Text

Attachment 1, Volume 9, Revision 1, Kewaunee Power Station Improved Technical Specifications Conversion, ITS Section 3.4 Reactor Coolant System (RCS)
	ML101890218
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Site:	Kewaunee
Issue date:	07/01/2010
From:	; Dominion Energy Kewaunee
To:	; Office of Nuclear Reactor Regulation
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Attachment 1, Volume 3, Rev. 1, Page i of i Summary of Changes ITS Section 3.4 Change Description Affected Pages The changes described in the KPS response to Pages 87, 90, and 98 question RPG-007 have been made. This change adds ISTS 3.4.5 Required Action D.1 into the KPS ITS.

The changes described in the KPS response to Pages 341, 343, 345, 346, 347, 348, 349, 351, 352, question RPG-008 have been made. This change 353, 354, 355, 359, 360, 361, 362, 363, 364, and deletes the containment humidity monitor 365 requirements from ITS 3.4.15. Note that these requirements are not currently in the CTS.

The changes described in the KPS response to Pages 372, 374, 375, 380, 383, 393, 394, and 395 question RPG-005 have been made. This change deletes the Notes from ITS SR 3.4.16.1 and 3.4.16.2.

Page 1 of 1 Attachment 1, Volume 3, Rev. 1, Page i of i

Attachment 1, Volume 9, Rev. 1, Page 1 of 478 ATTACHMENT 1 VOLUME 9 KEWAUNEE POWER STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.4 REACTOR COOLANT SYSTEM (RCS)

Revision 1 Attachment 1, Volume 9, Rev. 1, Page 1 of 478

Attachment 1, Volume 9, Rev. 1, Page 2 of 478 LIST OF ATTACHMENTS

1. ITS 3.4.1

2. ITS 3.4.2

3. ITS 3.4.3

4. ITS 3.4.4

5. ITS 3.4.5

6. ITS 3.4.6

7. ITS 3.4.7

8. ITS 3.4.8

9. ITS 3.4.9

10. ITS 3.4.10

11. ITS 3.4.11

12. ITS 3.4.12

13. ITS 3.4.13

14. ITS 3.4.14

15. ITS 3.4.15

16. ITS 3.4.16

17. ITS 3.4.17

18. Relocated/Deleted Current Technical Specifications (CTS)

19. ISTS Not Adopted Attachment 1, Volume 9, Rev. 1, Page 2 of 478

Attachment 1, Volume 9, Rev. 1, Page 3 of 478 ATTACHMENT 1 ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS Attachment 1, Volume 9, Rev. 1, Page 3 of 478

Attachment 1, Volume 9, Rev. 1, Page 4 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs)

Attachment 1, Volume 9, Rev. 1, Page 4 of 478

Attachment 1, Volume 9, Rev. 1, Page 5 of 478 ITS ITS 3.4.1 A01

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.

2. Not more than one inoperable full length rod shall be allowed at any time.

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 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 See ITS an analytically determined part power level which is consistent with the safety analysis. 3.1.4

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.

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 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 3.2.4 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 Applicability During steady-state power operation, Tavg, shall be maintained within the limits specified LCO 3.4.1.b in the COLR, except as provided by TS 3.10.n.

l. Reactor Coolant System Pressure Add proposed Applicability Notes a and b L01 Applicability During steady-state power operation, Reactor Coolant System pressure shall be LCO 3.4.1.a 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 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 5 of 478

Attachment 1, Volume 9, Rev. 1, Page 6 of 478 ITS ITS 3.4.1 A01

m. Reactor Coolant Flow Applicability

1. During steady-state power operation, reactor coolant total flow rate shall be 178,000 gallons per minute average and greater than or equal to the limit specified in the COLR.

LCO 3.4.1.c If reactor coolant flow rate is not within the limits as specified in the COLR, action shall be taken in accordance with TS 3.10.n.

LA01 SR 3.4.1.4

2. Compliance with this flow requirement shall be demonstrated by verifying the reactor coolant flow during initial power escalation following each REFUELING, at or above SR 3.4.1.4 Note 90% power with plant parameters as constant as practical. A02

n. DNBR Parameters ACTION A If, during power operation any of the conditions of TS 3.10.k, TS 3.10.l, or TS 3.10.m.1 are not met, restore the parameter in two hours or less to within limits or reduce power ACTION B to < 5% of thermal rated power within an additional six hours. Following analysis, M01 thermal power may be raised not to exceed a power level analyzed to maintain a DNBR greater than the minimum DNBR limit.

Add proposed SR 3.4.1.1, SR 3.4.1.2, and SR 3.4.1.3 M02 Amendment No. 181 TS 3.10-8 03/24/2005 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 6 of 478

Attachment 1, Volume 9, Rev. 1, Page 7 of 478 DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB 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.m.2 states, in part, that compliance with the reactor coolant total flow rate shall be demonstrated by verifying the reactor coolant flow during initial power escalation following each REFUELING, at or above 90% power "with plant parameters as constant as practical." ITS SR 3.4.1.4 requires measurement of the RCS total flow rate and is modified by a Note which states, "Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 90% RTP." This changes the CTS by explicitly specifying the time required to perform the Surveillance after entering MODE 1 conditions. That is, it defines how long the plant has to get the parameters constant.

This change is acceptable because the Note applies a specific period of time (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) to establish stable operating conditions. In CTS, the time period is not specific and can vary from outage to outage. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.10.m states, in part, that if the reactor coolant flow is not restored to within limits to reduce power to < 5% of thermal rated power. It further states that following analysis, thermal power may be raised not to exceed a power level analyzed to maintain a departure from nucleate boiling ratio (DNBR) greater than the minimum DNBR limit. ITS 3.4.1 ACTION B requires that if the RCS DNB parameters are not restored to be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . This changes the CTS by not allowing an analysis to be performed to raise THERMAL POWER.

The purpose of CTS 3.10.m is to ensure that the minimum DNBR will be met for each of the analyzed transients. ITS 3.4.1 ACTION B continues to maintain this assurance by exiting the MODE of Applicability when the RCS DNB parameters cannot be restored to within limits, thereby restoring the DNB margin and eliminating the potential for violation of the accident analysis bounds. Therefore, there is no need to perform an analysis as power will not be allowed to be increased until the parameters are within the limits. This change is designated as more restrictive because the ITS will no longer include an allowance for raising power based on an analysis.

M02 ITS SR 3.4.1.1 requires verification that the pressurizer pressure is greater than or equal to the limit specified in the COLR every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS SR 3.4.1.2 requires verification that the RCS average temperature is less than or equal to the limit specified in the COLR every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS SR 3.4.1.3 requires Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 7 of 478

Attachment 1, Volume 9, Rev. 1, Page 8 of 478 DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS verification that the RCS total flow rate is 178,000 gpm and greater than or equal to the limit specified in the COLR every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The CTS does not contain these Surveillance Requirements. This changes the CTS by adding Surveillance Requirements to verify that the pressurizer pressure is greater than or equal to the limit specified in the COLR, the RCS average temperature is less than or equal to the limit specified in the COLR, and the RCS total flow rate is 178,000 gpm and greater than or equal to the limit specified in the COLR.

This change is acceptable because the added Surveillance Requirement provide additional assurance that the RCS DNB parameters are within the limits specified in the COLR. This change is designated as more restrictive because new Surveillance Requirements are added.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.10.m.2 states, in part, that compliance with the reactor coolant flow rate requirement shall be demonstrated by verifying the reactor coolant flow during initial power escalation following each REFUELING.

ITS SR 3.4.1.4 requires verification by precision heat balance that RCS total flow rate is 178,000 gpm and greater than or equal to the limit specified in the COLR every 18 months. This SR is modified by a Note which states that the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 90% RTP. This changes the CTS by removing the procedural detail of "during initial power escalation" to the Bases.

The removal of this detail for performing the Surveillance Requirement from the Technical Specifications 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 perform a precision heat balance every 18 months. 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 Specification Bases Control Program in Chapter 5. This program provides for evaluation of changes to ensure the Bases are properly controlled.

This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 3.10.l states, in part, that during steady-state power operations, the Reactor Coolant System (RCS) pressure shall be maintained within the limits specified in the COLR. ITS 3.4.1.a requires Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 8 of 478

Attachment 1, Volume 9, Rev. 1, Page 9 of 478 DISCUSSION OF CHANGES ITS 3.4.1, RCS PRESSURE, TEMPERATURE, AND FLOW DNB LIMITS the RCS DNB parameters for pressurizer pressure to be within limits during MODE 1, but the Applicability is modified by a Note. The Note states that the pressurizer pressure limit does not apply during THERMAL POWER ramp > 5%

RTP per minute or THERMAL POWER step > 10% RTP. This changes the CTS by allowing the pressurizer pressure limit to be outside of its limit during THERMAL POWER ramp > 5% RTP per minute or THERMAL POWER step

> 10% RTP.

The purpose of CTS 3.10.l is to ensure the RCS pressure is consistent with operation within the nominal operational envelope. ITS 3.4.1 continues to ensure the RCS pressure is consistent with this same operational envelope. The addition of the Note is to allow short term perturbations where actions to control the pressure variations could be counterproductive. Furthermore, since these proposed Note allowances represent transients initiated from power levels

< 100 % RTP, an increased DNBR margin exists to offset the temporary pressure variations. This change is designated as less restrictive because an allowance for pressurizer pressure to be outside its limits is allowed in the ITS that is not allowed in the CTS.

Kewaunee Power Station Page 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 9 of 478

Attachment 1, Volume 9, Rev. 1, Page 10 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 10 of 478

Attachment 1, Volume 9, Rev. 1, Page 11 of 478 CTS RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

3.10.l a. Pressurizer pressure is greater than or equal to the limit specified in the COLR, 1 3.10.k b. RCS average temperature is less than or equal to the limit specified in the COLR, and 178,000 1

3.10.m.1 c. RCS total flow rate [284,000] gpm and greater than or equal to the 2 limit specified in the COLR.

3.10.l, APPLICABILITY: MODE 1.

3.10.k, 3.10.m.1

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

DOC L01 Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute or 1

b. THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.10.n A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months parameters not within parameter(s) to within limit.

limits.

3.10.n B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

WOG STS 3.4.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 11 of 478

Attachment 1, Volume 9, Rev. 1, Page 12 of 478 CTS RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M02 SR 3.4.1.1 Verify pressurizer pressure is greater than or equal 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to the limit specified in the COLR.

DOC M02 SR 3.4.1.2 Verify RCS average temperature is less than or 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months equal to the limit specified in the COLR.

178,000 DOC M02 SR 3.4.1.3 Verify RCS total flow rate is [284,000] gpm and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 2 greater than or equal to the limit specified in the COLR.

3.10.m.2 SR 3.4.1.4 -------------------------------NOTE------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after

[90]% RTP. 2 178,000 Verify by precision heat balance that RCS total flow [18] months 2 rate is [284,000] gpm and greater than or equal to the limit specified in the COLR.

WOG STS 3.4.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 12 of 478

Attachment 1, Volume 9, Rev. 1, Page 13 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4, RCS PRESSURE, TEMPERATURE AND FLOW DNB LIMITS

1. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

2. The ITS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and 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 9, Rev. 1, Page 13 of 478

Attachment 1, Volume 9, Rev. 1, Page 14 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 14 of 478

Attachment 1, Volume 9, Rev. 1, Page 15 of 478 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits BASES BACKGROUND These Bases address requirements for maintaining RCS pressure, temperature, and flow rate within limits assumed in the safety analyses.

The safety analyses (Ref. 1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on RCS pressure, temperature, and flow rate ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the transients 1

analyzed. Condition I and II The RCS pressure limit is consistent with operation within the nominal operational envelope. Pressurizer pressure indications are averaged to come up with a value for comparison to the limit. A lower pressure will cause the reactor core to approach DNB limits.

The RCS coolant average temperature limit is consistent with full power operation within the nominal operational envelope. Indications of temperature are averaged to determine a value for comparison to the limit. A higher average temperature will cause the core to approach DNB limits.

The RCS flow rate normally remains constant during an operational fuel cycle with all pumps running. The minimum RCS flow limit corresponds to that assumed for DNB analyses. Flow rate indications are averaged to come up with a value for comparison to the limit. A lower RCS flow will cause the core to approach DNB limits.

Operation for significant periods of time outside these DNB limits increases the likelihood of a fuel cladding failure in a DNB limited event.

APPLICABLE The requirements of this LCO represent the initial conditions for DNB Condition I and II 1 SAFETY limited transients analyzed in the plant safety analyses (Ref. 1). The ANALYSES safety analyses have shown that transients initiated from the limits of this LCO will result in meeting the DNBR criterion. This is the acceptance limit for the RCS DNB parameters. Changes to the unit that could impact these parameters must be assessed for their impact on the DNBR partial 1 RCCA criteria. The transients analyzed for include loss of coolant flow events misalignment and dropped or stuck rod events. A key assumption for the analysis of 1 these events is that the core power distribution is within the limits of LCO 3.1.6, "Control Bank Insertion Limits," LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)."

WOG STS B 3.4.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 15 of 478

Attachment 1, Volume 9, Rev. 1, Page 16 of 478 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued)

The pressurizer pressure limit and RCS average temperature limit specified in the COLR correspond to the analytical limits used in the safety analyses, with allowance for measurement uncertainty.

limits The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). 3 LCO This LCO specifies limits on the monitored process variables - pressurizer pressure, RCS average temperature, and RCS total flow rate - to ensure the core operates within the limits assumed in the safety analyses. These variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow, usually based on [maximum analyzed steam generator tube plugging], is retained 2 in the TS LCO. Operating within these limits will result in meeting the DNBR criterion in the event of a DNB limited transient.

RCS total flow rate contains a measurement error based on performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi, which might not be detected, could bias the result from the precision heat balance in a INSERT 1 nonconservative manner. Therefore, a penalty for undetected fouling of the feedwater venturi raises the nominal flow measurement allowance for no fouling.

Any fouling that might bias the flow rate measurement greater than the penalty for undetected fouling of the feedwater venturi can be detected by monitoring and trending various plant performance parameters. If detected, either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

pressurizer RCS average RCS total The numerical values for pressure, temperature, and flow rate specified in 1 the COLR are given for the measurement location and have been adjusted for instrument error.

APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern.

WOG STS B 3.4.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 16 of 478

Attachment 1, Volume 9, Rev. 1, Page 17 of 478 B 3.4.1 4 INSERT 1 RCS flow indication calibration must include appropriate considerations for the accuracy of feedwater flow measurement. KPS can employ either of two methods to measure feedwater flow; an installed Crossflow Ultrasonic Flow Measurement System (Crossflow System), or in-line feedwater flow venturis. Unlike the feedwater venturis, the Crossflow System is not susceptible to fouling during use and possesses a higher accuracy. These attributes make the Crossflow System the preferred method of measuring feedwater flow as an input to the determination of RCS flow.

In the event the Crossflow System is not available, the feedwater venturis are used to calibrate the RCS flow indicators. Because the feedwater venturis are susceptible to fouling during use, a nonconservative bias in RCS flow indicators might develop over time.

Therefore, when the feedwater venturis are used to calibrate the RCS flow rate indicators, a penalty for undetected fouling of the feedwater venturis may be applied to raise the nominal flow measurement. The penalty is described in Specification 5.6.3.b. The need for such a bias would be detected by monitoring and trending various plant performance parameters.

Insert Page B 3.4.1-2 Attachment 1, Volume 9, Rev. 1, Page 17 of 478

Attachment 1, Volume 9, Rev. 1, Page 18 of 478 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued)

A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.

The DNBR limit is provided in SL 2.1.1, "Reactor Core SLs." The conditions which define the DNBR limit are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.

ACTIONS A.1 RCS pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter(s).

RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS total flow rate is A

below the LCO limit, power must be reduced, as required by Required 3

Action B.1, to restore DNB margin and eliminate the potential for violation of the accident analysis bounds.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.

B.1 If Required Action A.1 is not met within the associated Completion Time, the 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 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis bounds. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable to reach the required plant conditions in an orderly manner.

WOG STS B 3.4.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 18 of 478

Attachment 1, Volume 9, Rev. 1, Page 19 of 478 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE SR 3.4.1.1 REQUIREMENTS Since Required Action A.1 allows a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore parameters that are not within limits, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Frequency for pressurizer pressure is sufficient to ensure the pressure can be restored to a normal operation, steady state condition following load changes and other expected transient operations. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

SR 3.4.1.2 Since Required Action A.1 allows a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore parameters that are not within limits, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Frequency for RCS average temperature is sufficient to ensure the temperature can be restored to a normal operation, steady state condition following load changes and other expected transient operations. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

SR 3.4.1.3 The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Frequency for RCS total flow rate is performed using the installed flow instrumentation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify operation within safety analysis assumptions.

SR 3.4.1.4 Measurement of RCS total flow rate by performance of a precision calorimetric heat balance once every [18] months allows the installed 2 RCS flow instrumentation to be calibrated and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.

The Frequency of [18] months reflects the importance of verifying flow 2 after a refueling outage when the core has been altered, which may have caused an alteration of flow resistance.

WOG STS B 3.4.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 19 of 478

Attachment 1, Volume 9, Rev. 1, Page 20 of 478 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)

This SR is modified by a Note that allows entry into MODE 1, without having performed the SR, and placement of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after [90%] RTP. This exception is appropriate 2 since the heat balance requires the plant to be at a minimum of

[90%] RTP to obtain the stated RCS flow accuracies. The Surveillance 2

shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching [90%] RTP.

U Chapter 14 REFERENCES 1. FSAR, Section [15]. 1 2 WOG STS B 3.4.1-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 20 of 478

Attachment 1, Volume 9, Rev. 1, Page 21 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.1 BASES, RCS PRESSURE, TEMPERATURE AND FLOW DNB LIMITS

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 ITS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and 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. Typographical error corrected.

4. KPS normally uses the Crossflow Ultrasonic Flow Measurement System Crossflow System) in lieu of flow venturis. Therefore, the Bases have been modified accordingly. The Crossflow System was approved for use at KPS by the NRC in License Amendment 168 (ADAMS Accession No. ML031530734).

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 21 of 478

Attachment 1, Volume 9, Rev. 1, Page 22 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 22 of 478

Attachment 1, Volume 9, Rev. 1, Page 23 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.1, RCS PRESSURE, TEMPERATURE AND FLOW DNB LIMITS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 23 of 478

Attachment 1, Volume 9, Rev. 1, Page 24 of 478 ATTACHMENT 2 ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY Attachment 1, Volume 9, Rev. 1, Page 24 of 478

, Volume 9, Rev. 1, Page 25 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 25 of 478

, Volume 9, Rev. 1, Page 26 of 478 ITS 3.4.2 Add Proposed ITS 3.4.2 M01 Page 1 of 1 , Volume 9, Rev. 1, Page 26 of 478

Attachment 1, Volume 9, Rev. 1, Page 27 of 478 DISCUSSION OF CHANGES ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M01 The CTS does not have any requirements for RCS minimum temperature for criticality. ITS 3.4.2 requires that each RCS loop average temperature (Tavg) shall be 540ºF in MODE 1 and in MODE 2 with keff 1.0. This changes the CTS by incorporating the requirements of ISTS 3.4.2. The ITS also provides an Action for when Tavg in one or more RCS loops is not within limits (ACTION A) and a Surveillance Requirement (SR 3.4.2.1).

The primary function of the RCS loop average temperature limit is to ensure that the reactor will not be made or maintained critical at a temperature less than that assumed in the safety analysis. This change is acceptable because each RCS loop average temperature is now required to be 540º when in MODE 1 and in MODE 2 with keff 1.0. All low power safety analyses assume initial RCS loop temperatures the HZP temperature of 547°F. The minimum temperature for criticality limitation provides a small band, 7°F, for critical operation below HZP.

This band allows critical operation below HZP during plant startup and does not adversely affect any safety analyses. In addition, the MTC is explicitly considered in the safety analysis process by verifying key input parameters to transient analyses at conditions which bound the MTC. This change has been designated as more restrictive because it adds a new requirement to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 27 of 478

Attachment 1, Volume 9, Rev. 1, Page 28 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 28 of 478

Attachment 1, Volume 9, Rev. 1, Page 29 of 478 CTS RCS Minimum Temperature for Criticality 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Minimum Temperature for Criticality 540 DOC M01 LCO 3.4.2 Each RCS loop average temperature (Tavg) shall be [541]°F. 1 DOC M01 APPLICABILITY: MODE 1, MODE 2 with keff 1.0.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC M01 A. Tavg in one or more RCS A.1 Be in MODE 2 with Keff 30 minutes 2 loops not within limit. < 1.0.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 540 DOC M01 SR 3.4.2.1 Verify RCS Tavg in each loop [541]°F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 1 WOG STS 3.4.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 29 of 478

Attachment 1, Volume 9, Rev. 1, Page 30 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY

1. The ITS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and 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. Typographical/grammatical error corrected.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 30 of 478

Attachment 1, Volume 9, Rev. 1, Page 31 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 31 of 478

Attachment 1, Volume 9, Rev. 1, Page 32 of 478 RCS Minimum Temperature for Criticality B 3.4.2 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.2 RCS Minimum Temperature for Criticality BASES BACKGROUND This LCO is based upon meeting several major considerations before the reactor can be made critical and while the reactor is critical.

The first consideration is moderator temperature coefficient (MTC),

LCO 3.1.3, "Moderator Temperature Coefficient (MTC)." In the transient and accident analyses, the MTC is assumed to be in a range from slightly positive to negative and the operating temperature is assumed to be within the nominal operating envelope while the reactor is critical. The LCO on minimum temperature for criticality helps ensure the plant is operated consistent with these assumptions.

The second consideration is the protective instrumentation. Because certain protective instrumentation (e.g., excore neutron detectors) can be affected by moderator temperature, a temperature value within the nominal operating envelope is chosen to ensure proper indication and response while the reactor is critical.

The third consideration is the pressurizer operating characteristics. The transient and accident analyses assume that the pressurizer is within its normal startup and operating range (i.e., saturated conditions and steam bubble present). It is also assumed that the RCS temperature is within its normal expected range for startup and power operation. Since the density of the water, and hence the response of the pressurizer to transients, depends upon the initial temperature of the moderator, a minimum value for moderator temperature within the nominal operating envelope is chosen.

The fourth consideration is that the reactor vessel is above its minimum nil ductility reference temperature when the reactor is critical.

APPLICABLE Although the RCS minimum temperature for criticality is not itself an initial SAFETY condition assumed in Design Basis Accidents (DBAs), the closely aligned ANALYSES uncontrolled temperature for hot zero power (HZP) is a process variable that is an initial condition of DBAs, such as the rod cluster control assembly (RCCA) withdrawal, RCCA ejection, and main steam line break accidents performed at zero power that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier.

WOG STS B 3.4.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 32 of 478

Attachment 1, Volume 9, Rev. 1, Page 33 of 478 RCS Minimum Temperature for Criticality B 3.4.2 In addition, the MTC is explicitly considered in the safety analysis process by verifying key input parameters to transient analyses at conditions which bound the MTC.

BASES APPLICABLE SAFETY ANALYSES (continued)

All low power safety analyses assume initial RCS loop temperatures the HZP temperature of 547°F (Ref. 1). The minimum temperature for 7

criticality limitation provides a small band, 6°F, for critical operation below 3 HZP. This band allows critical operation below HZP during plant startup and does not adversely affect any safety analyses since the MTC is not significantly affected by the small temperature difference between HZP 3 and the minimum temperature for criticality.

The RCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO Compliance with the LCO ensures that the reactor will not be made or maintained critical (keff 1.0) at a temperature less than a small band below the HZP temperature, which is assumed in the safety analysis.

Failure to meet the requirements of this LCO may produce initial conditions inconsistent with the initial conditions assumed in the safety analysis.

APPLICABILITY In MODE 1 and MODE 2 with keff 1.0, LCO 3.4.2 is applicable since the reactor can only be critical (keff 1.0) in these MODES.

The special test exception of LCO 3.1.8, "PHYSICS TESTS Exceptions -

MODE 2," permits PHYSICS TESTS to be performed at 5% RTP with RCS loop average temperatures slightly lower than normally allowed so that fundamental nuclear characteristics of the core can be verified. In order for nuclear characteristics to be accurately measured, it may be necessary to operate outside the normal restrictions of this LCO. For example, to measure the MTC at beginning of cycle, it is necessary to allow RCS loop average temperatures to fall below Tno load, which may cause RCS loop average temperatures to fall below the temperature limit of this LCO.

ACTIONS A.1 If the parameters that are outside the limit cannot be restored, 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 MODE 2 with Keff < 1.0 1 within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period. The allowed time is reasonable, based on operating experience, to reach MODE 2 with Keff < 1.0 in an 1 orderly manner and without challenging plant systems.

WOG STS B 3.4.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 33 of 478

Attachment 1, Volume 9, Rev. 1, Page 34 of 478 RCS Minimum Temperature for Criticality B 3.4.2 BASES SURVEILLANCE SR 3.4.2.1 REQUIREMENTS 540 RCS loop average temperature is required to be verified at or above 2

[541]°F every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The SR to verify RCS loop average temperatures every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months takes into account indications and alarms that are continuously available to the operator in the control room and is consistent with other routine Surveillances which are typically performed once per shift. In addition, operators are trained to be sensitive to RCS temperature during approach to criticality and will ensure that the minimum temperature for criticality is met as criticality is approached.

U 3 2 Chapter 14 REFERENCES 1. FSAR, Section [15.0.3].

WOG STS B 3.4.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 34 of 478

Attachment 1, Volume 9, Rev. 1, Page 35 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.2 BASES, RCS MINIMUM TEMPERATURE FOR CRITICALITY

1. Typographical/grammatical error corrected.

2. The ITS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and 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. 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.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 35 of 478

Attachment 1, Volume 9, Rev. 1, Page 36 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 36 of 478

Attachment 1, Volume 9, Rev. 1, Page 37 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.2, RCS MINIMUM TEMPERATURE FOR CRITICALITY There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 37 of 478

Attachment 1, Volume 9, Rev. 1, Page 38 of 478 ATTACHMENT 3 ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS Attachment 1, Volume 9, Rev. 1, Page 38 of 478

, Volume 9, Rev. 1, Page 39 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 39 of 478

Attachment 1, Volume 9, Rev. 1, Page 40 of 478 ITS A01 ITS 3.4.3

b. Heatup and Cooldown Limit Curves for Normal Operation LCO 3.4.3

1. The reactor coolant temperature and pressure and system heatup and cooldown LA01 M01 rates (with the exception of the pressurizer) shall be limited in accordance with Add proposed Figures TS 3.1-1 and TS 3.1-2. Figures TS 3.1-1 and TS 3.1-2 are applicable for the Figures Applicability service period of up to 33 (1) effective full-power years.

3.4.3-1 and Add proposed ACTIONS, A, B, and C M02 3.4.3-2 A. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

LA02 B. Figures TS 3.1-1 and TS 3.1-2 define limits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g.,

pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

C. The isothermal curve in Figure TS 3.1-2 defines limits to assure prevention of non-ductile failure applicable to low temperature overpressurization events only. See ITS 3.4.12 Application of this curve is limited to evaluation of LTOP events whenever one or more of the RCS cold leg temperatures are less than or equal to the LTOP enabling temperature of 200°F.

Add proposed SR 3.4.3.1 M03

2. The secondary side of the steam generator must not be pressurized > 200 psig if the See CTS temperature of the steam generator is < 70°F. 3.1.b.2

3. The pressurizer cooldown and heatup rates shall not exceed 200°F/hr and 100°F/hr, See CTS respectively. The spray shall not be used if the temperature difference between the 3.1.b.3 pressurizer and the spray fluid is > 320°F.

4. The overpressure protection system for low temperature operation shall be OPERABLE whenever one or more of the RCS cold leg temperatures are 200°F, and the reactor vessel head is installed. The system shall be considered OPERABLE when at least one of the following conditions is satisfied:

A. The overpressure relief valve on the Residual Heat Removal System (RHR 33-1) shall have a set pressure of 500 psig and shall be aligned to the RCS by maintaining valves RHR 1A, 1B, 2A, and 2B open.

See ITS 3.4.12

1. With one flow path inoperable, the valves in the parallel flow path shall be verified open with the associated motor breakers for the valves locked in the off position. Restore the inoperable flow path within five days or complete depressurization and venting of the RCS through a 6.4 square inch vent within an additional eight hours.

2. With both flow paths or RHR 33-1 inoperable, complete depressurization and venting of the RCS through at least a 6.4 square inch vent pathway within eight hours.

Figures (1) 3.4.3-1 and The curves are limited to 31.1 EFPY due to changes in vessel fluence associated with 3.4.3-2 operation at uprated power.

Amendment No. 168 TS 3.1-6 07/08/2003 Page 1 of 4 Attachment 1, Volume 9, Rev. 1, Page 40 of 478

Attachment 1, Volume 9, Rev. 1, Page 41 of 478 ITS A01 ITS 3.4.3

f. Minimum Conditions for Criticality Applicability

1. The reactor shall not be brought to a critical condition until the pressure-temperature LCO 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.

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 maximum upper moderator temperature coefficient limit shall be 5 pcm/°F for 3.1.3 and 3.1.8 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:

A. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , develop and maintain administrative control rod withdrawal See ITS limits sufficient to restore the moderator temperature coefficient to within the 3.1.3 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Amendment No. 165 TS 3.1-10 03/11/2003 Page 2 of 4 Attachment 1, Volume 9, Rev. 1, Page 41 of 478

A01 ITS ITS 3.4.3 FIGURE TS 3.1-1 Figure 3.4.3-1 KEWAUNEE UNIT NO. 1 HEATUP LIM ITATION CURVES

[1]

APPLICABLE FOR PERIODS UP TO 33 EFFECTIVE FULL-POWER YEARS 2500 Material Property Basis In-Service Leak Test Intermediate Forging Cu = 0.06 wt% Ni = 0.71 wt% Minimum Temperature 2250 Initial RT NDT = 60°F CF = 37°F 2000 Margin = 34°F At 33 Effective Full Power Years Adj. RT NDT at 1/4T = 139°F 1750 Adj. RT NDT at 3/4T = 131°F Closure Flange Heatup Rates Initial RTNDT = 60°F up to 100°F/Hr 1500 Margins for Instrumentation s for Instrumentation Error Error Pressure andand Drop Pressure Unacceptable 1250 RV Core Acrosscross RV Core Acceptable

+13°F Instrumentation Operation F Instrumentation

-58 psi Instrumentation Operation si Instrumentation

-70 psi D P Indicated Pressure (psig) 1000 si P 750 Criticality Limit LA03 500 , Volume 9, Rev. 1, Page 42 of 478 Attachment 1, Volume 9, Rev. 1, Page 42 of 478 250 0

50 100 150 200 250 300 350 400 Amendment No. 168 Indicate d Te mpe rature (°F) 07/08/2003 NOTE:

[1)

The curves are limited to 31.1 EFPY due to changes in vessel fluence associated with operation at uprated power.

Page 3 of 4

A01 ITS ITS 3.4.3 Figure 3.4.3-2 FIGURE TS 3.1-2 KEWAUNEE UNIT NO. 1 COOLDOWN LIMITATION CURVES APPLICABLE FOR PERIODS UP TO 33 [1] EFFECTIVE FULL-POWER YEARS 2500 Material Property Basis Weld Metal 2250 Cu = 0.287 wt% Ni = 0.756 wt%

Initial RTNDT = -50°F CF = 192.3°F 2000 Margin = 219.9°F At 33 Effective Full Power Years 1750 Adj. RTNDT at 1/4T = 246°F Unacceptable Adj. RTNDT at 3/4T = 200°F Operation Intermediate Forging 0°F 20°F 1500 Cu = 0.06 wt% Ni = 0.71 wt%

40°F Initial RT NDT = 60°F 60°F CF = 37°F 100°F 1250 Margin = 34°F At 33 Effective Full Power Years Adj. RT NDT at 1/4T = 139°F 1000 Adj. RT NDT at 3/4T = 131°F Acceptable Indicated Pressure (psig)

Closure Flange Operation Initial RTNDT = 60°F 750 Margins for Instrumentation Error and 500 0°F Pressure Drop Across RV Core 20°F 40°F

+13°F Instrumentation LA03

-58 psi Instrumentation , Volume 9, Rev. 1, Page 43 of 478 Attachment 1, Volume 9, Rev. 1, Page 43 of 478 250 60°F 100°F -70 psi P 0

50 100 150 200 250 300 350 Indicated Temperature (°F)

Amendment No. 168 07/08/2003

[1]

The curves are limited to 31.1 EFPY due to changes in vessel fluence associated with operation at uprated power.

Page 4 of 4

Attachment 1, Volume 9, Rev. 1, Page 44 of 478 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) 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 3.1.b.1 does not provide any Applicability requirements for the Heatup and Cooldown Limit Curves. ITS 3.4.3 requires the RCS Pressure, RCS Temperature, and RCS heatup and cooldown rates to be maintained with limits "at all times." This changes the CTS by specifically stating that the RCS Pressure, RCS Temperature, and RCS heatup and cooldown rates to maintained with limits are required "at all times."

The purpose of the heatup and cooldown limit curves is to limit the pressure and temperature changes during RCS heatup and cooldown to within the design assumptions and the stress limits for cyclic operation. This change is acceptable because the RCS pressure and temperature (P/T) limits provide the limits for acceptable operation for the prevention of nonductile failures. Even though the P/T limits were developed to provide guidance for operations during heatup and cooldown, the Applicability of "at all times" is required because of concern for nonductile failure. This change is designated as more restrictive because specific Applicability requirements have been added.

M02 CTS 3.1.b.1 provides limits for reactor coolant temperature and pressure and system heatup and cooldown rates. It does not specify Actions to take when the limitations are not met, therefore, CTS 3.0.c would be entered. CTS 3.0.c requires action to be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and to be in HOT STANDBY (equivalent to MODE 3) within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , HOT SHUTDOWN (equivalent to MODE 4) within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and COLD SHUTDOWN (equivalent to MODE 5) within the subsequent 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . ITS 3.4.3 adds ACTION A, which states that if the requirements of the LCO are not met in MODE 1, 2, 3, or 4 to restore the parameters to within limits in 30 minutes and to determine if the RCS is acceptable for continued operation within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . It also adds ACTION B, which states that if the Required Action and associated Completion Time of Condition A is not met, then to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in MODE 5 with the RCS pressure less than 500 psig within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Additionally, it adds ACTION C, which states that if the requirements of the LCO are not met when not in MODE 1, 2, 3, or 4 to initiate action to restore the parameters to within limits immediately and to determine the RCS is acceptable for continued operation prior to entering MODE 4. This changes the CTS by adding specific ITS 3.4.3 ACTIONS A, B, and C.

Kewaunee Power Station Page 1 of 4 Attachment 1, Volume 9, Rev. 1, Page 44 of 478

Attachment 1, Volume 9, Rev. 1, Page 45 of 478 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS This purpose of ITS 3.4.3 ACTIONS A, B, and C is to provide specific compensatory actions for when RCS pressure and temperature are not maintained within the limits. This change is acceptable because it provides the necessary and specific actions to take when the requirements are not met and provides appropriate times to complete the actions. This change is designated as more restrictive because it adds specific compensatory actions and times that are more restrictive than the current action and times for when P/T limits are not met in all conditions.

M03 CTS 3.1.b.1 provides pressure and temperature (P/T) limits during heatup and cooldown. However, there is no specific Surveillance Requirement for verification that the CTS pressure and temperature are within the P/T limits. ITS SR 3.4.3.1 requires verification that the RCS Pressure, RCS Temperature, and RCS heatup and cooldown rates are within limits every 30 minutes during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

This changes the CTS by adding a specific Surveillance Requirement to verify the RCS heatup and cooldown rates are met.

The purpose of CTS 3.1.b.1 is to specify the P/T limits during heatup and cooldown. The CTS does not include a Surveillance Requirement for verification of the specified P/T limits. This change adds a specific Surveillance Requirement. This change is acceptable because the proposed Surveillance Requirement (ITS SR 3.4.3.1) will verify that heatup and cooldown limits are met.

This change is designated as more restrictive because it adds a specific Surveillance Requirement not currently required by the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.1.b.1 states, in part, that reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures TS 3.1-1 and TS 3.1-2.

ITS 3.4.3 states that RCS pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained with limits. This changes the CTS by moving the exclusion of the pressurizer from the LCO limit to the Bases.

The removal of these details, which are related to design, from the Technical Specifications, 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 P/T limits on the RCS. Neither the CTS or the ITS P/T limits apply to the pressurizer. It is the ITS convention to state this detail in the ITS Bases. This detail of the LCO is not required to be in the Technical Specifications in order to provide adequate protection of the public health and safety. Also, this change is acceptable because the removed Kewaunee Power Station Page 2 of 4 Attachment 1, Volume 9, Rev. 1, Page 45 of 478

Attachment 1, Volume 9, Rev. 1, Page 46 of 478 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS information will be adequately controlled in the ITS Bases. Furthermore, the pressurizer limits are currently covered by CTS 3.1.b.3. As allowed by the NUREG-1431, these limits will be relocated to the TRM as discussed in CTS 3.1.b.3 DOC R01 in this section. Changes to the Bases are controlled by the Technical Specification 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 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 3.1.b.1.A states that allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation. Furthermore, CTS 3.1.b.1.B states that Figures TS 3.1-1 and TS 3.1-2 define limits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges. ITS 3.4.3 does not contain these statements. This changes the CTS by moving this information to the Bases.

The removal of these details, which are related to design, from the Technical Specifications, 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. ITS Figures 3.4.3-1 and 3.4.3-2 clearly state the acceptable regions of the curve. Also, 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 Specification 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 information relating to system design is being removed from the Technical Specifications.

LA03 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS Figures TS 3.1-1 and TS 3.1-2 describe in text boxes the Margins for Instrumentation Error and Pressure Drop that are used to determine the P/T limits. The ITS Figures 3.4.3-1 and 3.4.3-2 do not include this information. This changes the CTS by moving this information to the Bases.

The removal of these details, which are related to design, from the Technical Specifications, 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 P/T limits on the RCS and the curves include the margins listed in the two text boxes. Also, 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 Specification 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 information relating to system design is being removed from the Technical Specifications.

Kewaunee Power Station Page 3 of 4 Attachment 1, Volume 9, Rev. 1, Page 46 of 478

Attachment 1, Volume 9, Rev. 1, Page 47 of 478 DISCUSSION OF CHANGES ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 4 of 4 Attachment 1, Volume 9, Rev. 1, Page 47 of 478

Attachment 1, Volume 9, Rev. 1, Page 48 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 48 of 478

Attachment 1, Volume 9, Rev. 1, Page 49 of 478 CTS RCS P/T Limits 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.1.b.1; LCO 3.4.3 RCS pressure, RCS temperature, and RCS heatup and cooldown rates 3.1.f.1 shall be maintained within the limits specified in the PTLR. 1 Figure 3.4.3-1 and Figure 3.4.3-2 DOC APPLICABILITY: At all times.

M01 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC A. ------------NOTE------------ A.1 Restore parameter(s) to 30 minutes M02 Required Action A.2 within limits.

shall be completed whenever this Condition AND is entered.

A.2 Determine RCS is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months acceptable for continued Requirements of operation.

LCO not met in MODE 1, 2, 3, or 4.

DOC B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months M02 associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5 with RCS 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months pressure < [500] psig. 2 DOC C. ------------NOTE------------ C.1 Initiate action to restore Immediately M02 Required Action C.2 parameter(s) to within shall be completed limits.

whenever this Condition is entered. AND C.2 Determine RCS is Prior to entering Requirements of acceptable for continued MODE 4 LCO not met any time in operation.

other than MODE 1, 2, 3, or 4.

WOG STS 3.4.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 49 of 478

Attachment 1, Volume 9, Rev. 1, Page 50 of 478 CTS RCS P/T Limits 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC SR 3.4.3.1 -------------------------------NOTE------------------------------

M03 Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and RCS 30 minutes heatup and cooldown rates are within the limits specified in the PTLR. Figure 3.4.3-1 and Figure 3.4.3-2 1 INSERTS 1 and 2 1 WOG STS 3.4.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 50 of 478

Attachment 1, Volume 9, Rev. 1, Page 51 of 478 3.4.3 1

INSERT 1 Figure 3.4.3-1 KPS Heatup, Criticality, and In-Service Leak Test Limitation Curves Applicable for Periods up to 33(1) Effective Full Power Years (EFPY)

(1) Curves limited to 31.1 EFPY due to changes in vessel fluence associated with operation at power uprate.

Insert Page 3.4.3-2a Attachment 1, Volume 9, Rev. 1, Page 51 of 478

Attachment 1, Volume 9, Rev. 1, Page 52 of 478 3.4.3 1

INSERT 2 Figure 3.4.3-2 KPS Cooldown and LTOP Event Limitation Curves Applicable for Periods up to 33(1) EFPY (1) Curves limited to 31.1 EFPY due to changes in vessel fluence associated with operation at power uprate.

Insert Page 3.4.3-2b Attachment 1, Volume 9, Rev. 1, Page 52 of 478

Attachment 1, Volume 9, Rev. 1, Page 53 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS

1. Kewaunee Power Station is not adopting a Pressure Temperature Limits Report (PTLR) and is retaining in the ITS the limits on heatup, cooldown, and inservice leak and hydrostatic testing, and data for maximum rate of change of reactor coolant temperature. Therefore, references to the PTLR have been changed to Figure 3.4.3-1 and Figure 3.4.3-2. Furthermore, Figure 3.4.3-1 and Figure 3.4.3-2 have been added to ITS 3.4.3.

2. The ITS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and 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 9, Rev. 1, Page 53 of 478

Attachment 1, Volume 9, Rev. 1, Page 54 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 54 of 478

Attachment 1, Volume 9, Rev. 1, Page 55 of 478 RCS P/T Limits B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

This LCO The PTLR contains P/T limit curves for heatup, cooldown, inservice leak 1 and hydrostatic (ISLH) testing, and data for the maximum rate of change criticality, of reactor coolant temperature (Ref. 1).

Each P/T limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits do not apply to the pressurizer, which has different design characteristics and operating functions.

10 CFR 50, Appendix G (Ref. 2), requires the establishment of P/T limits for specific material fracture toughness requirements of the RCPB materials. Reference 2 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of and ASME Code, Mechanical Engineers (ASME) Code,Section III, Appendix G (Ref. 3). 7 Section XI, Appendix G (Ref.

4) The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTNDT) as exposure to neutron fluence increases.

The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel 5 6 material specimens, in accordance with ASTM E 185 (Ref. 4) and Appendix H of 10 CFR 50 (Ref. 5). The operating P/T limit curves will be 7 adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99 (Ref. 6). 7 WOG STS B 3.4.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 55 of 478

Attachment 1, Volume 9, Rev. 1, Page 56 of 478 RCS P/T Limits B 3.4.3 BASES BACKGROUND (continued)

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The criticality limit curve includes the Reference 2 requirement that it be 40°F above the heatup curve or the cooldown curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.4.2, "RCS Minimum Temperature for Criticality."

The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.

The ASME Code,Section XI, Appendix E (Ref. 7), provides a 8 7 recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE The P/T limits are not derived from Design Basis Accident (DBA)

SAFETY analyses. They are prescribed during normal operation to avoid ANALYSES encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. Reference 1 establishes the methodology for determining the P/T limits. Although the P/T limits are not derived from any DBA, the P/T limits are acceptance limits since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

WOG STS B 3.4.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 56 of 478

Attachment 1, Volume 9, Rev. 1, Page 57 of 478 RCS P/T Limits B 3.4.3 BASES LCO The two elements of this LCO are:

criticality,

a. The limit curves for heatup, cooldown, and ISLH testing and 2 3 (maximum of b. Limits on the rate of change of temperature. 2 100ºF/hr for heatup and cooldown)

The LCO limits apply to all components of the RCS, except the pressurizer. These limits define allowable operating regions and permit a large number of operating cycles while providing a wide margin to nonductile failure.

The limits for the rate of change of temperature control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and ISLH testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.

INSERT 1 1 Violating the LCO limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follow:

a. The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature, 3

b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced), and 3

c. The existences, sizes, and orientations of flaws in the vessel material.

APPLICABILITY The RCS P/T limits LCO provides a definition of acceptable operation for prevention of nonductile failure in accordance with 10 CFR 50, Appendix G (Ref. 2). Although the P/T limits were developed to provide guidance for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, their Applicability is at all times in keeping with the concern for nonductile failure. The limits do not apply to the pressurizer.

During MODES 1 and 2, other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these P/T limits. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," LCO 3.4.2, "RCS Minimum Temperature for Criticality," and Safety Limit 2.1, "Safety Limits," also provide operational restrictions for pressure and temperature and WOG STS B 3.4.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 57 of 478

Attachment 1, Volume 9, Rev. 1, Page 58 of 478 ITS B 3.4.3 1

INSERT 1 Figure 3.4.3-1 and Figure 3.4.3-2 were originally developed for service periods of up to 33 effective full power years (EFPY). However, the curves are limited to 31.1 EFPY due to changes in the vessel fluence associated with operation at uprated power.

Figure 3.4.3-1 and Figure 3.4.3-2 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

Furthermore, the Figures include margins for instrumentation error and pressure drop

(+ 13ºF, -58 psi, and -70 psi P).

Insert Page B 3.4.3-3 Attachment 1, Volume 9, Rev. 1, Page 58 of 478

Attachment 1, Volume 9, Rev. 1, Page 59 of 478 RCS P/T Limits B 3.4.3 BASES APPLICABILITY (continued) maximum pressure. Furthermore, MODES 1 and 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent.

ACTIONS A.1 and A.2 Operation outside the P/T limits during MODE 1, 2, 3, or 4 must be corrected so that the RCPB is returned to a condition that has been 6 is within the verified by stress analyses.

limits of the applicable Figures (i.e., The 30 minute Completion Time reflects the urgency of restoring the Figures 3.4.3-1 and 3.4.3-2) parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable and must be completed before 4 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months continuing operation. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.

8 ASME Code,Section XI, Appendix E (Ref. 7), may be used to support the 7 evaluation. However, its use is restricted to evaluation of the vessel beltline.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable to accomplish the evaluation.

The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed before continuing 4 to operate.

Condition A is modified by a Note requiring Required Action A.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

WOG STS B 3.4.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 59 of 478

Attachment 1, Volume 9, Rev. 1, Page 60 of 478 RCS P/T Limits B 3.4.3 BASES ACTIONS (continued)

B.1 and B.2 any is If a Required Action and associated Completion Time of Condition A are 4 not met, the plant must be placed in a lower MODE because either the resulted in a RCS remained in an unacceptable P/T region for an extended period of determination time increased stress or a sufficiently severe event caused entry into an that the RCS 4 is or may be unacceptable region. Either possibility indicates a need for more careful for continued operation examination of the event, best accomplished with the RCS at reduced pressure and temperature. In reduced pressure and temperature conditions, the possibility of propagation with undetected flaws is decreased.

If the required restoration activity cannot be accomplished within 30 minutes, Required Action B.1 and Required Action B.2 must be implemented to reduce pressure and temperature.

If the required evaluation for continued operation cannot be accomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in Required Action B.1 and Required Action B.2. A favorable evaluation must be completed and documented before returning to operating pressure and temperature conditions.

Pressure and temperature are reduced by bringing the plant to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 with RCS pressure < [500] psig within 5 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 Actions must be initiated immediately to correct operation outside of the P/T limits at times other than when in MODE 1, 2, 3, or 4, so that the 6

is within the RCPB is returned to a condition that has been verified by stress analysis.

limits of the applicable Figures (i.e., The immediate Completion Time reflects the urgency of initiating action to Figures 3.4.3-1 and 3.4.3-2) restore the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.

WOG STS B 3.4.3-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 60 of 478

Attachment 1, Volume 9, Rev. 1, Page 61 of 478 RCS P/T Limits B 3.4.3 BASES ACTIONS (continued)

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and must be completed prior to entry into MODE 4. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, or inspection of the components.

8 ASME Code,Section XI, Appendix E (Ref. 7), may be used to support the 7 evaluation. However, its use is restricted to evaluation of the vessel beltline.

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is required every 1 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status.

Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.

REFERENCES 1. WCAP-7924-A, April 1975. 14278, Rev. 1, "Kewaunee Heatup and Cooldown 2

Limit Curves for Normal Operation."

2. 10 CFR 50, Appendix G.

3. ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.

4. AMSE, Boiler and Pressure Vessel Code,Section XI, Appendix G. 2 WOG STS B 3.4.3-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 61 of 478

Attachment 1, Volume 9, Rev. 1, Page 62 of 478 RCS P/T Limits B 3.4.3 BASES REFERENCES (continued) ASTM E 185-70 (removal) and (evaluation) 5

4. ASTM E 185-82, July 1982. 2 6

5. 10 CFR 50, Appendix H.

7 7

6. Regulatory Guide 1.99, Revision 2, May 1988.

8

7. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

WOG STS B 3.4.3-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 62 of 478

Attachment 1, Volume 9, Rev. 1, Page 63 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.3 BASES, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS

1. Changes are made to reflect those changes made to the Specification.

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. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

4. Changes are made to reflect the ISTS.

5. The ITS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and 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 phrase is changed to be consistent with the actual words in the Required Action.

The curves are not generated from stress analyses; they are generated from fracture mechanics as stated in the Background section.

7. 10 CFR 50 Appendix G also mandates the use of ASME Code,Section XI, Appendix G. Thus, this statement has been modified to include this reference. Due to this addition, subsequent References have been renumbered.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 63 of 478

Attachment 1, Volume 9, Rev. 1, Page 64 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 64 of 478

Attachment 1, Volume 9, Rev. 1, Page 65 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.3, RCS PRESSURE AND TEMPERATURE (P/T) LIMITS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 65 of 478

, Volume 9, Rev. 1, Page 66 of 478 ATTACHMENT 4 ITS 3.4.4, RCS LOOPS - MODES 1 AND 2 , Volume 9, Rev. 1, Page 66 of 478

, Volume 9, Rev. 1, Page 67 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 67 of 478

Attachment 1, Volume 9, Rev. 1, Page 68 of 478 ITS A01 ITS 3.4.4 3.1 REACTOR COOLANT SYSTEM APPLICABILITY Applies to the OPERATING status of the Reactor Coolant System (RCS).

OBJECTIVE To specify those LIMITING CONDITIONS FOR OPERATION of the Reactor Coolant System which must be met to ensure safe reactor operation.

SPECIFICATIONS

a. Operational Components

1. Reactor Coolant Pumps See ITS 3.4.5, 3.4.6, 3.4.7, 3.4.8, A. At least one reactor coolant pump or one residual heat removal pump shall be in 3.9.3 and 3.9.4 operation when a reduction is made in the boron concentration of the reactor coolant.

M01 Applicability B. When the reactor is in the OPERATING mode, except for low power tests, both M02 LCO 3.4.4 reactor coolant pumps shall be in operation. A02 OPERABLE and C. A reactor coolant pump shall not be started with one or more of the RCS cold leg See ITS temperatures 200°F unless the secondary water temperature of each steam 3.4.7 and 3.4.12 generator is < 100°F above each of the RCS cold leg temperatures.

2. Decay Heat Removal Capability A. At least two of the following four heat sinks shall be OPERABLE whenever the average reactor coolant temperature is 350°F but > 200°F.

1. Steam Generator 1A

2. Steam Generator 1B See ITS

3. Residual Heat Removal Train A 3.4.6

4. Residual Heat Removal Train B If less than the above number of required heat sinks are OPERABLE, then corrective action shall be taken immediately to restore the minimum number to the OPERABLE status.

Add proposed ACTION A M01 Add proposed SR 3.4.4.1 M03 Amendment No. 165 TS 3.1-1 03/11/2003 Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 68 of 478

Attachment 1, Volume 9, Rev. 1, Page 69 of 478 DISCUSSION OF CHANGES ITS 3.4.4, RCS LOOPS - MODES 1 AND 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.a.1.B states, in part, both reactor coolant pumps shall be in operation.

ITS LCO 3.4.4 states two RCS loops shall be OPERABLE and in operation. This changes the CTS by requiring the RCS loops to be OPERABLE.

This change is acceptable because it is consistent with the current use and understanding of the LCO. It is not sufficient for an RCS loop to be in operation if it is not capable of performing its safety function (i.e., OPERABLE). This change is designated as administrative as it clarifies the current understanding of a requirement.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.a.1.B states, in part, that both reactor coolant pumps shall be in operation when the reactor is in the OPERATING (equivalent to ITS MODE 1) mode. ITS 3.4.4 requires two RCS loops to be OPERABLE and in operation in MODE 1 (equivalent to CTS OPERATING) and MODE 2 (equivalent to CTS HOT STANDBY). Furthermore, in the event there are less than two reactor coolant pumps operating, CTS 3.1.a.1.B does not contain any actions to be taken; CTS 3.0.c would be entered. CTS 3.0.c requires action to be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and to be in HOT STANDBY (equivalent to ITS MODE 2) in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ITS 3.4.4 adds an ACTION statement to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if the requirements of the LCO (i.e., two RCS loops OPERABLE and in operation) are not met. This changes the CTS by requiring the RCS loops be OPERABLE and in operation in MODE 2 and an appropriate ACTION to exit the Applicability if the LCO is not met. The change concerning the word "OPERABLE" is discussed in DOC A02.

