetermining reactivity behavior and the RCS boron concentration requirements for reactivity control during fuel depletion.  
Design calculations and safety analyses are performed for each fuel cycle for the purpose of predetermining reactivity behavior and the RCS boron concentration requirements for reactivity control during fuel depletion.
 
The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted RCS boron concentrations for identical core conditions at beginning of cycle (BOC) do not agree, then the assumptions used in the reload cycle design analysis or the calculational models used to predict soluble boron requirements may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to WOG                                   B 3.1.2-2                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 38 of 213
The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted RCS boron concentrations for identical core conditions at beginning of cycle (BOC) do not agree, then the assumptions used in the reload cycle design analysis or the calculational models used to predict soluble boron requirements may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to Core Reactivity B 3.1.2    WOG B 3.1.2-3 Rev. 3.0, 03/31/04 BASES APPLICABLE SAFETY ANALYSES  (continued)
 
the measured boron concentration. Thereafter, any significant deviations in the measured boron concentration from the predicted boron letdown curve that develop during fuel depletion may be an indication that the calculational model is not adequate for core burnups beyond BOC, or that an unexpected change in core conditions has occurred.  
 
The normalization of predicted RCS boron concentration to the measured value is typically performed after reaching RTP following startup from a refueling outage, with the control rods in their normal positions for power operation. The normalization is performed at BOC conditions, so that
 
core reactivity relative to predicted values can be continually monitored and evaluated as core conditions change during the cycle.


Attachment 1, Volume 6, Rev. 0, Page 39 of 213 Core Reactivity B 3.1.2 BASES APPLICABLE SAFETY ANALYSES (continued) the measured boron concentration. Thereafter, any significant deviations in the measured boron concentration from the predicted boron letdown curve that develop during fuel depletion may be an indication that the calculational model is not adequate for core burnups beyond BOC, or that an unexpected change in core conditions has occurred.
The normalization of predicted RCS boron concentration to the measured value is typically performed after reaching RTP following startup from a refueling outage, with the control rods in their normal positions for power operation. The normalization is performed at BOC conditions, so that core reactivity relative to predicted values can be continually monitored and evaluated as core conditions change during the cycle.
Core reactivity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
Core reactivity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO Long term core reactivity behavior is a result of the core physics design and cannot be easily controlled once the core design is fixed. During operation, therefore, the LCO can only be ensured through measurement and tracking, and appropriate actions taken as necessary. Large differences between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the reactivity balance of +/- 1% k/k has been established based on engineering judgment. A 1% deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated.
LCO               Long term core reactivity behavior is a result of the core physics design and cannot be easily controlled once the core design is fixed. During operation, therefore, the LCO can only be ensured through measurement and tracking, and appropriate actions taken as necessary. Large differences between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the reactivity balance of +/- 1% k/k has been established based on engineering judgment. A 1% deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated.
When measured core reactivity is within 1% k/k of the predicted value at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Si nce deviations from the limit are normally detected by comparing predicted and measured steady state RCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached. These values are well within the uncertainty limits for analysis of boron concentration samples, so that spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely.
When measured core reactivity is within 1% k/k of the predicted value at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Since deviations from the limit are normally detected by comparing predicted and measured steady state 140 RCS critical boron concentrations, the difference between measured and 3
140 3 3much larger than APPLICABILITY The limits on core reactivity must be maintained during MODES 1 and 2 because a reactivity balance must exist when the reactor is critical or producing THERMAL POWER. As the fuel depletes, core conditions are changing, and confirmation of the reactivity balance ensures the core is operating as designed. This Specification does not apply in MODES 3, 4, and 5 because the reactor is shut down and the reactivity balance is not changing.
predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached. These values are well within the 3
Core Reactivity B 3.1.2    WOG B 3.1.2-4 Rev. 3.0, 03/31/04 BASES APPLICABILITY  (continued)
uncertainty limits for analysis of boron concentration samples, so that     much larger than spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely.
APPLICABILITY     The limits on core reactivity must be maintained during MODES 1 and 2 because a reactivity balance must exist when the reactor is critical or producing THERMAL POWER. As the fuel depletes, core conditions are changing, and confirmation of the reactivity balance ensures the core is operating as designed. This Specification does not apply in MODES 3, 4, and 5 because the reactor is shut down and the reactivity balance is not changing.
WOG                                       B 3.1.2-3                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 39 of 213


In MODE 6, fuel loading results in a continually changing core reactivity. Boron concentration requirements (LCO 3.9.1, "Boron Concentration") ensure that fuel movements are performed within the bounds of the safety analysis. An SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling).
Attachment 1, Volume 6, Rev. 0, Page 40 of 213 Core Reactivity B 3.1.2 BASES APPLICABILITY (continued)
ACTIONS A.1 and A.2
In MODE 6, fuel loading results in a continually changing core reactivity.
Boron concentration requirements (LCO 3.9.1, "Boron Concentration")
ensure that fuel movements are performed within the bounds of the safety analysis. An SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling).
ACTIONS           A.1 and A.2 Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine their consistency with input to design calculations. Measured core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 7 days is based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.
Following evaluations of the core design and safety analysis, the cause of the reactivity anomaly may be resolved. If the cause of the reactivity anomaly is a mismatch in core conditions at the time of RCS boron concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate that core reactivity is behaving as expected. If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may be renormalized and power operation may continue. If operational restriction or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined.
The required Completion Time of 7 days is adequate for preparing whatever operating restrictions or Surveillances that may be required to allow continued reactor operation.
WOG                                      B 3.1.2-4                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 40 of 213


Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine their consistency with input to design calculations. Measured core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 7 days is based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.  
Attachment 1, Volume 6, Rev. 0, Page 41 of 213 Core Reactivity B 3.1.2 BASES ACTIONS (continued)
B.1                                    If any Required Action and associated Completion Time is not met 2
If the core reactivity cannot be restored to within the 1% k/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 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 plant systems.
SURVEILLANCE      SR 3.1.2.1 REQUIREMENTS 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 control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization of predicted core reactivity to the 5
may    measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1, is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (QPTR, AFD, etc.) for prompt indication of an anomaly.
REFERENCES        1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.                                                          1 USAR, Section 3.1.2.3, GDC 27, "Redundancy of Reactivity Control."
5
: 2. FSAR, Chapter [15].                                                                                        3  4 U                          14
: 2. USAR, Section 3.1.2.4, GDC 28, "Reactivity Hot Shutdown Capability."
: 3. USAR, Section 3.1.2.5, GDC 29, "Reactivity Shutdown Capability."
: 4. USAR, Section 3.1.2.6, GDC 30, "Reactivity Holddown Capability."
WOG                                              B 3.1.2-5                                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 41 of 213


Following evaluations of the core design and safety analysis, the cause of the reactivity anomaly may be resolved. If the cause of the reactivity anomaly is a mismatch in core conditions at the time of RCS boron concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate that core
Attachment 1, Volume 6, Rev. 0, Page 42 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.2 BASES, CORE REACTIVITY
 
: 1. ISTS 3.1.2 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.
reactivity is behaving as expected. If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may be renormalized and power operation may continue. If operational restriction or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined.  
Therefore, ITS 3.1.2 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
Subsequent reference numbers have been renumbered.
: 2. ITS 3.1.2 ACTION B requires that when the Required Action and associated Completion Time of Condition A is not met, to be in MODE 3 within 6 hours. ISTS 3.1.2 Bases statement for ACTION B.1 which states "if the core reactivity cannot be restored to within the 1% k/k limit" has been changed to match the wording contained in Condition B. Furthermore, ACTION A contains two Required Actions, and the "restore" words in the ISTS Bases are not consistent with either of the Required Actions. Therefore, changes have been made to be consistent with the Specification.
: 3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 4. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 5. Change made to be consistent with the actual wording of the Note in ISTS SR 3.1.2.1. Also the proposed wording (use of "may") is consistent with a similar description concerning normalization in the Applicable Safety Analyses section of the Bases.
: 6. Clarification has been added since all the reactor safety analyses are not in USAR Chapter 14.
Kewaunee Power Station                  Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 42 of 213


The required Completion Time of 7 days is adequate for preparing whatever operating restrictions or Surveillances that may be required to allow continued reactor operation.  
Attachment 1, Volume 6, Rev. 0, Page 43 of 213 Specific No Significant Hazards Considerations (NSHCs)
Attachment 1, Volume 6, Rev. 0, Page 43 of 213


Core Reactivity B 3.1.2   WOG B 3.1.2-5 Rev. 3.0, 03/31/04  BASES ACTIONS  (continued)
Attachment 1, Volume 6, Rev. 0, Page 44 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.2, CORE REACTIVITY There are no specific NSHC discussions for this Specification.
Kewaunee Power Station              Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 44 of 213


B.1 If the core reactivity cannot be restored to within the 1% k/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 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 plant systems. If any Required Action and associated Completion Time is not met 1 2 4 3 5 SURVEILLANCE SR  3.1.2.1 REQUIREMENTS 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 control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization of predicted core reactivity to the measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1, is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (QPTR, AFD, etc.) for prompt indication of an anomaly.
Attachment 1, Volume 6, Rev. 0, Page 45 of 213 ATTACHMENT 3 ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)
may REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29. 
Attachment 1, Volume 6, Rev. 0, Page 45 of 213
: 2. FSAR, Chapter [15].
  , Volume 6, Rev. 0, Page 46 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 46 of 213
USAR, Section 3.1.2.3, GDC 27, "Redundancy of Reactivity Control." 14 U 5  2. USAR, Section 3.1.2.4, GDC 28, "Reactivity Hot Shutdown Capability."  3. USAR, Section 3.1.2.5, GDC 29, "Reactivity Shutdown Capability."  4. USAR, Section 3.1.2.6, GDC 30, "Reactivity Holddown Capability."
JUSTIFICATION FOR DEVIATIONS ITS 3.1.2 BASES, CORE REACTIVITY
: 1. ISTS 3.1.2 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A. Therefore, ITS 3.1.2 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
Subsequent reference numbers have been renumbered.
: 2. ITS 3.1.2 ACTION B requires that when the Required Action and associated Completion Time of Condition A is not met, to be in MODE 3 within 6 hours. ISTS


====3.1.2 Bases====
Attachment 1, Volume 6, Rev. 0, Page 47 of 213 ITS                                                       A01                                                ITS 3.1.3
statement for ACTION B.1 which states "if the core reactivity cannot be restored to within the 1% k/k limit-" has been changed to match the wording contained in Condition B. Furthermore, ACTION A contains two Required Actions, and the "restore" words in the ISTS Bases are not consistent with either of the Required Actions. Therefore, changes have been made to be consistent with the Specification.
: f. Minimum Conditions for Criticality
: 3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 1. The reactor shall not be brought to a critical condition until the pressure-temperature                 See ITS state is to the right of the criticality limit line shown in Figure TS 3.1-1.                             3.4.3
: 4. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 2. The reactor shall be maintained subcritical by at least 1% k/k until normal water                       See ITS level is established in the pressurizer.                                                                 3.4.9 Applicability                                                                                                                  M01
: 5. Change made to be consistent with the actual wording of the Note in ISTS SR 3.1.2.1. Also the proposed wording (use of "may") is consistent with a similar description concerning normalization in the Applicable Safety Analyses section of the Bases. 
: 3. When the reactor is critical the moderator temperature coefficient shall be as specified in the COLR, except during LOW POWER PHYSICS TESTING. The                                     See ITS 3.1.8 LCO 3.1.3          maximum upper moderator temperature coefficient limit shall be  5 pcm/&deg;F for power levels  60% RATED POWER and  0 pcm/&deg;F for power levels > 60% RATED POWER.
: 6. Clarification has been added since all the reactor safety analyses are not in USAR Chapter 14.
: 4. If the limits of 3.1.f.3 cannot be met, then power operation may continue provided the following actions are taken:
Kewaunee Power Station Page 1 of 1 Specific No Significant Haza rds Considerations (NSHCs)
ACTION A            A. Within 24 hours, develop and maintain administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in TS 3.1.f.3. These withdrawal limits shall be in addition to the                 LA01 insertion limits specified in TS 3.10.d.
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.2, CORE REACTIVITY There are no specific NSHC discussions for this Specification.
ACTION B B. If the actions specified in TS 3.1.f.4.A are not satisfied, then be in HOT                             M02 STANDBY within the next 6 hours.
Kewaunee Power Station Page 1 of 1 ATTACHMENT 3 ITS 3.1.3, MODERATOR TEMP ERATURE COEFFICIENT (MTC)
Add proposed ACTION C             M01 Add proposed SR 3.1.3.1 and SR 3.1.3.2         M03 Amendment No. 165 TS 3.1-10                                                 03/11/2003 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 47 of 213
Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
A01 ITS 3.1.3ITS f. Minimum Conditions for Criticality  
: 1. The reactor shall not be brought to a critical condition until the pressure-temperature state is to the right of the criticality limit line shown in Figure TS 3.1-1.  
: 2. The reactor shall be maintained subcritical by at least 1% k/k until normal water level is established in the pressurizer.  
: 3. When the reactor is critical the moderator temperature coefficient shall be as specified in the COLR, except during LOW POWER PHYSICS TESTING. The maximum upper moderator temperature coefficient limit shall be  5 pcm/ F for power levels  60% RATED POWER and  0 pcm/ F for power levels 60% RATED POWER. LCO 3.1.3 See ITS 3.1.8 M01 A pplicabilit y See ITS 3.4.9 See ITS 3.4.3 4. If the limits of 3.1.f.3 cannot be met, then power operation may continue provided the following actions are taken:
A. Within 24 hours, develop and maintain administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in TS 3.1.f.3. These withdrawal limits shall be in addition to the  
 
insertion limits specified in TS 3.10.d.
B. If the actions specified in TS 3.1.f.4.A are not satisfied, then be in HOT STANDBY within the next 6 hours.
LA01M02 A CTION A A CTION B  M01Add proposed ACTION C M03Add proposed SR 3.1.3.1 and SR 3.1.3.2 Amendment No. 165 TS 3.1-10 03/11/2003 Page 1 of 1 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)
ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).


Attachment 1, Volume 6, Rev. 0, Page 48 of 213 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)
ADMINISTRATIVE CHANGES A01  In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
MORE RESTRICTIVE CHANGES
MORE RESTRICTIVE CHANGES M01 CTS 3.1.f.3 requires, in part, that the moderator temperature coefficient (MTC) shall be as specified in the COLR when the reactor is critical. The COLR provides both an upper MTC limit and a lower MTC limit. If an MTC limit is not met, CTS 3.1.f.4 provides 24 hours to develop and maintain administrative control rod withdrawal limits sufficient to restore the MTC to within the limits.
 
ITS 3.1.3 requires the upper MTC Limit to be met in MODE 1 and MODE 2 with keff 1.0 and the lower MTC limit to be met in MODES 1, 2, and 3. Furthermore, ITS 3.1.3 does not provide any time to restore the lower MTC to within limits. If the lower MTC limit is not met, ITS 3.1.3 ACTION C requires the unit to be in MODE 4 within 12 hours, which places the unit outside the new proposed Applicability. This changes the CTS by requiring the lower MTC limit to be met in MODE 3 and when it cannot be met, to be in MODE 4 within 12 hours.
M01 CTS 3.1.f.3 requires, in part, that the moderator temperature coefficient (MTC) shall be as specified in the COLR when the reactor is critical. The COLR provides both an upper MTC limit and a lower MTC limit. If an MTC limit is not met, CTS 3.1.f.4 provides 24 hours to develop and maintain administrative control rod withdrawal limits sufficient to restore the MTC to within the limits.
This change is acceptable because the lower MTC limit is required in MODE 3, in addition to MODE 2, to ensure that core overcooling accidents will not violate the assumptions in the safety analyses. This change is designated as more restrictive because it expands the Applicability for the lower MTC limit and requires an immediate shutdown to MODE 4 in lieu of 24 hours to restore MTC to within its lower limit prior to requiring a unit shutdown.
ITS 3.1.3 requires the upper MTC Limit to be met in MODE 1 and MODE 2 with  
M02 CTS 3.10.f.4.B requires that if the development and maintenance of administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in the COLR can not be satisfied within 24 hours, then to be in HOT STANDBY within 6 hours. The reactivity condition (k/k) for HOT STANDBY, as defined in CTS 1.0.j, is
 
    < 0.25%. Under similar conditions (i.e., if the establishment of administrative withdrawal limits for control banks to maintain MTC within limit is not met within 24 hours) ITS 3.1.3 ACTION B requires the unit to be in MODE 2 with keff < 1.0 within 6 hours. This changes the CTS by requiring the unit to be in MODE 2 with keff < 1.0 (i.e., k/k < 0.0%) in lieu of k/k < 0.25%.
k eff 1.0 and the lower MTC limit to be met in MODES 1, 2, and 3. Furthermore, ITS 3.1.3 does not provide any time to restore the lower MTC to within limits. If the lower MTC limit is not met, ITS 3.1.3 ACTION C requires the unit to be in MODE 4 within 12 hours, which places the unit outside the new proposed Applicability. This changes the CTS by requiring the lower MTC limit to be met in MODE 3 and when it cannot be met, to be in MODE 4 within 12 hours.
The purpose of CTS 3.10.f.4.B is to place the unit in a condition in which it does not rely on the MTC upper limit. In ITS 3.1.3, it is necessary to place the unit in MODE 2 with keff < 1.0. At this point, the unit will be outside of the MODE of Applicability for upper MTC limit (MODE 1 and MODE 2 with keff 1.0). This change is designated as more restrictive because the ITS requires the unit to be Kewaunee Power Station                   Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 48 of 213
This change is acceptable because the lower MTC limit is required in MODE 3, in addition to MODE 2, to ensure that core overcooling accidents will not violate the assumptions in the safety analyses. This change is designated as more restrictive because it expands the Applicability for the lower MTC limit and requires an immediate shutdown to MODE 4 in lieu of 24 hours to restore MTC to within its lower limit prior to requiring a unit shutdown.
 
M02 CTS 3.10.f.4.B requires that if the development and maintenance of administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in the COLR can not be satisfied within 24 hours, then to be in HOT STANDBY within 6 hours. The reactivity condition (k/k) for HOT STANDBY, as defined in CTS 1.0.j, is < 0.25%. Under similar conditions (i.e., if the establishment of administrative withdrawal limits for control banks to maintain MTC within limit is not met within 24 hours) ITS 3.1.3 ACTION B requires the unit to be in MODE 2 with k eff < 1.0 within 6 hours. This changes the CTS by requiring the unit to be in MODE 2 with  
 
k eff < 1.0 (i.e., k/k < 0.0%) in lieu of k/k < 0.25%.  
 
The purpose of CTS 3.10.f.4.B is to place the unit in a condition in which it does not rely on the MTC upper limit. In ITS 3.1.3, it is necessary to place the unit in MODE 2 with k eff < 1.0. At this point, the unit will be outside of the MODE of Applicability for upper MTC limit (MODE 1 and MODE 2 with k eff 1.0). This change is designated as more restrictive because the ITS requires the unit to be Kewaunee Power Station Page 1 of 3 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) made subcritical (i.e., MODE 2 with k eff < 1.0, which is the same as k/k < 0.0%) in lieu of the CTS requirement to just be in HOT STANDBY (k/k < 0.25%).
M03 CTS 3.1.f.3 does not provide any Surveillance Requirements for verifying that the MTC is within the limits specified in the COLR. ITS SR 3.1.3.1 requires verification that the MTC is within the upper limit prior to entering MODE 1 after each refueling. ITS SR 3.1.3.2 requires that the MTC is within the lower limit once each cycle. This changes the CTS by adding new Surveillance Requirements to verify the upper and lower MTC are within limits.
 
This change is acceptable because the added Surveillance Requirements verify that the MTC requirements will be met in MODE 1 and MODE 2 with k eff  1.0 for the upper limit and MODES 1, 2, and 3 for the lower limit. This change is designated as more restrictive because new Surveillance Requirements are being added to the ITS that are not included in the CTS.
 
RELOCATED SPECIFICATIONS
 
None REMOVED DETAIL CHANGES
 
LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements)  CTS 3.10.f.4.A requires, in part, the development and maintenance of administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in the COLR.
Additionally, CTS 3.10.f.4.A states that the withdrawal limits shall be in addition to the insertion limits specified in rod insertion limits. ITS 3.1.3 Required Action A.1 requires the establishment of withdrawal limits for control banks to maintain MTC within limits. This changes the CTS by moving the statement that these withdrawal limits shall be in addition to the insertion limits specified in rod
 
insertion limits to the Bases.


Attachment 1, Volume 6, Rev. 0, Page 49 of 213 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) made subcritical (i.e., MODE 2 with keff < 1.0, which is the same as k/k < 0.0%)
in lieu of the CTS requirement to just be in HOT STANDBY (k/k < 0.25%).
M03  CTS 3.1.f.3 does not provide any Surveillance Requirements for verifying that the MTC is within the limits specified in the COLR. ITS SR 3.1.3.1 requires verification that the MTC is within the upper limit prior to entering MODE 1 after each refueling. ITS SR 3.1.3.2 requires that the MTC is within the lower limit once each cycle. This changes the CTS by adding new Surveillance Requirements to verify the upper and lower MTC are within limits.
This change is acceptable because the added Surveillance Requirements verify that the MTC requirements will be met in MODE 1 and MODE 2 with keff  1.0 for the upper limit and MODES 1, 2, and 3 for the lower limit. This change is designated as more restrictive because new Surveillance Requirements are being added to the ITS that are not included in the CTS.
RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.10.f.4.A requires, in part, the development and maintenance of administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in the COLR.
Additionally, CTS 3.10.f.4.A states that the withdrawal limits shall be in addition to the insertion limits specified in rod insertion limits. ITS 3.1.3 Required Action A.1 requires the establishment of withdrawal limits for control banks to maintain MTC within limits. This changes the CTS by moving the statement that these withdrawal limits shall be in addition to the insertion limits specified in rod insertion limits to the Bases.
The removal of these details for meeting the Required Action from the Technical Specification is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to establish withdrawal limits for control banks to maintain MTC within limits. Also, this change is acceptable because these type of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This change is designated as less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.
The removal of these details for meeting the Required Action from the Technical Specification is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to establish withdrawal limits for control banks to maintain MTC within limits. Also, this change is acceptable because these type of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This change is designated as less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.
Kewaunee Power Station Page 2 of 3 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)
Kewaunee Power Station                 Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 49 of 213
Kewaunee Power Station Page 3 of 3 LESS RESTRICTIVE CHANGES None Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
MTC 3.1.3  WOG STS 3.1.3-1 Rev. 3.0, 03/31/04  CTS 3.1  REACTIVITY CONTROL SYSTEMS
 
====3.1.3 Moderator====
Temperature Coefficient (MTC)
 
LCO  3.1.3 The MTC shall be maintained within the limits specified in the COLR. The maximum upper limit shall be [ [  ] k/k&deg;F at hot zero power] [that specified in Figure 3.1.3-1].
5 pcm/&#xba;F with THERMAL POWER  60% RTP and  0 pcm/&#xba;F with THERMAL POWER > 60% RTP 3.1.f.3 1 APPLICABILITY: MODE 1 and MODE 2 with k eff  1.0 for the upper MTC limit, 3.1.f.3 MODES 1, 2, and 3 for the lower MTC limit.


ACTIONS  CONDITION REQUIRED ACTION COMPLETION TIME A. MTC not within upper limit.
Attachment 1, Volume 6, Rev. 0, Page 50 of 213 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)
A.1 Establish administrative withdrawal limits for control
LESS RESTRICTIVE CHANGES None Kewaunee Power Station        Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 50 of 213


banks to maintain MTC within limit.
Attachment 1, Volume 6, Rev. 0, Page 51 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
24 hours B. Required Action and associated Completion
Attachment 1, Volume 6, Rev. 0, Page 51 of 213


Attachment 1, Volume 6, Rev. 0, Page 52 of 213 CTS                                                                                                              MTC 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3        Moderator Temperature Coefficient (MTC) 3.1.f.3  LCO 3.1.3              The MTC shall be maintained within the limits specified in the COLR. The maximum upper limit shall be [ [ ] k/k&deg;F at hot zero power] [that                  1 specified in Figure 3.1.3-1].
5 pcm/&#xba;F with THERMAL POWER  60% RTP and 0 pcm/&#xba;F with THERMAL POWER > 60% RTP 3.1.f.3  APPLICABILITY:        MODE 1 and MODE 2 with keff  1.0 for the upper MTC limit, MODES 1, 2, and 3 for the lower MTC limit.
ACTIONS CONDITION                      REQUIRED ACTION                        COMPLETION TIME 3.1.f.4,  A. MTC not within upper          A.1    Establish administrative              24 hours 3.1.f.4.A limit.                              withdrawal limits for control banks to maintain MTC within limit.
3.1.f.4.b  B. Required Action and            B.1    Be in MODE 2 with                    6 hours associated Completion                keff < 1.0.
Time of Condition A not met.
Time of Condition A not met.
B.1 Be in MODE 2 with k eff < 1.0.
DOC M01    C. MTC not within lower           C.1     Be in MODE 4.                         12 hours limit.
6 hours  C. MTC not within lower limit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                             FREQUENCY DOC M03    SR 3.1.3.1         Verify MTC is within upper limit.                               Prior to entering MODE 1 after each refueling WOG STS                                        3.1.3-1                              Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 52 of 213
C.1 Be in MODE 4.  
 
12 hours 3.1.f.4, 3.1.f.4.A 3.1.f.4.b DOC M01    SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY  
 
SR 3.1.3.1 Verify MTC is within upper limit.  


Prior to entering MODE 1 after each refueling
Attachment 1, Volume 6, Rev. 0, Page 53 of 213 CTS                                                                                                        MTC 3.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                                FREQUENCY DOC M03  SR 3.1.3.2    ---------------------------NOTES--------------------------------
: 1. Not required to be performed until 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.
: 2. If the MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle.
: 3. SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of  60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR.
Verify MTC is within lower limit.                                    Once each cycle WOG STS                                        3.1.3-2                                Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 53 of 213


DOC M03 MTC 3.1.3   WOG STS 3.1.3-2 Rev. 3.0, 03/31/04  CTS SURVEILLANCE REQUIREMENTS (continued)
Attachment 1, Volume 6, Rev. 0, Page 54 of 213 MTC 3.1.3
SURVEILLANCE FREQUENCY SR  3.1.3.2 ---------------------------NOTES--------------------------------  
    ~    .9~--+---~--~--~----~--+---~--~~~--~
: 1. Not required to be performed until 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.  
  ~1l.L
: 2. If the MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle.
  ~ o
: 3. SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of  60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR.    ---------------------------------------------------------------------  
  <I        ~__~__~__~__~__- +__- 4_ _ _ UNACCf PTABLE __
          .8                                                            OPERATION C>
I-Z W
          *7  1-------+----11-------+-----+-----+-----+-------+-\------+-----+_
o LL LL
          *6 i-----------+----Il----------+----+------+--------+------+-_4______++_
                                                                          \ ~
W                                ACCEPTABLE o                                  OPERATION o    .5~--+---n---~--~----~--+---~--~\+---~--~
w a:
I-
    <<
* 4 f-----+----++-----+--f-----+------+-----+-------+-I--4-I\---+-_
a:
w 0...
:::2:
w I-
* 3 f-----+--+-+-----+----+-+----+----+-----+-+--\-+--\~                  \
a:    .2 _TH!S F! ~ URE FOR !LLUSTRAT!ON ONL y~--+--+-__~~~~
o I-                 DO NOT USE FOR OPERATION                                          \
a:
w o    *1  !-----------+--t-----+---t-----+---t-------+---+--t-----------+--\                        o 1
o
:::2:
o0        10      20      30      40      50      60      70        80    90      100 PERCENT RTP Figure 3.1.3 - 1 (page 1 of 1)
Moderator Temperature Coefficient Vs. Rated Thermal Power WOG STS                                        3.1.3-3                              Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 54 of 213


Verify MTC is within lower limit.
Attachment 1, Volume 6, Rev. 0, Page 55 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)
 
Once each cycle DOC M03 MTC 3.1.3  WOG STS 3.1.3-3 Rev. 3.0, 03/31/04 
 
Figure 3.1.3 - 1 (page 1 of 1) Moderator Temperature Coefficient Vs. Rated Thermal Power 1<I C> Z W o LL LL W o o w a: ::::> << a: w 0... :::2: w I-a: o << a: w o o :::2: __ __ __ __ __ -+ __ -4 ___ UNACCf PTABLE __ .8 OPERATION
* 7 1-------+----11-------+-----+-----+-----+-------+-\
------+-----+
_
* 6 i-----------+--
--Il----------+
----+------+--------+------+-
_4______++_
\ ACCEPTABLE OPERATION
* 4 f-----+----++-----+--f-----+------+-----+-
------+-I--4-I\---+-_
\
* 3 70 80 90 100 .2 _TH!S F! FOR !LLUSTRAT!ON ONL __
DO NOT USE FOR OPERATION
\
* 1 !-----------+--
t-----+---t-----+---
t-------+---+--
t-----------+--\
o 0 10 20 30 40 50 60 PERCENT RTP o JUSTIFICATION FOR DEVIATIONS ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)  
: 1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
Additionally, since ISTS Figure 3.1.3-1 is a bracketed value in ISTS LCO 3.1.3 and it is not being used, ISTS Figure 3.1.3-1 is deleted.
Additionally, since ISTS Figure 3.1.3-1 is a bracketed value in ISTS LCO 3.1.3 and it is not being used, ISTS Figure 3.1.3-1 is deleted.
Kewaunee Power Station Page 1 of 1 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Kewaunee Power Station               Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 55 of 213
MTC B 3.1.3    WOG STS B 3.1.3-1 Rev. 3.0, 03/31/04 B 3.1  REACTIVITY CONTROL SYSTEMS


B 3.1.3  Moderator Temperature Coefficient (MTC)  
Attachment 1, Volume 6, Rev. 0, Page 56 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 56 of 213


BASES BACKGROUND According to GDC 11 (Ref. 1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases. INSERT 1 1 The MTC relates a change in core reactivity to a change in reactor coolant temperature (a positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature). The reactor is designed to operate with a negative MTC over the largest possible range of fuel cycle operation. Therefore, a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result.  
Attachment 1, Volume 6, Rev. 0, Page 57 of 213 MTC B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Moderator Temperature Coefficient (MTC)
BASES BACKGROUND           According to GDC 11 (Ref. 1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable 1
INSERT 1    power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.
The MTC relates a change in core reactivity to a change in reactor coolant temperature (a positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature). The reactor is designed to operate with a negative MTC over the largest possible range of fuel cycle operation. Therefore, a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result.
MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by The measurements. Both initial and reload cores are designed so that the          2 beginning of cycle (BOC) MTC is less than zero when THERMAL POWER is at RTP. The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons to yield an MTC at BOC within the range analyzed in the plant accident analysis. The end of cycle (EOC) MTC is also limited by the requirements of the accident analysis. Fuel cycles that are designed to        2 achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the EOC limit.
The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed in the FSAR        2 accident and transient analyses.                                        U WOG STS                                    B 3.1.3-1                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 57 of 213


MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by measurements. Both initial and reload cores are designed so that the beginning of cycle (BOC) MTC is less than zero when THERMAL POWER is at RTP. The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons to yield an MTC at BOC within the range analyzed in the plant accident analysis. The end of cycle (EOC) MTC is also limited by the requirements of the accident analysis. Fuel cycles that are designed to achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the EOC limit.
Attachment 1, Volume 6, Rev. 0, Page 58 of 213 B 3.1.3 o 1
The 2 2 The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed in the FSAR accident and transient analyses.
                                    -INSERT 1 The design of the reactor core with its related controls and protection systems shall ensure that power oscillations, the magnitude of which could cause damage in excess of acceptable fuel damage limits, are not possible or can be readily suppressed (Ref. 1).
2 U B 3.1.3 1 INSERT 1   The design of the reactor core with its related controls and protection systems shall ensure that power oscillations, the magnitude of which could cause damage in excess of acceptable fuel damage limits, are not possible or can be readily suppressed (Ref. 1).
Insert Page B 3.1.3-1 Attachment 1, Volume 6, Rev. 0, Page 58 of 213
Insert Page B 3.1.3-1 o -
MTC B 3.1.3    WOG STS B 3.1.3-2 Rev. 3.0, 03/31/04  BASES BACKGROUND  (continued)
 
If the LCO limits are not met, the unit response during transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity.


Attachment 1, Volume 6, Rev. 0, Page 59 of 213 MTC B 3.1.3 BASES BACKGROUND (continued)
If the LCO limits are not met, the unit response during transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity.
The SRs for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits, since this coefficient changes slowly, due principally to the reduction in RCS boron concentration associated with fuel burnup.
The SRs for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits, since this coefficient changes slowly, due principally to the reduction in RCS boron concentration associated with fuel burnup.
APPLICABLE The acceptance criteria for the specified MTC are:  
APPLICABLE     The acceptance criteria for the specified MTC are:
 
SAFETY ANALYSES       a. The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2) and                                               3
SAFETY   ANALYSES a. The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2) and  
: b. The MTC must be such that inherently stable power operations result during normal operation and accidents, such as overheating and overcooling events.
: b. The MTC must be such that inherently stable power operations result during normal operation and accidents, such as overheating and  
U                        14 The FSAR, Chapter 15 (Ref. 2), contains analyses of accidents that result           2 in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref. 3).
 
The consequences of accidents that cause core overheating must be evaluated when the MTC is positive. Such accidents include the rod withdrawal transient from either zero (Ref. 4) or RTP, loss of main         normal 2 feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature.
overcooling events.  
WOG STS                                 B 3.1.3-2                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 59 of 213
;3 The FSAR, Chapter 15 (Ref. 2), contains analyses of accidents that result in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are  
 
bounding (Ref. 3).
14 U 2 The consequences of accidents that cause core overheating must be evaluated when the MTC is positive. Such accidents include the rod withdrawal transient from either zero (Ref. 4) or RTP, loss of main feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature.
2 normal MTC B 3.1.3    WOG STS B 3.1.3-3 Rev. 3.0, 03/31/04 BASES APPLICABLE SAFETY ANALYSES  (continued)


Attachment 1, Volume 6, Rev. 0, Page 60 of 213 MTC B 3.1.3 BASES APPLICABLE SAFETY ANALYSES (continued)
In order to ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether it is the BOC or EOC life. The most conservative combination appropriate to the accident is then used for the analysis (Ref. 2).
In order to ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether it is the BOC or EOC life. The most conservative combination appropriate to the accident is then used for the analysis (Ref. 2).
MTC values are bounded in reload safety evaluations assuming steady state conditions at BOC and EOC. An EOC measurement is conducted at conditions when the RCS boron concentration reaches approximately 300 ppm. The measured value may be extrapolated to project the EOC value, in order to confirm reload design predictions.
MTC values are bounded in reload safety evaluations assuming steady state conditions at BOC and EOC. An EOC measurement is conducted at conditions when the RCS boron concentration reaches approximately 300 ppm. The measured value may be extrapolated to project the EOC value, in order to confirm reload design predictions.
4MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed and controlled from the control room, MTC is considered an initial condition process variable because of its dependence on boron concentration.
MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not           4 directly observed and controlled from the control room, MTC is considered an initial condition process variable because of its dependence on boron concentration.
LCO LCO 3.1.3 requires the MTC to be within specified limits of the COLR to ensure that the core operates within the assumptions of the accident analysis. During the reload core safety evaluation, the MTC is analyzed to determine that its values remain within the bounds of the original accident analysis during operation.  
LCO                         LCO 3.1.3 requires the MTC to be within specified limits of the COLR to ensure that the core operates within the assumptions of the accident analysis. During the reload core safety evaluation, the MTC is analyzed to determine that its values remain within the bounds of the original accident analysis during operation.
 
Assumptions made in safety analyses require that the MTC be less positive than a given upper bound and more positive than a given lower           near    2 bound. The MTC is most positive at BOC; this upper bound must not be exceeded. This maximum upper limit occurs at BOC, all rods out (ARO),
Assumptions made in safety analyses require that the MTC be less positive than a given upper bound and more positive than a given lower bound. The MTC is most positive at BOC; this upper bound must not be exceeded. This maximum upper limit occurs at BOC, all rods out (ARO), hot zero power conditions. At EOC the MTC takes on its most negative value, when the lower bound becomes important. This LCO exists to ensure that both the upper and lower bounds are not exceeded.
hot zero power conditions. At EOC the MTC takes on its most negative value, when the lower bound becomes important. This LCO exists to ensure that both the upper and lower bounds are not exceeded.
2near During operation, therefore, the conditions of the LCO can only be ensured through measurement. The Surveillance checks at BOC and EOC on MTC provide confirmation that the MTC is behaving as anticipated so that the acceptance criteria are met.
are Core models provide                                                                              additionally predicted values of MTC    During operation, therefore, the conditions of the LCO can only be                     2 that validate the upper and lower MTC limits in ensured through measurement. The Surveillance checks at BOC and the reload safety      EOC on MTC provide confirmation that the MTC is behaving as evaluation process for each core reload. The anticipated so that the acceptance criteria are met.
are additionallyCore models provide predicted values of MTC that validate the upper and lower MTC limits in the reload safety evaluation process for each core reload. The 2
WOG STS                                             B 3.1.3-3                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 60 of 213
MTC B 3.1.3    WOG STS B 3.1.3-4 Rev. 3.0, 03/31/04 BASES LCO  (continued)


The LCO establishes a maximum positive value that cannot be exceeded. The BOC positive limit and the EOC negative limit are established in the COLR to allow specifying limits for each particular cycle. This permits the unit to take advantage of improved fuel management and changes in unit operating schedule. (lower) (upper) 5 APPLICABILITY Technical Specifications place both LCO and SR values on MTC, based on the safety analysis assumptions described above.
Attachment 1, Volume 6, Rev. 0, Page 61 of 213 MTC B 3.1.3 BASES LCO (continued)
(upper)
The LCO establishes a maximum positive value that cannot be exceeded.
The BOC positive limit and the EOC negative limit are established in the       5 (lower)
COLR to allow specifying limits for each particular cycle. This permits the unit to take advantage of improved fuel management and changes in unit operating schedule.
APPLICABILITY                   Technical Specifications place both LCO and SR values on MTC, based on the safety analysis assumptions described above.
In MODE 1, the limits on MTC must be maintained to ensure that any accident initiated from THERMAL POWER operation will not violate the design assumptions of the accident analysis. In MODE 2 with the reactor critical, the upper limit must also be maintained to ensure that startup and subcritical accidents (such as the uncontrolled control rod assembly or group withdrawal) will not violate the assumptions of the accident analysis. The lower MTC limit must be maintained in MODES 2 and 3, in addition to MODE 1, to ensure that cooldown accidents will not violate the assumptions of the accident analysis. In MODES 4, 5, and 6, this LCO is not applicable, since no Design Basis Accidents using the MTC as an analysis assumption are initiated from these MODES.
In MODE 1, the limits on MTC must be maintained to ensure that any accident initiated from THERMAL POWER operation will not violate the design assumptions of the accident analysis. In MODE 2 with the reactor critical, the upper limit must also be maintained to ensure that startup and subcritical accidents (such as the uncontrolled control rod assembly or group withdrawal) will not violate the assumptions of the accident analysis. The lower MTC limit must be maintained in MODES 2 and 3, in addition to MODE 1, to ensure that cooldown accidents will not violate the assumptions of the accident analysis. In MODES 4, 5, and 6, this LCO is not applicable, since no Design Basis Accidents using the MTC as an analysis assumption are initiated from these MODES.
ACTIONS A.1 If the BOC MTC limit is violated, administrative withdrawal limits for control banks must be established to maintain the MTC within its limits.
ACTIONS                         A.1 upper These withdrawal limits shall  If the BOC MTC limit is violated, administrative withdrawal limits for         5 be in addition to the insertion limits specified in LCO 3.1.6, control banks must be established to maintain the MTC within its limits.         6 "Control Bank Insertion      The MTC becomes more negative with control bank insertion and Limits."
The MTC becomes more negative with control bank insertion and decreased boron concentration. A Completion Time of 24 hours provides enough time for evaluating the MTC measurement and computing the required bank withdrawal limits. These withdrawal limits shall be in addition to the insertion limits specified in LCO 3.1.6, "Control Bank Insertion Limits." upper 6 5 As cycle burnup is increased, the RCS boron concentration will be reduced. The reduced boron concentration causes the MTC to become more negative. Using physics calculations, the time in cycle life at which the calculated MTC will meet the LCO requirement can be determined. At this point in core life Condition A no longer exists. The unit is no longer  
decreased boron concentration. A Completion Time of 24 hours provides enough time for evaluating the MTC measurement and computing the required bank withdrawal limits.
 
As cycle burnup is increased, the RCS boron concentration will be reduced. The reduced boron concentration causes the MTC to become more negative. Using physics calculations, the time in cycle life at which the calculated MTC will meet the LCO requirement can be determined.
in the Required Action, so the administrative withdrawal limits are no longer in effect.
At this point in core life Condition A no longer exists. The unit is no longer in the Required Action, so the administrative withdrawal limits are no longer in effect.
MTC B 3.1.3    WOG STS B 3.1.3-5 Rev. 3.0, 03/31/04 BASES ACTIONS  (continued)
WOG STS                                                   B 3.1.3-4                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 61 of 213


B.1 If the required administrative withdrawal limits at BOC are not established within 24 hours, the unit must be brought to MODE 2 with k eff < 1.0 to prevent operation with an MTC that is more positive than that assumed in safety analyses.
Attachment 1, Volume 6, Rev. 0, Page 62 of 213 MTC B 3.1.3 BASES ACTIONS (continued)
B.1 If the required administrative withdrawal limits at BOC are not established within 24 hours, the unit must be brought to MODE 2 with keff < 1.0 to prevent operation with an MTC that is more positive than that assumed in safety analyses.
The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
C.1 Exceeding the EOC MTC limit means that the safety analysis assumptions for the EOC accidents that use a bounding negative MTC value may be invalid. If the EOC MTC limit is exceeded, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 4 within 12 hours. lower lower 5 5 The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
C.1 lower Exceeding the EOC MTC limit means that the safety analysis                   5 lower  assumptions for the EOC accidents that use a bounding negative MTC value may be invalid. If the EOC MTC limit is exceeded, the plant must       5 be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 4 within 12 hours.
SURVEILLANCE SR 3.1.3.1 REQUIREMENTS This SR requires measurement of the MTC at BOC prior to entering MODE 1 in order to demonstrate compliance with the most positive MTC LCO. Meeting the limit prior to entering MODE 1 ensures that the limit will also be met at higher power levels.  
The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE       SR 3.1.3.1 REQUIREMENTS This SR requires measurement of the MTC at BOC prior to entering MODE 1 in order to demonstrate compliance with the most positive MTC LCO. Meeting the limit prior to entering MODE 1 ensures that the limit will also be met at higher power levels.
upper The BOC MTC value for ARO will be inferred from isothermal                  5 temperature coefficient measurements obtained during the physics tests 2
after refueling. The ARO value can be directly compared to the BOC MTC limit of the LCO. If required, measurement results and predicted design values can be used to establish administrative withdrawal limits for control banks.
WOG STS                                  B 3.1.3-5                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 62 of 213


The BOC MTC value for ARO will be inferred from isothermal temperature coefficient measurements obtained during the physics tests after refueling. The ARO value can be directly compared to the BOC MTC limit of the LCO. If required, measurement results and predicted design values can be used to establish administrative withdrawal limits for
Attachment 1, Volume 6, Rev. 0, Page 63 of 213 MTC B 3.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)
 
SR 3.1.3.2 In similar fashion, the LCO demands that the MTC be less negative than the specified value for EOC full power conditions. This measurement may be performed at any THERMAL POWER, but its results must be extrapolated to the conditions of RTP and all banks withdrawn in order to make a proper comparison with the LCO value. Because the RTP MTC value will gradually become more negative with further core depletion and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOC LCO limit. The 300 ppm SR value is sufficiently less negative than the EOC LCO limit value to ensure that the LCO limit will be met when the 300 ppm Surveillance criterion is met.
control banks. upper 5 2 MTC B 3.1.3     WOG B 3.1.3-6 Rev. 3.0, 03/31/04  BASES SURVEILLANCE REQUIREMENTS (continued)  
SR 3.1.3.2 is modified by three Notes that include the following requirements:
 
: a. The SR is not required to be performed until 7 effective full power days (EFPDs) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.
SR 3.1.3.2  
 
In similar fashion, the LCO demands that the MTC be less negative than the specified value for EOC full power conditions. This measurement may be performed at any THERMAL POWER, but its results must be extrapolated to the conditions of RTP and all banks withdrawn in order to make a proper comparison with the LCO value. Because the RTP MTC value will gradually become more negative with further core depletion and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOC LCO limit. The 300 ppm SR value is sufficiently less negative than the EOC LCO limit value to ensure that the LCO limit will be met when the 300 ppm Surveillance criterion is met.  
 
SR 3.1.3.2 is modified by three Notes that include the following  
 
requirements:  
: a. The SR is not required to be performed until 7 effective full power days (EFPDs) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.  
: b. If the 300 ppm Surveillance limit is exceeded, it is possible that the EOC limit on MTC could be reached before the planned EOC.
: b. If the 300 ppm Surveillance limit is exceeded, it is possible that the EOC limit on MTC could be reached before the planned EOC.
Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the  
Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the EOC limit.
 
: c. The Surveillance limit for RTP boron concentration of 60 ppm is conservative. If the measured MTC at 60 ppm is more positive than the 60 ppm Surveillance limit, the EOC limit will not be exceeded because of the gradual manner in which MTC changes with core burnup.
EOC limit.  
REFERENCES         1. 10 CFR 50, Appendix A, GDC 11.                                                               3 U                                                      USAR, Section 3.1.2.2, GDC 7, 14                "Suppression of Power Oscillations."
: c. The Surveillance limit for RTP boron concentration of 60 ppm is conservative. If the measured MTC at 60 ppm is more positive than  
: 2. FSAR, Chapter [15].                                                                       3  7 9272-P-A
 
: 3. WCAP 9273-NP-A, "Westinghouse Reload Safety Evaluation                                   3 Methodology," July 1985.
the 60 ppm Surveillance limit, t he EOC limit will not be exceeded because of the gradual manner in which MTC changes with core  
U                                    Section 14.1.1 3  7
 
burnup. REFERENCES 1. 10 CFR 50, Appendix A, GDC 11.
3 USAR, Section 3.1.2.2, GDC 7, "Suppression of Power Oscillations." 2. FSAR, Chapter [15].
U 
: 3. WCAP 9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.  
: 4. FSAR, Chapter [15].
: 4. FSAR, Chapter [15].
14 3 7 9272-P-A 3 U Section 14.1.1 3 7 JUSTIFICATION FOR DEVIATIONS ITS 3.1.3 BASES, MODERATOR TEMPERATURE COEFFICIENT (MTC)
WOG                                        B 3.1.3-6                                  Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 63 of 213
: 1. ISTS 3.1.3 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A. Therefore, ITS 3.1.3 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 3. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
: 4. The Applicable Safety Analyses discussion states that MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). It also says that even though MTC is not directly observed and controlled from the control room, MTC is still considered an initial condition process variable because of its dependence on boron concentration. The additional sentence has been deleted. The NRC Final Policy Statement on Technical Improvements of July 22, 1993 (58 FR 39132) states that process variables captured
 
by Criterion 2 are not limited to only those directly monitored and controlled from the control room. It also states that Criterion 2 includes other features or characteristics that are specifically assumed in Design Basis Accident and Transient analyses even if they cannot be directly observed in the control room (e.g., moderator temperature coefficient and hot channel factors). Since the Final Policy Statement provides guidance on which types of parameters satisfy Criterion 2, there is no reason to


duplicate these words in the KPS CTS.  
Attachment 1, Volume 6, Rev. 0, Page 64 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.3 BASES, MODERATOR TEMPERATURE COEFFICIENT (MTC)
: 5. The ISTS Bases variously refer to the "upper MTC limit," the "BOC MTC limit," the "lower MTC limit," and the "EOC MTC limit." References to the BOC and EOC MTC limit are eliminated and "upper" and "lower" are substituted to eliminate confusion and to be consistent with the Specification.  
: 1. ISTS 3.1.3 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.
: 6. This requirement has been added to the Bases consistent with the requirement in CTS 3.10.f.4.A. This clarifies the ITS 3.1.3 Required Action A.1 requirement to ensure the administrative withdrawal limits are in addition to the normal limits required by LCO 3.1.6, "Control Bank Insertion Limits."
Therefore, ITS 3.1.3 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 3. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
: 4. The Applicable Safety Analyses discussion states that MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). It also says that even though MTC is not directly observed and controlled from the control room, MTC is still considered an initial condition process variable because of its dependence on boron concentration. The additional sentence has been deleted. The NRC Final Policy Statement on Technical Improvements of July 22, 1993 (58 FR 39132) states that process variables captured by Criterion 2 are not limited to only those directly monitored and controlled from the control room. It also states that Criterion 2 includes other features or characteristics that are specifically assumed in Design Basis Accident and Transient analyses even if they cannot be directly observed in the control room (e.g., moderator temperature coefficient and hot channel factors). Since the Final Policy Statement provides guidance on which types of parameters satisfy Criterion 2, there is no reason to duplicate these words in the KPS CTS.
: 5. The ISTS Bases variously refer to the "upper MTC limit," the "BOC MTC limit," the "lower MTC limit," and the "EOC MTC limit." References to the BOC and EOC MTC limit are eliminated and "upper" and "lower" are substituted to eliminate confusion and to be consistent with the Specification.
: 6. This requirement has been added to the Bases consistent with the requirement in CTS 3.10.f.4.A. This clarifies the ITS 3.1.3 Required Action A.1 requirement to ensure the administrative withdrawal limits are in addition to the normal limits required by LCO 3.1.6, "Control Bank Insertion Limits."
: 7. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 7. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
Kewaunee Power Station Page 1 of 1 Specific No Significant Haza rds Considerations (NSHCs)
Kewaunee Power Station                 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 64 of 213
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) There are no specific NSHC discussions for this Specification.
Kewaunee Power Station Page 1 of 1 ATTACHMENT 4 ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
ITS 3.1.4A01ITS e. Rod Misalignment Limitations NOTE: Individual RPIs may be outside their limits for
< 1 hour following substantial rod movement LCO 3.1.4 Note  This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.
 
Rod misalignment limitations do not apply during physics testing.
LA01See ITS 3.1.8 1. When reactor power is  85% of rating, the rod cluster control assembly shall be maintained within +/- 12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating. 2. When reactor power is < 85% but  50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is
< 85% but  50%, the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating. 3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.
LCO 3.1.4 part a M02Add proposed Required Actions B.1.1, B.1.2, and B.2 and associated Completion Times M01 Add proposed Required Actions C.1.1, C.1.2, C.2, C.3, C.4, and C.5 and associated Completion TimesM01Add proposed ApplicabilityM02 LCO 3.1.4
 
part b A CTION B A CTION C  f. Inoperable Rod Position Indicator Channels
 
NOTE: Individual RPIs may be outside their limits for
< 1 hour following substantial rod movement
: 1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate entry condition is allowed for


each inoperable individual rod position indicator.)
Attachment 1, Volume 6, Rev. 0, Page 65 of 213 Specific No Significant Hazards Considerations (NSHCs)
A. Verify the position of the rod cluster control indirectly by movable incore detectors each eight hours, or B. Within 8 hours reduce reactor thermal power to  50% of RATED POWER. 2. If more than one individual rod position indicator channel per group are inoperable, then:
Attachment 1, Volume 6, Rev. 0, Page 65 of 213
A. IMMEDIATELY place the control rods in manual, and B. Once per 1 hour, monitor and record RCS Tavg , and C. Verify the position of the rod by movable incore detectors each 8 hours, and See ITS 3.1.7 Add proposed Required Action D.1M03Add proposed ACTION E M03 LCO 3.1.4 Note See ITS 3.1.7 A CTION D  Amendment No. 181 TS 3.10-5 03/24/2005 Page 1 of 5 ITS 3.1.4A01ITS g. Inoperable Rod Limitations
: 1. An inoperable rod is a rod which does not trip or which is declared inoperable under TS 3.10.e or TS 3.10.h. LA02M01Add proposed LCO 3.1.4 and Applicability
: 2. Not more than one inoperable full length rod shall be allowed at any time.
: 3. If reactor operation is continued with one i noperable full length rod, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days unless the rod is made OPERABLE earlier. The analysis shall include due allowance for nonuniform fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the plant power level shall be reduced to an analytically determined part power level which is consistent with the safety analysis. made OPERABLE earlier. The analysis shall include due allowance for nonuniform fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the plant power level shall be reduced to an analytically determined part power level which is consistent with the safety analysis. 
: h. Rod Drop Time h. Rod Drop Time  At OPERATING temperature and full flow, the drop time of each full length rod cluster control shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage


to dashpot entry. If drop time is > 1.8 seconds, the rod shall be declared inoperable. At OPERATING temperature and full flow, the drop time of each full length rod cluster control shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage
Attachment 1, Volume 6, Rev. 0, Page 66 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)
There are no specific NSHC discussions for this Specification.
Kewaunee Power Station              Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 66 of 213
, Volume 6, Rev. 0, Page 67 of 213 ATTACHMENT 4 ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS , Volume 6, Rev. 0, Page 67 of 213
, Volume 6, Rev. 0, Page 68 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 68 of 213


to dashpot entry. If drop time is > 1.8 seconds, the rod shall be declared inoperable. SR 3.1.4.3 A CTION A Add proposed LCO 3.1.4 and Applicability M04Add proposed Required Actions A.1.1, A.1.2, and A.2 M04M01 A CTION A i. Rod Position Deviation Monitor i. Rod Position Deviation Monitor  If the rod position deviation monitor is inoperable, individual rod positions shall be logged at least once per eight hours and after a load change > 10% of rated power or
Attachment 1, Volume 6, Rev. 0, Page 69 of 213 ITS                                                                                                                              ITS 3.1.4 A01
: e. Rod Misalignment Limitations LCO 3.1.4 NOTE:    Individual RPIs may be outside their limits for < 1 hour following substantial rod movement Note This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog                                            LA01 individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.
Rod misalignment limitations do not apply during physics testing.                                                              See ITS 3.1.8 LCO 3.1.4        1. When reactor power is  85% of rating, the rod cluster control assembly shall be part a                maintained within +/- 12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is ACTION B 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If peaking factors are not M01 determined within four hours, the reactor power shall be reduced to < 85% of rating.
M02 Add proposed Required Actions B.1.1, B.1.2, and B.2 and associated Completion Times LCO 3.1.4         2. When reactor power is < 85% but  50% of rating, the rod cluster control assemblies                                          Add part b                shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control                              proposed Applicability ACTION C assembly is misaligned from its bank by more than +/- 24 steps when reactor power is
                      < 85% but  50%, the rod will be realigned or the core power peaking factors shall be M01 determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating.
Add proposed Required Actions C.1.1, C.1.2, C.2, C.3, C.4, and C.5 and associated Completion Times      M02 ACTION D
: 3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.
Add proposed Required Action D.1          M03
: f. Inoperable Rod Position Indicator Channels                                                      Add proposed ACTION E M03 LCO 3.1.4 NOTE: Individual RPIs may be outside their limits for < 1 hour following substantial rod Note movement                                                                                                                              See ITS 3.1.7
: 1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate entry condition is allowed for each inoperable individual rod position indicator.)
A. Verify the position of the rod cluster control indirectly by movable incore detectors See ITS each eight hours, or                                                                                                      3.1.7 B. Within 8 hours reduce reactor thermal power to  50% of RATED POWER.
: 2. If more than one individual rod position indicator channel per group are inoperable, then:
A. IMMEDIATELY place the control rods in manual, and B. Once per 1 hour, monitor and record RCS Tavg, and C. Verify the position of the rod by movable incore detectors each 8 hours, and Amendment No. 181 TS 3.10-5                                                          03/24/2005 Page 1 of 5 Attachment 1, Volume 6, Rev. 0, Page 69 of 213


after > 24 steps of control rod motion. If the rod position deviation monitor is inoperable, individual rod positions shall be logged at least once per eight hours and after a load change > 10% of rated power or  
Attachment 1, Volume 6, Rev. 0, Page 70 of 213 ITS                                                                                                        ITS 3.1.4 A01
: g. Inoperable Rod Limitations                                      Add proposed LCO 3.1.4 and Applicability        M01
: 1. An inoperable rod is a rod which does not trip or which is declared inoperable under LA02 TS 3.10.e or TS 3.10.h.
: 2. Not more than one inoperable full length rod shall be allowed at any time.                                    M04
: 3. If reactor operation is continued with one inoperable full length rod, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days unless the rod is made OPERABLE earlier. The ACTION A analysis shall include due allowance for nonuniform fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the plant power level shall be reduced to an analytically determined part power level which is consistent with the safety analysis.
Add proposed Required Actions A.1.1, A.1.2, and A.2    M04
: h. Rod Drop Time Add proposed LCO 3.1.4 and Applicability      M01 At OPERATING temperature and full flow, the drop time of each full length rod cluster SR 3.1.4.3      control shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to dashpot entry. If drop time is > 1.8 seconds, the rod shall be declared inoperable.
ACTION A
: i. Rod Position Deviation Monitor If the rod position deviation monitor is inoperable, individual rod positions shall be                       L01 logged at least once per eight hours and after a load change > 10% of rated power or after > 24 steps of control rod motion.
: j. Quadrant Power Tilt Monitor If one or both of the quadrant power tilt monitors is inoperable, individual upper and                      See ITS lower excore detector calibrated outputs and the quadrant tilt shall be logged once per                      3.2.4 shift and after a load change > 10% of rated power or after > 24 steps of control rod motion. The monitors shall be set to alarm at 2% tilt ratio.
: k. Core Average Temperature During steady-state power operation, Tavg, shall be maintained within the limits specified in the COLR, except as provided by TS 3.10.n.
See ITS 3.4.1
: l. Reactor Coolant System Pressure During steady-state power operation, Reactor Coolant System pressure shall be maintained within the limits specified in the COLR, except as provided by TS 3.10.n.
Amendment No. 181 TS 3.10-7                Revised by letter dated 11/03/06 Page 2 of 5 Attachment 1, Volume 6, Rev. 0, Page 70 of 213


after > 24 steps of control rod motion.
Attachment 1, Volume 6, Rev. 0, Page 71 of 213 ITS                                                                                              ITS 3.1.4 A01 4.1 OPERATIONAL SAFETY REVIEW APPLICABILITY Applies to items directly related to safety limits and LIMITING CONDITIONS FOR OPERATION.
L01j. Quadrant Power Tilt Monitor j. Quadrant Power Tilt Monitor  If one or both of the quadrant power tilt monitors is inoperable, individual upper and lower excore detector calibrated outputs and the quadrant tilt shall be logged once per shift and after a load change > 10% of rated power or after > 24 steps of control rod
OBJECTIVE To assure that instrumentation shall be checked, tested, and calibrated, and that equipment and sampling tests shall be conducted at sufficiently frequent intervals to ensure safe operation.
 
motion. The monitors shall be set to alarm at 2% tilt ratio. If one or both of the quadrant power tilt monitors is inoperable, individual upper and lower excore detector calibrated outputs and the quadrant tilt shall be logged once per shift and after a load change > 10% of rated power or after > 24 steps of control rod
 
motion. The monitors shall be set to alarm at 2% tilt ratio. 
: k. Core Average Temperature k. Core Average Temperature  During steady-state power operation, Tavg , shall be maintained within the limits specified in the COLR, except as provided by TS 3.10.n. During steady-state power operation, Tavg , shall be maintained within the limits specified in the COLR, except as provided by TS 3.10.n. l. Reactor Coolant System Pressure l. Reactor Coolant System Pressure  During steady-state power operation, Reactor Coolant System pressure shall be
 
maintained within the limits specified in t he COLR, except as provided by TS 3.10.n. During steady-state power operation, Reactor Coolant System pressure shall be
 
maintained within the limits specified in t he COLR, except as provided by TS 3.10.n. See ITS 3.2.4 See ITS 3.4.1  Amendment No. 181 TS 3.10-7 Revised by letter dated 11/03/06 Page 2 of 5 ITS 3.1.4 A01ITS 4.1 OPERATIONAL SAFETY REVIEW APPLICABILITY Applies to items directly related to safety limits and LIMITING CONDITIONS FOR OPERATION.
OBJECTIVE To assure that instrumentation shall be checked, tested, and calibrated, and that equipment and sampling tests shall be conducted at sufficiently frequent intervals to  
 
ensure safe operation.
SPECIFICATION
SPECIFICATION
: a. Calibration, testing, and checking of protective instrumentation channels and testing of logic channels shall be performed as specified in Table TS 4.1-1.  
: a. Calibration, testing, and checking of protective instrumentation channels and testing of   See other logic channels shall be performed as specified in Table TS 4.1-1.                             ITS SR 3.1.4.2,    b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and SR 3.1.4.3         TS 4.1-3.
: b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and TS 4.1-3.
: c. Deleted
See other ITS SR 3.1.4.2, SR 3.1.4.3
: d. Deleted
: c. Deleted  
: e. Deleted Amendment No. 119 TS 4.1-1                                       04/18/95 Page 3 of 5 Attachment 1, Volume 6, Rev. 0, Page 71 of 213
: d. Deleted  
: e. Deleted Amendment No. 119 TS 4.1-1 04/18/95 Page 3 of 5 TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Amendment No. 151 Page 2 of 7 02/12/2001 CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS 6. Pressurizer Water Level Each shift Each refueling cycle Monthly  7. Pressurizer Pressure Each shift Each refueling cycle  Monthly  8. a. 4-KV Voltage and Frequency Not applicable Each refueling cycle Monthly Reactor protection circuits only b. 4-KV Voltage  (Loss of Voltage)
Not applicable Each refueling cycle Monthly Safeguards buses only c. 4-KV Voltage (Degraded Grid)
Not applicable Each refueling cycle Monthly Safeguards buses only 9. Analog Rod Position Each shift(a,b)
Each refueling cycle Each refueling cycle (a) With step counters (b) Following rod motion in excess of 24 steps when computer is out of service 10. Rod Position Bank Counters Each shift(a,b)
Not applicable Each refueling cycle (a) With analog rod position (b) Following rod motion in excess of 24 steps when computer is out of service SR 3.1.4.1 L02 L02See ITS 3.3.5 See ITS 3.3.1 and 3.3.2 L02See ITS 3.1.7 ITS A01 ITS 3.1.4 Page 4 of 5 TABLE TS 4.1-3 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Amendment No. 125 Page 1 of 1 08/07/96 ITS 3.1.4ITS A01 Page 5 of 5 EQUIPMENT TESTS (1) TEST FREQUENCY 1. Control Rods Rod drop times of all full length rods Partial movement of all


rods not fully inserted in the
ITS                                                                                A01                                                                  ITS 3.1.4 TABLE TS 4.1-1                                                      See ITS 3.3.5 See ITS MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS                                                3.3.1 and 3.3.2 CHANNEL DESCRIPTION            CHECK                    CALIBRATE                  TEST                          REMARKS
: 6. Pressurizer Water Level    Each shift            Each refueling cycle        Monthly
: 7. Pressurizer Pressure      Each shift            Each refueling cycle        Monthly
: 8. a.      4-KV Voltage and  Not applicable        Each refueling cycle        Monthly        Reactor protection circuits only Frequency
: b. 4-KV Voltage          Not applicable        Each refueling cycle        Monthly        Safeguards buses only (Loss of Voltage)
: c. 4-KV Voltage          Not applicable        Each refueling cycle        Monthly        Safeguards buses only (Degraded Grid)
: 9. Analog Rod Position        Each shift(a,b)        Each refueling cycle        Each refueling (a) With step counters cycle          (b) Following rod motion in excess of 24 steps when computer is out of service
: 10. Rod Position Bank        Each shift(a,b)        Not applicable              Each refueling (a) With analog rod position Counters                                                                      cycle          (b) Following rod motion in excess of 24 steps when computer is out of service SR 3.1.4.1                                                                            See ITS                                              L02        L02 L02                                  3.1.7 , Volume 6, Rev. 0, Page 72 of 213                                                                                                                                                                  Attachment 1, Volume 6, Rev. 0, Page 72 of 213 Amendment No. 151 Page 2 of 7                                                          02/12/2001 Page 4 of 5


core Each REFUELING outage  
Attachment 1, Volume 6, Rev. 0, Page 73 of 213 ITS                                                    TABLE TS 4.1-3            A01 ITS 3.1.4 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS                                          L03 EQUIPMENT TESTS(1)                          TEST                          FREQUENCY
 
: 1. Control Rods                            Rod drop times of all full Each REFUELING outage SR 3.1.4.3                                            length rods                    Quarterly when at or above HOT Partial movement of all        STANDBY SR 3.1.4.2 rods not fully inserted in the core 1a. Reactor Trip Breakers                 Independent test(2) shunt Monthly and undervoltage trip attachments See ITS 1b. Reactor Coolant Pump Breakers-         OPERABILITY                    Each REFUELING outage                          3.3.1 Open-Reactor Trip 1c. Manual Reactor Trip                   Open trip reactor(3) trip and Each REFUELING outage bypass breaker
Quarterly when at or above HOT  
: 2. Deleted
 
: 3. Deleted See ITS
STANDBY 1a. Reactor Trip Breakers Independent test (2) shunt and undervoltage trip  
: 4. Containment Isolation Trip           OPERABILITY                   Each REFUELING outage                         3.6.3
 
: 5. Refueling System Interlocks           OPERABILITY                   Prior to fuel movement each See CTS REFUELING outage                             3.8.a.11
attachments Monthly 1b. Reactor Coolant Pump Breakers-Open-Reactor Trip OPERABILITY Each REFUELING outage 1c. Manual Reactor Trip Open trip reactor (3) trip and bypass breaker Each REFUELING outage  
: 6. Deleted                                                                                                             See ITS
: 2. Deleted  
: 7. Deleted                                                                                                               3.4.15 See ITS
: 3. Deleted  
: 8. RCS Leak Detection                     OPERABILITY                   Weekly(4)                                   3.8.1 and
: 4. Containment Isolation Trip OPERABILITY Each REFUELING outage 5. Refueling System Interlocks OPERABILITY Prior to fuel movement each REFUELING outage  
: 9. Diesel Fuel Supply                     Fuel Inventory(5)             Weekly                                         3.8.3
: 6. Deleted  
: 10. Deleted                                                                                                             See ITS 4.0
: 7. Deleted  
: 11. Fuel Assemblies                       Visual Inspection             Each REFUELING outage See ITS
: 8. RCS Leak Detection OPERABILITY Weekly (4) 9. Diesel Fuel Supply Fuel Inventory (5) Weekly 10. Deleted  
: 12. Guard Pipes                           Visual Inspection             Each REFUELING outage                         3.6.1
: 11. Fuel Assemblies Visual Inspection Each REFUELING outage  
: 13. Pressurizer PORVs                     OPERABILITY                   Each REFUELING cycle                         See ITS (6)                                3.4.11
: 12. Guard Pipes Visual Inspection Each REFUELING outage 13. Pressurizer PORVs OPERABILITY Each REFUELING cycle 14. Pressurizer PORV Block Valves OPERABILITY Quarterly (6) 15. Pressurizer Heaters OPERABILITY (7) Each REFUELING cycle 16. Containment Purge and Vent Isolation Valves OPERABILITY (8) Each REFUELING cycle  
: 14. Pressurizer PORV Block Valves         OPERABILITY                   Quarterly See ITS
 
: 15. Pressurizer Heaters                   OPERABILITY(7)                 Each REFUELING cycle                           3.4.9
(1) Following maintenance on equipment that could affect the operation of the equipment, tests should be performed to verify OPERABILITY. (2) Verify OPERABILITY of the bypass breaker undervoltage trip attachment prior to placing breaker into service. (3) Using the Control Room push-buttons, independently test the reactor trip breakers shunt trip and undervoltage trip attachments. The test shall also verify the undervoltage trip attachment  
: 16. Containment Purge and Vent             OPERABILITY(8)                 Each REFUELING cycle                         See ITS Isolation Valves                                                                                                    3.6.3 (1)
 
Following maintenance on equipment that could affect the operation of the equipment, tests                         L03 should be performed to verify OPERABILITY.
on the reactor trip bypass breakers.  
(2)
(4) When reactor is at power or in HOT SHUTDOWN condition.  
Verify OPERABILITY of the bypass breaker undervoltage trip attachment prior to placing breaker into service.
(5) Inventory of fuel required in all plant modes.  
(3)                                                                                                                 See ITS Using the Control Room push-buttons, independently test the reactor trip breakers shunt trip                     3.3.1 and undervoltage trip attachments. The test shall also verify the undervoltage trip attachment on the reactor trip bypass breakers.
(6) Not required when valve is administratively closed.  
(4)
(7) Test will verify OPERABILITY of heaters and availability of an emergency power supply.  
When reactor is at power or in HOT SHUTDOWN condition.                                                     See ITS 3.4.15 (5)                                                                                           See ITS 3.8.1 Inventory of fuel required in all plant modes.                                               and 3.8.3 (6)
(8) This test shall demonstrate that the valve(s) close in  5 seconds.
Not required when valve is administratively closed.                                                         See ITS 3.4.11 (7)
SR 3.1.4.2 See ITS 3.3.1 See ITS 3.4.15 See ITS 3.4.11 See ITS 3.4.9 See ITS 3.6.3 See ITS 3.6.1 See ITS 4.0 See ITS 3.8.1 and 3.8.3 See ITS 3.6.3 See CTS 3.8.a.11 L03 See ITS 3.3.1 See ITS 3.4.15 See ITS 3.4.9 See ITS 3.6.3 See ITS 3.4.11 See ITS 3.8.1 and 3.8.3 L03 SR 3.1.4.3
Test will verify OPERABILITY of heaters and availability of an emergency power supply.
 
(8)                                                                                                               See ITS This test shall demonstrate that the valve(s) close in  5 seconds.                                             3.4.9 See ITS         Amendment No. 125 Page 1 of 1                 3.6.3                   08/07/96 Page 5 of 5 Attachment 1, Volume 6, Rev. 0, Page 73 of 213
DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).


Attachment 1, Volume 6, Rev. 0, Page 74 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS ADMINISTRATIVE CHANGES A01  In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.
3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
MORE RESTRICTIVE CHANGES
MORE RESTRICTIVE CHANGES M01 CTS 3.10.e.2 states, in part, that when reactor power is < 85% but  50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. CTS 3.10.g.2 states that not more than one inoperable full length rod shall be allowed at any time. CTS 3.10.h states, in part, that at OPERATING temperature and full flow, the drop time of each full length rod cluster control shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to dashpot entry. ITS 3.1.4 requires all shutdown and control rods and the individual rod position indication to be OPERABLE in MODES 1 and 2 and requires the 24 step alignment limit to be met when < 85% RTP in MODES 1 and 2. This changes the CTS by requiring all shutdown and control rods and the individual rod position indication to be OPERABLE in MODES 1 and 2 and requires the 24 step alignment limit to be met when < 85% RTP in MODES 1 and 2, in lieu of when < 85% RTP and > 50% RTP.
 
The purpose of CTS 3.10.e.1 and CTS 3.10.e.2 is to maintain the rods in alignment with their respective banks to provide consistency with the assumption of the safety analyses, to maintain symmetric neutron flux and power distribution profiles, to provide assurance that peaking factors are within acceptable limits, and to assure adequate shutdown margin. The purpose of CTS 3.10.g and 3.10.h are to maintain the potential consequence of accidents consistent with the safety analyses. The proposed change is acceptable because it now states that all shutdown and control rods are required to be OPERABLE in MODES 1 and 2 and that alignment limits must be met in MODES 1 and 2. Additionally, it requires that individual rod position indication, dependent on percent of RTP, to be OPERABLE in MODES 1 and 2. This change is designated as more restrictive because specific LCO requirements and MODES of Applicability are specified.
M01 CTS 3.10.e.2 states, in part, that when reactor power is < 85% but  50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. CTS 3.10.g.2 states that not more than one inoperable full length rod shall be allowed at any time. CTS 3.10.h states, in part, that at OPERATING temperature and full flow, the drop time of each full length rod cluster control shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to dashpot entry. ITS 3.1.4 requires all shutdown and control rods and the individual rod position indication to be OPERABLE in MODES 1 and 2 and requires the 24 step alignment limit to be met when < 85% RTP in MODES 1 and 2. This changes the CTS by requiring all shutdown and control rods and the individual rod position indication to be OPERABLE in MODES 1 and 2 and requires the 24 step alignment limit to be met when < 85% RTP in MODES 1 and 2, in lieu of when < 85% RTP and  
M02  CTS 3.10.e.1 state, in part, that if a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is > 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating. ITS 3.1.4 ACTION B states that with one rod not within alignment limits when THERMAL POWER is > 85% RTP, to verify SHUTDOWN MARGIN (SDM) or initiate boration to restore SDM to within limits within 1 hour, and to reduce THERMAL POWER to < 85% RTP within 2 hours. CTS 3.10.e.2 states, in part, that if a rod Kewaunee Power Station                Page 1 of 6 Attachment 1, Volume 6, Rev. 0, Page 74 of 213
> 50% RTP.  
 
The purpose of CTS 3.10.e.1 and CTS 3.10.e.2 is to maintain the rods in alignment with their respective banks to provide consistency with the assumption of the safety analyses, to maintain symmetric neutron flux and power distribution profiles, to provide assurance that peaking factors are within acceptable limits, and to assure adequate shutdown margin. The purpose of CTS 3.10.g and 3.10.h are to maintain the potential consequence of accidents consistent with the safety analyses. The proposed change is acceptable because it now states that all shutdown and control rods are required to be OPERABLE in MODES 1 and 2 and that alignment limits must be met in MODES 1 and 2. Additionally, it requires that individual rod position indication, dependent on percent of RTP, to be OPERABLE in MODES 1 and 2. This change is designated as more restrictive because specific LCO requirements and MODES of Applicability are specified.  


M02 CTS 3.10.e.1 state, in part, that if a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is
Attachment 1, Volume 6, Rev. 0, Page 75 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is < 85 % but > 50%, the rod will be realigned or the core power peaking factors shall be determined within four hours and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating. ITS 3.1.4 ACTION C states that with one rod not within alignment limits when THERMAL POWER is < 85% RTP, to verify SHUTDOWN MARGIN (SDM) or initiate boration to restore SDM to within limit within 1 hour, to reduce THERMAL POWER to < 50% RTP within 2 hours, to verify SDM is within the limits specified in the COLR once per 12 hours, to perform SR 3.2.1.1 and SR 3.2.1.2 (verification that FQ(Z) is within the required limits) within 72 hours, to perform SR 3.2.2.1 (verification that F N H is within the required limits) within 72 hours and to re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions within 5 days. This changes the CTS by requiring the verification of SDM or initiation of boration to restore SDM to within limit within 1 hour, reduction of THERMAL POWER to < 85% RTP within 2 hours, verification of SDM is within the limits specified in the COLR once per 12 hours, the performance SR 3.2.1.1 and SR 3.2.1.2 within 72 hours, the performance of SR 3.2.2.1within 72 hours and the re-evaluation of the safety analyses and confirmation that the results remain valid for the duration of operation under these conditions within 5 days.
> 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating. ITS 3.1.4 ACTION B states that with one rod not within alignment limits when THERMAL POWER is > 85% RTP, to verify SHUTDOWN MARGIN (SDM) or initiate boration to restore SDM to within limits within 1 hour, and to reduce THERMAL POWER to < 85% RTP within 2 hours. CTS 3.10.e.2 states, in part, that if a rod Kewaunee Power Station Page 1 of 6 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is < 85 % but  
This change is acceptable because verification that the SDM is within its limits or the initiation of boration to restore the SDM to within limits is desirable. In contrast, it is not always desirable or possible to realign the misaligned control rod. Furthermore, if the misaligned rod continues to not meet the applicable alignment limits, in addition to the verification or restoration, the THERMAL POWER must be reduced (which in the case of the 12 step limit could result in meeting the alignment limits), periodic verification of the SDM must be accomplished along with the verification of FQ(Z) and F N H , and an evaluation must be performed to confirm that the accident analyses remain valid. All of these actions ensure that the core design criteria will not be exceeded.
> 50%, the rod will be realigned or the core power peaking factors shall be determined within four hours and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating. ITS 3.1.4 ACTION C states that with one rod not within alignment limits when THERMAL POWER is < 85% RTP, to verify SHUTDOWN MARGIN (SDM) or initiate boration to restore SDM to within limit within 1 hour, to reduce THERMAL POWER to < 50% RTP within 2 hours, to verify SDM is within the limits specified in the COLR once per 12 hours, to perform SR 3.2.1.1 and SR 3.2.1.2 (verification that F Q(Z) is within the required limits) within 72 hours, to perform SR 3.2.2.1 (verification that is within the required limits) within 72 hours and to re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions within 5 days. This changes the CTS by requiring the verification of SDM or initiation of boration to restore SDM to within limit within 1 hour, reduction of THERMAL POWER to < 85% RTP within 2 hours, verification of SDM is within the limits specified in the COLR once per 12 hours, the performance SR 3.2.1.1 and SR 3.2.1.2 within 72 hours, the performance of SR 3.2.2.1within 72 hours and the re-evaluation of the safety analyses and confirmation that the results remain valid for the duration of operation under these conditions within 5 days.
Additionally, the Completion Times are commensurate with safe operation of the plant and provides sufficient time to accomplish the Required Actions. This change is designated as more restrictive because more stringent Required Actions and shorter Completion Times have been added that were not in the CTS.
H N F  This change is acceptable because verification that the SDM is within its limits or the initiation of boration to restore the SDM to within limits is desirable. In contrast, it is not always desirable or possible to realign the misaligned control rod. Furthermore, if the misaligned rod continues to not meet the applicable alignment limits, in addition to the verification or restoration, the THERMAL POWER must be reduced (which in the case of the 12 step limit could result in meeting the alignment limits), periodic verification of the SDM must be accomplished along with the verification of F Q (Z) and , and an evaluation must be performed to confirm that the accident analyses remain valid. All of these actions ensure that the core design criteria will not be exceeded. Additionally, the Completion Times are commensurate with safe operation of the plant and provides sufficient time to accomplish the Required Actions. This change is designated as more restrictive because more stringent Required Actions and shorter Completion Times have been added that were not in the CTS. H N F M03 CTS 3.10.e.3 states that in addition to TS 3.10.e.1 (CTS 3.10.e.1) and TS 3.10.e.2 (CTS 3.10.e.2) if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable. CTS 3.10.e does not contain any requirements if 3.10.e.1 or 3.10.e.2 are not met. Therefore, CTS 3.0.c entry would be required. CTS 3.0.c requires action to be initiated within 1 hour and to be in HOT STANDBY (equivalent to ITS MODE 2) in the following 6 hours. ITS 3.1.4 ACTION D states, that if the Required Action and associated Completion Time of Condition B and C are not met, then be in MODE 3 within 6 hours. Additionally, CTS 3.10.e does not contain a requirement for when more than one rod is not within alignment limits. Therefore, CTS 3.0.c entry would also be required for this condition. ITS 3.1.4 ACTION E requires that if more than one rod is not within the alignment limit, then verify SDM is within the  
M03 CTS 3.10.e.3 states that in addition to TS 3.10.e.1 (CTS 3.10.e.1) and TS 3.10.e.2 (CTS 3.10.e.2) if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable. CTS 3.10.e does not contain any requirements if 3.10.e.1 or 3.10.e.2 are not met. Therefore, CTS 3.0.c entry would be required. CTS 3.0.c requires action to be initiated within 1 hour and to be in HOT STANDBY (equivalent to ITS MODE 2) in the following 6 hours. ITS 3.1.4 ACTION D states, that if the Required Action and associated Completion Time of Condition B and C are not met, then be in MODE 3 within 6 hours. Additionally, CTS 3.10.e does not contain a requirement for when more than one rod is not within alignment limits. Therefore, CTS 3.0.c entry would also be required for this condition. ITS 3.1.4 ACTION E requires that if more than one rod is not within the alignment limit, then verify SDM is within the limits specified in the COLR or initiate boration to restore the required SDM to Kewaunee Power Station                  Page 2 of 6 Attachment 1, Volume 6, Rev. 0, Page 75 of 213


limits specified in the COLR or initiate boration to restore the required SDM to Kewaunee Power Station Page 2 of 6 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS within limits within 1 hour and to be in MODE 3 within 6 hours. This changes the CTS by requiring the plant to be placed in MODE 3 if the Required Actions and associated Completion Time of B or C are not met (ITS 3.1.4 ACTION D) and adding a new ACTION (ITS 3.1.4 ACTION E).  
Attachment 1, Volume 6, Rev. 0, Page 76 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS within limits within 1 hour and to be in MODE 3 within 6 hours. This changes the CTS by requiring the plant to be placed in MODE 3 if the Required Actions and associated Completion Time of B or C are not met (ITS 3.1.4 ACTION D) and adding a new ACTION (ITS 3.1.4 ACTION E).
 
This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Times.
This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features, a reas onable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Times.
Allowing 1 hour to verify SDM is within the limits specified in the COLR or initiation of boration to restore the required SDM to within limits and 6 hours to be in MODE 3 ensures a unit shutdown is commenced and completed within a reasonable period of time upon failure to restore the SDM to within limits within the allowed Completion Times. This change is designated as more restrictive because the time to exit the Applicability of the LCO has been reduced.
Allowing 1 hour to verify SDM is within the limits specified in the COLR or initiation of boration to restore the required SDM to within limits and 6 hours to be in MODE 3 ensures a unit shutdown is commenced and completed within a reasonable period of time upon failure to restore the SDM to within limits within the allowed Completion Times. This change is designated as more restrictive because the time to exit the Applicability of the LCO has been reduced.
M04 CTS 3.10.g.2 states that not more than one inoperable full length rod shall be allowed at any time. As a result, if more than one full length rod is inoperable, then LCO 3.0.c would be entered which requires action to be initiated within 1 hour, to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours and to be in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours. CTS 3.10.g.3 states, in part, that with one inoperable full length rod the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days.
M04 CTS 3.10.g.2 states that not more than one inoperable full length rod shall be allowed at any time. As a result, if more than one full length rod is inoperable, then LCO 3.0.c would be entered which requires action to be initiated within 1 hour, to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours and to be in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours. CTS 3.10.g.3 states, in part, that with one inoperable full length rod the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days.
ITS 3.1.4 requires that with one or more rods inoperable to verify the SDM is within the limits specified in the COLR or to initiate boration to restore the SDM to within limits within 1 hour and to be in MODE 3 within 6 hours. This changes the CTS by requiring verification or restoration of the SDM and requiring the unit to be in MODE 3 within 6 hours.  
ITS 3.1.4 requires that with one or more rods inoperable to verify the SDM is within the limits specified in the COLR or to initiate boration to restore the SDM to within limits within 1 hour and to be in MODE 3 within 6 hours. This changes the CTS by requiring verification or restoration of the SDM and requiring the unit to be in MODE 3 within 6 hours.
 
The purpose of CTS 3.10.g.2 and 3.10.g.3 is to maintain the potential ejected rod worth and associated transient power distribution factor consistent with the safety analysis. This change is acceptable because it is not expected that more than one control rod will become misaligned from its group average position. When one or more control rods are misaligned, there is a potential for reduced SDM.
The purpose of CTS 3.10.g.2 and 3.10.g.3 is to maintain the potential ejected rod worth and associated transient power distribution factor consistent with the safety analysis. This change is acceptable because it is not expected that more than one control rod will become misaligned from its group average position. When one or more control rods are misaligned, there is a potential for reduced SDM.
Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity. The required Completion Time of 1 hour for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored. This change is acceptable because it is consistent with the requirements of the assumptions of the safety analyses to be within the SDM limit. Additionally, the requirement to be in MODE 2 is not required since the ITS requires the unit to be in MODE 3 in the same time (i.e., 6 hours). The change has been designated as more restrictive because the ITS requires verification or restoration of the SDM within 1 hour and the unit to be placed in MODE 3 within 6 hours in lieu of the CTS requirement to determine the potential ejected rod worth and associated transient Kewaunee Power Station Page 3 of 6 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS power distribution peaking factors within 30 days for one inoperable full length rod. 
Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity. The required Completion Time of 1 hour for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored. This change is acceptable because it is consistent with the requirements of the assumptions of the safety analyses to be within the SDM limit. Additionally, the requirement to be in MODE 2 is not required since the ITS requires the unit to be in MODE 3 in the same time (i.e., 6 hours). The change has been designated as more restrictive because the ITS requires verification or restoration of the SDM within 1 hour and the unit to be placed in MODE 3 within 6 hours in lieu of the CTS requirement to determine the potential ejected rod worth and associated transient Kewaunee Power Station                 Page 3 of 6 Attachment 1, Volume 6, Rev. 0, Page 76 of 213
 
RELOCATED SPECIFICATIONS None 
 
REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits)  CTS 3.10.e states, in part, that the magnitude of an indicated rod misalignment may be determined by comparison of the respective bank demand counter to the analog individual rod position indicator, the rod position on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage. ITS 3.1.4 does not contain this statement. This changes the CTS by moving the description of how to determine the magnitude of the rod misalignment to the Bases.
The removal of these details which relate to the system design from the Technical Specification is acceptable because this type of information is not necessary to be included to provide adequate protection of public health and safety. This change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specifications Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because the information relating to system design is being removed from the Technical Specifications.
 
LA02 (Type 1 - Removing Details of System Design and System Description, Including Design Limits)  CTS 3.10.g.1 states, in part, that an inoperable rod is a rod which does not trip or which is declared inoperable. ITS 3.1.4 does not contain this statement. This changes the CTS by moving the detail describing an inoperable rod to the Bases.
 
The removal of these details which relate to the system design from the Technical Specifications is acceptable because this type of information is not necessary to be included to provide adequate protection of public health and safety. This change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specifications Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because the information relating to system design is being removed from the Technical Specifications.
 
Kewaunee Power Station Page 4 of 6 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS LESS RESTRICTIVE CHANGES L01 (Category 7 - Relaxation of Surveillance Frequency)
CTS 3.10.i requires that if the rod position deviation monitor is inoperable, individual rod positions shall be logged at least once per eight hours and after a load change > 10% of rated power or after > 24 steps of control rod motion. ITS SR 3.1.4.1 requires verification that the individual rod positions are within the alignment limits every 12 hours. This changes the CTS by eliminating the requirement to log the individual rod positions once per 8 hours and after a load change > 10% of rated power or after > 24 steps of control rod motion when the Rod Position Deviation Monitor is inoperable.
 
The purpose of CTS 3.10.i is to periodically verify that the rods are within the alignment limits specified in the LCO. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Increasing the Frequency of rod position verification when the Rod Position Deviation Monitor is inoperable is unnecessary, since an inoperability of the alarm does not increase the probability that the rods are misaligned. The Rod Position Deviation Monitor alarm is for indication only. Its use is not credited in any safety analyses. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.
 
L02 (Category 7 - Relaxation of Surveillance Frequency)  Note (b) to CTS Table TS 4.1-1 Channel Description 9 requires an Analog Rod Position check following rod motion in excess of 24 steps when the computer is out of service. Note (b) to CTS TS Table 4.1-1 Channel Description 10 requires a Rod Position Bank Counters check following rod motion in excess of 24 steps when the computer is out of service. ITS SR 3.1.4.1 requires a verification that individual rod positions are within alignment every 12 hours. This changes the CTS by not requiring an Analog Rod Position check and a Rod Position Bank Counters check following rod motion in excess of 24 steps when the computer is out of service.
The purpose of CTS Table TS 4.1-1 Channel Descriptions 9 and 10 Channel Checks is to verify that the rods are within the alignment limits. This change is acceptable because the verification of the rod position following rod motion in excess of 24 steps when the computer is out of service is unnecessary. An inoperable computer does not increase the probability that the rods are misaligned. The routine 12 hour Frequency (ITS SR 3.1.4.1) continues to ensure that the control rods are aligned properly. Furthermore, ITS 3.1.7 Required Action A.1 requires the position of a rod with an inoperable position indicator to be verified once within 4 hours after moving the rod in excess of 24 steps. Thus, any response determined necessary by plant personnel due to an inoperable computer is more appropriately controlled by plant procedures, not Technical Specifications. This change is designated as less restrictive because Surveillances will be performed less frequently under ITS than under the CTS.
 
L03 (Category 5 - Deletion of Surveillance Requirement)
Note 1 to CTS Table TS 4.1-3 requires, in part, that the Control Rods be tested to verify OPERABILITY following maintenance on equipment that could affect the Kewaunee Power Station Page 5 of 6 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Kewaunee Power Station Page 6 of 6 operation. ITS 3.1.4 does not include this requirement. This changes the CTS by eliminating a post-maintenance Surveillance Requirement.
 
This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the equipment used to meet the LCO can perform its required functions. Thus, appropriate equipment continues to be tested in a manner and frequency necessary to give confidence that the equipment can perform its assumed safety function.
Whenever, the OPERABILITY of a system or component has been affected by repair, maintenance, modification, or replacement of a component, post maintenance testing is required to
 
demonstrate the OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under SR 3.0.1. In addition, the requirement of 10 CFR 50, Appendix B, Section XI (Test Control), provides adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria. Compliance with 10 CFR 50, Appendix B is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is designated as less restrictive because a Surveillance which is required in the CTS will not be performed in the ITS.
 
Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Rod Group Alignment Limits 3.1.4  WOG STS 3.1.4-1 Rev. 3.0, 03/31/04  CTS 3.1  REACTIVITY CONTROL SYSTEMS
 
3.1.4 Rod Group Alignment Limits
 
LCO  3.1.4  All shutdown and control rods shall be OPERABLE. 3.10.g, 3.10.h    AND Individual indicated rod positions shall be within 12 steps of their group step counter demand position.
 
INSERT 1 3.10.e.1, 3.10.e.2 as follows:
1 DOC M01 APPLICABILITY: MODES 1 and 2.
 
ACTIONS  CONDITION REQUIRED ACTION COMPLETION TIME A. One or more rod(s) inoperable.
 
A.1.1 Verify SDM to be within the limits specified in the COLR. OR A.1.2 Initiate boration to restore SDM to within limit.
 
AND  A.2 Be in MODE 3.
1 hour 
 
1 hour
 
6 hours  B. One rod not within alignment limits.
 
B.1 Restore rod to within alignment limits.
 
OR B.2.1.1 Verify SDM to be within the limits specified in the COLR. OR  1 hour
 
1 hour when THERMAL POWER is  85% RTP is 3 s 4 4 5 4 2 is3.10.g.3, DOC M04 3.10.e.1 3.1.4 CTS 1 INSERT 1    a. With THERMAL POWER  85% RTP, within 12 steps from their respective banks; and 3.10.e.1  3.10.e.2 b. With THERMAL POWER < 85% RTP, within 24 steps from their respective banks.
 
--------------------------------------------------NOTE------------------------------------------------- Individual rod position indicators may be outside of their limits for  1 hour following substantial rod movement.
3.10.e Note
------------------------------------------------------------------------------------------------------------
 
Insert Page 3.1.4-1 Rod Group Alignment Limits 3.1.4  WOG STS 3.1.4-2 Rev. 3.0, 03/31/04  CTS ACTIONS  (continued)
CONDITION REQUIRED ACTION COMPLETION TIME
 
B.2.1.2 Initiate boration to restore SDM to within limit.
AND  B.2.2 Reduce THERMAL POWER to  75% RTP.
 
AND B.2.3 Verify SDM is within the limits specified in the COLR. AND  B.2.4 Perform SR 3.2.1.1 and SR 3.2.1.2.
AND  B.2.5 Perform SR 3.2.2.1.
 
AND  B.2.6 Re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions.
 
1 hour
 
2 hours
 
Once per 12 hours
 
72 hours
 
72 hours
 
5 days C. Required Action and associated Completion Time of Condition B not met. C.1 Be in MODE 3.
 
6 hours  D. More than one rod not within alignment limit.
 
D.1.1 Verify SDM is within the limits specified in the COLR.
OR 1 hour 3.10.e.1 3 4 s 4 5 4< 85INSERT 2 5 5 3.10.e.3 D D 5or C E E 5DOC M03 3.1.4 CTS 5 INSERT 2 C. One rod not within alignment limits when THERMAL POWER is
 
< 85% RTP.
 
C.1.1 Verify SDM is within limits specified in the COLR.
 
OR  C.1.2 Initiate boration to restore SDM to within limits.
AND  C.2 Reduce THERMAL POWER to < 50% RTP.
AND  C.3 Verify SDM is within limits specified in the COLR.
AND C.4 Perform SR 3.2.1.1 and SR 3.2.1.2.
 
AND C.5 Perform SR 3.2.2.1.
AND C.6 Re-evaluate safety analyses and confirm
 
results remain valid for duration of operation under these conditions.


1 hour
Attachment 1, Volume 6, Rev. 0, Page 77 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS power distribution peaking factors within 30 days for one inoperable full length rod.
RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.10.e states, in part, that the magnitude of an indicated rod misalignment may be determined by comparison of the respective bank demand counter to the analog individual rod position indicator, the rod position on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage. ITS 3.1.4 does not contain this statement.
This changes the CTS by moving the description of how to determine the magnitude of the rod misalignment to the Bases.
The removal of these details which relate to the system design from the Technical Specification is acceptable because this type of information is not necessary to be included to provide adequate protection of public health and safety. This change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specifications Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because the information relating to system design is being removed from the Technical Specifications.
LA02 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.10.g.1 states, in part, that an inoperable rod is a rod which does not trip or which is declared inoperable. ITS 3.1.4 does not contain this statement. This changes the CTS by moving the detail describing an inoperable rod to the Bases.
The removal of these details which relate to the system design from the Technical Specifications is acceptable because this type of information is not necessary to be included to provide adequate protection of public health and safety. This change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specifications Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because the information relating to system design is being removed from the Technical Specifications.
Kewaunee Power Station                Page 4 of 6 Attachment 1, Volume 6, Rev. 0, Page 77 of 213


1 hour
Attachment 1, Volume 6, Rev. 0, Page 78 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS LESS RESTRICTIVE CHANGES L01  (Category 7 - Relaxation of Surveillance Frequency) CTS 3.10.i requires that if the rod position deviation monitor is inoperable, individual rod positions shall be logged at least once per eight hours and after a load change > 10% of rated power or after > 24 steps of control rod motion. ITS SR 3.1.4.1 requires verification that the individual rod positions are within the alignment limits every 12 hours. This changes the CTS by eliminating the requirement to log the individual rod positions once per 8 hours and after a load change > 10% of rated power or after > 24 steps of control rod motion when the Rod Position Deviation Monitor is inoperable.
The purpose of CTS 3.10.i is to periodically verify that the rods are within the alignment limits specified in the LCO. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Increasing the Frequency of rod position verification when the Rod Position Deviation Monitor is inoperable is unnecessary, since an inoperability of the alarm does not increase the probability that the rods are misaligned. The Rod Position Deviation Monitor alarm is for indication only. Its use is not credited in any safety analyses. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.
L02  (Category 7 - Relaxation of Surveillance Frequency) Note (b) to CTS Table TS 4.1-1 Channel Description 9 requires an Analog Rod Position check following rod motion in excess of 24 steps when the computer is out of service. Note (b) to CTS TS Table 4.1-1 Channel Description 10 requires a Rod Position Bank Counters check following rod motion in excess of 24 steps when the computer is out of service. ITS SR 3.1.4.1 requires a verification that individual rod positions are within alignment every 12 hours. This changes the CTS by not requiring an Analog Rod Position check and a Rod Position Bank Counters check following rod motion in excess of 24 steps when the computer is out of service.
The purpose of CTS Table TS 4.1-1 Channel Descriptions 9 and 10 Channel Checks is to verify that the rods are within the alignment limits. This change is acceptable because the verification of the rod position following rod motion in excess of 24 steps when the computer is out of service is unnecessary. An inoperable computer does not increase the probability that the rods are misaligned. The routine 12 hour Frequency (ITS SR 3.1.4.1) continues to ensure that the control rods are aligned properly. Furthermore, ITS 3.1.7 Required Action A.1 requires the position of a rod with an inoperable position indicator to be verified once within 4 hours after moving the rod in excess of 24 steps. Thus, any response determined necessary by plant personnel due to an inoperable computer is more appropriately controlled by plant procedures, not Technical Specifications. This change is designated as less restrictive because Surveillances will be performed less frequently under ITS than under the CTS.
L03  (Category 5 - Deletion of Surveillance Requirement) Note 1 to CTS Table TS 4.1-3 requires, in part, that the Control Rods be tested to verify OPERABILITY following maintenance on equipment that could affect the Kewaunee Power Station                  Page 5 of 6 Attachment 1, Volume 6, Rev. 0, Page 78 of 213


2 hours 
Attachment 1, Volume 6, Rev. 0, Page 79 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS operation. ITS 3.1.4 does not include this requirement. This changes the CTS by eliminating a post-maintenance Surveillance Requirement.
This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the equipment used to meet the LCO can perform its required functions. Thus, appropriate equipment continues to be tested in a manner and frequency necessary to give confidence that the equipment can perform its assumed safety function. Whenever, the OPERABILITY of a system or component has been affected by repair, maintenance, modification, or replacement of a component, post maintenance testing is required to demonstrate the OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under SR 3.0.1. In addition, the requirement of 10 CFR 50, Appendix B, Section XI (Test Control), provides adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria. Compliance with 10 CFR 50, Appendix B is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is designated as less restrictive because a Surveillance which is required in the CTS will not be performed in the ITS.
Kewaunee Power Station              Page 6 of 6 Attachment 1, Volume 6, Rev. 0, Page 79 of 213


Once per 12 hours
Attachment 1, Volume 6, Rev. 0, Page 80 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 80 of 213


72 hours   
Attachment 1, Volume 6, Rev. 0, Page 81 of 213 CTS                                                                                  Rod Group Alignment Limits 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4          Rod Group Alignment Limits 3.10.g,  LCO 3.1.4                All shutdown and control rods shall be OPERABLE.
3.10.h AND as follows:
3.10.e.1,                          Individual indicated rod positions shall be within 12 steps of their group 3.10.e.2 step counter demand position.                                                      1 INSERT 1 DOC M01  APPLICABILITY:          MODES 1 and 2.
ACTIONS CONDITION                      REQUIRED ACTION                  COMPLETION TIME is 3.10.g.3,  A. One or more rod(s)                A.1.1  Verify SDM to be within the      1 hour                      2 DOC M04 inoperable.                            limits specified in the COLR.
OR A.1.2  Initiate boration to restore    1 hour SDM to within limit.                                        3 s
AND A.2    Be in MODE 3.                    6 hours 4
3.10.e.1   B. One rod not within                B.1    Restore rod to within            1 hour alignment limits.                      alignment limits.                                              5 when THERMAL POWER is  85% RTP OR is B.2.1.1 Verify SDM to be within the      1 hour                      4 limits specified in the COLR.
OR                                                              4 WOG STS                                          3.1.4-1                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 81 of 213


72 hours 
Attachment 1, Volume 6, Rev. 0, Page 82 of 213 CTS                                                                                                              3.1.4 1
INSERT 1 3.10.e.1    a.        With THERMAL POWER  85% RTP, within 12 steps from their respective banks; and 3.10.e.2    b.        With THERMAL POWER < 85% RTP, within 24 steps from their respective banks.
            --------------------------------------------------NOTE-------------------------------------------------
3.10.e Note Individual rod position indicators may be outside of their limits for  1 hour following substantial rod movement.
Insert Page 3.1.4-1 Attachment 1, Volume 6, Rev. 0, Page 82 of 213


5 days 3.10.e.2
Attachment 1, Volume 6, Rev. 0, Page 83 of 213 CTS                                                                      Rod Group Alignment Limits 3.1.4 ACTIONS (continued)
CONDITION                  REQUIRED ACTION                  COMPLETION TIME 3.10.e.1 B.2.1.2 Initiate boration to restore    1 hour                  4 SDM to within limit.                                        3 s
AND                                                          4 4
B.2.2  Reduce THERMAL                  2 hours POWER to  75% RTP.                                        5
                                                                  < 85 AND B.2.3  Verify SDM is within the        Once per limits specified in the          12 hours COLR.
AND B.2.4  Perform SR 3.2.1.1 and          72 hours                5 SR 3.2.1.2.
AND B.2.5  Perform SR 3.2.2.1.              72 hours AND B.2.6  Re-evaluate safety              5 days analyses and confirm results remain valid for duration of operation under these conditions.
INSERT 2  5 3.10.e.3 C. Required Action and    C.1    Be in MODE 3.                    6 hours associated Completion        D                                                          5 D
Time of Condition B not met.          or C DOC M03  D. More than one rod not  D.1.1  Verify SDM is within the        1 hour                  5 E
within alignment limit. E  limits specified in the COLR.
OR WOG STS                                3.1.4-2                           Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 83 of 213


Insert Page 3.1.4-2 Rod Group Alignment Limits 3.1.4   WOG STS 3.1.4-3 Rev. 3.0, 03/31/04  CTS ACTIONS  (continued)
Attachment 1, Volume 6, Rev. 0, Page 84 of 213 CTS                                                                              3.1.4 5
CONDITION REQUIRED ACTION COMPLETION TIME
INSERT 2 3.10.e.2 C. One rod not within    C.1.1  Verify SDM is within limits  1 hour alignment limits when          specified in the COLR.
THERMAL POWER is
            < 85% RTP.                OR C.1.2  Initiate boration to restore 1 hour SDM to within limits.
AND C.2    Reduce THERMAL              2 hours POWER to < 50% RTP.
AND C.3    Verify SDM is within limits  Once per specified in the COLR.      12 hours AND C.4     Perform SR 3.2.1.1 and      72 hours SR 3.2.1.2.
AND C.5    Perform SR 3.2.2.1.          72 hours AND C.6    Re-evaluate safety          5 days analyses and confirm results remain valid for duration of operation under these conditions.
Insert Page 3.1.4-2 Attachment 1, Volume 6, Rev. 0, Page 84 of 213


D.1.2 Initiate boration to restore required SDM to within  
Attachment 1, Volume 6, Rev. 0, Page 85 of 213 CTS                                                                              Rod Group Alignment Limits 3.1.4 ACTIONS (continued)
CONDITION                      REQUIRED ACTION                    COMPLETION TIME DOC M03                                    D.1.2   Initiate boration to restore     1 hour                    5 E    required SDM to within limit.                                                      3 s
AND E
D.2      Be in MODE 3.                    6 hours                    5 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY Table 4.1-1 Channel      SR 3.1.4.1      Verify individual rod positions within alignment limit. 12 hours Description 9 and 10 Table 4.1-3 Equipment SR 3.1.4.2      Verify rod freedom of movement (trippability) by            92 days Test 1                      moving each rod not fully inserted in the core 10 steps in either direction.
3.10.h.1,    SR 3.1.4.3      Verify rod drop time of each rod, from the fully            Prior to criticality Table 4.1-3                                                                      1.8 Equipment                    withdrawn position, is  [2.2] seconds from the            after each              6 Test 1                      beginning of decay of stationary gripper coil voltage      removal of the to dashpot entry, with:                                    reactor head
: a. Tavg  500&deg;F and
: b. All reactor coolant pumps operating.
WOG STS                                    3.1.4-3                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 85 of 213


limit.
Attachment 1, Volume 6, Rev. 0, Page 86 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS
AND D.2 Be in MODE 3.  
: 1. LCO 3.1.4 (2nd part) has been modified to incorporate a Kewaunee Power Station (KPS) specific allowance. This change to the LCO has been made consistent with the allowance in License Amendment 51, dated December 28, 1983 (ADAMS Accession Number ML020740166), with one exception. The CTS does not require alignment limits to be met when < 50% RTP, as shown in CTS 3.10.e.2. However, as part of converting to the ITS, KPS will require alignment limits to be met at all times when in MODES 1 and 2. That is, the 24 step limit that is currently required when < 85% and > 50% RTP will be applicable in MODE 1 when < 85% RTP and in MODE 2.
: 2. Changes are made to be consistent with the format of the ITS. The Required Action normally states that the parameter shall be "within limits."
: 3. Typographical/grammatical error corrected.
: 4. ISTS 3.1.4 Required Action B.1 requires restoration of a rod not within alignment limits within 1 hour or performance of a number of other actions, such as verification of SHUTDOWN MARGIN, reduction in reactor power, measurement of hot channel factors, and re-evaluation of the safety analyses. The Writer's Guide for Improved Standard Technical Specifications, TSTF-GG-05-01, Section 4.1.6.g states, "A Required Action which requires restoration, such that the Condition is no longer met, is considered superfluous. It is only included if it would be the only Required Action for the Condition or it is needed for presentation clarity." Neither exception applies in this case. In fact, the inclusion of Required Action B.1 requires an additional level of indenting and numbering for the remaining Required Actions in Condition B, which reduces its clarity.
Therefore, Required Action B.1 is deleted and the subsequent Required Actions renumbered.
: 5. ISTS 3.1.4 ACTION B has been split into two ACTIONS, based on the KPS specific rod group alignment limits. As discussed in JFD 1, LCO 3.1.4 (2nd part) has been modified to be consistent with KPS License Amendment 51. The KPS CTS requires the rod alignment to be within 12 steps when > 85% RTP, but only 24 steps when < 85% RTP. Therefore, the following changes have been made due to this difference:
: a. ISTS 3.1.4 Condition B has been modified to only be applicable when THERMAL POWER is > 85% RTP. ITS 3.1.4 ACTION B will now provide the actions when the > 85% RTP alignment limit (12 steps) is not met. For ITS ACTION B, only ISTS 3.1.4 Required Actions B.2.1.1, B.2.1.2, and B.2.2 are being maintained (note that ISTS 3.1.4 Required Action B.1 has been deleted as discussed in JFD 4). ISTS 3.1.4 Required Action B.2.2 (ITS 3.1.4 Required Action B.2) has been modified to require reduction in THERMAL POWER to < 85% RTP. Thus, once power is reduced to < 85% RTP, the Condition B entry requirements are not met. Furthermore, once power is reduced, a new alignment limit, 24 steps, is applicable. Therefore, the follow-on Required Actions in the ISTS (ISTS 3.1.4 Required Actions B.2.3, B.2.4, B.2.5, and B.2.6) are not applicable and are not included in ITS 3.1.4 ACTION B.
Kewaunee Power Station                Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 86 of 213


1 hour
Attachment 1, Volume 6, Rev. 0, Page 87 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS
 
: b. A new ACTION has been added (ITS 3.1.4 ACTION C) to cover not meeting the < 85% RTP alignment limit. ITS 3.1.4 Condition C covers the condition of one rod not within alignment limits when THERMAL POWER is < 85% RTP.
6 hours 5DOC M03 E 3 s E 5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR  3.1.4.1 Verify individual rod positions within alignment limit.
The proposed Required Actions for this Condition are consistent with the Required Actions of ISTS 3.1.4 ACTION B, except the restore Required Action is not included (as described in JFD 4) and the power reduction value in ITS 3.1.4 Required Action C.2 is < 50% RTP (in lieu of the ISTS value of
12 hours  SR  3.1.4.2 Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core 10 steps in either direction.
        < 75% RTP), consistent with the current KPS CTS.
 
92 days SR  3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn position, is  [2.2] seconds from the beginning of decay of stationary gripper coil voltage to dashpot entry, with:
: a. Tavg  500&deg;F and
: b. All reactor coolant pumps operating.
 
Prior to criticality after each removal of the reactor head Table 4.1-1 Channel Description 9 and 10 Table 4.1-3
 
Equipment Test 1 3.10.h.1, Table 4.1-3
 
Equipment Test 1 1.8 6 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Kewaunee Power Station Page 1 of 2 1. LCO 3.1.4 (2nd part) has been modified to incorporate a Kewaunee Power Station (KPS) specific allowance. This change to the LCO has been made consistent with the allowance in License Amendment 51, dated December 28, 1983 (ADAMS Accession Number ML020740166), with one exception. The CTS does not require alignment limits to be met when < 50% RTP, as shown in CTS 3.10.e.2. However, as part of converting to the ITS, KPS will require alignment limits to be met at all times when in MODES 1 and 2. That is, the
 
24 step limit that is currently required when <
85% and > 50% RTP will be applicable in MODE 1 when < 85% RTP and in MODE
: 2. 
: 2. Changes are made to be consistent with the format of the ITS. The Required Action normally states that the parameter shall be "within limits."  3. Typographical/grammatical error corrected.
: 4. ISTS 3.1.4 Required Action B.1 requires restoration of a rod not within alignment limits within 1 hour or performance of a number of other actions, such as verification of SHUTDOWN MARGIN, reduction in reactor power, measurement of hot channel factors, and re
-evaluation of the safety analyses. The Writer's Guide for Improved Standard Technical Specifications, TSTF
-GG-05-01, Section 4.1.6.g states, "A Required Action which requires restoration, such that the Condition is no longer met, is considered superfluous. It is only included if it would be the only Required Action for the Condition or it is needed for presentation clarity."  Neither exception applies in this case. In fact, the inclusion of Required Action B.1 requires an additional level of indenting and numbering for the remaining Required Actions in Condition B, which reduces its clarity. Therefore, Required Action B.1 is deleted and the subsequent Required Actions renumbered.
: 5. ISTS 3.1.4 ACTION B has been split into two ACTIONS, based on the KPS specific rod group alignment limits. As discussed in JFD 1, LCO 3.1.4 (2nd part) has been modified to be consistent with KPS License Amendment 51. The KPS CTS requires the rod alignment to be within 12 steps when
> 85% RTP, but only 24 steps when < 85% RTP. Therefore, the following changes have been made due to this difference:
: a. ISTS 3.1.4 Condition B has been modified to only be applicable when THERMAL POWER is
> 85% RTP. ITS 3.1.4 ACTION B will now provide the actions when the
> 85% RTP alignment limit (12 steps) is not met. For ITS ACTION B, only ISTS 3.1.4 Required Actions B.2.1.1, B.2.1.2, and B.2.2 are being maintained (note that ISTS 3.1.4 Required Action B.1 has been deleted as discussed in JFD 4). ISTS 3.1.4 Required Action B.2.2 (ITS 3.1.4 Required Action B.2) has been modified to require reduction in THERMAL POWER to < 85% RTP. Thus, once power is reduced to < 85% RTP, the Condition B entry requirements are not met. Furthermore, once power is reduced, a new alignment limit, 24 steps, is applicable. Therefore, the follow
-
on Required Actions in the ISTS (ISTS 3.1.4 Required Actions B.2.3, B.2.4, B.2.5, and B.2.6) are not applicable and are not included in ITS 3.1.4 ACTION B.
JUSTIFICATION FOR DEVIATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS Kewaunee Power Station Page 2 of 2 b. A new ACTION has been added (ITS 3.1.4 ACTION C) to cover not meeting the < 85% RTP alignment limit. ITS 3.1.4 Condition C covers the condition of one rod not within alignment limits when THERMAL POWER is < 85% RTP.
The proposed Required Actions for this Condition are consistent with the Required Actions of ISTS 3.1.4 ACTION B, except the restore Required Action is not included (as described in JFD 4) and the power reduction value in ITS 3.1.4 Required Action C.2 is < 50% RTP (in lieu of the ISTS value of < 75% RTP), consistent with the current KPS CTS.
In addition, due to the above changes (the addition of new ACTION C),
In addition, due to the above changes (the addition of new ACTION C),
subsequent Conditions and Required Actions have been renumbered.
subsequent Conditions and Required Actions have been renumbered.
: 6. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 6. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
Kewaunee Power Station                Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 87 of 213


Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 88 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Rod Group Alignment Limits B 3.1.4    WOG STS B 3.1.4-1 Rev. 3.0, 03/31/04  B 3.3  INSTRUMENTATION
Attachment 1, Volume 6, Rev. 0, Page 88 of 213


B 3.1.4 Rod Group Alignment Limits  
Attachment 1, Volume 6, Rev. 0, Page 89 of 213 Rod Group Alignment Limits B 3.1.4 B 3.3 INSTRUMENTATION B 3.1.4 Rod Group Alignment Limits BASES BACKGROUND          The OPERABILITY (i.e., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.
The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design,"                  1 INSERT 1  GDC 26, "Reactivity Control System Redundancy and Capability" (Ref. 1),
and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2).
Mechanical or electrical failures may cause a control or shutdown rod to become inoperable or to become misaligned from its group. Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.
Limits on rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.
Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMs). Each CRDM moves its RCCA one step (approximately e inch) at a time, but at varying rates (steps per        2 minute) depending on the signal output from the Rod Control System.
The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. If a bank of RCCAs consists of two groups, the groups are moved in a staggered fashion, but always within one step of each other. All units have four control banks            2 and at least two shutdown banks.                        Kewaunee Power Station (KPS) has WOG STS                                      B 3.1.4-1                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 89 of 213


BASES  BACKGROUND The OPERABILITY (i.e., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.
Attachment 1, Volume 6, Rev. 0, Page 90 of 213 B 3.1.4 1
INSERT 1 USAR GDC 6, "Reactor Core Design," GDC 7, "Suppression of Power Oscillations," GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot Shutdown Capability," GDC 29, "Reactivity Shutdown Capability," and GDC 30, "Reactivity Holddown Capability" Insert Page B 3.1.4-1 Attachment 1, Volume 6, Rev. 0, Page 90 of 213


The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," GDC 26, "Reactivity Control System Redundancy and Capability" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2). INSERT 1 1 Mechanical or electrical failures may cause a control or shutdown rod to become inoperable or to become misaligned from its group. Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.
Attachment 1, Volume 6, Rev. 0, Page 91 of 213 Rod Group Alignment Limits B 3.1.4 BASES BACKGROUND (continued)
Limits on rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.  
The shutdown banks are maintained either in the fully inserted or fully withdrawn position. The control banks are moved in an overlap pattern, using the following withdrawal sequence: When control bank A reaches a predetermined height in the core, control bank B begins to move out with control bank A. Control bank A stops at the position of maximum withdrawal, and control bank B continues to move out. When control bank B reaches a predetermined height, control bank C begins to move out with control bank B. This sequence continues until control banks A, B, and C are at the fully withdrawn position, and control bank D is approximately halfway withdrawn. The insertion sequence is the opposite of the withdrawal sequence. The control rods are arranged in a radially symmetric pattern, so that control bank motion does not introduce radial asymmetries in the core power distributions.
The axial position of shutdown rods and control rods is indicated by two separate and independent systems, which are the Bank Demand Position Indication System (commonly called group step counters) and the Digital              2 Rod Position Indication (DRPI) System.                        Individual I
The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal INSERT 2 to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position 5/8                  2 Indication System is considered highly precise (+/- 1 step or +/- e inch). If a         7 rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.
I control The DRPI System provides a highly accurate indication of actual rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube. To increase the reliability of the system, the inductive coils            2 are connected alternately to data system A or B. Thus, if one data INSERT 3    system fails, the DRPI will go on half accuracy. The DRPI System is capable of monitoring rod position within at least +/- 12 steps with either full accuracy or half accuracy.
WOG STS                                    B 3.1.4-2                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 91 of 213


Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMs). Each CRDM moves its RCCA one step (approximately e inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System. 2 The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. If a bank of RCCAs consists of two groups, the groups are moved in a staggered fashion, but always within one step of each other. All units have four control banks and at least two shutdown banks.
Attachment 1, Volume 6, Rev. 0, Page 92 of 213 B 3.1.4 2
2Kewaunee Power Station (KPS) has B 3.1.4 1 INSERT 1    USAR GDC 6, "Reactor Core Design," GDC 7, "Suppression of Power Oscillations," GDC 27, "Redundancy of Reacti vity Control," GDC 28, "Reactivity Hot Shutdown Capability," GDC 29, "Reactivity Shutdown Capability," and
INSERT 2 The readout of the Bank Demand Position Indication System is in the form of an add and subtract counter reading. There is one display for each rod group.
2 INSERT 3 an analog signal that is proportional to the actual control rod position. This is accomplished by using an electrical coil stack linear variable differential transformer placed above the stepping mechanism of the control rod magnetic jacks which are external to the pressure housing. When the control rod is at the bottom of the core, the magnetic coupling between the primary and secondary windings is small and induces a small voltage in the secondary winding. As the control rod raises, the lift rod causes an increase in magnetic coupling which in turn provides an analog signal proportional to the control rod position. The individual rod position indicator channels are sufficiently accurate to detect a rod
+ 12 steps away from its demand position.
Insert Page B 3.1.4-2 Attachment 1, Volume 6, Rev. 0, Page 92 of 213


GDC 30, "Reactivity Holddown Capability"
Attachment 1, Volume 6, Rev. 0, Page 93 of 213 Rod Group Alignment Limits B 3.1.4 BASES APPLICABLE                   Control rod misalignment accidents are analyzed in the safety analysis SAFETY                       (Ref. 3). The acceptance criteria for addressing control rod inoperability ANALYSES                     or misalignment are that:
 
: a. There be no violations of:
Insert Page B 3.1.4-1 Rod Group Alignment Limits B 3.1.4    WOG STS B 3.1.4-2 Rev. 3.0, 03/31/04  BASES BACKGROUND  (continued)
: 1. Specified acceptable fuel design limits or                                     4
 
: 2. Reactor Coolant System (RCS) pressure boundary integrity and                   4
The shutdown banks are maintained either in the fully inserted or fully withdrawn position. The control banks are moved in an overlap pattern, using the following withdrawal sequence:  When control bank A reaches a predetermined height in the core, control bank B begins to move out with control bank A. Control bank A stops at the position of maximum withdrawal, and control bank B continues to move out. When control bank B reaches a predetermined height, control bank C begins to move out with control bank B. This sequence continues until control banks A, B, and C are at the fully withdrawn position, and control bank D is approximately halfway withdrawn. The insertion sequence is the opposite of the withdrawal sequence. The control rods are arranged in a radially symmetric pattern, so that control bank motion does not introduce radial asymmetries in the core power distributions.
 
The axial position of shutdown rods and control rods is indicated by two separate and independent systems, which are the Bank Demand Position Indication System (commonly called group step counters) and the Digital Rod Position Indication (DRPI) System.
 
The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or +/-
e inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.
The DRPI System provides a highly accurate indication of actual rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one data system fails, the DRPI will go on half accuracy. The DRPI System is capable of monitoring rod position within at least +/- 12 steps with either full accuracy or half accuracy.
I I Individual 2 control 7 5/8 2 2INSERT 3 INSERT 2 B 3.1.4 2 INSERT 2  The readout of the Bank Demand Position Indication System is in the form of an add and subtract counter reading. There is one display for each rod group.
 
2 INSERT 3 
 
an analog signal that is proportional to the actual control rod position. This is accomplished by using an electrical coil stack linear variable differential
 
transformer placed above the stepping mec hanism of the control rod magnetic jacks which are external to the pressure housing. When the control rod is at the bottom of the core, the magnetic coupling between the primary and secondary windings is small and induces a small voltage in the secondary winding. As the control rod raises, the lift rod causes an increase in magnetic coupling which in turn provides an analog signal proportional to the control rod position. The individual rod position indicator channels are sufficiently accurate to detect a rod
+ 12 steps away from its demand position.
Insert Page B 3.1.4-2 Rod Group Alignment Limits B 3.1.4   WOG STS B 3.1.4-3 Rev. 3.0, 03/31/04  BASES APPLICABLE Control rod misalignment accidents are analyzed in the safety analysis SAFETY (Ref. 3). The acceptance criteria for addressing control rod inoperability ANALYSES or misalignment are that:  
: a. There be no violations of:  
: 41. Specified acceptable fuel design limits or  
;  42. Reactor Coolant System (RCS) pressure boundary integrity and  
;
: b. The core remains subcritical after accident transients.
: b. The core remains subcritical after accident transients.
Two types of misalignment are distinguished. During movement of a control rod group, one rod may stop moving, while the other rods in the group continue. This condition may cause excessive power peaking.
INSERT 4 Two types of misalignment are distinguished. During movement of a control rod group, one rod may stop moving, while the other rods in the                   2 group continue. This condition may cause excessive power peaking.
The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.  
A statically misaligned  The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.
Two types of analysis are performed in regard to static rod misalignment (Ref. 4). With control banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core.
The second type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying 2
limits on departure from nucleate boiling ratio in both of these cases          INSERT 5 bounds the situation when a rod is misaligned from its group by 12 steps.
INSERT 6      2 Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed                5 in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5).
The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved.
WOG STS                                            B 3.1.4-3                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 93 of 213


Two types of analysis are performed in regard to static rod misalignment (Ref. 4). With control banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core. The second type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps.  
Attachment 1, Volume 6, Rev. 0, Page 94 of 213 B 3.1.4 2
INSERT 4 There are three RCCA misalignment accidents which are analyzed including one or more dropped RCCAs, a dropped RCCA bank, and a statically misaligned RCCA (Ref. 4).
2 INSERT 5 when THERMAL POWER is  85% RTP or by 24 steps when THERMAL POWER is < 85% RTP 2
INSERT 6 For the dropped RCCA(s) or dropped RCCA bank misalignment accidents, a negative reactivity insertion will result. Power will be reestablished either by reactivity feedback or control bank withdrawal. Following plant stabilization, normal rod retrieval or shutdown procedures are followed. For dropped RCCA events in the automatic rod control mode, the Rod Control System detects the drop in power and initiates control bank withdrawal. In all cases, the minimum departure from nucleate boiling ratio (DNBR) remains above the limit value.
Insert Page B 3.1.4-3 Attachment 1, Volume 6, Rev. 0, Page 94 of 213


Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully wi thdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5).  
Attachment 1, Volume 6, Rev. 0, Page 95 of 213 Rod Group Alignment Limits B 3.1.4 BASES APPLICABLE SAFETY ANALYSES (continued)
Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor ( FQ(Z)) and the nuclear enthalpy hot channel factor ( F N H ) are verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the              5 rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and FQ(Z) and F N H must be verified directly by incore mapping.
Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of FQ(Z) and F N H to the operating limits.
Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO              The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on control rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The control rod OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements, which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment. The rod OPERABILITY requirement is satisfied provided the rod will fully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability.
INSERT 7 The requirement to maintain the rod alignment to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment              2 from fully withdrawn to fully inserted is assumed.
INSERT 8 Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.
INSERT 8A    6 WOG STS                                  B 3.1.4-4                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 95 of 213


INSERT 6 2 2 5INSERT 5 2INSERT 4 A statically misaligned The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved.
Attachment 1, Volume 6, Rev. 0, Page 96 of 213 B 3.1.4 2
B 3.1.4 2 INSERT 4   There are three RCCA misalignment acci dents which are analyzed including one or more dropped RCCAs, a dropped RCCA bank, and a statically misaligned RCCA (Ref. 4).  
INSERT 7 with THERMAL POWER  85% RTP or within 24 steps of their group step counter demand position when THERMAL POWER is < 85% RTP 2
INSERT 8 The safety analysis assumes a misalignment of one or more RCCA(s) or, in some cases, an entire RCCA bank.
6 INSERT 8A The LCO is modified by a Note stating that individual control rod position indications may be outside the required limits for < 1 hour following subsequent control rod movement. This allows up to 1 hour of thermal soak time to allow the control rod drive shaft to reach thermal equilibrium and thus present a consistent position indication. Subsequent rod movement is considered to be 10 or more steps in one direction in < 1 hour.
Insert Page B 3.1.4-4 Attachment 1, Volume 6, Rev. 0, Page 96 of 213


2 INSERT 5
Attachment 1, Volume 6, Rev. 0, Page 97 of 213 Rod Group Alignment Limits B 3.1.4 BASES APPLICABILITY      The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant.
In MODES 3, 4, 5, and 6, the alignment limits do not apply because the normally control rods are bottomed and the reactor is shut down and not producing              8 fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.
ACTIONS            A.1.1 and A.1.2 When one or more rods are inoperable (i.e., untrippable), there is a possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration until the required SDM is recovered. The Completion Time of 1 hour is adequate for determining SDM and, if necessary, for initiating emergency boration and restoring SDM.
In this situation, SDM verification must include the worth of the untrippable rod, as well as a rod of maximum worth.
A.2                                      When one or more rods are inoperable 6
If the inoperable rod(s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 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 plant systems.
B.1 When a rod becomes misaligned, it can usually be moved and is still 6
trippable. If the rod can be realigned within the Completion Time of 1 hour, local xenon redistribution during this short interval will not be significant, and operation may proceed without further restriction.
WOG STS                                    B 3.1.4-5                                Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 97 of 213


when THERMAL POWER is 85% RTP or by 24 steps when THERMAL POWER is < 85% RTP  
Attachment 1, Volume 6, Rev. 0, Page 98 of 213 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued)
An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the                          6 bank sequence, overlap, and insertion limits specified in LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits." The Completion Time of 1 hour gives the operator sufficient time to adjust the rod positions in an orderly manner.
                                            , C.1.1, and C.1.2 6
B.2.1.1 and B.2.1.2 With a misaligned rod, SDM must be verified to be within limit or boration must be initiated to restore SDM to within limit.
In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully in and control bank C must be moved in to approximately 100 to 115 steps.
OPERABLE Power operation may continue with one RCCA trippable but misaligned,                              6 provided that SDM is verified within 1 hour. The Completion Time of 1 hour represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration.
C.2, C.3, C.4, C.5, and C.6 B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6                                    INSERT 8B 6
6 For continued operation with a misaligned rod, RTP must be reduced, SDM must periodically be verified within limits, hot channel factors (FQ(Z) and F N H ) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible.
                                    < 85                     INSERT 9 Reduction of power to 75% RTP ensures that local LHR increases due to                            6 a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 7). The Completion Time of 2 hours gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System.
WOG STS                                  B 3.1.4-6                                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 98 of 213


2 INSERT 6   For the dropped RCCA(s) or dropped RCCA bank misalignment accidents, a negative reactivity insertion will result. Power will be reestablished either by reactivity feedback or control bank withdrawal. Following plant stabilization, normal rod retrieval or shutdown procedures are followed. For dropped RCCA events in the automatic rod control mode, the Rod Control System detects the drop in power and initiates control bank withdrawal. In all cases, the minimum departure from nucleate boiling ratio (DNBR) remains above the limit value.  
Attachment 1, Volume 6, Rev. 0, Page 99 of 213 B 3.1.4 6    INSERT 8B when > 85% RTP, THERMAL POWER must be reduced to < 85% RTP. Once power is reduced, the alignment limits are now 24 steps instead of 12 steps.
Thus, once power is reduced, Condition B no longer applies. If a rod continues to be misaligned (i.e., it is greater than 24 steps misaligned), then as stated in Required Actions C.2, C.3, C.4, C.5 and C.6, 6
INSERT 9 when the indicated rod position exceeds 12 steps from their respective banks and < 50% RTP when the indicated rod position exceeds 24 steps from their respective banks Insert Page B 3.1.4-6 Attachment 1, Volume 6, Rev. 0, Page 99 of 213


Insert Page B 3.1.4-3 Rod Group Alignment Limits B 3.1.4    WOG STS B 3.1.4-4 Rev. 3.0, 03/31/04 BASES APPLICABLE SAFETY ANALYSES  (continued)
Attachment 1, Volume 6, Rev. 0, Page 100 of 213 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued)
When a rod is known to be misaligned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours is sufficient to ensure this requirement continues to be met.
Verifying that FQ(Z), as approximated by F QC ( Z) and F WQ ( Z ) , and F N H are
              < 85 INSERT 10 within the required limits ensures that current operation at 75% RTP with              6 a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate FQ(Z) and F N H .
Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of U
operation under these conditions. The accident analyses presented in FSAR Chapter 15 (Ref. 5) that may be adversely affected will be                        6 14 evaluated to ensure that the analysis results remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.
D C.1 associated When Required Actions cannot be completed within their Completion 6
Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit                3 6 must be brought to at least MODE 2 with Keff < 1.0 within 6 hours, which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.
E D.1.1 and D.1.2                                                                        6 More than one control rod becoming misaligned from its group average position is not expected, and has the potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires WOG STS                                     B 3.1.4-7                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 100 of 213


Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor ( F Q(Z)) and the nuclear enthalpy hot channel factor () are verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and F Q (Z) and F must be verified directly by incore mapping.
Attachment 1, Volume 6, Rev. 0, Page 101 of 213 B 3.1.4 o--
Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of F Q (Z) and F to the operating limits.
6 INSERT 10 when the indicated rod position exceeds 12 steps from their respective banks and < 50% RTP when the indicated rod position exceeds 24 steps from their respective banks Insert Page B 3.1.4-7 Attachment 1, Volume 6, Rev. 0, Page 101 of 213
H N FH NH N5 Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on control rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The control rod OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements, which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment. The rod OPERABILITY requirement is satisfied provided the rod will fully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability.
The requirement to maintain the rod alignment to within plus or minus


12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed.  
Attachment 1, Volume 6, Rev. 0, Page 102 of 213 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued) increasing the RCS boron concentration to provide negative reactivity, as of described in the Bases or LCO 3.1.1. The required Completion Time of            7 1 hour for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps.
Boration will continue until the required SDM is restored.
E                                                                                      6 D.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or            6 Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 2 with Keff < 1.0        6 within 6 hours.                                                            3 The allowed Completion Time is reasonable, based on operating 3 6
experience, for reaching MODE 2 with Keff < 1.0 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE        SR 3.1.4.1 REQUIREMENTS Verification that individual rod positions are within alignment limits at a INSERT 11 Frequency of 12 hours provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. The        2 specified Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected.
SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2 with Keff  1.0, tripping each control        6 rod would result in radial or axial power tilts, or oscillations. Exercising each individual control rod every 92 days 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 10 steps will not cause radial or axial power tilts, or oscillations, to occur.
The 92 day Frequency takes into consideration other information WOG STS                                    B 3.1.4-8                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 102 of 213


Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.
Attachment 1, Volume 6, Rev. 0, Page 103 of 213 B 3.1.4 2
INSERT 8 2 INSERT 8A 6INSERT 7 B 3.1.4 2 INSERT 7  with THERMAL POWER  85% RTP or within 24 steps of their group step counter demand position when THERMAL POWER is < 85% RTP
INSERT 11 The verification may be determined by the comparison of the bank demand step counter to the analog IRPI, by the rod position as noted on the plant process computer or through the conditioning module output voltage via a correlation of rod position versus voltage.
Insert Page B 3.1.4-8 Attachment 1, Volume 6, Rev. 0, Page 103 of 213


2 INSERT 8  
Attachment 1, Volume 6, Rev. 0, Page 104 of 213 Rod Group Alignment Limits B 3.1.4 BASES SURVEILLANCE REQUIREMENTS (continued) available to the operator in the control room and SR 3.1.4.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. 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.
SR 3.1.4.3 Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticality, after reactor vessel head removal, ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect control rod motion or drop time. This testing is performed with all RCPs operating and the average moderator temperature 500&deg;F to simulate a reactor trip under actual conditions.
This Surveillance is performed during a plant outage, due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power.
REFERENCES      1. 10 CFR 50, Appendix A, GDC 10 and GDC 26.
1 USAR, Sections 3.1.2.1, 3.1.2.2, 3.1.2.3, 3.1.2.4, 3.1.2.5, and 3.1.2.6.
: 2. 10 CFR 50.46.
U Section 14.1
: 3. FSAR, Chapter [15].                                                                                  2 4 U
Section 14.1.3                                                          2 4
: 4. FSAR, Chapter [15].
U 14                                                                      2 4
: 5. FSAR, Chapter [15].
: 6. FSAR, Chapter [15].
2 4
: 7. FSAR, Chapter [15].
WOG STS                                  B 3.1.4-9                                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 104 of 213


The safety analysis assumes a misalignment of one or more RCCA(s) or, in
Attachment 1, Volume 6, Rev. 0, Page 105 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4 BASES, ROD GROUP ALIGNMENT LIMITS
 
: 1. ISTS 3.1.4 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.
some cases, an entire RCCA bank.  
Therefore, ITS 3.1.4 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
 
Subsequent reference numbers have been renumbered.
6INSERT 8A The LCO is modified by a Note stating that individual control rod position indications may be outside the required limits for < 1 hour following subsequent control rod movement. This allows up to 1 hour of thermal soak time to allow the control rod drive shaft to reach thermal equilibrium and thus present a consistent position indication. Subsequent rod movement is considered to be 10 or more steps in one direction in <
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
1 hour. Insert Page B 3.1.4-4 Rod Group Alignment Limits B 3.1.4    WOG STS B 3.1.4-5 Rev. 3.0, 03/31/04  BASES APPLICABILITY The requirements on RCCA OPER ABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are bottomed and the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.
: 3. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
normally 8 ACTIONS A.1.1 and A.1.2 When one or more rods are inoperable (i.e., untrippable), there is a possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration until the required SDM is recovered. The Completion Time of 1 hour is adequate for determining SDM and, if necessary, for initiating emergency boration and restoring SDM.  
: 4. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
 
: 5. The discussion of the Required Actions when the LCO is not met has been deleted since it is not appropriate in the Applicable Analyses Section. This information is adequately discussed in the Bases for ACTIONS B.2, B.3, B.4, B.5, B.6, C.2, C.3, C.4, C.5 and C.6.
In this situation, SDM verification must include the worth of the untrippable rod, as well as a rod of maximum worth.
: 6. Changes made to be consistent with the Specification.
 