The addition of the MODE 2 Applicability is acceptable since KPS is currently limited to less than 2 percent rated thermal power when only one reactor coolant pump is in operation. Less than two percent of rated thermal power is the CTS Fission Power value for HOT STANDBY (equivalent to ITS MODE 2) mode. If the plant were to operate with only one reactor coolant pump in operation, there would be reverse flow through the inactive reactor coolant loop due to the pressure difference across the reactor vessel and because there are no isolation or check valves in the reactor coolant loops. The change is also 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 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 69 of 478

Attachment 1, Volume 9, Rev. 1, Page 70 of 478 DISCUSSION OF CHANGES ITS 3.4.4, RCS LOOPS - MODES 1 AND 2 in MODE 3 ensures a unit shutdown is commenced and completed within a reasonable period of time. This change is more restrictive because a new Applicability (MODE 2) and appropriate ACTION with a shorter completion time have been added.

M02 CTS 3.1.a.1.B states, in part, both reactor coolant pumps are required to be in operation except for low power tests. ITS 3.4.4 does not include this exception; the reactor coolant pumps are required during PHYSICS TESTS. This changes the CTS by requiring the reactor coolant pumps to be OPERABLE during PHYSICS TESTS.

The purpose of CTS 3.1.a.1.B is to ensure the reactor coolant pumps are OPERABLE under both normal operating and accident conditions. Since Physics Tests do not require the RCPs to be inoperable to perform the tests, there is no reason to maintain this current allowance. Therefore, this change is acceptable and is more restrictive because the reactor coolant pumps are now required to be OPERABLE under more conditions in the ITS than in the CTS.

M03 CTS 3.1.a.1.B does not contain a Surveillance Requirement to verify each RCS loop is in operation. ITS SR 3.4.4.1 requires verification that each RCS loop is in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by requiring the operation of each RCS loop be verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The purpose of CTS 3.1.a.1.B is to ensure that the RCS loops are in operation when the plant is in the OPERATING mode. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months frequency for the proposed Surveillance is selected based on operating experience and the need for operator awareness. This change is more restrictive because a new Surveillance Requirement has been added to ensure the RCS loops are periodically verified to be in operation.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 70 of 478

Attachment 1, Volume 9, Rev. 1, Page 71 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 71 of 478

Attachment 1, Volume 9, Rev. 1, Page 72 of 478 CTS RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops - MODES 1 and 2 Two 3.1.a.1.B LCO 3.4.4 [Four] RCS loops shall be OPERABLE and in operation. 1 3.1.a.1.B APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC M01 A. Requirements of A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months LCO not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M03 SR 3.4.4.1 Verify each RCS loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months WOG STS 3.4.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 72 of 478

Attachment 1, Volume 9, Rev. 1, Page 73 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.4, RCS LOOPS - MODES 1 AND 2

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.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 73 of 478

Attachment 1, Volume 9, Rev. 1, Page 74 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 74 of 478

Attachment 1, Volume 9, Rev. 1, Page 75 of 478 All changes are 1 RCS Loops - MODES 1 and 2 B 3.4.4 unless otherwise noted B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 RCS Loops - MODES 1 and 2 BASES BACKGROUND The primary function of the RCS is removal of the heat generated in the fuel due to the fission process, and transfer of this heat, via the steam generators (SGs), to the secondary plant.

The secondary functions of the RCS include:

a. Moderating the neutron energy level to the thermal state, to increase the probability of fission, ;

b. Improving the neutron economy by acting as a reflector, ;

c. Carrying the soluble neutron poison, boric acid, ;

2

d. Providing a second barrier against fission product release to the environment, and

e. Removing the heat generated in the fuel due to fission product decay following a unit shutdown.

two The reactor coolant is circulated through [four] loops connected in parallel to the reactor vessel, each containing an SG, a reactor coolant pump (RCP), and appropriate flow and temperature instrumentation for both control and protection. The reactor vessel contains the clad fuel. The SGs provide the heat sink to the isolated secondary coolant. The RCPs circulate the coolant through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and prevent fuel damage. This forced circulation of the reactor coolant ensures mixing of the coolant for proper boration and chemistry control.

APPLICABLE Safety analyses contain various assumptions for the design bases SAFETY accident initial conditions including RCS pressure, RCS temperature, ANALYSES reactor power level, core parameters, and safety system setpoints. The important aspect for this LCO is the reactor coolant forced flow rate, which is represented by the number of RCS loops in service.

Both transient and steady state analyses have been performed to establish the effect of flow on the departure from nucleate boiling (DNB).

The transient and accident analyses for the plant have been performed assuming [four] RCS loops are in operation. The majority of the plant two WOG STS B 3.4.4-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 75 of 478

Attachment 1, Volume 9, Rev. 1, Page 76 of 478 All changes are 1 RCS Loops - MODES 1 and 2 B 3.4.4 unless otherwise noted BASES APPLICABLE SAFETY ANALYSES (continued) safety analyses are based on initial conditions at high core power or zero power. The accident analyses that are most important to RCP operation two are the [four] pump coastdown, single pump locked rotor, single pump (broken shaft or coastdown), and rod withdrawal events (Ref. 1).

two Steady state DNB analysis has been performed for the [four] RCS loop operation. For [four] RCS loop operation, the steady state DNB analysis, which generates the pressure and temperature Safety Limit (SL) (i.e., the departure from nucleate boiling ratio (DNBR) limit) assumes a maximum power level of 109% RTP. This is the design overpower condition for

[four] RCS loop operation. The value for the accident analysis setpoint of the nuclear overpower (high flux) trip is 107% and is based on an analysis assumption that bounds possible instrumentation errors. The DNBR limit defines a locus of pressure and temperature points that result in a minimum DNBR greater than or equal to the critical heat flux correlation limit. INSERT 1 3 The plant is designed to operate with all RCS loops in operation to maintain DNBR above the SL, during all normal operations and anticipated transients. By ensuring heat transfer in the nucleate boiling region, adequate heat transfer is provided between the fuel cladding and the reactor coolant.

RCS Loops - MODES 1 and 2 satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The purpose of this LCO is to require an adequate forced flow rate for core heat removal. Flow is represented by the number of RCPs in operation for removal of heat by the SGs. To meet safety analysis acceptance criteria for DNB, [four] pumps are required at rated power.

two An OPERABLE RCS loop consists of an OPERABLE RCP in operation providing forced flow for heat transport and an OPERABLE SG.

APPLICABILITY In MODES 1 and 2, the reactor is critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the assumptions of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DNB and core damage. 3 unacceptable The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, and 5.

WOG STS B 3.4.4-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 76 of 478

Attachment 1, Volume 9, Rev. 1, Page 77 of 478 B 3.4.4 3

INSERT 1 These analyses establish allowable reactor coolant average temperature for the minimum measured flow and power distribution as a function of RCS pressure.

These analyses also establish a locus of power, pressure, and temperature conditions for which the departure from nucleate boiling ratio (DNBR) is equal to its Safety Limit value. The area of permissible operation is bounded by the following combination of reactor trips: high neutron flux (fixed setpoint), high pressurizer pressure (fixed setpoint), low pressurizer pressure (fixed setpoint),

overpower T (variable setpoint), and overtemperature T (variable setpoint).

The difference between the reactor trip values assumed in the safety analyses and the nominal reactor trip setpoints provides an allowance for instrumentation channel error and setpoint error.

Insert Page B 3.4.4-2 Attachment 1, Volume 9, Rev. 1, Page 77 of 478

Attachment 1, Volume 9, Rev. 1, Page 78 of 478 All changes are 1 RCS Loops - MODES 1 and 2 B 3.4.4 unless otherwise noted BASES APPLICABILITY (continued)

Operation in other MODES is covered by:

LCO 3.4.5, "RCS Loops - MODE 3," ;

LCO 3.4.6, "RCS Loops - MODE 4,"

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled,"

2 LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled,"

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation -

3 High Water Level" (MODE 6), and ;

LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - 4 4 Low Water Level" (MODE 6).

ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.

The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.

SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that each RCS loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to DNB. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS loop performance.

U s 3 14.1.1, 14.1.2, 14.

1.8 REFERENCES

1. FSAR, Section [ ].

WOG STS B 3.4.4-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 78 of 478

Attachment 1, Volume 9, Rev. 1, Page 79 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.4 BASES, RCS LOOPS - MODES 1 AND 2

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 punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis or licensing basis description.

4. The correct ITS number has been provided (changed to be consistent with a number change to the actual specification in Section 3.9).

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 79 of 478

Attachment 1, Volume 9, Rev. 1, Page 80 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 80 of 478

Attachment 1, Volume 9, Rev. 1, Page 81 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.4, RCS LOOPS - MODES 1 AND 2 There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 81 of 478

, Volume 9, Rev. 1, Page 82 of 478 ATTACHMENT 5 ITS 3.4.5, RCS LOOPS - MODE 3 , Volume 9, Rev. 1, Page 82 of 478

, Volume 9, Rev. 1, Page 83 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 83 of 478

Attachment 1, Volume 9, Rev. 1, Page 84 of 478 ITS A01 ITS 3.4.5 3.1 REACTOR COOLANT SYSTEM APPLICABILITY Applies to the OPERATING status of the Reactor Coolant System (RCS).

OBJECTIVE To specify those LIMITING CONDITIONS FOR OPERATION of the Reactor Coolant System which must be met to ensure safe reactor operation.

SPECIFICATIONS M01

a. Operational Components See ITS 3.4.6, 3.4.7, 3.4.8,

1. Reactor Coolant Pumps 3.9.3, and Two RCS loops shall be OPERABLE 3.9.4 A. At least one reactor coolant pump or one residual heat removal pump shall be in LCO 3.4.5 operation when a reduction is made in the boron concentration of the reactor M02 Applicability coolant.

B. When the reactor is in the OPERATING mode, except for low power tests, both See ITS 3.4.4 reactor coolant pumps shall be in operation.

C. A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures 200°F unless the secondary water temperature of each steam See ITS 3.4.7 and generator is < 100°F above each of the RCS cold leg temperatures. 3.4.12

2. Decay Heat Removal Capability A. At least two of the following four heat sinks shall be OPERABLE whenever the average reactor coolant temperature is 350°F but > 200°F.

1. Steam Generator 1A

2. Steam Generator 1B See ITS

3. Residual Heat Removal Train A 3.4.6

4. Residual Heat Removal Train B If less than the above number of required heat sinks are OPERABLE, then corrective action shall be taken immediately to restore the minimum number to the OPERABLE status.

Add proposed ACTION A M01 Add proposed ACTIONS B and C (for 1st condition)

Add proposed ACTION C (for 2nd condition) M03 Add proposed SR 3.4.5.1 M04 Add proposed SR 3.4.5.2, SR 3.4.5.3 and Note M01 Amendment No. 165 TS 3.1-1 03/11/2003 Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 84 of 478

Attachment 1, Volume 9, Rev. 1, Page 85 of 478 DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 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.a.1.A requires one reactor coolant pump (RCP) to be in operation under a certain condition, but it does not provide any requirements that the associated Reactor Coolant System (RCS) loop be OPERABLE. ITS 3.4.5 requires two RCS loops to be OPERABLE. OPERABILITY of an RCS loop is defined in the ITS Bases as an OPERABLE RCP and one OPERABLE steam generator, and an OPERABLE RCP is OPERABLE if it is capable of being powered and able to provide forced flow. Thus, to ensure the RCP loops are OPERABLE, ITS SR 3.4.5.2 requires verification of steam generator secondary side water levels are 5% for required RCS loops every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS SR 3.4.5.3 requires verification that each required RCP is OPERABLE every 7 days by verifying correct breaker alignment and indicated power are available to each required pump. A Note further explains that the Surveillance is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation. Furthermore, ITS 3.4.5 ACTIONS A and B provide the appropriate compensatory measures if one of the RCS loops is inoperable in that restoration of the inoperable RCS loop is required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (ITS 3.4.5 ACTION A) and if not restored, to be in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (ITS 3.4.5 ACTION B). If both RCS loops are inoperable, ITS 3.4.5 ACTION C requires immediate suspension of operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM limit of LCO 3.1.1 and initiation of action to restore one RCS loop to OPERABLE status. This changes the CTS by adding new OPERABILITY requirements for two RCS loops and appropriate ACTIONS and Surveillance Requirements.

The purpose of the OPERABILITY requirements for two RCS loops is to ensure adequate loops are available for forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. While only one RCS loop is required to meet the above purpose, two RCS loops are required for redundancy. The purpose of the proposed ACTIONS is to provide adequate compensatory measures if one or both of the RCS loops are inoperable. The purpose of ITS SR 3.4.5.2 is to ensure that the secondary side water level of the steam generator is sufficient to provide a heat sink for removal of decay heat. The purpose of ITS SR 3.4.5.3 is to ensure that each required pump is OPERABLE. Verification of proper breaker alignment and power availability ensures that an additional reactor coolant pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Both Surveillances help ensure the RCS loops are OPERABLE.

Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 85 of 478

Attachment 1, Volume 9, Rev. 1, Page 86 of 478 DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 These changes are designated as more restrictive because a new LCO requirement and appropriate ACTIONS and Surveillance Requirements have been added.

M02 CTS 3.1.a.1.A, which requires an RCP to be in operation, is applicable in MODE 3 only when a reduction is made in the boron concentration of the reactor coolant. ITS 3.4.5 is applicable at all times when in MODE 3, except for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided a) no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1; and, b) core outlet temperature is maintained at least 10°F below saturation temperature (ITS 3.4.5 LCO Note). This changes the CTS by specifying that the LCO requirement for an RCP to be in operation is applicable in MODE 3 at all times except for those conditions specified in the ITS 3.4.5 LCO Note.

CTS 3.1.a.1 requires at least one reactor coolant pump to be in operation when a reduction is made in the boron concentration of the reactor coolant. The CTS applicability statement applies during only those times when a change in the boron concentration of the reactor coolant is made; it is not for decay heat removal purposes. While a change in boron concentration may occur in essentially any MODE; this discussion is only applicable during those times when the plant is in HOT SHUTDOWN (equivalent to ITS MODE 3). Other MODES are discussed in other ITS discussions. The ITS 3.4.5 Applicability is MODE 3 at all times, except for the conditions in 3.4.5 LCO Note, and primarily addresses the decay heat removal function, but also covers the boron mixing issue.

Reactor coolant pumps are utilized in MODE 3 to provide for removal of residual heat from the reactor core, via the steam generator to the secondary plant, and to ensure proper boron mixing within the reactor coolant. The MODE 3 decay heat removal requirements are low enough that a single RCS loop with one RCP running is sufficient to remove core decay heat. The purpose of the Note is to permit all RCPs to be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period to perform certain infrequent tests which require reactor coolant flow to be stopped.

With the RCPs removed from operation, there is no forced flow of reactor coolant. As a result, the reactor coolant is in natural circulation and there is a risk of boron stratification or the formation of a vapor bubble that may cause a natural circulation flow obstruction should the RCPs be removed from service for longer than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. This change is acceptable because the time limitation to have the RCPs removed from operation ensures that both boron stratification and inadequate residual heat removal do not occur should multiple one hour periods be required to complete the testing. This change is designated more restrictive because the reactor coolant pump is now required to be in operation in MODE 3 at all times except for those conditions specified in the ITS 3.4.5 LCO Note.

M03 CTS 3.1.a.1 does not contain any ACTIONS to take should there be less than the required number of reactor coolant pumps in operation when required. As a result, CTS 3.0.c would be entered, which requires action to be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , to be in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and to be in COLD SHUTDOWN (equivalent to ITS MODE 5) within the subsequent 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . However, since the RCP only has to be in Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 86 of 478

Attachment 1, Volume 9, Rev. 1, Page 87 of 478 DISCUSSION OF CHANGES ITS 3.4.5, RCS LOOPS - MODE 3 operation when a reduction in boron concentration is being made, the CTS 3.0.c requirement does not provide any relevant compensatory measures (i.e., it does not require boron concentration reductions to be suspended). ITS 3.4.5 ACTION C specifies the Required Actions for a required RCS loop not in operation as immediately placing the Rod Control System in a condition incapable of rod withdrawal, immediate suspension of operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and initiation of action to restore one RCS loop to operation. This changes the CTS by adding a new ACTION.

The purpose of CTS LCO 3.0.c is to place the unit outside of the Applicability of the Specification. ITS 3.4.5 ACTION C effectively places the unit outside of the Applicability for a loop not in operation by requiring the plant to immediately place the Rod Control System in a condition incapable of rod withdrawal, suspend operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and to initiate action to restore one RCS loop to operation. These proposed Required Actions reflect the importance of maintaining operation for heat removal and boron mixing. This change is designated as more restrictive because a new proposed ACTION has been added.

M04 CTS 3.1.a.1.A requires one RCP to be in operation under certain conditions, but does not provide a Surveillance Requirement to periodically verify the RCP is in operation. This changes the CTS by adding a new Surveillance Requirement.

The purpose of ITS SR 3.4.5.1 is to ensure that the RCS loops are in operation providing forced flow of the reactor coolant for heat removal. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency for the proposed Surveillance is selected based on operating experience and the need for operator awareness. This will ensure the LCO requirement is periodically verified to be met. This change is designated as more restrictive because a new Surveillance Requirement has been added.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 87 of 478

Attachment 1, Volume 9, Rev. 1, Page 88 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 88 of 478

Attachment 1, Volume 9, Rev. 1, Page 89 of 478 CTS RCS Loops - MODE 3 3.4.5 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.5 RCS Loops - MODE 3 1

DOC M01, LCO 3.4.5 [Two] RCS loops shall be OPERABLE and either:

3.1.a.1.A ,

a. [Two] RCS loops shall be in operation when the Rod Control 2

System is capable of rod withdrawal or

b. One RCS loop shall be in operation when the Rod Control System is not capable of rod withdrawal.

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

All reactor coolant pumps may be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1; and

, "SHUTDOWN MARGIN" 6

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

DOC M02 APPLICABILITY: MODE 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC A. One required RCS loop A.1 Restore required RCS loop 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 3 M01 inoperable. to OPERABLE status.

DOC B. Required Action and B.1 Be in MODE 4. 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s 4 M01 associated Completion Time of Condition A not met.

WOG STS 3.4.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 89 of 478

Attachment 1, Volume 9, Rev. 1, Page 90 of 478 CTS RCS Loops - MODE 3 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. [ One required RCS loop C.1 Restore required RCS loop 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months not in operation with Rod to operation.

Control System capable 2

of rod withdrawal. OR C.2 Place the Rod Control 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ]

System in a condition incapable of rod withdrawal.

DOC M01, D. [Two] [required] RCS D.1 Place the Rod Control Immediately 5 1 2 DOC M03 loops inoperable. System in a condition C C.1 incapable of rod withdrawal.

OR AND 2

Required RCS loop(s) not in operation. D.2 Suspend operations that Immediately would cause introduction of 5 C.2 coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.

AND D.3 Initiate action to restore Immediately 5 one RCS loop to C.3 OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY one is 2

DOC M04 SR 3.4.5.1 Verify required RCS loops are in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months WOG STS 3.4.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 90 of 478

Attachment 1, Volume 9, Rev. 1, Page 91 of 478 CTS RCS Loops - MODE 3 3.4.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY DOC M01 SR 3.4.5.2 Verify steam generator secondary side water levels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months are [17]% for required RCS loops. 1 3 5 both DOC SR 3.4.5.3 -------------------------------NOTE------------------------------

M01 Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation. 3 Verify correct breaker alignment and indicated 7 days power are available to each required pump. 3 WOG STS 3.4.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 91 of 478

Attachment 1, Volume 9, Rev. 1, Page 92 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.5, RCS LOOPS - MODE 3

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 ISTS contains requirements and actions on the Rod Control System based on the assumption that the accident analysis for an uncontrolled Rod Control Cluster Assembly (RCCA) bank withdrawal requires two Reactor Coolant System (RCS) loops to be in operation. The KPS accident analysis for an uncontrolled RCCA bank withdrawal from a subcritical condition assumes only one RCS loop is in operation.

As a result, the ITS LCO does not contain requirements on the reactor trip breakers or the Rod Control System, and requires only one RCS loop to be in operation. This is acceptable since the basis for the change is consistent with the accident analysis assumptions.

3. The ISTS is generically written for Westinghouse plants that have four RCS loops.

For those plants that have three or more RCS loops, there are typically a minimum number of "required" loops (out of the total number of RCS loops) that need to be available to satisfy design and accident considerations. Since KPS is a two loop RCS plant, the use of "required" is not needed since there are a total of two loops, both of which are required to be OPERABLE.

4. Typographical error/omission corrected.

5. Changes are made to reflect those changes made to the ISTS. Subsequent requirements are renumbered or revised, where applicable, to reflect the changes.

6. The title of the LCO has been provided since this is the first reference to the LCO.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 92 of 478

Attachment 1, Volume 9, Rev. 1, Page 93 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 93 of 478

Attachment 1, Volume 9, Rev. 1, Page 94 of 478 All changes are 1 RCS Loops - MODE 3 B 3.4.5 unless otherwise noted B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.5 RCS Loops - MODE 3 BASES BACKGROUND In MODE 3, the primary function of the reactor coolant is removal of decay heat and transfer of this heat, via the steam generator (SG), to the secondary plant fluid. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

two The reactor coolant is circulated through [four] RCS loops, connected in parallel to the reactor vessel, each containing an SG, a reactor coolant pump (RCP), and appropriate flow, pressure, level, and temperature instrumentation for control, protection, and indication. The reactor vessel contains the clad fuel. The SGs provide the heat sink. The RCPs circulate the water through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and prevent fuel damage.

In MODE 3, RCPs are used to provide forced circulation for heat removal during heatup and cooldown. The MODE 3 decay heat removal requirements are low enough that a single RCS loop with one RCP running is sufficient to remove core decay heat. However, [two] RCS loops are required to be OPERABLE to ensure redundant capability for decay heat removal.

APPLICABLE Whenever the reactor trip breakers (RTBs) are in the closed position and SAFETY the control rod drive mechanisms (CRDMs) are energized, an ANALYSES inadvertent rod withdrawal from subcritical, resulting in a power excursion, is possible. Such a transient could be caused by a malfunction of the rod control system. In addition, the possibility of a power excursion 11 due to the ejection of an inserted control rod is possible with the breakers 2 closed or open. Such a transient could be caused by the mechanical failure of a CRDM.

3 one RCS loop shall be Therefore, in MODE 3 with the Rod Control System capable of rod withdrawal, accidental control rod withdrawal from subcritical is postulated and requires at least [two] RCS loops to be OPERABLE and in operation 9 analysis to ensure that the accident analyses limits are met. For those conditions when the Rod Control System is not capable of rod withdrawal, two RCS 4 loops are required to be OPERABLE, but only one RCS loop is required to be in operation to be consistent with MODE 3 accident analyses.

WOG STS B 3.4.5-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 94 of 478

Attachment 1, Volume 9, Rev. 1, Page 95 of 478 All changes are 1 RCS Loops - MODE 3 B 3.4.5 unless otherwise noted BASES APPLICABLE SAFETY ANALYSES (continued)

Failure to provide decay heat removal may result in challenges to a fission product barrier. The RCS loops are part of the primary success path that functions or actuates to prevent or mitigate a Design Basis Accident or transient that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier.

RCS Loops - MODE 3 satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

9 LCO and one RCS The purpose of this LCO is to require that at least [two] RCS loops be 5 loop in operation OPERABLE. In MODE 3 with the Rod Control System capable of rod withdrawal, [two] RCS loops must be in operation. [Two] RCS loops are 4

required to be in operation in MODE 3 with the Rod Control System capable of rod withdrawal due to the postulation of a power excursion because of an inadvertent control rod withdrawal. The required number of RCS loops in operation ensures that the Safety Limit criteria will be met 3

for all of the postulated accidents. accident analysis 4

When the Rod Control System is not capable of rod withdrawal, only one RCS loop in operation is necessary to ensure removal of decay heat from the core and homogenous boron concentration throughout the RCS. An additional RCS loop is required to be OPERABLE to ensure that safety 3 analyses limits are met. redundant capability for heat removal The Note permits all RCPs to be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to perform tests that are designed to validate various accident analyses values. One of these tests is validation of the pump coastdown curve used as input to a number of accident analyses including a loss of flow accident. This test is generally performed in MODE 3 during the initial startup testing program, and as such should only be performed once. If, however, changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values of the coastdown curve must be revalidated by conducting the test again. Another test performed during the startup testing program is the validation of rod drop times during cold conditions, both with and without flow.

The no flow test may be performed in MODE 3, 4, or 5 and requires that the pumps be stopped for a short period of time. The Note permits the stopping of the pumps in order to perform this test and validate the assumed analysis values. As with the validation of the pump coastdown curve, this test should be performed only once unless the flow characteristics of the RCS are changed. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period specified is adequate to perform the desired tests, and operating experience has shown that boron stratification is not a problem during this short period with no forced flow.

WOG STS B 3.4.5-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 95 of 478

Attachment 1, Volume 9, Rev. 1, Page 96 of 478 All changes are 1 RCS Loops - MODE 3 B 3.4.5 unless otherwise noted BASES LCO (continued)

Utilization of the Note is permitted provided the following conditions are met, along with any other conditions imposed by initial startup test procedures: 7 "SHUTDOWN MARGIN",

a. No operations are permitted that would dilute the RCS boron concentration with coolant at boron concentrations less than required to assure the SDM of LCO 3.1.1, thereby maintaining the margin to criticality. Boron reduction with coolant at boron concentrations less than required to assure SDM is maintained is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation and 6

b. Core outlet temperature is maintained at least 10°F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

An OPERABLE RCS loop consists of one OPERABLE RCP and one OPERABLE SG, which has the minimum water level specified in SR 3.4.5.2. An RCP is OPERABLE if it is capable of being powered and is able to provide forced flow if required. 5 APPLICABILITY In MODE 3, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing.

The most stringent condition of the LCO, that is, two RCS loops OPERABLE and two RCS loops in operation, applies to MODE 3 with the 4

Rod Control System capable of rod withdrawal. The least stringent condition, that is, two RCS loops OPERABLE and one RCS loop in operation, applies to MODE 3 with the Rod Control System not capable of rod withdrawal.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and 2,"

LCO 3.4.6, "RCS Loops - MODE 4," ;

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled,"

6 LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled,"

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation -

3 High Water Level" (MODE 6), and ;

LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation -

8 4 Low Water Level" (MODE 6).

WOG STS B 3.4.5-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 96 of 478

Attachment 1, Volume 9, Rev. 1, Page 97 of 478 All changes are 1 RCS Loops - MODE 3 B 3.4.5 unless otherwise noted BASES ACTIONS A.1 If one [required] RCS loop is inoperable, redundancy for heat removal is lost. The Required Action is restoration of the required RCS loop to 5 OPERABLE status within the Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This time allowance is a justified period to be without the redundant, nonoperating loop because a single loop in operation has a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core and because of the low probability of a failure in the remaining loop occurring during this period.

B.1 If restoration for Required Action A.1 is not possible within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the unit must be brought to MODE 4. In MODE 4, the unit may be placed on the Residual Heat Removal System. The additional Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is compatible with required operations to achieve cooldown and depressurization from the existing plant conditions in an orderly manner and without challenging plant systems.

[ C.1 and C.2 If one required RCS loop is not in operation, and the Rod Control System is capable of rod withdrawal, the Required Action is either to restore the required RCS loop to operation or to place the Rod Control System in a condition incapable of rod withdrawal (e.g., de-energize all CRDMs by opening the RTBs or de-energizing the motor generator (MG) sets).

4 When the Rod Control System is capable of rod withdrawal, it is postulated that a power excursion could occur in the event of an inadvertent control rod withdrawal. This mandates having the heat transfer capacity of two RCS loops in operation. If only one loop is in operation, the Rod Control System must be rendered incapable of rod withdrawal. The Completion Times of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , to restore the required RCS loop to operation or defeat the Rod Control System is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period. ]

WOG STS B 3.4.5-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 97 of 478

Attachment 1, Volume 9, Rev. 1, Page 98 of 478 All changes are 1 RCS Loops - MODE 3 B 3.4.5 unless otherwise noted BASES ACTIONS (continued)

C.1, C.2, and C.3 5

D.1, D.2, and D.3 the 4 If [two] [required] RCS loops are inoperable or a required RCS loop is not in operation, except as during conditions permitted by the Note in the LCO section, the Rod Control System must be placed in a condition incapable of rod withdrawal (e.g., all CRDMs must be de-energized by opening the RTBs or de-energizing the MG sets). All operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended, and action to restore one of the RCS loops to OPERABLE status and operation must be initiated. Boron dilution requires forced circulation for proper mixing, and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.5.1 one REQUIREMENTS is 4

This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loops are in operation. Verification includes flow rate, temperature, and pump status or 10 monitoring, which help ensure that forced flow is providing heat removal.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS loop performance.

SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY the is verified by ensuring that the secondary side narrow range water level is

[17]% for required RCS loops. If the SG secondary side narrow range 5 5

water level is < [17]%, the tubes may become uncovered and the 5

associated loop may not be capable of providing the heat sink for removal of the decay heat. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to a loss of SG level.

WOG STS B 3.4.5-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 98 of 478

Attachment 1, Volume 9, Rev. 1, Page 99 of 478 All changes are 1 RCS Loops - MODE 3 B 3.4.5 unless otherwise noted BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.5.3 Verification that each required RCP is OPERABLE ensures that safety 5 analyses limits are met. The requirement also ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to each required 5 RCP. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability.

This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation. 5 REFERENCES None.

WOG STS B 3.4.5-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 99 of 478

Attachment 1, Volume 9, Rev. 1, Page 100 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.5 BASES, RCS LOOPS - MODE 3

1. The ISTS Bases 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 ISTS Bases Applicable Safety Analyses section addresses the possibility of a power excursion due to the ejection of an inserted control rod as a result of a control rod drive mechanism (CRDM) mechanical failure. The rod ejection analysis for Kewaunee Power Station (KPS) is performed in MODES 1 and 2, not MODE 3.

Therefore, the discussion on rod ejection is not applicable and has been deleted.

3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis or licensing basis description.

4. The ISTS Bases contains a discussion for those conditions when the Rod Control System is not capable of rod withdrawal based on the assumption that the accident analysis for an uncontrolled Rod Control Cluster Assembly (RCCA) bank withdrawal requires two Reactor Coolant System (RCS) loops to be in operation.

The KPS accident analysis for an uncontrolled RCCA withdrawal from a subcritical condition assumes only one RCS loop is in operation. As a result, the ITS LCO does not contain requirements on the reactor trip breakers or the Rod Control System, and requires only one RCS loop to be in operation. Therefore, the discussion of rod ejection is not applicable and has been deleted.

5. Changes are made to reflect those changes made to the ISTS. Subsequent requirements are renumbered or revised, where applicable, to reflect the changes.

6. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

7. The title of the LCO has been provided since this is the first reference to the LCO.

8. The correct ITS number has been provided (changed to be consistent with a number change to the actual specification in Section 3.9).

9. Since KPS only has two RCS loops, the words "at least" are not necessary.

10. Typographical error corrected. Verification that the loop is in operation can be performed by monitoring any of the three parameters (flow rate, temperature, and pump status). As written, it implies all three must be observed. Also, use of "or" is consistent with similar SRs in the ITS 3.4.4, 3.4.6, 3.4.7, and 3.4.8 Bases.

11. Typographical error corrected. The title of a system is capitalized, as shown in the next paragraph.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 100 of 478

Attachment 1, Volume 9, Rev. 1, Page 101 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 101 of 478

Attachment 1, Volume 9, Rev. 1, Page 102 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.5, RCS LOOPS - MODE 3 There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 102 of 478

, Volume 9, Rev. 1, Page 103 of 478 ATTACHMENT 6 ITS 3.4.6, RCS LOOPS - MODE 4 , Volume 9, Rev. 1, Page 103 of 478

, Volume 9, Rev. 1, Page 104 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 104 of 478

Attachment 1, Volume 9, Rev. 1, Page 105 of 478 ITS A01 ITS 3.4.6 3.1 REACTOR COOLANT SYSTEM APPLICABILITY Applies to the OPERATING status of the Reactor Coolant System (RCS).

OBJECTIVE To specify those LIMITING CONDITIONS FOR OPERATION of the Reactor Coolant System which must be met to ensure safe reactor operation.

SPECIFICATIONS

a. Operational Components

1. Reactor Coolant Pumps A. At least one reactor coolant pump or one residual heat removal pump shall be in LCO 3.4.6 operation when a reduction is made in the boron concentration of the reactor M01 Applicability coolant.

B. When the reactor is in the OPERATING mode, except for low power tests, both See ITS 3.4.4 reactor coolant pumps shall be in operation.

C. A reactor coolant pump shall not be started with one or more of the RCS cold leg See ITS temperatures 200°F unless the secondary water temperature of each steam 3.4.7 and generator is < 100°F above each of the RCS cold leg temperatures. 3.4.12

2. Decay Heat Removal Capability LCO 3.4.6 A. At least two of the following four heat sinks shall be OPERABLE whenever the Applicability average reactor coolant temperature is 350°F but > 200°F.

1. Steam Generator 1A

2. Steam Generator 1B LA01

3. Residual Heat Removal Train A

4. Residual Heat Removal Train B If less than the above number of required heat sinks are OPERABLE, then ACTIONS corrective action shall be taken immediately to restore the minimum number to A and B the OPERABLE status. M02 Add proposed Required Actions A.2 and B.1 Add proposed ACTION B (for 2nd Condition) M03 Add proposed SR 3.4.6.1 M04 Add proposed SR 3.4.6.2 and SR 3.4.6.3 M05 Amendment No. 165 TS 3.1-1 03/11/2003 Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 105 of 478

Attachment 1, Volume 9, Rev. 1, Page 106 of 478 DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4 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.a.1.A, which requires a reactor coolant pump (RCP) or residual heat removal (RHR) pump to be in operation, is applicable in MODE 4 only when a reduction is made in the boron concentration of the reactor coolant. ITS 3.4.6 is applicable at all times when in MODE 4, except for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided a) no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1; and b) core outlet temperature is maintained at least 10ºF below saturation temperature (ITS 3.4.6 LCO Note). This changes the CTS by specifying that the LCO requirement for an RCP or RHR pump to be in operation is applicable in MODE 4 at all times except for those conditions specified in ITS 3.4.6 LCO Note.

CTS 3.1.a.1.A requires at least one RCP or one RHR pump to be in operation when a reduction is made in the boron concentration of the reactor coolant. The CTS applicability statement applies during only those times when a change in the boron concentration of the reactor coolant is made; it is not for decay heat removal purposes. While a change in boron concentration may occur in essentially any MODE; this discussion is only applicable during those times when the plant is in MODE 4 (i.e., > 200°F and 350°F). Other MODES are discussed in other ITS discussions. The ITS 3.4.6 Applicability is MODE 4 at all times, except for the conditions in ITS 3.4.6 LCO Note, and primarily addresses the decay heat removal function, but also covers the boron mixing issue. RCPs or RHR pumps are utilized in MODE 4 to provide for removal of decay heat from the reactor core, via the steam generator to the secondary plant or through the RHR heat exchangers, and to ensure proper boron mixing within the reactor coolant.

The MODE 4 decay heat removal requirements are low enough that a single RCS loop with one RCP running or a single RHR pump is sufficient to remove core decay heat. The purpose of the Note is to permit all RCPs and RHR pumps to be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period to perform certain infrequent tests which require reactor coolant flow to be stopped. With the RCPs and RHR pumps removed from operation, there is no forced flow of reactor coolant. As a result, the reactor coolant is in natural circulation and there is a risk of boron stratification or the formation of a vapor bubble that may cause a natural circulation flow obstruction should the RCPs and RHR pumps be removed from service for longer than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. This change is acceptable because the time limitation to have the RCPs and RHR pumps removed from operation ensures that adequate boron mixing and residual heat removal are Kewaunee Power Station Page 1 of 4 Attachment 1, Volume 9, Rev. 1, Page 106 of 478

Attachment 1, Volume 9, Rev. 1, Page 107 of 478 DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4 maintained should multiple one hour periods be required to complete the testing.

This change is designated more restrictive because one RCP or RHR pump is required to be in operation in MODE 4 at all times except for those conditions specified in the ITS 3.4.6 LCO Note.

M02 CTS 3.1.a.2.A requires that if less than the required number of heat sinks are OPERABLE (i.e., one or both of the required heat sinks are inoperable), then corrective action shall be taken immediately to restore the minimum number to OPERABLE status. ITS 3.4.6 ACTION A provides the actions when one required heat sink is inoperable and ITS 3.4.6 ACTION B provides the actions when both required heat sinks are inoperable. Both ACTIONS require immediate action to be taken to restore the inoperable loops to OPERABLE status. In addition, ITS 3.4.6 Required Action A.2 requires that the unit be placed in MODE 5 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the inoperable loop is an RHR loop. A Note for Required Action A.2 states that this action is only required if an RHR loop is OPERABLE. Furthermore, ITS 3.4.6 Required Action B.1 requires immediate suspension of operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1. This changes the CTS by adding new Required Actions when one or both required loops are inoperable.

The purpose of the CTS 3.1.a.2.A action is to provide compensatory measures for when one or both required heat sinks are inoperable. ITS 3.4.6 ACTION A places the unit outside of the Applicability of the Specification by requiring the plant to be in MODE 5 in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months should the inoperable loop not be restored to OPERABLE. The change is acceptable because allowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to be in MODE 5 is reasonable, based on operating experience, to achieve MODE 5 from MODE 4 in an orderly manner and without challenging plant systems. In addition, bringing the unit to MODE 5 is a conservative action, with regard to residual heat removal, and with only one RHR loop OPERABLE, redundancy for residual heat removal is lost. In the event of a loss of the remaining RHR loop, it would be safer to initiate that loss from MODE 5 rather than MODE 4. If both required loops are inoperable, ITS 3.4.6 Required Action B.1 ensures that a boron reduction cannot occur since there are no pumps running to mix the coolant. This change is designated as more restrictive because the time to exit the Applicability of the LCO has been reduced from 49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

M03 CTS 3.1.a.1.A does not contain any ACTIONS to take should there be less than the required number of RHR pumps in operation while in MODE 4. As a result, CTS 3.0.c would be entered, which requires action to be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , to be in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and to be in COLD SHUTDOWN (equivalent to ITS MODE 5) within the subsequent 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . However, since the RCP or RHR pump only has to be in operation when a reduction in boron concentration is being made, the CTS 3.0.c requirement does not really provide any relevant compensatory measures; i.e., it does not require boron concentration reductions to be suspended. ITS 3.4.6 ACTION B specifies the Required Actions for a required RCS or RHR loop not in operation, and requires immediate suspension of operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and initiation of action to restore one RCS or RHR loop to operation. This changes the CTS by adding a new ACTION.

Kewaunee Power Station Page 2 of 4 Attachment 1, Volume 9, Rev. 1, Page 107 of 478

Attachment 1, Volume 9, Rev. 1, Page 108 of 478 DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4 The purpose of CTS LCO 3.0.c is to place the unit outside of the Applicability of the Specification. ITS 3.4.6 ACTION B effectively places the unit outside of the Applicability for a loop not in operation by requiring the plant to immediately suspend operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and to initiate action to restore one RCS or RHR loop to operation. The proposed Required Actions reflect the importance of maintaining operation for decay heat removal and boron mixing. This change is designated as more restrictive because a new proposed ACTION has been added.

M04 CTS 3.1.a.1.A requires one RCP or RHR pump to be in operation under certain conditions, but does not provide a Surveillance Requirement to periodically verify the required pump is in operation. ITS SR 3.4.6.1 requires verification that the required RHR or RCS loop is in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by adding a new Surveillance Requirement.

The purpose of ITS SR 3.4.6.1 is to ensure that one RCS or RHR loop is in operation providing forced flow of the reactor coolant for heat removal. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency for the proposed Surveillance is selected based on operating experience and the need for operator awareness. This will ensure the LCO requirement is periodically verified to be met. This change is designated as more restrictive because a new Surveillance Requirement has been added.

M05 CTS 3.1.a.2.A requires two heat sinks to be OPERABLE, but does not provide any Surveillance Requirements to periodically verify the required loops are OPERABLE. ITS SR 3.4.6.2 requires verification of steam generator secondary side water levels are 5% for required RCS loops every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS SR 3.4.6.3 requires verification that each required pump is OPERABLE every 7 days by verifying correct breaker alignment and indicated power are available to each required pump. A Note further explains that the Surveillance is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation. This changes the CTS by adding new Surveillance Requirements to periodically verify the RCS or RHR loops are OPERABLE.

The purpose of ITS SR 3.4.6.2 is to ensure that the secondary side water level of the steam generator is sufficient to provide a heat sink for removal of decay heat.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency for the proposed Surveillance is selected based on operating experience and the need for operator awareness. The purpose of ITS SR 3.4.6.3 is to ensure that each required pump is OPERABLE. Verification of proper breaker alignment and power availability ensures that an additional reactor coolant pump can be placed in operation, if needed, to maintain residual heat removal and reactor coolant circulation. These changes are acceptable because the surveillance requirements ensure the availability of the system to remove reactor residual heat and to provide proper boron mixing via the forced circulation of the reactor coolant. This change is designated as more restrictive because new Surveillance Requirements have been added to ensure that the RCS or RHR loops are periodically verified to be OPERABLE.

Kewaunee Power Station Page 3 of 4 Attachment 1, Volume 9, Rev. 1, Page 108 of 478

Attachment 1, Volume 9, Rev. 1, Page 109 of 478 DISCUSSION OF CHANGES ITS 3.4.6, RCS LOOPS - MODE 4 RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.1.a.2.A contains a listing of four heat sinks that are utilized for decay heat removal. The heat sinks are Steam Generator 1A, Steam Generator 1B, Residual Heat Removal Train A, and Residual Heat Removal Train B. ITS LCO 3.4.6, in part, requires two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops to be OPERABLE. The ITS does not define the components that comprise an OPERABLE loop. This changes the CTS by moving the description of the RCS and RHR loops to the Bases.

The removal of these details, which are related to 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. The ITS retains all necessary requirements in the LCO to ensure OPERABILITY of the required RCS and RHR loops in MODE 4. Also, 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 Specification 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 information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 4 of 4 Attachment 1, Volume 9, Rev. 1, Page 109 of 478

Attachment 1, Volume 9, Rev. 1, Page 110 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 110 of 478

Attachment 1, Volume 9, Rev. 1, Page 111 of 478 CTS RCS Loops - MODE 4 All changes are 1 3.4.6 unless otherwise noted 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops - MODE 4 DOC M01, LCO 3.4.6 Two loops consisting of any combination of RCS loops and residual heat 3.1.a.2.A removal (RHR) loops shall be OPERABLE, and one loop shall be in operation.

NOTES------------------------------------------- 2

1. All reactor coolant pumps (RCPs) and RHR pumps may be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1; and

, "SHUTDOWN MARGIN" 3

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

2. No RCP shall be started with any RCS cold leg temperature

[275°F] [Low Temperature Overpressure Protection (LTOP) 2 arming temperature specified in the PTLR] unless the secondary side water temperature of each steam generator (SG) is [50]°F above each of the RCS cold leg temperatures.

3.1.a.2.A APPLICABILITY: MODE 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.2.A A. One required loop A.1 Initiate action to restore a Immediately inoperable. second loop to OPERABLE status.

AND WOG STS 3.4.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 111 of 478

Attachment 1, Volume 9, Rev. 1, Page 112 of 478 CTS RCS Loops - MODE 4 3.4.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.2.A A.2 --------------NOTE--------------

Only required if RHR loop is OPERABLE.

Be in MODE 5. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.1.a.2.A, B. Two required loops B.1 Suspend operations that Immediately DOC M03 inoperable. would cause introduction of coolant into the RCS with OR boron concentration less than required to meet SDM Required loop not in of LCO 3.1.1.

operation.

AND B.2 Initiate action to restore one Immediately loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC SR 3.4.6.1 Verify required RHR or RCS loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months M04 1

DOC SR 3.4.6.2 Verify SG secondary side water levels are [17]% 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months M05 for required RCS loops.

5 DOC SR 3.4.6.3 -------------------------------NOTE------------------------------

M05 Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated 7 days power are available to each required pump.

WOG STS 3.4.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 112 of 478

Attachment 1, Volume 9, Rev. 1, Page 113 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.6, RCS LOOPS - MODE 4

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. ISTS 3.4.6 LCO Note 2 provides a temperature restriction for starting a reactor coolant pump (RCP). The temperature value is the Low Temperature Overpressure Protection (LTOP) arming temperature, and for the ISTS is < 275ºF. However, the value for Kewaunee Power Station is < 200ºF, as shown in CTS 3.1.a.1.C. Since the minimum temperature for MODE 4 is defined as > 200ºF, the restriction is not needed in this Specification (i.e., the plant cannot be in MODE 4 and violate the RCP minimum temperature restriction for startup). Therefore, this Note has not been included in the Kewaunee ITS. Furthermore, since there is only one Note for the LCO, the term "NOTES" has been changed to "NOTE" and numbering of the first Note is not required.

3. The title of the LCO has been provided since this is the first reference to the LCO.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 113 of 478

Attachment 1, Volume 9, Rev. 1, Page 114 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 114 of 478

Attachment 1, Volume 9, Rev. 1, Page 115 of 478 All changes are 1 RCS Loops - MODE 4 B 3.4.6 unless otherwise noted B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.6 RCS Loops - MODE 4 BASES BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the steam generator (SG) secondary side coolant or the component cooling water via the residual heat removal (RHR) heat exchangers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

two The reactor coolant is circulated through [four] RCS loops connected in parallel to the reactor vessel, each loop containing an SG, a reactor coolant pump (RCP), and appropriate flow, pressure, level, and temperature instrumentation for control, protection, and indication. The RCPs circulate the coolant through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and to prevent boric acid stratification.

In MODE 4, either RCPs or RHR loops can be used to provide forced circulation. The intent of this LCO is to provide forced flow from at least one RCP or one RHR loop for decay heat removal and transport. The OPERABLE flow provided by one RCP loop or RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that two paths be 2 available to provide redundancy for decay heat removal.

APPLICABLE In MODE 4, RCS circulation is considered in the determination of the time 3 SAFETY INSERT 1 available for mitigation of the accidental boron dilution event. The RCS ANALYSES and RHR loops provide this circulation. an would affect RCS Loops - MODE 4 satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The purpose of this LCO is to require that at least two loops be OPERABLE in MODE 4 and that one of these loops be in operation. The LCO allows the two loops that are required to be OPERABLE to consist of any combination of RCS loops and RHR loops. Any one loop in operation provides enough flow to remove the decay heat from the core with forced circulation. An additional loop is required to be OPERABLE to provide redundancy for heat removal.

The 5

Note 1 permits all RCPs or RHR pumps to be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests that are designed to validate various accident analyses values. One of the tests performed during the startup testing program is the validation of rod drop times during cold conditions, both with and without flow. The no flow test may be performed in MODE 3, 4, or 5 and requires that the pumps be WOG STS B 3.4.6-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 115 of 478

Attachment 1, Volume 9, Rev. 1, Page 116 of 478 B 3.4.6 o 3

-INSERT 1 The current licensing basis of Kewaunee Power Station (KPS) does not consider boron dilution events during MODE 4 conditions. Therefore, no safety analyses related to the loss of RCS loops are performed for MODE 4.

Insert Page B 3.4.6-1 Attachment 1, Volume 9, Rev. 1, Page 116 of 478

Attachment 1, Volume 9, Rev. 1, Page 117 of 478 RCS Loops - MODE 4 B 3.4.6 BASES LCO (continued) stopped for a short period of time. The Note permits the stopping of the pumps in order to perform this test and validate the assumed analysis values. If changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values must be revalidated by conducting the test again. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period is adequate to perform the test, and operating experience has shown that boron stratification is not a problem during this short period with no forced flow.

the 5

Utilization of Note 1 is permitted provided the following conditions are met along with any other conditions imposed by initial startup test procedures:

"SHUTDOWN MARGIN",

a. No operations are permitted that would dilute the RCS boron 6 concentration with coolant with boron concentrations less than required to meet SDM of LCO 3.1.1, therefore maintaining the margin to criticality. Boron reduction with coolant at boron concentrations thereby less than required to assure SDM is maintained is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation and 4

b. Core outlet temperature is maintained at least 10°F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

Note 2 requires that the secondary side water temperature of each SG be

[50]°F above each of the RCS cold leg temperatures before the start of 5 an RCP with any RCS cold leg temperature [275°F] [Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR]. This restraint is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.

An OPERABLE RCS loop comprises an OPERABLE RCP and an OPERABLE SG, which has the minimum water level specified in SR 3.4.6.2.

Similarly for the RHR System, an OPERABLE RHR loop comprises an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RCPs and RHR pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required.

WOG STS B 3.4.6-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 117 of 478

Attachment 1, Volume 9, Rev. 1, Page 118 of 478 RCS Loops - MODE 4 B 3.4.6 BASES APPLICABILITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing.

One loop of either RCS or RHR provides sufficient circulation for these purposes. However, two loops consisting of any combination of RCS and RHR loops are required to be OPERABLE to meet single failure 3 considerations.

provide redundant capability for heat removal Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and 2,"

LCO 3.4.5, "RCS Loops - MODE 3,"

4 LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled,"

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled,"

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation -

3 High Water Level" (MODE 6), and ;

LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - 7 4 Low Water Level" (MODE 6).

ACTIONS A.1 If one required loop is inoperable, redundancy for heat removal is lost.

Action must be initiated to restore a second RCS or RHR loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

A.2 If restoration is not accomplished and an RHR loop is OPERABLE, the unit must be brought to MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Bringing the unit to MODE 5 is a conservative action with regard to decay heat removal. With only one RHR loop OPERABLE, redundancy for decay heat removal is lost and, in the event of a loss of the remaining RHR loop, it would be safer to initiate that loss from MODE 5 rather than MODE 4. The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time, based on operating experience, to reach MODE 5 from MODE 4 in an orderly manner and without challenging plant systems.

This Required Action is modified by a Note which indicates that the unit must be placed in MODE 5 only if a RHR loop is OPERABLE. With no RHR loop OPERABLE, the unit is in a condition with only limited cooldown capabilities. Therefore, the actions are to be concentrated on the restoration of a RHR loop, rather than a cooldown of extended duration.

WOG STS B 3.4.6-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 118 of 478

Attachment 1, Volume 9, Rev. 1, Page 119 of 478 RCS Loops - MODE 4 All changes are 1 B 3.4.6 unless otherwise noted BASES ACTIONS (continued)

B.1 and B.2 If two required loops are inoperable or a required loop is not in operation, 5

the except during conditions permitted by Note 1 in the LCO section, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action to restore one RCS or RHR loop to OPERABLE status and operation must be initiated. The required margin to criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.

The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required RCS or RHR loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS and RHR loop performance.

SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY. SG OPERABILITY 5

is verified by ensuring that the secondary side narrow range water level is 5

[17]%. If the SG secondary side narrow range water level is < [17]%,

the tubes may become uncovered and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to the loss of SG level.

WOG STS B 3.4.6-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 119 of 478

Attachment 1, Volume 9, Rev. 1, Page 120 of 478 RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.6.3 Verification that each required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

REFERENCES None.

WOG STS B 3.4.6-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 120 of 478

Attachment 1, Volume 9, Rev. 1, Page 121 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.6 BASES, RCS LOOPS - MODE 4

1. The ISTS Bases 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 ISTS Bases Background section states the other intent of the LCO is to require that two paths be available to provide redundancy for decay heat removal. This statement has been changed to state the other intent of the LCO is to require the two paths to be OPERABLE to provide redundancy for decay heat removal. This change to the Background statement places it in agreement with the Bases LCO and Specification LCO statements to ensure it is understood that the intent is for the two paths to be OPERABLE.

3. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis or licensing basis description.

4. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

5. Changes are made to be consistent with changes made to the Specification.

6. The title of the LCO has been provided since this is the first reference to the LCO within the LCO 3.4.6 Bases.

7. The correct ITS number has been provided (changed to be consistent with a number change to the actual specification in Section 3.9).

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 121 of 478

Attachment 1, Volume 9, Rev. 1, Page 122 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 122 of 478

Attachment 1, Volume 9, Rev. 1, Page 123 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.6, RCS LOOPS - MODE 4 There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 123 of 478

Attachment 1, Volume 9, Rev. 1, Page 124 of 478 ATTACHMENT 7 ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED Attachment 1, Volume 9, Rev. 1, Page 124 of 478

, Volume 9, Rev. 1, Page 125 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 125 of 478

Attachment 1, Volume 9, Rev. 1, Page 126 of 478 ITS A01 ITS 3.4.7 3.1 REACTOR COOLANT SYSTEM APPLICABILITY Applies to the OPERATING status of the Reactor Coolant System (RCS).

OBJECTIVE To specify those LIMITING CONDITIONS FOR OPERATION of the Reactor Coolant System which must be met to ensure safe reactor operation.

SPECIFICATIONS

a. Operational Components See ITS 3.4.5 and

1. Reactor Coolant Pumps 3.4.6 LCO 3.4.7 A. At least one reactor coolant pump or one residual heat removal pump shall be in operation when a reduction is made in the boron concentration of the reactor M01 Applicability coolant. Add proposed LCO 3.4.7 Notes 1 and 4 See ITS B. When the reactor is in the OPERATING mode, except for low power tests, both 3.4.4 reactor coolant pumps shall be in operation.

C. A reactor coolant pump shall not be started with one or more of the RCS cold leg A02 LCO 3.4.7 temperatures 200°F unless the secondary water temperature of each steam Note 3 generator is < 100°F above each of the RCS cold leg temperatures.

2. Decay Heat Removal Capability A. At least two of the following four heat sinks shall be OPERABLE whenever the average reactor coolant temperature is 350°F but > 200°F.

1. Steam Generator 1A

2. Steam Generator 1B See ITS

3. Residual Heat Removal Train A 3.4.6

4. Residual Heat Removal Train B If less than the above number of required heat sinks are OPERABLE, then corrective action shall be taken immediately to restore the minimum number to the OPERABLE status.