: 7. Typographical/grammatical error corrected.
A.2 When one or more rods are inoperable 6If the inoperable rod(s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours.
: 8. The word "normally" has been added since the shutdown banks are required by ITS LCO 3.1.5 to be withdrawn to their insertion limits prior to entering MODE 2 (i.e., in MODE 3). Thus the rods are not bottomed all the time in MODE 3.
The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
Kewaunee Power Station                  Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 105 of 213
 
B.1 When a rod becomes misaligned, it can usually be moved and is still trippable. If the rod can be realigned within the Completion Time of 1 hour, local xenon redistribution during this short interval will not be significant, and operation may proceed without further restriction.
6 Rod Group Alignment Limits B 3.1.4    WOG STS B 3.1.4-6 Rev. 3.0, 03/31/04  BASES ACTIONS  (continued)


An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits."  The Completion Time of 1 hour gives the operator sufficient time to adjust the rod positions in an orderly manner.
Attachment 1, Volume 6, Rev. 0, Page 106 of 213 Specific No Significant Hazards Considerations (NSHCs)
6 B.2.1.1 and B.2.1.2 With a misaligned rod, SDM must be verified to be within limit or boration must be initiated to restore SDM to within limit.
Attachment 1, Volume 6, Rev. 0, Page 106 of 213


In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully in and control bank C must be moved in to approximately 100 to 115 steps.  
Attachment 1, Volume 6, Rev. 0, Page 107 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS There are no specific NSHC discussions for this Specification.
Kewaunee Power Station              Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 107 of 213


Power operation may continue with one RCCA trippable but misaligned, provided that SDM is verified within 1 hour. The Completion Time of 1 hour represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration.  
Attachment 1, Volume 6, Rev. 0, Page 108 of 213 ATTACHMENT 5 ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Attachment 1, Volume 6, Rev. 0, Page 108 of 213
, Volume 6, Rev. 0, Page 109 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 109 of 213


B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6
Attachment 1, Volume 6, Rev. 0, Page 110 of 213 ITS                                                         A01                                            ITS 3.1.5
 
: c. Quadrant Power Tilt Limits
For continued operation with a misaligned rod, RTP must be reduced, SDM must periodically be verified within limits, hot channel factors (F Q (Z) and F) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible.
H N Reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 7). The Completion Time of 2 hours gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System.
,, C.1.1, and C.1.2 6OPERABLE 6C.2, C.3, C.4, C.5, and C.6 6INSERT 8B 6< 85 INSERT 9 6 B 3.1.4 INSERT 8B 6  when > 85% RTP, THERMAL POWER must be reduced to < 85% RTP. Once power is reduced, the alignment limits are now 24 steps instead of 12 steps. Thus, once power is reduced, Condition B no longer applies. If a rod continues to be misaligned (i.e., it is greater than 24 steps misaligned), then as stated in
 
Required Actions C.2, C.3, C.4, C.5 and C.6,
 
6 INSERT 9 when the indicated rod position exceeds 12 steps from their respective banks and < 50% RTP when the indicated rod position exceeds 24 steps from their respective banks
 
Insert Page B 3.1.4-6 Rod Group Alignment Limits B 3.1.4    WOG STS B 3.1.4-7 Rev. 3.0, 03/31/04  BASES ACTIONS  (continued)
 
When a rod is known to be misaligned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours is sufficient to ensure this requirement continues to be met.
 
Verifying that F Q(Z), as approximated by  and , and are within the required limits ensures that current operation at 75% RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate F Q(Z) and F. )Z(F C Q)Z(F W Q H N FH N Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. The accident analyses presented in FSAR Chapter 15 (Ref. 5) that may be adversely affected will be evaluated to ensure that the analysis results remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.
 
C.1  When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 2 with K eff < 1.0 within 6 hours, which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.
D.1.1 and D.1.2
 
More than one control rod becoming misaligned from its group average position is not expected, and has the potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires  INSERT 10 < 85 6 14 U 6 D associated 6 6 6 3 E B 3.1.4 6 INSERT 10  when the indicated rod position exceeds 12 steps from their respective banks and < 50% RTP when the indicated rod position exceeds 24 steps from their respective banks
 
Insert Page B 3.1.4-7 o --
Rod Group Alignment Limits B 3.1.4    WOG STS B 3.1.4-8 Rev. 3.0, 03/31/04  BASES ACTIONS  (continued)
 
increasing the RCS boron concentration to provide negative reactivity, as described in the Bases or LCO 3.1.1. The required Completion Time of 1 hour for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored.
of 7 D.2  E 6If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 2 with K eff < 1.0 within 6 hours.
6 6 3 The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 2 with K eff < 1.0  from full power conditions in an orderly manner and without challenging plant systems.
3 6 SURVEILLANCE SR  3.1.4.1 REQUIREMENTS Verification that individual rod positions are within alignment limits at a Frequency of 12 hours provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. The specified Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected. INSERT 11 2
SR  3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be
 
tripped. However, in MODES 1 and 2 with K eff  1.0, tripping each control rod would result in radial or axial power tilts, or oscillations. Exercising each individual control rod every 92 days 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 10 steps will not cause radial or axial power tilts, or oscillations, to occur.
The 92 day Frequency takes into consideration other information 6
B 3.1.4 2 INSERT 11  The verification may be determined by the comparison of the bank demand step counter to the analog IRPI, by the rod position as noted on the plant process computer or through the conditioning module output voltage via a correlation of rod position versus voltage.
 
Insert Page B 3.1.4-8 Rod Group Alignment Limits B 3.1.4    WOG STS B 3.1.4-9 Rev. 3.0, 03/31/04  BASES SURVEILLANCE REQUIREMENTS  (continued)
 
available to the operator in the control room and SR 3.1.4.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. 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.
 
SR  3.1.4.3
 
Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticality, after reactor vessel head removal, ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect control rod motion or drop time. This testing is performed with all RCPs operating and the average moderator
 
temperature  500&deg;F to simulate a reactor trip under actual conditions.
This Surveillance is performed during a plant outage, due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at
 
power. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26.
1
: 2. 10 CFR 50.46. USAR, Sections 3.1.2.1, 3.1.2.2, 3.1.2.3, 3.1.2.4, 3.1.2.5, and 3.1.2.6.
: 3. FSAR, Chapter [15].
: 4. FSAR, Chapter [15].
: 5. FSAR, Chapter [15].
U U U 2 4 14Section 14.1.3 Section 14.1 2 4 2 4  6. FSAR, Chapter [15].
: 7. FSAR, Chapter [15].
2 4 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4 BASES, ROD GROUP ALIGNMENT LIMITS
: 1. ISTS 3.1.4 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A. Therefore, ITS 3.1.4 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
Subsequent reference numbers have been renumbered.
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 3. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
: 4. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 5. The discussion of the Required Actions when the LCO is not met has been deleted since it is not appropriate in the Applicable Analyses Section. This information is adequately discussed in the Bases for ACTIONS B.2, B.3, B.4, B.5, B.6, C.2, C.3, C.4, C.5 and C.6.
: 6. Changes made to be consistent with the Specification.
: 7. Typographical/grammatical error corrected.
: 8. The word "normally" has been added since the shutdown banks are required by ITS LCO 3.1.5 to be withdrawn to their insertion limits prior to entering MODE 2 (i.e., in MODE 3). Thus the rods are not bottomed all the time in MODE 3.
Kewaunee Power Station Page 1 of 1 Specific No Significant Haza rds Considerations (NSHCs)
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS There are no specific NSHC discussions for this Specification.
Kewaunee Power Station Page 1 of 1 ATTACHMENT 5 ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
A01 ITS 3.1.5ITS c. Quadrant Power Tilt Limits  
: 1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:
: 1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:
A. Eliminate the tilt.
A. Eliminate the tilt.
B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio  
B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio
> 1.0. 2. If the tilt condition is not eliminated after 24 hours, then reduce power to 50% or lower.  
                          > 1.0.
: 3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is simultaneous indication of a misaligned rod:
: 2. If the tilt condition is not eliminated after 24 hours, then reduce power to 50% or lower.
See ITS
: 3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is               3.2.4 simultaneous indication of a misaligned rod:
A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.
A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.
B. If the tilt condition is not eliminated within 12 hours, then the reactor shall be brought to a minimum load condition (30 Mwe). 4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition  
B. If the tilt condition is not eliminated within 12 hours, then the reactor shall be brought to a minimum load condition ( 30 Mwe).
 
: 4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition
( 5% reactor power).  
( 5% reactor power).
: d. Rod Insertion Limits  
: d. Rod Insertion Limits LCO 3.1.5
: 1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when the reactor is critical or approaching criticality.  
: 1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when                   A02 Applicability      the reactor is critical or approaching criticality.
: 2. The control banks shall be limited in physical insertion; insertion limits are specified in the COLR. If any one of the control bank insertion limits is not met:
Add proposed ACTIONS A and B
A. Within one hour, initiate boration to restore control bank insertion to within the limits specified in the COLR, and B. Restore control bank insertion to within the limits specified in the COLR within two hours of exceeding the insertion limits.
: 2. The control banks shall be limited in physical insertion; insertion limits are specified in             L01 the COLR. If any one of the control bank insertion limits is not met:
C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then withinone hour action shall be initiated to:  
A. Within one hour, initiate boration to restore control bank insertion to within the limits specified in the COLR, and B. Restore control bank insertion to within the limits specified in the COLR within                   See ITS 3.1.6 two hours of exceeding the insertion limits.
- Achieve HOT STANDBY within 6 hours
C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then withinone hour action shall be initiated to:
- Achieve HOT SHUTDOWN within the following 6 hours  
                        - Achieve HOT STANDBY within 6 hours
: 3. Insertion limit does not apply during physics tests or during periodic exercise of individual rods. However, the shutdown margin, as specified in the COLR, must be maintained except for the Low Power Physics Test to measure control rod worth and shutdown margin. For this test, the reactor may be critical with all but one high worth  
                        - Achieve HOT SHUTDOWN within the following 6 hours See ITS 3.1.8 Applicability    3. Insertion limit does not apply during physics tests or during periodic exercise of Note                individual rods. However, the shutdown margin, as specified in the COLR, must be See ITS maintained except for the Low Power Physics Test to measure control rod worth and                       3.1.8 shutdown margin. For this test, the reactor may be critical with all but one high worth rod inserted.
 
Add proposed SR 3.1.5.1             M01 Amendment No. 165 TS 3.10-4                                         03/11/2003 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 110 of 213
rod inserted. L01Add proposed ACTIONS A and BA02 A pplicabilit y LCO 3.1.5 See ITS 3.1.8 See ITS 3.1.8 A pplicabilit y Note Add proposed SR 3.1.5.1 M01See ITS 3.1.6 See ITS 3.2.4  Amendment No. 165 TS 3.10-4 03/11/2003 Page 1 of 1 DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).


Attachment 1, Volume 6, Rev. 0, Page 111 of 213 DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS ADMINISTRATIVE CHANGES A01  In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.
3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
A02 CTS 3.10.d.1 requires the shutdown rods to be withdrawn to within the limits specified in the COLR when the reactor is critical or approaching criticality. ITS 3.1.5 requires each shutdown bank to be within the insertion limits specified in the COLR in MODES 1 and 2. This changes the CTS by clearly specifying the MODES in which the LCO is required.  
A02 CTS 3.10.d.1 requires the shutdown rods to be withdrawn to within the limits specified in the COLR when the reactor is critical or approaching criticality. ITS 3.1.5 requires each shutdown bank to be within the insertion limits specified in the COLR in MODES 1 and 2. This changes the CTS by clearly specifying the MODES in which the LCO is required.
 
The CTS condition of "reactor is critical" normally occurs only in ITS MODE 1 and MODE 2 with keff > 1.0. The CTS condition of "approaching criticality" normally occurs only in MODE 2 with keff < 1.0. Therefore, since the ITS 3.1.5 MODES 1 and 2 are equivalent to the CTS conditions, this change is acceptable and is designated as administrative.
The CTS condition of "reactor is critical" normally occurs only in ITS MODE 1 and MODE 2 with k eff > 1.0. The CTS condition of "approaching criticality" normally occurs only in MODE 2 with k eff < 1.0. Therefore, since the ITS 3.1.5 MODES 1 and 2 are equivalent to the CTS conditions, this change is acceptable and is designated as administrative.  
MORE RESTRICTIVE CHANGES M01  CTS 3.10.d.1 requires the shutdown rods to be withdrawn to within the limits specified in the COLR, but does not provide a Surveillance Requirement to verify that the shutdown rods are within the limits specified in the COLR. ITS SR 3.1.5.1 requires verification that the shutdown banks are within the insertion limits specified in the COLR every 12 hours. This changes the CTS by adding a new Surveillance Requirement to verify that the shutdown banks are within the insertion limits specified in the COLR every 12 hours.
 
The purpose of ITS SR 3.1.5.1 is to verify that the shutdown banks are within their insertion limits, available to shut down the reactor, and that the required SHUTDOWN MARGIN (SDM) will be maintained following a reactor trip. The 12 hour Frequency takes into account other information that is available in the control room for monitoring the status of the shutdown rods. This change is designated as more restrictive because a new Surveillance Requirement has been added to verify that the shutdown banks are within the insertion limits specified in the COLR every 12 hours.
MORE RESTRICTIVE CHANGES
RELOCATED SPECIFICATIONS None Kewaunee Power Station                  Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 111 of 213


M01 CTS 3.10.d.1 requires the shutdown rods to be withdrawn to within the limits specified in the COLR, but does not provide a Surveillance Requirement to verify that the shutdown rods are within the limits specified in the COLR. ITS SR 3.1.5.1 requires verification that the shutdown banks are within the insertion limits specified in the COLR every 12 hours. This changes the CTS by adding a  
Attachment 1, Volume 6, Rev. 0, Page 112 of 213 DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01  (Category 4 - Relaxation of Required Action) CTS 3.10.d.1 does not provide any explicit time to restore the shutdown rods to within their limits. As a result, LCO 3.0.c would be entered, which requires action to be initiated within 1 hour, to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours, and in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours.
Under similar conditions, ITS 3.1.5 ACTION A provides one hour to either verify that the SDM is within the limits specified in the COLR or to initiate boration to restore the SDM to within limits and 2 hours to restore the shutdown banks to within limits. Additionally, ITS 3.1.5 ACTION B provides 6 hours to be in MODE 3 if any Required Action and associated Completion Time of ACTION A are not met. This changes the CTS by providing specific ACTIONS when the shutdown banks are not within the limits specified in the COLR.
The purpose of CTS 3.10.d.1 is to ensure that the shutdown rods are within limits specified in the COLR. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the shutdown rods not being within the limits established in the COLR because the SDM may be significantly reduced. ITS 3.1.5 Required Action A.2 provides 2 hours in lieu of the 1 hour in CTS 3.10.d.1 to restore the shutdown bank to within limits. This is acceptable because ITS 3.1.5 Required Action A.2 provides sufficient time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time. The Completion Time of ITS 3.1.5 Required Action B.1 provides 6 hours to be in MODE 3 which is a reasonable time, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging the plant systems. Additionally, since ITS 3.1.5 Required Action B.1 requires the unit to be in MODE 3 within 6 hours, there is no need to maintain the requirement to be in MODE 2 within the same 6 hours. This change is designated as less restrictive because more time is provided in the ITS to restore the shutdown banks to within limits prior to requiring a unit shutdown.
Kewaunee Power Station                  Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 112 of 213


new Surveillance Requirement to verify that the shutdown banks are within the insertion limits specified in the COLR every 12 hours.  
Attachment 1, Volume 6, Rev. 0, Page 113 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 113 of 213


The purpose of ITS SR 3.1.5.1 is to verify that the shutdown banks are within their insertion limits, available to shut down the reactor, and that the required SHUTDOWN MARGIN (SDM) will be maintained following a reactor trip. The 12 hour Frequency takes into account other information that is available in the control room for monitoring the status of the shutdown rods. This change is designated as more restrictive because a new Surveillance Requirement has been added to verify that the shutdown banks are within the insertion limits specified in the COLR every 12 hours.  
Attachment 1, Volume 6, Rev. 0, Page 114 of 213 CTS                                                                                    Shutdown Bank Insertion Limits 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5      Shutdown Bank Insertion Limits 3.10.d.1 LCO 3.1.5              Each shutdown bank shall be within insertion limits specified in the COLR.
3.10.d.1 APPLICABILITY:        MODES 1 and 2.
                                ------------------------------------------NOTE-----------------------------------------------
3.10.d.3                        This LCO is not applicable while performing SR 3.1.4.2.
ACTIONS CONDITION                            REQUIRED ACTION                            COMPLETION TIME DOC L01  A. One or more shutdown             A.1.1      Verify SDM is within the                 1 hour banks not within limits.                  limits specified in the COLR.
OR A.1.2      Initiate boration to restore            1 hour SDM to within limit.
AND A.2        Restore shutdown banks to                2 hours within limits.
DOC L01  B. Required Action and              B.1        Be in MODE 3.                            6 hours associated Completion Time not met.
WOG STS                                              3.1.5-1                                    Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 114 of 213


RELOCATED SPECIFICATIONS None
Attachment 1, Volume 6, Rev. 0, Page 115 of 213 CTS                                                                Shutdown Bank Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY DOC M01 SR 3.1.5.1    Verify each shutdown bank is within the insertion    12 hours limits specified in the COLR.
WOG STS                                3.1.5-2                      Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 115 of 213


Kewaunee Power Station Page 1 of 2 DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Kewaunee Power Station Page 2 of 2 REMOVED DETAIL CHANGES None
Attachment 1, Volume 6, Rev. 0, Page 116 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS None Kewaunee Power Station           Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 116 of 213


LESS RESTRICTIVE CHANGES
Attachment 1, Volume 6, Rev. 0, Page 117 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 117 of 213


L01 (Category 4 - Relaxation of Required Action)
Attachment 1, Volume 6, Rev. 0, Page 118 of 213 Shutdown Bank Insertion Limits B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Shutdown Bank Insertion Limits BASES BACKGROUND             The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available ejected rod worth, SDM and initial reactivity insertion rate.
CTS 3.10.d.1 does not provide any explicit time to restore the shutdown rods to within their limits. As a result, LCO 3.0.c would be entered, which requires action to be initiated within 1 hour, to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours, and in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours. Under similar conditions, ITS 3.1.5 ACTION A provides one hour to either verify that the SDM is within the limits specified in the COLR or to initiate boration to restore the SDM to within limits and 2 hours to restore the shutdown banks to within limits. Additionally, ITS 3.1.5 ACTION B provides 6 hours to be in MODE 3 if any Required Action and associated Completion Time of ACTION A are not met. This changes the CTS by providing specific ACTIONS when the shutdown banks are not within the limits specified in the COLR. 
The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design,"                 1 INSERT 1    GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.
 
The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of If  two or more RCCAs that are electrically paralleled to step simultaneously.
The purpose of CTS 3.10.d.1 is to ensure that the shutdown rods are within limits specified in the COLR. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the shutdown rods not being within the limits established in the COLR because the SDM may be significantly reduced. ITS 3.1.5 Required Action A.2 provides 2 hours in lieu of the 1 hour in CTS 3.10.d.1 to restore the shutdown bank to within limits. This is acceptable because ITS 3.1.5 Required Action A.2 provides sufficient time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time. The Completion Time of ITS 3.1.5 Required Action B.1 provides 6 hours to be in MODE 3 which is a reasonable time, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging the plant systems. Additionally, since ITS 3.1.5 Required Action B.1 requires the unit to be in MODE 3 within 6 hours, there is no need to maintain the requirement to be in MODE 2 within the same 6 hours. This change is designated as less restrictive because more time is provided in the ITS to restore the shutdown banks to within limits prior to requiring a unit shutdown.
          , the groups A bank of RCCAs consists of two groups that are moved in a staggered               4 fashion, but always within one step of each other. All plants have four Kewaunee Power 2 control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Station Group Alignment Limits," for control and shutdown rod OPERABILITY (KPS) has and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements.
 
Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Shutdown Bank Insertion Limits 3.1.5  WOG STS 3.1.5-1 Rev. 3.0, 03/31/04  CTS 3.1  REACTIVITY CONTROL SYSTEMS
 
====3.1.5 Shutdown====
Bank Insertion Limits
 
LCO  3.1.5 Each shutdown bank shall be within insertion limits specified in the COLR. 3.10.d.1 APPLICABILITY: MODES 1 and 2.
3.10.d.1    ------------------------------------------NOTE-----------------------------------------------    This LCO is not applicable while performing SR 3.1.4.2.
3.10.d.3    --------------------------------------------------------------------------------------------------
 
ACTIONS  CONDITION REQUIRED ACTION COMPLETION TIME
 
A. One or more shutdown banks not within limits.
 
A.1.1 Verify SDM is within the limits specified in the COLR.        OR A.1.2 Initiate boration to restore SDM to within limit.
 
AND A.2 Restore shutdown banks to within limits.
 
1 hour
 
1 hour 
 
2 hours B. Required Action and associated Completion Time not met.
 
B.1 Be in MODE 3.
 
6 hours DOC L01 DOC L01 Shutdown Bank Insertion Limits 3.1.5  WOG STS 3.1.5-2 Rev. 3.0, 03/31/04  CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR  3.1.5.1 Verify each shutdown bank is within the insertion limits specified in the COLR.
 
12 hours DOC M01 JUSTIFICATION FOR DEVIATIONS ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS None  Kewaunee Power Station Page 1 of 1 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Shutdown Bank Insertion Limits B 3.1.5   WOG STS B 3.1.5-1 Rev. 3.0, 03/31/04 B 3.1  REACTIVITY CONTROL SYSTEMS  
 
B 3.1.5 Shutdown Bank Insertion Limits  
 
BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available ejected rod worth, SDM and initial reactivity insertion rate.  
 
The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. INSERT 1 1 The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. A bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one step of each other. All plants have four control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Group Alignment Limits," for control and shutdown rod OPERABILITY and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements.
The control banks are used for precise reactivity control of the reactor.
The control banks are used for precise reactivity control of the reactor.
The positions of the control banks are normally automatically controlled by the Rod Control System, but they can also be manually controlled. They are capable of adding negative reactivity very quickly (compared to borating). The control banks must be maintained above designed insertion limits and are typically near the fully withdrawn position during  
The positions of the control banks are normally automatically controlled 2
by the Rod Control System, but they can also be manually controlled.
They are capable of adding negative reactivity very quickly (compared to borating). The control banks must be maintained above designed insertion limits and are typically near the fully withdrawn position during normal full power operations.
WOG STS                                          B 3.1.5-1                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 118 of 213


normal full power operations.  
Attachment 1, Volume 6, Rev. 0, Page 119 of 213 B 3.1.5 o 1
                                    -INSERT 1 USAR GDC 6, "Reactor Core Design," GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot Shutdown Capability," GDC 29, "Reactivity Shutdown Capability," and GDC 30, "Reactivity Holddown Capability" (Ref. 1),
Insert Page B 3.1.5-1 Attachment 1, Volume 6, Rev. 0, Page 119 of 213


KewauneePower Station (KPS) has 4, the groups If 2 2 B 3.1.5 1 INSERT 1    USAR GDC 6, "Reactor Core Design," GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot S hutdown Capability," GDC 29, "Reactivity Shutdown Capability," and GDC 30, "Reactivity Holddown Capability" (Ref. 1),
Attachment 1, Volume 6, Rev. 0, Page 120 of 213 Shutdown Bank Insertion Limits B 3.1.5 BASES BACKGROUND (continued)
 
Hence, they are not capable of adding a large amount of positive reactivity. Boration or dilution of the Reactor Coolant System (RCS) compensates for the reactivity changes associated with large changes in RCS temperature. The design calculations are performed with the assumption that the shutdown banks are withdrawn first. The shutdown banks can be fully withdrawn without the core going critical. This provides available negative reactivity in the event of boration errors. The shutdown banks are controlled manually by the control room operator.
Insert Page B 3.1.5-1 o -
During normal unit operation, the shutdown banks are either fully              1 withdrawn or fully inserted. The shutdown banks must be completely withdrawn from the core, prior to withdrawing any control banks during an approach to criticality. The shutdown banks are then left in this position until the reactor is shut down. They affect core power and burnup distribution, and add negative reactivity to shut down the reactor upon receipt of a reactor trip signal.
Shutdown Bank Insertion Limits B 3.1.5    WOG STS B 3.1.5-2 Rev. 3.0, 03/31/04 BASES BACKGROUND  (continued)
APPLICABLE      On a reactor trip, all RCCAs (shutdown banks and control banks), except SAFETY          the most reactive RCCA, are assumed to insert into the core. The ANALYSES        shutdown banks shall be at or above their insertion limits and available to insert the maximum amount of negative reactivity on a reactor trip signal.
The control banks may be partially inserted in the core, as allowed by LCO 3.1.6, "Control Bank Insertion Limits." The shutdown bank and control bank insertion limits are established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") following a reactor trip from full power. The combination of control banks and shutdown banks (less the most reactive RCCA, which is assumed to be fully withdrawn) is sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 3). The shutdown bank insertion limit also limits the reactivity worth of an ejected shutdown rod.
The acceptance criteria for addressing shutdown and control rod bank insertion limits and inoperability or misalignment is that:
: a. There be no violations of:
: 1. Specified acceptable fuel design limits or                              3
: 2. RCS pressure boundary integrity and                                    3 WOG STS                                 B 3.1.5-2                           Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 120 of 213


Hence, they are not capable of adding a large amount of positive reactivity. Boration or dilution of the Reactor  Coolant System (RCS) compensates for the reactivity changes associated with large changes in RCS temperature. The design calculations are performed with the assumption that the shutdown banks are withdrawn first. The shutdown banks can be fully withdrawn without the core going critical. This provides available negative reactivity in the event of boration errors. The shutdown banks are controlled manually by the control room operator.
Attachment 1, Volume 6, Rev. 0, Page 121 of 213 Shutdown Bank Insertion Limits B 3.1.5 BASES APPLICABLE SAFETY ANALYSES (continued)
During normal unit operation, the shutdown banks are either fully withdrawn or fully inserted. The shutdown banks must be completely withdrawn from the core, prior to withdrawing any control banks during an approach to criticality. The shutdown banks are then left in this position until the reactor is shut down. They affect core power and burnup distribution, and add negative reactivity to shut down the reactor upon receipt of a reactor trip signal.
1 APPLICABLE On a reactor trip, all RCCAs (s hutdown banks and contro l banks), except SAFETY  the most reactive RCCA, are assumed to insert into the core. The ANALYSES shutdown banks shall be at or above their insertion limits and available to insert the maximum amount of negative reactivity on a reactor trip signal.
The control banks may be partially inserted in the core, as allowed by LCO 3.1.6, "Control Bank Insertion Limits."  The shutdown bank and control bank insertion limits are established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") following a reactor trip from full power. The combination of control banks and shutdown banks (less the most reactive RCCA, which is assumed to be fully withdrawn) is sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 3). The shutdown bank insertion limit also limits the reactivity worth of an ejected shutdown rod.
The acceptance criteria for addressing shutdown and control rod bank insertion limits and inoperability or misalignment is that:
: a. There be no violations of:
: 1. Specified acceptable fuel design limits or 3 2. RCS pressure boundary integrity and
; 3; Shutdown Bank Insertion Limits B 3.1.5   WOG STS B 3.1.5-3 Rev. 3.0, 03/31/04  BASES APPLICABLE SAFETY ANALYSES (continued)  
: b. The core remains subcritical after accident transients.
: b. The core remains subcritical after accident transients.
As such, the shutdown bank insertion limits affect safety analysis  
As such, the shutdown bank insertion limits affect safety analysis involving core reactivity and SDM (Ref. 3).
 
involving core reactivity and SDM (Ref. 3).  
 
The shutdown bank insertion limits preserve an initial condition assumed in the safety analyses and, as such, satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
The shutdown bank insertion limits preserve an initial condition assumed in the safety analyses and, as such, satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The shutdown banks must be within their insertion limits any time the reactor is critical or approaching criticality. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip.
LCO             The shutdown banks must be within their insertion limits any time the reactor is critical or approaching criticality. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip.
The shutdown bank insertion limits are defined in the COLR.
The shutdown bank insertion limits are defined in the COLR.
APPLICABILITY The shutdown banks must be within their insertion limits, with the reactor in MODES 1 and 2. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. The shutdown banks do not have to be within their insertion limits in MODE 3, unless an approach to criticality is being made. In MODE 3, 4, 5, or 6, the shutdown banks are fully inserted in the core and contribute to the SDM. Refer to LCO 3.1.1 for SDM requirements in MODES 3, 4, and 5. LCO 3.9.1, "Boron Concentration,"
APPLICABILITY   The shutdown banks must be within their insertion limits, with the reactor in MODES 1 and 2. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. The shutdown banks do not have to be within their insertion limits in MODE 3, unless an approach to criticality is being made. In MODE 3, 4, 5, or 6, the shutdown banks are fully inserted in the core and contribute to the SDM. Refer to LCO 3.1.1 for SDM requirements in MODES 3, 4, and 5. LCO 3.9.1, "Boron Concentration,"
ensures adequate SDM in MODE 6.
ensures adequate SDM in MODE 6.
The Applicability requirements have been modified by a Note indicating the LCO requirement is suspended during SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the shutdown bank to move below the LCO limits, which would normally violate the LCO.
The Applicability requirements have been modified by a Note indicating the LCO requirement is suspended during SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the shutdown bank to move below the LCO limits, which would normally violate the LCO.
ACTIONS A.1.1, A.1.2, and A.2 When one or more shutdown banks is not within insertion limits, 2 hours is allowed to restore the shutdown banks to within the insertion limits.
ACTIONS         A.1.1, A.1.2, and A.2 When one or more shutdown banks is not within insertion limits, 2 hours is allowed to restore the shutdown banks to within the insertion limits.
This is necessary because the available SDM may be significantly reduced, with one or more of the shutdown banks not within their insertion limits. Also, verification of SDM or initiation of boration within 1 hour is required, since the SDM in MODES 1 and 2 is ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1). If shutdown banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1.
This is necessary because the available SDM may be significantly reduced, with one or more of the shutdown banks not within their insertion limits. Also, verification of SDM or initiation of boration within 1 hour is required, since the SDM in MODES 1 and 2 is ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1). If shutdown banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1.                                                   5 WOG STS                                 B 3.1.5-3                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 121 of 213
5 Shutdown Bank Insertion Limits B 3.1.5   WOG STS B 3.1.5-4 Rev. 3.0, 03/31/04 BASES ACTIONS  (continued)


The allowed Completion Time of 2 hours provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.  
Attachment 1, Volume 6, Rev. 0, Page 122 of 213 Shutdown Bank Insertion Limits B 3.1.5 BASES ACTIONS (continued)
The allowed Completion Time of 2 hours provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.
B.1                                any Required Action and associated Completion Time is not met 6
If the shutdown banks cannot be restored to within their insertion limits within 2 hours, the unit must be brought to a MODE where the LCO is not applicable. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE      SR 3.1.5.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip.
This SR and Frequency ensure that the shutdown banks are withdrawn before the control banks are withdrawn during a unit startup.
Since the shutdown banks are positioned manually by the control room operator, a verification of shutdown bank position at a Frequency of 12 hours, after the reactor is taken critical, is adequate to ensure that they are within their insertion limits. Also, the 12 hour Frequency takes into account other information available in the control room for the purpose of monitoring the status of shutdown rods.
REFERENCES        1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28.                                            1 USAR, Sections 3.1.2.1, 3.1.2.3, 3.1.2.4, 3.1.2.5,
: 2. 10 CFR 50.46.                                              and 3.1.2.6.
U                  Section 14.1
: 3. FSAR, Chapter [15].                                                                            2 WOG STS                                  B 3.1.5-4                                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 122 of 213


B.1   If the shutdown banks cannot be restored to within their insertion limits within 2 hours, the unit must be brought to a MODE where the LCO is not applicable. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems. any Required Action and associated Completion Time is not met 6 SURVEILLANCE SR  3.1.5.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip.
Attachment 1, Volume 6, Rev. 0, Page 123 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.5 BASES, SHUTDOWN BANK INSERTION LIMITS
This SR and Frequency ensure that the shutdown banks are withdrawn before the control banks are withdrawn during a unit startup.
: 1. ISTS 3.1.5 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.
Since the shutdown banks are positi oned manually by the control room operator, a verification of shutdown bank position at a Frequency of 12 hours, after the reactor is taken critical, is adequate to ensure that they are within their insertion limits. Also, the 12 hour Frequency takes into account other information available in the control room for the purpose of monitoring the status of shutdown rods.
Therefore, ITS 3.1.5 Bases Background has been revised to discuss the design standards used by KPS. Additionally, Bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28.
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
1USAR, Sections 3.1.2.1, 3.1.2.3, 3.1.2.4, 3.1.2.5, and 3.1.2.6.
: 3. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
: 2. 10 CFR 50.46.
: 4. Words changed to be consistent with similar words in the ISTS 3.1.4 Bases. Not all banks are divided into two groups at KPS.
: 3. FSAR, Chapter [15].
: 5. Typographical error corrected.
U Section 14.1 2
JUSTIFICATION FOR DEVIATIONS ITS 3.1.5 BASES, SHUTDOWN BANK INSERTION LIMITS  
: 1. ISTS 3.1.5 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A. Therefore, ITS 3.1.5 Bases Background has been revised to discuss the design standards used by KPS. Additionally, Base s references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.  
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.  
: 3. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.  
: 4. Words changed to be consistent with similar words in the ISTS 3.1.4 Bases. Not all banks are divided into two groups at KPS.  
: 5. Typographical error corrected.  
: 6. Change made to be consistent with the Specification. The words in the ISTS Bases only describe Required Action A.2. ACTION B must be entered if any Required Action and associated Completion Time of Condition A is not met.
: 6. Change made to be consistent with the Specification. The words in the ISTS Bases only describe Required Action A.2. ACTION B must be entered if any Required Action and associated Completion Time of Condition A is not met.
Kewaunee Power Station Page 1 of 1 Specific No Significant Haza rds Considerations (NSHCs)
Kewaunee Power Station                 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 123 of 213
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS There are no specific NSHC discussions for this Specification.
 
Kewaunee Power Station Page 1 of 1 ATTACHMENT 6 ITS 3.1.6, CONTROL BANK INSERTION LIMITS Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
Attachment 1, Volume 6, Rev. 0, Page 124 of 213 Specific No Significant Hazards Considerations (NSHCs)
A01 ITS 3.1.6 ITS c. Quadrant Power Tilt Limits  
Attachment 1, Volume 6, Rev. 0, Page 124 of 213
 
Attachment 1, Volume 6, Rev. 0, Page 125 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS There are no specific NSHC discussions for this Specification.
Kewaunee Power Station             Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 125 of 213
, Volume 6, Rev. 0, Page 126 of 213 ATTACHMENT 6 ITS 3.1.6, CONTROL BANK INSERTION LIMITS , Volume 6, Rev. 0, Page 126 of 213
, Volume 6, Rev. 0, Page 127 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 127 of 213
 
Attachment 1, Volume 6, Rev. 0, Page 128 of 213 ITS                                                          A01                                                           ITS 3.1.6
: c. Quadrant Power Tilt Limits
: 1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:
: 1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:
A. Eliminate the tilt.
A. Eliminate the tilt.
B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio  
B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio
> 1.0. 2. If the tilt condition is not eliminated after 24 hours, then reduce power to 50% or lower.  
                          > 1.0.
: 3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is simultaneous indication of a misaligned rod:
: 2. If the tilt condition is not eliminated after 24 hours, then reduce power to 50% or lower.
See ITS
: 3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is                                 3.2.4 simultaneous indication of a misaligned rod:
A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.
A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.
B. If the tilt condition is not eliminated within 12 hours, then the reactor shall be brought to a minimum load condition (30 Mwe). 4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition
B. If the tilt condition is not eliminated within 12 hours, then the reactor shall be brought to a minimum load condition ( 30 Mwe).
 
: 4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition
( 5% reactor power).  
( 5% reactor power).
: d. Rod Insertion Limits  
: d. Rod Insertion Limits See ITS
: 1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when the reactor is critical or approaching criticality.  
: 1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when                                   3.1.5 the reactor is critical or approaching criticality.
: 2. The control banks shall be limited in physical insertion; insertion limits are specified in the COLR. If any one of the control bank insertion limits is not met:
                                                                                                  , sequence, and overlap limits          A02 LCO 3.1.6
A. Within one hour, initiate boration to restore control bank insertion to within the limits specified in the COLR, and B. Restore control bank insertion to within the limits specified in the COLR within two hours of exceeding the insertion limits. C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then withinone hour action shall be initiated to:  
: 2. The control banks shall be limited in physical insertion; insertion limits are specified in L01 the COLR. If any one of the control bank insertion limits is not met: Add proposed Applicability Add proposed Required Actions A.1.1 and B.1.1                                      L02 A. Within one hour, initiate boration to restore control bank insertion to within the limits specified in the COLR, and ACTION A, ACTION B B. Restore control bank insertion to within the limits specified in the COLR within two hours of exceeding the insertion limits.
- Achieve HOT STANDBY within 6 hours
C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then                                     L01 withinone hour action shall be initiated to:
- Achieve HOT SHUTDOWN within the following 6 hours  
ACTION C
: 3. Insertion limit does not apply during physics tests or during periodic exercise of individual rods. However, the shutdown margin, as specified in the COLR, must be maintained except for the Low Power Physics Test to measure control rod worth and shutdown margin. For this test, the reactor may be critical with all but one high worth  
                        - Achieve HOT STANDBY within 6 hours
 
                        - Achieve HOT SHUTDOWN within the following 6 hours See ITS 3.1.8 Applicability    3. Insertion limit does not apply during physics tests or during periodic exercise of Note                individual rods. However, the shutdown margin, as specified in the COLR, must be See ITS maintained except for the Low Power Physics Test to measure control rod worth and                                         3.1.8 shutdown margin. For this test, the reactor may be critical with all but one high worth rod inserted.
rod inserted. See ITS 3.2.4 See ITS 3.1.8 A CTION A, A CTION B A CTION C L01See ITS 3.1.8 M01Add proposed SR 3.1.6.1, SR 3.1.6.2, and SR 3.1.6.3 Applicability Note See ITS 3.1.5 , sequence, and overlap limits Add proposed Required Actions A.1.1 and B.1.1 L02 L01Add proposed ApplicabilityA02 LCO 3.1.6  Amendment No. 165 TS 3.10-4 03/11/2003 Page 1 of 1 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
Add proposed SR 3.1.6.1, SR 3.1.6.2, and SR 3.1.6.3       M01 Amendment No. 165 TS 3.10-4                                                         03/11/2003 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 128 of 213


Attachment 1, Volume 6, Rev. 0, Page 129 of 213 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS ADMINISTRATIVE CHANGES A01  In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.
3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
A02 CTS 3.10.d.2 requires, in part, that the control rod banks are to be limited in physical insertion as specified in the COLR. ITS 3.1.6 requires the control bank to be within the insertion, sequence, and overlap limits specified in the COLR. This changes the CTS by specifically stating that the control bank physical insertion limits are the control bank insertion, overlap, and sequence limits.  
A02 CTS 3.10.d.2 requires, in part, that the control rod banks are to be limited in physical insertion as specified in the COLR. ITS 3.1.6 requires the control bank to be within the insertion, sequence, and overlap limits specified in the COLR.
 
This changes the CTS by specifically stating that the control bank physical insertion limits are the control bank insertion, overlap, and sequence limits.
This change is acceptable because the LCO requirements have not changed. COLR Section 2.5 states, in part, that physical insertion limits are shown in COLR Figure 4. COLR Figure 4 is a graphical presentation of when each control bank is required to be withdrawn with respect to the Percent of Rated Thermal Power. As shown in the Figure, the sequence of control bank withdrawal is included. Additionally, the Note for COLR Figure 4 states that the Rod Bank Insertion Limits are based on a control bank tip-to-tip distance of 126 steps.
This change is acceptable because the LCO requirements have not changed.
COLR Section 2.5 states, in part, that physical insertion limits are shown in COLR Figure 4. COLR Figure 4 is a graphical presentation of when each control bank is required to be withdrawn with respect to the Percent of Rated Thermal Power. As shown in the Figure, the sequence of control bank withdrawal is included. Additionally, the Note for COLR Figure 4 states that the Rod Bank Insertion Limits are based on a control bank tip-to-tip distance of 126 steps.
Furthermore, the COLR Figure 4 is similar to the ISTS Bases Figure B 3.1.6.
Furthermore, the COLR Figure 4 is similar to the ISTS Bases Figure B 3.1.6.
Therefore, the physical insertion limits in CTS 3.10.d.2 include the insertion, sequence, and overlap limits. This change is designated as an administrative change since it does not result in any technical changes to the CTS.  
Therefore, the physical insertion limits in CTS 3.10.d.2 include the insertion, sequence, and overlap limits. This change is designated as an administrative change since it does not result in any technical changes to the CTS.
 
MORE RESTRICTIVE CHANGES M01 CTS 3.10.d does not provide any Surveillance Requirements for verifying control bank insertion limits. ITS SR 3.1.6.1 requires verifying that the estimated critical control bank position is within the insertion limits specified in the COLR within 4 hours prior to achieving criticality. ITS SR 3.1.6.2 requires verifying that each control bank insertion is within the insertion limits specified in the COLR every 12 hours. ITS SR 3.1.6.3 requires verification that the overlap and sequence limits specified in the COLR are met for control banks not fully withdrawn from the core every 12 hours. This changes the CTS by adding specific Surveillance Requirements to periodically verify that control banks will be within insertion, sequence and overlap limits specified in the COLR.
MORE RESTRICTIVE CHANGES
 
M01 CTS 3.10.d does not provide any Surveillance Requirements for verifying control bank insertion limits. ITS SR 3.1.6.1 requires verifying that the estimated critical control bank position is within the insertion limits specified in the COLR within 4 hours prior to achieving criticality. ITS SR 3.1.6.2 requires verifying that each control bank insertion is within the insertion limits specified in the COLR every 12 hours. ITS SR 3.1.6.3 requires verification that the overlap and sequence limits specified in the COLR are met for control banks not fully withdrawn from the core every 12 hours. This changes the CTS by adding specific Surveillance Requirements to periodically verify that control banks will be within insertion, sequence and overlap limits specified in the COLR.  
 
The purpose of ITS SR 3.1.6.1 is to ensure that the reactor does not achieve criticality with the control banks below their insertion limits. The purpose of ITS SR 3.1.6.2 is to detect control banks that may be approaching the insertion limits. The purpose of ITS SR 3.1.6.3 is to verify that control bank sequence and overlap is within the requirements specified in the COLR. Therefore, this change is considered acceptable and is more restrictive because new Surveillance Requirements have been added to periodically verify the LCO limits are met.
The purpose of ITS SR 3.1.6.1 is to ensure that the reactor does not achieve criticality with the control banks below their insertion limits. The purpose of ITS SR 3.1.6.2 is to detect control banks that may be approaching the insertion limits. The purpose of ITS SR 3.1.6.3 is to verify that control bank sequence and overlap is within the requirements specified in the COLR. Therefore, this change is considered acceptable and is more restrictive because new Surveillance Requirements have been added to periodically verify the LCO limits are met.
Kewaunee Power Station Page 1 of 3 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS RELOCATED SPECIFICATIONS None
Kewaunee Power Station                 Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 129 of 213
 
REMOVED DETAIL CHANGES
 
None
 
LESS RESTRICTIVE CHANGES
 
L01 (Category 2 - Relaxation of Applicability)  CTS 3.10.d.2 requires the control banks to be limited in physical insertion as specified in the COLR, but does not specifically state an applicability. CTS 3.10.d.2.A, B, and C provide actions to be taken if the control banks insertion limits are not met. If the control bank insertion limits are not restored, then the unit must be placed in HOT STANDBY (equivalent to ITS MODE 2) within 7 hours and in HOT SHUTDOWN (equivalent to ITS MODE 3) within 13 hours. Thus, the effective Applicability for the control bank insertion limits is OPERATING and HOT STANDBY MODES (equivalent to ITS MODES 1 and 2). ITS 3.1.6 requires the control bank insertion limits to be met in MODE 1 and in MODE 2 with k eff  1.0. This changes the CTS by only requiring the control bank limits to met in MODE 2 with k eff  1.0, in lieu of at all times in MODE 2. Consistent with this change in Applicability, the requirements in CTS 3.10.d.2.C to be in HOT STANDBY and HOT SHUTDOWN are changed
 
to be in MODE 2 with k eff < 1.0 within 6 hours.
The control bank insertion limits are required to prevent core power distributions that could result in fuel cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RPS trip function. This change is acceptable because in MODE 2 with k eff < 1.0, neither the assumed power distribution nor the ejected rod worth would be exceeded. Furthermore, allowing 6 hours to be in MODE 2 with k eff < 1.0 is a reasonable time, based on operating experience, for reaching the required MODE from full power in an orderly manner and without challenging plant systems. This proposed ACTION also places the unit outside the new Applicability. This change is designated as less restrictive because the Applicability has been changed to MODE 1 and MODE 2 with k eff  1.0.
L02 (Category 4 - Relaxation of Required Action)  CTS 3.10.d.2.A requires that if the control banks physical insertion limits (i.e., insertion, sequence, or overlap) are not met, then within 1 hour initiate boration to restore control bank insertion to within limits. ITS 3.1.6 ACTION A and ACTION B contain the same requirements but allows a choice between verifying that the SDM is within the limits specified in the COLR (ITS 3.1.6 Required Actions A.1.1 and B.1.1) within 1 hour or initiating boration to restore SDM to within limits within 1 hour (ITS 3.1.6 Required Actions A.1.2 and B.1.2). This changes the CTS by allowing either the verification that SDM is within the limits specified in the COLR or the initiation of boration to restore the required SDM within one hour when the control banks are not within the physical insertion limits.
Kewaunee Power Station Page 2 of 3 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS Kewaunee Power Station Page 3 of 3  This change is acceptable because it requires verification that the initial conditions of the analysis is maintained. In MODE 1 and MODE 2 with k eff > 1.0, SDM is normally ensured by adhering to the control and shutdown bank insertion limits. If the control banks are not within their insertion limits, then SDM must be verified to be within limits or boration must be initiated to restore SDM to within limits. This change is designated as less restrictive because an alternate allowance to verify SDM is within the limits specified in the COLR is being added to the CTS.
 
Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Control Bank Insertion Limits 3.1.6  WOG STS 3.1.6-1 Rev. 3.0, 03/31/04  CTS 3.1  REACTIVITY CONTROL SYSTEMS
 
====3.1.6 Control====
Bank Insertion Limits
 
LCO  3.1.6 Control banks shall be within the insertion, sequence, and overlap limits specified in the COLR.
 
APPLICABILITY: MODE 1,    MODE 2 with k eff 1.0.    -------------------------------------------NOTE----------------------------------------------
This LCO is not applicable while performing SR 3.1.4.2.
  --------------------------------------------------------------------------------------------------
 
ACTIONS  CONDITION REQUIRED ACTION COMPLETION TIME
 
A. Control bank insertion limits not met.
 
A.1.1 Verify SDM is within the limits specified in the COLR.
OR  A.1.2 Initiate boration to restore SDM to within limit.
AND A.2 Restore control bank(s) to within limits.
 
1 hour 
 
1 hour 
 
2 hours B. Control bank sequence or overlap limits not met.
 
B.1.1 Verify SDM is within the limits specified in the COLR.        OR  B.1.2 Initiate boration to restore SDM to within limit.
AND  1 hour


1 hour 3.10.d.3 3.10.d.2.A, 3.10.d.2.B s s 3.10.d.2.A, 3.10.d.2.B DOC L01 3.10.d.2 1 1 Control Bank Insertion Limits 3.1.6   WOG STS 3.1.6-2 Rev. 3.0, 03/31/04  CTS ACTIONS  (continued)
Attachment 1, Volume 6, Rev. 0, Page 130 of 213 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01  (Category 2 - Relaxation of Applicability) CTS 3.10.d.2 requires the control banks to be limited in physical insertion as specified in the COLR, but does not specifically state an applicability. CTS 3.10.d.2.A, B, and C provide actions to be taken if the control banks insertion limits are not met. If the control bank insertion limits are not restored, then the unit must be placed in HOT STANDBY (equivalent to ITS MODE 2) within 7 hours and in HOT SHUTDOWN (equivalent to ITS MODE 3) within 13 hours. Thus, the effective Applicability for the control bank insertion limits is OPERATING and HOT STANDBY MODES (equivalent to ITS MODES 1 and 2). ITS 3.1.6 requires the control bank insertion limits to be met in MODE 1 and in MODE 2 with keff  1.0. This changes the CTS by only requiring the control bank limits to met in MODE 2 with keff  1.0, in lieu of at all times in MODE 2. Consistent with this change in Applicability, the requirements in CTS 3.10.d.2.C to be in HOT STANDBY and HOT SHUTDOWN are changed to be in MODE 2 with keff < 1.0 within 6 hours.
CONDITION REQUIRED ACTION COMPLETION TIME
The control bank insertion limits are required to prevent core power distributions that could result in fuel cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RPS trip function. This change is acceptable because in MODE 2 with keff < 1.0, neither the assumed power distribution nor the ejected rod worth would be exceeded. Furthermore, allowing 6 hours to be in MODE 2 with keff < 1.0 is a reasonable time, based on operating experience, for reaching the required MODE from full power in an orderly manner and without challenging plant systems. This proposed ACTION also places the unit outside the new Applicability. This change is designated as less restrictive because the Applicability has been changed to MODE 1 and MODE 2 with keff  1.0.
L02  (Category 4 - Relaxation of Required Action) CTS 3.10.d.2.A requires that if the control banks physical insertion limits (i.e., insertion, sequence, or overlap) are not met, then within 1 hour initiate boration to restore control bank insertion to within limits. ITS 3.1.6 ACTION A and ACTION B contain the same requirements but allows a choice between verifying that the SDM is within the limits specified in the COLR (ITS 3.1.6 Required Actions A.1.1 and B.1.1) within 1 hour or initiating boration to restore SDM to within limits within 1 hour (ITS 3.1.6 Required Actions A.1.2 and B.1.2). This changes the CTS by allowing either the verification that SDM is within the limits specified in the COLR or the initiation of boration to restore the required SDM within one hour when the control banks are not within the physical insertion limits.
Kewaunee Power Station                  Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 130 of 213


B.2 Restore control bank sequence and overlap to within limits.  
Attachment 1, Volume 6, Rev. 0, Page 131 of 213 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS This change is acceptable because it requires verification that the initial conditions of the analysis is maintained. In MODE 1 and MODE 2 with keff > 1.0, SDM is normally ensured by adhering to the control and shutdown bank insertion limits. If the control banks are not within their insertion limits, then SDM must be verified to be within limits or boration must be initiated to restore SDM to within limits. This change is designated as less restrictive because an alternate allowance to verify SDM is within the limits specified in the COLR is being added to the CTS.
Kewaunee Power Station                Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 131 of 213


2 hours  C. Required Action and associated Completion Time not met.
Attachment 1, Volume 6, Rev. 0, Page 132 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
C.1 Be in MODE 2 with k eff < 1.0.
Attachment 1, Volume 6, Rev. 0, Page 132 of 213
6 hours 3.10.d.2.A, 3.10.d.2.B 3.10.d.2.C


SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
Attachment 1, Volume 6, Rev. 0, Page 133 of 213 CTS                                                                                            Control Bank Insertion Limits 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6        Control Bank Insertion Limits 3.10.d.2    LCO 3.1.6              Control banks shall be within the insertion, sequence, and overlap limits specified in the COLR.
DOC L01    APPLICABILITY:          MODE 1, MODE 2 with keff 1.0.
                                    -------------------------------------------NOTE----------------------------------------------
3.10.d.3                            This LCO is not applicable while performing SR 3.1.4.2.
ACTIONS CONDITION                            REQUIRED ACTION                            COMPLETION TIME 3.10.d.2.A, 3.10.d.2.B A. Control bank insertion            A.1.1      Verify SDM is within the                1 hour limits not met.                            limits specified in the COLR.
OR A.1.2      Initiate boration to restore            1 hour SDM to within limit.                                                    1 s
AND A.2        Restore control bank(s) to              2 hours within limits.
3.10.d.2.A,  B. Control bank sequence              B.1.1      Verify SDM is within the                1 hour 3.10.d.2.B or overlap limits not met.                  limits specified in the COLR.
OR B.1.2      Initiate boration to restore            1 hour SDM to within limit.                                                    1 s
AND WOG STS                                              3.1.6-1                                    Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 133 of 213


SR 3.1.6.1 Verify estimated critical control bank position is within the limits specified in the COLR.  
Attachment 1, Volume 6, Rev. 0, Page 134 of 213 CTS                                                                            Control Bank Insertion Limits 3.1.6 ACTIONS (continued)
CONDITION                          REQUIRED ACTION              COMPLETION TIME 3.10.d.2.A, 3.10.d.2.B                                  B.2        Restore control bank        2 hours sequence and overlap to within limits.
3.10.d.2.C  C. Required Action and        C.1        Be in MODE 2 with keff      6 hours associated Completion                  < 1.0.
Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY DOC M02 SR 3.1.6.1       Verify estimated critical control bank position is         Within 4 hours within the limits specified in the COLR.                   prior to achieving    2 insertion                                criticality DOC M02      SR 3.1.6.2      Verify each control bank insertion is within the          12 hours insertion limits specified in the COLR.
DOC M02      SR 3.1.6.3      Verify sequence and overlap limits specified in the        12 hours COLR are met for control banks not fully withdrawn from the core.
WOG STS                                        3.1.6-2                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 134 of 213


Within 4 hours prior to achieving criticality
Attachment 1, Volume 6, Rev. 0, Page 135 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6, CONTROL BANK INSERTION LIMITS
: 1. Typographical/grammatical error corrected to be consistent with the wording in the proceeding Required Action.
: 2. SR 3.1.6.1 is clarified to state that the estimated critical control bank position must be verified to be within the "insertion limits," specified in the COLR. This Specification requires three limits to be met; insertion, sequence, and overlap.
As stated in the ISTS SR 3.1.6.1 Bases, the limit to be checked is only the insertion limit.
Kewaunee Power Station                Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 135 of 213


SR  3.1.6.2 Verify each control bank insertion is within the insertion limits specified in the COLR.
Attachment 1, Volume 6, Rev. 0, Page 136 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 136 of 213


12 hours SR  3.1.6.3 Verify sequence and overlap limits specified in the COLR are met for control banks not fully withdrawn
Attachment 1, Volume 6, Rev. 0, Page 137 of 213 Control Bank Insertion Limits B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Control Bank Insertion Limits BASES BACKGROUND               The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available SDM, and initial reactivity insertion rate.
 
The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design,"                 1 INSERT 1        GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.
from the core.
The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of If two or more RCCAs that are electrically paralleled to step simultaneously.
 
A bank of RCCAs consists of two groups that are moved in a staggered               6
12 hours DOC M02 2insertion DOC M02 DOC M02 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6, CONTROL BANK INSERTION LIMITS
            , the groups fashion, but always within one step of each other. All plants have four Kewaunee Power 2 control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Station Group Alignment Limits," for control and shutdown rod OPERABILITY (KPS) has and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements.
: 1. Typographical/grammatical error corrected to be consistent with the wording in the proceeding Required Action.
The control bank insertion limits are specified in the COLR. An example           3 is provided for information only in Figure B 3.1.6-1. The control banks are required to be at or above the insertion limit lines.
: 2. SR 3.1.6.1 is clarified to state that the estimated critical control bank position must be verified to be within the "insertion limits," specified in the COLR. This Specification requires three limits to be met; insertion, sequence, and overlap. As stated in the ISTS SR 3.1.6.1 Bases, the limit to be checked is only the insertion limit.
INSERT 2      Figure B 3.1.6-1 also indicates how the control banks are moved in an             4 overlap pattern. Overlap is the distance travelled together by two control banks. The predetermined position of control bank C, at which control 3
Kewaunee Power Station Page 1 of 1 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
bank D will begin to move with bank C on a withdrawal, will be at 118 steps for a fully withdrawn position of 231 steps. The fully withdrawn position is defined in the COLR.
Control Bank Insertion Limits B 3.1.6   WOG STS B 3.1.6-1 Rev. 3.0, 03/31/04 B 3.1  REACTIVITY CONTROL SYSTEMS  
WOG STS                                        B 3.1.6-1                           Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 137 of 213
 
B 3.1.6 Control Bank Insertion Limits  
 
BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available SDM, and initial reactivity insertion rate.  
 
The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved. INSERT 1 1 The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. A bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one step of each other. All plants have four control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Group Alignment Limits," for control and shutdown rod OPERABILITY and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements. If , the groups 6KewauneePower Station (KPS) has 2 3The control bank insertion limits are specified in the COLR. An example is provided for information only in Figure B 3.1.6-1. The control banks are required to be at or above the insertion limit lines.
4Figure B 3.1.6-1 also indicates how the control banks are moved in an overlap pattern. Overlap is the distance travelled together by two control banks. The predetermined position of control bank C, at which control bank D will begin to move with bank C on a withdrawal, will be at 118 steps for a fully withdrawn position of 231 steps. The fully withdrawn position is defined in the COLR. INSERT 2 3 B 3.1.6 1 INSERT 1   USAR GDC 6, "Reactor Core Design," GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot S hutdown Capability," GDC 29, "Reactivity Shutdown Capability," and GDC 30, "Reactivity Holddown Capability" (Ref. 1),


Attachment 1, Volume 6, Rev. 0, Page 138 of 213 B 3.1.6 1
INSERT 1 USAR GDC 6, "Reactor Core Design," GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot Shutdown Capability," GDC 29, "Reactivity Shutdown Capability," and GDC 30, "Reactivity Holddown Capability" (Ref. 1),
4 INSERT 2 The control bank sequence and overlap limits are specified in the COLR.
4 INSERT 2 The control bank sequence and overlap limits are specified in the COLR.
Sequencing is the order in which the banks are moved.  
Sequencing is the order in which the banks are moved.
 