Amendment No. 165 TS 3.1-1 03/11/2003 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 126 of 478

Attachment 1, Volume 9, Rev. 1, Page 127 of 478 ITS A01 ITS 3.4.7 L02 LCO 3.4.7 B. Two residual heat removal trains shall be OPERABLE whenever the average A03 Applicability reactor coolant temperature is 200°F and irradiated fuel is in the reactor, except when in the REFUELING MODE with the minimum water level above the See ITS 3.9.3 top of the vessel flange 23 feet, one train may be inoperable for maintenance.

Add proposed LCO 3.4.7 Note 2

1. Each residual heat removal train shall be comprised of: L01 a) One OPERABLE residual heat removal pump b) One OPERABLE residual heat removal heat exchanger LA01 c) An OPERABLE flow path consisting of all valves and piping associated with the above train of components and required to remove decay heat from the core during normal shutdown situations. This flow path shall be capable of taking suction from the appropriate Reactor Coolant System hot leg and returning to the Reactor Coolant System.

2. If one residual heat removal train is inoperable, then corrective action shall ACTION A be taken immediately to return it to the OPERABLE status.

Add proposed Required Action A.2 L02 Add proposed ACTION B Add proposed ACTION C for loop not in operation M02 Add proposed ACTION C for two M03 inoperable RHR loops Add proposed SR 3.4.7.1 and M04 SR 3.4.7.3 Add proposed SR 3.4.7.2 L02 Amendment No. 165 TS 3.1-2 03/11/2003 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 127 of 478

Attachment 1, Volume 9, Rev. 1, Page 128 of 478 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED 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.a.1.C restricts starting a reactor coolant pump with one or more of the RCS cold leg temperatures < 200°F unless the secondary water temperature of each steam generator is < 100°F above each of the RCS cold leg temperatures.

This restriction is included in ITS 3.4.7 as LCO Note 3; however, the statement "with one or more of the RCS cold leg temperatures < 200°F" is not included in the Note. However, the ITS 3.4.7 Applicability is MODE 5 (which is essentially equivalent to the CTS wording) with RCS loops filled. This changes the CTS by not including the redundant words that are already covered by the MODE 5 Applicability in the Note restriction.

ITS MODE 5 is the equivalent MODE covering the condition of less than or equal to 200°F. This is shown in ITS Table 1.1-1, which defines that MODE 5 has a minimum average reactor coolant temperature of 200°F. Since the Applicability of LCO 3.4.7 is MODE 5, the words are redundant and are not necessary to be included in ITS LCO Note 3. This change is designated as administrative and is acceptable because it does not result in any technical change to the CTS.

A03 CTS 3.1.a.2.B, which requires two RHR trains to be OPERABLE, is applicable whenever the average reactor coolant temperature is 200ºF and irradiated fuel is in the reactor. ITS 3.4.7, which also includes similar RHR train OPERABILITY requirements (as modified by DOC L02), is applicable in MODE 5 with the RCS loops filled. RHR requirements in MODE 5 with the RCS loops not filled are provided in LCO 3.4.8 and the MODE 6 RHR requirements are provided in LCO 3.9.5 and LCO 3.9.6. This changes the CTS by splitting these RHR requirements into four separate LCOs.

This change is acceptable since all facets of MODES 5 and 6 operation are covered in the four ITS Specifications. This change is designated as administrative because it does not result in any technical changes.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.a.1.A, which (in MODE 5) requires a residual heat removal (RHR) pump to be in operation, is applicable in MODE 5 only when a reduction is made in the boron concentration of the reactor coolant. ITS 3.4.7, in part, requires an RHR loop to be in operation and is applicable at all times when in MODE 5 with the RCS loops filled, except as allowed in LCO Notes 1 and 4. Note 1 allows the required RHR pump to not be in operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided a) no operations are permitted that would cause introduction of coolant into the Kewaunee Power Station Page 1 of 6 Attachment 1, Volume 9, Rev. 1, Page 128 of 478

Attachment 1, Volume 9, Rev. 1, Page 129 of 478 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED RCS with boron concentration less than required to meet the SDM of LCO 3.1.1; and b) core outlet temperature is maintained at least 10ºF below saturation temperature. Note 4 allows the required RHR pump to not be in operation during planned heatup to MODE 4 when at least one RCS loop is in operation. This changes the CTS by specifying that the LCO requirement for an RHR pump to be in operation is applicable in MODE 5 with the RCS loops filled at all times except for those conditions specified in ITS 3.4.7 LCO Notes 1 and 4.

In MODE 5, CTS 3.1.a.1.A requires at least one RHR pump to be in operation when a reduction is made in the boron concentration of the reactor coolant. The CTS applicability statement applies during only those times when a change in the boron concentration of the reactor coolant is made; it is not for decay heat removal purposes. While a change in boron concentration may occur in essentially any MODE; this discussion is only applicable during those times when the plant is in COLD SHUTDOWN (equivalent to ITS MODE 5) with the RCS loops filled. Other MODES are discussed in other ITS discussions. The ITS 3.4.7 Applicability is MODE 5 at all times with the RCS loops filled, except for the conditions in ITS 3.4.7 LCO Notes 1 and 4, and primarily addresses the decay heat removal function, but also covers the boron mixing issue. RHR pumps are utilized in MODE 5 to provide for removal of residual heat from the reactor core to the RHR heat exchangers, and to ensure proper boron mixing within the reactor coolant. The MODE 5 decay heat removal requirements are low enough that a single RHR pump is sufficient to remove core decay heat. The purpose of Note 1 is to permit all RHR pumps to be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period to perform certain infrequent tests which require reactor coolant flow to be stopped. With the RHR pumps removed from operation, there is no forced flow of reactor coolant. As a result, the reactor coolant is in natural circulation and there is a risk of boron stratification or the formation of a vapor bubble that may cause a natural circulation flow obstruction should the RHR pumps be removed from service for longer than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. This change is acceptable because the time limitation to have the RHR pumps removed from operation ensures that both boron stratification and inadequate residual heat removal do not occur should multiple one hour periods be required to complete the testing. The purpose of Note 4 is to allow plant heatup to MODE

4. This Note requires though, that an RCS loop be in operation during this time period. This will ensure adequate boron mixing of the reactor coolant during the heatup. This change is designated more restrictive because one RHR pump is required to be in operation in MODE 5 at all times with the RCS loops filled, except for those conditions specified in ITS 3.4.7 LCO Notes 1 and 4.

M02 CTS 3.1.a.1.A does not contain any ACTIONS to take should there be less than the required number of RHR pumps in operation while in MODE 5. As a result, CTS 3.0.c would be normally entered. However, LCO 3.0.c states that it is not applicable in COLD SHUTDOWN or REFUELING. Since the RHR pump only has to be in operation when a reduction in boron concentration is being made, and for this Specification, the unit is already in MODE 5, the CTS does not provide any compensatory measures. Therefore, 10 CFR 50.36(c)(2)(i) would apply, which states to shutdown the unit. However, no times are provided to complete the shutdown and no further actions (i.e., suspend boron concentration reduction) are required. ITS 3.4.7 ACTION C specifies the Required Actions for a required RHR loop not in operation, and requires immediate suspension of Kewaunee Power Station Page 2 of 6 Attachment 1, Volume 9, Rev. 1, Page 129 of 478

Attachment 1, Volume 9, Rev. 1, Page 130 of 478 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and initiation of action to restore one RHR loop to operation. This changes the CTS by adding a new ACTION when the required RHR pump is not in operation.

The purpose of CTS LCO 3.0.c is to place the unit outside of the Applicability of the Specification. The change is acceptable since ITS 3.4.7 ACTION C effectively places the unit outside of the Applicability for a loop not in operation by requiring the plant to immediately suspend operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and to initiate action to restore one RHR loop to operation. The proposed Required Actions reflect the importance of maintaining operation for decay heat removal and boron mixing. This change is designated as more restrictive because a new proposed ACTIONS has been added M03 CTS 3.1.a.2.B does not contain any ACTIONS to take if both required RHR loops are inoperable. As a result, CTS 3.0.c would be normally entered. However, LCO 3.0.c states that it is not applicable in COLD SHUTDOWN or REFUELING.

Since the RHR pump only has to be in operation when a reduction in boron concentration is being made, and for this Specification, the unit is already in MODE 5, the CTS does not provide any compensatory measures. Therefore, 10 CFR 50.36(c)(2)(i) would apply, which states to shutdown the unit. However, no times are provided to complete the shutdown and no further actions (i.e.,

suspend boron concentration reductions) are required. ITS 3.4.7 ACTION C (ISTS 3.4.7 ACTION C) provides the Required Actions when no required loops are OPERABLE. The Required Actions are to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1 and to immediately initiate action to restore one RHR loop to OPERABLE status. This changes the CTS by adding a new ACTION when both required RHR loops are inoperable.

The change is acceptable because the Completion Times are 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 an event occurring during the allowed Completion Times. The immediate suspension of operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and immediately initiating action to restore one loop to OPERABLE status reflects the importance of maintaining decay heat removal and boron mixing capability. Also, coolant added to the RCS without forced circulation could introduce unmixed coolant into the core, which could reduce the required margin to criticality. This change is designated as more restrictive because a new ACTION is being added to the ITS that was not required by the CTS.

M04 CTS 3.1.a.1.A requires one RHR pump to be in operation under certain conditions, but does not provide a Surveillance Requirement to periodically verify the required pump is in operation. ITS SR 3.4.7.1 requires verification that the required RHR loop is in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . CTS 3.1.a.2.B requires two RHR trains to be OPERABLE, but does not provide a Surveillance Requirement to periodically verify the required loops are OPERABLE. ITS SR 3.4.7.3 requires Kewaunee Power Station Page 3 of 6 Attachment 1, Volume 9, Rev. 1, Page 130 of 478

Attachment 1, Volume 9, Rev. 1, Page 131 of 478 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED verification that each required RHR pump is OPERABLE every 7 days by verifying correct breaker alignment and indicated power are available to each required pump. A Note further explains that the Surveillance is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation. This changes the CTS by adding new Surveillance Requirements to periodically verify the required pump is in operation and the required pumps are OPERABLE.

The purpose of ITS SR 3.4.7.1 is to ensure that one RHR loop is in operation providing forced flow of the reactor coolant for heat removal. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months frequency for the proposed Surveillance is selected based on operating experience and the need for operator awareness. The purpose of ITS SR 3.4.7.3 is to ensure that each required RHR pump is OPERABLE. Verification of proper breaker alignment and power availability ensures that an RHR pump can be placed in operation, if needed, to maintain residual heat removal and reactor coolant circulation. These changes are acceptable because the Surveillance Requirements ensure the availability of the system to remove reactor residual heat and to provide proper boron mixing via the forced circulation of the reactor coolant. This change is designated as more restrictive because new SRs have been added to periodically verify the requirements of the LCO are met.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.1.a.2.B requires two residual heat removal (RHR) trains be OPERABLE with each train consisting of the following: 1) one OPERABLE residual heat removal pump; 2) one OPERABLE residual heat removal heat exchanger; and, 3) an OPERABLE flow path consisting of all valves and piping associated with the above train of components and required to remove decay heat from the core during normal shutdown situations. This flow path shall be capable of taking suction from the appropriate Reactor Coolant System hot leg and returning to the Reactor Coolant System. ITS LCO 3.4.7 requires two RHR loops to be OPERABLE, but does not define the components and the associated flow path that comprise an OPERABLE RHR train. This changes the CTS by moving the description of the RHR trains to the Bases.

The removal of these details which are related to 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. The ITS still retains all necessary requirements in the LCO to ensure OPERABILITY of the RHR loops in MODE 5. Also, 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 Specification 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 Kewaunee Power Station Page 4 of 6 Attachment 1, Volume 9, Rev. 1, Page 131 of 478

Attachment 1, Volume 9, Rev. 1, Page 132 of 478 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) ITS 3.4.7 LCO Note 2 allows one RHR loop to be inoperable for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for Surveillance testing, provided that the other RHR loop is OPERABLE and in operation. The CTS does not contain this allowance; CTS 3.1.a.2.B requires both RHR trains to be OPERABLE at all times when in MODE 5 with the RCS loops filled. This changes the CTS by providing an allowance for one of the RHR loops to be inoperable for a limited period of time to perform required Surveillance testing.

Note 2 allows one RHR loop to be inoperable for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for Surveillance testing, provided that the other RHR loop is OPERABLE and in operation. The purpose of the Note is to permit periodic Surveillance tests to be performed on the inoperable loop during a time when such testing is safe and possible. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allowance is acceptable since the remaining RHR loop is OPERABLE and in operation. Thus the decay heat removal and boron mixing functions are still being met. This change is less restrictive because conditions that allow removal of one RHR loop from service have been added to the ITS that were not in the CTS.

L02 (Category 1 - Relaxation of LCO Requirements) CTS 3.1.a.2.B requires two RHR trains to be OPERABLE whenever the average reactor coolant temperature is < 200ºF and irradiated fuel is in the reactor (i.e., ITS MODE 5). ITS LCO 3.4.7 provides an allowance that a steam generator with secondary side water level

> 5% can replace one of the two RHR trains. Furthermore, due to this allowance, the ITS provides alternate actions (ITS 3.4.7 Required Action A.2 and ACTION B) when only one RHR train is OPERABLE. In addition, ITS SR 3.4.7.2 requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the secondary side water level of the required steam generator is > 5%. This changes the CTS by only requiring one RHR train to be OPERABLE, provided the secondary side water level of a steam generator is > 5%, and provides compensatory actions when this allowance is being used and is not met, as well as a Surveillance Requirement to periodically verify the limit is being met.

The purpose of CTS 3.1.a.2.B is to provide two methods to ensure decay heat can be removed from the reactor coolant. This ensures that if one method fails, a second method is still available. ITS LCO 3.4.7 continues to ensure two methods are required. The first method must be an RHR train. This ensures the primary method can remove the decay heat using forced circulation. The second heat removal method can be either another RHR train, or a steam generator via natural circulation. Even though steam generators cannot produce steam in this MODE, they are capable of being a heat sink due to their large containment volume of secondary water. As long as the secondary side water level is at a lower temperature than the reactor coolant, heat transfer will occur. This is stated in the NRC Information Notice 95-35, and in the ITS 3.4.7 Bases.

Therefore, this change is considered acceptable since a second method is still being required and an RHR loop OPERABLE and in operation is still the primary Kewaunee Power Station Page 5 of 6 Attachment 1, Volume 9, Rev. 1, Page 132 of 478

Attachment 1, Volume 9, Rev. 1, Page 133 of 478 DISCUSSION OF CHANGES ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED method required by LCO 3.4.7. Appropriate Actions have been added (ITS 3.4.7 Required Action A.2 and ACTION B) to immediately initiate action to restore the inoperable RHR loop or to restore the secondary side water level of the steam generator to within limit, when the second heat removal method (either a second RHR loop or a steam generator) is inoperable. Furthermore, a new Surveillance has been added to ensure secondary side water level of the required steam generator is periodically verified to be within the limit (> 5%) when the steam generator is being used as the second heat removal method. This change is designated as less restrictive because the LCO requirements in the ITS are less restrictive than the LCO requirements in the CTS.

Kewaunee Power Station Page 6 of 6 Attachment 1, Volume 9, Rev. 1, Page 133 of 478

Attachment 1, Volume 9, Rev. 1, Page 134 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 134 of 478

Attachment 1, Volume 9, Rev. 1, Page 135 of 478 CTS RCS Loops - MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops - MODE 5, Loops Filled 3.1.a.1.A, LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in 3.1.a.2.B operation, and either:

4

a. One additional RHR loop shall be OPERABLE or

b. The secondary side water level of at least [two] steam generators 5 (SGs) shall be [17]%. one 5

NOTES-------------------------------------------

DOC M01 1. The RHR pump of the loop in operation may be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1; and , "SHUTDOWN 2

MARGIN (SDM)"

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

DOC L01 2. One required RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for S surveillance testing provided that the other RHR loop is OPERABLE 1 and in operation.

3.1.a.1.C 3. No reactor coolant pump shall be started with one or more RCS cold 3

leg temperatures [275°F] [Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR]

unless the secondary side water temperature of each SG is [50]°F above each of the RCS cold leg temperatures. < 100 DOC M01 4. All RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

3.1.a.1.A, APPLICABILITY: MODE 5 with RCS Loops Filled.

3.1.a.2.B WOG STS 3.4.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 135 of 478

Attachment 1, Volume 9, Rev. 1, Page 136 of 478 CTS All changes are 1 RCS Loops - MODE 5, Loops Filled unless otherwise noted 3.4.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.2.B.2 A. One required RHR loop A.1 Initiate action to restore a Immediately inoperable. second RHR loop to OPERABLE status. 2 AND OR One RHR loop OPERABLE. A.2 Initiate action to restore Immediately 4 required SGs secondary side water level to within limit.

DOC L02 4 B. One or more required B.1 Initiate action to restore a Immediately SGs with secondary side second RHR loop to water level not within OPERABLE status.

limit.

OR AND B.2 Initiate action to restore Immediately One RHR loop required SGs secondary 4 OPERABLE. side water level to within limit.

DOCS M02 and C. No required RHR loops C.1 Suspend operations that Immediately M03 OPERABLE. would cause introduction of coolant into the RCS with OR boron concentration less than required to meet SDM Required RHR loop not of LCO 3.1.1.

in operation.

AND C.2 Initiate action to restore one Immediately RHR loop to OPERABLE status and operation.

WOG STS 3.4.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 136 of 478

Attachment 1, Volume 9, Rev. 1, Page 137 of 478 All changes are 1 RCS Loops - MODE 5, Loops Filled unless otherwise noted 3.4.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M04 SR 3.4.7.1 Verify required RHR loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months DOC SR 3.4.7.2 Verify SG secondary side water level is [17]% in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 5 4 L02 required SGs. 5 DOC SR 3.4.7.3 -------------------------------NOTE------------------------------

M04 Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated 7 days power are available to each required RHR pump.

WOG STS 3.4.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 137 of 478

Attachment 1, Volume 9, Rev. 1, Page 138 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED

1. Typographical error corrected.

2. The title of the LCO has been provided since this is the first reference to the LCO.

3. The ISTS LCO 3.4.7 Note 3 provides a bracketed allowance to either specify the minimum RCS cold leg temperature a reactor coolant pump can be started ( 275ºF in the ISTS) or to specify that the RCS cold leg temperature limit is the Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR. Kewaunee Power Station (KPS) has decided not to incorporate a PTLR allowance into the ITS, therefore the Note should reference the actual LTOP arming temperature. However, the current LTOP enabling temperature is 200°F, which is the exact temperature value at which MODE 5 starts. Thus, the words "with one or more RCS cold legs temperatures " and the KPS enabling temperature of 200°F are redundant and have not been included (i.e., the words describing the restriction is needed). In addition, the ISTS LCO 3.4.7 Note 3 provides a bracketed value for the SG secondary side water temperature limit ( 50ºF in the ISTS). The bracketed value in the ISTS has been replaced with the KPS CTS value for this limit (< 100ºF).

4. Changes made to be consistent with changes made to the LCO.

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 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 138 of 478

Attachment 1, Volume 9, Rev. 1, Page 139 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 139 of 478

Attachment 1, Volume 9, Rev. 1, Page 140 of 478 RCS Loops - MODE 5, Loops Filled B 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.7 RCS Loops - MODE 5, Loops Filled BASES BACKGROUND In MODE 5 with the RCS loops filled, the primary function of the reactor of 2 coolant is the removal of decay heat and transfer this heat either to the steam generator (SG) secondary side coolant via natural circulation (Ref. 1) or the component cooling water via the residual heat removal (RHR) heat exchangers. While the principal means for decay heat removal is via the RHR System, the SGs via natural circulation (Ref. 1) are specified as a backup means for redundancy. Even though the SGs cannot produce steam in this MODE, they are capable of being a heat sink due to their large contained volume of secondary water. As long as the SG secondary side water is at a lower temperature than the reactor coolant, heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

In MODE 5 with RCS loops filled, the reactor coolant is circulated by means of two RHR loops connected to the RCS, each loop containing an RHR heat exchanger, an RHR pump, and appropriate flow and temperature instrumentation for control, protection, and indication. One RHR pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratification.

The number of loops in operation can vary to suit the operational needs.

The intent of this LCO is to provide forced flow from at least one RHR loop for decay heat removal and transport. The flow provided by one RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that a second path be available to provide redundancy for heat removal. OPERABLE 4 is The LCO provides for redundant paths of decay heat removal capability.

The first path can be an RHR loop that must be OPERABLE and in one operation. The second path can be another OPERABLE RHR loop or narrow range 3

maintaining two SGs with secondary side water levels [17]% to provide an alternate method for decay heat removal via natural circulation 5 5 (Ref. 1).

would affect APPLICABLE In MODE 5, RCS circulation is considered in the determination of the time 5

SAFETY INSERT 1 available for mitigation of the accidental boron dilution event. The RHR ANALYSES loops provide this circulation. an RCS Loops - MODE 5 (Loops Filled) satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

WOG STS B 3.4.7-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 140 of 478

Attachment 1, Volume 9, Rev. 1, Page 141 of 478 B 3.4.7 o 5

-INSERT 1 The current licensing basis of Kewaunee Power Station (KPS) does not consider boron dilution events during MODE 5 conditions. Therefore, no safety analyses related to the loss of RCS loops are performed for MODE 5.

Insert Page B 3.4.7-1 Attachment 1, Volume 9, Rev. 1, Page 141 of 478

Attachment 1, Volume 9, Rev. 1, Page 142 of 478 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES narrow range LCO The purpose of this LCO is to require that at least one of the RHR loops one be OPERABLE and in operation with an additional RHR loop OPERABLE 5 or two SGs with secondary side water level [17]%. One RHR loop 3 1 provides sufficient forced circulation to perform the safety functions of the provide redundant reactor coolant under these conditions. An additional RHR loop is capability for heat 3 removal required to be OPERABLE to meet single failure considerations.

one However, if the standby RHR loop is not OPERABLE, an acceptable its 3 1 alternate method is two SGs with their secondary side water levels

[17]%. Should the operating RHR loop fail, the SGs could be used to 5 narrow range remove the decay heat via natural circulation.

Note 1 permits all RHR pumps to be removed from operation 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests designed to validate various accident analyses values. One of the tests performed during the startup testing program is the validation of rod drop times during cold conditions, both with and without flow. The no flow test may be performed in MODE 3, 4, or 5 and requires that the pumps be stopped for a short period of time. The Note permits stopping of the pumps in order to perform this test and validate the assumed analysis values. If changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values must be revalidated by conducting the test again. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period is adequate to perform the test, and operating experience has shown that boron stratification is not likely during this short period with no forced flow.

Utilization of Note 1 is permitted provided the following conditions are met, along with any other conditions imposed by initial startup test procedures: "SHUTDOWN 8

MARGIN (SDM),"

a. No operations are permitted that would dilute the RCS boron concentration with coolant with boron concentrations less than required to meet SDM of LCO 3.1.1, therefore maintaining the margin by to criticality. Boron reduction with coolant at boron concentrations less than required to assure SDM is maintained is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation and 6

b. Core outlet temperature is maintained at least 10°F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

Note 2 allows one RHR loop to be inoperable for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , provided that the other RHR loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when such testing is safe and 7 possible. a WOG STS B 3.4.7-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 142 of 478

Attachment 1, Volume 9, Rev. 1, Page 143 of 478 All changes are 1 RCS Loops - MODE 5, Loops Filled B 3.4.7 unless otherwise noted BASES LCO (continued)

< Note 3 requires that the secondary side water temperature of each SG be 100

[50]°F above each of the RCS cold leg temperatures before the start of a reactor coolant pump (RCP) with an RCS cold leg temperature 7

[275°F] [Low Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR]. This restriction is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.

Note 4 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation. This Note provides for the transition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.

RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. A SG can perform as a heat sink via natural circulation when it has an adequate water level and is OPERABLE.

APPLICABILITY In MODE 5 with RCS loops filled, this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or the secondary side water level of at least [two] SGs is 3 1 required to be [17]%. narrow range one 5

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and 2;"

LCO 3.4.5, "RCS Loops - MODE 3;"

LCO 3.4.6, "RCS Loops - MODE 4;" 6 LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled;"

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation -

3 High Water Level" (MODE 6), and ;

LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - 9 4 Low Water Level" (MODE 6).

WOG STS B 3.4.7-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 143 of 478

Attachment 1, Volume 9, Rev. 1, Page 144 of 478 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES ACTIONS A.1, A.2, B.1 and B.2 has narrow range If one RHR loop is OPERABLE and either the required SGs have 3 1 secondary side water levels < [17]%, or one required RHR loop is 5

inoperable, redundancy for heat removal is lost. Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to 3

restore the required SG secondary side water levels. Either Required Action will restore redundant heat removal paths. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

C.1 and C.2 If a required RHR loop is not in operation, except during conditions permitted by Note 1, or if no required loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Times reflect the importance of maintaining operation for heat removal.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance.

SR 3.4.7.2 one is its Verifying that at least two SGs are OPERABLE by ensuring their 3 1 secondary side narrow range water levels are [17]% ensures an 5 is alternate decay heat removal method via natural circulation in the event that the second RHR loop is not OPERABLE. If both RHR loops are OPERABLE, this Surveillance is not needed. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to the loss of SG level.

WOG STS B 3.4.7-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 144 of 478

Attachment 1, Volume 9, Rev. 1, Page 145 of 478 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.7.3 Verification that each required RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required RHR pump. Alternatively, verification that a pump is in operation also 5

verifies proper breaker alignment and power availability. If secondary side water level is [17]% in at least two SGs, this Surveillance is not one 3 1 narrow range needed. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

REFERENCES 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation."

WOG STS B 3.4.7-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 145 of 478

Attachment 1, Volume 9, Rev. 1, Page 146 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.7 BASES, RCS LOOPS - MODE 5, LOOPS FILLED

1. The ISTS Bases 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. Typographical error/omission corrected.

3. Changes made to be consistent with changes made to the ITS.

4. The ISTS Bases Background section states the other intent of the LCO is to require that two paths be available to provide redundancy for decay heat removal.

This statement has been changed to state the other intent of the LCO is to require the two paths to be OPERABLE to provide redundancy for decay heat removal.

This change to the Background statement places it in agreement with the Bases LCO and Specification LCO statements to ensure it is understood that the intent is for the two paths to be OPERABLE.

5. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis or licensing basis description.

6. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

7. Words have been changed to allow for there being other conditions wherein the test might be run safely.

8. The title of the LCO has been provided since this is the first reference to the LCO.

9. The correct ITS number has been provided (changed to be consistent with a number change to the actual specification in Section 3.9).

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 146 of 478

Attachment 1, Volume 9, Rev. 1, Page 147 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 147 of 478

Attachment 1, Volume 9, Rev. 1, Page 148 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.7, RCS LOOPS - MODE 5, LOOPS FILLED There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 148 of 478

Attachment 1, Volume 9, Rev. 1, Page 149 of 478 ATTACHMENT 8 ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED Attachment 1, Volume 9, Rev. 1, Page 149 of 478

, Volume 9, Rev. 1, Page 150 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 150 of 478

Attachment 1, Volume 9, Rev. 1, Page 151 of 478 ITS A01 ITS 3.4.8 3.1 REACTOR COOLANT SYSTEM APPLICABILITY Applies to the OPERATING status of the Reactor Coolant System (RCS).

OBJECTIVE To specify those LIMITING CONDITIONS FOR OPERATION of the Reactor Coolant System which must be met to ensure safe reactor operation.

SPECIFICATIONS

a. Operational Components See ITS

1. Reactor Coolant Pumps 3.4.5 and 3.4.6 LCO 3.4.8 A. At least one reactor coolant pump or one residual heat removal pump shall be in operation when a reduction is made in the boron concentration of the reactor M01 Applicability coolant. Add proposed LCO 3.4.8 Note 1 See ITS B. When the reactor is in the OPERATING mode, except for low power tests, both 3.4.4 reactor coolant pumps shall be in operation.

C. A reactor coolant pump shall not be started with one or more of the RCS cold leg See ITS 3.4.7 temperatures 200°F unless the secondary water temperature of each steam and 3.4.12 generator is < 100°F above each of the RCS cold leg temperatures.

2. Decay Heat Removal Capability A. At least two of the following four heat sinks shall be OPERABLE whenever the average reactor coolant temperature is 350°F but > 200°F.

1. Steam Generator 1A

2. Steam Generator 1B See ITS

3. Residual Heat Removal Train A 3.4.6

4. Residual Heat Removal Train B If less than the above number of required heat sinks are OPERABLE, then corrective action shall be taken immediately to restore the minimum number to the OPERABLE status.

Amendment No. 165 TS 3.1-1 03/11/2003 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 151 of 478

Attachment 1, Volume 9, Rev. 1, Page 152 of 478 ITS A01 ITS 3.4.8 LCO 3.4.8 B. Two residual heat removal trains shall be OPERABLE whenever the average A02 Applicability reactor coolant temperature is 200°F and irradiated fuel is in the reactor, except when in the REFUELING MODE with the minimum water level above the See ITS 3.9.3 top of the vessel flange 23 feet, one train may be inoperable for maintenance.

Add proposed LCO 3.4.8 Note 2

1. Each residual heat removal train shall be comprised of: L01 a) One OPERABLE residual heat removal pump b) One OPERABLE residual heat removal heat exchanger LA01 c) An OPERABLE flow path consisting of all valves and piping associated with the above train of components and required to remove decay heat from the core during normal shutdown situations. This flow path shall be capable of taking suction from the appropriate Reactor Coolant System hot leg and returning to the Reactor Coolant System.

2. If one residual heat removal train is inoperable, then corrective action shall ACTION A be taken immediately to return it to the OPERABLE status.

Add proposed ACTION B for loop not in operation M02 Add proposed ACTION B for two M03 inoperable RHR loops Add proposed SR 3.4.8.1 and M04 SR 3.4.8.2 Amendment No. 165 TS 3.1-2 03/11/2003 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 152 of 478

Attachment 1, Volume 9, Rev. 1, Page 153 of 478 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED 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.a.2.B, which requires two RHR trains to be OPERABLE, is applicable whenever the average reactor coolant temperature is 200°F and irradiated fuel is in the reactor. ITS 3.4.8, which also includes similar RHR train OPERABILITY requirements, is applicable in MODE 5 with the RCS loops not filled. RHR requirements in MODE 5 with the RCS loops filled are provided in LCO 3.4.7 and the MODE 6 RHR requirements are provided in LCO 3.9.5 and LCO 3.9.6. This changes the CTS by splitting these RHR requirements into four separate LCOs.

This change is acceptable since all facets of MODES 5 and 6 operation are covered in the four ITS Specifications. This change is designated as administrative because it does not result in any technical changes.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.a.1.A, which (in MODE 5) requires a residual heat removal (RHR) pump to be in operation, is applicable in MODE 5 only when a reduction is made in the boron concentration of the reactor coolant. ITS 3.4.8, in part, requires an RHR loop to be in operation and is applicable at all times when in MODE 5 with the RCS loops not filled, except as allowed in LCO 3.4.8 Note. Note 1 allows the required RHR pump to not be in operation for 15 minutes when switching from one loop to the other provided a) core outlet temperature is maintained > 10°F below saturation temperature; b) no operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1; and c) no draining operations to further reduce the RCS water volume are permitted. This changes the CTS by specifying that the LCO requirement for an RHR pump to be in operation is applicable in MODE 5 with the RCS loops not filled at all times except for those conditions specified in ITS LCO 3.4.8 Note 1.

In MODE 5, CTS 3.1.a.1.A requires at least one RHR pump to be in operation when a reduction is made in the boron concentration of the reactor coolant. The CTS applicability statement applies during only those times when a reduction in the boron concentration of the reactor coolant is made; it is not for decay heat removal purposes. While a change in boron concentration may occur in essentially any MODE; this discussion is only applicable during those times when the plant is in COLD SHUTDOWN (equivalent to ITS MODE 5) with the RCS loops not filled. Other MODES are discussed in other ITS discussions. The ITS 3.4.8 Applicability is MODE 5 at all times with the RCS loops not filled, except for the conditions in ITS 3.4.8 LCO Note 1, and primarily addresses the decay heat Kewaunee Power Station Page 1 of 5 Attachment 1, Volume 9, Rev. 1, Page 153 of 478

Attachment 1, Volume 9, Rev. 1, Page 154 of 478 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED removal function, but also covers the boron mixing issue. RHR pumps are utilized in MODE 5 to provide for removal of residual heat from the reactor core to the RHR heat exchangers, and to ensure proper boron mixing within the reactor coolant. The MODE 5 decay heat removal requirements are low enough that a single RHR pump is sufficient to remove core decay heat. The purpose of Note 1 is to permit all RHR pumps to be removed from operation for 15 minutes to allow switching from one loop to the other. With the RHR pumps removed from operation, there is no forced flow of reactor coolant. As a result, the reactor coolant is in natural circulation and there is a risk of boron stratification or the formation of a vapor bubble that may cause a natural circulation flow obstruction should the RHR pumps be removed from service for longer than the 15 minute period. This change is acceptable because the short time limitation to have the RHR pumps removed from operation ensures that both boron stratification and inadequate residual heat removal do not occur. This change is designated more restrictive because one RHR pump is required to be in operation in MODE 5 at all times with the RCS loops not filled, except for the condition specified in ITS 3.4.8 LCO Note 1.

M02 CTS 3.1.a.1.A does not contain any ACTIONS to take should there be less than the required number of RHR pumps in operation. As a result, CTS 3.0.c would normally be entered. However, LCO 3.0.c states that it is not applicable in COLD SHUTDOWN or REFUELING. Since the RHR pump only has to be in operation when a reduction in boron concentration is being made, and, for this Specification, the unit is already in MODE 5, the CTS does not provide any compensatory measures. Therefore, 10 CFR 50.36(c)(2)(i) would apply, which states to shutdown the unit. However, no times are provided to complete the shutdown and no further actions (i.e., suspend boron concentration reductions are required. Note that while no ACTIONS are required, KPS in all likelihood would suspend dilution if this occurred. ITS 3.4.8 ACTION B specifies the Required Actions for a required RHR loop not in operation, and requires immediate suspension of operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and initiation of action to restore one RHR loop to operation. This changes the CTS by adding a new ACTION when the required RHR pump is not in operation.

The purpose of the ACTIONS should be to place the unit outside of the Applicability of the Specification. ITS 3.4.8 ACTION B effectively places the unit in an equivalent condition by requiring the plant to immediately suspend operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and to initiate action to restore one RHR loop to operation. The proposed Required Actions reflect the importance of maintaining operation for decay heat removal and boron mixing. This change is designated as more restrictive because a new proposed ACTION has been added.

M03 CTS 3.1.a.2.B does not contain any ACTIONS to take if both required RHR loops are inoperable in MODE 5. As a result, CTS 3.0.c would be entered, which requires action to be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , to be in HOT SHUTDOWN (equivalent to ITS MODE 3) within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and to be in COLD SHUTDOWN Kewaunee Power Station Page 2 of 5 Attachment 1, Volume 9, Rev. 1, Page 154 of 478

Attachment 1, Volume 9, Rev. 1, Page 155 of 478 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED (equivalent to ITS MODE 5) within the subsequent 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . However, since the unit is already in MODE 5, the CTS 3.0.c requirement does not really provide any relevant compensatory measures. ITS 3.4.8 ACTION B provides the Required Actions when no required RHR loops are OPERABLE. The Required Actions are to immediately suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1 and to immediately initiate action to restore one RHR loop to OPERABLE status. This changes the CTS by adding a new ACTION when both required RHR loops are inoperable.

The change is acceptable because the Completion Times are 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 an event occurring during the allowed Completion Times. The immediate suspension of operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the SDM of LCO 3.1.1 and immediately initiating action to restore one loop to OPERABLE status reflects the importance of maintaining operation for decay heat removal and boron mixing capability. Also, coolant added to the RCS without forced circulation could introduce unmixed coolant into the core, which could reduce the required margin to criticality. This change is designated as more restrictive because a new ACTION is being added to the ITS that was not required by the CTS.

M04 CTS 3.1.a.1.A requires one RHR pump to be in operation under certain conditions, but does not provide a Surveillance Requirement to periodically verify the required pump is in operation. ITS SR 3.4.8.1 requires verification that the required RHR loop is in operation every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . CTS 3.1.a.2.B requires two RHR trains to be OPERABLE, but does not provide a Surveillance Requirement to periodically verify the required loops are OPERABLE. ITS SR 3.4.8.2 requires verification that each required RHR pump is OPERABLE every 7 days by verifying correct breaker alignment and indicated power are available to each required pump. A Note further explains that the Surveillance is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation. This changes the CTS by adding new Surveillance Requirements to periodically verify the required pump is in operation and the required pumps are OPERABLE.

The purpose of ITS SR 3.4.8.1 is to ensure that one RHR loop is in operation providing forced flow of the reactor coolant for heat removal. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months frequency for the proposed Surveillance is selected based on operating experience and the need for operator awareness. The purpose of ITS SR 3.4.8.2 is to ensure that each required RHR pump is OPERABLE. Verification of proper breaker alignment and power availability ensures that an RHR pump can be placed in operation, if needed, to maintain residual heat removal and reactor coolant circulation. These changes are acceptable because the Surveillance Requirements ensure the availability of the system to remove reactor residual heat and to provide proper boron mixing via the forced circulation of the reactor coolant. This change is designated as more restrictive because new SRs have been added to periodically verify the requirements of the LCO are met.

Kewaunee Power Station Page 3 of 5 Attachment 1, Volume 9, Rev. 1, Page 155 of 478

Attachment 1, Volume 9, Rev. 1, Page 156 of 478 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.1.a.2.B requires two residual heat removal (RHR) trains be OPERABLE with each train consisting of the following: 1) one OPERABLE residual heat removal pump; 2) one OPERABLE residual heat removal heat exchanger; and, 3) an OPERABLE flow path consisting of all valves and piping associated with the above train of components and required to remove decay heat from the core during normal shutdown situations. This flow path shall be capable of taking suction from the appropriate Reactor Coolant System hot leg and returning to the Reactor Coolant System. ITS LCO 3.4.8 requires two RHR loops to be OPERABLE, but does not define the components and the associated flow path that comprise an OPERABLE RHR train. This changes the CTS by moving the description of the RHR trains to the Bases.

The removal of these details which are related to 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. The ITS still retains all necessary requirements in the LCO to ensure OPERABILITY of the RHR loops in MODE 5. Also, 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 Specification 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 information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) ITS 3.4.8 Note 2 allows one RHR loop to be inoperable for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for Surveillance testing provided that the other RHR loop is OPERABLE and in operation. The CTS does not contain these allowances; CTS 3.1.a.2.B requires both RHR trains to be OPERABLE at all times when in MODE 5 with the RCS loops not filled. This changes the CTS by providing an allowance for one of the RHR loops to be inoperable for a limited period of time to perform required Surveillance testing.

Note 2 allows one RHR loop to be inoperable for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for Surveillance testing, provided that the other RHR loop is OPERABLE and in operation. The purpose of the Note is to permit periodic Surveillance tests to be performed on the inoperable loop during a time when such testing is safe and possible. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allowance is acceptable since the remaining RHR loop is OPERABLE and in operation. Thus the decay heat removal and boron mixing functions are still being met. This change is less restrictive because conditions Kewaunee Power Station Page 4 of 5 Attachment 1, Volume 9, Rev. 1, Page 156 of 478

Attachment 1, Volume 9, Rev. 1, Page 157 of 478 DISCUSSION OF CHANGES ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED that allow removal of one RHR loop have been added to the ITS that were not in the CTS.

Kewaunee Power Station Page 5 of 5 Attachment 1, Volume 9, Rev. 1, Page 157 of 478

Attachment 1, Volume 9, Rev. 1, Page 158 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 158 of 478

Attachment 1, Volume 9, Rev. 1, Page 159 of 478 CTS RCS Loops - MODE 5, Loops Not Filled All changes are 1 3.4.8 unless otherwise noted 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 RCS Loops - MODE 5, Loops Not Filled 3.1.a.1.A, LCO 3.4.8 Two residual heat removal (RHR) loops shall be OPERABLE and one 3.1.a.2.B RHR loop shall be in operation.

NOTES-------------------------------------------

DOC M01 1. All RHR pumps may be removed from operation for 15 minutes when switching from one loop to another provided:

at least 2

[ a. The core outlet temperature is maintained > 10°F below saturation temperature, ]

3

b. No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1; and , "SHUTDOWN 4

MARGIN (SDM)"

c. No draining operations to further reduce the RCS water volume are permitted.

S 5

DOC L01 2. One RHR loop may be inoperable for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other RHR loop is OPERABLE and in operation.

3.1.a.1.A, APPLICABILITY: MODE 5 with RCS loops not filled.

3.1.a.2.B ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.2.B.2 A. One required RHR loop A.1 Initiate action to restore Immediately inoperable. RHR loop to OPERABLE status.

WOG STS 3.4.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 159 of 478

Attachment 1, Volume 9, Rev. 1, Page 160 of 478 CTS RCS Loops - MODE 5, Loops Not Filled 3.4.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME DOCS M02 B. No required RHR loop B.1 Suspend operations that Immediately and M03 OPERABLE. would cause introduction of coolant into the RCS with OR boron concentration less than required to meet SDM Required RHR loop not of LCO 3.1.1.

in operation.

AND B.2 Initiate action to restore one Immediately RHR loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M04 SR 3.4.8.1 Verify required RHR loop is in operation. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months DOC M04 SR 3.4.8.2 -------------------------------NOTE------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

Verify correct breaker alignment and indicated 7 days power are available to each required RHR pump.

WOG STS 3.4.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 160 of 478

Attachment 1, Volume 9, Rev. 1, Page 161 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED

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. ISTS 3.4.8 Note 1.a states that the core outlet temperature is maintained > 10°F below saturation temperature. This same Note appears in ISTS 3.4.6 and ISTS 3.4.7 and states that the core outlet temperature is maintained at least 10°F below saturation temperature. The ">" 10°F term has been revised to read "at least" 10°F to be consistent with the conditions provided in the Notes for ISTS 3.4.6 and ISTS 3.4.7.

3. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

4. The title of the LCO has been provided since this is the first reference to the LCO.

5. Typographical error corrected.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 161 of 478

Attachment 1, Volume 9, Rev. 1, Page 162 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 162 of 478

Attachment 1, Volume 9, Rev. 1, Page 163 of 478 RCS Loops - MODE 5, Loops Not Filled All changes are 1 B 3.4.8 unless otherwise noted B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.8 RCS Loops - MODE 5, Loops Not Filled BASES BACKGROUND In MODE 5 with the RCS loops not filled, the primary function of the reactor coolant is the removal of decay heat generated in the fuel, and the transfer of this heat to the component cooling water via the residual heat removal (RHR) heat exchangers. The steam generators (SGs) are not available as a heat sink when the loops are not filled. The secondary function of the reactor coolant is to act as a carrier for the soluble neutron poison, boric acid.

In MODE 5 with loops not filled, only RHR pumps can be used for coolant circulation. The number of pumps in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at least one RHR pump for decay heat removal and transport and to require that two paths be available to provide redundancy for heat removal.

APPLICABLE In MODE 5, RCS circulation is considered in the determination of the time 2

SAFETY INSERT 1 available for mitigation of the accidental boron dilution event. The RHR ANALYSES loops provide this circulation. The flow provided by one RHR loop is an adequate for heat removal and for boron mixing. would affect RCS loops in MODE 5 (loops not filled) satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO The purpose of this LCO is to require that at least two RHR loops be OPERABLE and one of these loops be in operation. An OPERABLE loop is one that has the capability of transferring heat from the reactor coolant at a controlled rate. Heat cannot be removed via the RHR System unless forced flow is used. A minimum of one running RHR pump meets the LCO requirement for one loop in operation. An additional RHR loop is required to be OPERABLE to meet single failure considerations.

Note 1 permits all RHR pumps to be removed from operation for 15 minutes when switching from one loop to another. The circumstances for stopping both RHR pumps are to be limited to at least situations when the outage time is short [and core outlet temperature is 3

maintained > 10°F below saturation temperature]. The Note prohibits boron dilution with coolant at boron concentrations less than required to assure SDM of LCO 3.1.1 is maintained or draining operations when RHR forced flow is stopped.

"SHUTDOWN MARGIN (SDM)," 5 WOG STS B 3.4.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 163 of 478

Attachment 1, Volume 9, Rev. 1, Page 164 of 478 B 3.4.8 o 2

-INSERT 1 The current licensing basis of Kewaunee Power Station (KPS) does not consider boron dilution events during MODE 5 conditions. Therefore, no safety analyses related to the loss of RCS loops are performed for MODE 5.

Insert Page B 3.4.8-1 Attachment 1, Volume 9, Rev. 1, Page 164 of 478

Attachment 1, Volume 9, Rev. 1, Page 165 of 478 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES LCO (continued)

Note 2 allows one RHR loop to be inoperable for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when these tests are safe and possible. 7 a

An OPERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger.

RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required.

APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal and coolant circulation by the RHR System.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and 2,"

LCO 3.4.5, "RCS Loops - MODE 3,"

LCO 3.4.6, "RCS Loops - MODE 4," 4 LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled,"

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation -

3 High Water Level" (MODE 6), and ;

LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - 6 4 Low Water Level" (MODE 6).

ACTIONS A.1 If one required RHR loop is inoperable, redundancy for RHR is lost.

Action must be initiated to restore a second loop to OPERABLE status.

The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

B.1 and B.2 If no required loop is OPERABLE or the required loop is not in operation, except during conditions permitted by Note 1, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action must be initiated immediately to restore an RHR loop to OPERABLE status and operation. The required margin to WOG STS B 3.4.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 165 of 478

Attachment 1, Volume 9, Rev. 1, Page 166 of 478 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES ACTIONS (continued) criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must continue until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance.

SR 3.4.8.2 Verification that each required pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to each required pump. Alternatively, verification that a pump is in operation also verifies proper breaker alignment and power availability. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

This SR is modified by a Note that states the SR is not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required pump is not in operation.

REFERENCES None.

WOG STS B 3.4.8-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 166 of 478

Attachment 1, Volume 9, Rev. 1, Page 167 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.8 BASES, RCS LOOPS - MODE 5, LOOPS NOT FILLED

1. The ISTS Bases 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. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis or licensing basis description.

3. Changes are made to reflect those changes made to the ISTS. In this case, the

">" 10°F term has been revised to read "at least" 10°F to be consistent with the conditions provided in the Notes to ISTS 3.4.6 and ISTS 3.4.7.

4. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

5. The title of the LCO has been provided since this is the first reference to the LCO.

6. The correct ITS number has been provided (changed to be consistent with a number change to the actual specification in Section 3.9).

7. Words have been changed to allow for there being other conditions wherein the test might be run safely.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 167 of 478

Attachment 1, Volume 9, Rev. 1, Page 168 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 168 of 478

Attachment 1, Volume 9, Rev. 1, Page 169 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.8, RCS LOOPS - MODE 5, LOOPS NOT FILLED There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 169 of 478

, Volume 9, Rev. 1, Page 170 of 478 ATTACHMENT 9 ITS 3.4.9, PRESSURIZER , Volume 9, Rev. 1, Page 170 of 478

, Volume 9, Rev. 1, Page 171 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 171 of 478

Attachment 1, Volume 9, Rev. 1, Page 172 of 478 ITS A01 ITS 3.4.9

4. With one block valve inoperable, within one hour restore the block valve to OPERABLE status or place its associated PORV in manual control. Restore the block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months See ITS 3.4.11

5. With both block valves inoperable, within one hour restore the block valves to OPERABLE status or place their associated PORVs in manual control.

Restore at least one block valve to OPERABLE status within the next hour; otherwise, action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

6. Pressurizer Heaters LCO 3.4.9.b A. At least one group of pressurizer heaters shall have an emergency power supply available when the average RCS temperature is > 350°F.

Applicability Add proposed ACTIONS B and C L01

7. Reactor Coolant Vent System A. A reactor coolant vent path from both the reactor vessel head and pressurizer steam space shall be OPERABLE and closed prior to the average RCS temperature being heated > 200°F except as specified in TS 3.1.a.7.B and TS 3.1.a.7.C below.

B. When the average RCS temperature is > 200°F, any one of the following conditions of inoperability may exist:

1. Both of the parallel vent valves in the reactor vessel vent path are inoperable.

2. Both of the parallel vent valves in the pressurizer vent path are inoperable.

See CTS If OPERABILITY is not restored within 30 days, then within one hour action shall 3.1.a.7 be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve COLD SHUTDOWN within an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. If no Reactor Coolant System vent paths are OPERABLE, then restore at least one vent path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If OPERABILITY is not restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , then within one hour action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve COLD SHUTDOWN within an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Amendment No. 165 TS 3.1-5 03/11/2003 Page 1 of 4 Attachment 1, Volume 9, Rev. 1, Page 172 of 478

Attachment 1, Volume 9, Rev. 1, Page 173 of 478 A01 ITS ITS 3.4.9

f. Minimum Conditions for Criticality See ITS

1. The reactor shall not be brought to a critical condition until the pressure-temperature 3.4.3 state is to the right of the criticality limit line shown in Figure TS 3.1-1.

Add proposed MODE 3 Applicability M01 Applicability 2. The reactor shall be maintained subcritical by at least 1% k/k until normal water LC0 3.4.9 level is established in the pressurizer. M02 Add proposed ACTION A

3. When the reactor is critical the moderator temperature coefficient shall be as M01 specified in the COLR, except during LOW POWER PHYSICS TESTING. The maximum upper moderator temperature coefficient limit shall be 5 pcm/°F for M03 power levels 60% RATED POWER and 0 pcm/°F for power levels > 60% RATED POWER. See ITS 3.1.3 and 3.1.8

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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , develop and maintain administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the See ITS limits specified in TS 3.1.f.3. These withdrawal limits shall be in addition to the 3.1.3 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Amendment No. 165 TS 3.1-10 03/11/2003 Page 2 of 4 Attachment 1, Volume 9, Rev. 1, Page 173 of 478

Attachment 1, Volume 9, Rev. 1, Page 174 of 478 ITS A01 ITS 3.4.9 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 3.4.9.2, b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and SR 3.4.9.3 TS 4.1-3.

c. Deleted

d. Deleted

e. Deleted Amendment No. 119 TS 4.1-1 04/18/95 Page 3 of 4 Attachment 1, Volume 9, Rev. 1, Page 174 of 478

Attachment 1, Volume 9, Rev. 1, Page 175 of 478 ITS A01 ITS 3.4.9 TABLE TS 4.1-3 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS L02 (1)

EQUIPMENT TESTS TEST FREQUENCY

1. Control Rods Rod drop times of all full Each REFUELING outage length rods Quarterly when at or above HOT See ITS Partial movement of all STANDBY 3.1.4 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 3.6.3

4. Containment Isolation Trip OPERABILITY Each REFUELING outage

5. Refueling System Interlocks OPERABILITY Prior to fuel movement each See CTS REFUELING outage 3.8.a.11

6. Deleted

7. Deleted See ITS (4) 3.4.15

8. RCS Leak Detection OPERABILITY Weekly See ITS

9. Diesel Fuel Supply Fuel Inventory(5) Weekly 3.8.1 and

10. Deleted 3.8.3

11. Fuel Assemblies Visual Inspection Each REFUELING outage See ITS 4.0

12. Guard Pipes Visual Inspection Each REFUELING outage See ITS

13. Pressurizer PORVs OPERABILITY Each REFUELING cycle 3.6.1

14. Pressurizer PORV Block Valves OPERABILITY Quarterly(6) See ITS SR 3.4.9.2, 3.4.11

15. Pressurizer Heaters OPERABILITY(7) Each REFUELING cycle SR 3.4.9.3

16. Containment Purge and Vent OPERABILITY(8) Each REFUELING cycle See ITS Isolation Valves 3.6.3 Add proposed SR 3.4.9.1 M02 (1)

Following maintenance on equipment that could affect the operation of the equipment, tests L02 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 See ITS 3.3.1 and undervoltage trip attachments. The test shall also verify the undervoltage trip attachment on the reactor trip bypass breakers.

(4) See ITS When reactor is at power or in HOT SHUTDOWN condition. 3.4.15 (5)

Inventory of fuel required in all plant modes. See ITS 3.8.1 (6) and 3.8.3 Not required when valve is administratively closed. See ITS SR 3.4.9.2, (7) 3.4.11 SR 3.4.9.3 Test will verify OPERABILITY of heaters and availability of an emergency power supply.

(8)

This test shall demonstrate that the valve(s) close in 5 seconds. See ITS 3.6.3 Amendment No. 125 Page 1 of 1 08/07/96 Page 4 of 4 Attachment 1, Volume 9, Rev. 1, Page 175 of 478

Attachment 1, Volume 9, Rev. 1, Page 176 of 478 DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER 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 requires that the reactor remain subcritical by at least 1% k/k (ITS MODES 1 and 2) until normal water level is established in the pressurizer.

ITS 3.4.9 requires, in part, that the pressurizer water level is 90% not only in MODES 1 and 2, but also in MODE 3. Due to this new Applicability, when the pressurizer water level is not within the limit, ITS 3.4.9 Required Action A.4 will also require the unit to be taken to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , which is outside the new MODE of Applicability. This changes the CTS by adding a new MODE of Applicability (MODE 3) and a commensurate Required Action. The change in the water level requirement is discussed in DOC M02.