Insert Page B 3.1.6-1 Attachment 1, Volume 6, Rev. 0, Page 138 of 213
Insert Page B 3.1.6-1 Control Bank Insertion Limits B 3.1.6   WOG STS B 3.1.6-2 Rev. 3.0, 03/31/04  BASES BACKGROUND  (continued)
 
The control banks are used for precise reactivity control of the reactor. The positions of the control banks are normally controlled automatically by the Rod Control System, but can also be manually controlled. They are capable of adding reactivity very quickly (compared to borating or diluting).
 
The power density at any point in the core must be limited, so that the fuel design criteria are maintained. Together, LCO 3.1.4, LCO 3.1.5, "Shutdown Bank Insertion Limits," LCO 3.1.6, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.
 
The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained.


Attachment 1, Volume 6, Rev. 0, Page 139 of 213 Control Bank Insertion Limits B 3.1.6 BASES BACKGROUND (continued)
The control banks are used for precise reactivity control of the reactor.
The positions of the control banks are normally controlled automatically by the Rod Control System, but can also be manually controlled. They are capable of adding reactivity very quickly (compared to borating or diluting).
The power density at any point in the core must be limited, so that the fuel design criteria are maintained. Together, LCO 3.1.4, LCO 3.1.5, "Shutdown Bank Insertion Limits," LCO 3.1.6, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.
The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained.
Operation within the subject LCO limits will prevent fuel cladding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), loss of flow, ejected rod, or other accident requiring termination by a Reactor Trip System (RTS) trip function.
Operation within the subject LCO limits will prevent fuel cladding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), loss of flow, ejected rod, or other accident requiring termination by a Reactor Trip System (RTS) trip function.
APPLICABLE The shutdown and control bank insertion limits, AFD, and QPTR LCOs SAFETY are required to prevent power distributions that could result in fuel ANALYSES cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RTS trip function.
APPLICABLE     The shutdown and control bank insertion limits, AFD, and QPTR LCOs SAFETY         are required to prevent power distributions that could result in fuel ANALYSES       cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RTS trip function.
The acceptance criteria for addressing shutdown and control bank insertion limits and inoperability or misalignment are that:  
The acceptance criteria for addressing shutdown and control bank insertion limits and inoperability or misalignment are that:
: a. There be no violations of:
: a. There be no violations of:
5  1. Specified acceptable fuel design limits or  
: 1. Specified acceptable fuel design limits or                         5
: 2. Reactor Coolant System pressure boundary integrity and  
: 2. Reactor Coolant System pressure boundary integrity and             5
: b. The core remains subcritical after accident transients.  
: b. The core remains subcritical after accident transients.
;; 5 Control Bank Insertion Limits B 3.1.6    WOG STS B 3.1.6-3 Rev. 3.0, 03/31/04 BASES APPLICABLE SAFETY ANALYSES  (continued)
WOG STS                               B 3.1.6-2                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 139 of 213
 
As such, the shutdown and control bank insertion limits affect safety analysis involving core reactivity and power distributions (Ref. 3).
 
The SDM requirement is ensured by limiting the control and shutdown bank insertion limits so that allowable inserted worth of the RCCAs is such that sufficient reactivity is available in the rods to shut down the reactor to hot zero power with a reactivity margin that assumes the maximum worth RCCA remains fully withdrawn upon trip (Ref. 4).
 
Operation at the insertion limits or AFD limits may approach the maximum allowable linear heat generation rate or peaking factor with the allowed QPTR present. Operation at the insertion limit may also indicate the maximum ejected RCCA worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected RCCA worths.


Attachment 1, Volume 6, Rev. 0, Page 140 of 213 Control Bank Insertion Limits B 3.1.6 BASES APPLICABLE SAFETY ANALYSES (continued)
As such, the shutdown and control bank insertion limits affect safety analysis involving core reactivity and power distributions (Ref. 3).
The SDM requirement is ensured by limiting the control and shutdown bank insertion limits so that allowable inserted worth of the RCCAs is such that sufficient reactivity is available in the rods to shut down the reactor to hot zero power with a reactivity margin that assumes the maximum worth RCCA remains fully withdrawn upon trip (Ref. 4).
Operation at the insertion limits or AFD limits may approach the maximum allowable linear heat generation rate or peaking factor with the allowed QPTR present. Operation at the insertion limit may also indicate the maximum ejected RCCA worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected RCCA worths.
The control and shutdown bank insertion limits ensure that safety analyses assumptions for SDM, ejected rod worth, and power distribution peaking factors are preserved (Ref. 5).
The control and shutdown bank insertion limits ensure that safety analyses assumptions for SDM, ejected rod worth, and power distribution peaking factors are preserved (Ref. 5).
The insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii), in that they are initial conditions assumed in the safety analysis.
control bank The insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii), in that     2 they are initial conditions assumed in the safety analysis.
control bank 2 LCO The limits on control banks sequence, overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion is available on trip. The overlap between control banks provides more uniform rates of reactivity insertion and withdrawal and is imposed to maintain acceptable power peaking during control bank motion.
LCO                       The limits on control banks sequence, overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion is available on trip. The overlap between control banks provides more uniform rates of reactivity insertion and withdrawal and is imposed to maintain acceptable power peaking during control bank motion.
APPLICABILITY The control bank sequence, overl ap, and physical insertion limits shall be maintained with the reactor in MODES 1 and 2 with k eff 1.0. These limits must be maintained, since they preserve the assumed power distribution, ejected rod worth, SDM, and reactivity rate insertion assumptions. Applicability in MODES 3, 4, and 5 is not required, since neither the power distribution nor ejected rod worth assumptions would be  
APPLICABILITY             The control bank sequence, overlap, and physical insertion limits shall be maintained with the reactor in MODES 1 and 2 with keff 1.0. These limits must be maintained, since they preserve the assumed power in MODE 2 with      distribution, ejected rod worth, SDM, and reactivity rate insertion keff < 1.0 and                                                                                    2 assumptions. Applicability in MODES 3, 4, and 5 is not required, since neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES.
The applicability requirements have been modified by a Note indicating the LCO requirements are suspended during the performance of SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the control bank to move below the LCO limits, which would violate the LCO.
WOG STS                                            B 3.1.6-3                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 140 of 213


exceeded in these MODES. in MODE 2 with k eff < 1.0 and 2 The applicability requirements have been modified by a Note indicating the LCO requirements are suspended during the performance of SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the control bank to move below the LCO limits, which would violate the LCO.
Attachment 1, Volume 6, Rev. 0, Page 141 of 213 Control Bank Insertion Limits B 3.1.6 BASES ACTIONS     A.1.1, A.1.2, A.2, B.1.1, B.1.2, and B.2 When the control banks are outside the acceptable insertion limits, they must be restored to within those limits. This restoration can occur in two ways:
Control Bank Insertion Limits B 3.1.6   WOG STS B 3.1.6-4 Rev. 3.0, 03/31/04  BASES ACTIONS A.1.1, A.1.2, A.2, B.1.1, B.1.2, and B.2 When the control banks are outside the acceptable insertion limits, they must be restored to within those limits. This restoration can occur in two ways: 5 a. Reducing power to be consistent with rod position or  
: a. Reducing power to be consistent with rod position or                                     5
;  b. Moving rods to be consistent with power.
: b. Moving rods to be consistent with power.
Also, verification of SDM or initiation of boration to regain SDM is required within 1 hour, since the SDM in MODES 1 and 2 normally ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") has been upset. If control banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1.
Also, verification of SDM or initiation of boration to regain SDM is required within 1 hour, since the SDM in MODES 1 and 2 normally ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") has been upset. If control banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed                   7 in the BASES for SR 3.1.1.1.
7 Similarly, if the control banks are found to be out of sequence or in the wrong overlap configuration, they must be restored to meet the limits.  
Similarly, if the control banks are found to be out of sequence or in the wrong overlap configuration, they must be restored to meet the limits.
Operation beyond the LCO limits is allowed for a short time period in order to take conservative action because the simultaneous occurrence of either a LOCA, loss of flow accident, ejected rod accident, or other accident during this short time period, together with an inadequate power distribution or reactivity capability, has an acceptably low probability.
The allowed Completion Time of 2 hours for restoring the banks to within the insertion, sequence, and overlaps limits provides an acceptable time                      7 for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.
any Required Action and associated Completion Time is not met C.1 If Required Actions A.1 and A.2, or B.1 and B.2 cannot be completed                            8 within the associated Completion Times, the plant must be brought to MODE 2 with keff < 1.0, where the LCO is not applicable. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
WOG STS                            B 3.1.6-4                                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 141 of 213


Operation beyond the LCO limits is allowed for a short time period in order to take conservative action because the simultaneous occurrence of either a LOCA, loss of flow accident, ejected rod accident, or other accident during this short time period, together with an inadequate power distribution or reactivity capability, has an acceptably low probability.
Attachment 1, Volume 6, Rev. 0, Page 142 of 213 Control Bank Insertion Limits B 3.1.6 BASES SURVEILLANCE   SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.
 
The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. If the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Performing the ECP calculation within 4 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.
The allowed Completion Time of 2 hours for restoring the banks to within the insertion, sequence, and overlaps limits provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.
SR 3.1.6.2 Verification of the control bank insertion limits at a Frequency of 12 hours is sufficient to detect control banks that may be approaching the insertion limits since, normally, very little rod motion occurs in 12 hours.
7  any Required Action and associated Completion Time is not met C.1  8If Required Actions A.1 and A.2, or B.1 and B.2 cannot be completed within the associated Completion Times, the plant must be brought to MODE 2 with k eff < 1.0, where the LCO is not applicable. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
SR 3.1.6.3 When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. A Frequency of 12 hours is consistent with the insertion limit check above in SR 3.1.6.2.
Control Bank Insertion Limits B 3.1.6   WOG STS B 3.1.6-5 Rev. 3.0, 03/31/04  BASES SURVEILLANCE SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.  
REFERENCES    1. 10 CFR 50, Appendix A, GDC 10, GDC 26, GDC 28.
 
1 USAR, Sections 3.1.2.1, 3.1.2.3, 3.1.2.4, 3.1.2.5,
The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. If the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Performing the ECP calculation within 4 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.  
: 2. 10 CFR 50.46.                                        and 3.1.2.6.
 
U                  Section 14.1
SR 3.1.6.2
: 3. FSAR, Chapter [15].                                                                          2
 
: 4. FSAR, Chapter [15].
Verification of the control bank insertion limits at a Frequency of 12 hours is sufficient to detect control banks that may be approaching the insertion limits since, normally, very little rod motion occurs in 12 hours.  
2
: 5. FSAR, Chapter [15].
WOG STS                                B 3.1.6-5                                    Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 142 of 213


SR  3.1.6.3
Attachment 1, Volume 6, Rev. 0, Page 143 of 213 Control Bank Insertion Limits B 3.1.6 (17,231) 231 200 (100,190)
...---        (0,191 )
U)
:z: 150 o
I-en o
a...
~
~      100 co
---.J o
a:
I-
:z:
o C>
THIS FIGURE FOR ILLUSTRATION ONLY.
3 DO NOT USE FOR OPERATION.
o o                                              60                              100 PERCENT OF RTP Figure B 3.1.6 (page 1 of 1)
Control Bank Insertion vs. Percent RTP WOG STS                                    B 3.1.6-6                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 143 of 213


When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. A Frequency of 12 hours is consistent with the insertion limit check above in SR 3.1.6.2.
Attachment 1, Volume 6, Rev. 0, Page 144 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6 BASES, CONTROL BANK INSERTION LIMITS
REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, GDC 28.
: 1. ISTS 3.1.6 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.
1
Therefore, ITS 3.1.6 Bases Background has been revised to discuss the design standards used by KPS. Additionally, Bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
: 2. 10 CFR 50.46.
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 3. FSAR, Chapter [15]. USAR, Sections 3.1.2.1, 3.1.2.3, 3.1.2.4, 3.1.2.5, and 3.1.2.6. Section 14.1 U 2  4. FSAR, Chapter [15].
: 3. Since the ITS states the actual control bank insertion, sequence, and overlap limits are specified in the COLR, the example provided for information only is not needed in the Bases and has been deleted. Furthermore, since the overlap limits are specified in the COLR, the description of the actual value of the limit provided in the paragraph is also not needed and has been deleted.
: 5. FSAR, Chapter [15].
: 4. LCO 3.1.6 governs control bank insertion, sequence, and overlap limits. The Background section of ITS 3.1.6 Bases discusses insertion and overlaps, but does not discuss sequence. A discussion of control bank sequence is added for completeness.
2 Control Bank Insertion Limits B 3.1.6    WOG STS B 3.1.6-6 Rev. 3.0, 03/31/04 Figure B 3.1.6 (page 1 of 1) Control Bank Insertion vs. Percent RTP 3 231 200 ...---U) :z: 150 o I-en o a... 100 co ---.J o a: :z: o C> o o (17,231) (0,191 ) (100,190) 60 THIS FIGURE FOR ILLUSTRATION ONLY. DO NOT USE FOR OPERATION.
: 5. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
PERCENT OF RTP 100 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6 BASES, CONTROL BANK INSERTION LIMITS  
: 6. The wording has been changed to be consistent with the wording used in Bases of ISTS 3.1.4.
: 1. ISTS 3.1.6 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A. Therefore, ITS 3.1.6 Bases Background has been revised to discuss the design standards used by KPS. Additionally, Base s references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.  
: 7. Typographical/grammatical error corrected.
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.  
: 3. Since the ITS states the actual control bank insertion, sequence, and overlap limits are specified in the COLR, the example provided for information only is not needed in the Bases and has been deleted. Furthermore, since the overlap limits are specified in the COLR, the description of the actual value of the limit provided in the paragraph is also not needed and has been deleted.  
: 4. LCO 3.1.6 governs control bank insertion, sequence, and overlap limits. The Background section of ITS 3.1.6 Bases discusses insertion and overlaps, but does not discuss sequence. A discussion of control bank sequence is added for completeness.  
: 5. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.  
: 6. The wording has been changed to be consistent with the wording used in Bases of ISTS 3.1.4.  
: 7. Typographical/grammatical error corrected.  
: 8. Change made to be consistent with the Specification. The words in the ISTS Bases implies that Condition C is entered only when both Required Actions A.1 and A.2, or both Required Actions B.1 and B.2 are not met. Condition C is to be entered if any Required Action of ACTION A or B is not met. In addition, there is no Required Action A.1 or B.1.
: 8. Change made to be consistent with the Specification. The words in the ISTS Bases implies that Condition C is entered only when both Required Actions A.1 and A.2, or both Required Actions B.1 and B.2 are not met. Condition C is to be entered if any Required Action of ACTION A or B is not met. In addition, there is no Required Action A.1 or B.1.
Kewaunee Power Station Page 1 of 1 Specific No Significant Haza rds Considerations (NSHCs)
Kewaunee Power Station                 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 144 of 213
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.6, CONTROL BANK INSERTION LIMITS There are no specific NSHC discussions for this Specification.
Kewaunee Power Station Page 1 of 1 ATTACHMENT 7 ITS 3.1.7, ROD POSITION INDICATION Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
ITS 3.1.7A01ITS e. Rod Misalignment Limitations NOTE: Individual RPIs may be outside their limits for
< 1 hour following substantial rod movement This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.


Rod misalignment limitations do not apply during physics testing.
Attachment 1, Volume 6, Rev. 0, Page 145 of 213 Specific No Significant Hazards Considerations (NSHCs)
See ITS 3.1.8 See ITS 3.1.4 1. When reactor power is  85% of rating, the rod cluster control assembly shall be maintained within +/- 12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating. 2. When reactor power is < 85% but  50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is
Attachment 1, Volume 6, Rev. 0, Page 145 of 213
< 85% but  50%, the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating. 3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.
See ITS 3.1.4 f. Inoperable Rod Position Indicator Channels Add proposed ApplicabilityA02Add proposed LCO statement NOTE: Individual RPIs may be outside their limits for
< 1 hour following substantial rod movement See ITS 3.1.4 1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate entry condition is allowed for


each inoperable individual rod position indicator.)
Attachment 1, Volume 6, Rev. 0, Page 146 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.6, CONTROL BANK INSERTION LIMITS There are no specific NSHC discussions for this Specification.
A CTIONS Note A CTION A A. Verify the position of the rod cluster control indirectly by movable incore detectors each eight hours, or B. Within 8 hours reduce reactor thermal power to  50% of RATED POWER. A CTION A 2. If more than one individual rod position indicator channel per group are inoperable, then:
Kewaunee Power Station              Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 146 of 213
A. IMMEDIATELY place the control rods in manual, and B. Once per 1 hour, monitor and record RCS Tavg , and A CTION B C. Verify the position of the rod by movable incore detectors each 8 hours, and A CTION A  Amendment No. 181 TS 3.10-5 03/24/2005 Page 1 of 4 ITS 3.1.7A01ITS D. Within 24 hours restore the inoperable individual rod position indicators to OPERABLE status such that a maximum of one IRPI per group is inoperable or place the plant in HOT SHUTDOWN within the following 6 hours.
, Volume 6, Rev. 0, Page 147 of 213 ATTACHMENT 7 ITS 3.1.7, ROD POSITION INDICATION , Volume 6, Rev. 0, Page 147 of 213
such that a maximum of one IRPI per group is inoperable or place the plant in HOT SHUTDOWN within the following 6 hours.
, Volume 6, Rev. 0, Page 148 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 148 of 213
A CTION B  A CTION D 3. If one or more rods with inoperable individual rod position indicators have been moved in excess of 24 steps in one direction since the last determination of the rods position


then perform A or B below: 3. If one or more rods with inoperable individual rod position indicators have been moved in excess of 24 steps in one direction since the last determination of the rods position  
Attachment 1, Volume 6, Rev. 0, Page 149 of 213 ITS                                                      A01                                                ITS 3.1.7
: e. Rod Misalignment Limitations NOTE:  Individual RPIs may be outside their limits for < 1 hour following substantial rod movement See ITS 3.1.4 This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.
See ITS Rod misalignment limitations do not apply during physics testing.                                            3.1.8
: 1. When reactor power is  85% of rating, the rod cluster control assembly shall be maintained within +/- 12 steps from their respective banks. If a rod cluster control See ITS assembly is misaligned from its bank by more than +/- 12 steps when reactor power is                        3.1.4 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating.
: 2. When reactor power is < 85% but  50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is
                  < 85% but  50%, the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating.
: 3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.
: f. Inoperable Rod Position Indicator Channels            Add proposed LCO statement                                A02 Add proposed Applicability NOTE: Individual RPIs may be outside their limits for < 1 hour following substantial rod                          See ITS 3.1.4 movement ACTION A
: 1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate entry condition is allowed for ACTIONS            each inoperable individual rod position indicator.)
Note A. Verify the position of the rod cluster control indirectly by movable incore detectors ACTION A each eight hours, or B. Within 8 hours reduce reactor thermal power to  50% of RATED POWER.
: 2. If more than one individual rod position indicator channel per group are inoperable, then:
ACTION B          A. IMMEDIATELY place the control rods in manual, and B. Once per 1 hour, monitor and record RCS Tavg, and ACTION A          C. Verify the position of the rod by movable incore detectors each 8 hours, and Amendment No. 181 TS 3.10-5                                              03/24/2005 Page 1 of 4 Attachment 1, Volume 6, Rev. 0, Page 149 of 213


then perform A or B below:  A. Within 4 hours verify the position of the rod by movable incore detectors, or A. Within 4 hours verify the position of the rod by movable incore detectors, or B. Within 8 hours reduce reactor thermal power to  50% of RATED POWER. B. Within 8 hours reduce reactor thermal power to  50% of RATED POWER.
Attachment 1, Volume 6, Rev. 0, Page 150 of 213 ITS                                              A01                                        ITS 3.1.7 ACTION B   D. Within 24 hours restore the inoperable individual rod position indicators to OPERABLE status such that a maximum of one IRPI per group is inoperable or ACTION D        place the plant in HOT SHUTDOWN within the following 6 hours.
A CTION A 4. If one demand position indicator per bank for one or more banks is inoperable then perform either A or B below: (Note: Separate condition entry is allowed for each inoperable demand position indicator.) 4. If one demand position indicator per bank for one or more banks is inoperable then perform either A or B below: (Note: Separate condition entry is allowed for each
: 3. If one or more rods with inoperable individual rod position indicators have been moved in excess of 24 steps in one direction since the last determination of the rods position then perform A or B below:
ACTION A A. Within 4 hours verify the position of the rod by movable incore detectors, or B. Within 8 hours reduce reactor thermal power to  50% of RATED POWER.
ACTION C
: 4. If one demand position indicator per bank for one or more banks is inoperable then perform either A or B below: (Note: Separate condition entry is allowed for each ACTIONS Note inoperable demand position indicator.)
A. Each 8 hours verify,
: 1) All IRPIs for the affected banks are OPERABLE, by administrative means, and ACTION C            2) The most withdrawn rod and the least withdrawn rod of the affected bank(s) are  12 steps apart when > 85% RATED POWER or 24 steps when                A03 85% RATED POWER.
B. Within 8 hours, reduce reactor thermal power to  50% of RATED POWER.
: 5. If a rod cluster control assembly having a rod position indicator channel out of service is A04 found to be misaligned from TS 3.10.f.1.A, then TS 3.10.e will be applied.
Amendment No. 181 TS 3.10-6                                      03/24/2005 Page 2 of 4 Attachment 1, Volume 6, Rev. 0, Page 150 of 213


inoperable demand position indicator.)
Attachment 1, Volume 6, Rev. 0, Page 151 of 213 ITS                                                                                            ITS 3.1.7 A01 4.1 OPERATIONAL SAFETY REVIEW APPLICABILITY Applies to items directly related to safety limits and LIMITING CONDITIONS FOR OPERATION.
A CTIONS Note A CTION C A. Each 8 hours verify, A. Each 8 hours verify, 1) All IRPI's for the affected banks are OPERABLE, by administrative means, and 1) All IRPI's for the affected banks are OPERABLE, by administrative means, and 2) The most withdrawn rod and the least withdrawn rod of the affected bank(s) are  12 steps apart when > 85% RATED POWER or  24 steps when  85% RATED POWER. 2) The most withdrawn rod and the least withdrawn rod of the affected bank(s) are  12 steps apart when > 85% RATED POWER or  24 steps when  85% RATED POWER.
OBJECTIVE To assure that instrumentation shall be checked, tested, and calibrated, and that equipment and sampling tests shall be conducted at sufficiently frequent intervals to ensure safe operation.
B. Within 8 hours, reduce reactor thermal power to  50% of RATED POWER. B. Within 8 hours, reduce reactor thermal power to  50% of RATED POWER.
SPECIFICATION SR 3.1.7.1    a. Calibration, testing, and checking of protective instrumentation channels and testing of logic channels shall be performed as specified in Table TS 4.1-1.
A CTION C A03  5. If a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned from TS 3.10.f.1.A, then TS 3.10.e will be applied. 5. If a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned from TS 3.10.f.1.A, then TS 3.10.e will be applied. A04    Amendment No. 181 TS 3.10-6 03/24/2005 Page 2 of 4 ITS 3.1.7 A01ITS 4.1 OPERATIONAL SAFETY REVIEW APPLICABILITY Applies to items directly related to safety limits and LIMITING CONDITIONS FOR OPERATION.
: b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and        See other TS 4.1-3.                                                                                    ITS
OBJECTIVE To assure that instrumentation shall be checked, tested, and calibrated, and that equipment and sampling tests shall be conducted at sufficiently frequent intervals to  
: c. Deleted
: d. Deleted
: e. Deleted Amendment No. 119 TS 4.1-1                                      04/18/95 Page 3 of 4 Attachment 1, Volume 6, Rev. 0, Page 151 of 213


ensure safe operation.
ITS                                                                        A01                                                                  ITS 3.1.7 TABLE TS 4.1-1                                                       See ITS 3.3.5 See ITS MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS                                           3.3.1 and 3.3.2 CHANNEL DESCRIPTION             CHECK               CALIBRATE                 TEST                           REMARKS
SPECIFICATION
: 6. Pressurizer Water Level   Each shift       Each refueling cycle         Monthly
: a. Calibration, testing, and checking of protective instrumentation channels and testing of logic channels shall be performed as specified in Table TS 4.1-1.
: 7. Pressurizer Pressure     Each shift       Each refueling cycle         Monthly
SR 3.1.7.1
: 8. a. 4-KV Voltage and     Not applicable   Each refueling cycle         Monthly         Reactor protection circuits only Frequency
: b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and TS 4.1-3.  
: b. 4-KV Voltage           Not applicable   Each refueling cycle         Monthly         Safeguards buses only (Loss of Voltage)
: c. Deleted
: c. 4-KV Voltage           Not applicable   Each refueling cycle         Monthly         Safeguards buses only (Degraded Grid)
: d. Deleted
: 9. Analog Rod Position       Each shift(a,b)   Each refueling cycle         Each refueling (a) With step counters cycle          (b) Following rod motion in excess of 24 steps when computer is out of service
: e. Deleted See other ITS   Amendment No. 119 TS 4.1-1 04/18/95 Page 3 of 4 TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS Amendment No. 151 Page 2 of 7 02/12/2001 CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS 6. Pressurizer Water Level Each shift Each refueling cycle Monthly   7. Pressurizer Pressure Each shift Each refueling cycle Monthly   8. a. 4-KV Voltage and Frequency Not applicable Each refueling cycle Monthly Reactor protection circuits only b. 4-KV Voltage (Loss of Voltage)
: 10. Rod Position Bank       Each shift(a,b)   Not applicable               Each refueling (a) With analog rod position Counters                                                                cycle          (b) Following rod motion in excess of 24 steps when computer is out of service See ITS                                                                                     See ITS 3.1.4                                                                                       3.1.4 SR 3.1.7.1 SR 3.1.7.2 , Volume 6, Rev. 0, Page 152 of 213                                                                                                                                                          Attachment 1, Volume 6, Rev. 0, Page 152 of 213 Amendment No. 151 Page 2 of 7                                                          02/12/2001 Page 4 of 4
Not applicable Each refueling cycle Monthly Safeguards buses only c. 4-KV Voltage (Degraded Grid)
Not applicable Each refueling cycle Monthly Safeguards buses only 9. Analog Rod Position Each shift(a,b)
Each refueling cycle Each refueling cycle (a) With step counters (b) Following rod motion in excess of 24 steps when computer is out of service 10. Rod Position Bank Counters Each shift(a,b)
Not applicable Each refueling cycle (a) With analog rod position (b) Following rod motion in excess of 24 steps when computer is out of service SR 3.1.7.1 See ITS 3.3.5 See ITS 3.3.1 and 3.3.2 See ITS 3.1.4 See ITS 3.1.4 SR 3.1.7.2 ITS A01 ITS 3.1.7 Page 4 of 4 DISCUSSION OF CHANGES ITS 3.1.7, ROD POSITION INDICATION ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).  


Attachment 1, Volume 6, Rev. 0, Page 153 of 213 DISCUSSION OF CHANGES ITS 3.1.7, ROD POSITION INDICATION ADMINISTRATIVE CHANGES A01  In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.
3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
A02 CTS 3.10.f describes the ACTIONS to take for one inoperable individual rod position indicator channel per group, more than one inoperable individual rod position indicator channel per group, one or more rods with inoperable individual rod position indicators, and one demand position indicator bank inoperable for one or more banks. The lowest plant condition required by the CTS Actions (CTS 3.10.f.2.D) is HOT SHUTDOWN (ITS equivalent MODE 3). ITS 3.1.7 requires similar ACTIONS, but specifically states in the LCO that the Individual Rod Position Indication (IRPI) System and the Demand Position Indication System shall be OPERABLE in MODES 1 and 2. This changes the CTS by stating the specific Limiting Conditions for Operability (LCO) and the specific  
A02 CTS 3.10.f describes the ACTIONS to take for one inoperable individual rod position indicator channel per group, more than one inoperable individual rod position indicator channel per group, one or more rods with inoperable individual rod position indicators, and one demand position indicator bank inoperable for one or more banks. The lowest plant condition required by the CTS Actions (CTS 3.10.f.2.D) is HOT SHUTDOWN (ITS equivalent MODE 3). ITS 3.1.7 requires similar ACTIONS, but specifically states in the LCO that the Individual Rod Position Indication (IRPI) System and the Demand Position Indication System shall be OPERABLE in MODES 1 and 2. This changes the CTS by stating the specific Limiting Conditions for Operability (LCO) and the specific MODE of Applicability.
 
MODE of Applicability.  
 
This change is acceptable because the requirements for the IRPI System and the Demand Position Indication System have not changed. Actions will continue to be taken when the individual rod position indicators and the demand position indicator banks are inoperable. This change clearly states what is required to be OPERABLE and the MODE in which the Systems are required to be OPERABLE. Since the lowest plant condition required by the CTS Actions is HOT SHUTDOWN and the ITS Applicability is MODES 1 and 2. The proposed Applicability change is acceptable. This change is considered administrative because the technical requirements have not changed.
This change is acceptable because the requirements for the IRPI System and the Demand Position Indication System have not changed. Actions will continue to be taken when the individual rod position indicators and the demand position indicator banks are inoperable. This change clearly states what is required to be OPERABLE and the MODE in which the Systems are required to be OPERABLE. Since the lowest plant condition required by the CTS Actions is HOT SHUTDOWN and the ITS Applicability is MODES 1 and 2. The proposed Applicability change is acceptable. This change is considered administrative because the technical requirements have not changed.
A03 CTS 3.10.f.4 provides the Actions if one demand position indicator per bank for one or more banks is inoperable. CTS 3.10.f.4.A.2) requires a verification that the most withdrawn rod and the least withdrawn rod of the affected bank(s) are  
A03 CTS 3.10.f.4 provides the Actions if one demand position indicator per bank for one or more banks is inoperable. CTS 3.10.f.4.A.2) requires a verification that the most withdrawn rod and the least withdrawn rod of the affected bank(s) are
 
    < 12 steps apart when > 85% RATED POWER or < 24 steps apart when < 85%
< 12 steps apart when > 85% RATED POWER or < 24 steps apart when <
RATED POWER. Under similar conditions, ITS 3.1.7 Required Action C.1.2 requires a verification that the most withdrawn rod and the least withdrawn rod of the affected bank are within the required rod alignment limits of LCO 3.1.4, "Rod Group Alignment Limits." This changes the CTS by not specifying the actual rod alignment limits in this Action, but referencing the applicable LCO where the limits are controlled.
85% RATED POWER. Under similar conditions, ITS 3.1.7 Required Action C.1.2 requires a verification that the most withdrawn rod and the least withdrawn rod of the affected bank are within the required rod alignment limits of LCO 3.1.4, "Rod Group Alignment Limits." This changes the CTS by not specifying the actual rod alignment limits in this Action, but referencing the applicable LCO where the limits are controlled.
The purpose of this CTS Action is to ensure the rod alignment limits are maintained when a demand position indicator is inoperable. This change is acceptable since the actual limits are not changing. ITS LCO 3.1.4 includes the appropriate limits, and ITS 3.1.4 Required Action C.1.2 references the appropriate LCO that controls the limits. This change is administrative since the technical requirements are not being changed.
The purpose of this CTS Action is to ensure the rod alignment limits are maintained when a demand position indicator is inoperable. This change is acceptable since the actual limits are not changing. ITS LCO 3.1.4 includes the appropriate limits, and ITS 3.1.4 Required Action C.1.2 references the appropriate LCO that controls the limits. This change is administrative since the technical requirements are not being changed.
Kewaunee Power Station Page 1 of 2 DISCUSSION OF CHANGES ITS 3.1.7, ROD POSITION INDICATION Kewaunee Power Station Page 2 of 2 A04 CTS 3.10.f.5 states that if a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned when performing 3.10.f.1.A, then 3.10.e will be applied. ITS 3.1.7 does not contain the statement.
Kewaunee Power Station                 Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 153 of 213
This changes the CTS by not including the statement that if a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned, then 3.10.e will be applied.


This change is acceptable because CTS 3.10.f.5 is used to alert the user that a misaligned rod cluster control assembly is covered under a separate LCO requirement. It is an ITS convention to not include these types of cross references. ITS 3.1.4 provides the proper requirements for when a rod is misaligned, and a specific cross reference is not required. This change is designated as administrative as it incorporates an ITS convention with no technical change.  
Attachment 1, Volume 6, Rev. 0, Page 154 of 213 DISCUSSION OF CHANGES ITS 3.1.7, ROD POSITION INDICATION A04  CTS 3.10.f.5 states that if a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned when performing 3.10.f.1.A, then 3.10.e will be applied. ITS 3.1.7 does not contain the statement.
This changes the CTS by not including the statement that if a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned, then 3.10.e will be applied.
This change is acceptable because CTS 3.10.f.5 is used to alert the user that a misaligned rod cluster control assembly is covered under a separate LCO requirement. It is an ITS convention to not include these types of cross references. ITS 3.1.4 provides the proper requirements for when a rod is misaligned, and a specific cross reference is not required. This change is designated as administrative as it incorporates an ITS convention with no technical change.
MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station                Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 154 of 213


MORE RESTRICTIVE CHANGES None
Attachment 1, Volume 6, Rev. 0, Page 155 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 155 of 213


RELOCATED SPECIFICATIONS
Attachment 1, Volume 6, Rev. 0, Page 156 of 213 CTS                                                                                                        Rod Position Indication 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7          Rod Position Indication Individual                          I DOC        LCO 3.1.7                    The [Digital] Rod Position Indication ([D]RPI) System and the Demand                                1 A02 Position Indication System shall be OPERABLE.
DOC        APPLICABILITY:                MODES 1 and 2.
A02 ACTIONS
          ------------------------------------------------------------NOTE-----------------------------------------------------------
3.10.f.1,  Separate Condition entry is allowed for each inoperable rod position indicator and each demand 3.10.f.4 position indicator.
CONDITION                                  REQUIRED ACTION                          COMPLETION TIME I
3.10.f.1,    A. One [D]RPI per group                  A.1        Verify the position of the an          Once per 8 hours                  1 3.10.f.2.c                                                                                                                                2  4 inoperable for one or                            rods with inoperable                                    INSERT 1          3 more groups.                                    position indicators indirectly by using movable incore detectors.
OR A.2        Reduce THERMAL                          8 hours POWER to  50% RTP.
I 3.10.f.2    B. More than one [D]RPI                  B.1        Place the control rods                  Immediately                      1 per group inoperable.                            under manual control.
AND B.2        Monitor and record Reactor              Once per 1 hour Coolant System Tavg.
AND WOG STS                                                    3.1.7-1                                    Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 156 of 213


None
Attachment 1, Volume 6, Rev. 0, Page 157 of 213 3.1.7 3
INSERT 1 AND 4 hours after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position Insert Page 3.1.7-1 Attachment 1, Volume 6, Rev. 0, Page 157 of 213


REMOVED DETAIL CHANGES
Attachment 1, Volume 6, Rev. 0, Page 158 of 213 CTS                                                                                        Rod Position Indication 3.1.7 ACTIONS (continued)
CONDITION                    REQUIRED ACTION                            COMPLETION TIME 3.10.f.2 B.3    Verify the position of the                Once per 8 hours 3
rods with inoperable position indicators indirectly by using the movable incore detectors.
AND B.4    Restore inoperable position              24 hours                  3 3 indicators to OPERABLE status such that a                  I maximum of one [D]RPI per                                          1 group is inoperable.
C. One or more rods with    C.1    Verify the position of the                [4] hours inoperable position              rods with inoperable indicators have been              position indicators indirectly moved in excess of                by using movable incore 24 steps in one direction        detectors.                                                          3 since the last determination of the      OR rod's position.
C.2    Reduce THERMAL                            8 hours POWER to  50% RTP.
3.10.f.4    D. One demand position      D.1.1  Verify by administrative                  Once per 8 hours          3 I
C    indicator per bank            C  means all [D]RPIs for the                                            1 inoperable for one or            affected banks are                                                  2 more banks.                      OPERABLE.
AND D.1.2  Verify the most withdrawn                Once per 8 hours          3 C    rod and the least withdrawn rod of the affected banks                                          2 are  12 steps apart.
within the required rod alignment                          4 OR              limits of LCO 3.1.4, "Rod Group Alignment Limits."
WOG STS                                  3.1.7-2                                      Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 158 of 213


None
Attachment 1, Volume 6, Rev. 0, Page 159 of 213 CTS                                                                                          Rod Position Indication 3.1.7 ACTIONS (continued)
CONDITION                        REQUIRED ACTION                        COMPLETION TIME 3.10.f.4 3
D.2    Reduce THERMAL                          8 hours C  POWER to  50% RTP.
3 3.10.f.2.d    E. Required Action and          E.1    Be in MODE 3.                          6 hours D    associated Completion            D Time not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                              FREQUENCY Table 4.1-1 SR 3.1.7.1        Verify each [D]RPI agrees within [12] steps of the              Once prior to Item 9                          group demand position for the [full indicated range]            criticality after of rod travel.                                                  each removal of 5
Perform CHANNEL CALIBRATION of each IRPI. the reactor head INSERT 2 18 months WOG STS                                      3.1.7-3                                Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 159 of 213


LESS RESTRICTIVE CHANGES None Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 160 of 213 CTS                                                                    3.1.7 0 _INSERT 2 5
Rod Position Indication 3.1.7   WOG STS 3.1.7-1 Rev. 3.0, 03/31/04  CTS 3.1  REACTIVITY CONTROL SYSTEMS
Table    SR 3.1.7.2    Perform system functional test of the 18 months TS 4.1-1 Item 10                Demand Position Indication System.
Insert Page 3.1.7-3 Attachment 1, Volume 6, Rev. 0, Page 160 of 213


3.1.7 Rod Position Indication
Attachment 1, Volume 6, Rev. 0, Page 161 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.7, ROD POSITION INDICATION
 
: 1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
LCO  3.1.7 The [Digital] Rod Position Indication ([D]RPI) System and the Demand Position Indication System shall be OPERABLE. Individual I 1 DOC  A 02 APPLICABILITY: MODES 1 and 2.
: 2. The words in ISTS 3.1.7 Required Action A.1, ISTS 3.1.7 Required Action D.1.1 (ITS 3.1.7 Required Action C.1.1), and ISTS 3.1.7 Required Action D.1.2 (ITS 3.1.7 Required Action C.1.2) have been modified to be singular, versus plural, when referring to a rod or a bank. This has been done since the ISTS 3.1.7 ACTIONS Note allows separate Condition entry for each rod position indicator and each demand position indicator; thus the Required Action only applies to the individual rod or bank whose indicator in inoperable. Furthermore, this change was approved during the DC Cook ITS conversion.
DOC  A 02 ACTIONS
: 3. ISTS 3.1.7 ACTION C has two Required Actions that are connected with an OR.
------------------------------------------------------------NOTE----------------------------------------------------------- Separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator.
However, the stated Completion Time for these two Required Actions are different (4 hours and 8 hours, respectively). Due to the convention in the ISTS, as described in Section 1.3, the two Completion Times associated with the two Required Actions OR logical connector must be the same, since either Required Action can be chosen. Therefore, to be consistent with the format of the ISTS, ISTS 3.1.7 ACTION C has been deleted and a new, conditional Completion Time has been added to Required Action A.1. This ensures that the intent of the ISTS is maintained, in that a verification of the position of the rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position. In addition, since the unit is in both Conditions A and B when more than one rod position indicators per group are inoperable, and Required Action A.1 requires the identical position check required by Required Action B.3, there is no reason to include the position check as Required Action B.3. This is also consistent with the format of ISTS.
-------------------------------------------------------------------------------------------------------------------------------
CONDITION REQUIRED ACTION COMPLETION TIME A. One [D]RPI per group inoperable for one or more groups.
A.1 Verify the position of the rods with inoperable position indicators indirectly
 
by using movable incore detectors.
 
OR A.2 Reduce THERMAL POWER to  50% RTP.
Once per 8 hours
 
8 hours  B. More than one [D]RPI per group inoperable.
 
B.1 Place the control rods under manual control.
AND B.2 Monitor and record Reactor Coolant System Tavg.
AND  Immediately
 
Once per 1 hour an INSERT 1 I I 3.10.f.2 3.10.f.1, 3.10.f.2.c 3.10.f.1, 3.10.f.4 1 4 2 3 1 3.1.7 3 INSERT 1  AND  4 hours after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position
 
Insert Page 3.1.7-1 Rod Position Indication 3.1.7  WOG STS 3.1.7-2 Rev. 3.0, 03/31/04  CTS ACTIONS  (continued)
CONDITION REQUIRED ACTION COMPLETION TIME
 
B.3 Verify the position of the rods with inoperable position indicators indirectly by using the movable incore detectors.
 
AND B.4 Restore inoperable position indicators to OPERABLE status such that a maximum of one [D]RPI per group is inoperable.
 
Once per 8 hours
 
24 hours C. One or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of the rod's position.
 
C.1 Verify the position of the rods with inoperable position indicators indirectly
 
by using movable incore detectors.
OR  C.2 Reduce THERMAL POWER to  50% RTP.
 
[4] hours
 
8 hours D. One demand position indicator per bank inoperable for one or more banks.
 
D.1.1 Verify by administrative means all [D]RPIs for the affected banks are OPERABLE.
 
AND  D.1.2 Verify the most withdrawn rod and the least withdrawn rod of the affected banks
 
are  12 steps apart.
OR  Once per 8 hours
 
Once per 8 hours within the required rod alignment limits of LCO 3.1.4, "Rod Group Alignment Limits." I I C C 3 C 3.10.f.4 3.10.f.2 3 3 1 3 3 2 1 3 2 4 Rod Position Indication 3.1.7  WOG STS 3.1.7-3 Rev. 3.0, 03/31/04  CTS ACTIONS  (continued)
CONDITION REQUIRED ACTION COMPLETION TIME
 
D.2 Reduce THERMAL POWER to  50% RTP.
 
8 hours E. Required Action and associated Completion Time not met.
 
E.1 Be in MODE 3.
 
6 hours 3.10.f.4 3 C 3 D 3.10.f.2.d D
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
 
SR  3.1.7.1 Verify each [D]RPI agrees within [12] steps of the group demand position for the [full indicated range]
of rod travel.
Once prior to criticality after each removal of the reactor head
 
Perform CHANNEL CALIBRATION of each IRPI.
18 months INSERT 2 5 Table  4.1-1  Item 9 3.1.7 CTS 5 INSERT 2 SR  3.1.7.2 Perform system functional test of the Demand Position Indication System.
 
18 months Table  TS 4.1-1 Item 10  Insert Page 3.1.7-3 0_
JUSTIFICATION FOR DEVIATIONS ITS 3.1.7, ROD POSITION INDICATION  
: 1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.  
: 2. The words in ISTS 3.1.7 Required Action A.1, ISTS 3.1.7 Required Action D.1.1 (ITS 3.1.7 Required Action C.1.1), and ISTS 3.1.7 Required Action D.1.2 (ITS  
 
====3.1.7 Required====
Action C.1.2) have been modified to be singular, versus plural, when referring to a rod or a bank. This has been done since the ISTS 3.1.7 ACTIONS Note allows separate Condition entry for each rod position indicator and each demand position indicator; thus the Required Action only applies to the individual rod or bank whose indicator in inoperable. Furthermore, this change was approved during the DC Cook ITS conversion.  
: 3. ISTS 3.1.7 ACTION C has two Required Actions that are connected with an OR. However, the stated Completion Time for these two Required Actions are different (4 hours and 8 hours, respectively). Due to the convention in the ISTS, as described in Section 1.3, the two Completion Times associated with the two  
 
Required Actions OR logical connector must be the same, since either Required Action can be chosen. Therefore, to be consistent with the format of the ISTS, ISTS 3.1.7 ACTION C has been deleted and a new, conditional Completion Time has been added to Required Action A.1. This ensures that the intent of the ISTS is maintained, in that a verification of the position of the rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position. In addition, since the unit is in both Conditions A and B when more than one rod position indicators per group are inoperable, and Required Action A.1 requires the identical position check required by Required Action B.3, there is no reason to include the position check as Required Action B.3. This is also consistent with the format of ISTS.
Furthermore, this change was approved for the DC Cook ITS conversion.
Furthermore, this change was approved for the DC Cook ITS conversion.
Appropriate renumbering changes have also been made due to these deletions.  
Appropriate renumbering changes have also been made due to these deletions.
: 4. ISTS Required Action D.1.2 has been revised to be consistent with the current KPS licensing basis requirements. The CTS allows a different alignment criterion  
: 4. ISTS Required Action D.1.2 has been revised to be consistent with the current KPS licensing basis requirements. The CTS allows a different alignment criterion for  85% RATED POWER than it does for > 85% RATED POWER. Therefore, ITS 3.1.7 Required Action C.1.2 has been changed to met the alignment criteria specified in CTS 3.10.e (ITS 3.1.4).
: 5. The ISTS SR 3.1.7.1 requirement to verify each [D]RPI agrees within [12] steps of the group demand position for the full indicated range of rod travel prior to criticality after each removal of the reactor vessel head has been replaced by two new Surveillance Requirements. The first Surveillance is ITS SR 3.1.7.1 which requires performance of a CHANNEL CALIBRATION of each IRPI every 18 months. This Surveillance is acceptable because of the thermal drift characteristics of the IRPIs, performing a full range comparison of IRPI and demand position before criticality is not useful, as the IRPI response will change with IRPI temperature. The second Surveillance is ITS SR 3.1.7.2 which requires performance of a system functional test of the Demand Position Indication System (equivalent to CTS Rod Position Bank Counters) every 18 months. The performance of the CHANNEL CALIBRATION on the IRPIs and the system Kewaunee Power Station                Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 161 of 213


for  85% RATED POWER than it does for > 85% RATED POWER. Therefore, ITS 3.1.7 Required Action C.1.2 has been changed to met the alignment criteria specified in CTS 3.10.e (ITS 3.1.4).
Attachment 1, Volume 6, Rev. 0, Page 162 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.7, ROD POSITION INDICATION functional test of the Demand Position Indication System will align the ITS with the current licensing bases for Kewaunee Power Station.
: 5. The ISTS SR 3.1.7.1 requirement to verify each [D]RPI agrees within [12] steps of the group demand position for the full indicated range of rod travel prior to criticality after each removal of the reactor vessel head has been replaced by two new Surveillance Requirements. The first Surveillance is ITS SR 3.1.7.1 which requires performance of a CHANNEL CALIBRATION of each IRPI every 18 months. This Surveillance is acceptable because of the thermal drift characteristics of the IRPIs, performing a full range comparison of IRPI and demand position before criticality is not useful, as the IRPI response will change with IRPI temperature. The second Surveillance is ITS SR 3.1.7.2 which requires performance of a system functional test of the Demand Position Indication System (equivalent to CTS Rod Position Bank Counters) every 18 months. The performance of the CHANNEL CALIBRATION on the IRPIs and the system Kewaunee Power Station Page 1 of 2 JUSTIFICATION FOR DEVIATIONS ITS 3.1.7, ROD POSITION INDICATION Kewaunee Power Station Page 2 of 2 functional test of the Demand Position Indication System will align the ITS with the current licensing bases for Kewaunee Power Station.
Kewaunee Power Station              Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 162 of 213
Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Rod Position Indication B 3.1.7    WOG STS B 3.1.7-1 Rev. 3.0, 03/31/04 B 3.1  REACTIVITY CONTROL SYSTEMS


B 3.1.7  Rod Position Indication
Attachment 1, Volume 6, Rev. 0, Page 163 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Attachment 1, Volume 6, Rev. 0, Page 163 of 213


BASES  USAR, Section  
Attachment 1, Volume 6, Rev. 0, Page 164 of 213 Rod Position Indication B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Rod Position Indication USAR, Section 7.1 BASES BACKGROUND          According to GDC 13 (Ref. 1), instrumentation to monitor variables and INSERT 1 systems over their operating ranges during normal operation, anticipated          1 operational occurrences, and accident conditions must be OPERABLE.
LCO 3.1.7 is required to ensure OPERABILITY of the control rod position          6 indicators to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.
The OPERABILITY, including position indication, of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM. Rod position indication is required to assess OPERABILITY and misalignment.
Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group. Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.
Limits on control rod alignment and OPERABILITY have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.
Rod cluster control assemblies (RCCAs), or rods, are moved out of the core (up or withdrawn) or into the core (down or inserted) by their control rod drive mechanisms. The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control.
The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position Indication System (commonly called group step counters) and the [Digital]        3 Rod Position Indication ([D]RPI) System.                      Individual I
WOG STS                                      B 3.1.7-1                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 164 of 213


==7.1 BACKGROUND==
Attachment 1, Volume 6, Rev. 0, Page 165 of 213 B 3.1.7 o 1
According to GDC 13 (Ref. 1), instrumentation to monitor variables and systems over their operating ranges during normal operation, anticipated operational occurrences, and accident conditions must be OPERABLE. LCO 3.1.7 is required to ensure OPERABILITY of the control rod position indicators to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. INSERT 1 1 6 The OPERABILITY, including position indication, of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM. Rod position indication is required to assess OPERABILITY and misalignment.
                                    -INSERT 1 instrumentation and controls shall be provided as required to monitor and maintain within prescribed operating ranges essential reactor operating variables.
Insert Page B 3.1.7-1 Attachment 1, Volume 6, Rev. 0, Page 165 of 213


Mechanical or electrical failures may cause a control rod to become inoperable or to become misaligned from its group. Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.  
Attachment 1, Volume 6, Rev. 0, Page 166 of 213 Rod Position Indication B 3.1.7 BASES BACKGROUND (continued)
The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal INSERT 1A to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position 5/8                2 Indication System is considered highly precise (+/- 1 step or +/- e inch). If a      8 rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.
I The [D]RPI System provides a highly accurate indication of actual control           2 rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. To increase the reliability of the system, the inductive coils are INSERT 2 connected alternately to data system A or B. Thus, if one system fails,            2 the [D]RPI will go on half accuracy with an effective coil spacing of 7.5 inches, which is 12 steps. Therefore, the normal indication accuracy of the [D]RPI System is +/- 6 steps (+/- 3.75 inches), and the maximum uncertainty is +/- 12 steps (+/- 7.5 inches). With an indicated deviation of I 3
up to 24      12 steps between the group step counter and [D]RPI, the maximum                  5 deviation between actual rod position and the demand position could be 36    24 steps, or 15 inches.                                                         5  2 APPLICABLE            Control and shutdown rod position accuracy is essential during power SAFETY                operation. Power peaking, ejected rod worth, or SDM limits may be ANALYSES              violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM (LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits"). The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, "Rod Group Alignment Limits"). Control rod positions are continuously monitored to provide operators with information that ensures the plant is operating within the bounds of the accident analysis assumptions.
ion      ies The control rod position indicator channels satisfy Criterion 2 of                  6 10 CFR 50.36(c)(2)(ii). The control rod position indicators monitor control rod position, which is an initial condition of the accident.
WOG STS                                      B 3.1.7-2                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 166 of 213


Limits on control rod alignment and OPERABILITY have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.  
Attachment 1, Volume 6, Rev. 0, Page 167 of 213 B 3.1.7 2
INSERT 1A The readout of the Bank Demand Position Indication System is in the form of an add and subtract counter reading. There is only one display for each rod group.
2 INSERT 2 an analog signal that is proportional to the actual control rod position. This is accomplished by using an electrical coil stack linear variable differential transformer placed above the stepping mechanism of the control rod magnetic jacks which are external to the pressure housing. When the control rod is at the bottom of the core, the magnetic coupling between the primary and secondary windings is small which induces a small voltage in the secondary. As the control rod raises, the lift rod causes an increase in magnetic coupling which in turn provides an analog signal proportional to the control rod position. The individual rod position indicator channels are sufficiently accurate to detect a rod + 12 steps away from its demand position.
Insert Page B 3.1.7-2 Attachment 1, Volume 6, Rev. 0, Page 167 of 213


Rod cluster control assemblies (RCCAs), or rods, are moved out of the core (up or withdrawn) or into the core (down or inserted) by their control rod drive mechanisms. The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control.
Attachment 1, Volume 6, Rev. 0, Page 168 of 213 Rod Position Indication B 3.1.7 BASES                                   the         I 3
 
LCO             LCO 3.1.7 specifies that one [D]RPI System and one Bank Demand 6
The axial position of shutdown rods and control rods are determined by two separate and independent systems:  the Bank Demand Position Indication System (commonly called group step counters) and the [Digital] Rod Position Indication ([D]RPI) System.
Position Indication System be OPERABLE for each control rod. For the control rod position indicators to be OPERABLE requires meeting the SR           6 of the LCO and the following:
3Individual I
: a. The [D]RPI System indicates within 12 steps of the group step 7
B 3.1.7 1 INSERT 1  instrumentation and controls shall be provided as required to monitor and maintain within prescribed operating ranges essential reactor operating variables.
counter demand position as required by LCO 3.1.4, "Rod Group Alignment Limits,"
 
I a  b. For the [D]RPI System there are no failed coils, and                     2 3   4 b c. The Bank Demand Indication System has been calibrated either in the fully inserted position or to the [D]RPI System.                       3 I
Insert Page B 3.1.7-1 o -
The 12 step agreement limit between the Bank Demand Position Indication System and the [D]RPI System indicates that the Bank Demand Position Indication System is adequately calibrated, and can be used for indication of the measurement of control rod bank position.
Rod Position Indication B 3.1.7    WOG STS B 3.1.7-2 Rev. 3.0, 03/31/04  BASES BACKGROUND  (continued)
7 A deviation of less than the allowable limit, given in LCO 3.1.4, in position indication for a single control rod, ensures high confidence that the position uncertainty of the corresponding control rod group is within the assumed values used in the analysis (that specified control rod group insertion limits).
 
These requirements ensure that control rod position indication during           6 power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged.
The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or +/-
rod 6
e inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod. INSERT 1A 2 The [D]RPI System provides a highly accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one system fails, the [D]RPI will go on half accuracy with an effective coil spacing of 7.5 inches, which is 12 steps. Therefore, the normal indication accuracy of the [D]RPI System is +/- 6 steps (+/- 3.75 inches), and the maximum uncertainty is +/- 12 steps (+/- 7.5 inches). With an indicated deviation of 12 steps between the group step counter and [D]RPI, the maximum deviation between actual rod position and the demand position could be 24 steps, or 15 inches.
OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned control rods can be detected. Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits.
 
I APPLICABILITY   The requirements on the [D]RPI and step counters are only applicable in         3 MODES 1 and 2 (consistent with LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6),
APPLICABLE Control and shutdown rod positi on accuracy is essential during power SAFETY  operation. Power peaking, ejected rod worth, or SDM limits may be ANALYSES violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM (LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits"). The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, "Rod Group Alignment Limits"). Control rod positions are continuously monitored to provide operators with information that ensures the plant is operating within the bounds of the accident
 
analysis assumptions.
The control rod position indicator channels satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). The control rod position indicators monitor control rod position, which is an initial condition of the accident. INSERT 2 2 2 I ion ies 3 2 6 I up to 24 36 5 5 5/8 8 B 3.1.7 2INSERT 1A  The readout of the Bank Demand Position Indication System is in the form of an add and subtract counter reading. There is only one display for each rod group.
 
2 INSERT 2 
 
an analog signal that is proportional to the actual control rod position. This is accomplished by using an electrical coil stack linear variable differential
 
transformer placed above the stepping mec hanism of the control rod magnetic jacks which are external to the pressure housing. When the control rod is at the bottom of the core, the magnetic coupling between the primary and secondary windings is small which induces a small voltage in the secondary. As the control rod raises, the lift rod causes an increase in magnetic coupling which in turn provides an analog signal proportional to the control rod position. The individual rod position indicator channels are sufficiently accurate to detect a rod + 12 steps away from its demand position.
Insert Page B 3.1.7-2 Rod Position Indication B 3.1.7    WOG STS B 3.1.7-3 Rev. 3.0, 03/31/04  BASES LCO LCO 3.1.7 specifies that one [D]RPI System and one Bank Demand Position Indication System be OPERABLE for each control rod. For the control rod position indicators to be OPERABLE requires meeting the SR of the LCO and the following:
the I 3 6 6  a. The [D]RPI System indicates within 12 steps of the group step counter demand position as required by LCO 3.1.4, "Rod Group  
 
Alignment Limits," 7
: b. For the [D]RPI System there are no failed coils, and I 2 3 4  
: c. The Bank Demand Indication System has been calibrated either in the fully inserted position or to the [D]RPI System.
b a ; 3 I The 12 step agreement limit between the Bank Demand Position Indication System and the [D]RPI System indicates that the Bank Demand Position Indication System is adequately calibrated, and can be used for indication of the measurement of control rod bank position.  
 
A deviation of less than the allowable limit, given in LCO 3.1.4, in position indication for a single control rod, ensures high confidence that the position uncertainty of the corresponding control rod group is within the assumed values used in the analysis (that specified control rod group  
 
insertion limits).
7 6These requirements ensure that control rod position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged.  
 
OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned control rods can be detected. Therefore, power peaking, ejected rod worth, and SDM can be controlled within  
 
acceptable limits.
rod 6 APPLICABILITY The requirements on the [D]RPI and step counters are only applicable in MODES 1 and 2 (consistent with LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6),
because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of the plant. In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.
because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of the plant. In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.
I 3 Rod Position Indication B 3.1.7    WOG STS B 3.1.7-4 Rev. 3.0, 03/31/04 BASES ACTIONS  The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator.  
WOG STS                                 B 3.1.7-3                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 168 of 213


A.1 When one [D]RPI channel per group fails, the position of the rod may still be determined indirectly by use of the movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours that F Q satisfies LCO 3.2.1, satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks.
Attachment 1, Volume 6, Rev. 0, Page 169 of 213 Rod Position Indication B 3.1.7 BASES ACTIONS    The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator.
If a bank has been significantly moved, the Required Action of C.1 or C.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is  
A.1 I                        in a 3
When one [D]RPI channel per group fails, the position of the rod may still         6 be determined indirectly by use of the movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours that FQ satisfies LCO 3.2.1, F N H satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks.
5 If a bank has been significantly moved, the Required Action of C.1 or C.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.
INSERT 3  5 A.2 Reduction of THERMAL POWER to  50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 3).                                                                2 2
The allowed Completion Time of 8 hours is reasonable, based on operating experience, for reducing power to  50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above.
3 5
B.1, B.2, B.3, and B.4 in a    s I
When more than one [D]RPI per group fail, additional actions are                3  6 necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion will not occur. Together with the indirect position determination available via WOG STS                            B 3.1.7-4                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 169 of 213


small. H N Fin a I 6 3 5 5  INSERT 3 A.2  Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 3).
Attachment 1, Volume 6, Rev. 0, Page 170 of 213 B 3.1.7 5
2 The allowed Completion Time of 8 hours is reasonable, based on operating experience, for reducing power to  50% RTP from full power conditions without challenging plant systems and allowing for rod position  
INSERT 3 If a rod has been significantly moved (in excess of 24 steps in one direction, since the position was last determined), Required Action A.1 is still appropriate but must be initiated promptly to begin verifying that the rod is still properly positioned, relative to their group positions. The allowed Completion Time of 4 hours provides an acceptable period of time to verify the rod position with inoperable position indicator indirectly by using movable incore detectors. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." In this Required Action, the Completion Time only begins on discovery that both:
: a. One RPI per group inoperable for one or more groups; and
: b. A rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rods position.
If at any time during the existence of Condition A (one RPI per group inoperable for one or more groups) a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rods position, this Completion Time begins to be tracked.
Insert Page B 3.1.7-4 Attachment 1, Volume 6, Rev. 0, Page 170 of 213


determination by Required Action A.1 above.
Attachment 1, Volume 6, Rev. 0, Page 171 of 213 Rod Position Indication B 3.1.7 BASES ACTIONS (continued) movable incore detectors will minimize the potential for rod misalignment.
2
The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition.                                                    8 Monitoring and recording reactor coolant Tavg help assure that significant changes in power distribution and SDM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions.
(Required Action A.1)
The position of the rods may be determined indirectly by use of the 6
movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours that FQ satisfies LCO 3.2.1, F N H satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been 5
moved. Verification of control rod position once per 8 hours is adequate INSERT 4 for allowing continued full power operation for a limited, 24 hour period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. The 24 hour Completion Time provides sufficient time to troubleshoot and restore the I  [D]RPI system to operation while avoiding the plant challenges                        3 associated with the shutdown without full rod position indication.
Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly                    5 moved, the Required Action of C.1 or C.2 below is required.
C.1 and C.2 These Required Actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2, [or B.1, as applicable] are still appropriate but must be initiated promptly under Required Action C.1 to begin verifying              5 that these rods are still properly positioned, relative to their group positions.
If, within [4] hours, the rod positions have not been determined, THERMAL POWER must be reduced to  50% RTP within 8 hours to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps. The allowed Completion Time of [4] hours provides an acceptable period of time to verify the rod positions.
WOG STS                                            B 3.1.7-5                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 171 of 213


B.1, B.2, B.3, and B.4
Attachment 1, Volume 6, Rev. 0, Page 172 of 213 B 3.1.7 o-5 INSERT 4 and once within 4 hours after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position Insert Page B 3.1.7-5 Attachment 1, Volume 6, Rev. 0, Page 172 of 213


When more than one [D]RPI per group fail, additional actions are necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion will not occur. Together with the indirect position determination available via in a 3 I 5 s 6 3 B 3.1.7 5 INSERT 3  If a rod has been significantly moved (in excess of 24 steps in one direction, since the position was last determined), Required Action A.1 is still appropriate but must be initiated promptly to begin verifying that the rod is still properly positioned, relative to their group positions. The allowed Completion Time of 4 hours provides an acceptable period of time to verify the rod position with inoperable position indicator indirectly by using movable incore detectors. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." In this Required Action, the Completion Time only begins on discovery that both:
Attachment 1, Volume 6, Rev. 0, Page 173 of 213 Rod Position Indication B 3.1.7 BASES ACTIONS (continued)          C 5
: a. One RPI per group inoperable for one or more groups; and
D.1.1 and D.1.2 I
: b. A rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position.  
With one demand position indicator per bank inoperable, the rod positions can be determined by the [D]RPI System. Since normal power operation                    3 does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are 12 steps apart                  5 within the allowed Completion Time of once every 8 hours is adequate.
within the required rod alignment limits of LCO 3.1.4 C
D.2                                                                                      5 Reduction of THERMAL POWER to  50% RTP puts the core into a 2
condition where rod position is not significantly affecting core peaking factor limits (Ref. 3). The allowed Completion Time of 8 hours provides                 2 an acceptable period of time to verify the rod positions per Required Actions C.1.1 and C.1.2 or reduce power to  50% RTP.
D  E.1                                        and 5
any is not met If the Required Actions cannot be completed within the associated                        6 Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE          SR 3.1.7.1 REQUIREMENTS Verification that the [D]RPI agrees with the demand position within
[12] steps ensures that the [D]RPI is operating correctly. Since the INSERT 5
[D]RPI does not display the actual shutdown rod positions between 18                    5 and 210 steps, only points within the indicated ranges are required in comparison.
This Surveillance is performed prior to reactor criticality after each removal of the reactor head, as there is the potential for unnecessary plant transients if the SR were performed with the reactor at power.
WOG STS                                      B 3.1.7-6                              Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 173 of 213


If at any time during the existence of Condition A (one RPI per group inoperable for one or more groups) a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position, this Completion Time begins to be tracked.  
Attachment 1, Volume 6, Rev. 0, Page 174 of 213 B 3.1.7 5
INSERT 5 SR 3.1.7.1 is the performance of a CHANNEL CALIBRATION for each IRPI channel. The calibration verifies the accuracy of each IRPI channel.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for unnecessary plant transients if the SR were performed with the reactor at power. Operating experience has shown these components usually pass the SR when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.1.7.2 SR 3.1.7.2 is the performance of a system functional test of the Demand Position Indication System. The system functional test verifies the OPERABILITY of the Demand Position Indication System.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for unnecessary plant transients if the SR were performed with the reactor at power. Operating experience has shown these components usually pass the SR when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
Insert Page B 3.1.7-6 Attachment 1, Volume 6, Rev. 0, Page 174 of 213


Insert Page B 3.1.7-4 Rod Position Indication B 3.1.7   WOG STS B 3.1.7-5 Rev. 3.0, 03/31/04 BASES ACTIONS  (continued)
Attachment 1, Volume 6, Rev. 0, Page 175 of 213 Rod Position Indication B 3.1.7 BASES REFERENCES      1. 10 CFR 50, Appendix A, GDC 13.
U USAR, Section 7.1, General Design Criteria12,        2
: 2. FSAR, Chapter [15]. 14 "Instrumentation and Control Systems."
: 3. FSAR, Chapter [15].
WOG STS                           B 3.1.7-7                                Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 175 of 213


movable incore detectors will minimize the potential for rod misalignment. The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition.
Attachment 1, Volume 6, Rev. 0, Page 176 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.7 BASES, ROD POSITION INDICATION
8 Monitoring and recording reactor coolant Tavg help assure that significant changes in power distribution and SDM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions.  
: 1. ISTS 3.1.7 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.
Therefore, ITS 3.1.7 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 4. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
: 5. The Bases are changed to reflect changes made to the Specification.
: 6. Changes are made to reflect the actual Specification.
: 7. The requirement that the IRPI indicates within the agreement limit of the group step counter demand position has been deleted since the requirement is already covered by ITS LCO 3.1.4. If the agreement limit is not met, then the ACTIONS of LCO 3.1.4, "Rod Group Alignment Limits," should be entered. As written in these Bases, both the ACTIONS of ITS LCO 3.1.4 and ITS LCO 3.1.7 would have to be entered if not within the agreement limit. The appropriate ACTIONS are those of ITS LCO 3.1.4.
ITS LCO 3.1.7 should only cover the actual IRPI System, not the agreement limits.
: 8. Typographical error corrected.
Kewaunee Power Station                  Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 176 of 213


The position of the rods may be determined indirectly by use of the movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours that F Q satisfies LCO 3.2.1, satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. Verification of control rod position once per 8 hours is adequate for allowing continued full power operation for a limited, 24 hour period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. The 24 hour Completion Time provides sufficient time to troubleshoot and restore the
Attachment 1, Volume 6, Rev. 0, Page 177 of 213 Specific No Significant Hazards Considerations (NSHCs)
[D]RPI system to operation while avoiding the plant challenges associated with the shutdown without full rod position indication.
Attachment 1, Volume 6, Rev. 0, Page 177 of 213
H N F(Required Action A.1) 6 5 INSERT 4 Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly moved, the Required Action of C.1 or C.2 below is required.


C.1 and C.2
Attachment 1, Volume 6, Rev. 0, Page 178 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.7, ROD POSITION INDICATION There are no specific NSHC discussions for this Specification.
Kewaunee Power Station              Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 178 of 213


These Required Actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2, [or B.1, as applicable] are still appropriate but must be initiated promptly under Required Action C.1 to begin verifying that these rods are still properly positioned, relative to their group positions.  
Attachment 1, Volume 6, Rev. 0, Page 179 of 213 ATTACHMENT 8 ITS 3.1.8, PHYSICS TESTS EXCEPTIONS - MODE 2 Attachment 1, Volume 6, Rev. 0, Page 179 of 213
, Volume 6, Rev. 0, Page 180 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 180 of 213


If, within [4] hours, the rod positions have not been determined, THERMAL POWER must be reduced to 50% RTP within 8 hours to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps. The allowed Completion Time of [4] hours provides an acceptable period of time to verify the rod positions.
Attachment 1, Volume 6, Rev. 0, Page 181 of 213 ITS                                                        A01                                                          ITS 3.1.8
5 5 3 I B 3.1.7 5 INSERT 4   and once within 4 hours after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position
: f. Minimum Conditions for Criticality                          Add proposed suspension of LCO 3.4.2 requirement            L01
: 1. The reactor shall not be brought to a critical condition until the pressure-temperature                            See ITS 3.4.3 state is to the right of the criticality limit line shown in Figure TS 3.1-1.
: 2. The reactor shall be maintained subcritical by at least 1% k/k until normal water                                  See ITS 3.4.9 level is established in the pressurizer.
LCO 3.1.8 and
: 3. When the reactor is critical the moderator temperature coefficient shall be as                                  initiated in MODE 2 Applicability      specified in the COLR, except during LOW POWER PHYSICS TESTING. The maximum upper moderator temperature coefficient limit shall be  5 pcm/&deg;F for A03 power levels  60% RATED POWER and  0 pcm/&#xba;F for power levels > 60% RATED POWER.
: 4. If the limits of 3.1.f.3 cannot be met, then power operation may continue provided the following actions are taken:
See ITS 3.1.3 A. Within 24 hours, develop and maintain administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in TS 3.1.f.3. These withdrawal limits shall be in addition to the insertion limits specified in TS 3.10.d.
B. If the actions specified in TS 3.1.f.4.A are not satisfied, then be in HOT STANDBY within the next 6 hours.
Add proposed LCO 3.1.8.a                                                   L01 Add proposed LCO 3.1.8.b and LCO 3.1.8.c                                  M01 During performance of PHYSIC TESTS, the number of required channels for LCO 3.3.1, Functions 2, 3, 6, and 18.e may be reduced to 3.        A02 Add proposed ACTIONS A and B                                              M01 Add proposed ACTIONS C and D                                                L01 Add proposed SR 3.1.8.1                                                    L01 Add proposed SR 3.1.8.2 and SR 3.1.8.3                                    M01 Amendment No. 165 TS 3.1-10                                                        03/11/2003 Page 1 of 5 Attachment 1, Volume 6, Rev. 0, Page 181 of 213


Insert Page B 3.1.7-5 o -
Attachment 1, Volume 6, Rev. 0, Page 182 of 213 ITS                                                         A01                                       ITS 3.1.8 C. Quadrant Power Tilt Limits
Rod Position Indication B 3.1.7    WOG STS B 3.1.7-6 Rev. 3.0, 03/31/04 BASES ACTIONS  (continued)
D.1.1 and D.1.2
 
With one demand position indicator per bank inoperable, the rod positions can be determined by the [D]RPI System. Since normal power operation does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are  12 steps apart within the allowed Completion Time of once every 8 hours is adequate.
 
D.2  Reduction of THERMAL POWER to  50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factor limits (Ref. 3). The allowed Completion Time of 8 hours provides an acceptable period of time to verify the rod positions per Required Actions C.1.1 and C.1.2 or reduce power to  50% RTP.
 
E.1 If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours. The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.
I D and is not met 5 6 any 2 2 C C 5 3 5within the required rod alignment limits of LCO 3.1.4 5 SURVEILLANCE SR  3.1.7.1 REQUIREMENTS Verification that the [D]RPI agrees with the demand position within
[12] steps ensures that the [D]RPI is operating correctly. Since the [D]RPI does not display the actual shutdown rod positions between 18 and 210 steps, only points within the indicated ranges are required in
 
comparison. 
 
This Surveillance is performed prior to reactor criticality after each removal of the reactor head, as there is the potential for unnecessary plant transients if the SR were performed with the reactor at power.
INSERT 5 5 B 3.1.7 5 INSERT 5  SR 3.1.7.1 is the performance of a CHANNEL CALIBRATION for each IRPI channel. The calibration verifies the accuracy of each IRPI channel. 
 
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for unnecessary plant transients if the SR were performed with the reactor at power. Operating experience has shown these components usually pass the SR when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
 
SR  3.1.7.2 SR 3.1.7.2 is the performance of a system functional test of the Demand Position Indication System. The system functional test verifies the OPERABILITY of the Demand Position Indication System.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for unnecessary plant transients if the SR were performed with the reactor at power. Operating experience has shown these components usually pass the SR when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
Insert Page B 3.1.7-6 Rod Position Indication B 3.1.7    WOG STS B 3.1.7-7 Rev. 3.0, 03/31/04  BASES 2REFERENCES 1. 10 CFR 50, Appendix A, GDC 13.
: 2. FSAR, Chapter [15]. USAR, Section 7.1, General Design Criteria12, "Instrumentation and Control Systems." U 14  3. FSAR, Chapter [15].
 
JUSTIFICATION FOR DEVIATIONS ITS 3.1.7 BASES, ROD POSITION INDICATION
: 1. ISTS 3.1.7 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A. Therefore, ITS 3.1.7 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 4. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
: 5. The Bases are changed to reflect changes made to the Specification.
: 6. Changes are made to reflect the actual Specification.
: 7. The requirement that the IRPI indicates within the agreement limit of the group step counter demand position has been deleted since the requirement is already covered by ITS LCO 3.1.4. If the agreement limit is not met, then the ACTIONS of LCO 3.1.4, "Rod Group Alignment Limits," should be entered. As written in these Bases, both the ACTIONS of ITS LCO 3.1.4 and ITS LCO 3.1.7 would have to be entered if not within the agreement limit. The appropriate ACTIONS are those of ITS LCO 3.1.4. ITS LCO 3.1.7 should only cover the actual IRPI System, not the agreement limits.
: 8. Typographical error corrected.
Kewaunee Power Station Page 1 of 1 Specific No Significant Haza rds Considerations (NSHCs)
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.7, ROD POSITION INDICATION There are no specific NSHC discussions for this Specification.
Kewaunee Power Station Page 1 of 1 ATTACHMENT 8 ITS 3.1.8, PHYSICS TESTS EXCEPTIONS - MODE 2 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
A01 ITS 3.1.8 ITS f. Minimum Conditions for Criticality Add proposed suspension of LCO 3.4.2 requirement L01 1. The reactor shall not be brought to a critical condition until the pressure-temperature state is to the right of the criticality limit line shown in Figure TS 3.1-1.
: 2. The reactor shall be maintained subcritical by at least 1% k/k until normal water level is established in the pressurizer. See ITS 3.4.9 See ITS 3.4.3  3. When the reactor is critical the moderator temperature coefficient shall be as specified in the COLR, except during LOW POWER PHYSICS TESTING. The maximum upper moderator temperature coefficient limit shall be  5 pcm/ F for power levels  60% RATED POWER and  0 pcm/&#xba;F for power levels > 60% RATED POWER. 4. If the limits of 3.1.f.3 cannot be met, then power operation may continue provided the following actions are taken:
A. Within 24 hours, develop and maintain administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in TS 3.1.f.3. These withdrawal limits shall be in addition to the insertion limits specified in TS 3.10.d.
B. If the actions specified in TS 3.1.f.4.A are not satisfied, then be in HOT STANDBY within the next 6 hours. See ITS 3.1.3 A03 initiated in MODE 2 LCO 3.1.8 and A pplicabilit y  Add proposed LCO 3.1.8.a L01Add proposed LCO 3.1.8.b and LCO 3.1.8.c M01During performance of PHYSIC TESTS, the number of required channels for LCO 3.3.1, Functions 2, 3, 6, and 18.e may be reduced to 3. A02M01Add proposed ACTIONS A and B Add proposed ACTIONS C and D L01 L01Add proposed SR 3.1.8.1 M01Add proposed SR 3.1.8.2 and SR 3.1.8.3 Amendment No. 165 TS 3.1-10 03/11/2003 Page 1 of 5 A01ITS ITS 3.1.8 C. Quadrant Power Tilt Limits  
: 1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:
: 1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:
A. Eliminate the tilt.
A. Eliminate the tilt.
B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio  
B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio
> 1.0. 2. If the tilt condition is not eliminated after 24 hours, then reduce power to 50% or lower.  
                      > 1.0.
: 3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is simultaneous indication of a misaligned rod:
: 2. If the tilt condition is not eliminated after 24 hours, then reduce power to 50% or lower.
See ITS
: 3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is               3.2.4 simultaneous indication of a misaligned rod:
A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.
A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.
B. If the tilt condition is not eliminated within 12 hours, then the reactor shall be brought to a minimum load condition (30 Mwe). 4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition  
B. If the tilt condition is not eliminated within 12 hours, then the reactor shall be brought to a minimum load condition ( 30 Mwe).
 
: 4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition
(5% reactor power).
( 5% reactor power).
: d. Rod Insertion Limits  
: d. Rod Insertion Limits
: 1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when the reactor is critical or approaching criticality.  
: 1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when               See ITS the reactor is critical or approaching criticality.                                                     3.1.5
: 2. The control banks shall be limited in physical insertion; insertion limits are specified in the COLR. If any one of the control bank insertion limits is not met:
: 2. The control banks shall be limited in physical insertion; insertion limits are specified in the COLR. If any one of the control bank insertion limits is not met:
A. Within one hour, initiate boration to restore control bank insertion to within the limits specified in the COLR, and B. Restore control bank insertion to within the limits specified in the COLR within two hours of exceeding the insertion limits.
A. Within one hour, initiate boration to restore control bank insertion to within the limits specified in the COLR, and B. Restore control bank insertion to within the limits specified in the COLR within                   See ITS 3.1.6 two hours of exceeding the insertion limits.
C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then withinone hour action shall be initiated to:  
C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then withinone hour action shall be initiated to:
- Achieve HOT STANDBY within 6 hours - Achieve HOT SHUTDOWN within the following 6 hours  
                      - Achieve HOT STANDBY within 6 hours LCO 3.1.8
: 3. Insertion limit does not apply during physics tests or during periodic exercise of individual rods. However, the shutdown margin, as specified in the COLR, must be maintained except for the Low Power Physics Test to measure control rod worth and shutdown margin. For this test, the reactor may be critical with all but one high worth  
                      - Achieve HOT SHUTDOWN within the following 6 hours                                                 A03 and                                                                                            initiated in MODE 2 Applicability 3. Insertion limit does not apply during physics tests or during periodic exercise of                     See ITS 3.1.5 and individual rods. However, the shutdown margin, as specified in the COLR, must be                       3.1.6 LCO 3.1.8        maintained except for the Low Power Physics Test to measure control rod worth and and              shutdown margin. For this test, the reactor may be critical with all but one high worth Applicability rod inserted.
 
Amendment No. 165 TS 3.10-4                                       03/11/2003 Page 2 of 5 Attachment 1, Volume 6, Rev. 0, Page 182 of 213
rod inserted.
LCO 3.1.8 and A pplicabilit y LCO 3.1.8
 
and A pplicabilit y See ITS 3.2.4 See ITS 3.1.5 See ITS 3.1.6 See ITS 3.1.5 and 3.1.6 initiated in MODE 2 A03  Amendment No. 165 TS 3.10-4 03/11/2003 Page 2 of 5 A01ITS ITS 3.1.8 e. Rod Misalignment Limitations NOTE: Individual RPIs may be outside their limits for <
1 hour following substantial rod movement This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.
 
Rod misalignment limitations do not apply during physics testing.
: 1. When reactor power is  85% of rating, the rod cluster control assembly shall be maintained within +/- 12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating. 2. When reactor power is < 85% but  50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is
< 85% but  50%, the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating. 3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.
: f. Inoperable Rod Position Indicator Channels
 
NOTE: Individual RPIs may be outside their limits for <
1 hour following substantial rod movement
: 1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate entry condition is allowed for
 
each inoperable individual rod position indicator.)
A. Verify the position of the rod cluster control indirectly by movable incore detectors each eight hours, or B. Within 8 hours reduce reactor thermal power to  50% of RATED POWER. 2. If more than one individual rod position indicator channel per group are inoperable, then:
A. IMMEDIATELY place the control rods in manual, and B. Once per 1 hour, monitor and record RCS Tavg , and C. Verify the position of the rod by movable incore detectors each 8 hours, and See ITS 3.1.7 See ITS 3.1.7 See ITS 3.1.4 See ITS 3.1.4 See ITS 3.1.4 initiated in MODE 2 LCO 3.1.8 and A pplicabilit y A03  Amendment No. 181 TS 3.10-5 03/24/2005 Page 3 of 5 ITS 3.1.8 TABLE TS 3.5-2 INSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP Amendment No. 137  Page 1 of 4 06/09/98    1 2 3 4 5 6 NO. FUNCTIONAL UNIT NO. OF CHANNELS NO. OF CHANNELS TO TRIP MINIMUM OPERABLE CHANNELS MINIMUM DEGREE OF REDUNDANCY PERMISSIBLE BYPASS CONDITIONS OPERATOR ACTION IF CONDITIONS OF COLUMN 3 OR 4 CANNOT BE MET 1 Manual 2 1 1 -  Maintain HOT SHUTDOWN 2 Nuclear Flux Power Range (1)        Low setting 4  2 3 1 P-10 Maintain HOT SHUTDOWN High setting 4  2 3 1    Positive rate 4  2 3 1    Negative rate 4  2 3 1  3 Nuclear Flux 
 
Intermediate Range 2 1 1 - P-10 Maintain HOT SHUTDOWN (3) 4 Nuclear Flux
 
Source Range 2 1 1 - P-6 Maintain HOT SHUTDOWN (3) 5 Overtemperature T 4  2 3 1  Maintain HOT SHUTDOWN 6 Overpower T 4  2 3 1  Maintain HOT SHUTDOWN 7 Low Pressurizer Pressure 4  2 3 1 P-7 Maintain HOT SHUTDOWN 8 High Pressurizer Pressure 3 2 2 -  Maintain HOT SHUTDOWN
~See ITS 3.3.1 A01 LCO 3.1.8 Page 4 of 5 ITS 3.1.8 A01 TABLE TS 3.5-2 INSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP Amendment No. 195 Page 4 of 4 03/28/2008 NOTES (1)  One additional channel may be taken out of service for zero power physics testing.


(2) Deleted
Attachment 1, Volume 6, Rev. 0, Page 183 of 213 ITS                                                          A01                                          ITS 3.1.8
: e. Rod Misalignment Limitations NOTE:    Individual RPIs may be outside their limits for < 1 hour following substantial rod movement This specification defines allowable limits for misaligned rod cluster control assemblies. In See ITS TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may                3.1.4 be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or LCO 3.1.8      through the conditioning module output voltage via a correlation of rod position vs. voltage.
and            Rod misalignment limitations do not apply during physics testing.
Applicability                                                                                      initiated in MODE 2      A03
: 1. When reactor power is  85% of rating, the rod cluster control assembly shall be maintained within +/- 12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating.
: 2. When reactor power is < 85% but  50% of rating, the rod cluster control assemblies                  See ITS 3.1.4 shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is
                    < 85% but  50%, the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating.
: 3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.
See ITS
: f. Inoperable Rod Position Indicator Channels                                                                  3.1.7 NOTE: Individual RPIs may be outside their limits for < 1 hour following substantial rod                        See ITS 3.1.4 movement
: 1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate entry condition is allowed for each inoperable individual rod position indicator.)
A. Verify the position of the rod cluster control indirectly by movable incore detectors each eight hours, or B. Within 8 hours reduce reactor thermal power to  50% of RATED POWER.                              See ITS 3.1.7
: 2. If more than one individual rod position indicator channel per group are inoperable, then:
A. IMMEDIATELY place the control rods in manual, and B. Once per 1 hour, monitor and record RCS Tavg, and C. Verify the position of the rod by movable incore detectors each 8 hours, and Amendment No. 181 TS 3.10-5                                        03/24/2005 Page 3 of 5 Attachment 1, Volume 6, Rev. 0, Page 183 of 213


(3) When a block condition exists, maintain normal operation.  
ITS 3.1.8 A01 See ITS 3.3.1 TABLE TS 3.5-2 INSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP 1            2                3              4            5                  6 NO. OF          MINIMUM        MINIMUM    PERMISSIBLE    OPERATOR ACTION IF NO. OF    CHANNELS TO      OPERABLE        DEGREE OF      BYPASS      CONDITIONS OF COLUMN 3 NO.      FUNCTIONAL UNIT              CHANNELS      TRIP          CHANNELS      REDUNDANCY    CONDITIONS      OR 4 CANNOT BE MET 1    Manual                      2              1            1              -                              Maintain HOT SHUTDOWN Nuclear Flux Power 2                    LCO 3.1.8 Range(1)
Low setting              4              2            3              1                P-10          Maintain HOT SHUTDOWN High setting              4              2            3              1 Positive rate            4              2            3              1 Negative rate            4              2            3              1 3    Nuclear Flux 2              1            1              -                P-10          Maintain HOT SHUTDOWN(3)
Intermediate Range Nuclear Flux 4                                2              1            1              -                P-6            Maintain HOT SHUTDOWN(3)
Source Range 5    Overtemperature T          4              2            3              1                              Maintain HOT SHUTDOWN 6    Overpower T            4                  2            3              1                              Maintain HOT SHUTDOWN 7    Low Pressurizer Pressure 4                  2            3              1                P-7            Maintain HOT SHUTDOWN 8    High Pressurizer Pressure 3                2            2              -                              Maintain HOT SHUTDOWN , Volume 6, Rev. 0, Page 184 of 213                                                                                                                                                Attachment 1, Volume 6, Rev. 0, Page 184 of 213 Amendment No. 137 Page 1 of 4                                                06/09/98 Page 4 of 5


(4) Underfrequency on the 4-kV buses trips the Reactor Coolant Pump breakers, which in turn trips the reactor when power is ab ove P-7. Permissive/Interlock Channels Coincidence Setting Limit P-6 Intermediate Range Nuclear Instrument 1 of 2 > 10-5% RATED POWER P-7 Power Range Nuclear Instrument 3 of 4 AND 2 of 2  12.2% RATED POWER Turbine Impulse Pressure 12.2% RATED POWER (a) P-8 Power Range Nuclear Instrument 3 of 4 < 10% RATED POWER P-10 Power Range Nuclear Instrument 2 of 4 7.8% RATED POWER (a) Setting Limit is converted to an equivalent turbine impulse pressure See ITS 3.3.1 M02 Page 4 of 15 Page 5 of 5 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).  
A01                                                      ITS 3.1.8 TABLE TS 3.5-2 INSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP NOTES (1) One additional channel may be taken out of service for zero power physics testing.                                                    M02 (2) Deleted (3) When a block condition exists, maintain normal operation.
(4) Underfrequency on the 4-kV buses trips the Reactor Coolant Pump breakers, which in turn trips the reactor when power is above P-7.
Permissive/Interlock                       Channels                               Coincidence                     Setting Limit P-6                             Intermediate Range Nuclear Instrument                     1 of 2                 > 10-5% RATED POWER Power Range Nuclear Instrument                             3 of 4                 12.2% RATED POWER P-7                                                                                        AND Turbine Impulse Pressure                                   2 of 2                12.2% RATED POWER (a)
P-8                             Power Range Nuclear Instrument                           3 of 4                 < 10% RATED POWER P-10                             Power Range Nuclear Instrument                             2 of 4                 7.8% RATED POWER (a)
Setting Limit is converted to an equivalent turbine impulse pressure See ITS 3.3.1 , Volume 6, Rev. 0, Page 185 of 213                                                                                                                                                          Attachment 1, Volume 6, Rev. 0, Page 185 of 213 Amendment No. 195 Page 4 of 4                                              03/28/2008 Page 5 4 of 5 15


Attachment 1, Volume 6, Rev. 0, Page 186 of 213 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 ADMINISTRATIVE CHANGES A01  In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.
3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.
A02 CTS 3.1.f.3, CTS 3.10.d.3, and CTS 3.10.e state that the limitations of certain specifications may be suspended during the performance of PHYSICS TESTS. CTS Table TS 3.5-2, Functional Units 2, Nuclear Flux Power Range, states the minimum OPERABLE channel requirement is 3, except as modified by Note (1). ITS LCO 3.1.8 includes an allowance to reduce the required number of channels for ITS LCO 3.3.1, "RPS Instrumentation," Functions 2 (Power Range Neutron Flux), 3 (Power Range Neutron Flux Rate), 6 (Overtemperature T), and 16.e (Power Range Neutron Flux, P-10) from "4" to "3" during PHYSICS TESTS. This changes the CTS by stating the specific number of RPS channels from that have to be OPERABLE during PHYSICS TESTS.
A02 CTS 3.1.f.3, CTS 3.10.d.3, and CTS 3.10.e state that the limitations of certain specifications may be suspended during the performance of PHYSICS TESTS.
The purpose of CTS 3.1.f.3, CTS 3.10.d.3, and CTS 3.10.e is to allow some flexibility during the performance of PHYSIC TESTS. This change is acceptable because the minimum number of channels required for OPERABILITY for the Reactor Trip Functions in CTS Table TS 3.5-2 is currently "3." This allowance is needed since the Required Channels in ITS 3.3.1, Reactor Trip System Instrumentation, is "4." This change from the CTS is discussed in the Discussion of Changes for ITS 3.3.1. Furthermore, the exception identified in CTS Table TS 3.5-2 Note (1) is discussed in DOC M02. This change is designated as administrative because it does not result in technical changes to the CTS.  
CTS Table TS 3.5-2, Functional Units 2, Nuclear Flux Power Range, states the minimum OPERABLE channel requirement is 3, except as modified by Note (1).
 
ITS LCO 3.1.8 includes an allowance to reduce the required number of channels for ITS LCO 3.3.1, "RPS Instrumentation," Functions 2 (Power Range Neutron Flux), 3 (Power Range Neutron Flux Rate), 6 (Overtemperature T), and 16.e (Power Range Neutron Flux, P-10) from "4" to "3" during PHYSICS TESTS. This changes the CTS by stating the specific number of RPS channels from that have to be OPERABLE during PHYSICS TESTS.
A03 CTS 3.1.f.3 states that the Moderator Temperature Coefficient requirement specified in the COLR may be suspended during the performance of LOW POWER PHYSICS TESTING. CTS 3.10.d.3 states that the Rod Insertion Limits do not apply during the performance of PHYSICS TESTS. CTS 3.10.e states that the Rod Misalignment Limitations do not apply during the performance of PHYSICS TESTS. ITS 3.1.8 is applicable "During PHYSICS TESTS initiated in MODE 2." This changes the CTS such that the Specification is applicable in MODE 2 only when a PHYSICS TEST is initiated.  
The purpose of CTS 3.1.f.3, CTS 3.10.d.3, and CTS 3.10.e is to allow some flexibility during the performance of PHYSIC TESTS. This change is acceptable because the minimum number of channels required for OPERABILITY for the Reactor Trip Functions in CTS Table TS 3.5-2 is currently "3." This allowance is needed since the Required Channels in ITS 3.3.1, Reactor Trip System Instrumentation, is "4." This change from the CTS is discussed in the Discussion of Changes for ITS 3.3.1. Furthermore, the exception identified in CTS Table TS 3.5-2 Note (1) is discussed in DOC M02. This change is designated as administrative because it does not result in technical changes to the CTS.
 
A03 CTS 3.1.f.3 states that the Moderator Temperature Coefficient requirement specified in the COLR may be suspended during the performance of LOW POWER PHYSICS TESTING. CTS 3.10.d.3 states that the Rod Insertion Limits do not apply during the performance of PHYSICS TESTS. CTS 3.10.e states that the Rod Misalignment Limitations do not apply during the performance of PHYSICS TESTS. ITS 3.1.8 is applicable "During PHYSICS TESTS initiated in MODE 2." This changes the CTS such that the Specification is applicable in MODE 2 only when a PHYSICS TEST is initiated.
The purpose of ITS 3.1.8 Applicability is to ensure that the Actions contained in the Specification are followed. The CTS applicability does not specifically state what MODE the PHYSICS TEST is performed in. The ITS Applicability states emphatically the MODE that the PHYSICS TEST will be performed. This change is designated as administrative because it clarifies the current wording in the Specification with no change in intent.  
The purpose of ITS 3.1.8 Applicability is to ensure that the Actions contained in the Specification are followed. The CTS applicability does not specifically state what MODE the PHYSICS TEST is performed in. The ITS Applicability states emphatically the MODE that the PHYSICS TEST will be performed. This change is designated as administrative because it clarifies the current wording in the Specification with no change in intent.
 
Kewaunee Power Station               Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 186 of 213
Kewaunee Power Station Page 1 of 3 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 MORE RESTRICTIVE CHANGES M01 TS 3.1.f, CTS 3.10.d.3, and CTS 3.10.e state that the limitations of certain specifications may be suspended during the performance of PHYSICS TESTS but does not provide restrictions that must be followed when utilizing the CTS exceptions. ITS 3.1.8 provides the requirements and restrictions for performing testing during PHYSICS TESTS initiated in MODE 2. A Surveillance (SR 3.1.8.3) to verify the SHUTDOWN MARGIN is within limits specified in the COLR every 24 hours and an ACTION (ACTION A) to follow if the SHUTDOWN MARGIN is not met are added. Additionally, a Surveillance (SR 3.1.8.2) to verify that THERMAL POWER is  5% RTP every 30 minutes and an ACTION (ACTION B) to follow if the THERMAL POWER is not within limit (i.e., it is not  5% RTP) has been added. This changes the CTS by imposing additional requirements on the application of the test exception LCO.
 
This change is acceptable because it imposes reasonable restrictions on the performance of PHYSICS TESTS when the control rod and RCS minimum temperature Specifications are allowed to be violated. When the control rod and RCS minimum temperature Specifications are allowed to be violated, additional actions must be taken to ensure that sufficient SHUTDOWN MARGIN is available to shutdown the reactor and keep it subcritical if needed when in MODE 2 with k eff > 1.0. This change is designated as more restrictive because it imposes additional restrictions not found in the CTS.


M02 CTS Table TS 3.5-2, Functional Unit 2, Nuclear Flux Power Range, requires three channels to be OPERABLE. However, Note (1) allows one additional channel to be taken out of service during zero power testing. Thus, during zero power PHYSICS TESTS, the CTS only requires two Functional Unit 2 channels to be OPERABLE. ITS LCO 3.1.8, in part, includes an allowance to reduce the required number of channels for ITS LCO 3.3.1, "RPS Instrumentation,"
Attachment 1, Volume 6, Rev. 0, Page 187 of 213 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 MORE RESTRICTIVE CHANGES M01  TS 3.1.f, CTS 3.10.d.3, and CTS 3.10.e state that the limitations of certain specifications may be suspended during the performance of PHYSICS TESTS but does not provide restrictions that must be followed when utilizing the CTS exceptions. ITS 3.1.8 provides the requirements and restrictions for performing testing during PHYSICS TESTS initiated in MODE 2. A Surveillance (SR 3.1.8.3) to verify the SHUTDOWN MARGIN is within limits specified in the COLR every 24 hours and an ACTION (ACTION A) to follow if the SHUTDOWN MARGIN is not met are added. Additionally, a Surveillance (SR 3.1.8.2) to verify that THERMAL POWER is  5% RTP every 30 minutes and an ACTION (ACTION B) to follow if the THERMAL POWER is not within limit (i.e., it is not  5% RTP) has been added. This changes the CTS by imposing additional requirements on the application of the test exception LCO.
This change is acceptable because it imposes reasonable restrictions on the performance of PHYSICS TESTS when the control rod and RCS minimum temperature Specifications are allowed to be violated. When the control rod and RCS minimum temperature Specifications are allowed to be violated, additional actions must be taken to ensure that sufficient SHUTDOWN MARGIN is available to shutdown the reactor and keep it subcritical if needed when in MODE 2 with keff > 1.0. This change is designated as more restrictive because it imposes additional restrictions not found in the CTS.
M02 CTS Table TS 3.5-2, Functional Unit 2, Nuclear Flux Power Range, requires three channels to be OPERABLE. However, Note (1) allows one additional channel to be taken out of service during zero power testing. Thus, during zero power PHYSICS TESTS, the CTS only requires two Functional Unit 2 channels to be OPERABLE. ITS LCO 3.1.8, in part, includes an allowance to reduce the required number of channels for ITS LCO 3.3.1, "RPS Instrumentation,"
Functions 2 (Power Range Neutron Flux) and 3 (Power Range Neutron Flux Rate) from "4" to "3" during PHYSICS TESTS. This changes the CTS by always requiring three channels of Nuclear Flux Power Range be OPERABLE during all MODE 2 PHYSICS TESTS, even zero power PHYSICS TESTS.
Functions 2 (Power Range Neutron Flux) and 3 (Power Range Neutron Flux Rate) from "4" to "3" during PHYSICS TESTS. This changes the CTS by always requiring three channels of Nuclear Flux Power Range be OPERABLE during all MODE 2 PHYSICS TESTS, even zero power PHYSICS TESTS.
The purpose of CTS Table TS 3.5-2 Note (1) is to allow flexibility during performance of zero power PHYSICS TESTS. However, there are four installed Nuclear Flux Power Range channels, and ITS LCO 3.1.8 already allows one of them to be inoperable during PHYSICS TESTS. Thus, flexibility is being provided in LCO 3.1.8, and an additional channel being inoperable is not necessary. This change is designated as more restrictive since an additional channel is required during zero power PHYSICS TESTS.  
The purpose of CTS Table TS 3.5-2 Note (1) is to allow flexibility during performance of zero power PHYSICS TESTS. However, there are four installed Nuclear Flux Power Range channels, and ITS LCO 3.1.8 already allows one of them to be inoperable during PHYSICS TESTS. Thus, flexibility is being provided in LCO 3.1.8, and an additional channel being inoperable is not necessary. This change is designated as more restrictive since an additional channel is required during zero power PHYSICS TESTS.
 
RELOCATED SPECIFICATIONS None Kewaunee Power Station               Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 187 of 213
RELOCATED SPECIFICATIONS None  
 
Kewaunee Power Station Page 2 of 3 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 Kewaunee Power Station Page 3 of 3 REMOVED DETAIL CHANGES None 


LESS RESTRICTIVE CHANGES
Attachment 1, Volume 6, Rev. 0, Page 188 of 213 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01  (Category 1 - Relaxation of LCO Requirements) CTS 3.10.f.1 (ITS 3.4.2, " RCS Minimum Temperature for Criticality") states that the reactor shall not be brought to a critical condition until the pressure-temperature state is to the right of the criticality limit line shown in Figure TS 3.1-1. ITS 3.1.8 provides an exception to LCO 3.4.2, "RCS Minimum Temperature for Criticality," provided the RCS lowest loop average temperature is  530&#xba;F. A Surveillance Requirement (SR 3.1.8.1) to verify the RCS lowest loop average temperature is  530&#xba;F every 30 minutes has been added. In addition, ACTION C has been added to cover the situation when RCS lowest loop average temperature is not within limit. The Required Action is to restore RCS lowest loop average temperature to within limit within 15 minutes. If this is not met, then ACTION D requires the unit to be in MODE 3 within 15 minutes. This changes the CTS by allowing the suspension of LCO 3.4.2, "RCS Minimum Temperature for Criticality." However, it places a limitation on the RCS lowest loop average temperature.
This change is acceptable because the LCO requirements continue to ensure that the process variables are maintained consistent with the safety analyses and licensing basis. This changes the CTS by allowing the suspension of LCO 3.4.2, "RCS Minimum Temperature for Criticality." However, it places a limitation on the RCS lowest loop average temperature. The limit on RCS average temperature is provided in the test exception LCO to ensure that the RCS temperature stays within the analyzed range. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS.
Kewaunee Power Station                  Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 188 of 213


L01 (Category 1 - Relaxation of LCO Requirements)  CTS 3.10.f.1 (ITS 3.4.2, " RCS Minimum Temperature for Criticality") states that the reactor shall not be brought to a critical condition until the pressure-temperature state is to the right of the criticality limit line shown in Figure TS 3.1-1. ITS 3.1.8 provides an exception to LCO 3.4.2, "RCS Minimum Temperature for Criticality," provided the RCS lowest loop average temperature is  530&#xba;F. A Surveillance Requirement (SR 3.1.8.1) to verify the RCS lowest loop average temperature is  530&#xba;F every 30 minutes has been added. In addition, ACTION C has been added to cover the situation when RCS lowest loop average temperature is not within limit. The Required Action is to restore RCS lowest loop average temperature to within limit within 15 minutes. If this is not met, then ACTION D requires the unit to be in MODE 3 within 15 minutes. This changes the CTS by allowing the suspension of LCO 3.4.2, "RCS Minimum Temperature for Criticality."  However, it places a limitation on the RCS lowest loop average temperature.
Attachment 1, Volume 6, Rev. 0, Page 189 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
This change is acceptable because the LCO requirements continue to ensure that the process variables are maintained consistent with the safety analyses and licensing basis. This changes the CTS by allowing the suspension of LCO 3.4.2, "RCS Minimum Temperature for Criticality."  However, it places a limitation on the RCS lowest loop average temperature. The limit on RCS average temperature is provided in the test exception LCO to ensure that the RCS temperature stays within the analyzed range. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS.
Attachment 1, Volume 6, Rev. 0, Page 189 of 213
Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)
PHYSICS TESTS Exceptions - MODE 2 3.1.8  WOG STS 3.1.8-1 Rev. 3.0, 03/31/04  CTS 3.1  REACTIVITY CONTROL SYSTEMS


====3.1.8 PHYSICS====
Attachment 1, Volume 6, Rev. 0, Page 190 of 213 CTS                                                                PHYSICS TESTS Exceptions - MODE 2 3.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8        PHYSICS TESTS Exceptions - MODE 2 LCO 3.1.8              During the performance of PHYSICS TESTS, the requirements of:
TESTS Exceptions - MODE 2  
3.1.f                          LCO 3.1.3, "Moderator Temperature Coefficient,"
3.10.e                          LCO 3.1.4, "Rod Group Alignment Limits,"              ;
3.10.d                          LCO 3.1.5, "Shutdown Bank Insertion Limits,"                                  1 3.10.d                          LCO 3.1.6, "Control Bank Insertion Limits," and            ;
DOC L01                        LCO 3.4.2, "RCS Minimum Temperature for Criticality" DOC A02                        may be suspended and the number of required channels for LCO 3.3.1, RPS "RTS Instrumentation," Functions 2, 3, 6 and 18.e, may be reduced to 3      4 required channels, provided:                  16  530 DOC L01                        a. RCS lowest loop average temperature is  [531]&deg;F,                      2
                                                                                                ;            1 DOC M01                        b. SDM is within the limits specified in the COLR, and DOC M01                        c. THERMAL POWER is < 5% RTP.
3.1.f, 3.10.e, 3.10.d, APPLICABILITY:          During PHYSICS TESTS initiated in MODE 2.
DOC L01 ACTIONS CONDITION                      REQUIRED ACTION                  COMPLETION TIME DOC M01  A. SDM not within limit.          A.1    Initiate boration to restore    15 minutes SDM to within limit.
AND A.2    Suspend PHYSICS TESTS          1 hour exceptions.
DOC M01  B. THERMAL POWER not              B.1    Open reactor trip breakers. Immediately within limit.
DOC L01  C. RCS lowest loop                C.1    Restore RCS lowest loop        15 minutes average temperature not              average temperature to within limit.                        within limit.
WOG STS                                        3.1.8-1                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 190 of 213


LCO 3.1.8 During the performance of PHYSICS TESTS, the requirements of:
Attachment 1, Volume 6, Rev. 0, Page 191 of 213 CTS                                                            PHYSICS TESTS Exceptions - MODE 2 3.1.8 ACTIONS (continued)
CONDITION                      REQUIRED ACTION              COMPLETION TIME DOC L01 D. Required Action and        D.1    Be in MODE 3.                15 minutes associated Completion Time of Condition C not met.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                  FREQUENCY SR 3.1.8.1        Perform a CHANNEL OPERATIONAL TEST on                Prior to initiation power range and intermediate range channels per      of PHYSICS               3
[SR 3.3.1.7, SR 3.3.1.8, and Table 3.3.1-1].        TESTS DOC L01  SR 3.1.8.2        Verify the RCS lowest loop average temperature is    30 minutes              3 1      [531]&deg;F.                                                                      2 530 DOC M01  SR 3.1.8.3        Verify THERMAL POWER is < 5% RTP.                    30 minutes                3 2
DOC M01  SR 3.1.8.4        Verify SDM is within the limits specified in the    24 hours                  3 3    COLR.
WOG STS                                    3.1.8-2                      Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 191 of 213


LCO 3.1.3, "Moderator Temperature Coefficient," 1 2 1 4  LCO 3.1.4, "Rod Group Alignment Limits,"
Attachment 1, Volume 6, Rev. 0, Page 192 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2
LCO 3.1.5, "Shutdown Bank Insertion Limits,"    LCO 3.1.6, "Control Bank Insertion Limits," and    LCO 3.4.2, "RCS Minimum Temperature for Criticality"
: 1. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
 
: 2. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 18.e, may be reduced to 3 required channels, provided:
: 3. ISTS SR 3.1.8.1 requires a CHANNEL OPERATIONAL TEST be performed on the Intermediate and Power Range Channels "prior to initiation of PHYSICS TESTS." However, no finite time as to how soon prior to the PHYSICS TESTS is stated. The ITS Applicability for the Intermediate and Power Range channels includes MODE 2, thus the normal, periodic Frequencies for SR 3.3.1.7 and SR 3.3.1.8 must be met prior to entering MODE 2. Therefore, the normal periodic Frequencies already ensure the "prior to initiation of PHYSICS TEST" is met, and ISTS SR 3.1.8.1 is not necessary and has been deleted. Due to this deletion, the remaining SRs have been renumbered.
: a. RCS lowest loop average temperature is  [531]&deg;F,  b. SDM is within the limits specified in the COLR, and
: c. THERMAL POWER is < 5% RTP. 
 
APPLICABILITY: During PHYSICS TESTS initiated in MODE 2.
 
ACTIONS  CONDITION REQUIRED ACTION COMPLETION TIME
 
A. SDM not within limit.
 
A.1 Initiate boration to restore SDM to within limit.
 
AND  A.2 Suspend PHYSICS TESTS exceptions.
15 minutes
 
1 hour  B. THERMAL POWER not within limit.
 
B.1 Open reactor trip breakers.
 
Immediately
 
C. RCS lowest loop average temperature not within limit.
 
C.1 Restore RCS lowest loop average temperature to within limit.
 
15 minutes 530 16 RPS ;; ; DOC L01 DOC M01 DOC M01 3.1.f,  3.10.e,  3.10.d, DOC L01 DOC A02 DOC M01 DOC M01 DOC L01 DOC L01 3.10.d 3.10.d 3.10.e 3.1.f PHYSICS TESTS Exceptions - MODE 2 3.1.8  WOG STS 3.1.8-2 Rev. 3.0, 03/31/04  CTS ACTIONS  (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and associated Completion
 
Time of Condition C not met. D.1 Be in MODE 3.
 
15 minutes DOC L01 
 
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
 
SR  3.1.8.1 Perform a CHANNEL OPERATIONAL TEST on power range and intermediate range channels per
[SR 3.3.1.7, SR 3.3.1.8, and Table 3.3.1-1].
 
Prior to initiation of PHYSICS TESTS  SR  3.1.8.2 Verify the RCS lowest loop average temperature is  [531]&deg;F.
 
30 minutes SR  3.1.8.3 Verify THERMAL POWER is  5% RTP. 30 minutes
 
SR  3.1.8.4 Verify SDM is within the limits specified in the COLR.
24 hours 3 3 2DOC L01 2 1 530 3DOC M01 3DOC M01 3 JUSTIFICATION FOR DEVIATIONS ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2  
: 1. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.  
: 2. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.  
: 3. ISTS SR 3.1.8.1 requires a CHANNEL OPERATIONAL TEST be performed on the Intermediate and Power Range Channels "prior to initiation of PHYSICS TESTS." However, no finite time as to how soon prior to the PHYSICS TESTS is stated. The ITS Applicability for the Intermediate and Power Range channels includes MODE 2, thus the normal, periodic Frequencies for SR 3.3.1.7 and SR 3.3.1.8 must be met prior to entering MODE 2. Therefore, the normal periodic Frequencies already ensure the "prior to initiation of PHYSICS TEST" is met, and ISTS SR 3.1.8.1 is not necessary and has been deleted. Due to this deletion, the remaining SRs have been renumbered.  
: 4. Changes made to be consistent with changes made to LCO 3.3.1.
: 4. Changes made to be consistent with changes made to LCO 3.3.1.
Kewaunee Power Station Page 1 of 1 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
Kewaunee Power Station               Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 192 of 213
PHYSICS TESTS Exceptions - MODE 2 B 3.1.8    WOG STS B 3.1.8-1 Rev. 3.0, 03/31/04 B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.8  PHYSICS TESTS Exceptions - MODE 2
 
BASES  BACKGROUND The primary purpose of the MODE 2 PHYSICS TESTS exceptions is to permit relaxations of existing LCOs to allow certain PHYSICS TESTS to
 
be performed.
Section XI of 10 CFR 50, Appendix B (Ref. 1), requires that a test program be established to ensure that structures, systems, and components will perform satisfactorily in service. All functions necessary to ensure that the specified design conditions are not exceeded during normal operation and anticipated operational occurrences must be tested. This testing is an integral part of the design, construction, and operation of the plant. Requirements for notification of the NRC, for the purpose of conducting tests and experiments, are specified in 10 CFR 50.59 (Ref. 2).
 
The key objectives of a test program are to (Ref. 3):
: a. Ensure that the facility has been adequately designed, 1
: b. Validate the analytical models used in the design and analysis, ; 1  c. Verify the assumptions used to predict unit response,  d. Ensure that installation of equipment in the facility has been accomplished in accordance with the design, and
: e. Verify that the operating and emergency procedures are adequate.
; ; ;1 1 To accomplish these objectives, testing is performed prior to initial criticality, during startup, during low power operations, during power ascension, at high power, and after each refueling. The PHYSICS TESTS requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions and that the core can be operated as designed (Ref. 4).
 
PHYSICS TESTS procedures are written and approved in accordance with established formats. The procedures include all information necessary to permit a detailed execution of the testing required to ensure that the design intent is met. PHYSICS TESTS are performed in accordance with these procedures and test results are approved prior to continued power escalation and long term power operation.
 
PHYSICS TESTS Exceptions - MODE 2 B 3.1.8    WOG STS B 3.1.8-2 Rev. 3.0, 03/31/04  BASES BACKGROUND  (continued)


The PHYSICS TESTS required for reload fuel cycles (Ref. 4) in MODE 2 are listed below:
Attachment 1, Volume 6, Rev. 0, Page 193 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)
: a. Critical Boron Concentration - Control Rods Withdrawn,
Attachment 1, Volume 6, Rev. 0, Page 193 of 213
: b. Critical Boron Concentration - Control Rods Inserted,
: c. Control Rod Worth,
: d. Isothermal Temperature Coefficient (ITC), and  
: e. Neutron Flux Symmetry.
; ; ;.1 1 1 1 2Differential Boron Worth 2 2The first four tests are performed in MODE 2, and the last test can be performed in either MODE 1 or 2. These and other supplementary tests may be required to calibrate the nuclear instrumentation or to diagnose operational problems. These tests may cause the operating controls and process variables to deviate from their LCO requirements during their performance.
  [ a. The Critical Boron Concentration - Control Rods Withdrawn Test measures the critical boron concentration at hot zero power (HZP). With all rods out, the lead control bank is at or near its fully withdrawn position. HZP is where the core is critical (k eff = 1.0), and the Reactor Coolant System (RCS) is at design temperature and pressure for zero power. Performance of this test should not violate any of the referenced LCOs.
3  b. The Critical Boron Concentration - Control Rods Inserted Test measures the critical boron concentration at HZP, with a bank having


a worth of at least 1% k/k when fully inserted into the core. This test is used to measure the boron reactivity coefficient. With the core at HZP and all banks fully withdrawn, the boron concentration of the reactor coolant is gradually lowered in a continuous manner. The selected bank is then inserted to make up for the decreasing boron concentration until the selected bank has been moved over its entire
Attachment 1, Volume 6, Rev. 0, Page 194 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 PHYSICS TESTS Exceptions - MODE 2 BASES BACKGROUND        The primary purpose of the MODE 2 PHYSICS TESTS exceptions is to permit relaxations of existing LCOs to allow certain PHYSICS TESTS to be performed.
Section XI of 10 CFR 50, Appendix B (Ref. 1), requires that a test program be established to ensure that structures, systems, and components will perform satisfactorily in service. All functions necessary to ensure that the specified design conditions are not exceeded during normal operation and anticipated operational occurrences must be tested.
This testing is an integral part of the design, construction, and operation of the plant. Requirements for notification of the NRC, for the purpose of conducting tests and experiments, are specified in 10 CFR 50.59 (Ref. 2).
The key objectives of a test program are to (Ref. 3):
: a. Ensure that the facility has been adequately designed,                    1
: b. Validate the analytical models used in the design and analysis,          1
: c. Verify the assumptions used to predict unit response,                    1
: d. Ensure that installation of equipment in the facility has been accomplished in accordance with the design, and                          1
: e. Verify that the operating and emergency procedures are adequate.
To accomplish these objectives, testing is performed prior to initial criticality, during startup, during low power operations, during power ascension, at high power, and after each refueling. The PHYSICS TESTS requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions and that the core can be operated as designed (Ref. 4).
PHYSICS TESTS procedures are written and approved in accordance with established formats. The procedures include all information necessary to permit a detailed execution of the testing required to ensure that the design intent is met. PHYSICS TESTS are performed in accordance with these procedures and test results are approved prior to continued power escalation and long term power operation.
WOG STS                                    B 3.1.8-1                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 194 of 213


range of travel. The reactivity re sulting from each incremental bank movement is measured with a reactivity computer. The difference between the measured critical boron concentration with all rods fully withdrawn and with the bank inserted is determined. The boron INSERT 1 3 B 3.1.8 1 INSERT 1  The Differential Boron Worth Test determines if the measured differential boron worth is consistent with the predicted value. With the core at HZP, the change in equilibrium boron concentration is determined at different rod bank positions. As the rod bank or banks are moved, the reactivity change is measured using a reactivity computer. The measured reactivity change is divided by the difference in measured critical boron concentrations to determine the differential boron worth.      Insert Page B 3.1.8-2 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8    WOG STS B 3.1.8-3 Rev. 3.0, 03/31/04 BASES BACKGROUND  (continued)
Attachment 1, Volume 6, Rev. 0, Page 195 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES BACKGROUND (continued)
The PHYSICS TESTS required for reload fuel cycles (Ref. 4) in MODE 2 are listed below:
: a. Critical Boron Concentration - Control Rods Withdrawn,                            1
: b. Critical Boron Concentration - Control Rods Inserted,                          2  1 Differential Boron Worth
: c. Control Rod Worth,                                                                1
: d. Isothermal Temperature Coefficient (ITC), and                                    1
: e. Neutron Flux Symmetry.                                                           2 The first four tests are performed in MODE 2, and the last test can be                2 performed in either MODE 1 or 2. These and other supplementary tests may be required to calibrate the nuclear instrumentation or to diagnose operational problems. These tests may cause the operating controls and process variables to deviate from their LCO requirements during their performance.
[ a. The Critical Boron Concentration - Control Rods Withdrawn Test                    3 measures the critical boron concentration at hot zero power (HZP).
With all rods out, the lead control bank is at or near its fully withdrawn position. HZP is where the core is critical (keff = 1.0), and the Reactor Coolant System (RCS) is at design temperature and pressure for zero power. Performance of this test should not violate any of the referenced LCOs.
: b. The Critical Boron Concentration - Control Rods Inserted Test measures the critical boron concentration at HZP, with a bank having a worth of at least 1% k/k when fully inserted into the core. This test is used to measure the boron reactivity coefficient. With the core at HZP and all banks fully withdrawn, the boron concentration of the reactor coolant is gradually lowered in a continuous manner. The INSERT 1                                                                                            3 selected bank is then inserted to make up for the decreasing boron concentration until the selected bank has been moved over its entire range of travel. The reactivity resulting from each incremental bank movement is measured with a reactivity computer. The difference between the measured critical boron concentration with all rods fully withdrawn and with the bank inserted is determined. The boron WOG STS                                     B 3.1.8-2                                  Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 195 of 213


reactivity coefficient is determined by dividing the measured bank worth by the measured boron concentration difference. Performance of this test could violate LCO 3.1.4, "Rod Group Alignment Limits,"
Attachment 1, Volume 6, Rev. 0, Page 196 of 213 B 3.1.8 1
LCO 3.1.5, "Shutdown Bank Insertion Limit," or LCO 3.1.6, "Control
INSERT 1 The Differential Boron Worth Test determines if the measured differential boron worth is consistent with the predicted value. With the core at HZP, the change in equilibrium boron concentration is determined at different rod bank positions. As the rod bank or banks are moved, the reactivity change is measured using a reactivity computer. The measured reactivity change is divided by the difference in measured critical boron concentrations to determine the differential boron worth.
Insert Page B 3.1.8-2 Attachment 1, Volume 6, Rev. 0, Page 196 of 213


Bank Insertion Limits." 3  c. The Control Rod Worth Test is used to measure the reactivity worth of selected control banks. This test is performed at HZP and has three alternative methods of performance. The first method, the Boron Exchange Method, varies the reactor coolant boron concentration and moves the selected control bank in response to the changing boron concentration. The reactivity changes are measured with a reactivity computer. This sequence is repeated for the remaining control banks. The second method, the Rod Swap Method, measures the worth of a predetermined reference bank using the Boron Exchange Method above. The reference bank is then nearly fully inserted into the core. The selected bank is then inserted into the core as the reference bank is withdrawn. The HZP critical conditions are then determined with the selected bank fully inserted into the core. The worth of the selected bank is inferred, based on the position of the reference bank with respect to the selected bank. This sequence is repeated as necessary for the remaining control banks. The third method, the Boron Endpoint Method, moves the selected control bank over its entire length of travel and then varies the reactor coolant boron concentration to achieve HZP criticality again. The difference in boron concentration is the worth of the selected control bank. This sequence is repeated for the remaining control banks. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6.  
Attachment 1, Volume 6, Rev. 0, Page 197 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES BACKGROUND (continued) reactivity coefficient is determined by dividing the measured bank      3 worth by the measured boron concentration difference. Performance of this test could violate LCO 3.1.4, "Rod Group Alignment Limits,"
: d. The ITC Test measures the ITC of the reactor. This test is performed at HZP and has two methods of performance. The first method, the Slope Method, varies RCS temperature in a slow and continuous manner. The reactivity change is measured with a reactivity computer as a function of the temperature change. The ITC is the slope of the reactivity versus the temperature plot. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. The second method, the Endpoint Method, changes the RCS temperature and measures the reactivity at the beginning and end of the
LCO 3.1.5, "Shutdown Bank Insertion Limit," or LCO 3.1.6, "Control Bank Insertion Limits."
: c. The Control Rod Worth Test is used to measure the reactivity worth of selected control banks. This test is performed at HZP and has three alternative methods of performance. The first method, the Boron Exchange Method, varies the reactor coolant boron concentration and moves the selected control bank in response to the changing boron concentration. The reactivity changes are measured with a reactivity computer. This sequence is repeated for the remaining control banks. The second method, the Rod Swap Method, measures the worth of a predetermined reference bank using the Boron Exchange Method above. The reference bank is then nearly fully inserted into the core. The selected bank is then inserted into the core as the reference bank is withdrawn. The HZP critical conditions are then determined with the selected bank fully inserted into the core. The worth of the selected bank is inferred, based on the position of the reference bank with respect to the selected bank. This sequence is repeated as necessary for the remaining control banks. The third method, the Boron Endpoint Method, moves the selected control bank over its entire length of travel and then varies the reactor coolant boron concentration to achieve HZP criticality again. The difference in boron concentration is the worth of the selected control bank. This sequence is repeated for the remaining control banks. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6.
: d. The ITC Test measures the ITC of the reactor. This test is performed at HZP and has two methods of performance. The first method, the Slope Method, varies RCS temperature in a slow and continuous manner. The reactivity change is measured with a reactivity computer as a function of the temperature change. The ITC is the slope of the reactivity versus the temperature plot. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. The second method, the Endpoint Method, changes the RCS temperature and measures the reactivity at the beginning and end of the WOG STS                                B 3.1.8-3                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 197 of 213


PHYSICS TESTS Exceptions - MODE 2 B 3.1.8   WOG STS B 3.1.8-4 Rev. 3.0, 03/31/04  BASES BACKGROUND (continued)  
Attachment 1, Volume 6, Rev. 0, Page 198 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES BACKGROUND (continued) temperature change. The ITC is the total reactivity change divided by the total temperature change. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. Performance of this test could violate LCO 3.4.2, "RCS Minimum Temperature for Criticality."
 
: e. The Flux Symmetry Test measures the degree of azimuthal symmetry of the neutron flux at as low a power level as practical, depending on the test method employed. This test can be performed at HZP (Control Rod Worth Symmetry Method) or at  30% RTP (Flux Distribution Method). The Control Rod Worth Symmetry Method inserts a control bank, which can then be withdrawn to compensate for the insertion of a single control rod from a symmetric           3 set. The symmetric rods of each set are then tested to evaluate the symmetry of the control rod worth and neutron flux (power distribution). A reactivity computer is used to measure the control rod worths. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6. The Flux Distribution Method uses the incore flux detectors to measure the azimuthal flux distribution at selected locations with the core at  30% RTP. ]
temperature change. The ITC is the total reactivity change divided by the total temperature change. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. Performance of this test could violate LCO 3.4.2, "RCS Minimum Temperature for Criticality."   e. The Flux Symmetry Test measures the degree of azimuthal symmetry of the neutron flux at as low a power level as practical, depending on the test method employed. This test can be performed at HZP (Control Rod Worth Symmetry Method) or at  30% RTP (Flux Distribution Method). The Control Rod Worth Symmetry Method inserts a control bank, which can then be withdrawn to compensate for the insertion of a single control rod from a symmetric set. The symmetric rods of each set are then tested to evaluate the symmetry of the control rod worth and neutron flux (power distribution). A reactivity computer is used to measure the control rod worths. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6. The Flux Distribution Method uses the incore flux detectors to measure the azimuthal flux distribution at selected locations with the core at  30% RTP. ]
APPLICABLE           The fuel is protected by LCOs that preserve the initial conditions of the SAFETY               core assumed during the safety analyses. The methods for development ANALYSES             of the LCOs that are excepted by this LCO are described in the                     2 Reference 5 Westinghouse Reload Safety Evaluation Methodology Report (Ref. 5).
3 APPLICABLE The fuel is protected by LCOs that preserve the initial conditions of the SAFETY core assumed during the safety analyses. The methods for development ANALYSES of the LCOs that are excepted by this LCO are described in the Westinghouse Reload Safety Evaluation Methodology Report (Ref. 5). The above mentioned PHYSICS TESTS, and other tests that may be required to calibrate nuclear instrumentation or to diagnose operational problems, may require the operating control or process variables to deviate from their LCO limitations.
The above mentioned PHYSICS TESTS, and other tests that may be required to calibrate nuclear instrumentation or to diagnose operational problems, may require the operating control or process variables to deviate from their LCO limitations.
2Reference 5 The FSAR defines requirements for initial testing of the facility, including PHYSICS TESTS. Tables [14.1-1 and 14.1-2] summarize the zero, low power, and power tests. Requirements for reload fuel cycle PHYSICS TESTS are defined in ANSI/ANS-19.6.1-1985 (Ref. 4). Although these PHYSICS TESTS are generally accomplished within the limits for all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. When one or more of the requirements specified in LCO 3.1.3, "Moderator Temperature Coefficient (MTC)," LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 are suspended for PHYSICS TESTS, the fuel design criteria are preserved as long as the power level is limited to  5% RTP, the reactor coolant temperature is kept  531&deg;F, and SDM is within the limits provided in the COLR.
U The FSAR defines requirements for initial testing of the facility, including 13.3-1  2 PHYSICS TESTS. Tables [14.1-1 and 14.1-2] summarize the zero, low power, and power tests. Requirements for reload fuel cycle PHYSICS s TESTS are defined in ANSI/ANS-19.6.1-1985 (Ref. 4). Although these PHYSICS TESTS are generally accomplished within the limits for all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. When one or more of the requirements specified in LCO 3.1.3, "Moderator Temperature Coefficient (MTC)," LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 are suspended for PHYSICS TESTS, the fuel design criteria are preserved as long as the power level is limited to  5% RTP, the reactor coolant temperature is kept  531&deg;F, and SDM is within the limits               4 provided in the COLR.                       530 WOG STS                                       B 3.1.8-4                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 198 of 213
U 2 4 13.3-1 s 530 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8    WOG STS B 3.1.8-5 Rev. 3.0, 03/31/04 BASES APPLICABLE SAFETY ANALYSES  (continued)
 
The PHYSICS TESTS include measurement of core nuclear parameters or the exercise of control components that affect process variables. Among the process variables involved are AFD and QPTR, which represent initial conditions of the unit safety analyses. Also involved are the movable control components (control and shutdown rods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR.
 
As described in LCO 3.0.7, compliance with Test Exception LCOs is optional, and therefore no criteria of 10 CFR 50.36(c)(2)(ii) apply. Test Exception LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.


Attachment 1, Volume 6, Rev. 0, Page 199 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES APPLICABLE SAFETY ANALYSES (continued)
The PHYSICS TESTS include measurement of core nuclear parameters or the exercise of control components that affect process variables.
Among the process variables involved are AFD and QPTR, which represent initial conditions of the unit safety analyses. Also involved are the movable control components (control and shutdown rods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR.
As described in LCO 3.0.7, compliance with Test Exception LCOs is optional, and therefore no criteria of 10 CFR 50.36(c)(2)(ii) apply. Test Exception LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
Reference 6 allows special test exceptions (STEs) to be included as part of the LCO that they affect. It was decided, however, to retain this STE as a separate LCO because it was less cumbersome and provided additional clarity.
Reference 6 allows special test exceptions (STEs) to be included as part of the LCO that they affect. It was decided, however, to retain this STE as a separate LCO because it was less cumbersome and provided additional clarity.
LCO This LCO allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insert ion limits. One power range neutron flux channel may be bypassed, reducing the number of required channels from 4 to 3. Operation beyond specified limits is permitted for the purpose of performing PHYSICS TESTS and poses no threat to fuel  
LCO             This LCO allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits. One power range neutron flux channel may be bypassed, reducing the number of required channels from 4 to 3. Operation beyond specified limits is permitted for the purpose of performing PHYSICS TESTS and poses no threat to fuel integrity, provided the SRs are met.
The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 may be suspended and the number of required channels                16 RPS for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 18.e may be                4 reduced to 3 required channels during the performance of PHYSICS TESTS provided:
530
: a. RCS lowest loop average temperature is  [531]&deg;F,                              3  1
: b. SDM is within the limits provided in the COLR, and                                1
: c. THERMAL POWER is  5% RTP.
WOG STS                                  B 3.1.8-5                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 199 of 213


integrity, provided the SRs are met.
Attachment 1, Volume 6, Rev. 0, Page 200 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES APPLICABILITY      This LCO is applicable when performing low power PHYSICS TESTS.
The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 18.e may be reduced to 3 required channels during the performance of PHYSICS
The Applicability is stated as "during PHYSICS TESTS initiated in MODE 2" to ensure that the 5% RTP maximum power level is not exceeded. Should the THERMAL POWER exceed 5% RTP, and consequently the unit enter MODE 1, this Applicability statement prevents exiting this Specification and its Required Actions.
ACTIONS            A.1 and A.2 If the SDM requirement is not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. The operator should begin boration with the best source available for the plant conditions. Boration will be continued until SDM is within limit.
Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOs to within specification.
B.1 When THERMAL POWER is > 5% RTP, the only acceptable action is to open the reactor trip breakers (RTBs) to prevent operation of the reactor beyond its design limits. Immediately opening the RTBs will shut down the reactor and prevent operation of the reactor outside of its design limits.
C.1 530 When the RCS lowest Tavg is < 531&deg;F, the appropriate action is to restore      4 Tavg to within its specified limit. The allowed Completion Time of 15 minutes provides time for restoring Tavg to within limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Operation with the reactor critical and with temperature below 530                                                                                    4 531&deg;F could violate the assumptions for accidents analyzed in the safety analyses.
WOG STS                                    B 3.1.8-6                        Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 200 of 213


TESTS provided:
Attachment 1, Volume 6, Rev. 0, Page 201 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES ACTIONS (continued)
16 4 RPS  a. RCS lowest loop average temperature is  [531]&deg;F,   b. SDM is within the limits provided in the COLR, and 530 ; ; 3 1 1 c. THERMAL POWER is  5% RTP.
D.1 If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, for reaching MODE 3 in an orderly manner and without challenging plant systems.
PHYSICS TESTS Exceptions - MODE 2 B 3.1.8    WOG STS B 3.1.8-6 Rev. 3.0, 03/31/04  BASES APPLICABILITY This LCO is applicable when performing low power PHYSICS TESTS. The Applicability is stated as "during PHYSICS TESTS initiated in MODE 2" to ensure that the 5% RTP maximum power level is not exceeded. Should the THERMAL POWER exceed 5% RTP, and consequently the unit enter MODE 1, this Applicability statement prevents exiting this Specification and its Required Actions.
SURVEILLANCE     SR 3.1.8.1 REQUIREMENTS The power range and intermediate range neutron detectors must be verified to be OPERABLE in MODE 2 by LCO 3.3.1, "Reactor Trip                 4 System (RTS) Instrumentation." A CHANNEL OPERATIONAL TEST is performed on each power range and intermediate range channel prior to initiation of the PHYSICS TESTS. This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS.
ACTIONS A.1 and A.2
SR 3.1.8.2       1 530 4
 
Verification that the RCS lowest loop Tavg is  531&deg;F will ensure that the       4 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.
If the SDM requirement is not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. The operator should begin boration with the best source available for the plant conditions. Boration will be continued until SDM is within limit.
SR 3.1.8.3       2                                                           4 Verification that the THERMAL POWER is  5% RTP will ensure that the plant is not operating in a condition that could invalidate the safety analyses. Verification of the THERMAL POWER 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.
 