This change is acceptable because the analyses performed from a critical reactor condition assumes the existence of a steam bubble and saturated conditions in the pressurizer. Furthermore, the purpose of the MODE 3 Applicability is to prevent water solid reactor coolant system operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. This change is designated as more restrictive because the requirement that the water level of the pressurizer is 90% is now required in MODE 3.

M02 CTS 3.1.f requires the pressurizer water level to be within the "normal water level." In addition, the CTS does not provide any Surveillance Requirement to periodically verify the pressurizer water level. The CTS Bases states that the requirement that the pressurizer is partly voided ensures that the RCS will not be solid when criticality is achieved. ITS LCO 3.4.9.a requires the pressurizer level to be 90%. Furthermore, ITS SR 3.4.9.1 requires verification that the pressurizer water level is 90% every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by specifically stating the value for the normal water level ( 90%) and providing a periodic verification that the level is within the limit.

The purpose of the water level limit is to ensure the accident analysis is met and to prevent a water solid condition from existing during heatup and cooldown conditions. The proposed value is the value at which the pressurizer level provides a Reactor Protection System trip. The trip ensures a water solid condition does not occur with the reactor critical. Furthermore, the proposed Surveillance Requirement will ensure the water level is periodically monitored to ensure the limit is not exceeded. This change is designated as more restrictive Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 176 of 478

Attachment 1, Volume 9, Rev. 1, Page 177 of 478 DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER since the specific water level limit is being added to the CTS and a specific Surveillance Requirement to verify the limit is met is being added to the CTS.

M03 CTS 3.1.f does not provide any ACTIONS to be taken when the required pressurizer water level is not met when the reactor is not subcritical by at least 1% k/k. As a result, LCO 3.0.c would be entered, which requires action to be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to be in HOT SHUTDOWN (equivalent to ITS MODE

3) with the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Under similar conditions, ITS 3.4.9 ACTION A requires the unit to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (Required Action A.1), to fully insert all rods in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (Required Action A.2), and to place the Rod Control System in a condition incapable of rod withdrawal in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (Required Action A.3). This changes the CTS by providing specific Required Actions to take when the pressurizer water level is not within limits.

The purpose of CTS LCO 3.0.c is to place the unit outside of the Applicability of the Specification. ITS 3.4.9 ACTION A continues to require the unit to be placed outside the MODE of Applicability, but adds two additional requirements intended to minimize the core reactivity and any pressure transient which may result from an inadvertent withdrawal of control rods. This change is acceptable because it provides additional assurance that certain events will not occur during the transition out of the MODE of Applicability of the Specification. This change is designated as more restrictive because additional Required Actions are now required.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.1.a.6 does not provide any ACTIONS to take when the required group of pressurizer heaters are inoperable when the average RCS temperature is > 350°F (ITS MODES 1, 2, and 3). As a result, LCO 3.0.c would be entered, which requires action to be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , to be in HOT SHUTDOWN (equivalent to ITS MODE 3) with the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and to be in COLD SHUTDOWN (equivalent to ITS MODE 5) within the subsequent 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (although only CTS INTERMEDIATE SHUTDOWN is required to be entered since CTS 3.1.a.6 is required only when above 350ºF).

ITS 3.4.9 ACTION B allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the inoperable pressurizer heater group to OPERABLE status, and if not restored, ITS 3.4.9 ACTION C requires the unit to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 177 of 478

Attachment 1, Volume 9, Rev. 1, Page 178 of 478 DISCUSSION OF CHANGES ITS 3.4.9, PRESSURIZER changes the CTS by providing specific ACTIONS when the pressurizer heaters are inoperable.

The purpose of CTS 3.1.a.6 is to maintain pressurizer pressure. This is accomplished by having one pressurizer heater bank operable and capable of being powered from an emergency power supply. This change is acceptable because the Required Action is 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 time to repair inoperable features. 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. Under the ITS, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months are allowed to restore the inoperable pressurizer heater group before requiring a unit shutdown. This change is designated as less restrictive because less stringent Required Actions are being applied.

L02 (Category 5 - Deletion of Surveillance Requirement) Note 1 to CTS Table 4.1-3 requires, in part, that the pressurizer heaters be tested to verify OPERABILITY following maintenance on equipment that could affect the operation. ITS 3.4.9 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 a 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 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 178 of 478

Attachment 1, Volume 9, Rev. 1, Page 179 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 179 of 478

Attachment 1, Volume 9, Rev. 1, Page 180 of 478 CTS Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

90 3.1.f a. Pressurizer water level [92]% and 1 4 One 3.1.a.6 b. [Two groups of] pressurizer heaters OPERABLE with the capacity 158.4 1

[of each group] [125] kW [and capable of being powered from an the emergency power supply].

3.1.a.6, APPLICABILITY: MODES 1, 2, and 3.

3.1.f ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOCs M01 A. Pressurizer water level A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and M03 not within limit.

AND A.2 Fully insert all rods. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND A.3 Place Rod Control System 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months in a condition incapable of rod withdrawal.

AND A.4 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months DOC L01 2 B. One [required] group of B.1 Restore [required] group of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressurizer heaters pressurizer heaters to inoperable. OPERABLE status.

DOC L01 C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B not AND met.

WOG STS 3.4.9-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 180 of 478

Attachment 1, Volume 9, Rev. 1, Page 181 of 478 CTS Pressurizer 3.4.9 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 90 DOC M02 SR 3.4.9.1 Verify pressurizer water level is [92]%. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 1

REVIEWER'S NOTE---------------------------------

The frequency for performing Pressurizer heater capacity testing shall be either 18 months or 92 days, depending on whether or not the plant has dedicated safety-related heaters. For dedicated safety-related 3 heaters, which do not normally operate, 92 days is applied. For non-dedicated safety-related heaters, which normally operate, 18 months is applied.

Table 4.1-3 3 Equipment SR 3.4.9.2 Verify capacity of each required group of pressurizer [18] months Test 15, heaters is [125] kW. 1 including 158.4 Note 7 Table 4.1-3 Equipment SR 3.4.9.3 [ Verify required pressurizer heaters are capable of [18] months ] 1 Test 15, including being powered from an emergency power supply.

Note 7 WOG STS 3.4.9-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 181 of 478

Attachment 1, Volume 9, Rev. 1, Page 182 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.9, PRESSURIZER

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. ISTS 3.4.9 ACTION B contains bracketed information and/or values that are generic to all Westinghouse vintage plants. Since Kewaunee Power Station (KPS) has two groups of pressurizer heaters that are powered from an emergency power supply and only one group is required to be OPERABLE, the word "required" has been maintained in ITS 3.4.9 ACTION B. Furthermore, since only one group is required, the word "one" is not necessary and has been deleted.

3. ISTS SR 3.4.9.2 contains a Reviewers Note that is used by the NRC reviewer to determine the Frequency for ISTS SR 3.4.9.2. As stated in the Reviewers Note, the frequency for performing Pressurizer heater capacity testing is based on whether or not the plant has dedicated safety-related heaters. If the plant has safety-related heaters, which do not normally operate, 92 days is applied. For non-dedicated safety-related heaters, which normally operate, 18 months is applied. At KPS, the Pressurizer heaters are non-dedicated heaters which normally operate. Therefore, the Frequency for performing SR 3.4.9.2, at KPS, is 18 months, which is also consistent with the current Technical Specification requirement.

4. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 182 of 478

Attachment 1, Volume 9, Rev. 1, Page 183 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 183 of 478

Attachment 1, Volume 9, Rev. 1, Page 184 of 478 Pressurizer B 3.4.9 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.9 Pressurizer BASES BACKGROUND The pressurizer provides a point in the RCS where liquid and vapor are maintained in equilibrium under saturated conditions for pressure control purposes to prevent bulk boiling in the remainder of the RCS. Key functions include maintaining required primary system pressure during steady state operation, and limiting the pressure changes caused by reactor coolant thermal expansion and contraction during normal load transients.

The pressure control components addressed by this LCO include the pressurizer water level, the required heaters, and their controls and emergency power supplies. Pressurizer safety valves and pressurizer power operated relief valves are addressed by LCO 3.4.10, "Pressurizer Safety Valves," and LCO 3.4.11, "Pressurizer Power Operated Relief Valves (PORVs)," respectively.

The intent of the LCO is to ensure that a steam bubble exists in the pressurizer prior to power operation to minimize the consequences of potential overpressure transients. The presence of a steam bubble is consistent with analytical assumptions. Relatively small amounts of noncondensible gases can inhibit the condensation heat transfer between the pressurizer spray and the steam, and diminish the spray effectiveness for pressure control.

Electrical immersion heaters, located in the lower section of the pressurizer vessel, keep the water in the pressurizer at saturation temperature and maintain a constant operating pressure. A minimum required available capacity of pressurizer heaters ensures that the RCS pressure can be maintained. The capability to maintain and control system pressure is important for maintaining subcooled conditions in the RCS and ensuring the capability to remove core decay heat by either forced or natural circulation of reactor coolant. Unless adequate heater capacity is available, the hot, high pressure condition cannot be maintained indefinitely and still provide the required subcooling margin in the primary system. Inability to control the system pressure and maintain subcooling under conditions of natural circulation flow in the primary system could lead to a loss of single phase natural circulation and decreased capability to remove core decay heat.

WOG STS B 3.4.9-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 184 of 478

Attachment 1, Volume 9, Rev. 1, Page 185 of 478 Pressurizer B 3.4.9 BASES APPLICABLE In MODES 1, 2, and 3, the LCO requirement for a steam bubble is SAFETY reflected implicitly in the accident analyses. Safety analyses performed 1 ANALYSES for lower MODES are not limiting. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer. In making this assumption, the analyses neglect the small fraction of noncondensible gases normally present.

U Safety analyses presented in the FSAR (Ref. 1) do not take credit for 1 pressurizer heater operation; however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure.

The maximum pressurizer water level limit, which ensures that a steam bubble exists in the pressurizer, satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737 (Ref. 2), is the reason for providing an LCO.

LCO -----------------------------------REVIEWERS NOTE-----------------------------------

Plants licensed prior to the issuance of NUREG-0737 may not have a 2 requirement on the number of pressurizer groups.

level The LCO requirement for the pressurizer to be OPERABLE with a water 90 3 volume [1240] cubic feet, which is equivalent to [92]%, ensures that a 4 steam bubble exists. Limiting the LCO maximum operating water level which is equivalent to preserves the steam space for pressure control. The LCO has been approximately established to ensure the capability to establish and maintain pressure 900 cubic feet control for steady state operation and to minimize the consequences of potential overpressure transients. Requiring the presence of a steam bubble is also consistent with analytical assumptions.

one The LCO requires [two groups of] OPERABLE pressurizer heaters, [each]

3 158.4 with a capacity [125] kW, [capable of being powered from either the the emergency offsite power source or the emergency power supply]. The minimum buses heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions, a wide margin to subcooling can be obtained in the loops. 158.4 The exact design value of [125 kW is derived from the use of seven 3 heaters rated at 17.9 kW each]. The amount needed to maintain pressure is dependent on the heat losses.

data obtained during in-plant testing. The value is met by 13 OPERABLE heaters, since each heater has a rated capacity of 12.82 kW.

WOG STS B 3.4.9-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 185 of 478

Attachment 1, Volume 9, Rev. 1, Page 186 of 478 Pressurizer B 3.4.9 BASES APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature, resulting in the greatest effect on pressurizer level and RCS pressure control. Thus, applicability has been designated for MODES 1 and 2. The applicability is also provided for MODE 3. The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup.

In MODES 1, 2, and 3, there is need to maintain the availability of pressurizer heaters, capable of being powered from an emergency power supply. In the event of a loss of offsite power, the initial conditions of these MODES give the greatest demand for maintaining the RCS in a hot pressurized condition with loop subcooling for an extended period. For MODE 4, 5, or 6, it is not necessary to control pressure (by heaters) to ensure loop subcooling for heat transfer when the Residual Heat Removal (RHR) System is in service, and therefore, the LCO is not applicable.

ACTIONS A.1, A.2, A.3, and A.4 Pressurizer water level control malfunctions or other plant evolutions may result in a pressurizer water level above the nominal upper limit, even with the plant at steady state conditions. Normally the plant will trip in this event since the upper limit of this LCO is the same as the Pressurizer Water Level - High Trip.

If the pressurizer water level is not within the limit, action must be taken to bring the plant to a MODE in which the LCO does not apply. To achieve this status, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months the unit must be brought to MODE 3 with all rods fully inserted and incapable of withdrawal. Additionally, the unit must be brought to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This takes the unit out of the applicable MODES.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

B.1 the If one [required] group of pressurizer heaters is inoperable, restoration is 4 required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable considering the anticipation that a demand caused by loss of offsite power would be unlikely in this period. Pressure control may be maintained during this time using normal station powered heaters.

WOG STS B 3.4.9-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 186 of 478

Attachment 1, Volume 9, Rev. 1, Page 187 of 478 Pressurizer B 3.4.9 BASES ACTIONS (continued)

C.1 and C.2 4 the required If one group of pressurizer heaters are inoperable and cannot be restored in the allowed Completion Time of Required Action B.1, 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 MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months corresponds to verifying the parameter each shift. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess level for any deviation and verify that operation is within safety analyses assumption of ensuring that a steam bubble exists in the pressurizer. Alarms are also available for early detection of abnormal level indications.

SR 3.4.9.2

REVIEWERS NOTE-----------------------------------

The frequency for performing Pressurizer heater capacity testing shall be either 18 months or 92 days, depending on whether or not the plant has dedicated safety-related heaters. For dedicated safety-related heaters, 2 which do not normally operate, 92 days is applied. For non-dedicated safety-related heaters, which normally operate, 18 months is applied.

The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated pressurizer heaters are verified to be at their design rating. This may be done by testing the power supply output and by performing an electrical check on heater element continuity and resistance. The Frequency of [18] months 3 is considered adequate to detect heater degradation and has been shown by operating experience to be acceptable.

WOG STS B 3.4.9-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 187 of 478

Attachment 1, Volume 9, Rev. 1, Page 188 of 478 Pressurizer B 3.4.9 BASES SURVEILLANCE REQUIREMENTS (continued)

[ SR 3.4.9.3 3 This SR is not applicable if the heaters are permanently powered by 3 Class 1E power supplies.

verifies required are capable of 5

This Surveillance demonstrates that the heaters can be manually beingbypowered an transferred from the normal to the emergency power supply and energized. The Frequency of 18 months is based on a typical fuel cycle and is consistent with similar verifications of emergency power supplies. ] 3 REFERENCES 1. FSAR, Section [ ]. 14.1 1 3 U

2. NUREG-0737, November 1980.

The Surveillance can be met by verifying the required heaters are energized by its emergency power supply (if the heaters are normally 5 aligned to and energized by an emergency power supply) or by verifying that the required heater power supply can be transferred to an emergency power supply (if the heaters are not aligned to and energized by an emergency power supply).

WOG STS B 3.4.9-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 188 of 478

Attachment 1, Volume 9, Rev. 1, Page 189 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.9 BASES, PRESSURIZER

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 Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This is not meant to be retained in the final version of the plant specific submittal.

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 acceptable since the generic specific information/value is revised to reflect the current plant design.

4. Change made to be consistent with the Specification.

5. The KPS design for the pressurizer heaters includes one train that is only powered from an emergency power supply and one train that can be powered from either an emergency power supply or non-emergency power supply. Therefore, the Bases for SR 3.4.9.3 has been reworded to match the KPS design.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 189 of 478

Attachment 1, Volume 9, Rev. 1, Page 190 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 190 of 478

Attachment 1, Volume 9, Rev. 1, Page 191 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.9, PRESSURIZER There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 191 of 478

, Volume 9, Rev. 1, Page 192 of 478 ATTACHMENT 10 ITS 3.4.10, PRESSURIZER SAFETY VALVES , Volume 9, Rev. 1, Page 192 of 478

, Volume 9, Rev. 1, Page 193 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 193 of 478

Attachment 1, Volume 9, Rev. 1, Page 194 of 478 ITS A01 ITS 3.4.10 3.1.a.3 Pressurizer Safety Valves Add proposed lift settings M01 LCO 3.4.10 LCO 3.1.a.3 Two pressurizer safety valves shall be OPERABLE Applicability APPLICABILITY: Reactor Coolant System Temperature Greater than the Low Temperature A02 Overpressure Protection (LTOP) Enabling Temperature (200°F)

Add proposed Applicability Note M01 ACTIONS inoperable LA01

- NOTE -

LCO During a hydro test of the RCS, the pressurizer safety valves may be blanked provided the 3.4.10 power-operated relief valves and the safety valve on the discharge pump are set for the test Note pressure plus 35 psi to protect the system.

LA01 CONDITION REQUIRED ACTION COMPLETION TIME ACTION A A. One pressurizer safety A.1 Restore to OPERABLE 15 Minutes valve inoperable status OR 6 M02 Required Action B.1 A.2 Be in HOT SHUTDOWN 12 Hours Add proposed Required Action B.2 ACTION B B. Both pressurizer safety B.1 Restore one pressurizer 15 Minutes valves are inoperable safety valve to an OPERABLE status Add proposed Required Action B.1 M03 OR 36 B.2 Be in a condition with the 48 Hours LTOP system Be in MODE 5 A02 OPERABLE or reactor vessel head removed Add proposed SR 3.4.10.1 M01 Amendment No. 164 TS 3.1-3 11/07/2002 Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 194 of 478

Attachment 1, Volume 9, Rev. 1, Page 195 of 478 DISCUSSION OF CHANGES ITS 3.4.10, PRESSURIZER SAFETY VALVES 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 The Applicability for CTS 3.1.a.3 is when the Reactor Coolant System temperature is greater than the LTOP enabling temperature (200°F).

Furthermore, when LCO 3.1.a.3 is not being met, CTS 3.1.a.3 Required Action B.2 requires the unit to be in a condition with the LTOP system OPERABLE or reactor vessel head removed. The ITS 3.4.10 Applicability is MODES 1, 2, 3, and 4, and under similar conditions, ITS 3.4.10 Required Action B.2 requires the unit to be in MODE 5. This changes the CTS by clearly stating the MODES in which the pressurizer safety valves are required to be OPERABLE and the MODE the unit has to be placed in when a unit shutdown is required.

ITS MODES 1, 2, 3, and 4 are the equivalent MODES covering the condition of greater than the LTOP enabling temperature of 200°F. This is shown in ITS Table 1.1-1, which defines that MODE 4 has a minimum average reactor coolant temperature of > 200°F. Thus, changing the Applicability to be MODES 1, 2, 3, and 4 is the same as the current requirement. In addition, since the 200°F requirement is part of the ITS Table 1.1-1 requirements for MODE 4, there is no reason to include the value in the ITS 3.4.10 Applicability. ITS Table 1.1-1 also defines MODE 5 as having a maximum reactor coolant temperature of 200°F.

Furthermore, the Applicability of ITS 3.4.12, "LTOP," is also MODE 5. Therefore, requiring the unit to be in MODE 5 is the same as when the LTOP system is required to be OPERABLE. The current requirement to remove the reactor head is equivalent to the condition the unit will be in when in ITS MODE 6. Since the ITS 3.4.10 Required Action B.2 states to be in MODE 5, per the ITS convention, this does not preclude the unit from going all the way to MODE 6. Therefore, the change in the Required Action is also acceptable. These changes are designated as administrative changes because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.a.3 requires two pressurizer safety valves to be OPERABLE, but does not specify their lift settings. ITS 3.4.10 requires two pressurizer safety valves to be OPERABLE with lift settings 2410.45 psig and 2559.55 psig.

Commensurate with the addition of the lift settings, a Note has been added to the Applicability. The Note states that the lift settings are not required to be within the LCO limits during MODE 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions. Additionally, the Note states that this exception is allowed for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months following entry into MODE 3 provided a Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 195 of 478

Attachment 1, Volume 9, Rev. 1, Page 196 of 478 DISCUSSION OF CHANGES ITS 3.4.10, PRESSURIZER SAFETY VALVES preliminary cold setting was made prior to heatup. Furthermore, a Surveillance Requirement (SR 3.4.10.1) has been added to verify that each pressurizer safety valve is OPERABLE in accordance with the Inservice Testing Program and that following testing the lift settings are within +/- 1%. This changes the CTS by requiring the pressurizer safety valves to be OPERABLE within specific lift settings, except as allowed in the Applicability Note, and by adding a Surveillance Requirement to ensure the lift settings requirements are met.

This change is acceptable because the addition of the pressurizer safety valves' lift settings ensures that the pressurizer safety valves will be capable of protecting the reactor coolant pressure boundary (RCPB) Safety Limit.

Furthermore, the addition of the Surveillance Requirement provides assurance that the pressurizer safety valves will be capable of limiting the RCS pressure to 110% of design pressure. Additionally, the applicability Note permits testing and examination of the pressurizer safety valves to be performed at high pressure and temperatures near their normal operating range. This change is designated as more restrictive because the pressurizer safety valves are required to be OPERABLE within specific lift settings.

M02 CTS 3.1.a.3 ACTION A states that with one pressurizer safety valve inoperable to restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN (equivalent to ITS MODE 3) within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ITS 3.4.10 ACTION B, which provides the ACTION when one pressurizer safety valve is inoperable and has not been restored to OPERABLE status within 15 minutes, requires the unit to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 with any RCS cold leg temperatures the LTOP arming temperature specified in the PTLR within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This changes CTS by requiring the unit to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months instead of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and by further requiring the unit to be in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The purpose of the shutdown requirement is to reduce the vulnerability of the plant during the time one pressurizer safety valve is inoperable. However, it does not currently require the unit to exit the Applicability of the LCO. This change is acceptable because the requirement to place the unit in MODE 3 ensures an intermediate shutdown condition is reached in a shorter period of time. Furthermore, the requirement to place the unit in MODE 5 is acceptable because it places the unit in a MODE outside the Applicability of the LCO. The Completion Times are based on operating experience and the need to reach the required conditions from full power in an orderly manner and without challenging unit systems. This change is designated as more restrictive because it imposes a more restrictive time requirement for when the unit must be in MODE 3 and adds a new requirement for the unit to be in MODE 5.

M03 CTS 3.1.a.3 ACTION B states that with two pressurizer safety valve inoperable to restore one inoperable valve to OPERABLE status within 15 minutes or be in a condition with the LTOP system OPERABLE or reactor head removed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . ITS 3.4.10 ACTION B, which provides the ACTION when two pressurizer safety valves are inoperable, requires the unit to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This changes CTS by not providing a short restoration time for one of the pressurizer safety valves since it is redundant to the time provided in ITS 3.4.10 ACTION A and by requiring the unit to be in Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 196 of 478

Attachment 1, Volume 9, Rev. 1, Page 197 of 478 DISCUSSION OF CHANGES ITS 3.4.10, PRESSURIZER SAFETY VALVES MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The change from requiring the unit to be in a condition with the LTOP system OPERABLE or reactor head removal is discussed in DOC A02.

The purpose of requiring a shutdown when both pressurizer safety valves are inoperable is to exit the Applicability of the LCO since the plant is not meeting the overpressure protection analysis assumptions. This change is acceptable because the requirement to place the unit in MODE 3 ensures an intermediate shutdown condition is reached in a shorter period of time. Furthermore, the requirement to place the unit in MODE 5 is acceptable because it places the unit in a MODE outside the Applicability of the LCO. The Completion Times are based on operating experience and the need to reach the required conditions from full power in an orderly manner and without challenging unit systems. This change has been designated as more restrictive because it deletes a 15 minute restoration time for one of the inoperable valves, adds a requirement to be in MODE 3, and reduces the Completion Time to be in MODE 5 (i.e., in a condition the LTOP is OPERABLE) from 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 2 - Removing Descriptions of System Operation) CTS 3.1.a.3 ACTIONS Note states, in part, that during a hydro test of the RCS, the pressurizer safety valves may be blanked provided that the relief valves and safety valves of the discharge pump are set to protect the system. The Note for ITS LCO 3.4.10 does not include the statements may be "blanked" or "to protect the system."

The LCO 3.4.10 Note uses the term "inoperable". This changes the CTS by moving the statements may be "blanked" and "to protect the system" to the Bases.

The removal of these details which are related to system operation 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. The ITS retains all necessary requirements in LCO to ensure OPERABILITY of the pressurizer safety valves. Also, 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 Specification 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 less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 197 of 478

Attachment 1, Volume 9, Rev. 1, Page 198 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 198 of 478

Attachment 1, Volume 9, Rev. 1, Page 199 of 478 CTS Pressurizer Safety Valves 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves Two LCO 3.1.a.3, LCO 3.4.10 [Three] pressurizer safety valves shall be OPERABLE with lift settings DOC M01

[2460] psig and [2510] psig. 1 2410.45 2559.55 INSERT 1 2 and 4 3.1.a.3 APPLICABILITY: MODES 1, 2, and 3, MODE 4 with all RCS cold leg temperatures > [275°F] [Low Temperature 3 Overpressure Protection (LTOP) arming temperature specified in the PTLR].

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

DOC M01 The lift settings are not required to be within the LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves 36 4 under ambient (hot) conditions. This exception is allowed for [54] hours following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ACTION A A. One pressurizer safety A.1 Restore valve to 15 minutes valve inoperable. OPERABLE status.

ACTION A. B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ACTION B associated Completion Time not met. AND 5 36 5

OR B.2 Be in MODE 4 with any [24] hours RCS cold leg temperatures Two or more pressurizer [275°F] [LTOP arming 1 3 safety valves inoperable. temperature specified in the PTLR].

WOG STS 3.4.10-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 199 of 478

Attachment 1, Volume 9, Rev. 1, Page 200 of 478 CTS 3.4.10 2

INSERT 1

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

ACTIONS The pressurizer safety valves may be inoperable during a hydro test of the RCS Note provided the pressurizer power operated relief valves and the safety valves on the discharge pump are set at > the test pressure +35 psi.

Insert Page 3.4.10-1 Attachment 1, Volume 9, Rev. 1, Page 200 of 478

Attachment 1, Volume 9, Rev. 1, Page 201 of 478 CTS Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M01 SR 3.4.10.1 Verify each pressurizer safety valve is OPERABLE In accordance in accordance with the Inservice Testing Program. with the Inservice Following testing, lift settings shall be within +/- 1%. Testing Program WOG STS 3.4.10-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 201 of 478

Attachment 1, Volume 9, Rev. 1, Page 202 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.10, PRESSURIZER SAFETY VALVES

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. Kewaunee Power Station (KPS) has two pressurizer safety valves, and their setpoints are allowed to be +3% of the nomimal set pressure (i.e., 2410.45 psig to 2559.55 psig). The +3% tolerance band is consistent with the allowance in the ISTS SR 3.4.10.1 Bases, which states that the pressurizer safety valve setpoint is +3% for OPERABILITY. Furthermore, the words "or more" in the second Condition of ISTS 3.4.10 Condition B have been deleted since there are only two safety valves in the KPS design.

2. A Note has been added to LCO 3.4.10 that allows the pressurizer safety valves to be inoperable during a hydro test of the RCS provided that the pressurizer power operated relief valves and the safety valves on the discharge pump (of the hydrostatic test rig) are set for the test pressure plus 35 psi (i.e., > the test pressure

+35 psi). This change is consistent with the current Technical Specifications (CTS 3.1.a.3) and was approved in License Amendment 164, dated 11/07/02 (ADAMS Accession No. ML023120408).

3. ISTS 3.4.10 Applicability for MODE 4 allows a choice of either specifying the actual Low Temperature Overpressure Protection (LTOP) arming temperature value (e.g.,

275ºF in the ISTS) or stating that the value is specified in the PTLR (i.e., > LTOP arming temperature specified in the PTLR). Kewaunee Power Station (KPS) has decided not to incorporate a PTLR allowance into the ITS, therefore the Applicability should reference the actual LTOP arming temperature. However, the current LTOP enabling temperature is 200°F, which is the exact temperature value at which MODE 4 starts. Thus, the words "with all RCS cold legs temperatures >" and the KPS enabling temperature of 200°F are redundant and have not been included (i.e., only MODE 4 is needed). Furthermore, ISTS 3.4.10 Required Action B.2 uses similar words as in the Applicability. These words have been changed to "Be in MODE 5" for the same reason (i.e., MODE 5 is 200°F).

4. The Note for the ISTS 3.4.10 Applicability allows the lift settings to not be met for 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months . This time is bracketed, and the ISTS Bases states that the 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months is based on an 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months outage time for each of the three pressurizer safety valves. The KPS design only includes two pressurizer safety valves, thus the time allowed in the Applicability Note has been decreased to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

5. The Completion Time provided in ISTS 3.4.10 Required Action B.2 is a bracketed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For KPS, the LTOP arming temperature is 200°F. Therefore, since this is the MODE 5 entry temperature, the Completion Time has been changed to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months consistent with the time normally provided in the ISTS to reach MODE 5.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 202 of 478

Attachment 1, Volume 9, Rev. 1, Page 203 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 203 of 478

Attachment 1, Volume 9, Rev. 1, Page 204 of 478 Pressurizer Safety Valves B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.10 Pressurizer Safety Valves BASES BACKGROUND The pressurizer safety valves provide, in conjunction with the Reactor Protection System, overpressure protection for the RCS. The pressurizer safety valves are totally enclosed pop type, spring loaded, self actuated valves with backpressure compensation. The safety valves are designed to prevent the system pressure from exceeding the system Safety Limit (SL), [2735] psig, which is 110% of the design pressure. 1 2 approximately Because the safety valves are totally enclosed and self actuating, they 345,000 are considered independent components. The relief capacity for each valve, [380,000] lb/hr, is based on postulated overpressure transient 1 conditions resulting from a complete loss of steam flow to the turbine.

This event results in the maximum surge rate into the pressurizer, which specifies the minimum relief capacity for the safety valves. The discharge flow from the pressurizer safety valves is directed to the pressurizer relief tank. This discharge flow is indicated by an increase in temperature downstream of the pressurizer safety valves or increase in the pressurizer relief tank temperature or level.

Overpressure protection is required in MODES 1, 2, 3, 4, and 5; however, in MODE 4, with one or more RCS cold leg temperatures [275°F] [Low 5 Temperature Overpressure Protection (LTOP) arming temperature specified in the PTLR], and MODE 5 and MODE 6 with the reactor vessel head on, overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."

3 The upper and lower pressure limits are based on the +/- 1% tolerance 4 requirement (Ref. 1) for lifting pressures above 1000 psig. The lift setting is for the ambient conditions associated with MODES 1, 2, and 3. This requires either that the valves be set hot or that a correlation between hot and cold settings be established.

The pressurizer safety valves are part of the primary success path and mitigate the effects of postulated accidents. OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design pressure. The consequences of exceeding the American Society of Mechanical Engineers (ASME) pressure limit (Ref. 1) could include damage to RCS components, increased leakage, or a requirement to perform additional stress analyses prior to resumption of reactor operation.

WOG STS B 3.4.10-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 204 of 478

Attachment 1, Volume 9, Rev. 1, Page 205 of 478 Pressurizer Safety Valves B 3.4.10 BASES U

APPLICABLE All accident and safety analyses in the FSAR (Ref. 2) that require safety 2 two 1 SAFETY valve actuation assume operation of three pressurizer safety valves to ANALYSES limit increases in RCS pressure. The overpressure protection analysis two (Ref. 3) is also based on operation of [three] safety valves. Accidents that 1 could result in overpressurization if not properly terminated include:

a. Uncontrolled rod withdrawal from full power, 2 3 RCCA

b. Loss of reactor coolant flow, 3

c. Loss of external electrical load, 3

d. Loss of normal feedwater, 3

e. Loss of all AC power to station auxiliaries, and 3

f. Locked rotor.

Detailed analyses of the above transients are contained in Reference 2.

Safety valve actuation is required in events c, d, and e (above) to limit the 2 during pressure increase. Compliance with this LCO is consistent with the design bases and accident analyses assumptions.

Pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

two LCO The [three] pressurizer safety valves are set to open at the RCS design 1 pressure (2500 psia), and within the ASME specified tolerance, to avoid exceeding the maximum design pressure SL, to maintain accident analyses assumptions, and to comply with ASME requirements. The 4

upper and lower pressure tolerance limits are based on the +/- 1% 3 tolerance requirements (Ref. 1) for lifting pressures above 1000 psig.

The limit protected by this Specification is the reactor coolant pressure boundary (RCPB) SL of 110% of design pressure. Inoperability of one or more valves could result in exceeding the SL if a transient were to occur.

The consequences of exceeding the ASME pressure limit could include damage to one or more RCS components, increased leakage, or additional stress analysis being required prior to resumption of reactor operation. INSERT 1 4 APPLICABILITY In MODES 1, 2, and 3, and portions of MODE 4 above the LTOP arming 4 temperature, OPERABILITY of [three] valves is required because the two 3 combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents. MODE 3 and portions 4 of MODE 4 are conservatively included, although the listed accidents may not require the safety valves for protection.

WOG STS B 3.4.10-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 205 of 478

Attachment 1, Volume 9, Rev. 1, Page 206 of 478 B 3.4.10 4

INSERT 1 The LCO is modified by a Note which allows the pressurizer safety valves to be inoperable during a hydro test of the RCS. During the hydro test of the RCS, the pressurizer safety valves may be blanked provided the pressurizer power operated relief valves and the safety valve on the discharge pump are set at the test pressure +35 psi to protect the system during this test.

Insert Page B 3.4.10-2 Attachment 1, Volume 9, Rev. 1, Page 206 of 478

Attachment 1, Volume 9, Rev. 1, Page 207 of 478 Pressurizer Safety Valves B 3.4.10 BASES APPLICABILITY (continued)

The LCO is not applicable in MODE 4 when any RCS cold leg temperatures are [275°F] [Low Temperature Overpressure Protection 5 (LTOP) arming temperature specified in the PTLR] or in MODE 5 because LTOP is provided. Overpressure protection is not required in MODE 6 with reactor vessel head detensioned.

The Note allows entry into MODES 3 and 4 with the lift settings outside the LCO limits. This permits testing and examination of the safety valves at high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design 36 condition. Only one valve at a time will be removed from service for two testing. The [54] hour exception is based on 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months outage time for 1

each of the [three] valves. The 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months period is derived from operating experience that hot testing can be performed in this timeframe.

ACTIONS A.1 With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the RCS Overpressure Protection System. An inoperable safety valve coincident with an RCS overpressure event could challenge the integrity of the pressure boundary.

B.1 and B.2 If the Required Action of A.1 cannot be met within the required Completion Time or if two or more pressurizer safety valves are 4 inoperable, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be 5 brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 with any RCS cold leg temperatures [275°F] [Low Temperature Overpressure 4 Protection (LTOP) arming temperature specified in the PTLR] within 36 [24] hours. The allowed Completion Times are reasonable, based on 4 operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant In MODE 5 systems. With any RCS cold leg temperatures at or below [275°F] [Low 4 Temperature Overpressure (LTOP) arming temperature specified in the PTLR], overpressure protection is provided by the LTOP System. The 5 4

change from MODE 1, 2, or 3 to MODE 4 reduces the RCS energy (core

, or 4 power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by [three] two 1 pressurizer safety valves.

WOG STS B 3.4.10-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 207 of 478

Attachment 1, Volume 9, Rev. 1, Page 208 of 478 Pressurizer Safety Valves B 3.4.10 BASES SURVEILLANCE SR 3.4.10.1 REQUIREMENTS SRs are specified in the Inservice Testing Program. Pressurizer safety valves are to be tested in accordance with the requirements of the ASME Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs. No additional requirements are specified.

The pressurizer safety valve setpoint is +/- [3]% for OPERABILITY; 1 however, the valves are reset to +/- 1% during the Surveillance to allow for drift.

REFERENCES 1. ASME, Boiler and Pressure Vessel Code,Section III.

14 1

2. FSAR, Chapter [15]. 2 U

3. WCAP-7769, Rev. 1, June 1972. October 1971 2

4. ASME Code for Operation and Maintenance of Nuclear Power Plants.

WOG STS B 3.4.10-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 208 of 478

Attachment 1, Volume 9, Rev. 1, Page 209 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.10 BASES, PRESSURIZER SAFETY VALVES

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. 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. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

4. Changes are made to be consistent with the Specification.

5. The Applicability of ITS 3.4.12 does not include MODE 4. Therefore, the MODE 4 Applicability reference has been deleted.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 209 of 478

Attachment 1, Volume 9, Rev. 1, Page 210 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 210 of 478

Attachment 1, Volume 9, Rev. 1, Page 211 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.10, PRESSURIZER SAFETY VALVES There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 211 of 478

Attachment 1, Volume 9, Rev. 1, Page 212 of 478 ATTACHMENT 11 ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVS)

Attachment 1, Volume 9, Rev. 1, Page 212 of 478

, Volume 9, Rev. 1, Page 213 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 213 of 478

Attachment 1, Volume 9, Rev. 1, Page 214 of 478 ITS A01 ITS 3.4.11

4. Pressure Isolation Valves A. All pressure isolation valves listed in Table TS 3.1-2 shall be functional as a pressure isolation device during OPERATING and HOT STANDBY MODES, except as specified in 3.1.a.4.B. Valve leakage shall not exceed the amounts indicated.

B. In the event that integrity of any pressure isolation valve as specified in See ITS Table TS 3.1-2 cannot be demonstrated, reactor operation may continue, 3.4.14 provided that at least two valves in each high pressure line having a non-functional valve are in, and remain in, the mode corresponding to the isolated condition. (1)

C. If TS 3.1.a.4.A and TS 3.1.a.4.B cannot be met, then an orderly shutdown shall be initiated and the reactor shall be in the HOT SHUTDOWN condition within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the INTERMEDIATE SHUTDOWN condition in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the COLD SHUTDOWN condition within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5. Pressurizer Power-Operated Relief Valves (PORV) and PORV Block Valves Add proposed MODE 3 M01 LCO 3.4.11 A. Two PORVs and their associated block valves shall be OPERABLE during HOT Applicability A02 STANDBY and OPERATING modes. Add proposed ACTIONS Note and capable of being manually cycled A03

1. With one or both PORVs inoperable because of excessive seat leakage, ACTION A within one hour either restore the PORV(s) to OPERABLE status or close A04 the associated block valve(s) with power maintained to the block valve(s);

otherwise, action shall be initiated to: M02 M01 ACTION D

- Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Add proposed Required Action D.2

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and not capable of being manually cycled A03

2. With one PORV inoperable due to causes other than excessive seat ACTION B leakage, within one hour either restore the PORV to OPERABLE status or A04 close its associated block valve and remove power from the block valve.

Restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or action shall be initiated to: M02 M01 ACTION D

- Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Add proposed Required Action D.2

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months A03 and not capable of being manually cycled

3. With both PORVs inoperable due to causes other than excessive seat leakage, within one hour either restore at least one PORV to OPERABLE A04 status or close its associated block valve and remove power from the block ACTION E valve and M02

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months M01 Add proposed Required Action E.4 (1)

Manual valves shall be locked in the closed position. Motor operated valves shall be placed See ITS in the closed position with their power breakers locked out. 3.4.14 Amendment No. 165 TS 3.1-4 03/11/2003 Page 1 of 4 Attachment 1, Volume 9, Rev. 1, Page 214 of 478

Attachment 1, Volume 9, Rev. 1, Page 215 of 478 ITS A01 ITS 3.4.11 Add proposed ACTION C Note A05

4. With one block valve inoperable, within one hour restore the block valve to ACTION C OPERABLE status or place its associated PORV in manual control. Restore A04 the block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise action shall be initiated to: M02 ACTION D M01

- Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Add proposed Required Action D.2

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Add proposed ACTION F Note

5. With both block valves inoperable, within one hour restore the block valves A05 ACTION F to OPERABLE status or place their associated PORVs in manual control.

Restore at least one block valve to OPERABLE status within the next hour; A04 otherwise, action shall be initiated to:

M02 ACTION G

- Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Add proposed Required M01 Action G.2

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

6. Pressurizer Heaters See ITS A. At least one group of pressurizer heaters shall have an emergency power supply 3.4.9 available when the average RCS temperature is > 350°F.

7. Reactor Coolant Vent System A. A reactor coolant vent path from both the reactor vessel head and pressurizer steam space shall be OPERABLE and closed prior to the average RCS temperature being heated > 200°F except as specified in TS 3.1.a.7.B and TS 3.1.a.7.C below.

B. When the average RCS temperature is > 200°F, any one of the following conditions of inoperability may exist:

1. Both of the parallel vent valves in the reactor vessel vent path are inoperable.

See CTS 3.1.a.7

2. Both of the parallel vent valves in the pressurizer vent path are inoperable.

If OPERABILITY is not restored within 30 days, then within one hour action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve COLD SHUTDOWN within an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. If no Reactor Coolant System vent paths are OPERABLE, then restore at least one vent path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If OPERABILITY is not restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , then within one hour action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve COLD SHUTDOWN within an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Amendment No. 165 TS 3.1-5 03/11/2003 Page 2 of 4 Attachment 1, Volume 9, Rev. 1, Page 215 of 478

Attachment 1, Volume 9, Rev. 1, Page 216 of 478 ITS A01 ITS 3.4.11 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 3.4.11.1, b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and SR 3.4.11.2 TS 4.1-3.

c. Deleted

d. Deleted

e. Deleted Amendment No. 119 TS 4.1-1 04/18/95 Page 3 of 4 Attachment 1, Volume 9, Rev. 1, Page 216 of 478

Attachment 1, Volume 9, Rev. 1, Page 217 of 478 ITS A01 ITS 3.4.11 TABLE TS 4.1-3 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS L01 (1)

EQUIPMENT TESTS TEST FREQUENCY

1. Control Rods Rod drop times of all full Each REFUELING outage length rods Quarterly when at or above HOT See ITS Partial movement of all STANDBY 3.1.4 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 3.6.3

4. Containment Isolation Trip OPERABILITY Each REFUELING outage

5. Refueling System Interlocks OPERABILITY Prior to fuel movement each See CTS REFUELING outage 3.8.a.11

6. Deleted

7. Deleted See ITS (4) 3.4.15

8. RCS Leak Detection OPERABILITY Weekly See ITS

9. Diesel Fuel Supply Fuel Inventory(5) Weekly 3.8.1 and

10. Deleted 3.8.3

11. Fuel Assemblies Visual Inspection Each REFUELING outage See ITS 4.0

12. Guard Pipes Visual Inspection Each REFUELING outage See ITS SR 3.4.11.2 13. Pressurizer PORVs OPERABILITY Each REFUELING cycle 3.6.1 SR 3.4.11.1 14. Pressurizer PORV Block Valves OPERABILITY Quarterly(6)

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 Add proposed SR 3.4.11.1 Note 2 A06 Add proposed SR 3.4.11.2 Note (1)

Following maintenance on equipment that could affect the operation of the equipment, tests L01 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 See ITS 3.3.1 and undervoltage trip attachments. The test shall also verify the undervoltage trip attachment on the reactor trip bypass breakers.

(4) See ITS When reactor is at power or in HOT SHUTDOWN condition. 3.4.15 (5) See ITS 3.8.1 Inventory of fuel required in all plant modes. and 3.8.3 SR 3.4.11.1 (6)

Not required when valve is administratively closed. See ITS Note 1 (7) 3.4.9 Test will verify OPERABILITY of heaters and availability of an emergency power supply.

(8)

This test shall demonstrate that the valve(s) close in 5 seconds. See ITS 3.6.3 Amendment No. 125 Page 1 of 1 08/07/96 Page 4 of 4 Attachment 1, Volume 9, Rev. 1, Page 217 of 478

Attachment 1, Volume 9, Rev. 1, Page 218 of 478 DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVS)

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.a.5.A describes the compensatory actions to take when PORV(s) and/or block valve(s) are inoperable. ITS 3.4.11 ACTIONS A, B, C, D, E, F, and G also state the appropriate compensatory actions under the same conditions, however, an ITS 3.4.11 ACTIONS Note has been added. The ITS 3.4.11 ACTION Note allows separate Condition entry for each Pressurizer PORV and PORV block valve. This changes the CTS by explicitly stating the Actions are to be taken separately for each inoperable Pressurizer PORV and PORV block valve.

The purpose of the Note is to provide explicit instructions for proper application of the Action for Technical Specification compliance. In conjunction with proposed Specification 1.3, "Completion Times," this Note provides direction consistent with the intent of the existing Action for inoperable Pressurizer PORVs and PORV block valves. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS 3.1.a.5.A.1 applies to one or both PORVs inoperable solely due to excessive seat leakage. CTS 3.1.a.5.A.2 and 3.1.a.5.A.3 applies to one or both PORVs inoperable, respectively, due to causes other than excessive seat leakage. ITS 3.4.11 ACTIONS divide the conditions of PORV inoperability into those in which the PORV is capable of being manually cycled and those which the PORV is not capable of being manually cycled. ITS 3.4.11 ACTION A applies to one or more PORVs inoperable and capable of being manually cycled.

ITS 3.4.11 ACTION B applies to one PORV inoperable and not capable of being manually cycled. ITS 3.4.11 ACTION E applies to two PORVs inoperable and not capable of being manually cycled. This changes the CTS by dividing the existing conditions into those in which the PORV can, and cannot, be manually cycled.

This change is acceptable because the requirements have not changed. A PORV inoperable due to excessive seat leakage can still be manually cycled.

PORVs inoperable for other reasons cannot be manually cycled. Therefore, the conditions under which the Required Actions are applied have not changed. This change is designated as administrative because it does not result in a technical change to the CTS.

A04 When PORVs or block valves are inoperable, CTS 3.1.a.5.A.1, 3.1.a.5.A.2, and 3.1.a.5.A.3 provide an option to restore inoperable PORVs to OPERABLE status.

CTS 3.1.a.5.A.4 and 3.1.a.5.A.5 provide an option to restore inoperable block valves to OPERABLE status. ITS 3.4.11 does not include this explicit option to Kewaunee Power Station Page 1 of 4 Attachment 1, Volume 9, Rev. 1, Page 218 of 478

Attachment 1, Volume 9, Rev. 1, Page 219 of 478 DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVS) restore the valves to OPERABLE status. This changes the CTS by not explicitly stating the option to restore the valves to OPERABLE status.

The purpose of the CTS actions is to provide all of the acceptable options for inoperable PORVs and block valves. This change is acceptable because the requirements have not changed. LCO 3.0.3 states that upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. Therefore, it is not necessary to state the option to restore the inoperable valves to OPERABLE status. When they are restored LCO 3.0.2 allows exiting from the Condition. This change is designated as administrative as it allows a change required by the ITS usage rules that does not result in a technical change to the CTS.

A05 CTS 3.1.a.5.A.4 specifies the compensatory actions for one inoperable block valve. CTS 3.1.a.5.A.5 specifies the compensatory actions for two inoperable block valves. ITS 3.4.11 ACTION C specifies the Required Actions for one inoperable block valve and ITS 3.4.11 ACTION F specifies the Required Actions for two inoperable block valves. ITS 3.4.11 ACTION C Required Actions are preceded by a Note that states that the specified Required Actions (C.1 and C.2) do not apply when the block valve is inoperable solely as a result of complying with Required Action B.2 or E.2 and ITS 3.4.11 ACTION F Required Action is preceded by a Note that states that the specified Required Action (F.1) does not apply when the block valve is inoperable solely as a result of complying with Required Action B.2 or E.2. ITS 3.4.11 Required Actions B.2 and E.2 require the removal of power from the applicable block valve when a PORV is inoperable.

This changes the CTS by adding the clarification Note that the Required Action to place the PORV in manual control (ITS 3.4.11 Required Action C.1) and to restore a block valve to OPERABLE status (ITS 3.4.11 Required Actions C.2 and F.1) are not applicable when the block valve is inoperable solely due to complying with the ACTIONS for an inoperable PORV.

The purpose of the CTS 3.1.a.5.A Actions is to ensure the appropriate compensatory measures are taken with inoperable PORVs or inoperable block valves. The Note clarifies that the applicable Required Actions for block valves are not necessary when entry into the Condition is made as a result of application of the Required Actions for inoperable PORVs that are not capable of being manually cycled. This clarification is acceptable since these actions (place associated PORV in manual control or restore one block valve to OPERABLE status) are not appropriate for the block valve inoperability, and are not required in the CTS (since the individual PORV actions in CTS 3.1.a.5.A.1, 3.1.a.5.A.2, and 3.1.a.5.A.3 do not require anything to be done to the associated block valves as part of the actions). This change is designated as administrative since the change does not result in a technical change to the CTS.

A06 CTS Table 4.1-3 Equipment Test 13 requires an operability test of the Pressurizer PORVs every REFUELING outage. CTS Table 4.1-3 Equipment Test 14 requires an operability test of the Pressurizer PORV Block Valves quarterly, with the exception that the Pressurizer PORV Block Valves do not require testing when the valve is administratively closed. As stated in CTS Kewaunee Power Station Page 2 of 4 Attachment 1, Volume 9, Rev. 1, Page 219 of 478

Attachment 1, Volume 9, Rev. 1, Page 220 of 478 DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVS) 3.1.a.5.A, the PORVs and associated block valves are only required to be OPERABLE in the OPERATING (equivalent to ITS MODE 1) and HOT STANDBY (equivalent to ITS MODE 2) MODES. Thus, these Surveillances are only required to be performed in the same MODES. ITS 3.4.11 is Applicable in MODES 1, 2, and 3 (as described in Discussion of Change M01), thus the ITS Surveillances are also normally required in MODES 1, 2, and 3. ITS SR 3.4.11.1 requires performance of a complete cycle of each block valve, but Note 2 allows entry into and operation in MODE 3 prior to performing the Surveillance Requirement. ITS SR 3.4.11.2 requires performance of a complete cycle of each PORV, but the Note allows entry into and operation in MODE 3 prior to performing the Surveillance Requirement. This changes the CTS by more clearly stating in the individual Surveillances that testing is not required to be performed in MODE 3.

The purpose of the Notes is to allow entry into MODE 3 in order to perform the required tests. ITS SR 3.0.1 and SR 3.0.4 would normally preclude entering MODE 3 with the Surveillance not current. This change is acceptable because it is consistent with the current allowances in the CTS. Since the PORVs and block valves are not currently required to be OPERABLE in MODE 3, it is acceptable to enter MODE 3 prior to performing the two tests. The Notes are needed since the ITS has added a new MODE 3 Applicability as discussed in Discussion of Change M01. This change is designated as administrative because it does not result in any technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.a.5.A requires the PORVs and associated block valves to be OPERABLE in the OPERATING (equivalent to ITS MODE 1) and HOT STANDBY (equivalent to ITS MODE 2) modes. Furthermore, when a unit shutdown is required by CTS 3.1.a.5.A, the unit is required to be shut down to HOT SHUTDOWN (equivalent to ITS MODE 3). ITS 3.4.11 requires the PORVs and associated block valves to be OPERABLE in MODES 1, 2, and 3.

Consistent with the change in Applicability, the requirement to be in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is added to all cases where a unit shutdown is required as indicated in Required Actions D.2, E.4, and G.2. This changes the CTS by requiring the PORVs and associated block valves to be OPERABLE in MODE 3 and providing a Required Action to place the unit outside the Applicability.

The purpose of the PORVs and associated block valves is to help minimize the consequences of a steam generator tube rupture event. The addition of the MODE 3 Applicability is acceptable since the reactor is pressurized and near normal operating temperature in MODE 3, and the need to depressurize the reactor to MODE 4 conditions following a tube rupture is necessary. This change is more restrictive because a new Applicability containing MODE 3 and a commensurate Required Action to shut down to MODE 4 has been added.

M02 When a unit shutdown is required by CTS 3.1.a.5.A.1, 3.1.a.5.A.2, 3.1.a.5.A.3, 3.1.a.5.A.4, or 3.1.a.5.A.5, the unit is required to be in HOT STANDBY (equivalent to ITS MODE 2) in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN (equivalent to ITS MODE 3) in the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Under similar conditions in the ITS, the unit is Kewaunee Power Station Page 3 of 4 Attachment 1, Volume 9, Rev. 1, Page 220 of 478

Attachment 1, Volume 9, Rev. 1, Page 221 of 478 DISCUSSION OF CHANGES ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVS) required to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by requiring the unit to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months in lieu of the current 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time. The requirement to be in MODE 4 is discussed in Discussion of Change M01.

The purpose of the shutdown actions is to place the unit in a MODE in which the Specification is not required in an expeditious but safe manner. The change is acceptable since the allowed Completion Time to reach MODE 3 is reasonable, based on operating experience, to reach the required plant condition from full power conditions in an orderly manner and without challenging plant systems.

This change is designated as more restrictive since less time is allowed to reach MODE 3 in the ITS than is allowed in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 5 - Deletion of Surveillance Requirement) Note 1 to CTS Table 4.1-3 requires, in part, that the pressurizer PORVs and pressurizer PORV Block Valves be tested to verify OPERABILITY following maintenance on equipment that could affect the operation. ITS 3.4.11 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 a 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 4 of 4 Attachment 1, Volume 9, Rev. 1, Page 221 of 478

Attachment 1, Volume 9, Rev. 1, Page 222 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 222 of 478

Attachment 1, Volume 9, Rev. 1, Page 223 of 478 CTS Pressurizer PORVs 3.4.11 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) 3.1.a.5.A LCO 3.4.11 Each PORV and associated block valve shall be OPERABLE.

3.1.a.5.A APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

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

DOC A02 Separate Condition entry is allowed for each PORV and each block valve.

CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.5.A.1 A. One or more PORVs A.1 Close and maintain power 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable and capable to associated block valve.

of being manually cycled.

3.1.a.5.A.2 B. One [or two] PORV[s] B.1 Close associated block 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 inoperable and not valve[s].

capable of being manually cycled. AND B.2 Remove power from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated block valve[s]. 1 AND 1

B.3 Restore PORV[s] to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

WOG STS 3.4.11-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 223 of 478

Attachment 1, Volume 9, Rev. 1, Page 224 of 478 CTS Pressurizer PORVs 3.4.11 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.5.A.4 C. One [or two] block --------------------NOTE-------------------

1 valve(s) inoperable. Required Actions C.1 and C.2 do not apply when block valve is inoperable solely as a result of 5 complying with Required Actions B.2 or E.2.