WOG STS                                   B 3.1.8-7                          Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 201 of 213
Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOs to within specification.
 
B.1  When THERMAL POWER is > 5% RTP, the only acceptable action is to open the reactor trip breakers (RTBs) to prevent operation of the reactor beyond its design limits. Immediately opening the RTBs will shut down the reactor and prevent operation of the reactor outside of its design limits.
 
C.1  When the RCS lowest Tavg is < 531&deg;F, the appropriate action is to restore Tavg to within its specified limit. The allowed Completion Time of 15 minutes provides time for restoring Tavg to within limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Operation with the reactor critical and with temperature below 531&deg;F could violate the assumptions for accidents analyzed in the safety analyses.
530 4 530 4 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8   WOG STS B 3.1.8-7 Rev. 3.0, 03/31/04  BASES ACTIONS (continued)  
 
D.1   If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, for reaching MODE 3 in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.1.8.1 REQUIREMENTS  
 
The power range and intermediate range neutron detectors must be verified to be OPERABLE in MODE 2 by LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." A CHANNEL OPERATIONAL TEST is performed on each power range and intermediate range channel prior to initiation of the PHYSICS TESTS. This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS.  
 
SR 3.1.8.2 Verification that the RCS lowest loop Tavg is  531&deg;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.
4 530 4 1 4 SR 3.1.8.3 2 4 Verification that the THERMAL POWER is  5% RTP will ensure that the plant is not operating in a condition that could invalidate the safety analyses. Verification of the THERMAL POWER 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.  
 
PHYSICS TESTS Exceptions - MODE 2 B 3.1.8    WOG STS B 3.1.8-8 Rev. 3.0, 03/31/04 BASES SURVEILLANCE REQUIREMENTS  (continued)
 
SR  3.1.8.4 3 4 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:
: a. RCS boron concentration, ;1  b. Control bank position, 1  c. RCS average temperature, ;1
: d. Fuel burnup based on gross thermal energy generation, ;1 ;  e. Xenon concentration, 1
: f. Samarium concentration, ;1  g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH), ;1  h. Moderate defect, when above the POAH, and
;1 ; 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.


Attachment 1, Volume 6, Rev. 0, Page 202 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.1.8.4          3                                                                          4 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:
: a. RCS boron concentration,                                                                    1
: b. Control bank position,                                                                      1
: c. RCS average temperature,                                                                    1
: d. Fuel burnup based on gross thermal energy generation,                                      1
: e. Xenon concentration,                                                                        1
: f. Samarium concentration,                                                                    1
: g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH),                                                                        1
: h. Moderate defect, when above the POAH, and                                                  1
: 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.
The Frequency of 24 hours is based on the generally slow change in required boron concentration and on the low probability of an accident occurring without the required SDM.
The Frequency of 24 hours is based on the generally slow change in required boron concentration and on the low probability of an accident occurring without the required SDM.
REFERENCES 1. 10 CFR 50, Appendix B, Section XI.  
REFERENCES     1. 10 CFR 50, Appendix B, Section XI.
: 2. 10 CFR 50.59.  
: 2. 10 CFR 50.59.
: 3. Regulatory Guide 1.68, Revision 2, August, 1978.  
: 3. Regulatory Guide 1.68, Revision 2, August, 1978.
: 4. ANSI/ANS-19.6.1-1985, December 13, 1985.
: 4. ANSI/ANS-19.6.1-1985, December 13, 1985.
2 5. WCAP-9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology Report," July 1985.  
: 5. WCAP-9273-NP-A, "Westinghouse Reload Safety Evaluation                                     2 Methodology Report," July 1985.
: 6. WCAP-11618, including Addendum 1, April 1989.
: 6. WCAP-11618, including Addendum 1, April 1989.
DOM-NAF-5, Revision 0.0, "Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station," July 2006.
DOM-NAF-5, Revision 0.0, Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station, July 2006.
JUSTIFICATION FOR DEVIATIONS ITS 3.1.8 BASES, PHYSICS TESTS Exceptions - MODE 2  
WOG STS                                      B 3.1.8-8                                      Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 202 of 213
: 1. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.  
 
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.  
Attachment 1, Volume 6, Rev. 0, Page 203 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.8 BASES, PHYSICS TESTS Exceptions - MODE 2
: 3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.  
: 1. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
: 2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
: 3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
: 4. The Bases are changed to reflect changes made to the Specification.
: 4. The Bases are changed to reflect changes made to the Specification.
Kewaunee Power Station Page 1 of 1 Specific No Significant Haza rds Considerations (NSHCs)
Kewaunee Power Station                 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 203 of 213
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 There are no specific NSHC discussions for this Specification.
 
Kewaunee Power Station Page 1 of 1 ATTACHMENT 9 RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS
Attachment 1, Volume 6, Rev. 0, Page 204 of 213 Specific No Significant Hazards Considerations (NSHCs)
Attachment 1, Volume 6, Rev. 0, Page 204 of 213
 
Attachment 1, Volume 6, Rev. 0, Page 205 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 There are no specific NSHC discussions for this Specification.
Kewaunee Power Station             Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 205 of 213
, Volume 6, Rev. 0, Page 206 of 213 ATTACHMENT 9 RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS , Volume 6, Rev. 0, Page 206 of 213


CTS 3.2.a, CHEMICAL AND VOLUME CONTROL SYSTEM Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)
Attachment 1, Volume 6, Rev. 0, Page 207 of 213 CTS 3.2.a, CHEMICAL AND VOLUME CONTROL SYSTEM Attachment 1, Volume 6, Rev. 0, Page 207 of 213
CTS 3.2.a3.2 CHEMICAL AND VOLUME CONTROL SYSTEM APPLICABILITY Amendment No. 116 TS 3.2-1 03/28/95 Applies to the operational status of the Chemical and Volume Control System.
, Volume 6, Rev. 0, Page 208 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 208 of 213


OBJECTIVE   To define those conditions of the Chemical and Volume Control System necessary to ensure safe reactor operation.
Attachment 1, Volume 6, Rev. 0, Page 209 of 213 CTS 3.2.a 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM APPLICABILITY Applies to the operational status of the Chemical and Volume Control System.
OBJECTIVE R01 To define those conditions of the Chemical and Volume Control System necessary to ensure safe reactor operation.
SPECIFICATIONS
SPECIFICATIONS
: a. When fuel is in the reactor there shall be at least one flow path to the core for boric acid injection.
: a. When fuel is in the reactor there shall be at least one flow path to the core for boric acid injection.
R01 Page 1 of 1 DISCUSSION OF CHANGES CTS 3.2.a, CHEMICAL VOLUME CONTROL SYSTEM ADMINISTRATIVE CHANGES None
Amendment No. 116 TS 3.2-1                                  03/28/95 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 209 of 213
 
MORE RESTRICTIVE CHANGES
 
None
 
RELOCATED SPECIFICATIONS


R01 CTS 3.2.a provides requirements on boric acid flow paths to the reactor core although the specification is labeled "Chemical and Volume Control System."
Attachment 1, Volume 6, Rev. 0, Page 210 of 213 DISCUSSION OF CHANGES CTS 3.2.a, CHEMICAL VOLUME CONTROL SYSTEM ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.2.a provides requirements on boric acid flow paths to the reactor core although the specification is labeled "Chemical and Volume Control System.
The purpose for a boric acid flow path is for control of the chemical neutron absorber (boron) concentration in the Reactor Coolant System (RCS) and to help maintain the SHUTDOWN MARGIN. To accomplish this, the CTS requires at least one flow path to the core for boric acid injection. The Chemical and Volume Control System is not assumed to be OPERABLE to mitigate the consequences of a design basis accident (DBA) or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event, the automatic response, or that required by the operator, is to close the appropriate valves in the reactor makeup system. This action is required before the shutdown margin is lost. Operation of the boration subsystem is not assumed to mitigate this event (KPS USAR Section 14.1.4). The ITS does not include this Specification. This changes the CTS by relocating this Specification to the Technical Requirements Manual (TRM).
The purpose for a boric acid flow path is for control of the chemical neutron absorber (boron) concentration in the Reactor Coolant System (RCS) and to help maintain the SHUTDOWN MARGIN. To accomplish this, the CTS requires at least one flow path to the core for boric acid injection. The Chemical and Volume Control System is not assumed to be OPERABLE to mitigate the consequences of a design basis accident (DBA) or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event, the automatic response, or that required by the operator, is to close the appropriate valves in the reactor makeup system. This action is required before the shutdown margin is lost. Operation of the boration subsystem is not assumed to mitigate this event (KPS USAR Section 14.1.4). The ITS does not include this Specification. This changes the CTS by relocating this Specification to the Technical Requirements Manual (TRM).
This change is acceptable because CTS 3.2.a does not meet the 10 CFR 50.36(c)(2)(ii) criteria for inclusion into the ITS.  
This change is acceptable because CTS 3.2.a does not meet the 10 CFR 50.36(c)(2)(ii) criteria for inclusion into the ITS.
 
10 CFR 50.36(c)(2)(ii) Criteria Evaluation:
10 CFR 50.36(c)(2)(ii) Criteria Evaluation:  
: 1. The Chemical and Volume Control System is not used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary. The Chemical and Volume Control System does not satisfy criterion 1.
: 1. The Chemical and Volume Control System is not used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary. The Chemical and Volume Control System does not satisfy criterion 1.  
: 2. The Chemical and Volume Control System is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The Chemical and Volume Control System does not satisfy criterion 2.
: 2. The Chemical and Volume Control System is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The Chemical and Volume Control System does not satisfy criterion 2.  
: 3. The Chemical and Volume Control System is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The Chemical and Volume Control System does not satisfy criterion 3.
: 3. The Chemical and Volume Control System is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The  
Kewaunee Power Station                 Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 210 of 213
 
Chemical and Volume Control System does not satisfy criterion 3.
Kewaunee Power Station Page 1 of 2 DISCUSSION OF CHANGES CTS 3.2.a, CHEMICAL VOLUME CONTROL SYSTEM Kewaunee Power Station Page 2 of 2
: 4. As discussed in Section 4.0 (Appendix A, page A-6) and summarized in Table 1 of WCAP-11618, the loss of the Chemical and Volume Control System (i.e., boric acid flow path to the core - operating and shutdown)was found to be a non-significant risk contributor to core damage frequency and offsite releases. Dominion Energy Kewaunee (DEK) has reviewed this evaluation, considers it applicable to Kewaunee Power Station (KPS), and concurs with the assessment. The Chemical and Volume Control System does not satisfy criterion 4.
 
Since 10 CFR 50.36(c)(2)(ii) criteria have not been met, the Chemical and Volume Control System may be relocated out of the Technical Specifications. The Chemical and Volume Control System will be relocated to the TRM. Changes to the TRM will be controlled by the provisions of 10 CFR 50.59. This change is designated as relocation because the Specification did not meet the criteria in 10 CFR 50.36(c)(2)(ii) and has been relocated to the TRM.
REMOVED DETAIL CHANGES


None  
Attachment 1, Volume 6, Rev. 0, Page 211 of 213 DISCUSSION OF CHANGES CTS 3.2.a, CHEMICAL VOLUME CONTROL SYSTEM
: 4. As discussed in Section 4.0 (Appendix A, page A-6) and summarized in Table 1 of WCAP-11618, the loss of the Chemical and Volume Control System (i.e., boric acid flow path to the core - operating and shutdown)was found to be a non-significant risk contributor to core damage frequency and offsite releases. Dominion Energy Kewaunee (DEK) has reviewed this evaluation, considers it applicable to Kewaunee Power Station (KPS), and concurs with the assessment. The Chemical and Volume Control System does not satisfy criterion 4.
Since 10 CFR 50.36(c)(2)(ii) criteria have not been met, the Chemical and Volume Control System may be relocated out of the Technical Specifications.
The Chemical and Volume Control System will be relocated to the TRM.
Changes to the TRM will be controlled by the provisions of 10 CFR 50.59. This change is designated as relocation because the Specification did not meet the criteria in 10 CFR 50.36(c)(2)(ii) and has been relocated to the TRM.
REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station                  Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 211 of 213


LESS RESTRICTIVE CHANGES None
Attachment 1, Volume 6, Rev. 0, Page 212 of 213 Specific No Significant Hazards Considerations (NSHCs)
Attachment 1, Volume 6, Rev. 0, Page 212 of 213


Specific No Significant Haza rds Considerations (NSHCs)
Attachment 1, Volume 6, Rev. 0, Page 213 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.2.a, CHEMICAL AND VOLUME CONTROL SYSTEM There are no specific NSHC discussions for this Specification.
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.2.a, CHEMICAL AND VOLUME CONTROL SYSTEM There are no specific NSHC discussions for this Specification.
Kewaunee Power Station             Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 213 of 213}}
Kewaunee Power Station Page 1 of 1}}

Latest revision as of 09:18, 12 March 2020

Attachment 1 - Volume 6, Kewaunee Power Station, Improved Technical Specifications Conversion, ITS Section 3.1, Reactivity Control Systems, Revision 0
ML092440416
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Site: Kewaunee Dominion icon.png
Issue date: 08/24/2009
From:
Dominion Energy Kewaunee
To:
Office of Nuclear Reactor Regulation
References
Download: ML092440416 (213)
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Attachment 1, Volume 6, Rev. 0, Page 1 of 213 ATTACHMENT 1 VOLUME 6 KEWAUNEE POWER STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.1 REACTIVITY CONTROL SYSTEMS Revision 0 Attachment 1, Volume 6, Rev. 0, Page 1 of 213

Attachment 1, Volume 6, Rev. 0, Page 2 of 213 LIST OF ATTACHMENTS

1. ITS 3.1.1
2. ITS 3.1.2
3. ITS 3.1.3
4. ITS 3.1.4
5. ITS 3.1.5
6. ITS 3.1.6
7. ITS 3.1.7
8. ITS 3.1.8
9. Relocated/Deleted Current Technical Specifications (CTS)

Attachment 1, Volume 6, Rev. 0, Page 2 of 213

, Volume 6, Rev. 0, Page 3 of 213 ATTACHMENT 1 ITS 3.1.1, SHUTDOWN MARGIN , Volume 6, Rev. 0, Page 3 of 213

Attachment 1, Volume 6, Rev. 0, Page 4 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

Attachment 1, Volume 6, Rev. 0, Page 4 of 213

Attachment 1, Volume 6, Rev. 0, Page 5 of 213 ITS A01 ITS 3.1.1 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS APPLICABILITY Applies to the limits on core fission power distributions and to the limits on control rod operations.

OBJECTIVE To ensure: 1) core subcriticality after reactor trip, 2) acceptable core power distribution during power operation in order to maintain fuel integrity in normal operation transients associated with faults of moderate frequency, supplemented by automatic protection and by administrative procedures, and to maintain the design basis initial conditions for limiting faults, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.

SPECIFICATION

a. Shutdown Reactivity A02 Applicability When the reactor is subcritical prior to reactor startup, the SHUTDOWN MARGIN shall be at LCO 3.1.1 least that as specified in the COLR Add proposed ACTION A M01
b. Power Distribution Limits See ITS 3.2.1 and
1. At all times, except during Low Power Physics Tests, the hot channel factors defined in 3.2.2 the basis must meet the following limits:

See ITS A. FQN(Z) Limits shall be as specified in the COLR. 3.2.1 B. FHN Limits shall be as specified in the COLR.

2. If FHN not within limits:

A. Perform the following:

i. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either, restore FHN to within its limit or reduce thermal power to less than 50% of RATED POWER. See ITS 3.2.2 ii. Reduce the Power Range Neutron Flux-High Trip Setpoint to 55% of RATED POWER within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

iii. Verify FHN within limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. If the actions of TS 3.10.b.2.A are not completed within the specified time, then reduce thermal power to 5% of rated power within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Amendment No. 167 TS 3.10-1 04/04/2003 Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 5 of 213

Attachment 1, Volume 6, Rev. 0, Page 6 of 213 ITS A01 ITS 3.1.1 4.1 OPERATIONAL SAFETY REVIEW APPLICABILITY Applies to items directly related to safety limits and LIMITING CONDITIONS FOR OPERATION.

OBJECTIVE To assure that instrumentation shall be checked, tested, and calibrated, and that equipment and sampling tests shall be conducted at sufficiently frequent intervals to ensure safe operation.

SPECIFICATION

a. Calibration, testing, and checking of protective instrumentation channels and testing of See other ITS logic channels shall be performed as specified in Table TS 4.1-1.

SR 3.1.1.1 b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and TS 4.1-3.

c. Deleted
d. Deleted
e. Deleted Amendment No. 119 TS 4.1-1 04/18/95 Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 6 of 213

ITS A01 ITS 3.1.1 TABLE TS 4.1-2 See CTS See ITS 3.1.e MINIMUM FREQUENCIES FOR SAMPLING TESTS 3.4.16 SAMPLING TESTS TEST FREQUENCY (1)

1. Reactor Coolant a. Gross Radioactivity Determination (excluding tritium) 5/week Samples b. DOSE EQUIVALENT I-131 Concentration 1/14 days(2)
c. Tritium activity Monthly
d. Chemistry (Cl, F, O2)(3) 3/week(4)
e. Determination 1/6 months(5)
f. RCS isotopic analysis for Iodine Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in accordance with TS 3.1.c.2.C.
2. Reactor Coolant Boron(6) Boron Concentration(3) 2/week M02 LA01 Add proposed SR 3.1.1.1 M02 , Volume 6, Rev. 0, Page 7 of 213 Attachment 1, Volume 6, Rev. 0, Page 7 of 213 (1) See ITS Maximum time between tests is 3 days. 3.4.16 (2) L01 Sample required only when in the OPERATING MODE. A02 (3) See CTS Test required in all plant modes. 3.1.e (4)

Maximum time between tests is 4 days.

(5)

Sample after a minimum of 2 EFPD and 20 days of OPERATING MODE operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

(6)

A reactor coolant boron concentration sample does not have to be taken when the core is completely unloaded. See ITS 3.9.1 Amendment No. 119 Page 1 of 2 04/18/95 Page 3 of 3

Attachment 1, Volume 6, Rev. 0, Page 8 of 213 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM)

ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.10.a states that when the reactor is subcritical prior to startup, the SHUTDOWN MARGIN (SDM) shall be at least that as specified in the COLR.

CTS Table 4.1-2 Sampling Test 2 footnote (3) states that the boron concentration test is required in all plant modes. ITS 3.1.1 is applicable in MODE 2 with keff

< 1.0 and in MODES 3, 4, and 5. This changes the CTS by specifically stating that the applicability is in MODE 2 with keff < 1.0 and in MODES 3, 4, and 5. The change in Applicability for MODE 1 and MODE 2 with keff 1.0 is discussed in DOC L01. In MODE 6, the reactor head is detensioned or removed, so a reactor startup cannot occur. In addition, ITS 3.9.1 provides the MODE 6 boron concentration limits.

This change is acceptable because the applicability has not changed. This change results in a format change only to comply with the ISTS presentation of the Applicability. The purpose of CTS Table 4.1-2 Sampling Test 2 is to verify that the boron concentration is within limits to satisfy the SDM. In MODE 2 with keff < 1.0 and in MODES 3, 4, and 5 the boron concentration is taken into consideration as part of the reactivity balance calculation required in ITS SR 3.1.1.1. The deletion of the MODE 1 and MODE 2 with keff > 1.0 is discussed in DOC L01 and the MODE 6 requirement is covered in ITS 3.9.1.

Therefore, the change is acceptable because the boron concentration requirements have not changed. This change is designated as an administrative change since it does not result in any technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.10.a does not supply any explicit time to restore the SDM to within limits when the SDM is not within its limit. As a result, LCO 3.0.c would be entered, which requires action to be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and to place the unit in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN (equivalent to ITS MODE 5) within the subsequent 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Under similar conditions, ITS 3.1.1 provides 15 minutes to initiate boration to restore SDM to within limits. This changes the CTS by providing a maximum time limit of 15 minutes to initiate boration to restore SDM to within limits prior to entering ITS LCO 3.0.3.

The purpose of CTS 3.10.a SDM requirements is to provide sufficient reactivity margin to ensure that fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). This change is acceptable Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 8 of 213

Attachment 1, Volume 6, Rev. 0, Page 9 of 213 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) because the Completion Time is consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the allowed Completion Time. The ITS 3.1.1 Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. In addition, the ITS Bases for the ACTION states that boration must be initiated promptly. The current action (LCO 3.0.c) allows up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to start adding negative reactivity. The proposed action requires negative reactivity addition to begin in 15 minutes, and to continue until SDM is restored. This change is designated as more restrictive because a shorter time is allowed in ITS to restore parameters to within the LCO limits than was allowed in the CTS.

M02 CTS 3.10.a does not provide any Surveillance Requirements for verifying that the SDM is within the limits specified in the COLR. CTS Table 4.1-2 provides one parameter of the SDM in the boron concentration test which is performed two times per week. ITS SR 3.1.1.1 requires verifying that the SDM is within the limits specified in the COLR every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This changes the CTS by adding a new Surveillance Requirement to verify the SDM.

This change is acceptable because the added Surveillance Requirement proves that the SDM requirements are being met. Furthermore, the boron concentration test is a portion of the reactivity effects that are necessary to perform a reactivity balance calculation but is not included in the reactivity balance calculation. In addition to the RCS boron concentration, the control bank position, RCS average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, samarium concentration and isothermal temperature must be taken into account in the performance of the reactivity balance calculation. This change is designated as more restrictive because a new Surveillance Requirement is being added that was not specified in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS Table 4.1-2 Sampling Test 2 requires sampling tests of RCS boron concentration. ITS 3.1.1 does not have a specific requirement to test RCS boron concentration. This changes the CTS by removing the details of the RCS boron concentration test to the Bases.

The removal of these details for performing Surveillance Requirement from the Technical Specification is acceptable since this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. In ITS 3.1.1, RCS boron concentration is a parameter, in addition to control bank position, RCS average temperature, fuel Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 9 of 213

Attachment 1, Volume 6, Rev. 0, Page 10 of 213 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) burnup based on gross thermal energy generation, xenon concentration, samarium concentration and isothermal temperature, used in performing the reactivity balance calculation. Therefore, the details of how SDM is calculated do not need to appear in the Specification in order for the requirement to apply.

Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specifications Bases Control Program in Chapter 5. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 5 - Deletion of Surveillance Requirement) CTS Table TS 4.1-2 Sampling Test 2, footnote (3) states that the boron concentration test is required in all plant modes. ITS 3.1.1 is applicable in MODE 2 with keff < 1.0 and in MODES 3, 4, and 5. This changes CTS by not requiring the boron concentration test in MODE 1 and MODE 2 with keff 1.0. ITS 3.9.1 provides the MODE 6 boron concentration limits.

The purpose of CTS Table TS 4.1-2, Sampling Test 2 is to verify the boron concentration. When the reactor is critical (MODE 1 and MODE 2 with keff 1.0),

SDM is verified by ensuring that the control rods are within the insertion limits (see ITS 3.1.5 and ITS 3.1.6). If the control banks are not within their insertion limits in MODE 1 and MODE 2 with keff 1.0, ITS 3.1.6, Required Action A.1.1 and B.1.1 require the SDM to be verified within limits. This requires performing a reactivity balance calculation considering Reactor Coolant System (RCS) boron concentration, control bank position, RCS average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, samarium concentration and isothermal temperature. Therefore, the boron concentration sampling test is not needed in MODE 1 and MODE 2 with keff 1.0. This change has been designated as less restrictive because a Surveillance that was required in CTS is no longer required in ITS.

Kewaunee Power Station Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 10 of 213

Attachment 1, Volume 6, Rev. 0, Page 11 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 11 of 213

Attachment 1, Volume 6, Rev. 0, Page 12 of 213 CTS SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) 3.10.a LCO 3.1.1 SDM shall be within the limits specified in the COLR.

3.10.a APPLICABILITY: MODE 2 with keff < 1.0, MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC M01 A. SDM not within limits. A.1 Initiate boration to restore 15 minutes SDM to within limits.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M02 SR 3.1.1.1 Verify SDM to be within the limits specified in the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> COLR.

WOG STS 3.1.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 12 of 213

Attachment 1, Volume 6, Rev. 0, Page 13 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1, SHUTDOWN MARGIN (SDM)

None Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 13 of 213

Attachment 1, Volume 6, Rev. 0, Page 14 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 14 of 213

Attachment 1, Volume 6, Rev. 0, Page 15 of 213 SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND According to GDC 26 (Ref. 1), the reactivity control systems must be INSERT 1 redundant and capable of holding the reactor core subcritical when shut 8 down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.

SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or scram of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth 1

is fully withdrawn. control (RCCA)

The system design requires that two independent reactivity control 8 INSERT 2 systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor Coolant System (RCS). The Control Rod System can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the Control Rod System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn. The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions.

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

APPLICABLE The minimum required SDM is assumed as an initial condition in safety SAFETY analyses. The safety analysis (Ref. 2) establishes an SDM that ensures 9 ANALYSES specified acceptable fuel design limits are not exceeded for normal (which includes reactor operation and AOOs, with the assumption of the highest worth rod stuck safety analyses not in out on scram. For MODE 5, the primary safety analysis that relies on the 10 USAR Chapter 14)

SDM limits is the boron dilution analysis.

WOG STS B 3.1.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 15 of 213

Attachment 1, Volume 6, Rev. 0, Page 16 of 213 B 3.1.1 8

INSERT 1 Two independent reactivity control systems, preferably from different principles, shall be provided (Ref. 1). The reactivity control systems shall be capable of making and holding the core sub-critical from any hot standby or hot operating condition (Ref. 2).

8 INSERT 2 Furthermore, USAR GDC 30, "Reactivity Holddown Capability," requires that the reactivity control system provided shall be capable of making the core subcritical under credible accident conditions with appropriate margins for contingencies and limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public (Ref. 3).

Insert Page B 3.1.1-1 Attachment 1, Volume 6, Rev. 0, Page 16 of 213

Attachment 1, Volume 6, Rev. 0, Page 17 of 213 SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)

The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:

a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events, 2
b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature limits for AOOs, and 280 cal/gm energy deposition for the rod ejection 1 accident), and 200 2
c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

The most limiting accident for the SDM requirements is based on a main steam line break (MSLB), as described in the accident analysis (Ref. 2). 1 4

The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. As RCS temperature decreases, the severity of an MSLB decreases until the MODE 5 value is reached. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs, is a guillotine break of a main steam line inside containment initiated at the end of core life. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, no fuel damage occurs as a result of the post trip return to power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1.

In addition to the limiting MSLB transient, the SDM requirement must also protect against:

a. Inadvertent boron dilution, 2
b. An uncontrolled rod withdrawal from subcritical or low power condition, 2
and WOG STS B 3.1.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 17 of 213

Attachment 1, Volume 6, Rev. 0, Page 18 of 213 SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)

c. Startup of an inactive reactor coolant pump (RCP), and 3

c d. Rod ejection.

Each of these events is discussed below.

In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest.

Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high an overtemperature T 1 power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits.

The startup of an inactive RCP will not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur 3

due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition.

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

The ejection of a rod also produces a time dependent redistribution of core power.

SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed from the control room, SDM is considered an initial 4

condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.

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

WOG STS B 3.1.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 18 of 213

Attachment 1, Volume 6, Rev. 0, Page 19 of 213 SDM B 3.1.1 BASES LCO (continued) 4 5 The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most 1 limiting analyses that establish the SDM value of the LCO. For MSLB 50.67 accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4). 1 Accident Source Term For the boron dilution accident, if the LCO is violated, the minimum 6 required time assumed for operator action to terminate dilution may no longer be applicable.

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

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

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

In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly refueling concentrated solution, such as that normally found in the boric acid storage tank, or the borated water storage tank. The operator should 1 borate with the best source available for the plant conditions.

In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm. Assuming that a value of 1% k/k must be recovered and a boration flow rate of [ ] gpm, it is 40 5 possible to increase the boron concentration of the RCS by 100 ppm in 12 1 approximately 35 minutes. If a boron worth of 10 pcm/ppm is assumed, 40 this combination of parameters will increase the SDM by 1% k/k. These 13,000 boration parameters of [ ] gpm and [ ] ppm represent typical values and 5 1

are provided for the purpose of offering a specific example.

are the typical values when borating from the boric acid storage tank WOG STS B 3.1.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 19 of 213

Attachment 1, Volume 6, Rev. 0, Page 20 of 213 SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2 with Keff 1.0, SDM is verified by observing that the 7 requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum MODE 2 with worth.

keff <1.0 and in In MODES 3, 4, and 5, the SDM is verified by performing a reactivity 6 balance calculation, considering the listed reactivity effects:

a. RCS boron concentration, 2
b. Control bank position, 2
c. RCS average temperature, 2
d. Fuel burnup based on gross thermal energy generation, 2
e. Xenon concentration, 2
f. Samarium concentration, and 2
g. Isothermal temperature coefficient (ITC).

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

The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on the generally slow change in required boron concentration and the low probability of an accident occurring without the required SDM. This allows time for the operator to collect the required data, which includes performing a boron concentration analysis, and complete the calculation.

USAR, Section 3.1.2.3, GDC 27, 8

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26. "Redundancy of Reactivity Control."

4 2. USAR, Section 3.1.2.4, GDC 28, 8

2. FSAR, Chapter [15]. Section 14.2.5 "Reactivity Hot Shutdown Capability." 5 U
3. USAR, Section 3.1.2.6, GDC 30,
3. FSAR, Chapter [15]. Section 14.1.4 "Reactivity Holddown Capability." 8 5 5

6

4. 10 CFR 100. 50.67 1 WOG STS B 3.1.1-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 20 of 213

Attachment 1, Volume 6, Rev. 0, Page 21 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1 BASES, SHUTDOWN MARGIN (SDM)

1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
2. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
3. ISTS 3.1.1 Applicable Safety Analyses states, in part, that in addition to the limiting Main Steam Line Break (MSLB), the SDM requirements must also protect against inadvertent boron dilution, an uncontrolled rod withdrawal from subcritical or low power condition, startup of an inactive reactor coolant pump (RCP), and rod ejection.

For the description of the event, the ISTS Bases states that the startup of an inactive RCP will not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition. USAR, Section 14.1.5, the Startup of an Inactive Reactor Coolant Loop, states that if the plant were to operate with one reactor coolant pump (RCP) out-of-service, there would be reverse flow through the inactive loop due to the pressure difference across the reactor vessel and because there are no isolation valves or check valves in the reactor coolant loops. The cold leg temperature in the inactive loop is identical to the cold leg temperature of the active loop (the reactor core inlet temperature). Therefore, a startup of an inactive RCP would not cause a discernable power excursion, since water colder than the operating loop is not being added to the reactor vessel. Thus, the ITS 3.1.1 Bases has been modified to delete the startup of an inactive RCP requirements.

4. The Applicable Safety Analyses discussion states SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). It also says that even though SDM is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of the accident analysis assumptions. The additional sentence has been deleted. The NRC Final Policy Statement on Technical Improvements of July 22, 1993 (58 FR 39132) states that process variable captured by Criterion 2 are not limited to only those directly monitored and controlled from the control room. It also states that Criterion 2 includes other features or characteristics that are specifically assumed in Design Basis Accident and Transient analyses even if they cannot be directly observed in the control room (e.g., moderator temperature coefficient and hot channel factors). Since the Final Policy Statement provides guidance on which types of parameters satisfy Criterion 2, there is no reason to duplicate these words in the KPS ITS.
5. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
6. ISTS SR 3.1.1.1 Bases states, in part, that in MODES 1 and 2 with keff 1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. It continues by stating, in part, that in MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation. ITS SR 3.1.1.1 Bases states, in part, that Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 21 of 213

Attachment 1, Volume 6, Rev. 0, Page 22 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1 BASES, SHUTDOWN MARGIN (SDM) in MODES 1 and 2 with keff 1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. It continues by stating, in part, that in MODE 2 with keff < 1.0 and in MODES 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation. This changes the ISTS by stating that in MODE 2 with keff < 1.0 the SDM is verified by performing a reactivity balance calculation. This is acceptable because the ISTS did not state that a reactivity balance calculation is performed when the keff < 1.0 in MODE 2.

7. Typographical error corrected.
8. The ISTS lists GDC 26 of Appendix A to 10 CFR 50 as the reference document for the requirement that there be two independent reactivity control systems of different design principles. Per the information contained in USAR Section 1.8, Kewaunee Power Station (KPS) was designed, constructed, and is being operated to comply with the Atomic Energy Commission (AEC) General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, as proposed on July 10, 1967. Since the plant was approximately 50% complete prior to the February 20, 1971 issuance of 10 CFR 50 Appendix A General Design Criteria, KPS was not required to be reanalyzed and the Final Safety Analysis Report (FSAR) was not required to be revised to reflect these later criteria. However, the AEC Safety Evaluation Report (SER), issued July 24, 1972, acknowledged that the AEC staff assessed the plant, as described in the FSAR (Amendment No. 7), against the Appendix A design criteria and determined that the plant design generally conforms to the intent of the Appendix A criteria. As a result, KPS utilizes AEC GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot Shutdown Capability," and GDC 30, "Reactivity Holddown Capability", as the licensing reference documents for the design principles.
9. Clarification has been added since all the reactor safety analyses are not in USAR Chapter 14.
10. KPS does not analyze a boron dilution event in MODE 5. As stated in KPS USAR, Section 14.1.4.2, boron dilutions during cold shutdown are not part of the KPS licensing basis.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 22 of 213

Attachment 1, Volume 6, Rev. 0, Page 23 of 213 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 6, Rev. 0, Page 23 of 213

Attachment 1, Volume 6, Rev. 0, Page 24 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.1, SHUTDOWN MARGIN (SDM)

There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 24 of 213

, Volume 6, Rev. 0, Page 25 of 213 ATTACHMENT 2 ITS 3.1.2, CORE REACTIVITY , Volume 6, Rev. 0, Page 25 of 213

, Volume 6, Rev. 0, Page 26 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 26 of 213

Attachment 1, Volume 6, Rev. 0, Page 27 of 213 ITS A01 ITS 3.1.2 4.9 REACTIVITY ANOMALIES APPLICABILITY Applies to potential reactivity anomalies.

OBJECTIVE To require evaluation of reactivity anomalies within the reactor.

SPECIFICATION A02 Add proposed Applicability Following a normalization of the computed boron concentration as a function of burnup, the M01 SR 3.1.2.1 actual boron concentration of the coolant shall be periodically compared with the predicted value. If the difference between the observed and predicted steady-state concentrations LCO 3.1.2 reaches the equivalent of 1% in reactivity, an evaluation as to the cause of the discrepancy shall be made and reported to the Commission within 30 days.

Add proposed ACTION A and B M02 Amendment No. 122 TS 4.9-1 12/21/95 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 27 of 213

Attachment 1, Volume 6, Rev. 0, Page 28 of 213 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 4.9 requires a comparison of the actual boron concentration with the predicted (computer) value, and requires the difference between the computed boron concentration, as a function of burnup, and the actual boron concentration to be +/- 1% k/k. ITS 3.1.2 also requires the comparison of the actual boron concentration with the predicted value and requires the difference to be within

+/- 1% k/k, but specifically states that the Applicability is in MODES 1 and 2.

This changes the CTS by stating the specific MODES.

This change is acceptable because the required MODES have not changed.

When full power is initially reached, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted concentration and the slope of the actual curve relating burn-up and reactivity is compared with that of the predicted curve. Therefore, since the comparison of the actual boron concentration to the predicted concentration and the computed boron concentration is a function of burnup (which can only occur when the reactor is critical) it can only be performed in MODES 1 and 2. This change is considered administrative because the technical requirements have not changed.

MORE RESTRICTIVE CHANGES M01 CTS 4.9 states, in part, that after normalization, the actual boron concentration of the coolant shall be periodically compared with the predicted value. ITS SR 3.1.2.1 requires verification of the measured core reactivity is within

+/- 1% k/k of predicted values at least once prior to entering MODE 1 after each refueling and after 60 effective full power days (EFPD) it should be verified every 31 EFPD thereafter. This changes the CTS by specifying Frequencies for verifying measured core reactivity is within +/- 1% k/k of predicted values.

This change is acceptable because it requires a verification that core reactivity is within +/- 1% k/k of predicted values thus assuring agreement between the actual core design and the core design predictions prior to entering MODE 1 after each refueling, and every 31 EFPD after 60 EFPD. This verification provides additional confidence that the core design is acceptable for operation at full power. This change is designated as more restrictive because it adds Surveillance Requirement frequencies that do not appear in the CTS.

M02 CTS 4.9 requires that if the difference between the observed and predicted steady-state concentrations reach the equivalent of 1% in reactivity, that an Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 28 of 213

Attachment 1, Volume 6, Rev. 0, Page 29 of 213 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY evaluation as to the cause of the discrepancy shall be made and reported to the Commission within 30 days. ITS 3.1.2 requires that if the measured core reactivity is not within limits, within 7 days a re-evaluation of the core design and safety analysis, a determination that the reactor core is acceptable for continued operation, and establishment of appropriate operating restrictions and SRs are required. If the re-evaluation of the core design and safety analysis to determine that the reactor core is acceptable for continued operation, and the establishment of appropriate operating restrictions and SRs can not be accomplished within 7 days, then the reactor is required to be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This changes the CTS by requiring a re-evaluation of the core design and safety analysis, a determination that the reactor core is acceptable for continued operation, and establishment of appropriate operating restrictions and SRs instead of requiring an evaluation of the cause of the discrepancy shall be made and reported to the Commission. This also changes the Completion Time from 30 days to 7 days.

This change is acceptable because the proposed Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing an appropriate amount of time to ensure continued safe operation and re-evaluate the core design and safety analyses. The Required Actions are consistent with safe operation under the specified Condition, considering the operability status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the repair period. Should the core reactivity balance requirement not be met, time is required to determine the cause of the disagreement and what, if any, adjustments are needed to the operating conditions of the core. The startup physics testing program is used to verify most of the critical core design parameters, such as control rod worth, boron worth, and moderator temperature coefficient. In addition, there is considerable conservatism in the application of these values in the accident analysis. Therefore, allowing a time to evaluate the difference and make any adjustments to the operational controls is acceptable.

The 7 day Completion time is reasonable considering the complexity of the evaluations and the time to meet administrative requirements, such as 10 CFR 50.59 safety evaluation preparation and approval. If it cannot be determined within 7 days that the core is acceptable for continued operation, the unit must be shutdown. This change is designated as more restrictive because more stringent Required Actions are being applied in the ITS than were applied in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 29 of 213

Attachment 1, Volume 6, Rev. 0, Page 30 of 213 DISCUSSION OF CHANGES ITS 3.1.2, CORE REACTIVITY LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 30 of 213

Attachment 1, Volume 6, Rev. 0, Page 31 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 31 of 213

Attachment 1, Volume 6, Rev. 0, Page 32 of 213 CTS Core Reactivity 3.1.2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Core Reactivity 4.9 LCO 3.1.2 The measured core reactivity shall be within +/- 1% k/k of predicted values.

DOC A02 APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 4.9 A. Measured core reactivity A.1 Re-evaluate core design 7 days not within limit. and safety analysis, and determine that the reactor core is acceptable for continued operation.

AND A.2 Establish appropriate 7 days operating restrictions and SRs.

4.9 B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

WOG STS 3.1.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 32 of 213

Attachment 1, Volume 6, Rev. 0, Page 33 of 213 CTS Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.9 SR 3.1.2.1 ---------------------------NOTE----------------------------------

The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.

Verify measured core reactivity is within +/- 1% k/k Once prior to of predicted values. entering MODE 1 after each refueling AND

NOTE--------

Only required after 60 EFPD 31 EFPD thereafter WOG STS 3.1.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 33 of 213

Attachment 1, Volume 6, Rev. 0, Page 34 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.2, CORE REACTIVITY None Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 34 of 213

Attachment 1, Volume 6, Rev. 0, Page 35 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 35 of 213

Attachment 1, Volume 6, Rev. 0, Page 36 of 213 Core Reactivity B 3.1.2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Core Reactivity BASES BACKGROUND According to GDC 26, GDC 28, and GDC 29 (Ref. 1), reactivity shall be INSERT 1 controllable, such that subcriticality is maintained under cold conditions, 1 and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity balance is used as a measure of the predicted versus measured core reactivity during power operation. The periodic confirmation of core reactivity is necessary to ensure that Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity difference could be the result of unanticipated changes in fuel, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus (which includes reactor measured core reactivity validates the nuclear methods used in the safety 6 safety analyses not in analysis and supports the SDM demonstrations (LCO 3.1.1, USAR Chapter 14)

"SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, subcritical conditions.

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

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

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

In order to achieve the required fuel cycle energy output, the uranium enrichment, in the new fuel loading and in the fuel remaining from the previous cycle, provides excess positive reactivity beyond that required to sustain steady state operation throughout the cycle. When the reactor is WOG B 3.1.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 36 of 213

Attachment 1, Volume 6, Rev. 0, Page 37 of 213 B 3.1.2 1

INSERT 1 Two independent reactivity control systems, preferably from different principles, shall be provided (Ref. 1). The reactivity control systems provided shall be capable of making and holding the core sub-critical from any hot standby or hot operation condition (Ref. 2). One of the reactivity control systems provided shall be capable of making the core sub-critical under any anticipated operating condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margin should assure sub-criticality with the most reactive control rod fully withdrawn (Ref. 3). The reactivity control system provided shall be capable of making the core sub-critical under credible accident conditions with appropriate margins for contingencies and limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public (Ref. 4).

Insert Page B 3.1.2-1 Attachment 1, Volume 6, Rev. 0, Page 37 of 213

Attachment 1, Volume 6, Rev. 0, Page 38 of 213 Core Reactivity B 3.1.2 BASES BACKGROUND (continued) critical at RTP and moderator temperature, the excess positive reactivity is compensated by burnable absorbers (if any), control rods, whatever neutron poisons (mainly xenon and samarium) are present in the fuel, and the RCS boron concentration.

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

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

Accurate prediction of core reactivity is either an explicit or implicit 5

assumption in the accident analysis evaluations. Every accident evaluation (Ref. 2) is, therefore, dependent upon accurate evaluation of 1 core reactivity. In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod ejection accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity balance additionally ensures that the nuclear methods provide an accurate representation of the core reactivity.

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

The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted RCS boron concentrations for identical core conditions at beginning of cycle (BOC) do not agree, then the assumptions used in the reload cycle design analysis or the calculational models used to predict soluble boron requirements may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to WOG B 3.1.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 38 of 213

Attachment 1, Volume 6, Rev. 0, Page 39 of 213 Core Reactivity B 3.1.2 BASES APPLICABLE SAFETY ANALYSES (continued) the measured boron concentration. Thereafter, any significant deviations in the measured boron concentration from the predicted boron letdown curve that develop during fuel depletion may be an indication that the calculational model is not adequate for core burnups beyond BOC, or that an unexpected change in core conditions has occurred.

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

Core reactivity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

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

When measured core reactivity is within 1% k/k of the predicted value at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Since deviations from the limit are normally detected by comparing predicted and measured steady state 140 RCS critical boron concentrations, the difference between measured and 3

predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached. These values are well within the 3

uncertainty limits for analysis of boron concentration samples, so that much larger than spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely.

APPLICABILITY The limits on core reactivity must be maintained during MODES 1 and 2 because a reactivity balance must exist when the reactor is critical or producing THERMAL POWER. As the fuel depletes, core conditions are changing, and confirmation of the reactivity balance ensures the core is operating as designed. This Specification does not apply in MODES 3, 4, and 5 because the reactor is shut down and the reactivity balance is not changing.

WOG B 3.1.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 39 of 213

Attachment 1, Volume 6, Rev. 0, Page 40 of 213 Core Reactivity B 3.1.2 BASES APPLICABILITY (continued)

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

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

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

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

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

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

WOG B 3.1.2-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 40 of 213

Attachment 1, Volume 6, Rev. 0, Page 41 of 213 Core Reactivity B 3.1.2 BASES ACTIONS (continued)

B.1 If any Required Action and associated Completion Time is not met 2

If the core reactivity cannot be restored to within the 1% k/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the SDM for MODE 3 is not met, then the boration required by 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 plant systems.

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS 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 control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization of predicted core reactivity to the 5

may measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1, is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (QPTR, AFD, etc.) for prompt indication of an anomaly.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29. 1 USAR, Section 3.1.2.3, GDC 27, "Redundancy of Reactivity Control."

5

2. FSAR, Chapter [15]. 3 4 U 14
2. USAR, Section 3.1.2.4, GDC 28, "Reactivity Hot Shutdown Capability."
3. USAR, Section 3.1.2.5, GDC 29, "Reactivity Shutdown Capability."
4. USAR, Section 3.1.2.6, GDC 30, "Reactivity Holddown Capability."

WOG B 3.1.2-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 41 of 213

Attachment 1, Volume 6, Rev. 0, Page 42 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.2 BASES, CORE REACTIVITY

1. ISTS 3.1.2 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.

Therefore, ITS 3.1.2 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.

Subsequent reference numbers have been renumbered.

2. ITS 3.1.2 ACTION B requires that when the Required Action and associated Completion Time of Condition A is not met, to be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ISTS 3.1.2 Bases statement for ACTION B.1 which states "if the core reactivity cannot be restored to within the 1% k/k limit" has been changed to match the wording contained in Condition B. Furthermore, ACTION A contains two Required Actions, and the "restore" words in the ISTS Bases are not consistent with either of the Required Actions. Therefore, changes have been made to be consistent with the Specification.
3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
4. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
5. Change made to be consistent with the actual wording of the Note in ISTS SR 3.1.2.1. Also the proposed wording (use of "may") is consistent with a similar description concerning normalization in the Applicable Safety Analyses section of the Bases.
6. Clarification has been added since all the reactor safety analyses are not in USAR Chapter 14.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 42 of 213

Attachment 1, Volume 6, Rev. 0, Page 43 of 213 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 6, Rev. 0, Page 43 of 213

Attachment 1, Volume 6, Rev. 0, Page 44 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.2, CORE REACTIVITY There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 44 of 213

Attachment 1, Volume 6, Rev. 0, Page 45 of 213 ATTACHMENT 3 ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)

Attachment 1, Volume 6, Rev. 0, Page 45 of 213

, Volume 6, Rev. 0, Page 46 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 46 of 213

Attachment 1, Volume 6, Rev. 0, Page 47 of 213 ITS A01 ITS 3.1.3

f. Minimum Conditions for Criticality
1. The reactor shall not be brought to a critical condition until the pressure-temperature See ITS state is to the right of the criticality limit line shown in Figure TS 3.1-1. 3.4.3
2. The reactor shall be maintained subcritical by at least 1% k/k until normal water See ITS level is established in the pressurizer. 3.4.9 Applicability M01
3. When the reactor is critical the moderator temperature coefficient shall be as specified in the COLR, except during LOW POWER PHYSICS TESTING. The See ITS 3.1.8 LCO 3.1.3 maximum upper moderator temperature coefficient limit shall be 5 pcm/°F for power levels 60% RATED POWER and 0 pcm/°F for power levels > 60% RATED POWER.
4. If the limits of 3.1.f.3 cannot be met, then power operation may continue provided the following actions are taken:

ACTION A A. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, develop and maintain administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in TS 3.1.f.3. These withdrawal limits shall be in addition to the LA01 insertion limits specified in TS 3.10.d.

ACTION B B. If the actions specified in TS 3.1.f.4.A are not satisfied, then be in HOT M02 STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Add proposed ACTION C M01 Add proposed SR 3.1.3.1 and SR 3.1.3.2 M03 Amendment No. 165 TS 3.1-10 03/11/2003 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 47 of 213

Attachment 1, Volume 6, Rev. 0, Page 48 of 213 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)

ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev. 3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.f.3 requires, in part, that the moderator temperature coefficient (MTC) shall be as specified in the COLR when the reactor is critical. The COLR provides both an upper MTC limit and a lower MTC limit. If an MTC limit is not met, CTS 3.1.f.4 provides 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to develop and maintain administrative control rod withdrawal limits sufficient to restore the MTC to within the limits.

ITS 3.1.3 requires the upper MTC Limit to be met in MODE 1 and MODE 2 with keff 1.0 and the lower MTC limit to be met in MODES 1, 2, and 3. Furthermore, ITS 3.1.3 does not provide any time to restore the lower MTC to within limits. If the lower MTC limit is not met, ITS 3.1.3 ACTION C requires the unit to be in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, which places the unit outside the new proposed Applicability. This changes the CTS by requiring the lower MTC limit to be met in MODE 3 and when it cannot be met, to be in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This change is acceptable because the lower MTC limit is required in MODE 3, in addition to MODE 2, to ensure that core overcooling accidents will not violate the assumptions in the safety analyses. This change is designated as more restrictive because it expands the Applicability for the lower MTC limit and requires an immediate shutdown to MODE 4 in lieu of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore MTC to within its lower limit prior to requiring a unit shutdown.

M02 CTS 3.10.f.4.B requires that if the development and maintenance of administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in the COLR can not be satisfied within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then to be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The reactivity condition (k/k) for HOT STANDBY, as defined in CTS 1.0.j, is

< 0.25%. Under similar conditions (i.e., if the establishment of administrative withdrawal limits for control banks to maintain MTC within limit is not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) ITS 3.1.3 ACTION B requires the unit to be in MODE 2 with keff < 1.0 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This changes the CTS by requiring the unit to be in MODE 2 with keff < 1.0 (i.e., k/k < 0.0%) in lieu of k/k < 0.25%.

The purpose of CTS 3.10.f.4.B is to place the unit in a condition in which it does not rely on the MTC upper limit. In ITS 3.1.3, it is necessary to place the unit in MODE 2 with keff < 1.0. At this point, the unit will be outside of the MODE of Applicability for upper MTC limit (MODE 1 and MODE 2 with keff 1.0). This change is designated as more restrictive because the ITS requires the unit to be Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 48 of 213

Attachment 1, Volume 6, Rev. 0, Page 49 of 213 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC) made subcritical (i.e., MODE 2 with keff < 1.0, which is the same as k/k < 0.0%)

in lieu of the CTS requirement to just be in HOT STANDBY (k/k < 0.25%).

M03 CTS 3.1.f.3 does not provide any Surveillance Requirements for verifying that the MTC is within the limits specified in the COLR. ITS SR 3.1.3.1 requires verification that the MTC is within the upper limit prior to entering MODE 1 after each refueling. ITS SR 3.1.3.2 requires that the MTC is within the lower limit once each cycle. This changes the CTS by adding new Surveillance Requirements to verify the upper and lower MTC are within limits.

This change is acceptable because the added Surveillance Requirements verify that the MTC requirements will be met in MODE 1 and MODE 2 with keff 1.0 for the upper limit and MODES 1, 2, and 3 for the lower limit. This change is designated as more restrictive because new Surveillance Requirements are being added to the ITS that are not included in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.10.f.4.A requires, in part, the development and maintenance of administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in the COLR.

Additionally, CTS 3.10.f.4.A states that the withdrawal limits shall be in addition to the insertion limits specified in rod insertion limits. ITS 3.1.3 Required Action A.1 requires the establishment of withdrawal limits for control banks to maintain MTC within limits. This changes the CTS by moving the statement that these withdrawal limits shall be in addition to the insertion limits specified in rod insertion limits to the Bases.

The removal of these details for meeting the Required Action from the Technical Specification is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains the requirement to establish withdrawal limits for control banks to maintain MTC within limits. Also, this change is acceptable because these type of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This change is designated as less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 49 of 213

Attachment 1, Volume 6, Rev. 0, Page 50 of 213 DISCUSSION OF CHANGES ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)

LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 50 of 213

Attachment 1, Volume 6, Rev. 0, Page 51 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 51 of 213

Attachment 1, Volume 6, Rev. 0, Page 52 of 213 CTS MTC 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Moderator Temperature Coefficient (MTC) 3.1.f.3 LCO 3.1.3 The MTC shall be maintained within the limits specified in the COLR. The maximum upper limit shall be [ [ ] k/k°F at hot zero power] [that 1 specified in Figure 3.1.3-1].

5 pcm/ºF with THERMAL POWER 60% RTP and 0 pcm/ºF with THERMAL POWER > 60% RTP 3.1.f.3 APPLICABILITY: MODE 1 and MODE 2 with keff 1.0 for the upper MTC limit, MODES 1, 2, and 3 for the lower MTC limit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.f.4, A. MTC not within upper A.1 Establish administrative 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.1.f.4.A limit. withdrawal limits for control banks to maintain MTC within limit.

3.1.f.4.b B. Required Action and B.1 Be in MODE 2 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion keff < 1.0.

Time of Condition A not met.

DOC M01 C. MTC not within lower C.1 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M03 SR 3.1.3.1 Verify MTC is within upper limit. Prior to entering MODE 1 after each refueling WOG STS 3.1.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 52 of 213

Attachment 1, Volume 6, Rev. 0, Page 53 of 213 CTS MTC 3.1.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY DOC M03 SR 3.1.3.2 ---------------------------NOTES--------------------------------

1. Not required to be performed until 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.
2. If the MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle.
3. SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR.

Verify MTC is within lower limit. Once each cycle WOG STS 3.1.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 53 of 213

Attachment 1, Volume 6, Rev. 0, Page 54 of 213 MTC 3.1.3

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Moderator Temperature Coefficient Vs. Rated Thermal Power WOG STS 3.1.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 54 of 213

Attachment 1, Volume 6, Rev. 0, Page 55 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)

1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.

Additionally, since ISTS Figure 3.1.3-1 is a bracketed value in ISTS LCO 3.1.3 and it is not being used, ISTS Figure 3.1.3-1 is deleted.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 55 of 213

Attachment 1, Volume 6, Rev. 0, Page 56 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 56 of 213

Attachment 1, Volume 6, Rev. 0, Page 57 of 213 MTC B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Moderator Temperature Coefficient (MTC)

BASES BACKGROUND According to GDC 11 (Ref. 1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable 1

INSERT 1 power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.

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

MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by The measurements. Both initial and reload cores are designed so that the 2 beginning of cycle (BOC) MTC is less than zero when THERMAL POWER is at RTP. The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons to yield an MTC at BOC within the range analyzed in the plant accident analysis. The end of cycle (EOC) MTC is also limited by the requirements of the accident analysis. Fuel cycles that are designed to 2 achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the EOC limit.

The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed in the FSAR 2 accident and transient analyses. U WOG STS B 3.1.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 57 of 213

Attachment 1, Volume 6, Rev. 0, Page 58 of 213 B 3.1.3 o 1

-INSERT 1 The design of the reactor core with its related controls and protection systems shall ensure that power oscillations, the magnitude of which could cause damage in excess of acceptable fuel damage limits, are not possible or can be readily suppressed (Ref. 1).

Insert Page B 3.1.3-1 Attachment 1, Volume 6, Rev. 0, Page 58 of 213

Attachment 1, Volume 6, Rev. 0, Page 59 of 213 MTC B 3.1.3 BASES BACKGROUND (continued)

If the LCO limits are not met, the unit response during transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity.

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

APPLICABLE The acceptance criteria for the specified MTC are:

SAFETY ANALYSES a. The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2) and 3

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

U 14 The FSAR, Chapter 15 (Ref. 2), contains analyses of accidents that result 2 in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref. 3).

The consequences of accidents that cause core overheating must be evaluated when the MTC is positive. Such accidents include the rod withdrawal transient from either zero (Ref. 4) or RTP, loss of main normal 2 feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature.

WOG STS B 3.1.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 59 of 213

Attachment 1, Volume 6, Rev. 0, Page 60 of 213 MTC B 3.1.3 BASES APPLICABLE SAFETY ANALYSES (continued)

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

MTC values are bounded in reload safety evaluations assuming steady state conditions at BOC and EOC. An EOC measurement is conducted at conditions when the RCS boron concentration reaches approximately 300 ppm. The measured value may be extrapolated to project the EOC value, in order to confirm reload design predictions.

MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not 4 directly observed and controlled from the control room, MTC is considered an initial condition process variable because of its dependence on boron concentration.

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

Assumptions made in safety analyses require that the MTC be less positive than a given upper bound and more positive than a given lower near 2 bound. The MTC is most positive at BOC; this upper bound must not be exceeded. This maximum upper limit occurs at BOC, all rods out (ARO),

hot zero power conditions. At EOC the MTC takes on its most negative value, when the lower bound becomes important. This LCO exists to ensure that both the upper and lower bounds are not exceeded.

are Core models provide additionally predicted values of MTC During operation, therefore, the conditions of the LCO can only be 2 that validate the upper and lower MTC limits in ensured through measurement. The Surveillance checks at BOC and the reload safety EOC on MTC provide confirmation that the MTC is behaving as evaluation process for each core reload. The anticipated so that the acceptance criteria are met.

WOG STS B 3.1.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 60 of 213

Attachment 1, Volume 6, Rev. 0, Page 61 of 213 MTC B 3.1.3 BASES LCO (continued)

(upper)

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

The BOC positive limit and the EOC negative limit are established in the 5 (lower)

COLR to allow specifying limits for each particular cycle. This permits the unit to take advantage of improved fuel management and changes in unit operating schedule.

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

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

ACTIONS A.1 upper These withdrawal limits shall If the BOC MTC limit is violated, administrative withdrawal limits for 5 be in addition to the insertion limits specified in LCO 3.1.6, control banks must be established to maintain the MTC within its limits. 6 "Control Bank Insertion The MTC becomes more negative with control bank insertion and Limits."

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

As cycle burnup is increased, the RCS boron concentration will be reduced. The reduced boron concentration causes the MTC to become more negative. Using physics calculations, the time in cycle life at which the calculated MTC will meet the LCO requirement can be determined.

At this point in core life Condition A no longer exists. The unit is no longer in the Required Action, so the administrative withdrawal limits are no longer in effect.

WOG STS B 3.1.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 61 of 213

Attachment 1, Volume 6, Rev. 0, Page 62 of 213 MTC B 3.1.3 BASES ACTIONS (continued)

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

The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

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

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

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

upper The BOC MTC value for ARO will be inferred from isothermal 5 temperature coefficient measurements obtained during the physics tests 2

after refueling. The ARO value can be directly compared to the BOC MTC limit of the LCO. If required, measurement results and predicted design values can be used to establish administrative withdrawal limits for control banks.

WOG STS B 3.1.3-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 62 of 213

Attachment 1, Volume 6, Rev. 0, Page 63 of 213 MTC B 3.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

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

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

a. The SR is not required to be performed until 7 effective full power days (EFPDs) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.
b. If the 300 ppm Surveillance limit is exceeded, it is possible that the EOC limit on MTC could be reached before the planned EOC.

Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the EOC limit.

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

REFERENCES 1. 10 CFR 50, Appendix A, GDC 11. 3 U USAR, Section 3.1.2.2, GDC 7, 14 "Suppression of Power Oscillations."

2. FSAR, Chapter [15]. 3 7 9272-P-A
3. WCAP 9273-NP-A, "Westinghouse Reload Safety Evaluation 3 Methodology," July 1985.

U Section 14.1.1 3 7

4. FSAR, Chapter [15].

WOG B 3.1.3-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 63 of 213

Attachment 1, Volume 6, Rev. 0, Page 64 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.3 BASES, MODERATOR TEMPERATURE COEFFICIENT (MTC)

1. ISTS 3.1.3 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.

Therefore, ITS 3.1.3 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
4. The Applicable Safety Analyses discussion states that MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). It also says that even though MTC is not directly observed and controlled from the control room, MTC is still considered an initial condition process variable because of its dependence on boron concentration. The additional sentence has been deleted. The NRC Final Policy Statement on Technical Improvements of July 22, 1993 (58 FR 39132) states that process variables captured by Criterion 2 are not limited to only those directly monitored and controlled from the control room. It also states that Criterion 2 includes other features or characteristics that are specifically assumed in Design Basis Accident and Transient analyses even if they cannot be directly observed in the control room (e.g., moderator temperature coefficient and hot channel factors). Since the Final Policy Statement provides guidance on which types of parameters satisfy Criterion 2, there is no reason to duplicate these words in the KPS CTS.
5. The ISTS Bases variously refer to the "upper MTC limit," the "BOC MTC limit," the "lower MTC limit," and the "EOC MTC limit." References to the BOC and EOC MTC limit are eliminated and "upper" and "lower" are substituted to eliminate confusion and to be consistent with the Specification.
6. This requirement has been added to the Bases consistent with the requirement in CTS 3.10.f.4.A. This clarifies the ITS 3.1.3 Required Action A.1 requirement to ensure the administrative withdrawal limits are in addition to the normal limits required by LCO 3.1.6, "Control Bank Insertion Limits."
7. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 64 of 213

Attachment 1, Volume 6, Rev. 0, Page 65 of 213 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 6, Rev. 0, Page 65 of 213

Attachment 1, Volume 6, Rev. 0, Page 66 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.3, MODERATOR TEMPERATURE COEFFICIENT (MTC)

There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 66 of 213

, Volume 6, Rev. 0, Page 67 of 213 ATTACHMENT 4 ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS , Volume 6, Rev. 0, Page 67 of 213

, Volume 6, Rev. 0, Page 68 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 68 of 213

Attachment 1, Volume 6, Rev. 0, Page 69 of 213 ITS ITS 3.1.4 A01

e. Rod Misalignment Limitations LCO 3.1.4 NOTE: Individual RPIs may be outside their limits for < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following substantial rod movement Note This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog LA01 individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.

Rod misalignment limitations do not apply during physics testing. See ITS 3.1.8 LCO 3.1.4 1. When reactor power is 85% of rating, the rod cluster control assembly shall be part a maintained within +/- 12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is ACTION B 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If peaking factors are not M01 determined within four hours, the reactor power shall be reduced to < 85% of rating.

M02 Add proposed Required Actions B.1.1, B.1.2, and B.2 and associated Completion Times LCO 3.1.4 2. When reactor power is < 85% but 50% of rating, the rod cluster control assemblies Add part b shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control proposed Applicability ACTION C assembly is misaligned from its bank by more than +/- 24 steps when reactor power is

< 85% but 50%, the rod will be realigned or the core power peaking factors shall be M01 determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating.

Add proposed Required Actions C.1.1, C.1.2, C.2, C.3, C.4, and C.5 and associated Completion Times M02 ACTION D

3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.

Add proposed Required Action D.1 M03

f. Inoperable Rod Position Indicator Channels Add proposed ACTION E M03 LCO 3.1.4 NOTE: Individual RPIs may be outside their limits for < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following substantial rod Note movement See ITS 3.1.7
1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate entry condition is allowed for each inoperable individual rod position indicator.)

A. Verify the position of the rod cluster control indirectly by movable incore detectors See ITS each eight hours, or 3.1.7 B. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reduce reactor thermal power to 50% of RATED POWER.

2. If more than one individual rod position indicator channel per group are inoperable, then:

A. IMMEDIATELY place the control rods in manual, and B. Once per 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, monitor and record RCS Tavg, and C. Verify the position of the rod by movable incore detectors each 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and Amendment No. 181 TS 3.10-5 03/24/2005 Page 1 of 5 Attachment 1, Volume 6, Rev. 0, Page 69 of 213

Attachment 1, Volume 6, Rev. 0, Page 70 of 213 ITS ITS 3.1.4 A01

g. Inoperable Rod Limitations Add proposed LCO 3.1.4 and Applicability M01
1. An inoperable rod is a rod which does not trip or which is declared inoperable under LA02 TS 3.10.e or TS 3.10.h.
2. Not more than one inoperable full length rod shall be allowed at any time. M04
3. If reactor operation is continued with one inoperable full length rod, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days unless the rod is made OPERABLE earlier. The ACTION A analysis shall include due allowance for nonuniform fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the plant power level shall be reduced to an analytically determined part power level which is consistent with the safety analysis.

Add proposed Required Actions A.1.1, A.1.2, and A.2 M04

h. Rod Drop Time Add proposed LCO 3.1.4 and Applicability M01 At OPERATING temperature and full flow, the drop time of each full length rod cluster SR 3.1.4.3 control shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to dashpot entry. If drop time is > 1.8 seconds, the rod shall be declared inoperable.

ACTION A

i. Rod Position Deviation Monitor If the rod position deviation monitor is inoperable, individual rod positions shall be L01 logged at least once per eight hours and after a load change > 10% of rated power or after > 24 steps of control rod motion.
j. Quadrant Power Tilt Monitor If one or both of the quadrant power tilt monitors is inoperable, individual upper and See ITS lower excore detector calibrated outputs and the quadrant tilt shall be logged once per 3.2.4 shift and after a load change > 10% of rated power or after > 24 steps of control rod motion. The monitors shall be set to alarm at 2% tilt ratio.
k. Core Average Temperature During steady-state power operation, Tavg, shall be maintained within the limits specified in the COLR, except as provided by TS 3.10.n.

See ITS 3.4.1

l. Reactor Coolant System Pressure During steady-state power operation, Reactor Coolant System pressure shall be maintained within the limits specified in the COLR, except as provided by TS 3.10.n.

Amendment No. 181 TS 3.10-7 Revised by letter dated 11/03/06 Page 2 of 5 Attachment 1, Volume 6, Rev. 0, Page 70 of 213

Attachment 1, Volume 6, Rev. 0, Page 71 of 213 ITS ITS 3.1.4 A01 4.1 OPERATIONAL SAFETY REVIEW APPLICABILITY Applies to items directly related to safety limits and LIMITING CONDITIONS FOR OPERATION.

OBJECTIVE To assure that instrumentation shall be checked, tested, and calibrated, and that equipment and sampling tests shall be conducted at sufficiently frequent intervals to ensure safe operation.

SPECIFICATION

a. Calibration, testing, and checking of protective instrumentation channels and testing of See other logic channels shall be performed as specified in Table TS 4.1-1. ITS SR 3.1.4.2, b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and SR 3.1.4.3 TS 4.1-3.
c. Deleted
d. Deleted
e. Deleted Amendment No. 119 TS 4.1-1 04/18/95 Page 3 of 5 Attachment 1, Volume 6, Rev. 0, Page 71 of 213

ITS A01 ITS 3.1.4 TABLE TS 4.1-1 See ITS 3.3.5 See ITS MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS 3.3.1 and 3.3.2 CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

6. Pressurizer Water Level Each shift Each refueling cycle Monthly
7. Pressurizer Pressure Each shift Each refueling cycle Monthly
8. a. 4-KV Voltage and Not applicable Each refueling cycle Monthly Reactor protection circuits only Frequency
b. 4-KV Voltage Not applicable Each refueling cycle Monthly Safeguards buses only (Loss of Voltage)
c. 4-KV Voltage Not applicable Each refueling cycle Monthly Safeguards buses only (Degraded Grid)
9. Analog Rod Position Each shift(a,b) Each refueling cycle Each refueling (a) With step counters cycle (b) Following rod motion in excess of 24 steps when computer is out of service
10. Rod Position Bank Each shift(a,b) Not applicable Each refueling (a) With analog rod position Counters cycle (b) Following rod motion in excess of 24 steps when computer is out of service SR 3.1.4.1 See ITS L02 L02 L02 3.1.7 , Volume 6, Rev. 0, Page 72 of 213 Attachment 1, Volume 6, Rev. 0, Page 72 of 213 Amendment No. 151 Page 2 of 7 02/12/2001 Page 4 of 5

Attachment 1, Volume 6, Rev. 0, Page 73 of 213 ITS TABLE TS 4.1-3 A01 ITS 3.1.4 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS L03 EQUIPMENT TESTS(1) TEST FREQUENCY

1. Control Rods Rod drop times of all full Each REFUELING outage SR 3.1.4.3 length rods Quarterly when at or above HOT Partial movement of all STANDBY SR 3.1.4.2 rods not fully inserted in the core 1a. Reactor Trip Breakers Independent test(2) shunt Monthly and undervoltage trip attachments See ITS 1b. Reactor Coolant Pump Breakers- OPERABILITY Each REFUELING outage 3.3.1 Open-Reactor Trip 1c. Manual Reactor Trip Open trip reactor(3) trip and Each REFUELING outage bypass breaker
2. Deleted
3. Deleted See ITS
4. Containment Isolation Trip OPERABILITY Each REFUELING outage 3.6.3
5. Refueling System Interlocks OPERABILITY Prior to fuel movement each See CTS REFUELING outage 3.8.a.11
6. Deleted See ITS
7. Deleted 3.4.15 See ITS
8. RCS Leak Detection OPERABILITY Weekly(4) 3.8.1 and
9. Diesel Fuel Supply Fuel Inventory(5) Weekly 3.8.3
10. Deleted See ITS 4.0
11. Fuel Assemblies Visual Inspection Each REFUELING outage See ITS
12. Guard Pipes Visual Inspection Each REFUELING outage 3.6.1
13. Pressurizer PORVs OPERABILITY Each REFUELING cycle See ITS (6) 3.4.11
14. Pressurizer PORV Block Valves OPERABILITY Quarterly See ITS
15. Pressurizer Heaters OPERABILITY(7) Each REFUELING cycle 3.4.9
16. Containment Purge and Vent OPERABILITY(8) Each REFUELING cycle See ITS Isolation Valves 3.6.3 (1)

Following maintenance on equipment that could affect the operation of the equipment, tests L03 should be performed to verify OPERABILITY.

(2)

Verify OPERABILITY of the bypass breaker undervoltage trip attachment prior to placing breaker into service.

(3) See ITS Using the Control Room push-buttons, independently test the reactor trip breakers shunt trip 3.3.1 and undervoltage trip attachments. The test shall also verify the undervoltage trip attachment on the reactor trip bypass breakers.

(4)

When reactor is at power or in HOT SHUTDOWN condition. See ITS 3.4.15 (5) See ITS 3.8.1 Inventory of fuel required in all plant modes. and 3.8.3 (6)

Not required when valve is administratively closed. See ITS 3.4.11 (7)

Test will verify OPERABILITY of heaters and availability of an emergency power supply.

(8) See ITS This test shall demonstrate that the valve(s) close in 5 seconds. 3.4.9 See ITS Amendment No. 125 Page 1 of 1 3.6.3 08/07/96 Page 5 of 5 Attachment 1, Volume 6, Rev. 0, Page 73 of 213

Attachment 1, Volume 6, Rev. 0, Page 74 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.10.e.2 states, in part, that when reactor power is < 85% but 50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. CTS 3.10.g.2 states that not more than one inoperable full length rod shall be allowed at any time. CTS 3.10.h states, in part, that at OPERATING temperature and full flow, the drop time of each full length rod cluster control shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to dashpot entry. ITS 3.1.4 requires all shutdown and control rods and the individual rod position indication to be OPERABLE in MODES 1 and 2 and requires the 24 step alignment limit to be met when < 85% RTP in MODES 1 and 2. This changes the CTS by requiring all shutdown and control rods and the individual rod position indication to be OPERABLE in MODES 1 and 2 and requires the 24 step alignment limit to be met when < 85% RTP in MODES 1 and 2, in lieu of when < 85% RTP and > 50% RTP.

The purpose of CTS 3.10.e.1 and CTS 3.10.e.2 is to maintain the rods in alignment with their respective banks to provide consistency with the assumption of the safety analyses, to maintain symmetric neutron flux and power distribution profiles, to provide assurance that peaking factors are within acceptable limits, and to assure adequate shutdown margin. The purpose of CTS 3.10.g and 3.10.h are to maintain the potential consequence of accidents consistent with the safety analyses. The proposed change is acceptable because it now states that all shutdown and control rods are required to be OPERABLE in MODES 1 and 2 and that alignment limits must be met in MODES 1 and 2. Additionally, it requires that individual rod position indication, dependent on percent of RTP, to be OPERABLE in MODES 1 and 2. This change is designated as more restrictive because specific LCO requirements and MODES of Applicability are specified.

M02 CTS 3.10.e.1 state, in part, that if a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is > 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating. ITS 3.1.4 ACTION B states that with one rod not within alignment limits when THERMAL POWER is > 85% RTP, to verify SHUTDOWN MARGIN (SDM) or initiate boration to restore SDM to within limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and to reduce THERMAL POWER to < 85% RTP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. CTS 3.10.e.2 states, in part, that if a rod Kewaunee Power Station Page 1 of 6 Attachment 1, Volume 6, Rev. 0, Page 74 of 213

Attachment 1, Volume 6, Rev. 0, Page 75 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is < 85 % but > 50%, the rod will be realigned or the core power peaking factors shall be determined within four hours and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating. ITS 3.1.4 ACTION C states that with one rod not within alignment limits when THERMAL POWER is < 85% RTP, to verify SHUTDOWN MARGIN (SDM) or initiate boration to restore SDM to within limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, to reduce THERMAL POWER to < 50% RTP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, to verify SDM is within the limits specified in the COLR once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, to perform SR 3.2.1.1 and SR 3.2.1.2 (verification that FQ(Z) is within the required limits) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, to perform SR 3.2.2.1 (verification that F N H is within the required limits) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and to re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions within 5 days. This changes the CTS by requiring the verification of SDM or initiation of boration to restore SDM to within limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, reduction of THERMAL POWER to < 85% RTP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, verification of SDM is within the limits specified in the COLR once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the performance SR 3.2.1.1 and SR 3.2.1.2 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the performance of SR 3.2.2.1within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the re-evaluation of the safety analyses and confirmation that the results remain valid for the duration of operation under these conditions within 5 days.

This change is acceptable because verification that the SDM is within its limits or the initiation of boration to restore the SDM to within limits is desirable. In contrast, it is not always desirable or possible to realign the misaligned control rod. Furthermore, if the misaligned rod continues to not meet the applicable alignment limits, in addition to the verification or restoration, the THERMAL POWER must be reduced (which in the case of the 12 step limit could result in meeting the alignment limits), periodic verification of the SDM must be accomplished along with the verification of FQ(Z) and F N H , and an evaluation must be performed to confirm that the accident analyses remain valid. All of these actions ensure that the core design criteria will not be exceeded.

Additionally, the Completion Times are commensurate with safe operation of the plant and provides sufficient time to accomplish the Required Actions. This change is designated as more restrictive because more stringent Required Actions and shorter Completion Times have been added that were not in the CTS.

M03 CTS 3.10.e.3 states that in addition to TS 3.10.e.1 (CTS 3.10.e.1) and TS 3.10.e.2 (CTS 3.10.e.2) if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable. CTS 3.10.e does not contain any requirements if 3.10.e.1 or 3.10.e.2 are not met. Therefore, CTS 3.0.c entry would be required. CTS 3.0.c requires action to be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and to be in HOT STANDBY (equivalent to ITS MODE 2) in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ITS 3.1.4 ACTION D states, that if the Required Action and associated Completion Time of Condition B and C are not met, then be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Additionally, CTS 3.10.e does not contain a requirement for when more than one rod is not within alignment limits. Therefore, CTS 3.0.c entry would also be required for this condition. ITS 3.1.4 ACTION E requires that if more than one rod is not within the alignment limit, then verify SDM is within the limits specified in the COLR or initiate boration to restore the required SDM to Kewaunee Power Station Page 2 of 6 Attachment 1, Volume 6, Rev. 0, Page 75 of 213

Attachment 1, Volume 6, Rev. 0, Page 76 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS within limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and to be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This changes the CTS by requiring the plant to be placed in MODE 3 if the Required Actions and associated Completion Time of B or C are not met (ITS 3.1.4 ACTION D) and adding a new ACTION (ITS 3.1.4 ACTION E).

This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Times.

Allowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to verify SDM is within the limits specified in the COLR or initiation of boration to restore the required SDM to within limits and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to be in MODE 3 ensures a unit shutdown is commenced and completed within a reasonable period of time upon failure to restore the SDM to within limits within the allowed Completion Times. This change is designated as more restrictive because the time to exit the Applicability of the LCO has been reduced.

M04 CTS 3.10.g.2 states that not more than one inoperable full length rod shall be allowed at any time. As a result, if more than one full length rod is inoperable, then LCO 3.0.c would be entered which requires action to be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to be in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. CTS 3.10.g.3 states, in part, that with one inoperable full length rod the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days.

ITS 3.1.4 requires that with one or more rods inoperable to verify the SDM is within the limits specified in the COLR or to initiate boration to restore the SDM to within limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and to be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This changes the CTS by requiring verification or restoration of the SDM and requiring the unit to be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The purpose of CTS 3.10.g.2 and 3.10.g.3 is to maintain the potential ejected rod worth and associated transient power distribution factor consistent with the safety analysis. This change is acceptable because it is not expected that more than one control rod will become misaligned from its group average position. When one or more control rods are misaligned, there is a potential for reduced SDM.

Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity. The required Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored. This change is acceptable because it is consistent with the requirements of the assumptions of the safety analyses to be within the SDM limit. Additionally, the requirement to be in MODE 2 is not required since the ITS requires the unit to be in MODE 3 in the same time (i.e., 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />). The change has been designated as more restrictive because the ITS requires verification or restoration of the SDM within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and the unit to be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in lieu of the CTS requirement to determine the potential ejected rod worth and associated transient Kewaunee Power Station Page 3 of 6 Attachment 1, Volume 6, Rev. 0, Page 76 of 213

Attachment 1, Volume 6, Rev. 0, Page 77 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS power distribution peaking factors within 30 days for one inoperable full length rod.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.10.e states, in part, that the magnitude of an indicated rod misalignment may be determined by comparison of the respective bank demand counter to the analog individual rod position indicator, the rod position on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage. ITS 3.1.4 does not contain this statement.

This changes the CTS by moving the description of how to determine the magnitude of the rod misalignment to the Bases.

The removal of these details which relate to the system design from the Technical Specification is acceptable because this type of information is not necessary to be included to provide adequate protection of public health and safety. This change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specifications Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because the information relating to system design is being removed from the Technical Specifications.

LA02 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.10.g.1 states, in part, that an inoperable rod is a rod which does not trip or which is declared inoperable. ITS 3.1.4 does not contain this statement. This changes the CTS by moving the detail describing an inoperable rod to the Bases.

The removal of these details which relate to the system design from the Technical Specifications is acceptable because this type of information is not necessary to be included to provide adequate protection of public health and safety. This change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specifications Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because the information relating to system design is being removed from the Technical Specifications.

Kewaunee Power Station Page 4 of 6 Attachment 1, Volume 6, Rev. 0, Page 77 of 213

Attachment 1, Volume 6, Rev. 0, Page 78 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS LESS RESTRICTIVE CHANGES L01 (Category 7 - Relaxation of Surveillance Frequency) CTS 3.10.i requires that if the rod position deviation monitor is inoperable, individual rod positions shall be logged at least once per eight hours and after a load change > 10% of rated power or after > 24 steps of control rod motion. ITS SR 3.1.4.1 requires verification that the individual rod positions are within the alignment limits every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This changes the CTS by eliminating the requirement to log the individual rod positions once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and after a load change > 10% of rated power or after > 24 steps of control rod motion when the Rod Position Deviation Monitor is inoperable.

The purpose of CTS 3.10.i is to periodically verify that the rods are within the alignment limits specified in the LCO. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Increasing the Frequency of rod position verification when the Rod Position Deviation Monitor is inoperable is unnecessary, since an inoperability of the alarm does not increase the probability that the rods are misaligned. The Rod Position Deviation Monitor alarm is for indication only. Its use is not credited in any safety analyses. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

L02 (Category 7 - Relaxation of Surveillance Frequency) Note (b) to CTS Table TS 4.1-1 Channel Description 9 requires an Analog Rod Position check following rod motion in excess of 24 steps when the computer is out of service. Note (b) to CTS TS Table 4.1-1 Channel Description 10 requires a Rod Position Bank Counters check following rod motion in excess of 24 steps when the computer is out of service. ITS SR 3.1.4.1 requires a verification that individual rod positions are within alignment every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This changes the CTS by not requiring an Analog Rod Position check and a Rod Position Bank Counters check following rod motion in excess of 24 steps when the computer is out of service.

The purpose of CTS Table TS 4.1-1 Channel Descriptions 9 and 10 Channel Checks is to verify that the rods are within the alignment limits. This change is acceptable because the verification of the rod position following rod motion in excess of 24 steps when the computer is out of service is unnecessary. An inoperable computer does not increase the probability that the rods are misaligned. The routine 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency (ITS SR 3.1.4.1) continues to ensure that the control rods are aligned properly. Furthermore, ITS 3.1.7 Required Action A.1 requires the position of a rod with an inoperable position indicator to be verified once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after moving the rod in excess of 24 steps. Thus, any response determined necessary by plant personnel due to an inoperable computer is more appropriately controlled by plant procedures, not Technical Specifications. This change is designated as less restrictive because Surveillances will be performed less frequently under ITS than under the CTS.

L03 (Category 5 - Deletion of Surveillance Requirement) Note 1 to CTS Table TS 4.1-3 requires, in part, that the Control Rods be tested to verify OPERABILITY following maintenance on equipment that could affect the Kewaunee Power Station Page 5 of 6 Attachment 1, Volume 6, Rev. 0, Page 78 of 213

Attachment 1, Volume 6, Rev. 0, Page 79 of 213 DISCUSSION OF CHANGES ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS operation. ITS 3.1.4 does not include this requirement. This changes the CTS by eliminating a post-maintenance Surveillance Requirement.

This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the equipment used to meet the LCO can perform its required functions. Thus, appropriate equipment continues to be tested in a manner and frequency necessary to give confidence that the equipment can perform its assumed safety function. Whenever, the OPERABILITY of a system or component has been affected by repair, maintenance, modification, or replacement of a component, post maintenance testing is required to demonstrate the OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under SR 3.0.1. In addition, the requirement of 10 CFR 50, Appendix B, Section XI (Test Control), provides adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria. Compliance with 10 CFR 50, Appendix B is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is designated as less restrictive because a Surveillance which is required in the CTS will not be performed in the ITS.

Kewaunee Power Station Page 6 of 6 Attachment 1, Volume 6, Rev. 0, Page 79 of 213

Attachment 1, Volume 6, Rev. 0, Page 80 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 80 of 213

Attachment 1, Volume 6, Rev. 0, Page 81 of 213 CTS Rod Group Alignment Limits 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits 3.10.g, LCO 3.1.4 All shutdown and control rods shall be OPERABLE.

3.10.h AND as follows:

3.10.e.1, Individual indicated rod positions shall be within 12 steps of their group 3.10.e.2 step counter demand position. 1 INSERT 1 DOC M01 APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME is 3.10.g.3, A. One or more rod(s) A.1.1 Verify SDM to be within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2 DOC M04 inoperable. limits specified in the COLR.

OR A.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limit. 3 s

AND A.2 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 4

3.10.e.1 B. One rod not within B.1 Restore rod to within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> alignment limits. alignment limits. 5 when THERMAL POWER is 85% RTP OR is B.2.1.1 Verify SDM to be within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 4 limits specified in the COLR.

OR 4 WOG STS 3.1.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 81 of 213

Attachment 1, Volume 6, Rev. 0, Page 82 of 213 CTS 3.1.4 1

INSERT 1 3.10.e.1 a. With THERMAL POWER 85% RTP, within 12 steps from their respective banks; and 3.10.e.2 b. With THERMAL POWER < 85% RTP, within 24 steps from their respective banks.

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

3.10.e Note Individual rod position indicators may be outside of their limits for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following substantial rod movement.

Insert Page 3.1.4-1 Attachment 1, Volume 6, Rev. 0, Page 82 of 213

Attachment 1, Volume 6, Rev. 0, Page 83 of 213 CTS Rod Group Alignment Limits 3.1.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.10.e.1 B.2.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 4 SDM to within limit. 3 s

AND 4 4

B.2.2 Reduce THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> POWER to 75% RTP. 5

< 85 AND B.2.3 Verify SDM is within the Once per limits specified in the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> COLR.

AND B.2.4 Perform SR 3.2.1.1 and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 5 SR 3.2.1.2.

AND B.2.5 Perform SR 3.2.2.1. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND B.2.6 Re-evaluate safety 5 days analyses and confirm results remain valid for duration of operation under these conditions.

INSERT 2 5 3.10.e.3 C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion D 5 D

Time of Condition B not met. or C DOC M03 D. More than one rod not D.1.1 Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 5 E

within alignment limit. E limits specified in the COLR.

OR WOG STS 3.1.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 83 of 213

Attachment 1, Volume 6, Rev. 0, Page 84 of 213 CTS 3.1.4 5

INSERT 2 3.10.e.2 C. One rod not within C.1.1 Verify SDM is within limits 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> alignment limits when specified in the COLR.

THERMAL POWER is

< 85% RTP. OR C.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limits.

AND C.2 Reduce THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> POWER to < 50% RTP.

AND C.3 Verify SDM is within limits Once per specified in the COLR. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND C.4 Perform SR 3.2.1.1 and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> SR 3.2.1.2.

AND C.5 Perform SR 3.2.2.1. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND C.6 Re-evaluate safety 5 days analyses and confirm results remain valid for duration of operation under these conditions.

Insert Page 3.1.4-2 Attachment 1, Volume 6, Rev. 0, Page 84 of 213

Attachment 1, Volume 6, Rev. 0, Page 85 of 213 CTS Rod Group Alignment Limits 3.1.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME DOC M03 D.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 5 E required SDM to within limit. 3 s

AND E

D.2 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Table 4.1-1 Channel SR 3.1.4.1 Verify individual rod positions within alignment limit. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Description 9 and 10 Table 4.1-3 Equipment SR 3.1.4.2 Verify rod freedom of movement (trippability) by 92 days Test 1 moving each rod not fully inserted in the core 10 steps in either direction.

3.10.h.1, SR 3.1.4.3 Verify rod drop time of each rod, from the fully Prior to criticality Table 4.1-3 1.8 Equipment withdrawn position, is [2.2] seconds from the after each 6 Test 1 beginning of decay of stationary gripper coil voltage removal of the to dashpot entry, with: reactor head

a. Tavg 500°F and
b. All reactor coolant pumps operating.

WOG STS 3.1.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 85 of 213

Attachment 1, Volume 6, Rev. 0, Page 86 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS

1. LCO 3.1.4 (2nd part) has been modified to incorporate a Kewaunee Power Station (KPS) specific allowance. This change to the LCO has been made consistent with the allowance in License Amendment 51, dated December 28, 1983 (ADAMS Accession Number ML020740166), with one exception. The CTS does not require alignment limits to be met when < 50% RTP, as shown in CTS 3.10.e.2. However, as part of converting to the ITS, KPS will require alignment limits to be met at all times when in MODES 1 and 2. That is, the 24 step limit that is currently required when < 85% and > 50% RTP will be applicable in MODE 1 when < 85% RTP and in MODE 2.
2. Changes are made to be consistent with the format of the ITS. The Required Action normally states that the parameter shall be "within limits."
3. Typographical/grammatical error corrected.
4. ISTS 3.1.4 Required Action B.1 requires restoration of a rod not within alignment limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or performance of a number of other actions, such as verification of SHUTDOWN MARGIN, reduction in reactor power, measurement of hot channel factors, and re-evaluation of the safety analyses. The Writer's Guide for Improved Standard Technical Specifications, TSTF-GG-05-01, Section 4.1.6.g states, "A Required Action which requires restoration, such that the Condition is no longer met, is considered superfluous. It is only included if it would be the only Required Action for the Condition or it is needed for presentation clarity." Neither exception applies in this case. In fact, the inclusion of Required Action B.1 requires an additional level of indenting and numbering for the remaining Required Actions in Condition B, which reduces its clarity.

Therefore, Required Action B.1 is deleted and the subsequent Required Actions renumbered.

5. ISTS 3.1.4 ACTION B has been split into two ACTIONS, based on the KPS specific rod group alignment limits. As discussed in JFD 1, LCO 3.1.4 (2nd part) has been modified to be consistent with KPS License Amendment 51. The KPS CTS requires the rod alignment to be within 12 steps when > 85% RTP, but only 24 steps when < 85% RTP. Therefore, the following changes have been made due to this difference:
a. ISTS 3.1.4 Condition B has been modified to only be applicable when THERMAL POWER is > 85% RTP. ITS 3.1.4 ACTION B will now provide the actions when the > 85% RTP alignment limit (12 steps) is not met. For ITS ACTION B, only ISTS 3.1.4 Required Actions B.2.1.1, B.2.1.2, and B.2.2 are being maintained (note that ISTS 3.1.4 Required Action B.1 has been deleted as discussed in JFD 4). ISTS 3.1.4 Required Action B.2.2 (ITS 3.1.4 Required Action B.2) has been modified to require reduction in THERMAL POWER to < 85% RTP. Thus, once power is reduced to < 85% RTP, the Condition B entry requirements are not met. Furthermore, once power is reduced, a new alignment limit, 24 steps, is applicable. Therefore, the follow-on Required Actions in the ISTS (ISTS 3.1.4 Required Actions B.2.3, B.2.4, B.2.5, and B.2.6) are not applicable and are not included in ITS 3.1.4 ACTION B.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 86 of 213

Attachment 1, Volume 6, Rev. 0, Page 87 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS

b. A new ACTION has been added (ITS 3.1.4 ACTION C) to cover not meeting the < 85% RTP alignment limit. ITS 3.1.4 Condition C covers the condition of one rod not within alignment limits when THERMAL POWER is < 85% RTP.

The proposed Required Actions for this Condition are consistent with the Required Actions of ISTS 3.1.4 ACTION B, except the restore Required Action is not included (as described in JFD 4) and the power reduction value in ITS 3.1.4 Required Action C.2 is < 50% RTP (in lieu of the ISTS value of

< 75% RTP), consistent with the current KPS CTS.

In addition, due to the above changes (the addition of new ACTION C),

subsequent Conditions and Required Actions have been renumbered.

6. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 87 of 213

Attachment 1, Volume 6, Rev. 0, Page 88 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 88 of 213

Attachment 1, Volume 6, Rev. 0, Page 89 of 213 Rod Group Alignment Limits B 3.1.4 B 3.3 INSTRUMENTATION B 3.1.4 Rod Group Alignment Limits BASES BACKGROUND The OPERABILITY (i.e., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," 1 INSERT 1 GDC 26, "Reactivity Control System Redundancy and Capability" (Ref. 1),

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

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

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

Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMs). Each CRDM moves its RCCA one step (approximately e inch) at a time, but at varying rates (steps per 2 minute) depending on the signal output from the Rod Control System.

The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. If a bank of RCCAs consists of two groups, the groups are moved in a staggered fashion, but always within one step of each other. All units have four control banks 2 and at least two shutdown banks. Kewaunee Power Station (KPS) has WOG STS B 3.1.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 89 of 213

Attachment 1, Volume 6, Rev. 0, Page 90 of 213 B 3.1.4 1

INSERT 1 USAR GDC 6, "Reactor Core Design," GDC 7, "Suppression of Power Oscillations," GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot Shutdown Capability," GDC 29, "Reactivity Shutdown Capability," and GDC 30, "Reactivity Holddown Capability" Insert Page B 3.1.4-1 Attachment 1, Volume 6, Rev. 0, Page 90 of 213

Attachment 1, Volume 6, Rev. 0, Page 91 of 213 Rod Group Alignment Limits B 3.1.4 BASES BACKGROUND (continued)

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

The axial position of shutdown rods and control rods is indicated by two separate and independent systems, which are the Bank Demand Position Indication System (commonly called group step counters) and the Digital 2 Rod Position Indication (DRPI) System. Individual I

The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal INSERT 2 to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position 5/8 2 Indication System is considered highly precise (+/- 1 step or +/- e inch). If a 7 rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.

I control The DRPI System provides a highly accurate indication of actual rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube. To increase the reliability of the system, the inductive coils 2 are connected alternately to data system A or B. Thus, if one data INSERT 3 system fails, the DRPI will go on half accuracy. The DRPI System is capable of monitoring rod position within at least +/- 12 steps with either full accuracy or half accuracy.

WOG STS B 3.1.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 91 of 213

Attachment 1, Volume 6, Rev. 0, Page 92 of 213 B 3.1.4 2

INSERT 2 The readout of the Bank Demand Position Indication System is in the form of an add and subtract counter reading. There is one display for each rod group.

2 INSERT 3 an analog signal that is proportional to the actual control rod position. This is accomplished by using an electrical coil stack linear variable differential transformer placed above the stepping mechanism of the control rod magnetic jacks which are external to the pressure housing. When the control rod is at the bottom of the core, the magnetic coupling between the primary and secondary windings is small and induces a small voltage in the secondary winding. As the control rod raises, the lift rod causes an increase in magnetic coupling which in turn provides an analog signal proportional to the control rod position. The individual rod position indicator channels are sufficiently accurate to detect a rod

+ 12 steps away from its demand position.

Insert Page B 3.1.4-2 Attachment 1, Volume 6, Rev. 0, Page 92 of 213

Attachment 1, Volume 6, Rev. 0, Page 93 of 213 Rod Group Alignment Limits B 3.1.4 BASES APPLICABLE Control rod misalignment accidents are analyzed in the safety analysis SAFETY (Ref. 3). The acceptance criteria for addressing control rod inoperability ANALYSES or misalignment are that:

a. There be no violations of:
1. Specified acceptable fuel design limits or 4
2. Reactor Coolant System (RCS) pressure boundary integrity and 4
b. The core remains subcritical after accident transients.

INSERT 4 Two types of misalignment are distinguished. During movement of a control rod group, one rod may stop moving, while the other rods in the 2 group continue. This condition may cause excessive power peaking.

A statically misaligned The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.

Two types of analysis are performed in regard to static rod misalignment (Ref. 4). With control banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core.

The second type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying 2

limits on departure from nucleate boiling ratio in both of these cases INSERT 5 bounds the situation when a rod is misaligned from its group by 12 steps.

INSERT 6 2 Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed 5 in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5).

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

WOG STS B 3.1.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 93 of 213

Attachment 1, Volume 6, Rev. 0, Page 94 of 213 B 3.1.4 2

INSERT 4 There are three RCCA misalignment accidents which are analyzed including one or more dropped RCCAs, a dropped RCCA bank, and a statically misaligned RCCA (Ref. 4).

2 INSERT 5 when THERMAL POWER is 85% RTP or by 24 steps when THERMAL POWER is < 85% RTP 2

INSERT 6 For the dropped RCCA(s) or dropped RCCA bank misalignment accidents, a negative reactivity insertion will result. Power will be reestablished either by reactivity feedback or control bank withdrawal. Following plant stabilization, normal rod retrieval or shutdown procedures are followed. For dropped RCCA events in the automatic rod control mode, the Rod Control System detects the drop in power and initiates control bank withdrawal. In all cases, the minimum departure from nucleate boiling ratio (DNBR) remains above the limit value.

Insert Page B 3.1.4-3 Attachment 1, Volume 6, Rev. 0, Page 94 of 213

Attachment 1, Volume 6, Rev. 0, Page 95 of 213 Rod Group Alignment Limits B 3.1.4 BASES APPLICABLE SAFETY ANALYSES (continued)

Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor ( FQ(Z)) and the nuclear enthalpy hot channel factor ( F N H ) are verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the 5 rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and FQ(Z) and F N H must be verified directly by incore mapping.

Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of FQ(Z) and F N H to the operating limits.

Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on control rod OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The control rod OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements, which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment. The rod OPERABILITY requirement is satisfied provided the rod will fully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability.

INSERT 7 The requirement to maintain the rod alignment to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment 2 from fully withdrawn to fully inserted is assumed.

INSERT 8 Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

INSERT 8A 6 WOG STS B 3.1.4-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 95 of 213

Attachment 1, Volume 6, Rev. 0, Page 96 of 213 B 3.1.4 2

INSERT 7 with THERMAL POWER 85% RTP or within 24 steps of their group step counter demand position when THERMAL POWER is < 85% RTP 2

INSERT 8 The safety analysis assumes a misalignment of one or more RCCA(s) or, in some cases, an entire RCCA bank.

6 INSERT 8A The LCO is modified by a Note stating that individual control rod position indications may be outside the required limits for < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following subsequent control rod movement. This allows up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of thermal soak time to allow the control rod drive shaft to reach thermal equilibrium and thus present a consistent position indication. Subsequent rod movement is considered to be 10 or more steps in one direction in < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Insert Page B 3.1.4-4 Attachment 1, Volume 6, Rev. 0, Page 96 of 213

Attachment 1, Volume 6, Rev. 0, Page 97 of 213 Rod Group Alignment Limits B 3.1.4 BASES APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant.

In MODES 3, 4, 5, and 6, the alignment limits do not apply because the normally control rods are bottomed and the reactor is shut down and not producing 8 fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.

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

In this situation, SDM verification must include the worth of the untrippable rod, as well as a rod of maximum worth.

A.2 When one or more rods are inoperable 6

If the inoperable rod(s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

B.1 When a rod becomes misaligned, it can usually be moved and is still 6

trippable. If the rod can be realigned within the Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, local xenon redistribution during this short interval will not be significant, and operation may proceed without further restriction.

WOG STS B 3.1.4-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 97 of 213

Attachment 1, Volume 6, Rev. 0, Page 98 of 213 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued)

An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the 6 bank sequence, overlap, and insertion limits specified in LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits." The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> gives the operator sufficient time to adjust the rod positions in an orderly manner.

, C.1.1, and C.1.2 6

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

In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully in and control bank C must be moved in to approximately 100 to 115 steps.

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

C.2, C.3, C.4, C.5, and C.6 B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 INSERT 8B 6

6 For continued operation with a misaligned rod, RTP must be reduced, SDM must periodically be verified within limits, hot channel factors (FQ(Z) and F N H ) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible.

< 85 INSERT 9 Reduction of power to 75% RTP ensures that local LHR increases due to 6 a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 7). The Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System.

WOG STS B 3.1.4-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 98 of 213

Attachment 1, Volume 6, Rev. 0, Page 99 of 213 B 3.1.4 6 INSERT 8B when > 85% RTP, THERMAL POWER must be reduced to < 85% RTP. Once power is reduced, the alignment limits are now 24 steps instead of 12 steps.

Thus, once power is reduced, Condition B no longer applies. If a rod continues to be misaligned (i.e., it is greater than 24 steps misaligned), then as stated in Required Actions C.2, C.3, C.4, C.5 and C.6, 6

INSERT 9 when the indicated rod position exceeds 12 steps from their respective banks and < 50% RTP when the indicated rod position exceeds 24 steps from their respective banks Insert Page B 3.1.4-6 Attachment 1, Volume 6, Rev. 0, Page 99 of 213

Attachment 1, Volume 6, Rev. 0, Page 100 of 213 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued)

When a rod is known to be misaligned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to ensure this requirement continues to be met.

Verifying that FQ(Z), as approximated by F QC ( Z) and F WQ ( Z ) , and F N H are

< 85 INSERT 10 within the required limits ensures that current operation at 75% RTP with 6 a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate FQ(Z) and F N H .

Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of U

operation under these conditions. The accident analyses presented in FSAR Chapter 15 (Ref. 5) that may be adversely affected will be 6 14 evaluated to ensure that the analysis results remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.

D C.1 associated When Required Actions cannot be completed within their Completion 6

Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit 3 6 must be brought to at least MODE 2 with Keff < 1.0 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.

E D.1.1 and D.1.2 6 More than one control rod becoming misaligned from its group average position is not expected, and has the potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires WOG STS B 3.1.4-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 100 of 213

Attachment 1, Volume 6, Rev. 0, Page 101 of 213 B 3.1.4 o--

6 INSERT 10 when the indicated rod position exceeds 12 steps from their respective banks and < 50% RTP when the indicated rod position exceeds 24 steps from their respective banks Insert Page B 3.1.4-7 Attachment 1, Volume 6, Rev. 0, Page 101 of 213

Attachment 1, Volume 6, Rev. 0, Page 102 of 213 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued) increasing the RCS boron concentration to provide negative reactivity, as of described in the Bases or LCO 3.1.1. The required Completion Time of 7 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for initiating boration is reasonable, based on the time required for potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps.

Boration will continue until the required SDM is restored.

E 6 D.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or 6 Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 2 with Keff < 1.0 6 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 3 The allowed Completion Time is reasonable, based on operating 3 6

experience, for reaching MODE 2 with Keff < 1.0 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Verification that individual rod positions are within alignment limits at a INSERT 11 Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. The 2 specified Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected.

SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2 with Keff 1.0, tripping each control 6 rod would result in radial or axial power tilts, or oscillations. Exercising each individual control rod every 92 days 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 10 steps will not cause radial or axial power tilts, or oscillations, to occur.

The 92 day Frequency takes into consideration other information WOG STS B 3.1.4-8 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 102 of 213

Attachment 1, Volume 6, Rev. 0, Page 103 of 213 B 3.1.4 2

INSERT 11 The verification may be determined by the comparison of the bank demand step counter to the analog IRPI, by the rod position as noted on the plant process computer or through the conditioning module output voltage via a correlation of rod position versus voltage.

Insert Page B 3.1.4-8 Attachment 1, Volume 6, Rev. 0, Page 103 of 213

Attachment 1, Volume 6, Rev. 0, Page 104 of 213 Rod Group Alignment Limits B 3.1.4 BASES SURVEILLANCE REQUIREMENTS (continued) available to the operator in the control room and SR 3.1.4.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. 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.

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

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

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

1 USAR, Sections 3.1.2.1, 3.1.2.2, 3.1.2.3, 3.1.2.4, 3.1.2.5, and 3.1.2.6.

2. 10 CFR 50.46.

U Section 14.1

3. FSAR, Chapter [15]. 2 4 U

Section 14.1.3 2 4

4. FSAR, Chapter [15].

U 14 2 4

5. FSAR, Chapter [15].
6. FSAR, Chapter [15].

2 4

7. FSAR, Chapter [15].

WOG STS B 3.1.4-9 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 104 of 213

Attachment 1, Volume 6, Rev. 0, Page 105 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4 BASES, ROD GROUP ALIGNMENT LIMITS

1. ISTS 3.1.4 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.

Therefore, ITS 3.1.4 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.

Subsequent reference numbers have been renumbered.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
4. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
5. The discussion of the Required Actions when the LCO is not met has been deleted since it is not appropriate in the Applicable Analyses Section. This information is adequately discussed in the Bases for ACTIONS B.2, B.3, B.4, B.5, B.6, C.2, C.3, C.4, C.5 and C.6.
6. Changes made to be consistent with the Specification.
7. Typographical/grammatical error corrected.
8. The word "normally" has been added since the shutdown banks are required by ITS LCO 3.1.5 to be withdrawn to their insertion limits prior to entering MODE 2 (i.e., in MODE 3). Thus the rods are not bottomed all the time in MODE 3.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 105 of 213

Attachment 1, Volume 6, Rev. 0, Page 106 of 213 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 6, Rev. 0, Page 106 of 213

Attachment 1, Volume 6, Rev. 0, Page 107 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.4, ROD GROUP ALIGNMENT LIMITS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 107 of 213

Attachment 1, Volume 6, Rev. 0, Page 108 of 213 ATTACHMENT 5 ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS Attachment 1, Volume 6, Rev. 0, Page 108 of 213

, Volume 6, Rev. 0, Page 109 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 109 of 213

Attachment 1, Volume 6, Rev. 0, Page 110 of 213 ITS A01 ITS 3.1.5

c. Quadrant Power Tilt Limits
1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:

A. Eliminate the tilt.

B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio

> 1.0.

2. If the tilt condition is not eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then reduce power to 50% or lower.

See ITS

3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is 3.2.4 simultaneous indication of a misaligned rod:

A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.

B. If the tilt condition is not eliminated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, then the reactor shall be brought to a minimum load condition ( 30 Mwe).

4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition

( 5% reactor power).

d. Rod Insertion Limits LCO 3.1.5
1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when A02 Applicability the reactor is critical or approaching criticality.

Add proposed ACTIONS A and B

2. The control banks shall be limited in physical insertion; insertion limits are specified in L01 the COLR. If any one of the control bank insertion limits is not met:

A. Within one hour, initiate boration to restore control bank insertion to within the limits specified in the COLR, and B. Restore control bank insertion to within the limits specified in the COLR within See ITS 3.1.6 two hours of exceeding the insertion limits.

C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then withinone hour action shall be initiated to:

- Achieve HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> See ITS 3.1.8 Applicability 3. Insertion limit does not apply during physics tests or during periodic exercise of Note individual rods. However, the shutdown margin, as specified in the COLR, must be See ITS maintained except for the Low Power Physics Test to measure control rod worth and 3.1.8 shutdown margin. For this test, the reactor may be critical with all but one high worth rod inserted.

Add proposed SR 3.1.5.1 M01 Amendment No. 165 TS 3.10-4 03/11/2003 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 110 of 213

Attachment 1, Volume 6, Rev. 0, Page 111 of 213 DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.10.d.1 requires the shutdown rods to be withdrawn to within the limits specified in the COLR when the reactor is critical or approaching criticality. ITS 3.1.5 requires each shutdown bank to be within the insertion limits specified in the COLR in MODES 1 and 2. This changes the CTS by clearly specifying the MODES in which the LCO is required.

The CTS condition of "reactor is critical" normally occurs only in ITS MODE 1 and MODE 2 with keff > 1.0. The CTS condition of "approaching criticality" normally occurs only in MODE 2 with keff < 1.0. Therefore, since the ITS 3.1.5 MODES 1 and 2 are equivalent to the CTS conditions, this change is acceptable and is designated as administrative.

MORE RESTRICTIVE CHANGES M01 CTS 3.10.d.1 requires the shutdown rods to be withdrawn to within the limits specified in the COLR, but does not provide a Surveillance Requirement to verify that the shutdown rods are within the limits specified in the COLR. ITS SR 3.1.5.1 requires verification that the shutdown banks are within the insertion limits specified in the COLR every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This changes the CTS by adding a new Surveillance Requirement to verify that the shutdown banks are within the insertion limits specified in the COLR every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The purpose of ITS SR 3.1.5.1 is to verify that the shutdown banks are within their insertion limits, available to shut down the reactor, and that the required SHUTDOWN MARGIN (SDM) will be maintained following a reactor trip. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into account other information that is available in the control room for monitoring the status of the shutdown rods. This change is designated as more restrictive because a new Surveillance Requirement has been added to verify that the shutdown banks are within the insertion limits specified in the COLR every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

RELOCATED SPECIFICATIONS None Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 111 of 213

Attachment 1, Volume 6, Rev. 0, Page 112 of 213 DISCUSSION OF CHANGES ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.10.d.1 does not provide any explicit time to restore the shutdown rods to within their limits. As a result, LCO 3.0.c would be entered, which requires action to be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Under similar conditions, ITS 3.1.5 ACTION A provides one hour to either verify that the SDM is within the limits specified in the COLR or to initiate boration to restore the SDM to within limits and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to restore the shutdown banks to within limits. Additionally, ITS 3.1.5 ACTION B provides 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to be in MODE 3 if any Required Action and associated Completion Time of ACTION A are not met. This changes the CTS by providing specific ACTIONS when the shutdown banks are not within the limits specified in the COLR.

The purpose of CTS 3.10.d.1 is to ensure that the shutdown rods are within limits specified in the COLR. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the shutdown rods not being within the limits established in the COLR because the SDM may be significantly reduced. ITS 3.1.5 Required Action A.2 provides 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in lieu of the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in CTS 3.10.d.1 to restore the shutdown bank to within limits. This is acceptable because ITS 3.1.5 Required Action A.2 provides sufficient time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time. The Completion Time of ITS 3.1.5 Required Action B.1 provides 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to be in MODE 3 which is a reasonable time, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging the plant systems. Additionally, since ITS 3.1.5 Required Action B.1 requires the unit to be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, there is no need to maintain the requirement to be in MODE 2 within the same 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. This change is designated as less restrictive because more time is provided in the ITS to restore the shutdown banks to within limits prior to requiring a unit shutdown.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 112 of 213

Attachment 1, Volume 6, Rev. 0, Page 113 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 113 of 213

Attachment 1, Volume 6, Rev. 0, Page 114 of 213 CTS Shutdown Bank Insertion Limits 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Shutdown Bank Insertion Limits 3.10.d.1 LCO 3.1.5 Each shutdown bank shall be within insertion limits specified in the COLR.

3.10.d.1 APPLICABILITY: MODES 1 and 2.

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

3.10.d.3 This LCO is not applicable while performing SR 3.1.4.2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC L01 A. One or more shutdown A.1.1 Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> banks not within limits. limits specified in the COLR.

OR A.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limit.

AND A.2 Restore shutdown banks to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> within limits.

DOC L01 B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

WOG STS 3.1.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 114 of 213

Attachment 1, Volume 6, Rev. 0, Page 115 of 213 CTS Shutdown Bank Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M01 SR 3.1.5.1 Verify each shutdown bank is within the insertion 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> limits specified in the COLR.

WOG STS 3.1.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 115 of 213

Attachment 1, Volume 6, Rev. 0, Page 116 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS None Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 116 of 213

Attachment 1, Volume 6, Rev. 0, Page 117 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 117 of 213

Attachment 1, Volume 6, Rev. 0, Page 118 of 213 Shutdown Bank Insertion Limits B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Shutdown Bank Insertion Limits BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available ejected rod worth, SDM and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," 1 INSERT 1 GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of If two or more RCCAs that are electrically paralleled to step simultaneously.

, the groups A bank of RCCAs consists of two groups that are moved in a staggered 4 fashion, but always within one step of each other. All plants have four Kewaunee Power 2 control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Station Group Alignment Limits," for control and shutdown rod OPERABILITY (KPS) has and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements.

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

The positions of the control banks are normally automatically controlled 2

by the Rod Control System, but they can also be manually controlled.

They are capable of adding negative reactivity very quickly (compared to borating). The control banks must be maintained above designed insertion limits and are typically near the fully withdrawn position during normal full power operations.

WOG STS B 3.1.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 118 of 213

Attachment 1, Volume 6, Rev. 0, Page 119 of 213 B 3.1.5 o 1

-INSERT 1 USAR GDC 6, "Reactor Core Design," GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot Shutdown Capability," GDC 29, "Reactivity Shutdown Capability," and GDC 30, "Reactivity Holddown Capability" (Ref. 1),

Insert Page B 3.1.5-1 Attachment 1, Volume 6, Rev. 0, Page 119 of 213

Attachment 1, Volume 6, Rev. 0, Page 120 of 213 Shutdown Bank Insertion Limits B 3.1.5 BASES BACKGROUND (continued)

Hence, they are not capable of adding a large amount of positive reactivity. Boration or dilution of the Reactor Coolant System (RCS) compensates for the reactivity changes associated with large changes in RCS temperature. The design calculations are performed with the assumption that the shutdown banks are withdrawn first. The shutdown banks can be fully withdrawn without the core going critical. This provides available negative reactivity in the event of boration errors. The shutdown banks are controlled manually by the control room operator.

During normal unit operation, the shutdown banks are either fully 1 withdrawn or fully inserted. The shutdown banks must be completely withdrawn from the core, prior to withdrawing any control banks during an approach to criticality. The shutdown banks are then left in this position until the reactor is shut down. They affect core power and burnup distribution, and add negative reactivity to shut down the reactor upon receipt of a reactor trip signal.

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

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

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

a. There be no violations of:
1. Specified acceptable fuel design limits or 3
2. RCS pressure boundary integrity and 3 WOG STS B 3.1.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 120 of 213

Attachment 1, Volume 6, Rev. 0, Page 121 of 213 Shutdown Bank Insertion Limits B 3.1.5 BASES APPLICABLE SAFETY ANALYSES (continued)

b. The core remains subcritical after accident transients.

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

The shutdown bank insertion limits preserve an initial condition assumed in the safety analyses and, as such, satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

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

The shutdown bank insertion limits are defined in the COLR.

APPLICABILITY The shutdown banks must be within their insertion limits, with the reactor in MODES 1 and 2. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. The shutdown banks do not have to be within their insertion limits in MODE 3, unless an approach to criticality is being made. In MODE 3, 4, 5, or 6, the shutdown banks are fully inserted in the core and contribute to the SDM. Refer to LCO 3.1.1 for SDM requirements in MODES 3, 4, and 5. LCO 3.9.1, "Boron Concentration,"

ensures adequate SDM in MODE 6.

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

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

This is necessary because the available SDM may be significantly reduced, with one or more of the shutdown banks not within their insertion limits. Also, verification of SDM or initiation of boration within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required, since the SDM in MODES 1 and 2 is ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1). If shutdown banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1. 5 WOG STS B 3.1.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 121 of 213

Attachment 1, Volume 6, Rev. 0, Page 122 of 213 Shutdown Bank Insertion Limits B 3.1.5 BASES ACTIONS (continued)

The allowed Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

B.1 any Required Action and associated Completion Time is not met 6

If the shutdown banks cannot be restored to within their insertion limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the unit must be brought to a MODE where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

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

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

Since the shutdown banks are positioned manually by the control room operator, a verification of shutdown bank position at a Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, after the reactor is taken critical, is adequate to ensure that they are within their insertion limits. Also, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into account other information available in the control room for the purpose of monitoring the status of shutdown rods.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28. 1 USAR, Sections 3.1.2.1, 3.1.2.3, 3.1.2.4, 3.1.2.5,

2. 10 CFR 50.46. and 3.1.2.6.

U Section 14.1

3. FSAR, Chapter [15]. 2 WOG STS B 3.1.5-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 122 of 213

Attachment 1, Volume 6, Rev. 0, Page 123 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.5 BASES, SHUTDOWN BANK INSERTION LIMITS

1. ISTS 3.1.5 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.

Therefore, ITS 3.1.5 Bases Background has been revised to discuss the design standards used by KPS. Additionally, Bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
4. Words changed to be consistent with similar words in the ISTS 3.1.4 Bases. Not all banks are divided into two groups at KPS.
5. Typographical error corrected.
6. Change made to be consistent with the Specification. The words in the ISTS Bases only describe Required Action A.2. ACTION B must be entered if any Required Action and associated Completion Time of Condition A is not met.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 123 of 213

Attachment 1, Volume 6, Rev. 0, Page 124 of 213 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 6, Rev. 0, Page 124 of 213

Attachment 1, Volume 6, Rev. 0, Page 125 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.5, SHUTDOWN BANK INSERTION LIMITS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 125 of 213

, Volume 6, Rev. 0, Page 126 of 213 ATTACHMENT 6 ITS 3.1.6, CONTROL BANK INSERTION LIMITS , Volume 6, Rev. 0, Page 126 of 213

, Volume 6, Rev. 0, Page 127 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 127 of 213

Attachment 1, Volume 6, Rev. 0, Page 128 of 213 ITS A01 ITS 3.1.6

c. Quadrant Power Tilt Limits
1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:

A. Eliminate the tilt.

B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio

> 1.0.

2. If the tilt condition is not eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then reduce power to 50% or lower.

See ITS

3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is 3.2.4 simultaneous indication of a misaligned rod:

A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.

B. If the tilt condition is not eliminated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, then the reactor shall be brought to a minimum load condition ( 30 Mwe).

4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition

( 5% reactor power).

d. Rod Insertion Limits See ITS
1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when 3.1.5 the reactor is critical or approaching criticality.

, sequence, and overlap limits A02 LCO 3.1.6

2. The control banks shall be limited in physical insertion; insertion limits are specified in L01 the COLR. If any one of the control bank insertion limits is not met: Add proposed Applicability Add proposed Required Actions A.1.1 and B.1.1 L02 A. Within one hour, initiate boration to restore control bank insertion to within the limits specified in the COLR, and ACTION A, ACTION B B. Restore control bank insertion to within the limits specified in the COLR within two hours of exceeding the insertion limits.

C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then L01 withinone hour action shall be initiated to:

ACTION C

- Achieve HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> See ITS 3.1.8 Applicability 3. Insertion limit does not apply during physics tests or during periodic exercise of Note individual rods. However, the shutdown margin, as specified in the COLR, must be See ITS maintained except for the Low Power Physics Test to measure control rod worth and 3.1.8 shutdown margin. For this test, the reactor may be critical with all but one high worth rod inserted.

Add proposed SR 3.1.6.1, SR 3.1.6.2, and SR 3.1.6.3 M01 Amendment No. 165 TS 3.10-4 03/11/2003 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 128 of 213

Attachment 1, Volume 6, Rev. 0, Page 129 of 213 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.10.d.2 requires, in part, that the control rod banks are to be limited in physical insertion as specified in the COLR. ITS 3.1.6 requires the control bank to be within the insertion, sequence, and overlap limits specified in the COLR.

This changes the CTS by specifically stating that the control bank physical insertion limits are the control bank insertion, overlap, and sequence limits.

This change is acceptable because the LCO requirements have not changed.

COLR Section 2.5 states, in part, that physical insertion limits are shown in COLR Figure 4. COLR Figure 4 is a graphical presentation of when each control bank is required to be withdrawn with respect to the Percent of Rated Thermal Power. As shown in the Figure, the sequence of control bank withdrawal is included. Additionally, the Note for COLR Figure 4 states that the Rod Bank Insertion Limits are based on a control bank tip-to-tip distance of 126 steps.

Furthermore, the COLR Figure 4 is similar to the ISTS Bases Figure B 3.1.6.

Therefore, the physical insertion limits in CTS 3.10.d.2 include the insertion, sequence, and overlap limits. This change is designated as an administrative change since it does not result in any technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.10.d does not provide any Surveillance Requirements for verifying control bank insertion limits. ITS SR 3.1.6.1 requires verifying that the estimated critical control bank position is within the insertion limits specified in the COLR within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving criticality. ITS SR 3.1.6.2 requires verifying that each control bank insertion is within the insertion limits specified in the COLR every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ITS SR 3.1.6.3 requires verification that the overlap and sequence limits specified in the COLR are met for control banks not fully withdrawn from the core every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This changes the CTS by adding specific Surveillance Requirements to periodically verify that control banks will be within insertion, sequence and overlap limits specified in the COLR.

The purpose of ITS SR 3.1.6.1 is to ensure that the reactor does not achieve criticality with the control banks below their insertion limits. The purpose of ITS SR 3.1.6.2 is to detect control banks that may be approaching the insertion limits. The purpose of ITS SR 3.1.6.3 is to verify that control bank sequence and overlap is within the requirements specified in the COLR. Therefore, this change is considered acceptable and is more restrictive because new Surveillance Requirements have been added to periodically verify the LCO limits are met.

Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 129 of 213

Attachment 1, Volume 6, Rev. 0, Page 130 of 213 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 3.10.d.2 requires the control banks to be limited in physical insertion as specified in the COLR, but does not specifically state an applicability. CTS 3.10.d.2.A, B, and C provide actions to be taken if the control banks insertion limits are not met. If the control bank insertion limits are not restored, then the unit must be placed in HOT STANDBY (equivalent to ITS MODE 2) within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and in HOT SHUTDOWN (equivalent to ITS MODE 3) within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Thus, the effective Applicability for the control bank insertion limits is OPERATING and HOT STANDBY MODES (equivalent to ITS MODES 1 and 2). ITS 3.1.6 requires the control bank insertion limits to be met in MODE 1 and in MODE 2 with keff 1.0. This changes the CTS by only requiring the control bank limits to met in MODE 2 with keff 1.0, in lieu of at all times in MODE 2. Consistent with this change in Applicability, the requirements in CTS 3.10.d.2.C to be in HOT STANDBY and HOT SHUTDOWN are changed to be in MODE 2 with keff < 1.0 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The control bank insertion limits are required to prevent core power distributions that could result in fuel cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RPS trip function. This change is acceptable because in MODE 2 with keff < 1.0, neither the assumed power distribution nor the ejected rod worth would be exceeded. Furthermore, allowing 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to be in MODE 2 with keff < 1.0 is a reasonable time, based on operating experience, for reaching the required MODE from full power in an orderly manner and without challenging plant systems. This proposed ACTION also places the unit outside the new Applicability. This change is designated as less restrictive because the Applicability has been changed to MODE 1 and MODE 2 with keff 1.0.

L02 (Category 4 - Relaxation of Required Action) CTS 3.10.d.2.A requires that if the control banks physical insertion limits (i.e., insertion, sequence, or overlap) are not met, then within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate boration to restore control bank insertion to within limits. ITS 3.1.6 ACTION A and ACTION B contain the same requirements but allows a choice between verifying that the SDM is within the limits specified in the COLR (ITS 3.1.6 Required Actions A.1.1 and B.1.1) within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or initiating boration to restore SDM to within limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (ITS 3.1.6 Required Actions A.1.2 and B.1.2). This changes the CTS by allowing either the verification that SDM is within the limits specified in the COLR or the initiation of boration to restore the required SDM within one hour when the control banks are not within the physical insertion limits.

Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 130 of 213

Attachment 1, Volume 6, Rev. 0, Page 131 of 213 DISCUSSION OF CHANGES ITS 3.1.6, CONTROL BANK INSERTION LIMITS This change is acceptable because it requires verification that the initial conditions of the analysis is maintained. In MODE 1 and MODE 2 with keff > 1.0, SDM is normally ensured by adhering to the control and shutdown bank insertion limits. If the control banks are not within their insertion limits, then SDM must be verified to be within limits or boration must be initiated to restore SDM to within limits. This change is designated as less restrictive because an alternate allowance to verify SDM is within the limits specified in the COLR is being added to the CTS.

Kewaunee Power Station Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 131 of 213

Attachment 1, Volume 6, Rev. 0, Page 132 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 132 of 213

Attachment 1, Volume 6, Rev. 0, Page 133 of 213 CTS Control Bank Insertion Limits 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Control Bank Insertion Limits 3.10.d.2 LCO 3.1.6 Control banks shall be within the insertion, sequence, and overlap limits specified in the COLR.

DOC L01 APPLICABILITY: MODE 1, MODE 2 with keff 1.0.

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

3.10.d.3 This LCO is not applicable while performing SR 3.1.4.2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.10.d.2.A, 3.10.d.2.B A. Control bank insertion A.1.1 Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limits not met. limits specified in the COLR.

OR A.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limit. 1 s

AND A.2 Restore control bank(s) to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> within limits.

3.10.d.2.A, B. Control bank sequence B.1.1 Verify SDM is within the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 3.10.d.2.B or overlap limits not met. limits specified in the COLR.

OR B.1.2 Initiate boration to restore 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SDM to within limit. 1 s

AND WOG STS 3.1.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 133 of 213

Attachment 1, Volume 6, Rev. 0, Page 134 of 213 CTS Control Bank Insertion Limits 3.1.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.10.d.2.A, 3.10.d.2.B B.2 Restore control bank 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> sequence and overlap to within limits.

3.10.d.2.C C. Required Action and C.1 Be in MODE 2 with keff 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion < 1.0.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M02 SR 3.1.6.1 Verify estimated critical control bank position is Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> within the limits specified in the COLR. prior to achieving 2 insertion criticality DOC M02 SR 3.1.6.2 Verify each control bank insertion is within the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> insertion limits specified in the COLR.

DOC M02 SR 3.1.6.3 Verify sequence and overlap limits specified in the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> COLR are met for control banks not fully withdrawn from the core.

WOG STS 3.1.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 134 of 213

Attachment 1, Volume 6, Rev. 0, Page 135 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6, CONTROL BANK INSERTION LIMITS

1. Typographical/grammatical error corrected to be consistent with the wording in the proceeding Required Action.
2. SR 3.1.6.1 is clarified to state that the estimated critical control bank position must be verified to be within the "insertion limits," specified in the COLR. This Specification requires three limits to be met; insertion, sequence, and overlap.

As stated in the ISTS SR 3.1.6.1 Bases, the limit to be checked is only the insertion limit.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 135 of 213

Attachment 1, Volume 6, Rev. 0, Page 136 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 136 of 213

Attachment 1, Volume 6, Rev. 0, Page 137 of 213 Control Bank Insertion Limits B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Control Bank Insertion Limits BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available SDM, and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," 1 INSERT 1 GDC 26, "Reactivity Control System Redundancy and Protection," GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of If two or more RCCAs that are electrically paralleled to step simultaneously.

A bank of RCCAs consists of two groups that are moved in a staggered 6

, the groups fashion, but always within one step of each other. All plants have four Kewaunee Power 2 control banks and at least two shutdown banks. See LCO 3.1.4, "Rod Station Group Alignment Limits," for control and shutdown rod OPERABILITY (KPS) has and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements.

The control bank insertion limits are specified in the COLR. An example 3 is provided for information only in Figure B 3.1.6-1. The control banks are required to be at or above the insertion limit lines.

INSERT 2 Figure B 3.1.6-1 also indicates how the control banks are moved in an 4 overlap pattern. Overlap is the distance travelled together by two control banks. The predetermined position of control bank C, at which control 3

bank D will begin to move with bank C on a withdrawal, will be at 118 steps for a fully withdrawn position of 231 steps. The fully withdrawn position is defined in the COLR.

WOG STS B 3.1.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 137 of 213

Attachment 1, Volume 6, Rev. 0, Page 138 of 213 B 3.1.6 1

INSERT 1 USAR GDC 6, "Reactor Core Design," GDC 27, "Redundancy of Reactivity Control," GDC 28, "Reactivity Hot Shutdown Capability," GDC 29, "Reactivity Shutdown Capability," and GDC 30, "Reactivity Holddown Capability" (Ref. 1),

4 INSERT 2 The control bank sequence and overlap limits are specified in the COLR.

Sequencing is the order in which the banks are moved.

Insert Page B 3.1.6-1 Attachment 1, Volume 6, Rev. 0, Page 138 of 213

Attachment 1, Volume 6, Rev. 0, Page 139 of 213 Control Bank Insertion Limits B 3.1.6 BASES BACKGROUND (continued)

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

The positions of the control banks are normally controlled automatically by the Rod Control System, but can also be manually controlled. They are capable of adding reactivity very quickly (compared to borating or diluting).

The power density at any point in the core must be limited, so that the fuel design criteria are maintained. Together, LCO 3.1.4, LCO 3.1.5, "Shutdown Bank Insertion Limits," LCO 3.1.6, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.

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

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

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

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

a. There be no violations of:
1. Specified acceptable fuel design limits or 5
2. Reactor Coolant System pressure boundary integrity and 5
b. The core remains subcritical after accident transients.