C.1 Place associated PORV in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months manual control.

AND C.2 Restore block valve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

3.1.a.5.A.1, D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 3.1.a.5.A.2, 3.1.a.5.A.4 associated Completion Time of Condition A, B, AND or C not met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 3.1.a.5.A.3 E. Two [or three] PORVs E.1 Close associated block 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 inoperable and not valves.

capable of being manually cycled. AND E.2 Remove power from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated block valves.

AND E.3 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND E.4 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months WOG STS 3.4.11-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 224 of 478

Attachment 1, Volume 9, Rev. 1, Page 225 of 478 CTS Pressurizer PORVs 3.4.11 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.5.A.5 F. Two [or three] block --------------------NOTE------------------- 1 valves inoperable. Required Action F.1 does not apply when block valve is inoperable solely as a result of complying with 5

Required Actions B.2 or E.2.

F.1 Restore one block valve to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s]

OPERABLE status [if three 1

block valves are inoperable].

[

3.1.a.5.A.5 G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition F not AND met.

G.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months WOG STS 3.4.11-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 225 of 478

Attachment 1, Volume 9, Rev. 1, Page 226 of 478 CTS Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 ------------------------------NOTES-----------------------------

Table 4.1.3 1. Not required to be performed with block valve Note 6 closed in accordance with the Required Actions of this LCO.

DOC A06 2. Only required to be performed in MODES 1 and 2.

Table 4.1.3 Equipment Perform a complete cycle of each block valve. 92 days Test 14 SR 3.4.11.2 -------------------------------NOTE------------------------------

DOC A06 Only required to be performed in MODES 1 and 2.

Table 4.1.3 Equipment Perform a complete cycle of each PORV. [18] months 2 Test 13 SR 3.4.11.3 [ Perform a complete cycle of each solenoid air [18] months ] 3 control valve and check valve on the air accumulators in PORV control systems.

SR 3.4.11.4 [ Verify PORVs and block valves are capable of [18] months ] 4 being powered from emergency power sources.

WOG STS 3.4.11-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 226 of 478

Attachment 1, Volume 9, Rev. 1, Page 227 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVS)

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. The Kewaunee Power Station design includes only two PORVs and associated block valves.

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 refueling outage cycle length (18 months).

3. KPS does not require the air accumulators for the PORVS to be OPERABLE. This is documented in the NRC Safety Evaluation for Amendment 108, dated April 7, 1994 (ADAMS Accession No. ML020770581). Therefore, ISTS SR 3.4.11.3, which tests the solenoid air valves and check valves of the accumulators, has not been added.

4. The Bases for ISTS SR 3.4.11.4 contains a statement which says, "This Surveillance is not required for plants with permanent 1E power supplies to the valves." KPS has permanent 1E power supplies to the valves as described in the ITS 3.4.11 Bases, therefore, the Surveillance is not required.

5. Grammatical error corrected.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 227 of 478

Attachment 1, Volume 9, Rev. 1, Page 228 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 228 of 478

Attachment 1, Volume 9, Rev. 1, Page 229 of 478 Pressurizer PORVs B 3.4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)

BASES BACKGROUND The pressurizer is equipped with two types of devices for pressure relief:

pressurizer safety valves and PORVs. The PORVs are air operated valves that are controlled to open at a specific set pressure when the pressurizer pressure increases and close when the pressurizer pressure decreases. The PORVs may also be manually operated from the control room.

Block valves, which are normally open, are located between the pressurizer and the PORVs. The block valves are used to isolate the PORVs in case of excessive leakage or a stuck open PORV. Block valve closure is accomplished manually using controls in the control room. A stuck open PORV is, in effect, a small break loss of coolant accident (LOCA). As such, block valve closure terminates the RCS depressurization and coolant inventory loss.

The PORVs and their associated block valves may be used by plant operators to depressurize the RCS to recover from certain transients if normal pressurizer spray is not available. Additionally, the series arrangement of the PORVs and their block valves permit performance of surveillances on the valves during power operation. 11 block The PORVs may also be used for feed and bleed core cooling in the case of multiple equipment failure events that are not within the design basis, such as a total loss of feedwater.

The PORVs, their block valves, and their controls are powered from the vital buses that normally receive power from offsite power sources, but are also capable of being powered from emergency power sources in the event of a loss of offsite power. Two PORVs and their associated block valves are powered from two separate safety trains (Ref. 1).

The plant has two PORVs, each having a relief capacity of 210,000 lb/hr 1 at 2335 psig. The functional design of the PORVs is based on 179,000 maintaining pressure below the Pressurizer Pressure - High reactor trip setpoint following a step reduction of 50% of full load with steam dump.

In addition, the PORVs minimize challenges to the pressurizer safety valves and also may be used for low temperature overpressure protection 2

(LTOP). See LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System."

WOG STS B 3.4.11-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 229 of 478

Attachment 1, Volume 9, Rev. 1, Page 230 of 478 Pressurizer PORVs B 3.4.11 BASES APPLICABLE Plant operators employ the PORVs to depressurize the RCS in response SAFETY to certain plant transients if normal pressurizer spray is not available.

ANALYSES For the Steam Generator Tube Rupture (SGTR) event, the safety analysis assumes that manual operator actions are required to mitigate the event. A loss of offsite power is assumed to accompany the event, and thus, normal pressurizer spray is unavailable to reduce RCS pressure. The PORVs are assumed to be used for RCS depressurization, which is one of the steps performed to equalize the primary and secondary pressures in order to terminate the primary to secondary break flow and the radioactive releases from the affected steam generator.

The PORVs are also modeled in safety analyses for events that result in increasing RCS pressure for which departure from nucleate boiling ratio (DNBR) criteria are critical (Ref. 2). By assuming PORV actuation, the primary pressure remains below the high pressurizer pressure trip setpoint; thus, the DNBR calculation is more conservative. As such, this actuation is not required to mitigate these events, and PORV automatic operation is, therefore, not an assumed safety function.

Pressurizer PORVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires the PORVs and their associated block valves to be OPERABLE for manual operation to mitigate the effects associated with an SGTR.

By maintaining two PORVs and their associated block valves redundancy is maintained consistent OPERABLE, the single failure criterion is satisfied. An OPERABLE block 10 with the applicable design basis.

valve may be either open and energized with the capability to be closed, or closed and energized with the capability to be opened, since the required safety function is accomplished by manual operation. Although typically open to allow PORV operation, the block valves may be OPERABLE when closed to isolate the flow path of an inoperable PORV that is capable of being manually cycled (e.g., as in the case of excessive PORV leakage). Similarly, isolation of an OPERABLE PORV does not render that PORV or block valve inoperable provided the relief function remains available with manual action.

An OPERABLE PORV is required to be capable of manually opening and closing, and not experiencing excessive seat leakage. Excessive seat leakage, although not associated with a specific acceptance criteria, exists when conditions dictate closure of the block valve to limit leakage.

Satisfying the LCO helps minimize challenges to fission product barriers.

WOG STS B 3.4.11-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 230 of 478

Attachment 1, Volume 9, Rev. 1, Page 231 of 478 Pressurizer PORVs B 3.4.11 BASES APPLICABILITY In MODES 1, 2, and 3, the PORV and its block valve are required to be OPERABLE to limit the potential for a small break LOCA through the flow path. The most likely cause for a PORV small break LOCA is a result of a pressure increase transient that causes the PORV to open. Imbalances in the energy output of the core and heat removal by the secondary system can cause the RCS pressure to increase to the PORV opening setpoint. The most rapid increases will occur at the higher operating power and pressure conditions of MODES 1 and 2. The PORVs are also required to be OPERABLE in MODES 1, 2, and 3 for manual actuation to mitigate a steam generator tube rupture event.

Pressure increases are less prominent in MODE 3 because the core input energy is reduced, but the RCS pressure is high. Therefore, the LCO is applicable in MODES 1, 2, and 3. The LCO is not applicable in MODES 4, 5, and 6 with the reactor vessel head in place when both pressure and core energy are decreased and the pressure surges become much less significant. LCO 3.4.12 addresses the PORV 2 requirements in these MODES.

3 ACTIONS A Note 1 has been added to clarify that all pressurizer PORVs and block valves are treated as separate entities, each with separate Completion Times (i.e., the Completion Time is on a component basis).

REVIEWERS NOTE-----------------------------------

The bracketed options in Conditions B, C, E, and F are to accommodate 4 plants with three PORVs and associated block valves.

A.1 PORVs may be inoperable and capable of being manually cycled (e.g.,

excessive seat leakage). In this condition, either the PORVs must be 3 restored or the flow path isolated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The associated block valve is required to be closed, but power must be maintained to the associated block valve, since removal of power would render the block valve inoperable. This permits operation of the plant until the next 5

refueling outage (MODE 6) so that maintenance can be performed on the PORVs to eliminate the problem condition.

Quick access to the PORV for pressure control can be made when power remains on the closed block valve. The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is based on plant operating experience that has shown that minor problems can be corrected or closure accomplished in this time period.

WOG STS B 3.4.11-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 231 of 478

Attachment 1, Volume 9, Rev. 1, Page 232 of 478 Pressurizer PORVs B 3.4.11 BASES ACTIONS (continued)

B.1, B.2, and B.3 6

If one [or two] PORV[s] is inoperable and not capable of being manually 3

cycled, it must be either restored, or isolated by closing the associated block valve and removing the power to the associated block valve. The Completion Times of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months are reasonable, based on challenges to the PORVs during this time period, and provide the operator adequate time to correct the situation. If the inoperable valve cannot be restored to 3 OPERABLE status, it must be isolated within the specified time. Because there is at least one PORV that remains OPERABLE, an additional 3 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is provided to restore the inoperable PORV to OPERABLE status. If the PORV cannot be restored within this additional time, the 3 plant must be brought to a MODE in which the LCO does not apply, as required by Condition D.

C.1 and C.2 is If one [or two] block valve(s) are inoperable, then it is necessary to either 6 restore the block valve(s) to OPERABLE status within the Completion 3 Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or place the associated PORV in manual control. The prime importance for the capability to close the block valve(s) is to isolate a stuck open PORV. Therefore, if the block valve(s) cannot be restored 3 to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the Required Action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck open PORV at a time that the block valve(s) are inoperable. The Completion Time of is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is reasonable, based on the small potential for challenges to the system during this time period, and provides the operator time to correct the situation. Because at least one PORV remains OPERABLE, the operator is permitted a Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the inoperable block valve(s) to OPERABLE status. The time allowed to restore the block valve(s) is based upon the Completion Time for restoring an inoperable PORV in Condition B, since the PORVs may not be capable of mitigating an event if the inoperable block valve(s) are not is full open. If the block valve(s) are restored within the Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the PORV may be restored to automatic operation. If it cannot 3 be restored within this additional time, the plant must be brought to a MODE in which the LCO does not apply, as required by Condition D.

WOG STS B 3.4.11-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 232 of 478

Attachment 1, Volume 9, Rev. 1, Page 233 of 478 Pressurizer PORVs B 3.4.11 BASES ACTIONS (continued)

The Required Actions C.1 and C.2 are modified by a Note stating that the Required Actions do not apply if the sole reason for the block valve being declared inoperable is as a result of power being removed to comply with other Required Actions. In this event, the Required Actions for inoperable PORV(s) (which require the block valve power to be removed once it is closed) are adequate to address the condition. While it may be desirable to also place the PORV(s) in manual control, this may not be possible for all causes of Condition B or E entry with PORV(s) inoperable and not capable of being manually cycled (e.g., as a result of failed control power fuse(s) or control switch malfunctions(s)).

D.1 and D.2 If the Required Action of Condition A, B, or C is not met, then 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4 and 5, automatic PORV 2

OPERABILITY may be required. See LCO 3.4.12.

E.1, E.2, E.3, and E.4 s are two 6

If more than one PORV is inoperable and not capable of being manually cycled, it is necessary to either restore at least one valve within the 3 Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or isolate the flow path by closing and removing the power to the associated block valves. The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is reasonable, based on the small potential for challenges to the system during this time and provides the operator time to correct 3

the situation. If no PORVs are restored within the Completion Time, then the plant must be brought to a MODE in which the LCO does not apply.

also To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4 and 5, automatic 2 PORV OPERABILITY may be required. See LCO 3.4.12.

WOG STS B 3.4.11-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 233 of 478

Attachment 1, Volume 9, Rev. 1, Page 234 of 478 Pressurizer PORVs B 3.4.11 BASES ACTIONS (continued)

F.1 If two [or three] block valve(s) are inoperable, it is necessary to restore at 6 least one block valve within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The Completion Time is reasonable, based on the small potential for challenges to the system during this time and provide the operator time to correct the situation.

s Required Action F.1 is modified by a Note stating that the Required Action does not apply if the sole reason for the block valve being declared inoperable is a result of power being removed to comply with other Required Actions. In this event, the Required Actions for inoperable PORV(s) (which require the block valve power to be removed once it is closed) are adequate to address the condition. While it may be desirable to also place the PORV(s) in manual control, this may not be possible for all causes of Condition B or E entry with PORV(s) inoperable and not capable of being manually cycled (e.g., as a result of failed control power fuse(s) or control switch malfunctions(s)).

G.1 and G.2 If the Required Action of Condition F is not met, then 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4 and 5, automatic PORV 2 OPERABILITY may be required. See LCO 3.4.12.

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block valve cycling verifies that the valve(s) can be opened and closed if needed. The basis for the Frequency of 92 days is the ASME Code (Ref. 3).

This SR is modified by two Notes. Note 1 modifies this SR by stating that it is not required to be performed with the block valve closed in accordance with the Required Actions of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable. Note 2 modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR.

This allows the test to be performed in MODE 3 under operating WOG STS B 3.4.11-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 234 of 478

Attachment 1, Volume 9, Rev. 1, Page 235 of 478 Pressurizer PORVs B 3.4.11 BASES SURVEILLANCE REQUIREMENTS (continued) temperature and pressure conditions, prior to entering MODE 1 or 2. [In 7 accordance with Reference 4, administrative controls require this test be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation.] 7 SR 3.4.11.2 SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. The Frequency of

[18] months is based on a typical refueling cycle and industry accepted 7 practice.

The Note modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. [In accordance with Reference 4, administrative 7 controls require this test be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation.] 7

[ SR 3.4.11.3 Operating the solenoid air control valves and check valves on the air accumulators ensures the PORV control system actuates properly when called upon. The Frequency of [18] months is based on a typical 10 refueling cycle and the Frequency of the other Surveillances used to demonstrate PORV OPERABILITY. ]

[ SR 3.4.11.4 This Surveillance is not required for plants with permanent 1E power 8 supplies to the valves.

The Surveillance demonstrates that emergency power can be provided and is performed by transferring power from normal to emergency supply and cycling the valves. The Frequency of [18] months is based on a typical refueling cycle and industry accepted practice. ]

WOG STS B 3.4.11-7 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 235 of 478

Attachment 1, Volume 9, Rev. 1, Page 236 of 478 Pressurizer PORVs B 3.4.11 BASES REFERENCES 1. Regulatory Guide 1.32, February 1977. 9 U 14.1 USAR, Section 4.2.2.4

2. FSAR, Section [15.2]. 9 7

3. ASME Code for Operation and Maintenance of Nuclear Power Plants.

[ 4. Generic Letter 90-06, "Resolution of Generic Issue 70, 'Power- 7 Operated Relief Valve and Block Valve Reliability,' and Generic Issue 94, 'Additional Low-Temperature Overpressure for Light-Water Reactors,' Pursuant to 10 CFR 50.54(f)," June 25, 1990.] 7 WOG STS B 3.4.11-8 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 236 of 478

Attachment 1, Volume 9, Rev. 1, Page 237 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.11 BASES, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVS)

1. ISTS 3.4.11 Bases states that the PORV relief capacity is 210,000 lb/hr at 2335 psig. Kewaunee Power Station (KPS) has a relief capacity of 179,000 lb/hr at 2335 psig. Therefore, the relief capacity was changed to the KPS specific value.

2. ISTS 3.4.11 Bases makes reference that the Pressurizer Power Operated Relief Valves (PORVs) may be used for low temperature overpressure protection (LTOP) and references the ISTS 3.4.12, LTOP specification. KPS does not credit the PORVs in the LTOP. Therefore, the reference to LTOP has been removed.

3. Changes made to reflect the ISTS.

4. The Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed into what is needed to meet this requirement. This is not meant to be retained in the final version of the plant specific submittal.

5. This statement has been deleted since the statement is not valid. The Required Action does not preclude the unit from starting up without performing the maintenance on the valve(s).

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. As stated in the Reviewers note, the bracketed options are for plants with three PORVs and associated block valves. Kewaunee Power Station only has two PORVs and associated block valves.

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.

8. ISTS SR 3.4.11.4 contains a statement which says, "This Surveillance is not required for plants with permanent 1E power supplies to the valves." KPS has permanent 1E power supplies to the valves; therefore, the Surveillance is not required.

9. 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.

10. KPS does not require the air accumulators for the PORVS to be OPERABLE. This is documented in the NRC Safety Evaluation for Amendment 108, dated April 7, 1994 (ADAMS Accession No. ML020770581). Therefore, ISTS SR 3.4.11.3, which tests the solenoid air valves and check valves of the accumulators, has not been added. Furthermore, the PORVs utilize the normal instrument air supply to operate. This air system is highly reliable, but is not a safety grade supply.

Therefore, the PORVs do not meet the single failure criteria, and this wording has been changed in the LCO Bases section to be consistent with the KPS design basis.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 237 of 478

Attachment 1, Volume 9, Rev. 1, Page 238 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.11 BASES, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVS)

11. The ISTS (and ITS) SRs for PORV cycling (SR 3.4.11.2) only requires the test every 18 months. Furthermore, NUREG-1316 emphasizes the importance of not stroking the PORVs during power operation. Therefore, the word "valves" has been replaced with "block valves" to clarify that only block valves can be tested online.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 238 of 478

Attachment 1, Volume 9, Rev. 1, Page 239 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 239 of 478

Attachment 1, Volume 9, Rev. 1, Page 240 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.11, PRESSURIZER POWER OPERATED RELIEF VALVES (PORVS)

There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 240 of 478

Attachment 1, Volume 9, Rev. 1, Page 241 of 478 ATTACHMENT 12 ITS 3.4.12, LOW PRESSURE OVERPRESSURE PROTECTION (LTOP) SYSTEM Attachment 1, Volume 9, Rev. 1, Page 241 of 478

, Volume 9, Rev. 1, Page 242 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 242 of 478

Attachment 1, Volume 9, Rev. 1, Page 243 of 478 ITS A01 ITS 3.4.12

b. Heatup and Cooldown Limit Curves for Normal Operation

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures TS 3.1-1 and TS 3.1-2. Figures TS 3.1-1 and TS 3.1-2 are applicable for the service period of up to 33 (1) effective full-power years.

A. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for See ITS 3.4.3 cooldown rates between those presented may be obtained by interpolation.

B. Figures TS 3.1-1 and TS 3.1-2 define limits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g.,

pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

C. The isothermal curve in Figure TS 3.1-2 defines limits to assure prevention of non-ductile failure applicable to low temperature overpressurization events only.

LA01 Application of this curve is limited to evaluation of LTOP events whenever one or more of the RCS cold leg temperatures are less than or equal to the LTOP enabling temperature of 200°F.

2. The secondary side of the steam generator must not be pressurized > 200 psig if the See CTS temperature of the steam generator is < 70°F. 3.1.b.2

3. The pressurizer cooldown and heatup rates shall not exceed 200°F/hr and 100°F/hr, See CTS respectively. The spray shall not be used if the temperature difference between the 3.1.b.3 pressurizer and the spray fluid is > 320°F.

4. The overpressure protection system for low temperature operation shall be LCO 3.4.12 OPERABLE whenever one or more of the RCS cold leg temperatures are 200°F, A02 Applicability and the reactor vessel head is installed. The system shall be considered LCO 3.4.12 OPERABLE when at least one of the following conditions is satisfied:

A. The overpressure relief valve on the Residual Heat Removal System LA02 LCO 3.4.12.a (RHR 33-1) shall have a set pressure of 500 psig and shall be aligned to the RCS by maintaining valves RHR 1A, 1B, 2A, and 2B open.

two RHR suction flow paths OPERABLE

1. With one flow path inoperable, the valves in the parallel flow path shall be ACTION A verified open with the associated motor breakers for the valves locked in the off position. Restore the inoperable flow path within five days or complete ACTION B depressurization and venting of the RCS through a 6.4 square inch vent within an additional eight hours.

RHR System LTOP overpressure relief valve LA02

2. With both flow paths or RHR 33-1 inoperable, complete depressurization ACTION B and venting of the RCS through at least a 6.4 square inch vent pathway within eight hours.

(1)

The curves are limited to 31.1 EFPY due to changes in vessel fluence associated with See ITS 3.4.3 operation at uprated power.

Amendment No. 168 TS 3.1-6 07/08/2003 Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 243 of 478

Attachment 1, Volume 9, Rev. 1, Page 244 of 478 ITS A01 ITS 3.4.12 LCO 3.4.12.b B. A vent pathway shall be provided with an effective flow cross section 6.4 square inches.

Add proposed SR 3.4.12.1 M01

1. When low temperature overpressure protection is provided via a vent pathway, verify the vent pathway at least once per 31 days when the SR 3.4.12.2 pathway is provided by a valve(s) that is locked, sealed, or otherwise secured in the open position. If the vent path is provided by any other means, then verify the vent pathway every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Maximum Coolant Activity

1. The specific activity of the reactor coolant shall be limited to:

A. 1.0 µCi/gram DOSE EQUIVALENT I-131, and 91 Ci B. gross radioactivity due to nuclides with half-lives > 30 minutes E cc excluding tritium ( E is the average sum of the beta and gamma energies in Mev per disintegration) whenever the reactor is critical or the average coolant temperature is > 500°F.

2. If the reactor is critical or the average temperature is > 500°F:

A. With the specific activity of the reactor coolant > 1.0 µCi/gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, See ITS or exceeding 20 µCi/gram DOSE EQUIVALENT I-131, be in at least 3.4.16 INTERMEDIATE SHUTDOWN with an average coolant temperature of < 500°F within six hours.

91 Ci B. With the specific activity of the reactor coolant > of gross radioactivity, E cc be in at least INTERMEDIATE SHUTDOWN with an average coolant temperature < 500°F within six hours.

C. With the specific activity of the reactor coolant > 1.0 µCi/gram DOSE 91 Ci EQUIVALENT I-131 o r > perform the sample and analysis requirements E cc of Table TS 4.1-2, item 1.f, once every four hours until restored to within its limits.

3. Annual reporting requirements are identified in TS 6.9.a.2.D.

Amendment No. 190 TS 3.1-7 03/08/2007 Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 244 of 478

Attachment 1, Volume 9, Rev. 1, Page 245 of 478 ITS ITS 3.4.12 A01 3.1 REACTOR COOLANT SYSTEM APPLICABILITY Applies to the OPERATING status of the Reactor Coolant System (RCS).

OBJECTIVE To specify those LIMITING CONDITIONS FOR OPERATION of the Reactor Coolant System which must be met to ensure safe reactor operation.

SPECIFICATIONS

a. Operational Components See

1. Reactor Coolant Pumps ITS 3.4.5, ITS 3.4.6, ITS 3.4.7, A. At least one reactor coolant pump or one residual heat removal pump shall be in ITS 3.4.8, operation when a reduction is made in the boron concentration of the reactor ITS 3.9.3, and ITS coolant. 3.9.4 See ITS B. When the reactor is in the OPERATING mode, except for low power tests, both 3.4.4 reactor coolant pumps shall be in operation.

C. A reactor coolant pump shall not be started with one or more of the RCS cold leg LCO 3.4.12 temperatures 200°F unless the secondary water temperature of each steam Note generator is < 100°F above each of the RCS cold leg temperatures.

2. Decay Heat Removal Capability A. At least two of the following four heat sinks shall be OPERABLE whenever the average reactor coolant temperature is 350°F but > 200°F.

1. Steam Generator 1A

2. Steam Generator 1B See ITS

3. Residual Heat Removal Train A 3.4.6

4. Residual Heat Removal Train B If less than the above number of required heat sinks are OPERABLE, then corrective action shall be taken immediately to restore the minimum number to the OPERABLE status.

Amendment No. 165 TS 3.1-1 03/11/2003 Page 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 245 of 478

Attachment 1, Volume 9, Rev. 1, Page 246 of 478 DISCUSSION OF CHANGES ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM 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.b.4 states, in part, that the overpressure protection system for low temperature overprotection operation shall be OPERABLE whenever one or more of the RCS cold leg temperatures are 200ºF and the reactor vessel head is installed. ITS 3.4.12 requires the low temperature overpressure protection (LTOP) System to be OPERABLE in MODE 5 and MODE 6 whenever the reactor vessel head is on. This changes the CTS by clearly stating the MODES of Applicability, consistent with the ITS terminology.

ITS MODE 5 and MODE 6 whenever the reactor vessel head is on are the MODES equivalent to the CTS requirements for the LTOP System. This is shown in ITS Table 1.1-1, which defines that MODE 5 has a minimum average reactor coolant temperature of 200°F. In addition, since the 200°F requirement is part of the ITS Table 1.1-1 requirements for MODE 5, there is no reason to include the temperature value in the ITS 3.4.12 Applicability.

Furthermore, ITS MODE 6 is shown on ITS Table 1.1-1 as having one or more reactor vessel head closure bolts less than fully tensioned. Since it possible to have all of the reactor vessel head closure bolts less than fully tensioned and the vessel head removed, the ITS Applicability in MODE 6 for LTOP is changed to clearly state that it is applicable only when the reactor vessel head is on. These changes are designated as administrative changes because they do not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.b.4.A states, in part, that the overpressure relief valve shall be aligned to the RCS by maintaining valves RHR 1A, 1B, 2A, and 2B open. However, the CTS does not provide any Surveillance Requirement to periodically verify this alignment. ITS SR 3.4.12.1 requires verification that the RHR suction valves are open for each RHR suction flow path every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by requiring verification that the RHR suction valves are open for each RHR suction flow path every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The purpose of SR 3.4.12.1 is to verify the required pathways are open by performing a verification that the RHR suction valves are open every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

This change is designated as more restrictive because a Surveillance has been added to the ITS that is not required by the CTS.

Kewaunee Power Station Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 246 of 478

Attachment 1, Volume 9, Rev. 1, Page 247 of 478 DISCUSSION OF CHANGES ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.1.b.1.C states that "the isothermal curve in Figure TS 3.1-2 defines limits to assure prevention of non-ductile failure applicable to low temperature overpressurization events only. Application of this curve is limited to evaluation of LTOP events whenever one or more RCS cold leg temperatures are less than or equal to the LTOP enabling temperature of 200ºF. ITS 3.4.12 does not contain this information. This changes the CTS by moving this information to the Bases.

The removal of these details, which are related to design, from the Technical Specifications 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. Furthermore, ITS Figure 3.4.1-2 states in the title that it is applicable for LTOP events. ITS 3.4.12 requires the LTOP System to be OPERABLE. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. This change is designated as a less restrictive removal of detail change because 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 3.1.b.4.A states, in part, that the overpressure relief valve on the Residual Heat Removal System (RHR 33-1) shall have a set pressure of 500 psig and shall be aligned to the RCS by maintaining valves RHR 1A, 1B, 2A, and 2B open. CTS 3.1.b.4.A.2 also identifies the RHR System overpressure relief valve number. ITS LCO 3.4.12.a states that an LTOP System shall be OPERABLE with the RHR System LTOP overpressure relief valve with a lift setting 500 psig and two RHR suction flow paths OPERABLE. Furthermore, the ITS Actions use a similar description for the associated valves. This changes the CTS by moving the details of the overpressure relief valve number and the RHR suction valve numbers to the Bases.

The removal of these details which are related to system design from the Technical Specifications is acceptable because this type of information is not necessary to be included to provide protection of public health and safety. The ITS still requires that the LTOP overpressure relief valve have a lift setting 500 psig and requires the RHR suction valves to be open by requiring two RHR suction flow paths to be OPERABLE. Also, this change is acceptable because the removed information will be adequately controlled by the Technical Specification 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 less restrictive because information relating to system design is being removed from the Technical Specifications.

Kewaunee Power Station Page 2 of 3 Attachment 1, Volume 9, Rev. 1, Page 247 of 478

Attachment 1, Volume 9, Rev. 1, Page 248 of 478 DISCUSSION OF CHANGES ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 248 of 478

Attachment 1, Volume 9, Rev. 1, Page 249 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 249 of 478

Attachment 1, Volume 9, Rev. 1, Page 250 of 478 CTS All changes are 1 LTOP System unless otherwise noted 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) System 3.1.b.4 LCO 3.4.12 An LTOP System shall be OPERABLE with a maximum of [one] [high pressure injection (HPI)] pump [and one charging pump] capable of injecting into the RCS and the accumulators isolated and one of the following pressure relief capabilities:

a. Two power operated relief valves (PORVs) with lift settings within the limits specified in the PTLR, The System LTOP overpressure a lift setting 3.1.b.4.A a [ b. Two residual heat removal (RHR) suction relief valves with setpoints 3 500 [436.5] psig and [463.5] psig, ] two RHR suction flow paths OPERABLE;

[ c. One PORV with a lift setting within the limits specified in the PTLR and one RHR suction relief valve with a setpoint [436.5] psig and

[463.5] psig, ] or 6.4 2

3.1.b.4.B b d. The RCS depressurized and an RCS vent of [2.07] square inches.

NOTES-------------------------------------------

1. [Two charging pumps] may be made capable of injecting for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for pump swap operations.

INSERT 1

2. Accumulator may be unisolated when accumulator pressure is less than the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR.

3.1.b.4. APPLICABILITY: MODE 4 when any RCS cold leg temperature is [275°F] [LTOP arming temperature specified in the PTLR],

MODE 5, MODE 6 when the reactor vessel head is on.

WOG STS 3.4.12-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 250 of 478

Attachment 1, Volume 9, Rev. 1, Page 251 of 478 CTS 3.4.12 o 1 INSERT 1 3.1.a.1.C A reactor coolant pump shall not be started with one or more RCS cold leg temperatures 200ºF unless the secondary water temperature of each steam generator is < 100 ºF above each of the RCS cold leg temperatures.

Insert Page 3.4.12-1 Attachment 1, Volume 9, Rev. 1, Page 251 of 478

Attachment 1, Volume 9, Rev. 1, Page 252 of 478 CTS All changes are 1 LTOP System unless otherwise noted 3.4.12 ACTIONS

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

LCO 3.0.4.b is not applicable when entering MODE 4.

CONDITION REQUIRED ACTION COMPLETION TIME A. Two or more [HPI] A.1 Initiate action to verify a Immediately pumps capable of maximum of [one] [HPI]

injecting into the RCS. pump is capable of injecting into the RCS.

B. [ Two or more charging B.1 Initiate action to verify a Immediately ]

pumps capable of maximum of [one] charging injecting into the RCS. pump is capable of injecting into the RCS.

C. An accumulator not C.1 Isolate affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolated when the accumulator.

accumulator pressure is greater than or equal to the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

D. Required Action and D.1 Increase RCS cold leg 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion temperature to > [275°F]

Time of Condition [C] not [LTOP arming temperature met. specified in the PTLR].

OR D.2 Depressurize affected 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months accumulator to less than the maximum RCS pressure for existing cold leg temperature allowed in the PTLR.

E. One required RCS relief E.1 Restore required RCS relief 7 days valve inoperable in valve to OPERABLE status.

MODE 4.

WOG STS 3.4.12-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 252 of 478

Attachment 1, Volume 9, Rev. 1, Page 253 of 478 CTS All changes are 1 LTOP System unless otherwise noted 3.4.12 ACTIONS ACTINS (continued) 3 CONDITION REQUIRED ACTION COMPLETION TIME RHR suction flow path INSERT 2 3.1.b.4.A.1 F. One required RCS relief F.1 Restore required RCS relief 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months A valve inoperable in A.2 valve to OPERABLE status. 5 days MODE 5 or 6. RHR suction flow path 3.1.b.4.A.1, 3.1.b.4.A.2 G. Two required RCS relief G.1 Depressurize RCS and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B

valves inoperable. B establish RCS vent of 8 2

RHR suction flow paths [2.07] square inches.

OR 6.4 Required Action and associated Completion Time of Condition A, [B,]

D, E, or F not met.

RHR System LTOP overpressure relief OR valve inoperable.

LTOP System inoperable for any reason other than Condition A, [B,] C, D, E, or F.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of [one] [HPI] pump is capable of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months injecting into the RCS.

SR 3.4.12.2 [ Verify a maximum of one charging pump is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ]

capable of injecting into the RCS.

SR 3.4.12.3 Verify each accumulator is isolated. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months DOC M01 SR 3.4.12.4 [ Verify RHR suction valve is open for each required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ]

1 RHR suction relief valve. s are flow path WOG STS 3.4.12-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 253 of 478

Attachment 1, Volume 9, Rev. 1, Page 254 of 478 CTS 3.4.12 1

INSERT 2 3.1.b.4.A.1 A.1 Verify suction valves in the Immediately other RHR suction flow path are locked open with the motive power removed.

AND Insert Page 3.4.12-3 Attachment 1, Volume 9, Rev. 1, Page 254 of 478

Attachment 1, Volume 9, Rev. 1, Page 255 of 478 CTS All changes are 1 LTOP System unless otherwise noted 3.4.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 3.1.b.4.B.1 3 2 SR 3.4.12.5 Verify required RCS vent [2.07] square inches 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for 2 open. 6.4 is unlocked open vent valve(s)

AND 31 days for other vent path(s)

SR 3.4.12.6 Verify PORV block valve is open for each required 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months PORV.

SR 3.4.12.7 [ Verify associated RHR suction isolation valve is 31 days ]

locked open with operator power removed for each required RHR suction relief valve.

SR 3.4.12.8 -------------------------------NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperature to [275°F]

[LTOP arming temperature specified in the PTLR].

Perform a COT on each required PORV, excluding 31 days actuation.

SR 3.4.12.9 Perform CHANNEL CALIBRATION for each [18] months required PORV actuation channel.

WOG STS 3.4.12-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 255 of 478

Attachment 1, Volume 9, Rev. 1, Page 256 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.12, LOW PRESSURE OVERPRESSURE PROTECTION (LTOP) SYSTEM

1. ISTS 3.4.12 has been changed to be consistent with the Kewaunee Power Station (KPS) current licensing basis. The KPS low temperature overpressure protection is provided by a single LTOP overpressure relief valve in the RHR System or a vent pathway with an effective flow cross section of 6.4 square inches. In order for the LTOP overpressure relief System to be OPERABLE, the RHR System LTOP overpressure relief valve must have a lift setting 500 psig and be aligned to the RCS with the suction valves on the RHR lines open (i.e.,

two RHR suction flow paths OPERABLE). This current licensing basis was approved in Amendment 108 (ADAMS Accession No. ML020770581) and revised in Amendment 144 (ADAMS Accession No. ML020770334). Additionally, the following changes have also been made:

a. Proper plant specific information/values/nomenclature have been provided.

b. KPS has elected not to adopt a Pressure Temperature Limits Report (PTLR),

so all references to a PTLR have not been adopted.

c. The KPS LTOP analysis shows that LTOP is required when the RCS temperature is < 200ºF. The ISTS includes an Applicability of MODE 4 when any RCS cold leg temperature is [275ºF] [LTOP arming temperature specified in the PTLR]. However, since the ITS MODE 4 average reactor coolant temperature is > 200ºF and < 350ºF, the KPS LTOP requirements are not Applicable in MODE 4. Therefore, the MODE 4 Applicability has not been included in the KPS ITS.

d. The limitations of the reactor coolant pump startups have been added to the ITS LCO 3.4.12 as a Note, consistent with the requirements of the CTS.

These limits are also in ITS 3.4.7 as Note 3, but it is actually an LTOP limitation, not RCS loop limitations, thus they are more appropriate to be included in this Specification. Furthermore, a similar change to this was approved at DC Cook in response to question 200409081618.

Based on the above changes, the ISTS LCO, Applicability, ACTIONS and SRs have been modified accordingly.

2. The ITS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and 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. Typographical error corrected.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 256 of 478

Attachment 1, Volume 9, Rev. 1, Page 257 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 257 of 478

Attachment 1, Volume 9, Rev. 1, Page 258 of 478 All changes are 1 LTOP System unless otherwise noted B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 Low Temperature Overpressure Protection (LTOP) System BASES BACKGROUND The LTOP System controls RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting RCPB component for demonstrating such protection. The PTLR provides the LCO 3.4.3, "RCS Pressure and Temperature (P/T) maximum allowable actuation logic setpoints for the power operated relief Limits," provides valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the Reference 1 requirements during the LTOP MODES.

INSERT 1 2 The reactor vessel material is less tough at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2). RCS pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased.

The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only while shutdown; a pressure fluctuation can occur more quickly than an operator can react to relieve the condition.

Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel. LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the PTLR limits.

P/T This LCO provides RCS overpressure protection by having a minimum coolant input capability and having adequate pressure relief capacity.

Limiting coolant input capability requires all but [one] [high pressure injection (HPI)] pump [and one charging pump] incapable of injection into one RHR System LTOP the RCS and isolating the accumulators. The pressure relief capacity overpressure relief valve (RHR 33-1) requires either two redundant RCS relief valves or a depressurized RCS and two RHR suction flow paths and an RCS vent of sufficient size. One RCS relief valve or the open RHR System LTOP RCS vent is the overpressure protection device that acts to terminate an overpressure increasing pressure event.

With minimum coolant input capability, the ability to provide core coolant addition is restricted. The LCO does not require the makeup control 2 system deactivated or the safety injection (SI) actuation circuits blocked.

Due to the lower pressures in the LTOP MODES and the expected core WOG STS B 3.4.12-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 258 of 478

Attachment 1, Volume 9, Rev. 1, Page 259 of 478 B 3.4.12 2

INSERT 1 The isothermal curve (0ºF curve) in Figure 3.4.3-2 defines the limits to assure prevention of non-ductile failure applicable to low temperature overpressurization events only.

Application of this curve is limited to evaluation of LTOP events whenever one or more of the RCS cold leg temperatures are less than or equal to the LTOP enabling temperature of 200ºF.

Insert Page B 3.4.12-1 Attachment 1, Volume 9, Rev. 1, Page 259 of 478

Attachment 1, Volume 9, Rev. 1, Page 260 of 478 LTOP System B 3.4.12 BASES BACKGROUND (continued) decay heat levels, the makeup system can provide adequate flow via the makeup control valve. If conditions require the use of more than one [HPI 2 or] charging pump for makeup in the event of loss of inventory, then pumps can be made available through manual actions.

an RHR System LTOP overpressure relief valve The LTOP System for pressure relief consists of two PORVs with reduced 2 (RHR 33-1) and two RHR lift settings, or two residual heat removal (RHR) suction relief valves, or suction flow paths OPERABLE one PORV and one RHR suction relief valve, or a depressurized RCS An OPERABLE RHR and an RCS vent of sufficient size. Two RCS relief valves are required suction flow path is for redundancy. One RCS relief valve has adequate relieving capability 2 accomplished by maintaining either valves to keep from overpressurization for the required coolant input capability.

RHR 1A and RHR 2A or RHR 1B and RHR 2B open.

PORV Requirements As designed for the LTOP System, each PORV is signaled to open if the RCS pressure approaches a limit determined by the LTOP actuation logic. The LTOP actuation logic monitors both RCS temperature and RCS pressure and determines when a condition not acceptable in the PTLR limits is approached. The wide range RCS temperature indications are auctioneered to select the lowest temperature signal.

The lowest temperature signal is processed through a function generator that calculates a pressure limit for that temperature. The calculated 2 pressure limit is then compared with the indicated RCS pressure from a wide range pressure channel. If the indicated pressure meets or exceeds the calculated value, a PORV is signaled to open.

The PTLR presents the PORV setpoints for LTOP. The setpoints are normally staggered so only one valve opens during a low temperature overpressure transient. Having the setpoints of both valves within the limits in the PTLR ensures that the Reference 1 limits will not be exceeded in any analyzed event.

When a PORV is opened in an increasing pressure transient, the release of coolant will cause the pressure increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close. The pressure continues to decrease below the reset pressure as the valve closes.

WOG STS B 3.4.12-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 260 of 478

Attachment 1, Volume 9, Rev. 1, Page 261 of 478 All changes are 1 LTOP System unless otherwise noted B 3.4.12 BASES BACKGROUND (continued)

System LTOP Overpressure and RHR Suction Flow Path

[ RHR Suction Relief Valve Requirements 3 During LTOP MODES, the RHR System is operated for decay heat removal and low pressure letdown control. Therefore, the RHR suction isolation valves are open in the piping from the RCS hot legs to the inlets System LTOP of the RHR pumps. While these valves are open and the RHR suction is overpressure valves are open, the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS.

is The RHR suction isolation valves and the RHR suction valves must be 5 open to make the RHR suction relief valves OPERABLE for RCS overpressure mitigation. Autoclosure interlocks are not permitted to 2 cause the RHR suction isolation valves to close. The RHR suction relief is a valves are spring loaded, bellows type water relief valves with pressure System LTOP overpressure tolerances and accumulation limits established by Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 3) for Class 2 relief valves. ] 3 RCS Vent Requirements Once the RCS is depressurized, a vent exposed to the containment or near atmosphere will maintain the RCS at containment ambient pressure in an 2 RCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass or heat input transient, and maintaining pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.

removal of For an RCS vent to meet the flow capacity requirement, it requires or steam generator manway removing a pressurizer safety valve, removing a PORV's internals, and 2 disabling its block valve in the open position, or similarly establishing a that has an effective flow cross vent by opening an RCS vent valve. The vent path(s) must be above the section > 6.4 square inches level of reactor coolant, so as not to drain the RCS when open.

APPLICABLE Safety analyses (Ref. 4) demonstrate that the reactor vessel is SAFETY adequately protected against exceeding the Reference 1 P/T limits. In ANALYSES MODES 1, 2, and 3, and in MODE 4 with RCS cold leg temperature exceeding [275°F] [LTOP arming temperature specified in the PTLR], the 200 pressurizer safety valves will prevent RCS pressure from exceeding the 2 9 Reference 1 limits. At about [275°F] [LTOP arming temperature specified in the PTLR] and below, overpressure prevention falls to two OPERABLE the RHR System LTOP overpressure RCS relief valves or to a depressurized RCS and a sufficient sized RCS vent. Each of these means has a limited overpressure relief capability.

WOG STS B 3.4.12-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 261 of 478

Attachment 1, Volume 9, Rev. 1, Page 262 of 478 All changes are 1 LTOP System unless otherwise noted B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued)

RHR System LTOP overpressure relief The actual temperature at which the pressure in the P/T limit curve falls 2

valve below the pressurizer safety valve setpoint increases as the reactor P/T limit vessel material toughness decreases due to neutron embrittlement. Each time the PTLR curves are revised, the LTOP System must be re-evaluated to ensure its functional requirements can still be met using the RHR RCS relief valve method or the depressurized and vented RCS condition.

The PTLR contains the acceptance limits that define the LTOP requirements. Any change to the RCS must be evaluated against the Reference 4 analyses to determine the impact of the change on the LTOP acceptance limits.

energy Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input transients, examples of which follow:

Mass Input Type Transients

a. Inadvertent safety injection or 2

b. Charging/letdown flow mismatch.

INSERT 2 Heat Input Type Transients

a. Inadvertent actuation of pressurizer heaters,

b. Loss of RHR cooling, or

c. Reactor coolant pump (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.

, 2 The following are required during the LTOP MODES to ensure that mass energy and heat input transients do not occur, which either of the LTOP unanalyzed overpressure protection means cannot handle:

a. Rendering all but [one] [HPI] pump [and one charging pump]

incapable of injection, 2 INSERT 3

b. Deactivating the accumulator discharge isolation valves in their closed positions, and

c. Disallowing start of an RCP if secondary temperature is more than

[50]°F above primary temperature in any one loop. LCO 3.4.6, "RCS Loops - MODE 4," and LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled," provide this protection.

WOG STS B 3.4.12-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 262 of 478

Attachment 1, Volume 9, Rev. 1, Page 263 of 478 B 3.4.12 2

INSERT 2

. The mass input transient assumes an inadvertent safety injection (SI) pump start with two RHR and two reactor coolant pumps operating. The energy input transient assumes an initial reactor coolant pump start with steam generator to RCS temperature difference of 100ºF and two RHR pumps operating.

2 INSERT 3 a reactor coolant pump shall not be started with one or more RCS cold leg temperatures 200ºF unless the secondary water temperature of each steam generator is < 100ºF above each RCS cold leg temperature.

Insert Page B 3.4.12-4 Attachment 1, Volume 9, Rev. 1, Page 263 of 478

Attachment 1, Volume 9, Rev. 1, Page 264 of 478 LTOP System B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued)

The Reference 4 analyses demonstrate that either one RCS relief valve or the depressurized RCS and RCS vent can maintain RCS pressure below limits when only one [HPI] pump [and one charging pump are] is

[are] actuated. Thus, the LCO allows only [one] [HPI] pump [and one charging pump] OPERABLE during the LTOP MODES. Since neither one RCS relief valve nor the RCS vent can handle the pressure transient need from accumulator injection, when RCS temperature is low, the LCO also requires the accumulators isolation when accumulator pressure is 2 greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR.

The isolated accumulators must have their discharge valves closed and the valve power supply breakers fixed in their open positions. The analyses show the effect of accumulator discharge is over a narrower ASME, RCS temperature range ([175]°F and below) than that of the Section XI, Appendix G LCO ([275]°F and below).

(Ref. 8)

Fracture mechanics analyses established the temperature of LTOP 2 200 Applicability at [275°F] [LTOP arming temperature specified in the PTLR]. 3 2 (i.e., the greater of The consequences of a small break loss of coolant accident (LOCA) in RTNDT + 50ºF or 200ºF)

LTOP MODE 4 conform to 10 CFR 50.46 and 10 CFR 50, Appendix K 2

(Refs. 5 and 6), requirements by having a maximum of [one] [HPI] pump

[and one charging pump] OPERABLE and SI actuation enabled.

PORV Performance The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the limit shown in the PTLR. The setpoints are derived by analyses that model the performance of the LTOP System, assuming the limiting LTOP transient of [one] [HPI] pump

[and one charging pump] injecting into the RCS. These analyses consider pressure overshoot and undershoot beyond the PORV opening 2 and closing, resulting from signal processing and valve stroke times. The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met.

The PORV setpoints in the PTLR will be updated when the revised P/T limits conflict with the LTOP analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are determined using neutron fluence projections and the results of WOG STS B 3.4.12-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 264 of 478

Attachment 1, Volume 9, Rev. 1, Page 265 of 478 LTOP System B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued) examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," discuss these examinations. 2 The PORVs are considered active components. Thus, the failure of one PORV is assumed to represent the worst case, single active failure.

System LTOP Overpressure

[ RHR Suction Relief Valve Performance System LTOP does overpressure the 2 3 The RHR suction relief valves do not have variable pressure and temperature lift setpoints like the PORVs. Analyses must show that one RHR suction relief valve with a setpoint at or between [436.5] psig and lift setting

< 500 psig

[463.5] psig will pass flow greater than that required for the limiting LTOP transient while maintaining RCS pressure less than the P/T limit curve.

Assuming all relief flow requirements during the limiting LTOP event, an INSERT 3A RHR suction relief valve will maintain RCS pressure to within the valve rated lift setpoint, plus an accumulation 10% of the rated lift setpoint.

the Although each RHR suction relief valve may itself meet single failure criteria, its inclusion and location within the RHR System does not allow it to meet single failure criteria when spurious RHR suction isolation valve closure is postulated. Also, as the RCS P/T limits are decreased to reflect the loss of toughness in the reactor vessel materials due to neutron embrittlement, the RHR suction relief valves must be analyzed to still accommodate the design basis transients for LTOP. transients is an The RHR suction relief valves are considered active components. Thus, the failure of one valve is assumed to represent the worst case single active failure. ]

RCS Vent Performance 6.4 (equivalent to that of the LTOP With the RCS depressurized, analyses show a vent size of 2.07 square overpressure relief valve) inches is capable of mitigating the allowed LTOP overpressure transient.

The capacity of a vent this size is greater than the flow of the limiting 2 s

transient for the LTOP configuration, [one] HPI pump [and one charging pump] OPERABLE, maintaining RCS pressure less than the maximum pressure on the P/T limit curve. INSERT 3B The RCS vent size will be re-evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel material surveillance.

WOG STS B 3.4.12-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 265 of 478

Attachment 1, Volume 9, Rev. 1, Page 266 of 478 B 3.4.12 2 3 INSERT 3A The RHR System LTOP overpressure relief valve set pressure specified includes consideration for the opening setpoint tolerance of +/- 3% (+/- 15 psig) as defined in ASME Boiler and Pressure Vessel Code,Section III, Subsection NC: Class 2 Components for Safety Relief Valves (Ref. 3). The analysis of pressure transient conditions has demonstrated acceptable relieving capability at the upper tolerance limit of 515 psig.

2 INSERT 3B The licensing basis mass input transient assumes an inadvertent SI pump start with two RHR pumps and two reactor coolant pumps (RCPs) operating. The licensing basis energy input transient assumes an initial RCP start with a steam generator to RCS temperature differential of 100ºF and two RHR pumps operating.

Insert Page B 3.4.12-6 Attachment 1, Volume 9, Rev. 1, Page 266 of 478

Attachment 1, Volume 9, Rev. 1, Page 267 of 478 1

All changes are LTOP System unless otherwise noted B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued)

The RCS vent is passive and is not subject to active failure.

The LTOP System satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO requires that the LTOP System is OPERABLE. The LTOP System is OPERABLE when the minimum coolant input and pressure relief capabilities are OPERABLE. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient.

To limit the coolant input capability, the LCO requires that a maximum of

[one] [HPI] pump [and one charging pump] be capable of injecting into the RCS, and all accumulator discharge isolation valves be closed and immobilized (when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR).

The LCO is modified by two Notes. Note 1 allows [two charging pumps]

to be made capable of injecting for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during pump swap operations. One hour provides sufficient time to safely complete the actual transfer and to complete the administrative controls and Surveillance Requirements associated with the swap. The intent is to minimize the actual time that more than [one] charging pump is physically capable of injection. Note 2 states that accumulator isolation is only required when the accumulator pressure is more than or at the maximum RCS pressure for the existing temperature, as allowed by the P/T limit curves. This Note permits the accumulator discharge isolation valve Surveillance to be performed only under these pressure and temperature conditions.

The elements of the LCO that provide low temperature overpressure mitigation through pressure relief are:

a. Two OPERABLE PORVs, A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set to the limit required by the PTLR and testing proves its ability to open at this setpoint, and motive power is available to the two valves and their control circuits.

WOG STS B 3.4.12-7 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 267 of 478

Attachment 1, Volume 9, Rev. 1, Page 268 of 478 All changes are 1 LTOP System B 3.4.12 unless otherwise noted BASES LCO (continued) The System LTOP and two RHR suction overpressure flow paths OPERABLE; a [ b. Two OPERABLE RHR suction relief valves, An RHR suction relief valve is OPERABLE for LTOP when its RHR INSERT 4 suction isolation valve and its RHR suction valve are open, its setpoint is at or between [436.5] psig and [463.5] psig, and testing has proven its ability to open at this setpoint.

c. One OPERABLE PORV and one OPERABLE RHR suction relief valve, or ]

b d. A depressurized RCS and an RCS vent. 6.4 An RCS vent is OPERABLE when open with an area of [2.07] square 3 inches.

Each of these methods of overpressure prevention is capable of mitigating the limiting LTOP transient.

APPLICABILITY This LCO is applicable in MODE 4 when any RCS cold leg temperature is

[275°F] [LTOP arming temperature specified in the PTLR], in MODE 5, and in MODE 6 when the reactor vessel head is on. The pressurizer 200 safety valves provide overpressure protection that meets the Reference 1 P/T limits above [275°F] [LTOP arming temperature specified in the PTLR]. When the reactor vessel head is off, overpressurization cannot occur.

LCO 3.4.3 provides the operational P/T limits for all MODES.

LCO 3.4.10, "Pressurizer Safety Valves," requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1, 2, and 3, and MODE 4 above [275°F] [LTOP arming temperature specified in the PTLR].

Low temperature overpressure prevention is most critical during energy shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event.

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable LTOP System. There is an increased risk associated with entering MODE 4 from MODE 5 with LTOP inoperable and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

WOG STS B 3.4.12-8 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 268 of 478

Attachment 1, Volume 9, Rev. 1, Page 269 of 478 B 3.4.12 2

INSERT 4 The RHR System LTOP overpressure relief valve (RHR 33-1) is required to be OPERABLE for LTOP. In addition, the RHR shall be aligned to the RCS by maintaining both RHR suction flow paths OPERABLE. Valves RHR 1A, RHR 1B, RHR 2A, and RHR 2B must be open for the suction flow paths to be OPERABLE.

Insert Page B 3.4.12-8 Attachment 1, Volume 9, Rev. 1, Page 269 of 478

Attachment 1, Volume 9, Rev. 1, Page 270 of 478 LTOP System All changes are 1 B 3.4.12 unless otherwise noted BASES ACTIONS (continued)

A.1 and [B.1]

With two or more HPI pumps capable of injecting into the RCS, RCS overpressurization is possible.

To immediately initiate action to restore restricted coolant input capability to the RCS reflects the urgency of removing the RCS from this condition.

C.1, D.1, and D.2 An unisolated accumulator requires isolation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This is only required when the accumulator pressure is at or more than the maximum RCS pressure for the existing temperature allowed by the P/T limit curves.

If isolation is needed and cannot be accomplished in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action D.1 and Required Action D.2 provide two options, either of which must be performed in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By increasing the RCS temperature to > [275°F] [LTOP arming temperature specified in the PTLR], an accumulator pressure of [600] psig cannot exceed the LTOP limits if the accumulators are fully injected. Depressurizing the accumulators below the LTOP limit from the PTLR also gives this protection.

The Completion Times are based on operating experience that these activities can be accomplished in these time periods and on engineering evaluations indicating that an event requiring LTOP is not likely in the allowed times.

E.1 In MODE 4 when any RCS cold leg temperature is [275°F] [LTOP arming temperature specified in the PTLR], with one required RCS relief valve inoperable, the RCS relief valve must be restored to OPERABLE status within a Completion Time of 7 days. Two RCS relief valves [in any combination of the PORVS and the RHR suction relief valves] are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.

The Completion Time considers the facts that only one of the RCS relief valves is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low.

WOG STS B 3.4.12-9 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 270 of 478

Attachment 1, Volume 9, Rev. 1, Page 271 of 478 All changes are 1 LTOP System unless otherwise noted B 3.4.12 BASES ACTIONS (continued) and A.2 A F.1 The consequences of operational events that will overpressurize the RCS 5 2

are more severe at lower temperature (Ref. 7). Thus, with one of the two RHR suction flow paths RCS relief valves inoperable in MODE 5 or in MODE 6 with the head on, the Completion Time to restore two valves to OPERABLE status is 5 days 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . the RHR suction flow path INSERT 5 for the inoperable RHR suction flow path The Completion Time represents a reasonable time to investigate and the RHR suction flow path repair several types of relief valve failures without exposure to a lengthy period with only one OPERABLE RCS relief valve to protect against overpressure events. RHR suction flow path 2

and is acceptable as described in Reference 6 B G.1 The RCS must be depressurized and a vent must be established within 8 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when:

RHR suction flow paths

a. Both required RCS relief valves are inoperable, ; 4 The

b. A Required Action and associated Completion Time of Condition A,

[B], D, E, or F is not met, or ; 4 RHR System overpressure relief valve

c. The LTOP System is inoperable for any reason other than Condition A, [B], C, D, E, or F.

6.4 The vent must be sized [2.07] square inches to ensure that the flow 3 capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.

The Completion Time considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of 2

administrative control requirements. The Completion Time was also approved in Amendment 108 (Ref. 6).

SURVEILLANCE SR 3.4.12.1, [SR 3.4.12.2], and SR 3.4.12.3 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, a maximum of [one] [HPI] pump [and a maximum of one charging pump] are verified incapable of injecting into the RCS and the accumulator discharge isolation valves are verified closed and locked out.

WOG STS B 3.4.12-10 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 271 of 478

Attachment 1, Volume 9, Rev. 1, Page 272 of 478 B 3.4.12 o-2 INSERT 5 action must be taken to immediately verify that the suction valves in the other RHR suction flow path are locked open with the motive power removed. Additionally, the inoperable RHR suction flow path must be restored to OPERABLE status.

Insert Page B 3.4.12-10 Attachment 1, Volume 9, Rev. 1, Page 272 of 478

Attachment 1, Volume 9, Rev. 1, Page 273 of 478 All changes are 1 LTOP System unless otherwise noted B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)

The [HPI] pump[s] and charging pump[s] are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control. An alternate method of LTOP control may be employed using at least two independent means to prevent a pump start such that a single failure or single action will not result in an injection into the RCS. This may be accomplished through the pump control switch being placed in [pull to lock] and at least one valve in the discharge flow path being closed.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering other indications and alarms available to the operator in the control room, to verify the required status of the equipment.

[ SR 3.4.12.4 1 3 Each required RHR suction relief valve shall be demonstrated the OPERABLE by verifying its RHR suction valve and RHR suction isolation valves are open and by testing it in accordance with the Inservice Testing is Program. (Refer to SR 3.4.12.7 for the RHR suction isolation valve Surveillance.) This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.

s are The RHR suction valve is verified to be opened every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The Frequency is considered adequate in view of other administrative controls such as valve status indications available to the operator in the control room that verify the RHR suction valve remains open.

7 s The ASME Code (Ref. 8), test per Inservice Testing Program verifies 2 OPERABILITY by proving proper relief valve mechanical motion and by measuring and, if required, adjusting the lift setpoint. ] 3 SR 3.4.12.5 2 6.4 The RCS vent of [2.07] square inches is proven OPERABLE by 3 verifying its open condition either:

a. Once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for a valve that is not locked (valves that are sealed or secured in the open position are considered "locked" in this context) or 4

b. Once every 31 days for other vent path(s) (e.g., a vent valve that is locked, sealed, or secured in position). A removed pressurizer safety valve or open manway also fits this category.

WOG STS B 3.4.12-11 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 273 of 478

Attachment 1, Volume 9, Rev. 1, Page 274 of 478 All changes are 1 LTOP System unless otherwise noted B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)

The passive vent path arrangement must only be open to be OPERABLE.

This Surveillance is required to be met if the vent is being used to satisfy the pressure relief requirements of the LCO 3.4.12d.

.b SR 3.4.12.6 The PORV block valve must be verified open every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to provide the flow path for each required PORV to perform its function when actuated. The valve must be remotely verified open in the main control room. [This Surveillance is performed if the PORV satisfies the LCO.]

The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required removed, and the manual operator is not required locked in the inactive position. Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is considered adequate in view of other administrative controls available to the operator in the control room, such as valve position indication, that verify that the PORV block valve remains open.

[ SR 3.4.12.7 Each required RHR suction relief valve shall be demonstrated OPERABLE by verifying its RHR suction valve and RHR suction isolation valve are open and by testing it in accordance with the Inservice Testing Program. (Refer to SR 3.4.12.4 for the RHR suction valve Surveillance and for a description of the requirements of the Inservice Testing Program.) This Surveillance is only performed if the RHR suction relief valve is being used to satisfy this LCO.]

Every 31 days the RHR suction isolation valve is verified locked open, with power to the valve operator removed, to ensure that accidental closure will not occur. The "locked open" valve must be locally verified in its open position with the manual actuator locked in its inactive position.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve position. ]

WOG STS B 3.4.12-12 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 274 of 478

Attachment 1, Volume 9, Rev. 1, Page 275 of 478 All changes are 1 LTOP System unless otherwise noted B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.12.8 Performance of a COT is required within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS temperature to [275°F] [LTOP arming temperature specified in the PTLR] and every 31 days on each required PORV to verify and, as necessary, adjust its lift setpoint. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable COT of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The COT will verify the setpoint is within the PTLR allowed maximum limits in the PTLR. PORV actuation could depressurize the RCS and is not required.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency considers the unlikelihood of a low temperature overpressure event during this time.

A Note has been added indicating that this SR is required to be performed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperature to

[275°F] [LTOP arming temperature specified in the PTLR]. The COT cannot be performed until in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP setting. The test must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering the LTOP MODES.

SR 3.4.12.9 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required every [18] months to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input.

REFERENCES 1. 10 CFR 50, Appendix G.

2. Generic Letter 88-11.

3. ASME, Boiler and Pressure Vessel Code,Section III.

U Section 9.3.4.3.2 3

4. FSAR, Chapter [15]. 2

5. 10 CFR 50, Section 50.46. 2 WOG STS B 3.4.12-13 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 275 of 478

Attachment 1, Volume 9, Rev. 1, Page 276 of 478 LTOP System B 3.4.12 BASES REFERENCES (continued)

6. 10 CFR 50, Appendix K. 2 5 7. Generic Letter 90-06. 6. Kewaunee Power Station Technical Specification 2

Amendment No. 108 (ADAMS Accession No. ML020770581) 7 8. ASME Code for Operation and Maintenance of Nuclear Power 2 Plants.

2

8. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix G.

2

9. ASME, Boiler and Pressure Vessel Code,Section III, Article NB-7000.

WOG STS B 3.4.12-14 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 276 of 478

Attachment 1, Volume 9, Rev. 1, Page 277 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.12 BASES, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP)

SYSTEM

1. Changes are made to reflect those changes made to the ISTS. The following requirements are renumbered or revised, where applicable to reflect the changes.

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 ITS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The brackets are removed and 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. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

5. The sentence has been deleted since the RHR suction flow paths are not part of RHR relief valve OPERABILITY; they are separate OPERABILITY requirements, as stated in ITS LCO 3.4.12.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 277 of 478

Attachment 1, Volume 9, Rev. 1, Page 278 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 278 of 478

Attachment 1, Volume 9, Rev. 1, Page 279 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.12, LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 279 of 478

, Volume 9, Rev. 1, Page 280 of 478 ATTACHMENT 13 ITS 3.4.13, RCS OPERATIONAL LEAKAGE , Volume 9, Rev. 1, Page 280 of 478

, Volume 9, Rev. 1, Page 281 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 281 of 478

Attachment 1, Volume 9, Rev. 1, Page 282 of 478 ITS ITS 3.4.13 A01

d. RCS Operational LEAKAGE Applicability 1. When the average RCS temperature is > 200°F, RCS operational leakage shall be LCO 3.4.13 limited to:

A. No pressure boundary LEAKAGE, B. 1 gpm unidentified LEAKAGE, C. 10 gpm identified LEAKAGE, and D. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

ACTION A 2. If the limits contained in TS 3.1.d.1 are exceeded for reasons other than pressure boundary LEAKAGE or primary-to-secondary LEAKAGE, then reduce the LEAKAGE to within their limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION B 3. If the limits contained in TS 3.1.d.1 for pressure boundary or primary to secondary LEAKAGE are exceeded, or the time limit contained in TS 3.1.d.2 is exceeded, then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4. When the reactor is critical and above 2% power, two reactor coolant leak detection See ITS systems of different operating principles shall be in operation with one of the two 3.4.15 systems sensitive to radioactivity. Either system may be out of operation for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provided at least one system is OPERABLE.

Amendment No. 188 TS 3.1-8 Revised by letter dated August 29, 2006 Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 282 of 478

Attachment 1, Volume 9, Rev. 1, Page 283 of 478 ITS A01 ITS 3.4.13 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 b. Equipment and sampling tests shall be conducted as specified in Table TS 4.1-2 and 3.4.13.1 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 9, Rev. 1, Page 283 of 478

Attachment 1, Volume 9, Rev. 1, Page 284 of 478 ITS A01 ITS 3.4.13 4.18 RCS Operational LEAKAGE APPLICABILITY Applies to the surveillance requirements for RCS operational LEAKAGE in TS 3.1.d.

OBJECTIVE To assure that the RCS operational LEAKAGE requirements are verified in a sufficient periodicity.

SPECIFICATION SR 3.4.13.1 Note 1: LEAKAGE surveillances are not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after Note 1, SR 3.4.13.2 establishment of steady state operation.

Note SR 3.4.13.1 Note 2: TS 4.18.a is not applicable to primary to secondary LEAKAGE Note 2 SR 3.4.13.1 a. Verify RCS operational LEAKAGE, except for primary to secondary LEAKAGE, is within limits by performance of RCS water inventory balance each 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

SR 3.4.13.2 b. Verify primary to secondary LEAKAGE is 150 gallons per day through any one SG each 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Amendment No. 188 TS 4.18-1 Revised by letter dated August 29, 2006 Page 3 of 3 Attachment 1, Volume 9, Rev. 1, Page 284 of 478

Attachment 1, Volume 9, Rev. 1, Page 285 of 478 DISCUSSION OF CHANGES ITS 3.4.13, RCS OPERATIONAL LEAKAGE 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 None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 285 of 478

Attachment 1, Volume 9, Rev. 1, Page 286 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 286 of 478

Attachment 1, Volume 9, Rev. 1, Page 287 of 478 RCS Operational LEAKAGE CTS 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE 3.1.d.1 LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE, 1

b. 1 gpm unidentified LEAKAGE,

c. 10 gpm identified LEAKAGE, and

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

3.1.d.1 APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.d.2 A. RCS operational A.1 Reduce LEAKAGE to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

3.1.d.3 B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

WOG STS 3.4.13-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 287 of 478

Attachment 1, Volume 9, Rev. 1, Page 288 of 478 CTS RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.18 SR 3.4.13.1 ------------------------------NOTES-----------------------------

Notes 1 1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 2, 4.18.a after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within limits by 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months performance of RCS water inventory balance.

4.18 SR 3.4.13.2 -------------------------------NOTE------------------------------

Note 1, Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 4.18.b establishment of steady state operation.

Verify primary to secondary LEAKAGE is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 150 gallons per day through any one SG.

WOG STS 3.4.13-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 288 of 478

Attachment 1, Volume 9, Rev. 1, Page 289 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.13, RCS OPERATIONAL LEAKAGE

1. The punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 289 of 478

Attachment 1, Volume 9, Rev. 1, Page 290 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 290 of 478

Attachment 1, Volume 9, Rev. 1, Page 291 of 478 RCS Operational LEAKAGE B 3.4.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting INSERT 1 and, to the extent practical, identifying the source of reactor coolant 1 LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

WOG STS B 3.4.13-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 291 of 478

Attachment 1, Volume 9, Rev. 1, Page 292 of 478 B 3.4.13 o1

-INSERT 1 USAR General Design Criteria (GDC) 16 (Ref. 1) states that means shall be provided to detect significant uncontrolled leakage from the reactor coolant pressure boundary. USAR, Section 6.5 (Ref. 2) describes the capabilities of the leakage monitoring indication systems.

Insert Page B 3.4.13-1 Attachment 1, Volume 9, Rev. 1, Page 292 of 478

Attachment 1, Volume 9, Rev. 1, Page 293 of 478 All changes are 1 RCS Operational LEAKAGE B 3.4.13 unless otherwise noted BASES APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY address operational LEAKAGE. However, other operational LEAKAGE ANALYSES is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an event the resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is [1 gallon per minute] or increases to [1 gallon per minute] as a result of accident 150 gallons per day per SG.

induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.

a Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam locked reactor coolant release to the atmosphere, such as a steam generator tube rupture pump rotor, and (SGTR). The leakage contaminates the secondary fluid.

control rod ejection primary to secondary The FSAR (Ref. 3) analysis for SGTR assumes the contaminated INSERT 2 secondary fluid is only briefly released via safety valves and the majority is steamed to the condenser. The [1 gpm] primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential.

less The SLB is more limiting for site radiation releases. The safety analysis 150 gallons for the SLB accident assumes the entire [1 gpm] primary to secondary per day LEAKAGE is through the affected generator as an initial condition. The 50.67 dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits).

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

WOG STS B 3.4.13-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 293 of 478

Attachment 1, Volume 9, Rev. 1, Page 294 of 478 B 3.4.13 1

INSERT 2 The radiological accident analysis (Ref. 3) for SGTR assumes the contaminated secondary fluid is released to the environment from the ruptured and the intact SGs. The release from the ruptured SG occurs until 30 minutes after the reactor trip and the release from the intact SG occurs until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the reactor trip when residual heat removal is placed in service. The 150 gallons per day SG primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential.

Insert Page B 3.4.13-2 Attachment 1, Volume 9, Rev. 1, Page 294 of 478

Attachment 1, Volume 9, Rev. 1, Page 295 of 478 RCS Operational LEAKAGE B 3.4.13 BASES LCO (continued)

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE Through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

WOG STS B 3.4.13-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 295 of 478

Attachment 1, Volume 9, Rev. 1, Page 296 of 478 RCS Operational LEAKAGE B 3.4.13 BASES APPLICABILITY (continued)

LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

ACTIONS A.1 Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

B.1 and B.2 If any pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within limit, or if unidentified or identified LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

WOG STS B 3.4.13-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 296 of 478

Attachment 1, Volume 9, Rev. 1, Page 297 of 478 All changes are 1 RCS Operational LEAKAGE unless otherwise noted B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued)

The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, [and RCP seal injection and return flows]). The Surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation."

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.

SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.20, "Steam Generator Tube 17 Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

WOG STS B 3.4.13-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 297 of 478

Attachment 1, Volume 9, Rev. 1, Page 298 of 478 All changes are 1 RCS Operational LEAKAGE B 3.4.13 unless otherwise noted BASES SURVEILLANCE REQUIREMENTS (continued)

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The Surveillance Frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5).

USAR, Section 4.1.3.2, GDC 16, "Monitoring REFERENCES 1. 10 CFR 50, Appendix A, GDC 30. Reactor Coolant Leakage."

2. Regulatory Guide 1.45, May 1973. USAR, Section 6.5, Leakage Detection and Provisions for the Primary and Auxiliary Coolant Loops

3. FSAR, Section [15]. Westinghouse calculation CN-CRA-99-36, Steam Generator Tube Rupture

4. NEI 97-06, "Steam Generator Program Guidelines."

5. EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

WOG STS B 3.4.13-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 298 of 478

Attachment 1, Volume 9, Rev. 1, Page 299 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.13 BASES, RCS OPERATIONAL LEAKAGE

1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis or licensing basis description. These changes were approved as part of License Amendment 188, dated July 17, 2006 (ADAMS Accession No. ML061700208).

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 299 of 478

Attachment 1, Volume 9, Rev. 1, Page 300 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 300 of 478

Attachment 1, Volume 9, Rev. 1, Page 301 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.13, RCS OPERATIONAL LEAKAGE There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 301 of 478

Attachment 1, Volume 9, Rev. 1, Page 302 of 478 ATTACHMENT 14 ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE Attachment 1, Volume 9, Rev. 1, Page 302 of 478

, Volume 9, Rev. 1, Page 303 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 303 of 478

Attachment 1, Volume 9, Rev. 1, Page 304 of 478 ITS ITS 3.4.14 A01

4. Pressure Isolation Valves LA01 LCO 3.4.14 A. All pressure isolation valves listed in Table TS 3.1-2 shall be functional as a pressure isolation device during OPERATING and HOT STANDBY MODES, M01 Applicability except as specified in 3.1.a.4.B. Valve leakage shall not exceed the amounts LCO 3.4.14 indicated.

Add proposed ACTIONS Notes 1 and 2 A02 ACTION A B. In the event that integrity of any pressure isolation valve as specified in Table TS 3.1-2 cannot be demonstrated, reactor operation may continue, LA01 provided that at least two valves in each high pressure line having a non-functional valve are in, and remain in, the mode corresponding to the isolated condition. (1) L01 6

ACTION B C. If TS 3.1.a.4.A and TS 3.1.a.4.B cannot be met, then an orderly shutdown shall L02 be initiated and the reactor shall be in the HOT SHUTDOWN condition within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the INTERMEDIATE SHUTDOWN condition in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the COLD SHUTDOWN condition within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5. Pressurizer Power-Operated Relief Valves (PORV) and PORV Block Valves A. Two PORVs and their associated block valves shall be OPERABLE during HOT STANDBY and OPERATING modes.

1. With one or both PORVs inoperable because of excessive seat leakage, within one hour either restore the PORV(s) to OPERABLE status or close the associated block valve(s) with power maintained to the block valve(s);

otherwise, action shall be initiated to:

- Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months See ITS

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 3.4.11

2. With one PORV inoperable due to causes other than excessive seat leakage, within one hour either restore the PORV to OPERABLE status or close its associated block valve and remove power from the block valve.

Restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

3. With both PORVs inoperable due to causes other than excessive seat leakage, within one hour either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ACTION A (1)

Manual valves shall be locked in the closed position. Motor operated valves shall be placed L03 in the closed position with their power breakers locked out.

Amendment No. 165 TS 3.1-4 03/11/2003 Page 1 of 5 Attachment 1, Volume 9, Rev. 1, Page 304 of 478

Attachment 1, Volume 9, Rev. 1, Page 305 of 478 ITS TABLE TS 3.1-2 A01 ITS 3.4.14 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES LA01 SYSTEM YSTEM VALVE NO. MAXIMUM(1)(2)

LA02 ALLOWABLE LEAKAGE BASED ON NORMAL OPERATING PRESSURE Reactor Vessel, Core Flooding Line SI-304A 5.0 gallons per minute (Upper Plenum Injection)

SI-303A 5.0 gallons per minute SI-304B 5.0 gallons per minute SI-303B 5.0 gallons per minute Loop B 12" Accumulator Discharge SI-22B 5.0 gallons per minute Line SR 3.4.14.1 (1)

Leakage rates 1.0 gpm are considered acceptable.

Leakage rates > 1.0 gpm but 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

SR 3.4.14.1 Leakage rates > 1.0 gpm but 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

Leakage rates greater than 5.0 gpm are considered unacceptable.

(2)

Minimum test differential pressure shall not be < 150 psid. LA02 Amendment No. 122 PAGE 1 of 1 12/21/95 Page 2 of 5 Attachment 1, Volume 9, Rev. 1, Page 305 of 478

Attachment 1, Volume 9, Rev. 1, Page 306 of 478 ITS A01 ITS 3.4.14 4.2 ASME CODE CLASS IN-SERVICE INSPECTION AND TESTING APPLICABILITY Applies to in-service structural surveillance of the ASME Code Class components and supports and functional testing of pumps and valves.

OBJECTIVE To assure the continued integrity and operational readiness of ASME Code Class 1, 2, 3, and MC components.

SPECIFICATION

a. ASME Code Class 1, 2, 3, and MC Components and Supports

1. In-service inspection of ASME Code Class 1, Class 2, Class 3, and Class MC components and supports shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g), except where relief has been granted by the Commission See ITS pursuant to 10 CFR 50.55a(g)(6)(i). The testing and surveillance of shock 5.5.6 suppressors (snubbers) is detailed in TS 3.14 and TS 4.14.

2. In-service testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants and applicable Addenda as required by 10 CFR 50.55a(f), except where relief has been granted by the Commission pursuant to 10 CFR 50.55a(f)(6)(i).

3. Surveillance testing of pressure isolation valves: LA01 SR 3.4.14.1 a. Periodic leakage testing 1 on each valve listed in Table TS 3.1-2 shall be accomplished prior to entering the OPERATING mode after every time the plant M02 is placed in the COLD SHUTDOWN condition for refueling, after each time the plant is placed in a COLD SHUTDOWN condition for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if testing has not been accomplished in the preceding 9 months, and prior to returning the valve to service after maintenance, repair, or replacement work is performed. L04 (1)

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators if accomplished in accordance with approved procedures LA02 and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

Amendment No. 189 TS 4.2-1 12/14/2006 Page 3 of 5 Attachment 1, Volume 9, Rev. 1, Page 306 of 478

Attachment 1, Volume 9, Rev. 1, Page 307 of 478 ITS A01 ITS 3.4.14

b. Whenever integrity of a pressure isolation valve listed in Table TS 3.1-2 cannot be demonstrated, the integrity of the remaining pressure isolation valve in each high pressure line having a leaking valve shall be determined and recorded daily. L05 In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily.

b. Deleted Amendment No. 188 TS 4.2-2 Revised by letter dated August 29, 2006 Page 4 of 5 Attachment 1, Volume 9, Rev. 1, Page 307 of 478

Attachment 1, Volume 9, Rev. 1, Page 308 of 478 ITS A01 ITS 3.4.14

3. Containment Fancoil Units See ITS Each fancoil unit shall be tested once every operating cycle or once every 18 3.6.6 months, whichever occurs first, to verify proper operation of the motor-operated service water outlet valves and the fancoil emergency discharge and associated backdraft dampers.

b. Component Tests

1. Pumps A. The safety injection pumps, residual heat removal pumps, and containment See ITS spray pumps shall be started and operated quarterly during power operation and 3.5.2 and ITS 3.6.6 within 1 week after the plant is returned to power operation, if the test was not performed during plant shutdown.

B. Acceptable levels of performance are demonstrated by the pumps' ability to start and develop head within an acceptable range.

2. Valves See ITS A. The containment sump outlet valves shall be tested during the pump tests. 3.5.2 B. The accumulator check valves shall be checked for OPERABILITY during each See ITS major REFUELING outage. The accumulator block valves shall be checked to 3.5.1 assure "valve open" requirements during each major REFUELING outage.

C. Deleted See ITS D. Spray additive tank valves shall be tested during each major REFUELING 3.6.7 outage.

E. Deleted Add proposed LCO 3.4.14 part 2 A03 SR 3.4.14.2 F. Residual Heat Removal System valve interlocks shall be tested once per operating cycle.

Add proposed setpoint and method M03 Add proposed ACTION C L06 Amendment No. 137 TS 4.5-2 06/09/98 Page 5 of 5 Attachment 1, Volume 9, Rev. 1, Page 308 of 478

Attachment 1, Volume 9, Rev. 1, Page 309 of 478 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 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.a.4.B and C specify the compensatory actions to take when the leakage through any RCS PIV is greater than the specified limit. ITS 3.4.14 ACTIONS A and B also state the appropriate compensatory actions under the same condition; however, ITS 3.4.14 ACTIONS Notes 1 and 2 have been added. ITS 3.4.14 ACTIONS Note 1 allows separate Condition entry for each RCS PIV flow path.

ITS 3.4.14 ACTIONS Note 2 states "Enter applicable Conditions and Required Actions for systems made inoperable by an inoperable RCS PIV." This changes the CTS by explicitly stating that the Actions are to be taken separately for each inoperable RCS PIV flow path and by explicitly stating that the Conditions and Required Actions for systems made inoperable by an inoperable RCS PIV must be entered.

The purpose of Note 1 is to provide explicit instructions for proper application of the Action for Technical Specification compliance. In conjunction with proposed Specification 1.3, "Completion Times," Note 1 provides direction consistent with the intent of the existing Action for inoperable PIVs. The purpose of Note 2 is to provide explicit instructions for proper application of the ACTION for Technical Specification compliance. Note 2 facilitates the use and understanding of the intent to consider any system affected by inoperable RCS PIVs, which is to have its ACTIONS also apply if it is determined to be inoperable. With the addition of ITS LCO 3.0.6, this intent would not be necessarily applied. This clarification is consistent with the intent and interpretation of the existing Technical Specifications, and is therefore considered an administrative presentation preference. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS 4.5.b.2.F requires testing the RHR System valve interlocks once per operating cycle. In the ITS, this Surveillance has been included as ITS SR 3.4.14.2. In addition, a new LCO has been added which requires the Residual Heat Removal System interlock function to be OPERABLE. This changes the CTS by including the Residual Heat Removal System interlock Surveillance Requirement with the RCS PIV leakage limits and adding a new LCO for the interlock function.

The purpose of CTS 4.5.b.2.F is to ensure the RHR low pressure piping is not overpressurized. This Surveillance is not directly related to the OPERABLITY of the ECCS function of the RHR System. The Operability of the RHR System is affected when this valve is open, not when the interlock is inoperable. Therefore, the transfer of this requirement to the RCS PIV Specification is appropriate. A discussion of a change to the Required Actions when the interlock is found to be Kewaunee Power Station Page 1 of 7 Attachment 1, Volume 9, Rev. 1, Page 309 of 478

Attachment 1, Volume 9, Rev. 1, Page 310 of 478 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE inoperable is discussed in DOC L06. This change is acceptable since the RHR interlock function is retained in the Technical Specifications. This change is designated as administrative because it does not result in technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.a.4.A requires the RCS PIVs to be OPERABLE in the OPERATING (equivalent to ITS MODE 1) and HOT STANDBY (equivalent to ITS MODE 2)

MODES. ITS 3.4.14 requires the RCS PIVs to be OPERABLE in MODES 1, 2, 3, and 4, except valves in the Residual Heat Removal (RHR) System flow path when in, or during the transition to or from, the RHR mode of operation. This changes the CTS by requiring the RCS PIVs to be OPERABLE in MODES 3 and 4 (except valves in the Residual Heat Removal (RHR) System flow path when in, or during the transition to or from, the RHR mode of operation).

The purpose of the leakage limits on the RCS PIVs is to prevent overpressurization of the low pressure side of the connected systems. The addition of the MODES 3 and 4 Applicability is acceptable since the reactor is pressurized in MODES 3 and 4, and excess leakage through the RCS PIVs could result in an overpressure condition in the low pressure portion of the piping. This change is more restrictive because a new Applicability containing MODES 3 and 4.

M02 CTS 4.2.a.3.a requires periodic leakage testing of each RCS PIV prior to entering the OPERATING mode (equivalent to ITS MODE1). ITS SR 3.4.14.1, which performs similar leakage testing, requires the testing to be performed prior to entering MODES 1 and 2 (i.e., the Note to the SR states it is required to be performed in MODES 1 and 2, which means that it has to be current prior to entering those MODES). This changes the CTS by requiring the RCS PIV leakage testing Surveillance to be current prior to entering MODE 2, in lieu of prior to entering MODE 1.

The purpose of the current allowance is to permit leakage testing at high differential pressures with stable conditions that are not possible in the MODES with lower pressures (i.e., MODES 4 and 5). Thus, the proposed Note allows entry into MODES 3 and 4 to establish the necessary differential pressures and stable conditions to allow performance of this Surveillance. This change is acceptable since the high end of MODE 3 provides adequate pressure to satisfactorily perform this Surveillance, thus an allowance to not perform the test until MODE 2 is entered is not required. This change is designated as more restrictive since the test will have to be performed prior to entering MODE 2, in lieu of the current prior to entering MODE 3 allowance.

M03 CTS 4.5.b.2.F requires testing the RHR System valve interlocks once per operating cycle. ITS SR 3.4.14.2 requires a similar test, but includes the specific method for performing the test (using a simulated or actual RCS pressure signal) and the actual pressure value the interlocks must function ( 450 psig). This changes the CTS by adding the specific method and setpoint for the RHR System interlock function Surveillance.

Kewaunee Power Station Page 2 of 7 Attachment 1, Volume 9, Rev. 1, Page 310 of 478

Attachment 1, Volume 9, Rev. 1, Page 311 of 478 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE The purpose of CTS 4.5.b.2.F is to ensure the RHR low pressure piping is not overpressurized. This change is acceptable since the specific pressure value at which the interlock must function is now included in the Surveillance, as well as the manner in which the test is to be performed (use of an actual or simulated RCS pressure signal). This change is more restrictive because new Surveillance acceptance criteria and methods for performing the Surveillance are required in the ITS that are not required in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.1.a.4.A requires the leakage from each RCS PIV specified in Table 3.1-2 to be within the specified limit. CTS 3.1.a.4.B, the specification which provides the action for an inoperable PIV, and CTS 4.2.a.3.a, the Surveillance which checks the RCS PIV leakage, also references Table 3.1-2. CTS Table 3.1-2 contains a list of the RCS PIVs and their associated valve numbers. ITS 3.4.14 does not contain a list of the RCS PIVs or their associated valve numbers. This changes the CTS by relocating the list of RCS PIVs and their associated valve numbers to the Bases.

The removal of these details, which are related to system design, from the Technical Specifications 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. ITS 3.4.14 still requires the RCS PIVs to be OPERABLE, and ITS SR 3.4.14.1 requires periodic Surveillances to determine RCS PIV leakage. It is not necessary for the list of RCS PIVs to be in the Technical Specifications in order to ensure that the RCS PIVs are OPERABLE. Other lists of components, such as containment isolation valves and equipment response time, have been relocated from the Technical Specification to licensee-controlled documents while retaining the requirements on these components in 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 Specification 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 information relating to system design is being removed from the Technical Specifications.

LA02 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS Table 3.1-2 is modified by Note (2). Note (2) describes the minimum test differential pressure at which the RCS PIVs are to be tested. CTS 4.2.a.3 Note (1) explains an alternative method of testing the RCS PIVs to satisfy the ALARA requirements. ITS 3.4.14 does not retain these Notes.

This changes the CTS by relocating the information in the Notes to the Bases.

Kewaunee Power Station Page 3 of 7 Attachment 1, Volume 9, Rev. 1, Page 311 of 478

Attachment 1, Volume 9, Rev. 1, Page 312 of 478 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE The removal of these details for performing Surveillance Requirements from the Technical Specifications 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. ITS 3.4.14 still retains the requirements that RCS PIV leakage must be within the limit and provides the appropriate Surveillance that includes the leakage limit. 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 Specification 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 procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 3 - Relaxation of Completion Time) CTS 3.1.a.4.B requires that if the RCS PIV leakage is not within the limit, operation can continue provided at least two valves in each high pressure line that has a non-functional valve are in and remain in, the mode corresponding to the isolated condition. The term "are in" the mode corresponding to the isolated condition implies that this is an immediate action; no time is provided to place the penetration in the isolated condition. ITS 3.4.14 ACTION A contains this same requirement, but allows 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate the first valve and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to isolate the second valve. This changes the CTS by extending the time requirement to close the first valve from immediately to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and the second valve from immediately to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The purpose of CTS 3.1.a.4.B action is to isolate the pathway. This change is acceptable because the Completion Times are consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The time to close the first valve has been changed from immediately to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and the time to close the second valve has been changed from immediately to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time to close the first valve ensures leakage in excess of the allowable limit is reduced. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time allows time for these actions and restricts the time of operation with leaking valves. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time to close the second valve considers the time required to complete the Required Action and the low probability of the first valve failing during this period. This change is designated as less restrictive because additional time is allowed to isolate the affected penetrations than was allowed in the CTS.

L02 (Category 3 - Relaxation of Completion Time) CTS 3.1.a.4.C, in part, requires that if CTS 3.1.a.4.A and B cannot be met (i.e., an RCS PIV is not within leakage limits and the associated high pressure side of the penetration is not isolated within the required time), the reactor is required to be placed in HOT SHUTDOWN (equivalent to ITS MODE 3) in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and in INTERMEDIATE SHUTDOWN in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Under similar conditions, ITS 3.4.14 Required Kewaunee Power Station Page 4 of 7 Attachment 1, Volume 9, Rev. 1, Page 312 of 478

Attachment 1, Volume 9, Rev. 1, Page 313 of 478 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE Action B.1 requires the unit to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . This changes the CTS by extending the time to be in MODE 3 from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The purpose of CTS 3.1.a.4.C action is to place the unit in a MODE outside the Applicability of the LCO. 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. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. The time to place the unit in MODE 3 has been extended from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Time is reasonable based on operating experience, to reach the required plant condition from full power conditions in an orderly manner and without challenging plant systems.

The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months time is also consistent with the time to reach MODE 3 from full power conditions in other ITS actions. Furthermore MODE 3 is now defined as 350°F.

In the CTS, the temperature break point between HOT SHUTDOWN and INTERMEDIATE SHUTDOWN is 540°F. Thus, by now defining HOT SHUTDOWN (ITS MODE 3) as 350°F, this essentially covers the CTS INTERMEDIATE SHUTDOWN requirement. Thus, it is acceptable to delete the INTERMEDIATE SHUTDOWN requirement. This change is designated as less restrictive because additional time is allowed to reach MODE 3 than was allowed in the CTS.

L03 (Category 4 - Relaxation of Required Action) CTS 3.1.a.4.B includes a Note (Note 1) that describes that the isolated condition for a manual valve includes locking the valve in the closed position and for an motor operated valve includes locking out the power breaker. For a manual valve, ITS 3.4.14 Required Action A.1 only requires closing the manual valve; locking the valve is not required. For an automatic valve, ITS 3.4.14 Required Action A.1 only requires deactivating the valve (i.e., opening the power breaker), locking out the power breaker is not required. This changes the CTS by deleting the requirement to lock the manual valve in the closed position and lock out the motor operated valve power breaker.

The purpose of the CTS action is to ensure the valves are placed in the isolated position and remain in the isolated position. ITS 3.4.14 Required Action A.1 requires manual valves to be closed and automatic valves to be deactivated.

This places the valves in the isolated position. The requirement to lock the valves in the close position or lock out the power breakers is an additional step to help ensure the valves or breakers are not placed back in service. However, this step is not necessary since the Required Actions will continue to require the penetrations remain isolated. Therefore, deletion of these additional locking requirements is acceptable. This change is designated as less restrictive since a Required Action has been deleted.

L04 (Category 5 - Deletion of Surveillance Requirement) CTS 4.2.a.3.a requires testing of RCS PIVs following maintenance, repair, or replacement work on the valve. ITS 3.4.14 does not include this requirement. This changes the CTS by eliminating a post-maintenance Surveillance Requirement.

Kewaunee Power Station Page 5 of 7 Attachment 1, Volume 9, Rev. 1, Page 313 of 478

Attachment 1, Volume 9, Rev. 1, Page 314 of 478 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 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 at a 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 a 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 requirements of 10 CFR 50, Appendix B, Section XI (Test Control), provide 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 Surveillances which are required in the CTS will not be required in the ITS.

L05 (Category 5 - Deletion of Surveillance Requirement) CTS 4.2.a.3.b provides additional compensatory measures to take, above those required by CTS 3.1.a.4.B, when leakage through an RCS PIV is not within the limit. The CTS requires a daily leakage test of the remaining OPERABLE RCS PIV in the flow path. In addition, the position of the second, non-RCS PIV valve is required to be recorded on a daily basis. ITS 3.4.14 does not include these additional compensatory measures. This changes the CTS by deleting the additional compensatory measures taken when leakage through an RCS PIV is not within limit.

The purpose of CTS 4.2.a.3.b is to help ensure that the leakage through the valves used to isolate the penetration with an inoperable RCS PIV is minimized so that an overpressurization event of the downstream piping cannot occur. The change is acceptable since the requirements to ensure the leakage through the two closed valves is within the RCS PIV leakage limit and to ensure closure of the valves are maintained in the ITS. The RCS PIV leakage is ensured prior to using each of the valves as an isolation boundary, as required by the ITS 3.4.14 Required Actions Note. Once leakage is checked, it is not expected to change since the valve cannot be manipulated (ITS 3.4.14 ACTION A requires the valves to be isolated - thus they must remain isolated to comply with the ACTION).

Manipulation of manual valves that have been closed and automatic valves that have been de-activated to comply with Technical Specification Actions is a controlled evolution and the valves are not expected to be inadvertently moved from the isolated condition. Furthermore, these valves will be verified to be in the correct position when first isolated to comply with ITS 3.4.14 ACTION A. This change is designated as less restrictive because a Surveillance required by the CTS will not be required in the ITS.

L06 (Category 4 - Relaxation of Required Action) CTS 4.5.b.2.F requires testing the RHR System valve interlocks once per operating cycle. When the interlock is inoperable, the CTS does not provide any actions to take. ITS 3.4.14 ACTION C has been added which requires the isolation of the penetration by use of one closed manual or deactivated power operated valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Kewaunee Power Station Page 6 of 7 Attachment 1, Volume 9, Rev. 1, Page 314 of 478

Attachment 1, Volume 9, Rev. 1, Page 315 of 478 DISCUSSION OF CHANGES ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE changes the CTS by allowing the penetration to be isolated and to continue operation of the unit for an unlimited amount of time without entry into CTS 3.0.c.

The purpose of ITS 3.4.14 ACTION C is to isolate the penetration to ensure the RHR System is not overpressurized by the RCS. This change is acceptable because the 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 time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the repair period. ITS 3.4.14 ACTION C has been added that requires the isolation of the penetration by use of one closed manual or deactivated power operated valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This change allows the unit to continue to operate and avoids an unnecessary entry into CTS 3.0.c.

Deactivating the power operated valve or closing a manual valve will ensure the function of the interlock is met. Therefore, since the penetration is isolated by closing and deactivating a power operated valve or by closing a manual valve, the function of the interlock is satisfied and this change is acceptable. In addition, the added ACTION avoids an unnecessary reduction in unit power to enter MODE 5. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS.

Kewaunee Power Station Page 7 of 7 Attachment 1, Volume 9, Rev. 1, Page 315 of 478

Attachment 1, Volume 9, Rev. 1, Page 316 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 316 of 478

Attachment 1, Volume 9, Rev. 1, Page 317 of 478 CTS RCS PIV Leakage 3.4.14 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage 3.1.a.4.A LCO 3.4.14 Leakage from each RCS PIV shall be within limit.

1 DOC A03 INSERT 1 3.1.a.4.A APPLICABILITY: MODES 1, 2, and 3, MODE 4, except valves in the residual heat removal (RHR) flow path when in, or during the transition to or from, the RHR mode of operation.

ACTIONS

NOTES----------------------------------------------------------

DOC A02 1. Separate Condition entry is allowed for each flow path.

DOC A02 2. Enter applicable Conditions and Required Actions for systems made inoperable by an inoperable PIV.

11 RCS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.4.B A. One or more flow paths --------------------NOTE-------------------

with leakage from one or Each valve used to satisfy Required more RCS PIVs not Action A.1 and Required Action A.2 within limit. must have been verified to meet .a 2 SR 3.4.14.1 and be in the reactor coolant pressure boundary [or the high pressure portion of the 3 system].

A.1 Isolate the high pressure 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve.

AND WOG STS 3.4.14-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 317 of 478

Attachment 1, Volume 9, Rev. 1, Page 318 of 478 CTS 3.4.14 1 INSERT 1 AND DOC A03 The Residual Heat Removal (RHR) System interlock function shall be OPERABLE.

Insert Page 3.4.14-1 Attachment 1, Volume 9, Rev. 1, Page 318 of 478

Attachment 1, Volume 9, Rev. 1, Page 319 of 478 CTS RCS PIV Leakage 3.4.14 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.4.B 4 A.2 [ Isolate the high pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve.

[or]

Restore RCS PIV to within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ]

4 limits.

3.1.a.4.C, DOC M01 B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months each of DOC L06 C. [ RHR System C.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ]

autoclosure interlock penetration by use of one 5 function inoperable. closed manual or lines deactivated automatic valve.

WOG STS 3.4.14-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 319 of 478

Attachment 1, Volume 9, Rev. 1, Page 320 of 478 CTS RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Only Table 3.1-2 SR 3.4.14.1 ------------------------------NOTES----------------------------- 9 6 including Note 1,

1. Not required to be performed in MODES 3 and 4.2.a.3.a 4. 2 1

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the 7 shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested 8 more than once if a repetitive testing loop cannot be avoided.

5.0 Verify leakage from each RCS PIV is equivalent to In accordance 2

0.5 gpm per nominal inch of valve size up to a with the Inservice a.

maximum of 5 gpm at an RCS pressure Testing Program,

[2215] psig and [2255] psig. and [18] months 3 of 2235 ; and AND INSERT 2 Prior to entering MODE 2 whenever the unit has been in MODE 5 for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 7 days or more, if 8 leakage testing has not been performed in the previous 9 months AND Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to 8 automatic or manual action or flow through the valve WOG STS 3.4.14-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 320 of 478

Attachment 1, Volume 9, Rev. 1, Page 321 of 478 CTS 3.4.14 2

INSERT 2 Table 3.1-2 b. When current measured rate is > 1 gpm, the current Note 1 measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and 5.0 gpm by 50%.

Insert Page 3.4.14-3 Attachment 1, Volume 9, Rev. 1, Page 321 of 478

Attachment 1, Volume 9, Rev. 1, Page 322 of 478 CTS RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 4.5.b.2.F SR 3.4.14.2 -------------------------------NOTE------------------------------ 10

[ Not required to be met when the RHR System autoclosure interlock is disabled in accordance with SR 3.4.12.7.

Verify RHR System autoclosure interlock prevents [18] months ] 5 3 the valves from being opened with a simulated or actual RCS pressure signal [425] psig. 3 450 SR 3.4.14.3 -------------------------------NOTE------------------------------

[ Not required to be met when the RHR System autoclosure interlock is disabled in accordance with SR 3.4.12.7.

5 Verify RHR System autoclosure interlock causes the [18] months ]

valves to close automatically with a simulated or actual RCS pressure signal [600] psig.

WOG STS 3.4.14-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 322 of 478

Attachment 1, Volume 9, Rev. 1, Page 323 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE

1. The second part of the LCO has been added to ensure consistency between the LCO, ACTIONS, and Surveillance Requirements. The ISTS LCO, ACTIONS, and Surveillances do not match up since there is no explicit statement in the LCO requiring the Residual Heat Removal (RHR) System interlock function to be OPERABLE. ISTS LCO 3.0.1 requires LCOs to be met during the MODES or other specified conditions in the Applicability. ISTS LCO 3.0.2 states that upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met. Currently, if the RHR System interlock function is inoperable, the LCO is still met. Thus, ISTS 3.4.14 ACTION C is not required to be entered since the LCO is still met. Therefore, the inclusion of the second portion of the LCO ensures consistency between the LCO, ACTIONS, and Surveillance Requirements. This change is also consistent with a similar change approved for the most recently approved ITS conversion (Davis-Besse).

2. The Kewaunee Power Station (KPS) RCS PIV leakage limits have been provided, consistent with current licensing basis (CTS Table 3.1-2 including Note 1). In addition, since ITS SR 3.4.14.1 includes two limits, only the first limit (a maximum limit) is applicable for the Required Actions A.1 and A.2 Note. Thus, the SR referenced in the ACTION A Note has been changed to "SR 3.4.14.1.a."

3. The brackets have been removed and the proper plant specific information/value has been provided.

4. The first option for ISTS 3.4.14 Required Action A.2 has been maintained in the KPS ITS. This first option (allowing both valves in a penetration to be closed) is consistent with CTS 3.1.a.4.B.

5. The RHR System interlock does not have an autoclosure feature; the interlock only prevents the valves from opening when RCS pressure is 450 psig. In addition, the interlock affects valves on both hot leg suction lines (two valves per line).

Therefore, the following changes have been made: a) the term "autoclosure" has been deleted from Condition C; b) Required Action C.1 has been changed to isolate "each of" the affected lines; c) term "autoclosure" has been deleted from SR 3.4.14.2; and d) ISTS SR 3.4.14.3 has not been included in the KPS ITS. Also, the RHR System interlock function affects a "line," not a penetration (which is generally used to describe requirements on lines that penetrate the containment). Thus the term "penetration" in ISTS 3.4.14 Required Action C.1 has been changed to "line."

6. Editorial changes have been made to be consistent with the Writers Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 4.1.7.g.

7. Note 2 to ISTS SR 3.4.14.1 has been deleted since it is not necessary. The ISTS 3.4.14 Applicability does not require leakage to be met for RHR valves in the flow path when in MODE 4 and when in, or during the transition to or from, the RHR mode of operation.

8. The third Frequency of ISTS SR 3.4.14.1 has been deleted since it is not required by the current licensing basis (as shown in CTS 4.2.a.3.a). The first two Frequencies are adequate to ensure the RCS PIV leakage is within the limit.

Furthermore, the second Frequency has been changed from 7 days to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 323 of 478

Attachment 1, Volume 9, Rev. 1, Page 324 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE also consistent with CTS 4.2.a.3.a. In addition, due to this deletion, Note 3 has also been deleted.

9. Due to the deletion of ISTS SR 3.4.14.1 Notes 2 and 3, the remaining Note has not been numbered and the word "NOTES" has been changed to "NOTE."

10. The ISTS 3.4.14.2 Note has not been included in the KPS ITS since the Low Temperature Overpressure Protection Specification (ITS 3.4.12) is only applicable in MODES 5 and 6. Thus, since ITS 3.4.14 (and hence, ITS SR 3.4.14.2) is not applicable in MODES 5 and 6, this allowance in the Note is not required.

11. The term "RCS" has been added to ACTIONS Note 2 to be consistent with both the LCO statement, Condition A, and SR 3.4.14.1.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 324 of 478

Attachment 1, Volume 9, Rev. 1, Page 325 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 325 of 478

Attachment 1, Volume 9, Rev. 1, Page 326 of 478 RCS PIV Leakage B 3.4.14 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage USAR, Section 1.8, Criterion 51 BASES BACKGROUND 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A 1 discuss reactor coolant pressure (Refs. 1, 2, and 3), define RCS PIVs as any two normally closed valves in boundary valves, which are series within the reactor coolant pressure boundary (RCPB), which that separate the high pressure RCS from an attached low pressure system.

INSERT 1 During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. The RCS PIV Leakage LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety.

The PIV leakage limit applies to each individual valve. Leakage through both series PIVs in a line must be included as part of the identified LEAKAGE, governed by LCO 3.4.13, "RCS Operational LEAKAGE." This is true during operation only when the loss of RCS mass through two series valves is determined by a water inventory balance (SR 3.4.13.1).

A known component of the identified LEAKAGE before operation begins is the least of the two individual leak rates determined for leaking series PIVs during the required surveillance testing; leakage measured through one PIV in a line is not RCS operational LEAKAGE if the other is leaktight.

Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components. Failure consequences could be a loss of coolant accident (LOCA) outside of containment, an unanalyzed accident, that could degrade the ability for low pressure injection.

The basis for this LCO is the 1975 NRC "Reactor Safety Study" (Ref. 4) that identified potential intersystem LOCAs as a significant contributor to 1 the risk of core melt. A subsequent study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs.

PIVs are provided to isolate the RCS from the following typically connected systems: 1

a. Residual Heat Removal (RHR) System, WOG STS B 3.4.14-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 326 of 478

Attachment 1, Volume 9, Rev. 1, Page 327 of 478 B 3.4.14 1

INSERT 1 The 1975 Reactor Safety Study, WASH-1400, (Ref. 4) identified intersystem loss of coolant accidents (LOCAs) as a significant contributor to the risk of core melt. The study considered designs containing two in-series check valves and two check valves in series with a motor operated valve that isolated the high pressure RCS from the low pressure safety injection system. The scenario considered is a failure of the two check valves leading to overpressurization and rupture of the low pressure injection piping which results in a LOCA that bypasses containment. A letter was issued (Ref. 5) by the NRC requiring plants to describe the PIV configuration of the plant. On April 20, 1981, the NRC issued an Order modifying the Kewaunee Power Station Technical Specifications to include testing requirements on PIVs and to specify the PIVs to be tested (Ref. 6).

Insert Page B 3.4.14-1 Attachment 1, Volume 9, Rev. 1, Page 327 of 478

Attachment 1, Volume 9, Rev. 1, Page 328 of 478 RCS PIV Leakage B 3.4.14 BASES BACKGROUND (continued)

b. Safety Injection System, and SI-22B, SI-303A, SI-303B, SI-304A, and SI-304B.

1

c. Chemical and Volume Control System.

The PIVs are listed in the FSAR, Section [ ] (Ref. 6).

Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system and the loss of 2

the integrity of a fission product barrier. INSERT 2 APPLICABLE Reference 4 identified potential intersystem LOCAs as a significant SAFETY contributor to the risk of core melt. The dominant accident sequence in ANALYSES the intersystem LOCA category is the failure of the low pressure portion of the RHR System outside of containment. The accident is the result of a postulated failure of the PIVs, which are part of the RCPB, and the subsequent pressurization of the RHR System downstream of the PIVs not from the RCS. Because the low pressure portion of the RHR System is typically designed for 600 psig, overpressurization failure of the RHR low 1 pressure line would result in a LOCA outside containment and to handle normal subsequent risk of core melt. RCS pressures Reference 5 evaluated various PIV configurations, leakage testing of the valves, and operational changes to determine the effect on the probability of intersystem LOCAs. This study concluded that periodic leakage testing 1 of the PIVs can substantially reduce the probability of an intersystem LOCA.

9 RCS PIV leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO RCS PIV leakage is identified LEAKAGE into closed systems connected to the RCS. Isolation valve leakage is usually on the order of drops per minute. Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken. INSERT 3 The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm. The previous criterion of 1 gpm for all valve sizes imposed an unjustified penalty on the larger valves without 3 providing information on potential valve degradation and resulted in higher personnel radiation exposures. A study concluded a leakage rate limit based on valve size was superior to a single allowable value.

WOG STS B 3.4.14-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 328 of 478

Attachment 1, Volume 9, Rev. 1, Page 329 of 478 B 3.4.14 1

INSERT 2 Two motor operated valves (which are not PIVs) are included in series in each hot leg suction line of the RHR System to isolate the high pressure RCS from the low pressure piping of the RHR System when the RCS pressure is above the design pressure of the RHR System piping and components. Ensuring the RHR System interlock function that prevents the valves from being opened is OPERABLE ensures that RCS pressure will not pressurize the RHR System beyond its design pressure at the pump discharge.

3 INSERT 3 5.0 gpm. However, when the current measured rate is > 1.0 gpm, the current measured rate shall not exceed the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate (5.0 gpm) by 50%.

Insert Page B 3.4.14-2 Attachment 1, Volume 9, Rev. 1, Page 329 of 478

Attachment 1, Volume 9, Rev. 1, Page 330 of 478 RCS PIV Leakage B 3.4.14 BASES LCO (continued)

Reference 7 permits leakage testing at a lower pressure differential than between the specified maximum RCS pressure and the normal pressure of the connected system during RCS operation (the maximum pressure differential) in those types of valves in which the higher service pressure will tend to diminish the overall leakage channel opening. In such cases, the observed rate may be adjusted to the maximum pressure differential INSERT 4 by assuming leakage is directly proportional to the pressure differential to 2 the one half power.

the RCS PIV APPLICABILITY In MODES 1, 2, 3, and 4, this LCO applies because the PIV leakage 4

potential is greatest when the RCS is pressurized. In MODE 4, valves in the RHR flow path are not required to meet the requirements of this LCO when in, or during the transition to or from, the RHR mode of operation.

In MODES 5 and 6, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment.

ACTIONS The Actions are modified by two Notes. Note 1 provides clarification that each flow path allows separate entry into a Condition. This is allowed based upon the functional independence of the flow path. Note 2 requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system operability, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function.

A.1 and A.2 The flow path must be isolated by two valves. Required Actions A.1 5 If the leakage from one and A.2 are modified by a Note that the valves used for isolation must or more RCS PIVs is not within limits, the meet the same leakage requirements as the PIVs and must be within the RCPB [or the high pressure portion of the system]. 6 Required Action A.1 requires that the isolation with one valve must be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the affected system if leakage cannot be reduced. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time allows the actions and restricts the operation with leaking isolation valves.