WOG STS B 3.1.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 139 of 213

Attachment 1, Volume 6, Rev. 0, Page 140 of 213 Control Bank Insertion Limits B 3.1.6 BASES APPLICABLE SAFETY ANALYSES (continued)

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

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

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

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

control bank The insertion limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii), in that 2 they are initial conditions assumed in the safety analysis.

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

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

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

WOG STS B 3.1.6-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 140 of 213

Attachment 1, Volume 6, Rev. 0, Page 141 of 213 Control Bank Insertion Limits B 3.1.6 BASES ACTIONS A.1.1, A.1.2, A.2, B.1.1, B.1.2, and B.2 When the control banks are outside the acceptable insertion limits, they must be restored to within those limits. This restoration can occur in two ways:

a. Reducing power to be consistent with rod position or 5
b. Moving rods to be consistent with power.

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

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

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

The allowed Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for restoring the banks to within the insertion, sequence, and overlaps limits provides an acceptable time 7 for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

any Required Action and associated Completion Time is not met C.1 If Required Actions A.1 and A.2, or B.1 and B.2 cannot be completed 8 within the associated Completion Times, the plant must be brought to MODE 2 with keff < 1.0, where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

WOG STS B 3.1.6-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 141 of 213

Attachment 1, Volume 6, Rev. 0, Page 142 of 213 Control Bank Insertion Limits B 3.1.6 BASES SURVEILLANCE SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.

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

SR 3.1.6.2 Verification of the control bank insertion limits at a Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to detect control banks that may be approaching the insertion limits since, normally, very little rod motion occurs in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SR 3.1.6.3 When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. A Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is consistent with the insertion limit check above in SR 3.1.6.2.

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

1 USAR, Sections 3.1.2.1, 3.1.2.3, 3.1.2.4, 3.1.2.5,

2. 10 CFR 50.46. and 3.1.2.6.

U Section 14.1

3. FSAR, Chapter [15]. 2
4. FSAR, Chapter [15].

2

5. FSAR, Chapter [15].

WOG STS B 3.1.6-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 142 of 213

Attachment 1, Volume 6, Rev. 0, Page 143 of 213 Control Bank Insertion Limits B 3.1.6 (17,231) 231 200 (100,190)

...--- (0,191 )

U)

z: 150 o

I-en o

a...

~

~ 100 co

---.J o

a:

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z:

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THIS FIGURE FOR ILLUSTRATION ONLY.

3 DO NOT USE FOR OPERATION.

o o 60 100 PERCENT OF RTP Figure B 3.1.6 (page 1 of 1)

Control Bank Insertion vs. Percent RTP WOG STS B 3.1.6-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 143 of 213

Attachment 1, Volume 6, Rev. 0, Page 144 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6 BASES, CONTROL BANK INSERTION LIMITS

1. ISTS 3.1.6 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.

Therefore, ITS 3.1.6 Bases Background has been revised to discuss the design standards used by KPS. Additionally, Bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. Since the ITS states the actual control bank insertion, sequence, and overlap limits are specified in the COLR, the example provided for information only is not needed in the Bases and has been deleted. Furthermore, since the overlap limits are specified in the COLR, the description of the actual value of the limit provided in the paragraph is also not needed and has been deleted.
4. LCO 3.1.6 governs control bank insertion, sequence, and overlap limits. The Background section of ITS 3.1.6 Bases discusses insertion and overlaps, but does not discuss sequence. A discussion of control bank sequence is added for completeness.
5. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
6. The wording has been changed to be consistent with the wording used in Bases of ISTS 3.1.4.
7. Typographical/grammatical error corrected.
8. Change made to be consistent with the Specification. The words in the ISTS Bases implies that Condition C is entered only when both Required Actions A.1 and A.2, or both Required Actions B.1 and B.2 are not met. Condition C is to be entered if any Required Action of ACTION A or B is not met. In addition, there is no Required Action A.1 or B.1.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 144 of 213

Attachment 1, Volume 6, Rev. 0, Page 145 of 213 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 6, Rev. 0, Page 145 of 213

Attachment 1, Volume 6, Rev. 0, Page 146 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.6, CONTROL BANK INSERTION LIMITS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 146 of 213

, Volume 6, Rev. 0, Page 147 of 213 ATTACHMENT 7 ITS 3.1.7, ROD POSITION INDICATION , Volume 6, Rev. 0, Page 147 of 213

, Volume 6, Rev. 0, Page 148 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 148 of 213

Attachment 1, Volume 6, Rev. 0, Page 149 of 213 ITS A01 ITS 3.1.7

e. Rod Misalignment Limitations NOTE: Individual RPIs may be outside their limits for < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following substantial rod movement See ITS 3.1.4 This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.

See ITS Rod misalignment limitations do not apply during physics testing. 3.1.8

1. When reactor power is 85% of rating, the rod cluster control assembly shall be maintained within +/- 12 steps from their respective banks. If a rod cluster control See ITS assembly is misaligned from its bank by more than +/- 12 steps when reactor power is 3.1.4 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating.
2. When reactor power is < 85% but 50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is

< 85% but 50%, the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating.

3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.
f. Inoperable Rod Position Indicator Channels Add proposed LCO statement A02 Add proposed Applicability NOTE: Individual RPIs may be outside their limits for < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following substantial rod See ITS 3.1.4 movement ACTION A
1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate entry condition is allowed for ACTIONS each inoperable individual rod position indicator.)

Note A. Verify the position of the rod cluster control indirectly by movable incore detectors ACTION A each eight hours, or B. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reduce reactor thermal power to 50% of RATED POWER.

2. If more than one individual rod position indicator channel per group are inoperable, then:

ACTION B A. IMMEDIATELY place the control rods in manual, and B. Once per 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, monitor and record RCS Tavg, and ACTION A C. Verify the position of the rod by movable incore detectors each 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and Amendment No. 181 TS 3.10-5 03/24/2005 Page 1 of 4 Attachment 1, Volume 6, Rev. 0, Page 149 of 213

Attachment 1, Volume 6, Rev. 0, Page 150 of 213 ITS A01 ITS 3.1.7 ACTION B D. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> restore the inoperable individual rod position indicators to OPERABLE status such that a maximum of one IRPI per group is inoperable or ACTION D place the plant in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3. If one or more rods with inoperable individual rod position indicators have been moved in excess of 24 steps in one direction since the last determination of the rods position then perform A or B below:

ACTION A A. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> verify the position of the rod by movable incore detectors, or B. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reduce reactor thermal power to 50% of RATED POWER.

ACTION C

4. If one demand position indicator per bank for one or more banks is inoperable then perform either A or B below: (Note: Separate condition entry is allowed for each ACTIONS Note inoperable demand position indicator.)

A. Each 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> verify,

1) All IRPIs for the affected banks are OPERABLE, by administrative means, and ACTION C 2) The most withdrawn rod and the least withdrawn rod of the affected bank(s) are 12 steps apart when > 85% RATED POWER or 24 steps when A03 85% RATED POWER.

B. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reduce reactor thermal power to 50% of RATED POWER.

5. If a rod cluster control assembly having a rod position indicator channel out of service is A04 found to be misaligned from TS 3.10.f.1.A, then TS 3.10.e will be applied.

Amendment No. 181 TS 3.10-6 03/24/2005 Page 2 of 4 Attachment 1, Volume 6, Rev. 0, Page 150 of 213

Attachment 1, Volume 6, Rev. 0, Page 151 of 213 ITS ITS 3.1.7 A01 4.1 OPERATIONAL SAFETY REVIEW APPLICABILITY Applies to items directly related to safety limits and LIMITING CONDITIONS FOR OPERATION.

OBJECTIVE To assure that instrumentation shall be checked, tested, and calibrated, and that equipment and sampling tests shall be conducted at sufficiently frequent intervals to ensure safe operation.

SPECIFICATION SR 3.1.7.1 a. Calibration, testing, and checking of protective instrumentation channels and testing of logic channels shall be performed as specified in Table TS 4.1-1.

b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and See other TS 4.1-3. ITS
c. Deleted
d. Deleted
e. Deleted Amendment No. 119 TS 4.1-1 04/18/95 Page 3 of 4 Attachment 1, Volume 6, Rev. 0, Page 151 of 213

ITS A01 ITS 3.1.7 TABLE TS 4.1-1 See ITS 3.3.5 See ITS MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS 3.3.1 and 3.3.2 CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

6. Pressurizer Water Level Each shift Each refueling cycle Monthly
7. Pressurizer Pressure Each shift Each refueling cycle Monthly
8. a. 4-KV Voltage and Not applicable Each refueling cycle Monthly Reactor protection circuits only Frequency
b. 4-KV Voltage Not applicable Each refueling cycle Monthly Safeguards buses only (Loss of Voltage)
c. 4-KV Voltage Not applicable Each refueling cycle Monthly Safeguards buses only (Degraded Grid)
9. Analog Rod Position Each shift(a,b) Each refueling cycle Each refueling (a) With step counters cycle (b) Following rod motion in excess of 24 steps when computer is out of service
10. Rod Position Bank Each shift(a,b) Not applicable Each refueling (a) With analog rod position Counters cycle (b) Following rod motion in excess of 24 steps when computer is out of service See ITS See ITS 3.1.4 3.1.4 SR 3.1.7.1 SR 3.1.7.2 , Volume 6, Rev. 0, Page 152 of 213 Attachment 1, Volume 6, Rev. 0, Page 152 of 213 Amendment No. 151 Page 2 of 7 02/12/2001 Page 4 of 4

Attachment 1, Volume 6, Rev. 0, Page 153 of 213 DISCUSSION OF CHANGES ITS 3.1.7, ROD POSITION INDICATION ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.10.f describes the ACTIONS to take for one inoperable individual rod position indicator channel per group, more than one inoperable individual rod position indicator channel per group, one or more rods with inoperable individual rod position indicators, and one demand position indicator bank inoperable for one or more banks. The lowest plant condition required by the CTS Actions (CTS 3.10.f.2.D) is HOT SHUTDOWN (ITS equivalent MODE 3). ITS 3.1.7 requires similar ACTIONS, but specifically states in the LCO that the Individual Rod Position Indication (IRPI) System and the Demand Position Indication System shall be OPERABLE in MODES 1 and 2. This changes the CTS by stating the specific Limiting Conditions for Operability (LCO) and the specific MODE of Applicability.

This change is acceptable because the requirements for the IRPI System and the Demand Position Indication System have not changed. Actions will continue to be taken when the individual rod position indicators and the demand position indicator banks are inoperable. This change clearly states what is required to be OPERABLE and the MODE in which the Systems are required to be OPERABLE. Since the lowest plant condition required by the CTS Actions is HOT SHUTDOWN and the ITS Applicability is MODES 1 and 2. The proposed Applicability change is acceptable. This change is considered administrative because the technical requirements have not changed.

A03 CTS 3.10.f.4 provides the Actions if one demand position indicator per bank for one or more banks is inoperable. CTS 3.10.f.4.A.2) requires a verification that the most withdrawn rod and the least withdrawn rod of the affected bank(s) are

< 12 steps apart when > 85% RATED POWER or < 24 steps apart when < 85%

RATED POWER. Under similar conditions, ITS 3.1.7 Required Action C.1.2 requires a verification that the most withdrawn rod and the least withdrawn rod of the affected bank are within the required rod alignment limits of LCO 3.1.4, "Rod Group Alignment Limits." This changes the CTS by not specifying the actual rod alignment limits in this Action, but referencing the applicable LCO where the limits are controlled.

The purpose of this CTS Action is to ensure the rod alignment limits are maintained when a demand position indicator is inoperable. This change is acceptable since the actual limits are not changing. ITS LCO 3.1.4 includes the appropriate limits, and ITS 3.1.4 Required Action C.1.2 references the appropriate LCO that controls the limits. This change is administrative since the technical requirements are not being changed.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 153 of 213

Attachment 1, Volume 6, Rev. 0, Page 154 of 213 DISCUSSION OF CHANGES ITS 3.1.7, ROD POSITION INDICATION A04 CTS 3.10.f.5 states that if a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned when performing 3.10.f.1.A, then 3.10.e will be applied. ITS 3.1.7 does not contain the statement.

This changes the CTS by not including the statement that if a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned, then 3.10.e will be applied.

This change is acceptable because CTS 3.10.f.5 is used to alert the user that a misaligned rod cluster control assembly is covered under a separate LCO requirement. It is an ITS convention to not include these types of cross references. ITS 3.1.4 provides the proper requirements for when a rod is misaligned, and a specific cross reference is not required. This change is designated as administrative as it incorporates an ITS convention with no technical change.

MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 154 of 213

Attachment 1, Volume 6, Rev. 0, Page 155 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 155 of 213

Attachment 1, Volume 6, Rev. 0, Page 156 of 213 CTS Rod Position Indication 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication Individual I DOC LCO 3.1.7 The [Digital] Rod Position Indication ([D]RPI) System and the Demand 1 A02 Position Indication System shall be OPERABLE.

DOC APPLICABILITY: MODES 1 and 2.

A02 ACTIONS

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

3.10.f.1, Separate Condition entry is allowed for each inoperable rod position indicator and each demand 3.10.f.4 position indicator.

CONDITION REQUIRED ACTION COMPLETION TIME I

3.10.f.1, A. One [D]RPI per group A.1 Verify the position of the an Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1 3.10.f.2.c 2 4 inoperable for one or rods with inoperable INSERT 1 3 more groups. position indicators indirectly by using movable incore detectors.

OR A.2 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> POWER to 50% RTP.

I 3.10.f.2 B. More than one [D]RPI B.1 Place the control rods Immediately 1 per group inoperable. under manual control.

AND B.2 Monitor and record Reactor Once per 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Coolant System Tavg.

AND WOG STS 3.1.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 156 of 213

Attachment 1, Volume 6, Rev. 0, Page 157 of 213 3.1.7 3

INSERT 1 AND 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position Insert Page 3.1.7-1 Attachment 1, Volume 6, Rev. 0, Page 157 of 213

Attachment 1, Volume 6, Rev. 0, Page 158 of 213 CTS Rod Position Indication 3.1.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.10.f.2 B.3 Verify the position of the Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3

rods with inoperable position indicators indirectly by using the movable incore detectors.

AND B.4 Restore inoperable position 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3 3 indicators to OPERABLE status such that a I maximum of one [D]RPI per 1 group is inoperable.

C. One or more rods with C.1 Verify the position of the [4] hours inoperable position rods with inoperable indicators have been position indicators indirectly moved in excess of by using movable incore 24 steps in one direction detectors. 3 since the last determination of the OR rod's position.

C.2 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> POWER to 50% RTP.

3.10.f.4 D. One demand position D.1.1 Verify by administrative Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3 I

C indicator per bank C means all [D]RPIs for the 1 inoperable for one or affected banks are 2 more banks. OPERABLE.

AND D.1.2 Verify the most withdrawn Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3 C rod and the least withdrawn rod of the affected banks 2 are 12 steps apart.

within the required rod alignment 4 OR limits of LCO 3.1.4, "Rod Group Alignment Limits."

WOG STS 3.1.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 158 of 213

Attachment 1, Volume 6, Rev. 0, Page 159 of 213 CTS Rod Position Indication 3.1.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.10.f.4 3

D.2 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C POWER to 50% RTP.

3 3.10.f.2.d E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D associated Completion D Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Table 4.1-1 SR 3.1.7.1 Verify each [D]RPI agrees within [12] steps of the Once prior to Item 9 group demand position for the [full indicated range] criticality after of rod travel. each removal of 5

Perform CHANNEL CALIBRATION of each IRPI. the reactor head INSERT 2 18 months WOG STS 3.1.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 159 of 213

Attachment 1, Volume 6, Rev. 0, Page 160 of 213 CTS 3.1.7 0 _INSERT 2 5

Table SR 3.1.7.2 Perform system functional test of the 18 months TS 4.1-1 Item 10 Demand Position Indication System.

Insert Page 3.1.7-3 Attachment 1, Volume 6, Rev. 0, Page 160 of 213

Attachment 1, Volume 6, Rev. 0, Page 161 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.7, ROD POSITION INDICATION

1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
2. The words in ISTS 3.1.7 Required Action A.1, ISTS 3.1.7 Required Action D.1.1 (ITS 3.1.7 Required Action C.1.1), and ISTS 3.1.7 Required Action D.1.2 (ITS 3.1.7 Required Action C.1.2) have been modified to be singular, versus plural, when referring to a rod or a bank. This has been done since the ISTS 3.1.7 ACTIONS Note allows separate Condition entry for each rod position indicator and each demand position indicator; thus the Required Action only applies to the individual rod or bank whose indicator in inoperable. Furthermore, this change was approved during the DC Cook ITS conversion.
3. ISTS 3.1.7 ACTION C has two Required Actions that are connected with an OR.

However, the stated Completion Time for these two Required Actions are different (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, respectively). Due to the convention in the ISTS, as described in Section 1.3, the two Completion Times associated with the two Required Actions OR logical connector must be the same, since either Required Action can be chosen. Therefore, to be consistent with the format of the ISTS, ISTS 3.1.7 ACTION C has been deleted and a new, conditional Completion Time has been added to Required Action A.1. This ensures that the intent of the ISTS is maintained, in that a verification of the position of the rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position. In addition, since the unit is in both Conditions A and B when more than one rod position indicators per group are inoperable, and Required Action A.1 requires the identical position check required by Required Action B.3, there is no reason to include the position check as Required Action B.3. This is also consistent with the format of ISTS.

Furthermore, this change was approved for the DC Cook ITS conversion.

Appropriate renumbering changes have also been made due to these deletions.

4. ISTS Required Action D.1.2 has been revised to be consistent with the current KPS licensing basis requirements. The CTS allows a different alignment criterion for 85% RATED POWER than it does for > 85% RATED POWER. Therefore, ITS 3.1.7 Required Action C.1.2 has been changed to met the alignment criteria specified in CTS 3.10.e (ITS 3.1.4).
5. The ISTS SR 3.1.7.1 requirement to verify each [D]RPI agrees within [12] steps of the group demand position for the full indicated range of rod travel prior to criticality after each removal of the reactor vessel head has been replaced by two new Surveillance Requirements. The first Surveillance is ITS SR 3.1.7.1 which requires performance of a CHANNEL CALIBRATION of each IRPI every 18 months. This Surveillance is acceptable because of the thermal drift characteristics of the IRPIs, performing a full range comparison of IRPI and demand position before criticality is not useful, as the IRPI response will change with IRPI temperature. The second Surveillance is ITS SR 3.1.7.2 which requires performance of a system functional test of the Demand Position Indication System (equivalent to CTS Rod Position Bank Counters) every 18 months. The performance of the CHANNEL CALIBRATION on the IRPIs and the system Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 161 of 213

Attachment 1, Volume 6, Rev. 0, Page 162 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.7, ROD POSITION INDICATION functional test of the Demand Position Indication System will align the ITS with the current licensing bases for Kewaunee Power Station.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 162 of 213

Attachment 1, Volume 6, Rev. 0, Page 163 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 163 of 213

Attachment 1, Volume 6, Rev. 0, Page 164 of 213 Rod Position Indication B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Rod Position Indication USAR, Section 7.1 BASES BACKGROUND According to GDC 13 (Ref. 1), instrumentation to monitor variables and INSERT 1 systems over their operating ranges during normal operation, anticipated 1 operational occurrences, and accident conditions must be OPERABLE.

LCO 3.1.7 is required to ensure OPERABILITY of the control rod position 6 indicators to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

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

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

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

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

The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position Indication System (commonly called group step counters) and the [Digital] 3 Rod Position Indication ([D]RPI) System. Individual I

WOG STS B 3.1.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 164 of 213

Attachment 1, Volume 6, Rev. 0, Page 165 of 213 B 3.1.7 o 1

-INSERT 1 instrumentation and controls shall be provided as required to monitor and maintain within prescribed operating ranges essential reactor operating variables.

Insert Page B 3.1.7-1 Attachment 1, Volume 6, Rev. 0, Page 165 of 213

Attachment 1, Volume 6, Rev. 0, Page 166 of 213 Rod Position Indication B 3.1.7 BASES BACKGROUND (continued)

The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal INSERT 1A to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position 5/8 2 Indication System is considered highly precise (+/- 1 step or +/- e inch). If a 8 rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.

I The [D]RPI System provides a highly accurate indication of actual control 2 rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is 6 steps. To increase the reliability of the system, the inductive coils are INSERT 2 connected alternately to data system A or B. Thus, if one system fails, 2 the [D]RPI will go on half accuracy with an effective coil spacing of 7.5 inches, which is 12 steps. Therefore, the normal indication accuracy of the [D]RPI System is +/- 6 steps (+/- 3.75 inches), and the maximum uncertainty is +/- 12 steps (+/- 7.5 inches). With an indicated deviation of I 3

up to 24 12 steps between the group step counter and [D]RPI, the maximum 5 deviation between actual rod position and the demand position could be 36 24 steps, or 15 inches. 5 2 APPLICABLE Control and shutdown rod position accuracy is essential during power SAFETY operation. Power peaking, ejected rod worth, or SDM limits may be ANALYSES violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM (LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits"). The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, "Rod Group Alignment Limits"). Control rod positions are continuously monitored to provide operators with information that ensures the plant is operating within the bounds of the accident analysis assumptions.

ion ies The control rod position indicator channels satisfy Criterion 2 of 6 10 CFR 50.36(c)(2)(ii). The control rod position indicators monitor control rod position, which is an initial condition of the accident.

WOG STS B 3.1.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 166 of 213

Attachment 1, Volume 6, Rev. 0, Page 167 of 213 B 3.1.7 2

INSERT 1A The readout of the Bank Demand Position Indication System is in the form of an add and subtract counter reading. There is only one display for each rod group.

2 INSERT 2 an analog signal that is proportional to the actual control rod position. This is accomplished by using an electrical coil stack linear variable differential transformer placed above the stepping mechanism of the control rod magnetic jacks which are external to the pressure housing. When the control rod is at the bottom of the core, the magnetic coupling between the primary and secondary windings is small which induces a small voltage in the secondary. As the control rod raises, the lift rod causes an increase in magnetic coupling which in turn provides an analog signal proportional to the control rod position. The individual rod position indicator channels are sufficiently accurate to detect a rod + 12 steps away from its demand position.

Insert Page B 3.1.7-2 Attachment 1, Volume 6, Rev. 0, Page 167 of 213

Attachment 1, Volume 6, Rev. 0, Page 168 of 213 Rod Position Indication B 3.1.7 BASES the I 3

LCO LCO 3.1.7 specifies that one [D]RPI System and one Bank Demand 6

Position Indication System be OPERABLE for each control rod. For the control rod position indicators to be OPERABLE requires meeting the SR 6 of the LCO and the following:

a. The [D]RPI System indicates within 12 steps of the group step 7

counter demand position as required by LCO 3.1.4, "Rod Group Alignment Limits,"

I a b. For the [D]RPI System there are no failed coils, and 2 3 4 b c. The Bank Demand Indication System has been calibrated either in the fully inserted position or to the [D]RPI System. 3 I

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

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

These requirements ensure that control rod position indication during 6 power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged.

rod 6

OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned control rods can be detected. Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits.

I APPLICABILITY The requirements on the [D]RPI and step counters are only applicable in 3 MODES 1 and 2 (consistent with LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6),

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

WOG STS B 3.1.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 168 of 213

Attachment 1, Volume 6, Rev. 0, Page 169 of 213 Rod Position Indication B 3.1.7 BASES ACTIONS The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator.

A.1 I in a 3

When one [D]RPI channel per group fails, the position of the rod may still 6 be determined indirectly by use of the movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that FQ satisfies LCO 3.2.1, F N H satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks.

5 If a bank has been significantly moved, the Required Action of C.1 or C.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

INSERT 3 5 A.2 Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 3). 2 2

The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is reasonable, based on operating experience, for reducing power to 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above.

3 5

B.1, B.2, B.3, and B.4 in a s I

When more than one [D]RPI per group fail, additional actions are 3 6 necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion will not occur. Together with the indirect position determination available via WOG STS B 3.1.7-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 169 of 213

Attachment 1, Volume 6, Rev. 0, Page 170 of 213 B 3.1.7 5

INSERT 3 If a rod has been significantly moved (in excess of 24 steps in one direction, since the position was last determined), Required Action A.1 is still appropriate but must be initiated promptly to begin verifying that the rod is still properly positioned, relative to their group positions. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides an acceptable period of time to verify the rod position with inoperable position indicator indirectly by using movable incore detectors. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." In this Required Action, the Completion Time only begins on discovery that both:

a. One RPI per group inoperable for one or more groups; and
b. A rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rods position.

If at any time during the existence of Condition A (one RPI per group inoperable for one or more groups) a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rods position, this Completion Time begins to be tracked.

Insert Page B 3.1.7-4 Attachment 1, Volume 6, Rev. 0, Page 170 of 213

Attachment 1, Volume 6, Rev. 0, Page 171 of 213 Rod Position Indication B 3.1.7 BASES ACTIONS (continued) movable incore detectors will minimize the potential for rod misalignment.

The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition. 8 Monitoring and recording reactor coolant Tavg help assure that significant changes in power distribution and SDM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions.

(Required Action A.1)

The position of the rods may be determined indirectly by use of the 6

movable incore detectors. The Required Action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that FQ satisfies LCO 3.2.1, F N H satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been 5

moved. Verification of control rod position once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is adequate INSERT 4 for allowing continued full power operation for a limited, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time provides sufficient time to troubleshoot and restore the I [D]RPI system to operation while avoiding the plant challenges 3 associated with the shutdown without full rod position indication.

Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly 5 moved, the Required Action of C.1 or C.2 below is required.

C.1 and C.2 These Required Actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2, [or B.1, as applicable] are still appropriate but must be initiated promptly under Required Action C.1 to begin verifying 5 that these rods are still properly positioned, relative to their group positions.

If, within [4] hours, the rod positions have not been determined, THERMAL POWER must be reduced to 50% RTP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to avoid undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps. The allowed Completion Time of [4] hours provides an acceptable period of time to verify the rod positions.

WOG STS B 3.1.7-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 171 of 213

Attachment 1, Volume 6, Rev. 0, Page 172 of 213 B 3.1.7 o-5 INSERT 4 and once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a rod with an inoperable position indicator has been moved in excess of 24 steps in one direction since the last determination of the rod's position Insert Page B 3.1.7-5 Attachment 1, Volume 6, Rev. 0, Page 172 of 213

Attachment 1, Volume 6, Rev. 0, Page 173 of 213 Rod Position Indication B 3.1.7 BASES ACTIONS (continued) C 5

D.1.1 and D.1.2 I

With one demand position indicator per bank inoperable, the rod positions can be determined by the [D]RPI System. Since normal power operation 3 does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are 12 steps apart 5 within the allowed Completion Time of once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is adequate.

within the required rod alignment limits of LCO 3.1.4 C

D.2 5 Reduction of THERMAL POWER to 50% RTP puts the core into a 2

condition where rod position is not significantly affecting core peaking factor limits (Ref. 3). The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provides 2 an acceptable period of time to verify the rod positions per Required Actions C.1.1 and C.1.2 or reduce power to 50% RTP.

D E.1 and 5

any is not met If the Required Actions cannot be completed within the associated 6 Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.7.1 REQUIREMENTS Verification that the [D]RPI agrees with the demand position within

[12] steps ensures that the [D]RPI is operating correctly. Since the INSERT 5

[D]RPI does not display the actual shutdown rod positions between 18 5 and 210 steps, only points within the indicated ranges are required in comparison.

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

WOG STS B 3.1.7-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 173 of 213

Attachment 1, Volume 6, Rev. 0, Page 174 of 213 B 3.1.7 5

INSERT 5 SR 3.1.7.1 is the performance of a CHANNEL CALIBRATION for each IRPI channel. The calibration verifies the accuracy of each IRPI channel.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for unnecessary plant transients if the SR were performed with the reactor at power. Operating experience has shown these components usually pass the SR when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.1.7.2 SR 3.1.7.2 is the performance of a system functional test of the Demand Position Indication System. The system functional test verifies the OPERABILITY of the Demand Position Indication System.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for unnecessary plant transients if the SR were performed with the reactor at power. Operating experience has shown these components usually pass the SR when performed at a Frequency of once every 18 months. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Insert Page B 3.1.7-6 Attachment 1, Volume 6, Rev. 0, Page 174 of 213

Attachment 1, Volume 6, Rev. 0, Page 175 of 213 Rod Position Indication B 3.1.7 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 13.

U USAR, Section 7.1, General Design Criteria12, 2

2. FSAR, Chapter [15]. 14 "Instrumentation and Control Systems."
3. FSAR, Chapter [15].

WOG STS B 3.1.7-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 175 of 213

Attachment 1, Volume 6, Rev. 0, Page 176 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.7 BASES, ROD POSITION INDICATION

1. ISTS 3.1.7 Bases Background references General Design Criteria. Kewaunee Power Station (KPS) was designed prior to promulgation of 10 CFR 50, Appendix A.

Therefore, ITS 3.1.7 Bases Background has been revised to discuss the design standards used by KPS. Additionally, bases references to 10 CFR 50, Appendix A have been replaced with references to the appropriate section of the USAR.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
4. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
5. The Bases are changed to reflect changes made to the Specification.
6. Changes are made to reflect the actual Specification.
7. The requirement that the IRPI indicates within the agreement limit of the group step counter demand position has been deleted since the requirement is already covered by ITS LCO 3.1.4. If the agreement limit is not met, then the ACTIONS of LCO 3.1.4, "Rod Group Alignment Limits," should be entered. As written in these Bases, both the ACTIONS of ITS LCO 3.1.4 and ITS LCO 3.1.7 would have to be entered if not within the agreement limit. The appropriate ACTIONS are those of ITS LCO 3.1.4.

ITS LCO 3.1.7 should only cover the actual IRPI System, not the agreement limits.

8. Typographical error corrected.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 176 of 213

Attachment 1, Volume 6, Rev. 0, Page 177 of 213 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 6, Rev. 0, Page 177 of 213

Attachment 1, Volume 6, Rev. 0, Page 178 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.7, ROD POSITION INDICATION There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 178 of 213

Attachment 1, Volume 6, Rev. 0, Page 179 of 213 ATTACHMENT 8 ITS 3.1.8, PHYSICS TESTS EXCEPTIONS - MODE 2 Attachment 1, Volume 6, Rev. 0, Page 179 of 213

, Volume 6, Rev. 0, Page 180 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 180 of 213

Attachment 1, Volume 6, Rev. 0, Page 181 of 213 ITS A01 ITS 3.1.8

f. Minimum Conditions for Criticality Add proposed suspension of LCO 3.4.2 requirement L01
1. The reactor shall not be brought to a critical condition until the pressure-temperature See ITS 3.4.3 state is to the right of the criticality limit line shown in Figure TS 3.1-1.
2. The reactor shall be maintained subcritical by at least 1% k/k until normal water See ITS 3.4.9 level is established in the pressurizer.

LCO 3.1.8 and

3. When the reactor is critical the moderator temperature coefficient shall be as initiated in MODE 2 Applicability specified in the COLR, except during LOW POWER PHYSICS TESTING. The maximum upper moderator temperature coefficient limit shall be 5 pcm/°F for A03 power levels 60% RATED POWER and 0 pcm/ºF for power levels > 60% RATED POWER.
4. If the limits of 3.1.f.3 cannot be met, then power operation may continue provided the following actions are taken:

See ITS 3.1.3 A. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, develop and maintain administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in TS 3.1.f.3. These withdrawal limits shall be in addition to the insertion limits specified in TS 3.10.d.

B. If the actions specified in TS 3.1.f.4.A are not satisfied, then be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Add proposed LCO 3.1.8.a L01 Add proposed LCO 3.1.8.b and LCO 3.1.8.c M01 During performance of PHYSIC TESTS, the number of required channels for LCO 3.3.1, Functions 2, 3, 6, and 18.e may be reduced to 3. A02 Add proposed ACTIONS A and B M01 Add proposed ACTIONS C and D L01 Add proposed SR 3.1.8.1 L01 Add proposed SR 3.1.8.2 and SR 3.1.8.3 M01 Amendment No. 165 TS 3.1-10 03/11/2003 Page 1 of 5 Attachment 1, Volume 6, Rev. 0, Page 181 of 213

Attachment 1, Volume 6, Rev. 0, Page 182 of 213 ITS A01 ITS 3.1.8 C. Quadrant Power Tilt Limits

1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within two hours:

A. Eliminate the tilt.

B. Restrict maximum core power level 2% for every 1% of indicated power tilt ratio

> 1.0.

2. If the tilt condition is not eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then reduce power to 50% or lower.

See ITS

3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is 3.2.4 simultaneous indication of a misaligned rod:

A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.

B. If the tilt condition is not eliminated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, then the reactor shall be brought to a minimum load condition ( 30 Mwe).

4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition

( 5% reactor power).

d. Rod Insertion Limits
1. The shutdown rods shall be withdrawn to within the limits, specified in the COLR, when See ITS the reactor is critical or approaching criticality. 3.1.5
2. The control banks shall be limited in physical insertion; insertion limits are specified in the COLR. If any one of the control bank insertion limits is not met:

A. Within one hour, initiate boration to restore control bank insertion to within the limits specified in the COLR, and B. Restore control bank insertion to within the limits specified in the COLR within See ITS 3.1.6 two hours of exceeding the insertion limits.

C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then withinone hour action shall be initiated to:

- Achieve HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> LCO 3.1.8

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> A03 and initiated in MODE 2 Applicability 3. Insertion limit does not apply during physics tests or during periodic exercise of See ITS 3.1.5 and individual rods. However, the shutdown margin, as specified in the COLR, must be 3.1.6 LCO 3.1.8 maintained except for the Low Power Physics Test to measure control rod worth and and shutdown margin. For this test, the reactor may be critical with all but one high worth Applicability rod inserted.

Amendment No. 165 TS 3.10-4 03/11/2003 Page 2 of 5 Attachment 1, Volume 6, Rev. 0, Page 182 of 213

Attachment 1, Volume 6, Rev. 0, Page 183 of 213 ITS A01 ITS 3.1.8

e. Rod Misalignment Limitations NOTE: Individual RPIs may be outside their limits for < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following substantial rod movement This specification defines allowable limits for misaligned rod cluster control assemblies. In See ITS TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may 3.1.4 be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or LCO 3.1.8 through the conditioning module output voltage via a correlation of rod position vs. voltage.

and Rod misalignment limitations do not apply during physics testing.

Applicability initiated in MODE 2 A03

1. When reactor power is 85% of rating, the rod cluster control assembly shall be maintained within +/- 12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is 85%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating.
2. When reactor power is < 85% but 50% of rating, the rod cluster control assemblies See ITS 3.1.4 shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is

< 85% but 50%, the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating.

3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within eight hours, the rod shall be declared inoperable.

See ITS

f. Inoperable Rod Position Indicator Channels 3.1.7 NOTE: Individual RPIs may be outside their limits for < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following substantial rod See ITS 3.1.4 movement
1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate entry condition is allowed for each inoperable individual rod position indicator.)

A. Verify the position of the rod cluster control indirectly by movable incore detectors each eight hours, or B. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reduce reactor thermal power to 50% of RATED POWER. See ITS 3.1.7

2. If more than one individual rod position indicator channel per group are inoperable, then:

A. IMMEDIATELY place the control rods in manual, and B. Once per 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, monitor and record RCS Tavg, and C. Verify the position of the rod by movable incore detectors each 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and Amendment No. 181 TS 3.10-5 03/24/2005 Page 3 of 5 Attachment 1, Volume 6, Rev. 0, Page 183 of 213

ITS 3.1.8 A01 See ITS 3.3.1 TABLE TS 3.5-2 INSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP 1 2 3 4 5 6 NO. OF MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF NO. OF CHANNELS TO OPERABLE DEGREE OF BYPASS CONDITIONS OF COLUMN 3 NO. FUNCTIONAL UNIT CHANNELS TRIP CHANNELS REDUNDANCY CONDITIONS OR 4 CANNOT BE MET 1 Manual 2 1 1 - Maintain HOT SHUTDOWN Nuclear Flux Power 2 LCO 3.1.8 Range(1)

Low setting 4 2 3 1 P-10 Maintain HOT SHUTDOWN High setting 4 2 3 1 Positive rate 4 2 3 1 Negative rate 4 2 3 1 3 Nuclear Flux 2 1 1 - P-10 Maintain HOT SHUTDOWN(3)

Intermediate Range Nuclear Flux 4 2 1 1 - P-6 Maintain HOT SHUTDOWN(3)

Source Range 5 Overtemperature T 4 2 3 1 Maintain HOT SHUTDOWN 6 Overpower T 4 2 3 1 Maintain HOT SHUTDOWN 7 Low Pressurizer Pressure 4 2 3 1 P-7 Maintain HOT SHUTDOWN 8 High Pressurizer Pressure 3 2 2 - Maintain HOT SHUTDOWN , Volume 6, Rev. 0, Page 184 of 213 Attachment 1, Volume 6, Rev. 0, Page 184 of 213 Amendment No. 137 Page 1 of 4 06/09/98 Page 4 of 5

A01 ITS 3.1.8 TABLE TS 3.5-2 INSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP NOTES (1) One additional channel may be taken out of service for zero power physics testing. M02 (2) Deleted (3) When a block condition exists, maintain normal operation.

(4) Underfrequency on the 4-kV buses trips the Reactor Coolant Pump breakers, which in turn trips the reactor when power is above P-7.

Permissive/Interlock Channels Coincidence Setting Limit P-6 Intermediate Range Nuclear Instrument 1 of 2 > 10-5% RATED POWER Power Range Nuclear Instrument 3 of 4 12.2% RATED POWER P-7 AND Turbine Impulse Pressure 2 of 2 12.2% RATED POWER (a)

P-8 Power Range Nuclear Instrument 3 of 4 < 10% RATED POWER P-10 Power Range Nuclear Instrument 2 of 4 7.8% RATED POWER (a)

Setting Limit is converted to an equivalent turbine impulse pressure See ITS 3.3.1 , Volume 6, Rev. 0, Page 185 of 213 Attachment 1, Volume 6, Rev. 0, Page 185 of 213 Amendment No. 195 Page 4 of 4 03/28/2008 Page 5 4 of 5 15

Attachment 1, Volume 6, Rev. 0, Page 186 of 213 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 ADMINISTRATIVE CHANGES A01 In the conversion of the Kewaunee Power Station (KPS) Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1431, Rev.

3.0, "Standard Technical Specifications-Westinghouse Plants" (ISTS).

These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS.

A02 CTS 3.1.f.3, CTS 3.10.d.3, and CTS 3.10.e state that the limitations of certain specifications may be suspended during the performance of PHYSICS TESTS.

CTS Table TS 3.5-2, Functional Units 2, Nuclear Flux Power Range, states the minimum OPERABLE channel requirement is 3, except as modified by Note (1).

ITS LCO 3.1.8 includes an allowance to reduce the required number of channels for ITS LCO 3.3.1, "RPS Instrumentation," Functions 2 (Power Range Neutron Flux), 3 (Power Range Neutron Flux Rate), 6 (Overtemperature T), and 16.e (Power Range Neutron Flux, P-10) from "4" to "3" during PHYSICS TESTS. This changes the CTS by stating the specific number of RPS channels from that have to be OPERABLE during PHYSICS TESTS.

The purpose of CTS 3.1.f.3, CTS 3.10.d.3, and CTS 3.10.e is to allow some flexibility during the performance of PHYSIC TESTS. This change is acceptable because the minimum number of channels required for OPERABILITY for the Reactor Trip Functions in CTS Table TS 3.5-2 is currently "3." This allowance is needed since the Required Channels in ITS 3.3.1, Reactor Trip System Instrumentation, is "4." This change from the CTS is discussed in the Discussion of Changes for ITS 3.3.1. Furthermore, the exception identified in CTS Table TS 3.5-2 Note (1) is discussed in DOC M02. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS 3.1.f.3 states that the Moderator Temperature Coefficient requirement specified in the COLR may be suspended during the performance of LOW POWER PHYSICS TESTING. CTS 3.10.d.3 states that the Rod Insertion Limits do not apply during the performance of PHYSICS TESTS. CTS 3.10.e states that the Rod Misalignment Limitations do not apply during the performance of PHYSICS TESTS. ITS 3.1.8 is applicable "During PHYSICS TESTS initiated in MODE 2." This changes the CTS such that the Specification is applicable in MODE 2 only when a PHYSICS TEST is initiated.

The purpose of ITS 3.1.8 Applicability is to ensure that the Actions contained in the Specification are followed. The CTS applicability does not specifically state what MODE the PHYSICS TEST is performed in. The ITS Applicability states emphatically the MODE that the PHYSICS TEST will be performed. This change is designated as administrative because it clarifies the current wording in the Specification with no change in intent.

Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 6, Rev. 0, Page 186 of 213

Attachment 1, Volume 6, Rev. 0, Page 187 of 213 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 MORE RESTRICTIVE CHANGES M01 TS 3.1.f, CTS 3.10.d.3, and CTS 3.10.e state that the limitations of certain specifications may be suspended during the performance of PHYSICS TESTS but does not provide restrictions that must be followed when utilizing the CTS exceptions. ITS 3.1.8 provides the requirements and restrictions for performing testing during PHYSICS TESTS initiated in MODE 2. A Surveillance (SR 3.1.8.3) to verify the SHUTDOWN MARGIN is within limits specified in the COLR every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and an ACTION (ACTION A) to follow if the SHUTDOWN MARGIN is not met are added. Additionally, a Surveillance (SR 3.1.8.2) to verify that THERMAL POWER is 5% RTP every 30 minutes and an ACTION (ACTION B) to follow if the THERMAL POWER is not within limit (i.e., it is not 5% RTP) has been added. This changes the CTS by imposing additional requirements on the application of the test exception LCO.

This change is acceptable because it imposes reasonable restrictions on the performance of PHYSICS TESTS when the control rod and RCS minimum temperature Specifications are allowed to be violated. When the control rod and RCS minimum temperature Specifications are allowed to be violated, additional actions must be taken to ensure that sufficient SHUTDOWN MARGIN is available to shutdown the reactor and keep it subcritical if needed when in MODE 2 with keff > 1.0. This change is designated as more restrictive because it imposes additional restrictions not found in the CTS.

M02 CTS Table TS 3.5-2, Functional Unit 2, Nuclear Flux Power Range, requires three channels to be OPERABLE. However, Note (1) allows one additional channel to be taken out of service during zero power testing. Thus, during zero power PHYSICS TESTS, the CTS only requires two Functional Unit 2 channels to be OPERABLE. ITS LCO 3.1.8, in part, includes an allowance to reduce the required number of channels for ITS LCO 3.3.1, "RPS Instrumentation,"

Functions 2 (Power Range Neutron Flux) and 3 (Power Range Neutron Flux Rate) from "4" to "3" during PHYSICS TESTS. This changes the CTS by always requiring three channels of Nuclear Flux Power Range be OPERABLE during all MODE 2 PHYSICS TESTS, even zero power PHYSICS TESTS.

The purpose of CTS Table TS 3.5-2 Note (1) is to allow flexibility during performance of zero power PHYSICS TESTS. However, there are four installed Nuclear Flux Power Range channels, and ITS LCO 3.1.8 already allows one of them to be inoperable during PHYSICS TESTS. Thus, flexibility is being provided in LCO 3.1.8, and an additional channel being inoperable is not necessary. This change is designated as more restrictive since an additional channel is required during zero power PHYSICS TESTS.

RELOCATED SPECIFICATIONS None Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 6, Rev. 0, Page 187 of 213

Attachment 1, Volume 6, Rev. 0, Page 188 of 213 DISCUSSION OF CHANGES ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.10.f.1 (ITS 3.4.2, " RCS Minimum Temperature for Criticality") states that the reactor shall not be brought to a critical condition until the pressure-temperature state is to the right of the criticality limit line shown in Figure TS 3.1-1. ITS 3.1.8 provides an exception to LCO 3.4.2, "RCS Minimum Temperature for Criticality," provided the RCS lowest loop average temperature is 530ºF. A Surveillance Requirement (SR 3.1.8.1) to verify the RCS lowest loop average temperature is 530ºF every 30 minutes has been added. In addition, ACTION C has been added to cover the situation when RCS lowest loop average temperature is not within limit. The Required Action is to restore RCS lowest loop average temperature to within limit within 15 minutes. If this is not met, then ACTION D requires the unit to be in MODE 3 within 15 minutes. This changes the CTS by allowing the suspension of LCO 3.4.2, "RCS Minimum Temperature for Criticality." However, it places a limitation on the RCS lowest loop average temperature.

This change is acceptable because the LCO requirements continue to ensure that the process variables are maintained consistent with the safety analyses and licensing basis. This changes the CTS by allowing the suspension of LCO 3.4.2, "RCS Minimum Temperature for Criticality." However, it places a limitation on the RCS lowest loop average temperature. The limit on RCS average temperature is provided in the test exception LCO to ensure that the RCS temperature stays within the analyzed range. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS.

Kewaunee Power Station Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 188 of 213

Attachment 1, Volume 6, Rev. 0, Page 189 of 213 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 189 of 213

Attachment 1, Volume 6, Rev. 0, Page 190 of 213 CTS PHYSICS TESTS Exceptions - MODE 2 3.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 PHYSICS TESTS Exceptions - MODE 2 LCO 3.1.8 During the performance of PHYSICS TESTS, the requirements of:

3.1.f LCO 3.1.3, "Moderator Temperature Coefficient,"

3.10.e LCO 3.1.4, "Rod Group Alignment Limits,"  ;

3.10.d LCO 3.1.5, "Shutdown Bank Insertion Limits," 1 3.10.d LCO 3.1.6, "Control Bank Insertion Limits," and  ;

DOC L01 LCO 3.4.2, "RCS Minimum Temperature for Criticality" DOC A02 may be suspended and the number of required channels for LCO 3.3.1, RPS "RTS Instrumentation," Functions 2, 3, 6 and 18.e, may be reduced to 3 4 required channels, provided: 16 530 DOC L01 a. RCS lowest loop average temperature is [531]°F, 2

1 DOC M01 b. SDM is within the limits specified in the COLR, and DOC M01 c. THERMAL POWER is < 5% RTP.

3.1.f, 3.10.e, 3.10.d, APPLICABILITY: During PHYSICS TESTS initiated in MODE 2.

DOC L01 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC M01 A. SDM not within limit. A.1 Initiate boration to restore 15 minutes SDM to within limit.

AND A.2 Suspend PHYSICS TESTS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> exceptions.

DOC M01 B. THERMAL POWER not B.1 Open reactor trip breakers. Immediately within limit.

DOC L01 C. RCS lowest loop C.1 Restore RCS lowest loop 15 minutes average temperature not average temperature to within limit. within limit.

WOG STS 3.1.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 190 of 213

Attachment 1, Volume 6, Rev. 0, Page 191 of 213 CTS PHYSICS TESTS Exceptions - MODE 2 3.1.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME DOC L01 D. Required Action and D.1 Be in MODE 3. 15 minutes associated Completion Time of Condition C not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform a CHANNEL OPERATIONAL TEST on Prior to initiation power range and intermediate range channels per of PHYSICS 3

[SR 3.3.1.7, SR 3.3.1.8, and Table 3.3.1-1]. TESTS DOC L01 SR 3.1.8.2 Verify the RCS lowest loop average temperature is 30 minutes 3 1 [531]°F. 2 530 DOC M01 SR 3.1.8.3 Verify THERMAL POWER is < 5% RTP. 30 minutes 3 2

DOC M01 SR 3.1.8.4 Verify SDM is within the limits specified in the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3 3 COLR.

WOG STS 3.1.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 191 of 213

Attachment 1, Volume 6, Rev. 0, Page 192 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2

1. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
2. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
3. ISTS SR 3.1.8.1 requires a CHANNEL OPERATIONAL TEST be performed on the Intermediate and Power Range Channels "prior to initiation of PHYSICS TESTS." However, no finite time as to how soon prior to the PHYSICS TESTS is stated. The ITS Applicability for the Intermediate and Power Range channels includes MODE 2, thus the normal, periodic Frequencies for SR 3.3.1.7 and SR 3.3.1.8 must be met prior to entering MODE 2. Therefore, the normal periodic Frequencies already ensure the "prior to initiation of PHYSICS TEST" is met, and ISTS SR 3.1.8.1 is not necessary and has been deleted. Due to this deletion, the remaining SRs have been renumbered.
4. Changes made to be consistent with changes made to LCO 3.3.1.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 192 of 213

Attachment 1, Volume 6, Rev. 0, Page 193 of 213 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 6, Rev. 0, Page 193 of 213

Attachment 1, Volume 6, Rev. 0, Page 194 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 PHYSICS TESTS Exceptions - MODE 2 BASES BACKGROUND The primary purpose of the MODE 2 PHYSICS TESTS exceptions is to permit relaxations of existing LCOs to allow certain PHYSICS TESTS to be performed.

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

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

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

a. Ensure that the facility has been adequately designed, 1
b. Validate the analytical models used in the design and analysis, 1
c. Verify the assumptions used to predict unit response, 1
d. Ensure that installation of equipment in the facility has been accomplished in accordance with the design, and 1
e. Verify that the operating and emergency procedures are adequate.

To accomplish these objectives, testing is performed prior to initial criticality, during startup, during low power operations, during power ascension, at high power, and after each refueling. The PHYSICS TESTS requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions and that the core can be operated as designed (Ref. 4).

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

WOG STS B 3.1.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 194 of 213

Attachment 1, Volume 6, Rev. 0, Page 195 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES BACKGROUND (continued)

The PHYSICS TESTS required for reload fuel cycles (Ref. 4) in MODE 2 are listed below:

a. Critical Boron Concentration - Control Rods Withdrawn, 1
b. Critical Boron Concentration - Control Rods Inserted, 2 1 Differential Boron Worth
c. Control Rod Worth, 1
d. Isothermal Temperature Coefficient (ITC), and 1
e. Neutron Flux Symmetry. 2 The first four tests are performed in MODE 2, and the last test can be 2 performed in either MODE 1 or 2. These and other supplementary tests may be required to calibrate the nuclear instrumentation or to diagnose operational problems. These tests may cause the operating controls and process variables to deviate from their LCO requirements during their performance.

[ a. The Critical Boron Concentration - Control Rods Withdrawn Test 3 measures the critical boron concentration at hot zero power (HZP).

With all rods out, the lead control bank is at or near its fully withdrawn position. HZP is where the core is critical (keff = 1.0), and the Reactor Coolant System (RCS) is at design temperature and pressure for zero power. Performance of this test should not violate any of the referenced LCOs.

b. The Critical Boron Concentration - Control Rods Inserted Test measures the critical boron concentration at HZP, with a bank having a worth of at least 1% k/k when fully inserted into the core. This test is used to measure the boron reactivity coefficient. With the core at HZP and all banks fully withdrawn, the boron concentration of the reactor coolant is gradually lowered in a continuous manner. The INSERT 1 3 selected bank is then inserted to make up for the decreasing boron concentration until the selected bank has been moved over its entire range of travel. The reactivity resulting from each incremental bank movement is measured with a reactivity computer. The difference between the measured critical boron concentration with all rods fully withdrawn and with the bank inserted is determined. The boron WOG STS B 3.1.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 195 of 213

Attachment 1, Volume 6, Rev. 0, Page 196 of 213 B 3.1.8 1

INSERT 1 The Differential Boron Worth Test determines if the measured differential boron worth is consistent with the predicted value. With the core at HZP, the change in equilibrium boron concentration is determined at different rod bank positions. As the rod bank or banks are moved, the reactivity change is measured using a reactivity computer. The measured reactivity change is divided by the difference in measured critical boron concentrations to determine the differential boron worth.

Insert Page B 3.1.8-2 Attachment 1, Volume 6, Rev. 0, Page 196 of 213

Attachment 1, Volume 6, Rev. 0, Page 197 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES BACKGROUND (continued) reactivity coefficient is determined by dividing the measured bank 3 worth by the measured boron concentration difference. Performance of this test could violate LCO 3.1.4, "Rod Group Alignment Limits,"

LCO 3.1.5, "Shutdown Bank Insertion Limit," or LCO 3.1.6, "Control Bank Insertion Limits."

c. The Control Rod Worth Test is used to measure the reactivity worth of selected control banks. This test is performed at HZP and has three alternative methods of performance. The first method, the Boron Exchange Method, varies the reactor coolant boron concentration and moves the selected control bank in response to the changing boron concentration. The reactivity changes are measured with a reactivity computer. This sequence is repeated for the remaining control banks. The second method, the Rod Swap Method, measures the worth of a predetermined reference bank using the Boron Exchange Method above. The reference bank is then nearly fully inserted into the core. The selected bank is then inserted into the core as the reference bank is withdrawn. The HZP critical conditions are then determined with the selected bank fully inserted into the core. The worth of the selected bank is inferred, based on the position of the reference bank with respect to the selected bank. This sequence is repeated as necessary for the remaining control banks. The third method, the Boron Endpoint Method, moves the selected control bank over its entire length of travel and then varies the reactor coolant boron concentration to achieve HZP criticality again. The difference in boron concentration is the worth of the selected control bank. This sequence is repeated for the remaining control banks. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6.
d. The ITC Test measures the ITC of the reactor. This test is performed at HZP and has two methods of performance. The first method, the Slope Method, varies RCS temperature in a slow and continuous manner. The reactivity change is measured with a reactivity computer as a function of the temperature change. The ITC is the slope of the reactivity versus the temperature plot. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. The second method, the Endpoint Method, changes the RCS temperature and measures the reactivity at the beginning and end of the WOG STS B 3.1.8-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 197 of 213

Attachment 1, Volume 6, Rev. 0, Page 198 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES BACKGROUND (continued) temperature change. The ITC is the total reactivity change divided by the total temperature change. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. Performance of this test could violate LCO 3.4.2, "RCS Minimum Temperature for Criticality."

e. The Flux Symmetry Test measures the degree of azimuthal symmetry of the neutron flux at as low a power level as practical, depending on the test method employed. This test can be performed at HZP (Control Rod Worth Symmetry Method) or at 30% RTP (Flux Distribution Method). The Control Rod Worth Symmetry Method inserts a control bank, which can then be withdrawn to compensate for the insertion of a single control rod from a symmetric 3 set. The symmetric rods of each set are then tested to evaluate the symmetry of the control rod worth and neutron flux (power distribution). A reactivity computer is used to measure the control rod worths. Performance of this test could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6. The Flux Distribution Method uses the incore flux detectors to measure the azimuthal flux distribution at selected locations with the core at 30% RTP. ]

APPLICABLE The fuel is protected by LCOs that preserve the initial conditions of the SAFETY core assumed during the safety analyses. The methods for development ANALYSES of the LCOs that are excepted by this LCO are described in the 2 Reference 5 Westinghouse Reload Safety Evaluation Methodology Report (Ref. 5).

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

U The FSAR defines requirements for initial testing of the facility, including 13.3-1 2 PHYSICS TESTS. Tables [14.1-1 and 14.1-2] summarize the zero, low power, and power tests. Requirements for reload fuel cycle PHYSICS s TESTS are defined in ANSI/ANS-19.6.1-1985 (Ref. 4). Although these PHYSICS TESTS are generally accomplished within the limits for all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. When one or more of the requirements specified in LCO 3.1.3, "Moderator Temperature Coefficient (MTC)," LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 are suspended for PHYSICS TESTS, the fuel design criteria are preserved as long as the power level is limited to 5% RTP, the reactor coolant temperature is kept 531°F, and SDM is within the limits 4 provided in the COLR. 530 WOG STS B 3.1.8-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 198 of 213

Attachment 1, Volume 6, Rev. 0, Page 199 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES APPLICABLE SAFETY ANALYSES (continued)

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

Among the process variables involved are AFD and QPTR, which represent initial conditions of the unit safety analyses. Also involved are the movable control components (control and shutdown rods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR.

As described in LCO 3.0.7, compliance with Test Exception LCOs is optional, and therefore no criteria of 10 CFR 50.36(c)(2)(ii) apply. Test Exception LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

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

LCO This LCO allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits. One power range neutron flux channel may be bypassed, reducing the number of required channels from 4 to 3. Operation beyond specified limits is permitted for the purpose of performing PHYSICS TESTS and poses no threat to fuel integrity, provided the SRs are met.

The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 may be suspended and the number of required channels 16 RPS for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6 and 18.e may be 4 reduced to 3 required channels during the performance of PHYSICS TESTS provided:

530

a. RCS lowest loop average temperature is [531]°F, 3 1
b. SDM is within the limits provided in the COLR, and 1
c. THERMAL POWER is 5% RTP.

WOG STS B 3.1.8-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 199 of 213

Attachment 1, Volume 6, Rev. 0, Page 200 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES APPLICABILITY This LCO is applicable when performing low power PHYSICS TESTS.

The Applicability is stated as "during PHYSICS TESTS initiated in MODE 2" to ensure that the 5% RTP maximum power level is not exceeded. Should the THERMAL POWER exceed 5% RTP, and consequently the unit enter MODE 1, this Applicability statement prevents exiting this Specification and its Required Actions.

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

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

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

C.1 530 When the RCS lowest Tavg is < 531°F, the appropriate action is to restore 4 Tavg to within its specified limit. The allowed Completion Time of 15 minutes provides time for restoring Tavg to within limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Operation with the reactor critical and with temperature below 530 4 531°F could violate the assumptions for accidents analyzed in the safety analyses.

WOG STS B 3.1.8-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 200 of 213

Attachment 1, Volume 6, Rev. 0, Page 201 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES ACTIONS (continued)

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

SURVEILLANCE SR 3.1.8.1 REQUIREMENTS The power range and intermediate range neutron detectors must be verified to be OPERABLE in MODE 2 by LCO 3.3.1, "Reactor Trip 4 System (RTS) Instrumentation." A CHANNEL OPERATIONAL TEST is performed on each power range and intermediate range channel prior to initiation of the PHYSICS TESTS. This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS.

SR 3.1.8.2 1 530 4

Verification that the RCS lowest loop Tavg is 531°F will ensure that the 4 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 3.1.8.3 2 4 Verification that the THERMAL POWER is 5% RTP will ensure that the plant is not operating in a condition that could invalidate the safety analyses. Verification of the THERMAL POWER 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.

WOG STS B 3.1.8-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 201 of 213

Attachment 1, Volume 6, Rev. 0, Page 202 of 213 PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.1.8.4 3 4 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:

a. RCS boron concentration, 1
b. Control bank position, 1
c. RCS average temperature, 1
d. Fuel burnup based on gross thermal energy generation, 1
e. Xenon concentration, 1
f. Samarium concentration, 1
g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH), 1
h. Moderate defect, when above the POAH, and 1
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.

The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on the generally slow change in required boron concentration and on the low probability of an accident occurring without the required SDM.

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-19.6.1-1985, December 13, 1985.
5. WCAP-9273-NP-A, "Westinghouse Reload Safety Evaluation 2 Methodology Report," July 1985.
6. WCAP-11618, including Addendum 1, April 1989.

DOM-NAF-5, Revision 0.0, Application of Dominion Nuclear Core Design and Safety Analysis Methods to the Kewaunee Power Station, July 2006.

WOG STS B 3.1.8-8 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 202 of 213

Attachment 1, Volume 6, Rev. 0, Page 203 of 213 JUSTIFICATION FOR DEVIATIONS ITS 3.1.8 BASES, PHYSICS TESTS Exceptions - MODE 2

1. The punctuation corrections have been made consistent with the Writer's Guide from the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.
2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.
3. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and the proper plant specific information/value is provided. This is acceptable since the generic specific information/value is revised to reflect the current plant design.
4. The Bases are changed to reflect changes made to the Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 203 of 213

Attachment 1, Volume 6, Rev. 0, Page 204 of 213 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 6, Rev. 0, Page 204 of 213

Attachment 1, Volume 6, Rev. 0, Page 205 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.8, PHYSICS TESTS Exceptions - MODE 2 There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 205 of 213

, Volume 6, Rev. 0, Page 206 of 213 ATTACHMENT 9 RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS , Volume 6, Rev. 0, Page 206 of 213

Attachment 1, Volume 6, Rev. 0, Page 207 of 213 CTS 3.2.a, CHEMICAL AND VOLUME CONTROL SYSTEM Attachment 1, Volume 6, Rev. 0, Page 207 of 213

, Volume 6, Rev. 0, Page 208 of 213 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 6, Rev. 0, Page 208 of 213

Attachment 1, Volume 6, Rev. 0, Page 209 of 213 CTS 3.2.a 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM APPLICABILITY Applies to the operational status of the Chemical and Volume Control System.

OBJECTIVE R01 To define those conditions of the Chemical and Volume Control System necessary to ensure safe reactor operation.

SPECIFICATIONS

a. When fuel is in the reactor there shall be at least one flow path to the core for boric acid injection.

Amendment No. 116 TS 3.2-1 03/28/95 Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 209 of 213

Attachment 1, Volume 6, Rev. 0, Page 210 of 213 DISCUSSION OF CHANGES CTS 3.2.a, CHEMICAL VOLUME CONTROL SYSTEM ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.2.a provides requirements on boric acid flow paths to the reactor core although the specification is labeled "Chemical and Volume Control System.

The purpose for a boric acid flow path is for control of the chemical neutron absorber (boron) concentration in the Reactor Coolant System (RCS) and to help maintain the SHUTDOWN MARGIN. To accomplish this, the CTS requires at least one flow path to the core for boric acid injection. The Chemical and Volume Control System is not assumed to be OPERABLE to mitigate the consequences of a design basis accident (DBA) or transient. In the case of a malfunction of the CVCS, which causes a boron dilution event, the automatic response, or that required by the operator, is to close the appropriate valves in the reactor makeup system. This action is required before the shutdown margin is lost. Operation of the boration subsystem is not assumed to mitigate this event (KPS USAR Section 14.1.4). The ITS does not include this Specification. This changes the CTS by relocating this Specification to the Technical Requirements Manual (TRM).

This change is acceptable because CTS 3.2.a does not meet the 10 CFR 50.36(c)(2)(ii) criteria for inclusion into the ITS.

10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. The Chemical and Volume Control System is not used to detect, and indicate in the control room, a significant degradation of the reactor coolant pressure boundary. The Chemical and Volume Control System does not satisfy criterion 1.
2. The Chemical and Volume Control System is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The Chemical and Volume Control System does not satisfy criterion 2.
3. The Chemical and Volume Control System is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The Chemical and Volume Control System does not satisfy criterion 3.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 210 of 213

Attachment 1, Volume 6, Rev. 0, Page 211 of 213 DISCUSSION OF CHANGES CTS 3.2.a, CHEMICAL VOLUME CONTROL SYSTEM

4. As discussed in Section 4.0 (Appendix A, page A-6) and summarized in Table 1 of WCAP-11618, the loss of the Chemical and Volume Control System (i.e., boric acid flow path to the core - operating and shutdown)was found to be a non-significant risk contributor to core damage frequency and offsite releases. Dominion Energy Kewaunee (DEK) has reviewed this evaluation, considers it applicable to Kewaunee Power Station (KPS), and concurs with the assessment. The Chemical and Volume Control System does not satisfy criterion 4.

Since 10 CFR 50.36(c)(2)(ii) criteria have not been met, the Chemical and Volume Control System may be relocated out of the Technical Specifications.

The Chemical and Volume Control System will be relocated to the TRM.

Changes to the TRM will be controlled by the provisions of 10 CFR 50.59. This change is designated as relocation because the Specification did not meet the criteria in 10 CFR 50.36(c)(2)(ii) and has been relocated to the TRM.

REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 211 of 213

Attachment 1, Volume 6, Rev. 0, Page 212 of 213 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 6, Rev. 0, Page 212 of 213

Attachment 1, Volume 6, Rev. 0, Page 213 of 213 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.2.a, CHEMICAL AND VOLUME CONTROL SYSTEM There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 213 of 213

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