WOG STS B 3.4.14-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 330 of 478

Attachment 1, Volume 9, Rev. 1, Page 331 of 478 B 3.4.14 0 INSERT 2

_ 4 Ensuring the RHR System interlock function that prevents the RHR hot leg suction valves (RHR-1A, 1B, 2A, and 2B) from being opened is OPERABLE ensures that RCS pressure will not pressurize the RHR System beyond its design pressure at the pump discharge.

Insert Page B 3.4.14-3 Attachment 1, Volume 9, Rev. 1, Page 331 of 478

Attachment 1, Volume 9, Rev. 1, Page 332 of 478 RCS PIV Leakage B 3.4.14 BASES ACTIONS (continued)

[ Required Action A.2 specifies that the double isolation barrier of two 6 valves be restored by closing some other valve qualified for isolation or restoring one leaking PIV. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time after exceeding the limit considers the time required to complete the Action and the low probability of a second valve failing during this time period.

[or]

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time after exceeding the limit allows for the 3 restoration of the leaking PIV to OPERABLE status. This timeframe considers the time required to complete this Action and the low probability of a second valve failing during this period. ]

REVIEWERS NOTE-----------------------------------

Two options are provided for Required Action A.2. The second option 7 (72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months restoration) is appropriate if isolation of a second valve would place the unit in an unanalyzed condition.

B.1 and B.2 6

If leakage cannot be reduced, [the system can not be isolated,] or the other Required Actions accomplished, 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 MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This Action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The allowed Completion Times are reasonable based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

C.1 function allows hot leg at the pump The inoperability of the RHR autoclosure interlock renders the RHR discharge to be inadvertently suction isolation valves incapable of isolating in response to a high 3

opened pressure condition and preventing inadvertent opening of the valves at function RCS pressures in excess of the RHR systems design pressure. If the hot leg RHR autoclosure interlock is inoperable, operation may continue as long each as the affected RHR suction penetration is closed by at least one closed manual or deactivated automatic valve within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Action isolated 5 accomplishes the purpose of the autoclosure function. line 3

interlock WOG STS B 3.4.14-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 332 of 478

Attachment 1, Volume 9, Rev. 1, Page 333 of 478 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each RCS PIV or isolation valve used to satisfy Required Action A.1 and Required Action A.2 is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up 3 INSERT 5 to 5 gpm maximum applies to each valve. Leakage testing requires a 8

stable pressure condition. INSERT 6 For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

6 Testing is to be performed every [18] months, a typical refueling cycle, if 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months the plant does not go into MODE 5 for at least 7 days. The [18 month] 3 6 Frequency is consistent with 10 CFR 50.55a(g) (Ref. 8) as contained in the Inservice Testing Program, is within frequency allowed by the f 9 American Society of Mechanical Engineers (ASME) Code (Ref. 7), and is based on the need to perform such surveillances under the conditions that apply during an outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating. PIVs disturbed in the performance of this Surveillance should also be tested unless 3 documentation shows that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has been reseated. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and practical time limit for performing this test after opening or reseating a valve.

5 performed The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance. The Note that allows this provision is complementary to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months the Frequency of prior to entry into MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed 3 in the previous 9 months. In addition, this Surveillance is not required to WOG STS B 3.4.14-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 333 of 478

Attachment 1, Volume 9, Rev. 1, Page 334 of 478 B 3.4.14 3 INSERT 5 The RCS PIV leakage limit is 5.0 gpm at 2235 psig. However, RCS PIV leakage is also limited when the current measured rate is > 1.0 gpm, such that the current measured rate shall not exceed the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and 5.0 gpm by 50%.

8 INSERT 6 The test differential pressure across each valve shall be 150 psid. Additionally, to satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

Insert Page B 3.4.14-5 Attachment 1, Volume 9, Rev. 1, Page 334 of 478

Attachment 1, Volume 9, Rev. 1, Page 335 of 478 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued) be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation. PIVs contained in 3

the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established.

[ SR 3.4.14.2 and SR 3.4.14.3 is 6 3

Verifying that the RHR autoclosure interlocks are OPERABLE ensures at the pump that RCS pressure will not pressurize the RHR system beyond 125% of discharge its design pressure of [600] psig. The interlock setpoint that prevents the 6 1 valves from being opened is set so the actual RCS pressure must be 450 < [425] psig to open the valves. This setpoint ensures the RHR design 6 1

pressure will not be exceeded and the RHR relief valves will not lift. The

[18] month Frequency is based on the need to perform the Surveillance 6

under conditions that apply during a plant outage. The [18] month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.

These SRs are modified by Notes allowing the RHR autoclosure function to be disabled when using the RHR System suction relief valves for cold 3 overpressure protection in accordance with SR 3.4.12.7. ]

REFERENCES 1. 10 CFR 50.2.

2. 10 CFR 50.55a(c).

3. 10 CFR 50, Appendix A, Section V, GDC 55.

4. WASH-1400 (NUREG-75/014), Appendix V, October 1975.

5. Letter from D. G. Eisenhut (NRC) to all LWR Licensees, LWR Primary 1 Coolant System Pressure Isolation 5. NUREG-0677, May 1980.

Valves, February 23, 1980.

6. Letter from D. G. Eisenhut (NRC) to [ 6. Document containing list of PIVs. ] 1 Kewaunee Power Station, Order for Modification of License, April 20, 1981.

7. ASME Code for Operation and Maintenance of Nuclear Power Plants.

f 9

8. 10 CFR 50.55a(g).

WOG STS B 3.4.14-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 335 of 478

Attachment 1, Volume 9, Rev. 1, Page 336 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.14 BASES, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE

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. A description of the RHR interlock function has been added, since the Specification also covers this function.

3. Changes are made to reflect changes made to the Specification.

4. The term "RCS PIV" has been added since the LCO covers valves that are not RCS PIVs (i.e., the RHR System interlock affects valves that are not RCS PIVs).

5. Changes made to be consistent with the Specification.

6. The brackets have been removed and the proper plant specific information/value has been provided.

7. The Reviewers Note is deleted because it is not intended to be included in the plant specific ITS submittal.

8. Changes made to be consistent with the Kewaunee Power Station CTS requirements, as described in Discussion of Change LA02.

9. Typographical error corrected.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 336 of 478

Attachment 1, Volume 9, Rev. 1, Page 337 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 337 of 478

Attachment 1, Volume 9, Rev. 1, Page 338 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.14, RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 338 of 478

Attachment 1, Volume 9, Rev. 1, Page 339 of 478 ATTACHMENT 15 ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION Attachment 1, Volume 9, Rev. 1, Page 339 of 478

, Volume 9, Rev. 1, Page 340 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 340 of 478

Attachment 1, Volume 9, Rev. 1, Page 341 of 478 ITS ITS 3.4.15 A01

d. RCS Operational LEAKAGE

1. When the average RCS temperature is > 200°F, RCS operational leakage shall be limited to:

A. No pressure boundary LEAKAGE, B. 1 gpm unidentified LEAKAGE, C. 10 gpm identified LEAKAGE, and D. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG). See ITS 3.4.13

2. If the limits contained in TS 3.1.d.1 are exceeded for reasons other than pressure boundary LEAKAGE or primary-to-secondary LEAKAGE, then reduce the LEAKAGE to within their limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. If the limits contained in TS 3.1.d.1 for pressure boundary or primary to secondary LEAKAGE are exceeded, or the time limit contained in TS 3.1.d.2 is exceeded, then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

M01 Applicability

4. When the reactor is critical and above 2% power, two reactor coolant leak detection LCO 3.4.15 systems of different operating principles shall be in operation with one of the two ACTION A, systems sensitive to radioactivity. Either system may be out of operation for up to ACTION B 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provided at least one system is OPERABLE. L01 L02 Add proposed ACTION C M03 Add proposed ACTION D A02 Amendment No. 188 TS 3.1-8 Revised by letter dated August 29, 2006 Page 1 of 4 Attachment 1, Volume 9, Rev. 1, Page 341 of 478

Attachment 1, Volume 9, Rev. 1, Page 342 of 478 ITS A01 ITS 3.4.15 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 3.4.15.1, SR 3.4.15.2, SR 3.4.15.3,

a. Calibration, testing, and checking of protective instrumentation channels and testing of SR 3.4.15.4 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 2 of 4 Attachment 1, Volume 9, Rev. 1, Page 342 of 478

ITS ITS 3.4.15 A01 TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL CHECK CALIBRATE TEST REMARKS DESCRIPTION

18. a. Containment Each shift Each refueling cycle Monthly(a) (a) Isolation Valve Signal Pressure (SIS signal)

b. Containment Each shift(a) Each refueling cycle(a) Monthly(a) (a) Narrow range containment pressure Pressure See ITS

(-3.0, +3.0 psig excluded) 3.3.2 (Steamline Isolation)

c. Containment Each shift Each refueling cycle Monthly Pressure (Containment Spray Act)

d. Annulus Not applicable Each refueling cycle Each refueling See ITS 3.6.9 Pressure cycle (Vacuum LA01 Breaker)

SR 3.4.15.1,

19. Radiation Monitoring Daily (a,b) A03 Each refueling cycle (a) Quarterly (a) (a) Includes only channels R11 thru R15, R19, SR 3.4.15.2, System R21, and R23 SR 3.4.15.4, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (b) Channel check required in all plant modes A03

20. Deleted

21. Containment Sump Not applicable Not applicable Each refueling M04 SR 3.4.15.3 Level cycle

22. Accumulator Level Each shift Deleted Not applicable See ITS , Volume 9, Rev. 1, Page 343 of 478 Attachment 1, Volume 9, Rev. 1, Page 343 of 478 and Pressure 3.5.1

23. Steam Generator Each shift Each refueling cycle Monthly See ITS Pressure 3.3.2 Discussion of Change LA01 is for channels R11, R12, and R21. For the other channels, see ITS 3.3.2, 3.3.6, 3.3.7, and CTS 3.8.a.9.

Amendment No. 182 Page 4 of 7 4/06/2005 Page 3 of 4

Attachment 1, Volume 9, Rev. 1, Page 344 of 478 A01 ITS 3.4.15 TABLE TS 4.1-3 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS A04 (1)

EQUIPMENT TESTS TEST FREQUENCY

1. Control Rods Rod drop times of all full Each REFUELING outage length rods Quarterly when at or above HOT See ITS Partial movement of all STANDBY 3.1.4 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 3.6.3

4. Containment Isolation Trip OPERABILITY Each REFUELING outage

5. Refueling System Interlocks OPERABILITY Prior to fuel movement each See CTS REFUELING outage 3.8.a.11

6. Deleted

7. Deleted A04

8. RCS Leak Detection OPERABILITY Weekly(4) See ITS 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

12. Guard Pipes Visual Inspection Each REFUELING outage See ITS 3.6.1

13. Pressurizer PORVs OPERABILITY Each REFUELING cycle See ITS (6) 3.4.11

14. Pressurizer PORV Block Valves OPERABILITY Quarterly

15. Pressurizer Heaters OPERABILITY(7) Each REFUELING cycle See ITS 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 A04 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 See ITS 3.3.1 and undervoltage trip attachments. The test shall also verify the undervoltage trip attachment on the reactor trip bypass breakers.

(4) A04 When reactor is at power or in HOT SHUTDOWN condition.

(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 (7) 3.4.11 Test will verify OPERABILITY of heaters and availability of an emergency power supply.

(8)

This test shall demonstrate that the valve(s) close in 5 seconds. See ITS 3.4.9 Amendment No. 125 Page 1 of 1 See ITS 08/07/96 3.6.3 Page 4 of 4 Attachment 1, Volume 9, Rev. 1, Page 344 of 478

Attachment 1, Volume 9, Rev. 1, Page 345 of 478 DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION 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.d.4 does not contain a specific ACTION for all required monitors inoperable. With all required monitors inoperable, CTS 3.0.c would be entered.

ITS 3.4.15 ACTION D directs entry into LCO 3.0.3 when all monitors are inoperable. This changes the CTS by specifically stating to enter LCO 3.0.3, which is the equivalent requirement in the ITS, in this System Specification.

Changes to CTS 3.0.c are discussed in Section 3.0.

This change is acceptable because the actions taken when all monitors are inoperable are unchanged. Adding this ACTION is consistent with the ITS conversion of directing entry into LCO 3.0.3 when multiple ACTIONS are presented in the ITS, and entry into these multiple ACTIONS could result in a loss of safety function. This change is designated as administrative because it does not result in any technical changes to the CTS.

A03 Note (b) to CTS Table TS 4.1-1, Item 19 in the Remarks Section states that the Channel Check is required in all plant modes. ITS SR 3.4.15.1, the equivalent CHANNEL CHECK requirement, is only applicable in MODES 1, 2, 3, and 4, consistent with the Applicability of the LCO. This changes the CTS by deleting this specific Note.

While the purpose of the Note appears to require the Channel Check requirement to be performed in all plant modes, CTS 4.0.a, which provides the general requirements of all Surveillance Requirements, specifically states that Surveillances are only required to be met during the operational MODES or other specified conditions in the LCO. While it does state that this can be modified as stated in an individual Surveillance, it further states that if the Surveillance is not met, then the actions of the LCO are to be taken. However, as shown in CTS 3.1.d.4, since the equipment is only required when critical above 2% RTP, when declared inoperable, the plant only has to get out of this specific Applicability, as stated in CTS 3.0.c. Thus, while this Note appears to require the Channel Check in all plant modes, in actuality, it only is required when the LCO has to be met; i.e., when critical above 2% RTP. Therefore, the deletion of this is acceptable and considered administrative, since the technical requirements have not been changed.

A04 CTS 4.1.b requires the equipment tests to be performed as specified in Table TS 4.1-3. CTS Table TS 4.1-3 Equipment Test 8 requires a weekly OPERABILITY test on the RCS Leakage Detection. The test is modified by Note 1, which states that following maintenance on equipment that could affect the operation of the equipment, tests should be performed to verify Kewaunee Power Station Page 1 of 5 Attachment 1, Volume 9, Rev. 1, Page 345 of 478

Attachment 1, Volume 9, Rev. 1, Page 346 of 478 DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION OPERABILITY. The Weekly Frequency is modified by a Note (Note 4) that states the Frequency is applicable when the reactor is at power or in the HOT SHUTDOWN condition. The ITS does not include this requirement. This changes the CTS by deleting this weekly OPERABILITY check of the RCS Leak Detection Instrumentation.

CTS 3.1.d.4 requires two RCS leak detection systems to be OPERABLE. Thus, this CTS Table 4.1-3 Equipment Test 8 is met by verifying that there are two RCS leak detection instruments OPERABLE. This is the manner in which Kewaunee Power Station is meeting this requirement. Since ITS LCO 3.4.15 continues to require RCS leak detection systems to be OPERABLE, and ACTIONS are provided when the required systems are inoperable, there is no purpose to having a requirement to verify weekly that the required systems are OPERABLE. Operations staff is constantly aware of equipment that has been declared inoperable. Thus, this requirement serves no purpose and is redundant to both the LCO and the applicable ACTIONS. Furthermore, LCO 3.0.2 requires entry into the ACTIONS when the LCO is no longer met. To meet this requirement, Operations staff would have to know at all times whether an LCO is being met, not just on a weekly basis. Therefore, deletion of this redundant check is acceptable. Furthermore, if a component is inoperable, prior to calling the component OPERABLE, KPS would have to determine if any tests are required to be performed as part of returning the component to OPERABLE status. ITS LCO 3.0.1 and SR 3.0.1 would ensure any required tests are performed or are current. Furthermore, Note (4) is unnecessary since it is duplicative of the Applicability statement for the RCS leak detection systems in CTS 3.1.d.4. This change is designated as administrative since it does not result in any technical changes.

MORE RESTRICTIVE CHANGES M01 The CTS Applicability of the RCS leak detection system is when the reactor is critical and above 2% power (equivalent to ITS MODE 1 and part of MODE 2).

ITS 3.4.15 requires the RCS leakage detection instrumentation to be OPERABLE in MODES 1, 2, 3, and 4. This changes the CTS by requiring the RCS leak detection instrumentation to be OPERABLE in more MODES in ITS than in CTS.

Leakage detection systems must have the capability to detect significant reactor coolant pressure boundary degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure. Due to the elevated RCS temperature and pressure conditions in MODES 1, 2, 3, and 4, the RCS leakage detection instrumentation is required to be OPERABLE. Since the RCS pressure and temperatures are much lower outside of MODES 1, 2, 3, and 4, the likelihood of leakage and crack propagation is much smaller. The addition of MODES 2, 3, and 4 is acceptable since the RCS pressure and temperature during these MODES increases the potential for reactor coolant pressure boundary leakage. This change is more restrictive because a new Applicability containing MODES 1, 2, 3, and 4 has been added.

M02 Not used.

Kewaunee Power Station Page 2 of 5 Attachment 1, Volume 9, Rev. 1, Page 346 of 478

Attachment 1, Volume 9, Rev. 1, Page 347 of 478 DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION M03 CTS 3.1.d.4 does not contain any ACTIONS to take if one of the required RCS leakage detection instruments are inoperable and not restored within the allowed Completion Time. As a result, CTS 3.0.c would be entered, which requires action to be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and to be in HOT STANDBY (equivalent to ITS MODE 2) within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ITS 3.4.15 ACTION C states that if the Required Action and associated Completion Time of ACTIONS A or B are not met, then the unit must be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

This changes the CTS by adding a specific shutdown action.

The purpose of CTS LCO 3.0.c is to place the unit outside of the Applicability of the Specification. With the ITS Applicability being MODES 1, 2, 3, and 4, ITS 3.4.15 ACTION C effectively places the unit outside of the Applicability by requiring the unit to be in MODE 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Furthermore, the unit is required to be in MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , which is less than the current time to be in HOT STANDBY in CTS 3.0.c. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. This change is designated as more restrictive because a new proposed ACTION has been added.

M04 Item 19 of CTS Table TS 4.1-1 requires a Daily instrument check of the radiation monitoring system. ITS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment radioactivity monitor every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This changes the CTS by requiring a check of the required containment radioactivity monitor more often in ITS than in CTS.

The purpose of the instrument check is to demonstrate that the required containment radioactivity monitor is OPERABLE and capable of providing an early indication of any abnormal leakage conditions in the containment. ITS SR 3.4.15.1 provides reasonable confidence that the channel is operating properly. This change is designated more restrictive because less time is allowed between performances of the CHANNEL CHECK than was allowed in the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) Note (a) to Item 19 of CTS Table TS 4.1-1 in the Remarks Section states that the instrument CHECK, CALIBRATE, and TEST Frequencies for the Radiation Monitoring System are applicable only to channels R11 thru R15, R19, R21, and R23. For the RCS Leakage Detection Specification, only instruments R11, R12, and R21 apply. ITS 3.4.15 does not contain this note.

This changes the CTS by removing the description of the applicable channels to the Bases.

Kewaunee Power Station Page 3 of 5 Attachment 1, Volume 9, Rev. 1, Page 347 of 478

Attachment 1, Volume 9, Rev. 1, Page 348 of 478 DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION The removal of these details (instruments R11, R12, and R21), which are related to 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. The ITS still retains the requirement to perform the instrument CHANNEL CHECK, CHANNEL CALIBRATION , and the CHANNEL OPERABILITY TEST (COT) for the containment atmosphere radioactivity monitors. Also, 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 Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. The removal of the other Instruments, R13, R14, R15, R19, and R23, will be discussed in other Discussion of Changes. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 4 - Relaxation of Required Action) CTS 3.1.d.4 states that either reactor coolant leak detection system may be out of operation for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provided at least one system is OPERABLE. When the containment sump monitor is inoperable, ITS 3.4.15 ACTION A allows 30 days to restore the required containment sump monitor to OPERABLE status (Required Action A.2).

In addition, ITS 3.4.15 ACTION A also requires the performance of an RCS water inventory balance (i.e., SR 3.4.13.1) once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during this 30 day period (Required Action A.1). This changes the CTS by allowing a longer period of time to restore the required containment sump to OPERABLE in the ITS than was allowed in the CTS.

The purpose of CTS 3.1.4.d is to ensure an approved method of monitoring RCS leakage is always available. The change is acceptable because ITS 3.4.15 ACTION A continues to provide assurance that an alternate and reliable means of detecting RCS leakage is available (since if both methods are inoperable the unit would also be in ITS 3.4.15 ACTION D, which requires a unit shutdown). In addition, the increase in the allowable time in which the required containment sump monitor is inoperable is acceptable since SR 3.4.13.1, which is normally performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , is now performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to provide periodic information that is adequate to detect leakage. The performance of the SR ensures that RCS Leakage is verified to be within limit, on a more frequent basis, thus performing the function of the instrumentation. This change is designated as less restrictive because a less restrictive Completion Time is being applied in the ITS than was applied in the CTS.

L02 (Category 4 - Relaxation of Required Action) CTS 3.1.d.4 states that either reactor coolant leak detection system may be out of operation for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provided at least one system is OPERABLE. When the required containment atmosphere radioactivity monitor is inoperable, ITS 3.4.15 ACTION B requires an analysis of containment atmosphere grab samples be performed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Required Action B.1.1) OR an RCS water inventory balance (i.e., SR 3.4.13.1) be performed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Required Action B.1.2). In addition, Kewaunee Power Station Page 4 of 5 Attachment 1, Volume 9, Rev. 1, Page 348 of 478

Attachment 1, Volume 9, Rev. 1, Page 349 of 478 DISCUSSION OF CHANGES ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION ACTION B requires restoration of the required containment atmosphere radioactivity monitor to OPERABLE status within 30 days (Required Action B.2).

During the 30 days allowed for restoration, Required Actions B.1.1 and B.1.2 will continue to be performed. This changes the CTS by allowing a longer period of time to restore the required containment atmosphere radioactivity monitor to OPERABLE in the ITS than was allowed in the CTS.

The purpose of CTS 3.1.4.d is to ensure an approved method of monitoring RCS leakage is always available. The change is acceptable because ITS 3.4.15 ACTION B continues to provide assurance that an alternate and reliable means of detecting RCS leakage is available (since if both methods are inoperable the unit would also be in ITS 3.4.15 ACTION D, which requires a unit shutdown).

Required Actions B.1.1 and B.1.2 are used to establish remedial measures taken in response to degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The analysis of the grab samples of the containment atmosphere or the check of the RCS water inventory balance (i.e., SR 3.4.13.1) once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides periodic information that is adequate to detect any change in containment conditions that may be attributed to RCS leakage. Thus, it is acceptable to allow continued operation for 30 days with the containment radioactivity monitor inoperable under these conditions. The increase in the allowable time in which the required containment atmosphere radioactivity monitor is inoperable (from 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to 30 days) is acceptable since SR 3.4.13.1, which is normally performed every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , is now performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (per Required Action B.1.2) to provide periodic information that is adequate to detect leakage, or an analysis of a containment atmosphere grab sample is performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (per required Action B.1.1). The analysis of the containment atmosphere grab samples or performance of the SR ensures that RCS Leakage is verified to be within limit, on a more frequent basis, thus performing the function of the instrumentation during the proposed 30 day restoration time. This change is designated as less restrictive because a less restrictive Completion Time is being applied in the ITS than was applied in the CTS.

Kewaunee Power Station Page 5 of 5 Attachment 1, Volume 9, Rev. 1, Page 349 of 478

Attachment 1, Volume 9, Rev. 1, Page 350 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 350 of 478

Attachment 1, Volume 9, Rev. 1, Page 351 of 478 CTS RCS Leakage Detection Instrumentation 3.4.15 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.15 RCS Leakage Detection Instrumentation 3.1.d.4 LCO 3.4.15 The following RCS leakage detection instrumentation shall be OPERABLE: hi or hi-hi

a. One containment sump (level or discharge flow) monitor,

and 4 1

b. One containment atmosphere radioactivity monitor (gaseous or particulate), and

[ c. One containment air cooler condensate flow rate monitor. ] 2 3.1.d.4 APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.d.4 A. Required containment A.1 --------------NOTE--------------

sump monitor Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable. after establishment of steady state operation.

Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND A.2 Restore required 30 days containment sump monitor to OPERABLE status.

WOG STS 3.4.15-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 351 of 478

Attachment 1, Volume 9, Rev. 1, Page 352 of 478 CTS RCS Leakage Detection Instrumentation 3.4.15 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.1.d.4 B. Required containment B.1.1 Analyze grab samples of Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months atmosphere radioactivity the containment monitor inoperable. atmosphere.

OR B.1.2 --------------NOTE--------------

Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

[ AND B.2.1 Restore required 30 days containment atmosphere radioactivity monitor to OPERABLE status.

OR 2

B.2.2 Verify containment air 30 days ]

cooler condensate flow rate monitor is OPERABLE.

C. [ Required containment C.1 Perform SR 3.4.15.1. Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months air cooler condensate flow rate monitor OR inoperable.

C.2 --------------NOTE-------------- 2 Not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Perform SR 3.4.13.1. Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ]

WOG STS 3.4.15-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 352 of 478

Attachment 1, Volume 9, Rev. 1, Page 353 of 478 CTS RCS Leakage Detection Instrumentation 3.4.15 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. [ Required containment D.1 Restore required 30 days atmosphere radioactivity containment atmosphere monitor inoperable. radioactivity monitor to OPERABLE status.

AND 2 OR Required containment air cooler condensate D.2 Restore required 30 days ]

flow rate monitor containment air cooler inoperable. condensate flow rate monitor to OPERABLE status.

C 2 DOC M03 E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion 6 C Time not met. AND C

of Condition 2 A or B E.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D

2 DOC A02 F. All required monitors F.1 Enter LCO 3.0.3. Immediately inoperable. D SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.1.a, Table TS 4.1-1, SR 3.4.15.1 Perform CHANNEL CHECK of the required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Item 19 containment atmosphere radioactivity monitor.

4.1.a, Table TS 4.1-1, SR 3.4.15.2 Perform COT of the required containment 92 days Item 19 atmosphere radioactivity monitor.

4.1.a, Table SR 3.4.15.3 Perform CHANNEL CALIBRATION of the required [18] months 4 TS 4.1-1, Item 21 containment sump monitor.

WOG STS 3.4.15-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 353 of 478

Attachment 1, Volume 9, Rev. 1, Page 354 of 478 CTS RCS Leakage Detection Instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 4.1.a, Table TS 4.1-1, SR 3.4.15.4 [ Perform CHANNEL CALIBRATION of the required [18] months ] 4 Item 19 containment atmosphere radioactivity monitor.

SR 3.4.15.5 [ Perform CHANNEL CALIBRATION of the required [18] months ] 2 containment air cooler condensate flow rate monitor.

WOG STS 3.4.15-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 354 of 478

Attachment 1, Volume 9, Rev. 1, Page 355 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION

1. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 5.1.3.

2. ISTS LCO 3.4.15.c is a bracketed requirement and states that one containment air cooler condensate flow rate monitor is required to be OPERABLE. Kewaunee Power Station (KPS) has no containment air cooler condensate flow rate monitor. The KPS CTS 3.1.d.4 only requires two types of leakage detection instrumentation; sump monitoring and containment atmospheric monitor. Therefore, the bracketed ITS 3.4.15.c requirement has not been included in the KPS ITS. Due to this deletion, ISTS ACTIONS C and D and optional Required Action B.2.2 have been deleted, since they apply to the air cooler condensate flow rate monitor. Subsequent ACTIONS have been renumbered due to these deletions. Furthermore, ISTS SR 3.4.15.5 has been deleted since it applies only to the air cooler condensate flow rate monitor.

3. Not used.

4. Changes are made (additions, deletions, and/or changes) to the ISTS Specification that reflect the plant specific nomenclature, number, reference, system description, analysis or licensing basis description.

5. Not used.

6. The specific Conditions the ACTION applies to have been added, since there is one ACTION it does not apply to (ISTS 3.4.15 ACTION F, ITS 3.4.15 ACTION D). This is consistent with the Writers Guide for the Improved Standard Technical Specifications, TSTF-GG-05-01, Section 4.1.6.i.5.ii. This is also consistent with the words in the ISTS Bases for this ACTION.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 355 of 478

Attachment 1, Volume 9, Rev. 1, Page 356 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 356 of 478

Attachment 1, Volume 9, Rev. 1, Page 357 of 478 RCS Leakage Detection Instrumentation B 3.4.15 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.15 RCS Leakage Detection Instrumentation Updated Safety Analysis Report (USAR) General Design Criteria (GDC) 16 (Ref. 1) requires that means shall be provided to detect significant 2 BASES uncontrolled leakage from the reactor coolant pressure boundary.

BACKGROUND GDC 30 of Appendix A to 10 CFR 50 (Ref. 1) requires means for USAR, Section 6.5, detecting and, to the extent practical, identifying the location of the source Leakage Detection and Provisions for of RCS LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable 1 the Primary and methods for selecting leakage detection systems.

Auxiliary Coolant Loops Leakage detection systems must have the capability to detect significant reactor coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for propagation to a gross failure. Thus, an early indication or warning signal is necessary to permit proper evaluation of all unidentified LEAKAGE.

Industry practice has shown that water flow changes of 0.5 to 1.0 gpm can be readily detected in contained volumes by monitoring changes in A is water level, in flow rate, or in the operating frequency of a pump. The .

containment sump used to collect unidentified LEAKAGE [is] [(or) and air cooler condensate flow rate monitor] [are] instrumented to alarm for 1

increases of 0.5 to 1.0 gpm in the normal flow rates. This sensitivity is 3 INSERT 1 acceptable for detecting increases in unidentified LEAKAGE.

The reactor coolant contains radioactivity that, when released to the containment, can be detected by radiation monitoring instrumentation.

Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel element cladding contamination or cladding defects. Instrument sensitivities of 10-9 µCi/cc radioactivity for particulate monitoring and of 10-6 µCi/cc radioactivity for gaseous monitoring are practical for these leakage detection systems.

Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS LEAKAGE.

An increase in humidity of the containment atmosphere would indicate release of water vapor to the containment. Dew point temperature measurements can thus be used to monitor humidity levels of the containment atmosphere as an indicator of potential RCS LEAKAGE.

A 1°F increase in dew point is well within the sensitivity range of available 4 instruments.

Although not as sensitive as the air particulate monitor, the humidity detection instrumentation has the characteristics of being sensitive to vapor originating from all sources within the containment and plots of air dew point variations should be sensitive to an incremental leakage equivalent to 2 to 10 gpm.

WOG STS B 3.4.15-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 357 of 478

Attachment 1, Volume 9, Rev. 1, Page 358 of 478 B 3.4.15 3

INSERT 1 Since the fan-coil units drain to the containment sump (sump A), all condensation from primary coolant leaks is directed to the containment sump. The difference between the volume of leakage within the containment sump at the high level alarm setpoint and the volume within the containment sump at the sump pump auto shutdown setpoint is used for determining leakage rates into the containment sump. The known volume, divided by the time it takes for leakage into the containment sump to refill the containment sump to its high level alarm setpoint indicates the average leakage rate into the containment sump during the time period under consideration. Detection of leakage rates greater than 10 gpm are possible within 30 to 40 minutes. Larger leakage rates are detectable in much shorter time periods. Leakage rates of less than 0.5 gpm can be detected by this method.

Insert Page B 3.4.15-1 Attachment 1, Volume 9, Rev. 1, Page 358 of 478

Attachment 1, Volume 9, Rev. 1, Page 359 of 478 RCS Leakage Detection Instrumentation B 3.4.15 BASES BACKGROUND (continued)

Since the humidity level is influenced by several factors, a quantitative evaluation of an indicated leakage rate by this means may be questionable and should be compared to observed increases in liquid flow into or from the containment sump [and condensate flow from air 1 coolers]. Humidity level monitoring is considered most useful as an indirect alarm or indication to alert the operator to a potential problem. 5 Humidity monitors are not required by this LCO.

Air temperature and pressure monitoring methods may also be used to infer unidentified LEAKAGE to the containment. Containment temperature and pressure fluctuate slightly during plant operation, but a rise above the normally indicated range of values may indicate RCS leakage into the containment. The relevance of temperature and pressure measurements are affected by containment free volume and, for temperature, detector location. Alarm signals from these instruments can be valuable in recognizing rapid and sizable leakage to the containment.

Temperature and pressure monitors are not required by this LCO.

APPLICABLE The need to evaluate the severity of an alarm or an indication is important SAFETY to the operators, and the ability to compare and verify with indications ANALYSES from other systems is necessary. The system response times and U 1 sensitivities are described in the FSAR (Ref. 3). Multiple instrument locations are utilized, if needed, to ensure that the transport delay time of the leakage from its source to an instrument location yields an acceptable overall response time.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leakage occur detrimental to the safety of the unit and the public.

RCS leakage detection instrumentation satisfies Criterion 1 of 10 CFR 50.36(c)(2)(ii).

LCO One method of protecting against large RCS leakage derives from the ability of instruments to rapidly detect extremely small leaks. This LCO requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition, when RCS LEAKAGE indicates possible RCPB degradation.

WOG STS B 3.4.15-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 359 of 478

Attachment 1, Volume 9, Rev. 1, Page 360 of 478 RCS Leakage Detection Instrumentation All changes are 1 B 3.4.15 unless otherwise noted BASES LCO (continued)

(the hi or hi-hi alarm )

(Channel R12 or (Channel R11)

The LCO is satisfied when monitors of diverse measurement means are Channel R21 when available. Thus, the containment sump monitor, in combination with a 7 aligned to the containment) gaseous or particulate radioactivity monitor [and a containment air cooler 6 condensate flow rate monitor], provides an acceptable minimum.

APPLICABILITY Because of elevated RCS temperature and pressure in MODES 1, 2, 3, and 4, RCS leakage detection instrumentation is required to be OPERABLE.

In MODE 5 or 6, the temperature is to be 200°F and pressure is maintained low or at atmospheric pressure. Since the temperatures and pressures are far lower than those for MODES 1, 2, 3, and 4, the likelihood of leakage and crack propagation are much smaller. Therefore, the requirements of this LCO are not applicable in MODES 5 and 6.

ACTIONS A.1 and A.2 With the required containment sump monitor inoperable, no other form of sampling can provide the equivalent information; however, the containment atmosphere radioactivity monitor will provide indications of changes in leakage. Together with the atmosphere monitor, the periodic surveillance for RCS water inventory balance, SR 3.4.13.1, must be performed at an increased frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to provide information that is adequate to detect leakage. A Note is added allowing that SR 3.4.13.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, [and RCP seal injection and return flows]). The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Restoration of the required sump monitor to OPERABLE status within a Completion Time of 30 days is required to regain the function after the monitor's failure. This time is acceptable, considering the Frequency and adequacy of the RCS water inventory balance required by Required Action A.1.

WOG STS B 3.4.15-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 360 of 478

Attachment 1, Volume 9, Rev. 1, Page 361 of 478 RCS Leakage Detection Instrumentation All changes are 1 B 3.4.15 unless otherwise noted BASES ACTIONS (continued) and 6

B.1.1, B.1.2, B.2.1, and B.2.2 With both gaseous and particulate containment atmosphere radioactivity monitoring instrumentation channels inoperable, alternative action is required. Either grab samples of the containment atmosphere must be for radioactivity taken and analyzed or water inventory balances, in accordance with SR 3.4.13.1, must be performed to provide alternate periodic information.

With a sample obtained and analyzed or water inventory balance performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor may be operated for up to 30 days to allow restoration of the required containment atmosphere radioactivity 6 monitors. Alternatively, continued operation is allowed if the air cooler condensate flow rate monitoring system is OPERABLE, provided grab samples are taken or water inventory balances performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adequate to detect leakage. A Note is added allowing that SR 3.4.13.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, [and RCP seal injection and return flows]). The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established. The 30 day Completion Time recognizes at least one other form of leakage detection is available.

[ C.1 and C.2 With the required containment air cooler condensate flow rate monitor inoperable, alternative action is again required. Either SR 3.4.15.1 must be performed or water inventory balances, in accordance with SR 3.4.13.1, must be performed to provide alternate periodic information. 6 Provided a CHANNEL CHECK is performed every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or a water inventory balance is performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , reactor operation may continue while awaiting restoration of the containment air cooler condensate flow rate monitor to OPERABLE status.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adequate to detect RCS LEAKAGE. A Note is added allowing that SR 3.4.13.1 is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, [and RCP seal injection and return flows]).

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established. ]

WOG STS B 3.4.15-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 361 of 478

Attachment 1, Volume 9, Rev. 1, Page 362 of 478 RCS Leakage Detection Instrumentation All changes are 1 B 3.4.15 unless otherwise noted BASES ACTIONS (continued)

[ D.1 and D.2 With the required containment atmosphere radioactivity monitor and the 6

required containment air cooler condensate flow rate monitor inoperable, the only means of detecting leakage is the containment sump monitor.

This Condition does not provide the required diverse means of leakage detection. The Required Action is to restore either of the inoperable required monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be operated in a reduced configuration for a lengthy time period. ]

C C E.1 and E.2 or 6 If a Required Action of Condition A, B, [C], or [D] cannot be met, 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

6 D F.1 With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is operating properly. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

WOG STS B 3.4.15-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 362 of 478

Attachment 1, Volume 9, Rev. 1, Page 363 of 478 RCS Leakage Detection Instrumentation 1

B 3.4.15 All changes are unless otherwise noted BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable COT of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The test verifies the alarm setpoint and relative accuracy of the instrument string. The Frequency of 92 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation.

and 6

SR 3.4.15.3, [SR 3.4.15.4, and SR 3.4.15.5]

These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of [18] months is a typical refueling cycle and considers channel reliability. Again, operating experience has proven that this Frequency is acceptable.

REFERENCES 1. 10 CFR 50, Appendix A, Section IV, GDC 30.

USAR, Section 4.1.3.2, GDC 16, 2

2. Regulatory Guide 1.45. "Monitoring Reactor Coolant Leakage."

U USAR, Section 6.5, Leakage Detection and Provisions

3. FSAR, Section [ ]. 6.5 for the Primary and Auxiliary Coolant Loops WOG STS B 3.4.15-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 363 of 478

Attachment 1, Volume 9, Rev. 1, Page 364 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.15 BASES, RCS LEAKAGE DETECTION INSTRUMENTATION

1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases that reflect the plant specific nomenclature, number, reference, system description, analysis or licensing basis description.

2. The ISTS lists GDC 30 of Appendix A to 10 CFR 50 as the reference document for the requirement that a means shall be provided for detecting, and to the extent practical, identifying the location of the source of reactor coolant leakage. Per the information contained in USAR Section 1.8, 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 16, Monitoring Reactor Coolant Pressure Boundary, as the licensing reference document for the requirement that a means shall be provided for detecting, and to the extent practical, identifying the location of the source of reactor coolant leakage.

3. Additional descriptive information has been added to the ITS Bases Background relative to the containment sump leakage measuring system.

4. The ISTS states that a 1°F increase in dew point is well within the sensitivity range of available instruments. The information in USAR Section 6.5.1.2.4 reflects that the humidity detection instrumentation should be sensitive to an incremental leakage equivalent of 2 to 10 gallons per minute (gpm).

5. The ISTS refers to the humidity level monitoring as an indirect alarm of indication to alert the operator to a potential problem. The humidity detection instrumentation at KPS is an indication only loop and provides no alarm function. Hence the deletion of the words "alarm or" in the sentence.

6. The ISTS states, in part, that the containment sump monitor in combination with a gaseous or particulate radioactivity monitor and a containment air cooler condensate flow rate monitor provide an acceptable minimum for meeting the LCO requirement of monitoring using diverse measurement means. At KPS, there is no air cooler condensate flow rate monitor. Therefore, all references to the air cooler condensate flow rate monitor, including ACTIONS and Surveillances, have been deleted. These changes are also consistent with changes made to the actual Specification.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 364 of 478

Attachment 1, Volume 9, Rev. 1, Page 365 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.15 BASES, RCS LEAKAGE DETECTION INSTRUMENTATION

7. The containment atmosphere radioactivity monitor channel designations for the gaseous monitor (Channel R12 or R21) and the particulate monitor (Channel R11) have been relocated from Item 19 Remarks column of CTS Table TS 4.1-1 to the ITS Bases LCO as a result of DOC LA01.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 365 of 478

Attachment 1, Volume 9, Rev. 1, Page 366 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 366 of 478

Attachment 1, Volume 9, Rev. 1, Page 367 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.15, RCS LEAKAGE DETECTION INSTRUMENTATION There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 367 of 478

, Volume 9, Rev. 1, Page 368 of 478 ATTACHMENT 16 ITS 3.4.16, RCS SPECIFIC ACTIVITY , Volume 9, Rev. 1, Page 368 of 478

, Volume 9, Rev. 1, Page 369 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 369 of 478

Attachment 1, Volume 9, Rev. 1, Page 370 of 478 A01 CTS ITS 3.4.16 B. A vent pathway shall be provided with an effective flow cross section 6.4 square inches.

See ITS

1. When low temperature overpressure protection is provided via a vent 3.4.12 pathway, verify the vent pathway at least once per 31 days when the pathway is provided by a valve(s) that is locked, sealed, or otherwise secured in the open position. If the vent path is provided by any other means, then verify the vent pathway every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Maximum Coolant Activity LCO 3.4.16

1. The specific activity of the reactor coolant shall be limited to:

LCO 3.4.16, SR 3.4.16.2 A. 1.0 µCi/gram DOSE EQUIVALENT I-131, and DOSE EQUIVILENT XE-133 within limits L01 91 Ci B. gross radioactivity due to nuclides with half-lives > 30 minutes E cc excluding tritium ( E is the average sum of the beta and gamma energies in Mev per disintegration) whenever the reactor is critical or the average coolant temperature is > 500°F.

Applicability

2. If the reactor is critical or the average temperature is > 500°F: M01 ACTION C A. With the specific activity of the reactor coolant > 1.0 µCi/gram DOSE Required EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, Action A.2 or exceeding 20 µCi/gram DOSE EQUIVALENT I-131, be in at least MODE 3 ACTION C INTERMEDIATE SHUTDOWN with an average coolant temperature of < 500°F within six hours. Add proposed Required Action C.2 M01 L01 DOSE EQUIVILENT XE-133 not within limits restore to within limit in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ACTION B Add proposed 91 Ci L01 ACTION B Note B. With the specific activity of the reactor coolant > of gross radioactivity, E cc be in at least INTERMEDIATE SHUTDOWN with an average coolant ACTION C temperature < 500°F within six hours. Add proposed Required Action C.2 MODE 3 M01 C. With the specific activity of the reactor coolant > 1.0 µCi/gram DOSE Add proposed 91 Ci L01 ACTION A ACTION A Note EQUIVALENT I-131 o r > perform the sample and analysis requirements E cc A04 of Table TS 4.1-2, item 1.f, once every four hours until restored to within its A02 limits.

3. Annual reporting requirements are identified in TS 6.9.a.2.D. A03 Amendment No. 190 TS 3.1-7 03/08/2007 Page 1 of 3 Attachment 1, Volume 9, Rev. 1, Page 370 of 478

Attachment 1, Volume 9, Rev. 1, Page 371 of 478 ITS A01 ITS 3.4.16 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 3.4.16.1, SR 3.4.16.2

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 9, Rev. 1, Page 371 of 478

ITS A01 ITS 3.4.16 See CTS SR 3.4.16.1 TABLE TS 4.1-2 3.1.e SR 3.4.16.2 MINIMUM FREQUENCIES FOR SAMPLING TESTS L01 DOSE EQUIVALENT XE-133 specific activity SAMPLING TESTS TEST FREQUENCY (1) Every 7 days

1. Reactor Coolant a. Gross Radioactivity Determination (excluding tritium) 5/week Samples b. DOSE EQUIVALENT I-131 Concentration 1/14 days(2) M02 L02 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in accordance with TS 3.1.c.2.C.

2. Reactor Coolant Boron(6) Boron Concentration(3) 2/week See ITS 3.1.1 and Required ITS 3.9.1 Action A1 L01 Add proposed SR 3.4.16.2 second Frequency M02 L01 , Volume 9, Rev. 1, Page 372 of 478 Attachment 1, Volume 9, Rev. 1, Page 372 of 478 L01 (1)

Maximum time between tests is 3 days. See ITS (2) M02 3.1.1, ITS Sample required only when in the OPERATING MODE. 3.9.1 and (3) CTS 3.1.e Test required in all plant modes. See CTS (4) 3.1.e 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

(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 9, Rev. 1, Page 373 of 478 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY 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.c.2.C requires the Table TS 4.1-2 Sampling Test 1.f, RCS isotopic analysis for iodine, to be performed every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months until the specific activity of the primary coolant system is restored to within limits. ITS 3.4.16 Required Action A.1 essentially requires this same analysis, however the explicit statement to perform the isotopic analysis for iodine "until restored to within its limits" has been deleted. This changes the CTS by deleting the explicit statement to perform the isotopic analysis for iodine until the limits are met.

The purpose of the CTS 3.1.c.2.C and Table TS 4.1-2 Sampling Test 1.f is to ensure the Surveillance is performed to determine whether the specific activity is met. This statement is not necessary in the ITS, because ITS LCO 3.0.2 requires the Required Actions of the associated Conditions to be met upon discovery of failure to meet an LCO. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. This change is acceptable since ITS LCO 3.0.4 will require the Required Action to be performed until the LCO is met. This change is designated as administrative because it does not result in technical changes to the CTS.

A03 CTS 3.1.c.3 provides a cross-reference to CTS 6.9.a.2.D, the Annual Reporting Requirements. ITS 3.4.16 does not contain this cross-reference. This changes the CTS by deleting a cross-reference to another CTS requirement.

The purpose of the reference is to alert the user that a report may need to be generated due to the specific activity being outside the limit. However, CTS 6.9.a.2.D has not been included in the KPS ITS. Therefore, the cross-reference is not needed. Furthermore, it is an ITS convention to not include these types of cross-references. This change is designated as administrative because it does not result in technical changes to the CTS.

A04 CTS 3.1.c does not preclude the unit from becoming critical or increasing average temperature to > 500°F when the DOSE EQUIVALENT I-131 is > 1.0

µCi/gm for < 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Thus, during this 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time, the reactor can be made critical and the average temperature can be increased to > 500°F. ITS 3.4.16 ACTION A Note specifies that LCO 3.0.4.c is applicable. This changes the CTS by specifying the applicable ITS LCO that is consistent with the CTS allowance.

This change is acceptable since LCO 3.0.4.c provides a similar allowance - the reactor can be made critical and average temperature increased to > 500°F with Kewaunee Power Station Page 1 of 5 Attachment 1, Volume 9, Rev. 1, Page 373 of 478

Attachment 1, Volume 9, Rev. 1, Page 374 of 478 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY the DOSE EQUIVALENT I-131 > 1.0 µCi/gm for < 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This change is administrative since it does not result in any technical changes to the CTS.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.c.2 essentially requires that the specific activity of the reactor coolant shall be limited whenever the reactor is critical or the average coolant temperature is > 500°F. ITS 3.4.16 Applicability, with TSTF-490-A incorporated, requires the RCS DOSE EQUIVALENT I-131 and RCS DOSE EQUIVALENT XE-133 specific activity to be within limits during MODES 1, 2, 3 and 4. In addition, when a unit shutdown is required by CTS 3.1.c.2.A and CTS 3.1.c.2.B, the CTS requires the unit to be in INTERMEDIATE SHUTDOWN with an average coolant temperature of < 500°F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ITS 3.4.16 Required Action C.1 requires the unit to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Required Action C.2 requires the unit to be in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This changes the CTS by applying the LCO in more MODES in ITS than in CTS and by adding commensurate Required Actions to exit the new Applicability. The change that deletes the E-bar requirement and replace it with a DOSE EQUIVALENT XE-133 requirement is discussed in DOC L01.

The purpose of CTS 3.1.c is to ensure that the specific activity of the RCS is within the assumptions of the Main Steam Line Break (MSLB) and Steam Generator Tube Rupture (SGTR) analyses. This change is acceptable because the requirements continue to ensure that the process variables are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. The CTS Applicability is "whenever the reactor is critical or the average coolant temperature is > 500°F". The reactor is considered critical in the OPERATING (ITS equivalent MODE 1) and HOT STANDBY (ITS equivalent MODE 2) MODES. The reactor coolant temperature ranges of HOT SHUTDOWN (ITS equivalent MODE 3) is 540°F and INTERMEDIATE SHUTDOWN (ITS equivalent MODE 4) is > 200°F and < 540°F. However, the MODE reactor coolant temperature ranges of CTS are not equivalent to those of ITS. The reactor coolant temperature range of ITS MODE 3 is 350°F and of ITS MODE 4 is > 200°F and < 350°F. As a result of the differences in the temperature ranges between the MODE definitions of CTS and ITS, the CTS Applicability of "the average reactor coolant temperature of > 500°F" is attainable in equivalent ITS MODES 1, 2, and 3 but not in ITS MODE 4. Therefore, the ITS 3.4.16 Applicability of MODES 1, 2, 3, and 4 is more restrictive than the CTS Applicability. During operation with RCS Tavg < 500°F, the release of activity is minimal should a steam generator tube rupture occur since the saturation pressure of the reactor coolant is below the lift pressure of the main steam safety valves. This condition is satisfied once the unit enters ITS MODE 3.

Furthermore, the proposed Required Actions for when a unit shutdown are required ensure the LCO Applicability is exited. This change is designated as more restrictive because the Applicability is applicable in more MODES than in CTS and commensurate actions to exit the proposed Applicability have been added.

M02 CTS Table TS 4.1-2 Item 1.b requires the performance of a DOSE EQUIVALENT I-131 concentration test of the reactor coolant sample every 14 days when in the Kewaunee Power Station Page 2 of 5 Attachment 1, Volume 9, Rev. 1, Page 374 of 478

Attachment 1, Volume 9, Rev. 1, Page 375 of 478 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY OPERATING MODE (i.e., ITS MODE 1). ITS SR 3.4.16.2 requires verification of reactor coolant DOSE EQUIVALENT I-131 specific activity be 1.0 µCi/gm every 14 days and between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of 15%

RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period. This changes the CTS by adding a new conditional Surveillance Requirement Frequency and deleting the allowance to only perform the 14 day routine Surveillance when in MODE 1.

The purpose of ITS SR 3.4.16.2 is to verify the reactor coolant DOSE EQUIVALENT I-131 specific activity is within the assumptions of the accident (MSLB and SGTR) analyses. This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during operation and following fast power changes when iodine spiking is more apt to occur. The proposed Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following iodine spike initiation; samples at other times following a spike would provide inaccurate results. This change is more restrictive because a new SR Frequency has been added and an allowance to only perform the routine 14 day Surveillance when in MODE 1 has been deleted.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.1.c.1.B requires the gross radioactivity due to nuclides with half-lives > 30 minutes excluding tritium to be < 91/ µCi/cc. CTS 3.1.c.2.B states that if the limit is not met, then the unit must be shut down to INTERMEDIATE SHUTDOWN with an average coolant temperature < 500°F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months - no restoration time prior to the shutdown is provided. Furthermore, if the limit is not met, CTS 3.1.c.2.C requires the sample and analysis requirements of Table TS 4.1-2, item 1.f (an isotopic analysis for iodine), to be performed every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Table TS 4.1-2, Item 1.a, requires a gross radioactivity determination (excluding tritium) 5 times per week, with a maximum time between tests of 3 days and item 1.e requires an determination every 6 months with the sample being required 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . ITS 3.4.16 does not include any requirements related to . ITS LCO 3.4.16 requires the DOSE EQUIVALENT XE-133 limit to be met. SR 3.4.16.1 states that the DOSE EQUIVALENT XE-133 must be

< 595 µCi/gm and requires verification of this limit every 7 days. If DOSE EQUIVALENT XE-133 is not within the limit, ITS 3.4.16 ACTION B provides 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to restore the DOSE EQUIVALENT XE-133 to within its limit prior to requiring a unit shutdown. It also allows LCO 3.0.4.c to be applicable when in Kewaunee Power Station Page 3 of 5 Attachment 1, Volume 9, Rev. 1, Page 375 of 478

Attachment 1, Volume 9, Rev. 1, Page 376 of 478 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY ACTION B. Furthermore, when DOSE EQUIVALENT XE-133 is not within its limit, the ITS does not require the isotopic analysis for iodine to be performed every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This changes the CTS by deleting the requirements on primary coolant gross specific activity and replacing it with the DOSE EQUIVALENT XE-133 requirements on primary coolant noble gas activity, consistent with Technical Specification Task Force (TSTF) change traveler TSTF-490-A.

The proposed changes are consistent with TSTF-490-A, Revision 0. TSTF change traveler TSTF-490-A, Revision 0, "Deletion of E Bar definition and Revision to RCS Specific Activity Tech Spec" was announced for availability in the Federal Register on March 15, 2007 as part of the consolidated line item improvement process (CLIIP). The changes were approved by the NRC staff Safety Evaluation (SE) dated March 8, 2007 (ADAMS Accession No. ML070250176). KPS has reviewed the NRC staff SE listed above, the Federal Notice for comment published November 20, 2006 (including the model SE), and the Federal Notice of availability published on March 15, 2007. KPS has concluded that the justifications presented in TSTF-490-A, Revision 0 and the model SE prepared by the NRC staff are applicable to Kewaunee Power Station and justify this change. The change incorporating the newly defined quantity DOSE EQUIVALENT XE-133 is acceptable from a radiological dose perspective since it will result in an LCO that more closely relates to non-iodine RCS activity limits to the dose consequence analyses which form the bases. The Dose Conversion Factors used in the determination of DOSE EQUIVALENT I-131 and XE-133 are consistent with the Dose Conversion factors used in the applicable dose consequence analysis. This change is less restrictive because the LCO is now being based on noble gas activity versus gross specific activity and a limited amount of time (48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ) is provided to restore the limit prior to requiring a unit shutdown.

L02 (Category 5 - Deletion of Surveillance Requirement) CTS Table TS 4.1-2 Item 1.c requires a monthly reactor coolant sample for tritium activity. ITS 3.4.16, including the incorporation of TSTF-490-A, does not include this requirement.

This changes the CTS by deleting this Surveillance Requirement.

The purpose of CTS Table TS 4.1-2 Item 1.c and 1.e is to ensure plant operation within the specified gross activity LCO limit (i.e., ). TSTF-490-A, incorporated into the ITS, changes the measurement of the gross specific activity of the reactor coolant to the primary coolant noble gas activity. The bases for this change lies in the fact that when is determined using a design basis approach in which it is assumed that 1% of the power is being generated by fuel rods having cladding defects and it is also assumed that there is no removal of fission gases from the letdown flow, the value of is dominated by XE-133. During normal plant operation there are typically only a small amount of fuel defects and the radioactive nuclide inventory can become dominated by tritium and corrosion and/or activation products, resulting in the determination of a value of that is very different than would be calculated using the design basis approach. The accident dose analyses become disconnected from plant operation and the LCO becomes essentially meaningless. Since the purpose of the LCO on gross specific activity is to support the dose analyses for design basis accidents, it would be more appropriate to have the LCO apply to the noble gas concentration in the primary coolant. Thus, the current LCO on gross coolant activity, which is Kewaunee Power Station Page 4 of 5 Attachment 1, Volume 9, Rev. 1, Page 376 of 478

Attachment 1, Volume 9, Rev. 1, Page 377 of 478 DISCUSSION OF CHANGES ITS 3.4.16, RCS SPECIFIC ACTIVITY based on , is replaced by an LCO on reactor coolant noble gas activity, which is based on DOSE EQUIVALENT XE-133. This change is described in DOC L01.

This change is designated as less restrictive because a Surveillance which is required in the CTS will not be required in the ITS.

Kewaunee Power Station Page 5 of 5 Attachment 1, Volume 9, Rev. 1, Page 377 of 478

Attachment 1, Volume 9, Rev. 1, Page 378 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 378 of 478

Attachment 1, Volume 9, Rev. 1, Page 379 of 478 CTS All changes are TSTF-RCS Specific Activity unless otherwise noted 490-A 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity 3.1.c.1 LCO 3.4.16 The specific activity of the reactor coolant shall be within limits.

RCS DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

1, 2, 3, and 4.

3.1.c.2 APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 500°F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Table TS 4.1-2, A. DOSE EQUIVALENT --------------------NOTE-------------------

Item 1.f I-131 > 1.0 µCi/gm. LCO 3.0.4.c is applicable.

not within limit A.1 Verify DOSE EQUIVALENT Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 3.1.c.2.C I-131 within the acceptable region of Figure 3.4.16-1.

[60] µCi/gm.

AND 1 20 3.1.c.2.C A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT I-131 to within limit.

48 3.1.c.2.B B. Gross specific activity of B.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months the reactor coolant not Tavg < 500°F.

within limit.

DOSE EQUIVALENT --------------------NOTE-------------------

Restore DOSE EQUIVALENT XE-133 not within limit. LCO 3.0.4.c is applicable.

XE-133 to within limit.

WOG STS 3.4.16-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 379 of 478

Attachment 1, Volume 9, Rev. 1, Page 380 of 478 CTS All changes are TSTF-RCS Specific Activity unless otherwise noted 490-A 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.1.c.2.A, C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 3.1.c.2.B associated Completion Tavg < 500°F.

Time of Condition A not AND met. or B 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C.2 Be in MODE 5.

OR DOSE EQUIVALENT I-131 in the unacceptable region of > [60] µCi/gm Figure 3.4.16-1. 20 1 SURVEILLANCE REQUIREMENTS

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

2 SURVEILLANCE Only required to be performed in MODE 1.

FREQUENCY Table TS 4.1-2, SR 3.4.16.1 Verify reactor coolant gross specific activity 7 days Item 1.a 100/ µCi/gm. Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity [280] µCi/gm.

1 Table TS 595 4.1-2, SR 3.4.16.2 -------------------------------NOTE------------------------------

2 Item 1.b Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-131 14 days 1

specific activity [1.0]µCi/gm.

AND Between 2 and DOC 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a M02 THERMAL POWER change of 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period WOG STS 3.4.16-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 380 of 478

Attachment 1, Volume 9, Rev. 1, Page 381 of 478 CTS All changes are TSTF-RCS Specific Activity unless otherwise noted 490-A 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.16.3 -------------------------------NOTE------------------------------

Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine from a sample taken in MODE 1 after a 184 days minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

WOG STS 3.4.16-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 381 of 478

Attachment 1, Volume 9, Rev. 1, Page 382 of 478 CTS

[

All changes are unless otherwise noted OJ TSTF-490-A RCS Specific Activity 3.4.16 300 I E THIS FIGURE FOR ILLUSTr AT ON ONLY.

g DO NOT USE FOR OPER TION.

u

1.

f-

250 \

t::

2:

f-U u 200 LL u

w (fJ

\ UNACCEPTABLE OPERATION

--'- CDE 3 f- 150 z

w

J a

w w

(fJ 100

\~

o I o

f-z ACCE:JI; BLE

<< OPER/ ION o

o u 50 I

cr:

o f-u w

cr:

o 20 30 o 50 60 70 80 90 100 PERC NT OF RATED THERMAL POWER Figure 3.4.16-1 (page 1 of 1)

Reactor Coo ant DOSE EQUIVALENT 1-131 Specific Activity Limit V rsus Percent of RATED THERMAL POWER WOG STS 3.4.16-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 382 of 478

Attachment 1, Volume 9, Rev. 1, Page 383 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.16, RCS SPECIFIC ACTIVITY

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 allowance that SR 3.4.16.1 and SR 3.4.16.2 only have to be performed in MODE 1 has been deleted, as requested by the NRC.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 383 of 478

Attachment 1, Volume 9, Rev. 1, Page 384 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 384 of 478

Attachment 1, Volume 9, Rev. 1, Page 385 of 478 All changes are TSTF-RCS Specific Activity unless otherwise noted 490-A B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND The maximum dose to the whole body and the thyroid that an individual at following exclusion area the site boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during an accident is specified in 10 CFR 100 (Ref. 1). The limits on specific activity ensure that the INSERT 1 offsite and doses are held to a small fraction of the 10 CFR 100 limits during INSERT 2 control room analyzed transients and accidents. appropriately limited The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident. steam line break (SLB) or DOSE EQUIVALENT XE-133 main M 3 The LCO contains specific activity limits for both DOSE EQUIVALENT I-131 and gross specific activity. The allowable levels are intended to limit the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose at the site boundary to a small fraction of the 10 CFR 100 dose guideline limits. The limits in the LCO are standardized, based on parametric evaluations of offsite radioactivity dose consequences for typical site locations. INSERT 3 The parametric evaluations showed the potential offsite dose levels for a SGTR accident were an appropriately small fraction of the 10 CFR 100 dose guideline limits. Each evaluation assumes a broad range of site applicable atmospheric dispersion factors in a parametric evaluation.

INSERT 4 APPLICABLE The LCO limits on the specific activity of the reactor coolant ensures that SAFETY the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed a small ANALYSES fraction of the 10 CFR 100 dose guideline limits following a SGTR 3 and 4 150 gallons accident. The SGTR safety analysis (Ref. 2) assumes the specific activity per day es is of the reactor coolant at the LCO limit and an existing reactor coolant s steam generator (SG) tube leakage rate of 1 gpm. The safety analysis [1] gpm exists. 1 assumes the specific activity of the secondary coolant at its limit of es

[0.1]Ci/gm DOSE EQUIVALENT I-131 from LCO 3.7.6, "Secondary 1 4 Specific Activity." 17 SLB and s es The analysis for the SGTR accident establishes the acceptance limits for M RCS specific activity. Reference to this analysis is used to assess these 3 changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.

WOG STS B 3.4.16-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 385 of 478

Attachment 1, Volume 9, Rev. 1, Page 386 of 478 B 3.4.16 TSTF-490-A INSERT 1

, or at the low population zone outer boundary for the radiological release duration, TSTF-490-A INSERT 2

[10 CFR 100.11][10 CFR 50.67] (Ref. 1). Doses to control room operators must be 1 limited per GDC 19. 4 10 CFR 50, TSTF-490-A INSERT 3 ensure that offsite and control room doses meet the appropriate acceptance criteria in the Standard Review Plan (Ref. 2). 3 Regulatory Guide (RG) 1.183 TSTF-490-A INSERT 4 RG 1.183 3

offsite and control room doses meet the appropriate SRP acceptance criteria following a SLB or SGTR accident.

3 M

Insert Page B 3.4.16-1 Attachment 1, Volume 9, Rev. 1, Page 386 of 478

Attachment 1, Volume 9, Rev. 1, Page 387 of 478 All changes are TSTF-RCS Specific Activity unless otherwise noted 490-A B 3.4.16 BASES APPLICABLE SAFETY ANALYSES (continued) safety analyses consider iodine The analysis is for two cases of reactor coolant specific activity. One an accident case assumes specific activity at [1.0]Ci/gm DOSE EQUIVALENT I-131 1 3

initiated with a concurrent large iodine spike that increases the I-131 activity in the INSERT 5 reactor coolant by a factor of about 50 immediately after the accident.

20.0 The second case assumes the initial reactor coolant iodine activity at an 1 60.0 Ci/gm DOSE EQUIVALENT I-131 due to a pre-accident iodine prior to the accident spike caused by an RCS transient. In both cases, the noble gas activity INSERT 6 3 a reactor or in the reactor coolant assumes 1% failed fuel, which closely equals the specific LCO limit of 100/ Ci/gm for gross specific activity. INSERT 7 SGTR considers a possible 3

The analysis also assumes a loss of offsite power at the same time as the reactor trip SGTR event. The SGTR causes a reduction in reactor coolant inventory.

The reduction initiates a reactor trip from a low pressurizer pressure signal[or an RCS overtemperature T signal. ] 1 would 3

A The coincident loss of offsite power causes the steam dump valves to is assumed close to protect the condenser. The rise in pressure in the ruptured SG ]

to discharges radioactively contaminated steam to the atmosphere through 3 the SG power operated relief valves [and the main steam safety valves. 1 The unaffected SGs remove core decay heat by venting steam to the s

atmosphere until the cooldown ends. event is terminated 4 The safety analysis shows the radiological consequences of an SGTR accident are within a small fraction of the Reference 1 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed the limits shown in Figure 3.4.16-1, in the applicable specification, for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The safety analysis has concurrent and pre-accident iodine spiking levels up to 60.0 Ci/gm DOSE EQUIVALENT I-131.

INSERT 8 The remainder of the above limit permissible iodine levels shown in Figure 3.4.16-1 are acceptable because of the low probability of a SGTR accident occurring during the established 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time limit. The occurrence of an SGTR accident at these permissible levels could increase the site boundary dose levels, but still be within 10 CFR 100 dose guideline limits.

The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

WOG STS B 3.4.16-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 387 of 478

Attachment 1, Volume 9, Rev. 1, Page 388 of 478 B 3.4.16 TSTF-490-A INSERT 5 rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB (by a factor of 500), or SGTR (by a factor of 335), 3 respectively. M or SGTR 3

INSERT 6 for the SLB accident and 20.0 µCi/gm DOSE EQUIVALENT I-131 for the SGTR accident 3 M

TSTF-490-A INSERT 7 1

is assumed to be [280] µCi/gm DOSE EQUIVALENT XE-133.

595 TSTF-490-A INSERT 8 M

3 The SLB radiological analysis assumes that offsite power is lost at the same time as the pipe break occurs outside containment. Reactor trip occurs after the generation of an SI signal on low steam line pressure. The affected SG blows down completely and steam 3 s

is vented directly to the atmosphere. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the RHR system is placed 4 in service.

Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed [60.0] µCi/gm for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

1 20 Insert Page B 3.4.16-2 Attachment 1, Volume 9, Rev. 1, Page 388 of 478

Attachment 1, Volume 9, Rev. 1, Page 389 of 478 All changes are TSTF-RCS Specific Activity unless otherwise noted 490-A B 3.4.16 BASES in the reactor coolant LCO noble The specific iodine activity is limited to [1.0 ]Ci/gm DOSE EQUIVALENT I- 1 gas 131, and the gross specific activity in the reactor coolant is limited to the number of Ci/gm equal to 100 divided by (average disintegration energy of the sum of the average beta and gamma energies of the coolant nuclides). The limit on DOSE EQUIVALENT I-131 ensures the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose to an individual at the site boundary during the Design Basis Accident (DBA) will be a small fraction of the allowed thyroid dose.

The limit on gross specific activity ensures the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months whole body dose to an individual at the site boundary during the DBA will be a small fraction of the allowed whole body dose. INSERT 9 3 and 4 calculated SLB and es s s The SGTR accident analysis (Ref. 2) shows that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site boundary 3 M

dose levels are within acceptable limits. Violation of the LCO may result SLB or in reactor coolant radioactivity levels that could, in the event of an SGTR, M 3 lead to site boundary doses that exceed the 10 CFR 100 dose guideline limits. RG 1.183 SRP acceptance criteria (Ref. 2). 3 APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS average temperature , 2, 3, and 4 500°F, operation within the LCO limits for DOSE EQUIVALENT I-131 DOSE EQUIVALENT XE-133 is and gross specific activity are necessary to contain the potential limit consequences of an SGTR to within the acceptable site boundary dose values. SLB or SRP acceptance criteria (Ref. 2).

3 M RG 1.183 For operation in MODE 3 with RCS average temperature < 500°F, and in INSERT 10 MODES 4 and 5, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves.

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT I-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the limits of specific activity is

[60.0] µCi/gm Figure 3.4.16-1 are not exceeded. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is done to continue to provide a trend. continued every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1

20 The DOSE EQUIVALENT I-131 must be restored to within limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is required, if the limit INSERT 11 violation resulted from normal iodine spiking.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS.

This allowance is acceptable due to the significant conservatism Required Actions A.1 and A.2 while the DOSE EQUIVALENT I-131 LCO limit is not met WOG STS B 3.4.16-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 389 of 478

Attachment 1, Volume 9, Rev. 1, Page 390 of 478 B 3.4.16 TSTF-490-A INSERT 9 595 1

[280.0] µCi/gm DOSE EQUIVALENT XE-133. The limits on specific activity ensure that offsite and control room doses will meet the appropriate SRP acceptance criteria (Ref. 3 2).

RG 1.183 TSTF-490-A INSERT 10 normally 6

In MODES 5 and 6, the steam generators are not being used for decay heat removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal. Therefore, the monitoring of RCS specific activity is not required.

TSTF-490-A INSERT 11 acceptable since it is expected that, if there were an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

3 M

Insert Page B 3.4.16-3 Attachment 1, Volume 9, Rev. 1, Page 390 of 478

Attachment 1, Volume 9, Rev. 1, Page 391 of 478 All changes are TSTF-RCS Specific Activity unless otherwise noted 490-A B 3.4.16 BASES ACTIONS (continued) incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.

INSERT 12 B.1 With the gross specific activity in excess of the allowed limit, the unit must be placed in a MODE in which the requirement does not apply.

The change within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to MODE 3 and RCS average temperature

< 500°F lowers the saturation pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 below 500°F from full power conditions in an orderly manner and without challenging plant systems.

C.1 and C.2 20 any the or B 1 5 If a Required Action and the associated Completion Time of Condition A

, is not met or if the DOSE EQUIVALENT I-131 is in the unacceptable 1

> [60.0] µCi/gm region of Figure 3.4.16-1, the reactor must be brought to MODE 3 with RCS average temperature < 500°F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time s are of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach allowed the required plant conditions MODE 3 below 500°F from full power conditions in an orderly manner and without challenging plant systems. and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE SR 3.4.16.1 REQUIREMENTS noble gas SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at least once every 7 days. While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in gross specific activity.

noble gas WOG STS B 3.4.16-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 391 of 478

Attachment 1, Volume 9, Rev. 1, Page 392 of 478 B 3.4.16 TSTF-490-A INSERT 12 B.1 With the DOSE EQUIVALENT XE-133 greater than the LCO limit, DOSE EQUIVALENT XE-133 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The allowed Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

3 M

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required Action B.1 while the DOSE EQUIVALENT XE-133 LCO limit is met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

Insert Page B 3.4.16-4 Attachment 1, Volume 9, Rev. 1, Page 392 of 478

Attachment 1, Volume 9, Rev. 1, Page 393 of 478 All changes are TSTF-RCS Specific Activity unless otherwise noted 490-A B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued)

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance is applicable in MODES 1 and 2, and in MODE 3 with Tavg at least 500°F. The 7 day Frequency considers the unlikelihood of a gross fuel failure during the time.

low probability INSERT 13 SR 3.4.16.2 iodine spiking the LCO specific activity This Surveillance is performed in MODE 1 only to ensure iodine remains 7

within limit during normal operation and following fast power changes noble gas when fuel failure is more apt to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering gross activity is monitored every 7 days. The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results. iodine spike initiation INSERT 14 SR 3.4.16.3 A radiochemical analysis for determination is required every 184 days (6 months) with the plant operating in MODE 1 equilibrium conditions.

The determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. The Frequency of 184 days recognizes does not change rapidly.

This SR has been modified by a Note that indicates sampling is required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This ensures that the radioactive materials are at equilibrium so the analysis for is representative and not skewed by a crud burst or other similar abnormal event.

REFERENCES 1. 10 CFR 100.11, 1973.

INSERT 15

2. FSAR, Section [15.6.3].

WOG STS B 3.4.16-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 393 of 478

Attachment 1, Volume 9, Rev. 1, Page 394 of 478 B 3.4.16 TSTF-490-A INSERT 13 Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and I-134, it is acceptable to include the minimum detectable activity for Kr-85 in the SR 3.4.16.1 calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity.

A Note modifies the SR to allow entry into and operation in MODE 4, MODE 3, and MODE 2 prior to performing the SR. This allows the Surveillance to be performed in 7 those MODES, prior to entering MODE 1.

TSTF-490-A INSERT 14 The Note modifies the SR to allow entry into and operation in MODE 4, MODE 3, and 7

MODE 2 prior to performing the SR. This allows the Surveillance to be performed in those MODES, prior to entering MODE 1.

TSTF-490-A INSERT 15

Reviewer's Note --------------------------------------------

The first listed References 1 and 2 are for plants that are licensed to 10 CFR 100.11. 2 The second set of References are for plants that are licensed to 10 CFR 50.67.

1

[ 1. 10 CFR 100.11.

2. Standard Review Plan (SRP) Section 15.1.5 Appendix A (SLB) and Section 1

15.6.3 (SGTR).

1. 10 CFR 50.67.

2. Standard Review Plan (SRP) Section 15.0.1 "Radiological Consequence Analyses Using Alternate Source Terms." ] 3 1 U

Regulatory Guide 1.183, July 2000.

3. FSAR, Section [15.1.5]. 14.2.4 3 1 U

4. FSAR, Section [15.6.3]. 14.2.5 3 1 Insert Page B 3.4.16-5 Attachment 1, Volume 9, Rev. 1, Page 394 of 478

Attachment 1, Volume 9, Rev. 1, Page 395 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.16 BASES, RCS SPECIFIC ACTIVITY

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. TSTF-490-A, Revision 0, contains an NRC Reviewer's Note. The NRC Reviewer's Note has been deleted. This information is for the NRC reviewer to be keyed in to what is needed to meet this requirement. The NRC Reviewer's Note is not meant to be retained in the final version of the plant specific submittal.

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. Typographical error corrected.

5. ACTION A has two Required Actions. Not meeting either of the Required Actions results in entering ACTION C. Thus the word "the" has been replaced with "any".

6. ITS 3.4.7, which is applicable in MODE 5, allows the use of a steam generator as a backup method for decay heat removal. Therefore, the word "normally" has been added.

7. The allowance that SR 3.4.16.1 and SR 3.4.16.2 only have to be performed in MODE 1 (i.e., "normal" operation) has been deleted, as requested by the NRC.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 395 of 478

Attachment 1, Volume 9, Rev. 1, Page 396 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 396 of 478

Attachment 1, Volume 9, Rev. 1, Page 397 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.16, RCS SPECIFIC ACTIVITY There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 397 of 478

Attachment 1, Volume 9, Rev. 1, Page 398 of 478 ATTACHMENT 17 ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY Attachment 1, Volume 9, Rev. 1, Page 398 of 478

, Volume 9, Rev. 1, Page 399 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 399 of 478

Attachment 1, Volume 9, Rev. 1, Page 400 of 478 ITS ITS 3.4.17 A01

g. Steam Generator (SG) Tube Integrity Applicability 1. When the average reactor coolant system temperature is > 200°F the following shall be maintained:

LCO 3.4.17 A. SG Tube integrity shall be maintained, and LCO 3.4.17 B. All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

ACTIONS Note Note: Separate entry condition is allowed for each SG tube.

ACTION A 2. If the requirements of TS 3.1.g.1.B are not met, then:

A. Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, and B. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering INTERMEDIATE SHUTDOWN following the next refueling outage or SG tube inspection.

ACTION B 3. If the requirements of TS 3.1.g.2.A or TS 3.1.g.1.A are not met, then initiate action to:

- Achieve HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve COLD SHUTDOWN within an additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Amendment No. 188 TS 3.1-11 Revised by letter dated August 29, 2006 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 400 of 478

Attachment 1, Volume 9, Rev. 1, Page 401 of 478 ITS ITS 3.4.17 A01 4.19 Steam Generator (SG) Tube Integrity APPLICABILITY Applies to the surveillance requirements for Steam Generator (SG) Tube Integrity in TS 3.1.g.

OBJECTIVE To assure that the Steam Generator Tube Integrity requirements are verified in a sufficient periodicity.

SPECIFICATION SR 3.4.17.1 a. Verify SG tube integrity in accordance with the Steam Generator Program.

SR 3.4.17.2 b. Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering INTERMEDIATE SHUTDOWN following a SG tube inspection.

Amendment No. 188 TS 4.19-1 Revised by letter dated August 29, 2006 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 401 of 478

Attachment 1, Volume 9, Rev. 1, Page 402 of 478 DISCUSSION OF CHANGES ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY 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 None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 402 of 478

Attachment 1, Volume 9, Rev. 1, Page 403 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 403 of 478

Attachment 1, Volume 9, Rev. 1, Page 404 of 478 CTS SG Tube Integrity 1

All changes are 3.4.20 unless otherwise noted 17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.20 Steam Generator (SG) Tube Integrity 17 3.1.g.1.A LCO 3.4.20 SG tube integrity shall be maintained.

17 AND 3.1.g.1.B All SG tubes satisfying the tube repair criteria shall be plugged [or repaired] in accordance with the Steam Generator Program.

3.1.g.1 APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

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

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME 3.1.g.2 A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next

[or repaired] in refueling outage or SG tube accordance with the inspection.

Steam Generator Program. AND A.2 Plug [or repair] the affected Prior to entering tube(s) in accordance with MODE 4 following the the Steam Generator next refueling outage Program. or SG tube inspection 3.1.g.3 B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR SG tube integrity not maintained.

WOG STS 3.4.20-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 404 of 478

Attachment 1, Volume 9, Rev. 1, Page 405 of 478 SG Tube Integrity CTS All changes are 1 3.4.20 unless otherwise noted 17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.1.9.a SR 3.4.20.1 Verify SG tube integrity in accordance with the In accordance 17 Steam Generator Program. with the Steam Generator Program 4.1.9.b SR 3.4.20.2 Verify that each inspected SG tube that satisfies the Prior to entering 17 tube repair criteria is plugged [or repaired] in MODE 4 following accordance with the Steam Generator Program. a SG tube inspection WOG STS 3.4.20-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 405 of 478

Attachment 1, Volume 9, Rev. 1, Page 406 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY

1. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. ISTS 5.5.9, Steam Generator Program, includes a Reviewer's Note that states the repair criteria currently permitted by plant technical specifications should be provided in the ITS. The bracketed allowance to repair a steam generator tube is not included since the current KPS Steam Generator Program does not allow repair; only plugging is allowed.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 406 of 478

Attachment 1, Volume 9, Rev. 1, Page 407 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 407 of 478

Attachment 1, Volume 9, Rev. 1, Page 408 of 478 SG Tube Integrity B 3.4.20 4 17 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.20 Steam Generator (SG) Tube Integrity 4

17 BASES BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.

The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary systems pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, "RCS Loops - MODES 1 and 2," LCO 3.4.5, "RCS Loops - MODE 3," LCO 3.4.6, "RCS Loops - MODE 4," and LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled."

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively.

The SG performance criteria are used to manage SG tube degradation.

7 5

Specification 5.5.9, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity 5

is maintained. Pursuant to Specification 5.5.9, tube integrity is 7 maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.5.9. Meeting the SG performance criteria provides 5 7

reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

WOG STS B 3.4.20-1 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 408 of 478

Attachment 1, Volume 9, Rev. 1, Page 409 of 478 All changes are 1 SG Tube Integrity unless otherwise noted B 3.4.20 4 17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The largely contained within accident analysis for a SGTR assumes the contaminated secondary fluid the secondary side of the affected SG and is only briefly released to the atmosphere via safety valves and the secondary majority is discharged to the main condenser. system leakage The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all of 150 gallons per day per SG SGs of [1 gallon per minute] or is assumed to increase to [1 gallon per 150 gallons minute] as a result of accident induced conditions. For accidents that do per day SG per not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC 2 approved licensing basis (e.g., a small fraction of these limits).

10 CFR 50.67 (Ref. 2) 3 Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged [or 4 repaired] in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is [repaired or] removed from service by 4 plugging. If a tube was determined to satisfy the repair criteria but was not plugged [or repaired], the tube may still have tube integrity. 4 In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall [and any repairs made to it], 4 between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria.

The SG performance criteria are defined in Specification 5.5.9, "Steam 7 5 Generator Program," and describe acceptable SG tube performance.

The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

WOG STS B 3.4.20-2 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 409 of 478

Attachment 1, Volume 9, Rev. 1, Page 410 of 478 All changes are 1 SG Tube Integrity unless otherwise noted B 3.4.20 4 17 BASES LCO (continued)

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.

The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.

This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory 3 Guide 1.121 (Ref. 5). 3 4

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed [1 gpm per SG, except for specific types of degradation at specific locations 150 gallons per where the NRC has approved greater accident induced leakage.] The day accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

WOG STS B 3.4.20-3 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 410 of 478

Attachment 1, Volume 9, Rev. 1, Page 411 of 478 SG Tube Integrity B 3.4.20 4 17 BASES LCO (continued)

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged [or repaired] in accordance with the Steam Generator 4 17 Program as required by SR 3.4.20.2. An evaluation of SG tube integrity 4 of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged

[or repaired] has tube integrity, an evaluation must be completed that 4 demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity WOG STS B 3.4.20-4 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 411 of 478

Attachment 1, Volume 9, Rev. 1, Page 412 of 478 SG Tube Integrity B 3.4.20 4 17 BASES ACTIONS (continued) determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged [or 4 repaired] prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.20.1 4 REQUIREMENTS 17 During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

WOG STS B 3.4.20-5 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 412 of 478

Attachment 1, Volume 9, Rev. 1, Page 413 of 478 SG Tube Integrity B 3.4.20 4 17 BASES SURVEILLANCE REQUIREMENTS (continued)

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation.

Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

17 The Steam Generator Program defines the Frequency of SR 3.4.20.1. 4 The Frequency is determined by the operational assessment and other 3 5

limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next 7 5 scheduled inspection. In addition, Specification 5.5.9 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SR 3.4.20.2 4 17 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is [repaired or] removed from service by 4 plugging. The tube repair criteria delineated in Specification 5.5.9 are 7 5 intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

[Steam generator tube repairs are only performed using approved repair 4 methods as described in the Steam Generator Program.]

The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged [or repaired] prior to subjecting the SG 4 tubes to significant primary to secondary pressure differential.

WOG STS B 3.4.20-6 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 413 of 478

Attachment 1, Volume 9, Rev. 1, Page 414 of 478 SG Tube Integrity B 3.4.20 4 17 BASES REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. 10 CFR 50 Appendix A, GDC 19. 10 CFR 50.67 2 3

3. 10 CFR 100.

3 4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

4 5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam 3 Generator Tubes," August 1976.

5 6. EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

WOG STS B 3.4.20-7 Rev. 3.1, 12/01/05 Attachment 1, Volume 9, Rev. 1, Page 414 of 478

Attachment 1, Volume 9, Rev. 1, Page 415 of 478 JUSTIFICATION FOR DEVIATIONS ITS 3.4.17 BASES, STEAM GENERATOR (SG) TUBE INTEGRITY

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. 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. Reference 2 was revised to 10 CFR 50.67 to reflect Kewaunee Power Station (KPS) current licensing basis. Reference 3 was deleted since it is not applicable to KPS current licensing basis. As a result, the remaining references were renumbered.

4. Changes made to be consistent with changes made to the Specification.

5. ISTS 5.5.3, "Post Accident Sampling," and the ISTS 5.5.6, "Pre-Stressed Concrete Containment Tendon Surveillance Program," are not included in the Kewaunee Power Station (KPS) ITS. As a result, subsequent programs in ITS Section 5.5 have been renumbered and Specification 5.5.9 is now 5.5.7.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 415 of 478

Attachment 1, Volume 9, Rev. 1, Page 416 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 416 of 478

Attachment 1, Volume 9, Rev. 1, Page 417 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.17, STEAM GENERATOR (SG) TUBE INTEGRITY There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 417 of 478

, Volume 9, Rev. 1, Page 418 of 478 ATTACHMENT 18 RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS , Volume 9, Rev. 1, Page 418 of 478

Attachment 1, Volume 9, Rev. 1, Page 419 of 478 CTS 3.1.a.7, REACTOR COOLANT VENT SYSTEM Attachment 1, Volume 9, Rev. 1, Page 419 of 478

, Volume 9, Rev. 1, Page 420 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 420 of 478

Attachment 1, Volume 9, Rev. 1, Page 421 of 478 CTS 3.1.a.7

4. With one block valve inoperable, within one hour restore the block valve to OPERABLE status or place its associated PORV in manual control. Restore the block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months See ITS 3.4.11

5. With both block valves inoperable, within one hour restore the block valves to OPERABLE status or place their associated PORVs in manual control.

Restore at least one block valve to OPERABLE status within the next hour; otherwise, action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

6. Pressurizer Heaters See ITS

. At least one group of pressurizer heaters shall have an emergency power supply 3.4.9 available when the average RCS temperature is > 350°F.

7. Reactor Coolant Vent System

. A reactor coolant vent path from both the reactor vessel head and pressurizer steam space shall be OPERABLE and closed prior to the average RCS temperature being heated > 200°F except as specified in TS 3.1.a.7.B and TS 3.1.a.7.C below.

A. When the average RCS temperature is > 200°F, any one of the following conditions of inoperability may exist:

1. Both of the parallel vent valves in the reactor vessel vent path are inoperable.

2. Both of the parallel vent valves in the pressurizer vent path are inoperable.

R01 If OPERABILITY is not restored within 30 days, then within one hour action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve COLD SHUTDOWN within an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months B. If no Reactor Coolant System vent paths are OPERABLE, then restore at least one vent path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If OPERABILITY is not restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , then within one hour action shall be initiated to:

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

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve COLD SHUTDOWN within an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Amendment No. 165 TS 3.1-5 03/11/2003 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 421 of 478

Attachment 1, Volume 9, Rev. 1, Page 422 of 478 CTS 3.1.a.7 4.16 Reactor Coolant Vent System Tests Applicability Applies to the surveillance testing requirements of the reactor coolant vent system.

Objectives To assure that the capability exists to vent non-condensible gases from the reactor coolant system, if required.

Specification

a. Vent Path Operability At least once per operating cycle or once every 18 months, whichever occurs first, each reactor coolant system vent path shall be demonstrated operable by:

1) Cycling each solenoid operated valve in each vent path through at least one R01 complete cycle of full travel.

2) Verifying that unobstructed flow exists through the reactor coolant vent system paths during the normal filling and venting operations following refueling.

Basis The cycling of each solenoid operated valve once each refueling ensures that the valves are capable of opening, if required, to vent the reactor coolant system. More frequent cycling of these valves is not practical since it would provide unnecessary challenges to the reactor coolant pressure boundary during plant operation.

Flow verification is performed to assure that there are no blockages in the reactor coolant system vent piping that would prevent venting of non-condensible gases from the reactor coolant system.

Flow verification is performed following each refueling by qualitatively assuring flow exists through the system during the postrefueling filling and venting of the RCS.

Amendment No. 59 TS 4.16-1 03/10/85 Page 2 of 2 Attachment 1, Volume 9, Rev. 1, Page 422 of 478

Attachment 1, Volume 9, Rev. 1, Page 423 of 478 DISCUSSION OF CHANGES CTS 3.1.a.7, REACTOR COOLANT VENT SYSTEM ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.1.a.7 provides requirements for the reactor coolant vent system.

CTS 4.16 provides the testing requirements for the reactor coolant vent system.

The reactor coolant vent system is provided to exhaust noncondensible gases and/or steam from the RCS which could inhibit natural circulation core cooling following any event involving a loss of offsite power and requiring long term cooling, such as a loss-of-coolant accident (LOCA). Their function, capabilities, and test requirements are consistent with the requirements of Item II.B.1 of NUREG-0737, "Clarification of TMI Action Plan Requirements," however; the operation of reactor vessel head vents is not part of the primary success path.

The operation of these vents is an operator action after the event has occurred, and is only required when there is indication that natural circulation is not occurring. This Specification does not meet the criteria for retention in the ITS; therefore, it will be retained in the Technical Requirements Manual (TRM).

This change is acceptable because CTS 3.1.a.7 and CTS 4.16 do 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 reactor coolant vent system is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. The reactor coolant vent system Specifications do not satisfy criterion 1.

2. The reactor coolant vent 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 reactor coolant vent system Specifications do not satisfy criterion 2.

3. The reactor coolant vent 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 reactor coolant vent system Specifications do not satisfy criterion 3.

4. The reactor coolant vent system is not a structure, system or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. As discussed in Section 4.0 Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 423 of 478

Attachment 1, Volume 9, Rev. 1, Page 424 of 478 DISCUSSION OF CHANGES CTS 3.1.a.7, REACTOR COOLANT VENT SYSTEM (Appendix A, page A-44) and summarized in Table 1 of WCAP-11618, the reactor vessel head vents were 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 reactor coolant vent system Specifications do not satisfy criterion 4.

Since 10 CFR 50.36(c)(2)(ii) criteria have not been met, the reactor coolant vent system Specifications may 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 9, Rev. 1, Page 424 of 478

Attachment 1, Volume 9, Rev. 1, Page 425 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 425 of 478

Attachment 1, Volume 9, Rev. 1, Page 426 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.1.a.7, REACTOR COOLANT VENT SYSTEM There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 426 of 478

Attachment 1, Volume 9, Rev. 1, Page 427 of 478 CTS 3.1.b.2, STEAM GENERATOR PRESSURE/TEMPERATURE LIMITS Attachment 1, Volume 9, Rev. 1, Page 427 of 478

, Volume 9, Rev. 1, Page 428 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 428 of 478

Attachment 1, Volume 9, Rev. 1, Page 429 of 478 CTS 3.1.b.2

b. Heatup and Cooldown Limit Curves for Normal Operation

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures TS 3.1-1 and TS 3.1-2. Figures TS 3.1-1 and TS 3.1-2 are applicable for the service period of up to 33 (1) effective full-power years.

See ITS 3.4.3 A. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

B. Figures TS 3.1-1 and TS 3.1-2 define limits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g.,

pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

C. The isothermal curve in Figure TS 3.1-2 defines limits to assure prevention of non-ductile failure applicable to low temperature overpressurization events only. See ITS 3.4.12 Application of this curve is limited to evaluation of LTOP events whenever one or more of the RCS cold leg temperatures are less than or equal to the LTOP enabling temperature of 200°F.

2. The secondary side of the steam generator must not be pressurized > 200 psig if the R01 temperature of the steam generator is < 70°F.

3. The pressurizer cooldown and heatup rates shall not exceed 200°F/hr and 100°F/hr, See CTS respectively. The spray shall not be used if the temperature difference between the 3.1.b.3 pressurizer and the spray fluid is > 320°F.

4. The overpressure protection system for low temperature operation shall be OPERABLE whenever one or more of the RCS cold leg temperatures are 200°F, and the reactor vessel head is installed. The system shall be considered OPERABLE when at least one of the following conditions is satisfied:

A. The overpressure relief valve on the Residual Heat Removal System (RHR 33-1) shall have a set pressure of 500 psig and shall be aligned to the RCS by maintaining valves RHR 1A, 1B, 2A, and 2B open.

See ITS

1. With one flow path inoperable, the valves in the parallel flow path shall be 3.4.12 verified open with the associated motor breakers for the valves locked in the off position. Restore the inoperable flow path within five days or complete depressurization and venting of the RCS through a 6.4 square inch vent within an additional eight hours.

2. With both flow paths or RHR 33-1 inoperable, complete depressurization and venting of the RCS through at least a 6.4 square inch vent pathway within eight hours.

(1)

The curves are limited to 31.1 EFPY due to changes in vessel fluence associated with See ITS operation at uprated power. 3.4.3 Amendment No. 168 TS 3.1-6 07/08/2003 Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 429 of 478

Attachment 1, Volume 9, Rev. 1, Page 430 of 478 DISCUSSION OF CHANGES CTS 3.1.b.2, STEAM GENERATOR PRESSURE/TEMPERATURE LIMITS ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.1.b.2 states that the secondary side of the steam generator must not be pressurized > 200 psig if the temperature of the steam generator is < 70ºF. The limits meet the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G. These limitations are consistent with structural analysis results. However, these limits are not initial condition assumptions of a DBA or transient. These limits represent operating restrictions and Criterion 2 includes operating restrictions. However, it should be noted that in the Final Policy Station the Criterion 2 discussion specified only those operating restrictions required to preclude unanalyzed accidents and transients be included in Technical Specifications. This Specification does not meet the criteria for retention in the ITS, therefore, it will be retained in the Technical Requirements Manual (TRM).

This change is acceptable because CTS 3.1.b.2 does not meet the 10CFR 50.36(c)(2)(ii) criteria for inclusion into the ITS.

10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. The steam generator pressure/temperature limit is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. The steam generator pressure/temperature limits Specification does not satisfy criterion 1.

2. The steam generator pressure/temperature limit 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 steam generator pressure/temperature limits Specification does not satisfy criterion 2.

3. The steam generator pressure/temperature limit is not a structure, system, or component that is part of 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 steam generator pressure/temperature limits Specification does not satisfy criterion 3.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 430 of 478

Attachment 1, Volume 9, Rev. 1, Page 431 of 478 DISCUSSION OF CHANGES CTS 3.1.b.2, STEAM GENERATOR PRESSURE/TEMPERATURE LIMITS

4. The steam generator pressure/temperature limit is not a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. As Discussed in Section 4.0 and summarized in Table 1 of WCAP-11618, the steam generator pressure/temperature limits were 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 steam generator pressure/temperature limits Specification does not satisfy criterion 4.

Since 10 CFR 50.36(c)(2)(ii) criteria have not been met, the Steam Generator temperature/pressure limits Specification 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 9, Rev. 1, Page 431 of 478

Attachment 1, Volume 9, Rev. 1, Page 432 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 432 of 478

Attachment 1, Volume 9, Rev. 1, Page 433 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.1.b.2, STEAM GENERATOR PRESSURE/TEMPERATURE LIMITS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 433 of 478

Attachment 1, Volume 9, Rev. 1, Page 434 of 478 CTS 3.1.b.3, PRESSURIZER PRESSURE/TEMPERATURE LIMITS Attachment 1, Volume 9, Rev. 1, Page 434 of 478

, Volume 9, Rev. 1, Page 435 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 435 of 478

Attachment 1, Volume 9, Rev. 1, Page 436 of 478 CTS 3.1.b.3

b. Heatup and Cooldown Limit Curves for Normal Operation

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures TS 3.1-1 and TS 3.1-2. Figures TS 3.1-1 and TS 3.1-2 are applicable for the service period of up to 33 (1) effective full-power years.

See ITS 3.4.3 A. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

B. Figures TS 3.1-1 and TS 3.1-2 define limits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g.,

pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

C. The isothermal curve in Figure TS 3.1-2 defines limits to assure prevention of non-ductile failure applicable to low temperature overpressurization events only. See ITS 3.4.12 Application of this curve is limited to evaluation of LTOP events whenever one or more of the RCS cold leg temperatures are less than or equal to the LTOP enabling temperature of 200°F.

2. The secondary side of the steam generator must not be pressurized > 200 psig if the See CTS temperature of the steam generator is < 70°F. 3.1.b.2

3. The pressurizer cooldown and heatup rates shall not exceed 200°F/hr and 100°F/hr, respectively. The spray shall not be used if the temperature difference between the R01 pressurizer and the spray fluid is > 320°F.

4. The overpressure protection system for low temperature operation shall be OPERABLE whenever one or more of the RCS cold leg temperatures are 200°F, and the reactor vessel head is installed. The system shall be considered OPERABLE when at least one of the following conditions is satisfied:

A. The overpressure relief valve on the Residual Heat Removal System (RHR 33-1) shall have a set pressure of 500 psig and shall be aligned to the RCS by maintaining valves RHR 1A, 1B, 2A, and 2B open.

See ITS

1. With one flow path inoperable, the valves in the parallel flow path shall be 3.4.12 verified open with the associated motor breakers for the valves locked in the off position. Restore the inoperable flow path within five days or complete depressurization and venting of the RCS through a 6.4 square inch vent within an additional eight hours.

2. With both flow paths or RHR 33-1 inoperable, complete depressurization and venting of the RCS through at least a 6.4 square inch vent pathway within eight hours.

(1)

The curves are limited to 31.1 EFPY due to changes in vessel fluence associated with See ITS operation at uprated power. 3.4.3 Amendment No. 168 TS 3.1-6 07/08/2003 Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 436 of 478

Attachment 1, Volume 9, Rev. 1, Page 437 of 478 DISCUSSION OF CHANGES CTS 3.1.b.3, PRESSURIZER PRESSURE/TEMPERATURE LIMITS ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.1.b.3 states that the pressurizer cooldown and heatup rates shall not exceed 200ºF/hr and 100ºF/hr, respectively. It also states that the spray shall not be used if the temperature difference between the pressurizer and the spray fluid is > 320ºF. The limits meet the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G. These limitations are consistent with structural analysis results. However, these limits are not initial condition assumptions of a DBA or transient. These limits represent operating restrictions and Criterion 2 includes operating restrictions. However, it should be noted that in the Final Policy Station the Criterion 2 discussion specified only those operating restrictions required to preclude unanalyzed accidents and transients be included in Technical Specifications. This Specification does not meet the criteria for retention in the ITS, therefore, it will be retained in the Technical Requirements Manual (TRM).

This change is acceptable because CTS 3.1.b.3 does not meet the 10CFR 50.36(c)(2)(ii) criteria for inclusion into the ITS.

10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. The pressurizer pressure/temperature limits are not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. The Pressurizer pressure/temperature limits Specification does not satisfy criterion 1.

2. The pressurizer pressure/temperature limits are 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 Pressurizer pressure/temperature limits Specification does not satisfy criterion 2.

3. The pressurizer pressure/temperature limits are not a structure, system, or component that is part of 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 Pressurizer pressure/temperature limits Specification does not satisfy criterion 3.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 437 of 478

Attachment 1, Volume 9, Rev. 1, Page 438 of 478 DISCUSSION OF CHANGES CTS 3.1.b.3, PRESSURIZER PRESSURE/TEMPERATURE LIMITS

4. The pressurizer pressure/temperature limits are not a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. As Discussed in Section 4.0 and summarized in Table 1 of WCAP-11618, the pressurizer pressure/temperature limits were 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 Pressurizer pressure/temperature limits Specification does not satisfy criterion 4.

Since 10 CFR 50.36(c)(2)(ii) criteria have not been met, the Pressurizer pressure/temperature limits Specification 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 9, Rev. 1, Page 438 of 478

Attachment 1, Volume 9, Rev. 1, Page 439 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 439 of 478

Attachment 1, Volume 9, Rev. 1, Page 440 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.1.b.3, PRESSURIZER PRESSURE/TEMPERATURE LIMITS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 440 of 478

, Volume 9, Rev. 1, Page 441 of 478 CTS 3.1.e, CHEMISTRY , Volume 9, Rev. 1, Page 441 of 478

, Volume 9, Rev. 1, Page 442 of 478 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 9, Rev. 1, Page 442 of 478

Attachment 1, Volume 9, Rev. 1, Page 443 of 478 CTS 3.1.e

e. Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration

1. Concentrations of contaminants in the reactor coolant shall not exceed the following limits when the reactor coolant temperature is > 250°F.

NORMAL STEADY-STATE TRANSIENT CONTAMINANT OPERATION (ppm) LIMITS (ppm)

A. Oxygen 0.10 1.00 B. Chloride 0.15 1.50 C. Fluoride 0.15 1.50

2. If any of the normal steady-state operating limits as specified in TS 3.1.e.1 above are exceeded, or if it is anticipated that they may be exceeded, then corrective action shall be taken immediately.

3. If the concentrations of any of the contaminants cannot be controlled within the transient limits of TS 3.1.e.1 above or returned to the normal steady-state limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then the reactor shall be brought to the COLD SHUTDOWN condition, utilizing normal operating procedures, and the cause shall be ascertained and corrected. The reactor may be restarted and operation resumed if the maximum R01 concentration of any of the contaminants did not exceed the permitted transient values. Otherwise a safety review by the Plant Operations Review Committee shall be made before starting.

4. Concentrations of contaminants in the reactor coolant shall not exceed the following maximum limits when the reactor coolant temperature is 250°F.

NORMAL TRANSIENT CONTAMINANT CONCENTRATION (ppm) LIMITS (ppm)

A. Oxygen Saturated Saturated B. Chloride 0.15 1.50 C. Fluoride 0.15 1.50

5. If the transient limits of TS 3.1.e.4 are exceeded or the concentrations cannot be returned to normal values within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , then the reactor shall be brought to the COLD SHUTDOWN condition and the cause shall be ascertained and corrected.

6. To meet TS 3.1.e.1 and TS 3.1.e.4 above, reactor coolant pump operation shall be permitted for short periods, provided the coolant temperature does not exceed 250°F.

Amendment No. 165 TS 3.1-9 03/11/2003 Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 443 of 478

CTS 3.1.e TABLE TS 4.1-2 MINIMUM FREQUENCIES FOR SAMPLING TESTS SAMPLING TESTS TEST FREQUENCY (1)

1. Reactor Coolant a. Gross Radioactivity Determination (excluding tritium) 5/week See ITS Samples b. DOSE EQUIVALENT I-131 Concentration 1/14 days(2) 3.4.16

c. Tritium activity Monthly R01

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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in accordance with TS 3.1.c.2.C.

2. Reactor Coolant Boron(6) Boron Concentration(3) 2/week See ITS See ITS 3.4.16 3.1.1 and 3.9.1 , Volume 9, Rev. 1, Page 444 of 478 Attachment 1, Volume 9, Rev. 1, Page 444 of 478 (1)

Maximum time between tests is 3 days. See ITS (2) 3.4.16 Sample required only when in the OPERATING MODE.

(3) See ITS Test required in all plant modes. R01 3.4.16 (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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

(6)

A reactor coolant boron concentration sample does not have to be taken when the core is completely unloaded.

See ITS Amendment No. 119 Page 1 of 2 3.9.1 04/18/95 Page 2 of 2

Attachment 1, Volume 9, Rev. 1, Page 445 of 478 DISCUSSION OF CHANGES CTS 3.1.e, CHEMISTRY ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.1.e provides limits on the oxygen, chloride and fluoride content in the RCS. CTS Table 4.1-2 Sampling Test 1.d provides the testing requirements for the oxygen, chloride and fluoride content in the RCS. Poor coolant chemistry contributes to the long term degradation of system materials of construction, and thus is not of immediate importance to the plant operator. One reason is to reduce the possibility of failures in the Reactor Coolant System pressure boundary caused by corrosion. However, the chemistry monitoring activity is of a long term preventative purpose rather than of a mitigative purpose. This Specification does not meet the criteria for retention in the ITS; it will be retained in the Technical Requirements Manual (TRM).

This change is acceptable because CTS 3.1.e and CTS Table 4.1-2 Sampling Test 1.d do 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 RCS chemistry limits are not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. The RCS Chemistry Specification does not satisfy criterion 1.

2. The RCS chemistry limits are 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 RCS Chemistry Specification does not satisfy criterion 2.

3. The RCS chemistry limits are 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 RCS Chemistry Specification does not satisfy criterion 3.

4. The RCS chemistry limits are not a structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. As discussed in Section 4.0 (Appendix A, Page A-40) and summarized in Table 1 of WCAP-11618, the RCS chemistry limits are found to be a non-significant risk contributor to core damage and offsite releases. Dominion Energy Kewaunee (DEK)

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 9, Rev. 1, Page 445 of 478

Attachment 1, Volume 9, Rev. 1, Page 446 of 478 DISCUSSION OF CHANGES CTS 3.1.e, CHEMISTRY has reviewed this evaluation, considers it applicable to Kewaunee Power Station (KPS) and concurs with this assessment. The RCS Chemistry Specification does not meet criterion 4.

Since the 10 CFR 50.36(c)(2)(ii) criteria have not been met, the RCS Chemistry may be relocated out of the Technical Specifications. The RCS Chemistry Specifications 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 specifications 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 9, Rev. 1, Page 446 of 478

Attachment 1, Volume 9, Rev. 1, Page 447 of 478 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 9, Rev. 1, Page 447 of 478

Attachment 1, Volume 9, Rev. 1, Page 448 of 478 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.1.e, CHEMISTRY There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 448 of 478

Attachment 1, Volume 9, Rev. 1, Page 449 of 478 ATTACHMENT 19 Improved Standard Technical Specifications (ISTS) not used in the Kewaunee Power Station ITS Attachment 1, Volume 9, Rev. 1, Page 449 of 478

Attachment 1, Volume 9, Rev. 1, Page 450 of 478 ISTS 3.4.17, REACTOR COOLANT SYSTEM (RCS) LOOP ISOLATION VALVES Attachment 1, Volume 9, Rev. 1, Page 450 of 478

Attachment 1, Volume 9, Rev. 1, Page 451 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 451 of 478

Attachment 1, Volume 9, Rev. 1, Page 452 of 478 RCS Loop Isolation Valves 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 RCS Loop Isolation Valves LCO 3.4.17 Each RCS hot and cold leg loop isolation valve shall be open with power removed from each isolation valve operator.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

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

Separate Condition entry is allowed for each RCS loop isolation valve.

CONDITION REQUIRED ACTION COMPLETION TIME A. Power available to one A.1 Remove power from loop 30 minutes or more loop isolation isolation valve operators.

valve operators.

B. ------------NOTE------------ B.1 Maintain valve(s) closed. Immediately All Required Actions shall be completed AND whenever this Condition is entered. B.2 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 1 AND One or more RCS loop isolation valves closed. B.3 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify each RCS loop isolation valve is open and 31 days power is removed from each loop isolation valve operator.

WOG STS 3.4.17-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 452 of 478

Attachment 1, Volume 9, Rev. 1, Page 453 of 478 JUSTIFICATION FOR DEVIATIONS ISTS 3.4.17, REACTOR COOLANT SYSTEM (RCS) LOOP ISOLATION VALVES

1. ISTS 3.4.17, "Reactor Coolant System (RCS) Loop Isolation Valves," is not included in the Kewaunee Power Station (KPS) ITS because the Reactor Coolant System hot and cold loops do not include isolation valves.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 453 of 478

Attachment 1, Volume 9, Rev. 1, Page 454 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 454 of 478

Attachment 1, Volume 9, Rev. 1, Page 455 of 478 RCS Loop Isolation Valves B 3.4.17 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.17 RCS Loop Isolation Valves BASES BACKGROUND The reactor coolant loops are equipped with loop isolation valves that permit any loop to be isolated from the reactor vessel. One valve is installed on each hot leg and one on each cold leg. The loop isolation valves are used to perform maintenance on an isolated loop. Power operation with a loop isolated is not permitted.

To ensure that inadvertent closure of a loop isolation valve does not occur, the valves must be open with power to the valve operators removed in MODES 1, 2, 3, and 4. If the valves are closed, a set of administrative controls and equipment interlocks must be satisfied prior to opening the isolation valves as described in LCO 3.4.18, "RCS Isolated Loop Startup."

APPLICABLE The safety analyses performed for the reactor at power assume that all SAFETY reactor coolant loops are initially in operation and the loop isolation ANALYSES valves are open. This LCO places controls on the loop isolation valves to ensure that the valves are not inadvertently closed in MODES 1, 2, 3, and

4. The inadvertent closure of a loop isolation valve when the Reactor 1 Coolant Pumps (RCPs) are operating will result in a partial loss of forced reactor coolant flow (Ref. 1). If the reactor is at power at the time of the event, the effect of the partial loss of forced coolant flow is a rapid increase in the coolant temperature which could result in DNB with subsequent fuel damage if the reactor is not tripped by the Low Flow reactor trip. If the reactor is shutdown and an RCS loop is in operation removing decay heat, closure of the loop isolation valve associated with the operating loop could also result in increasing coolant temperature and the possibility of fuel damage.

RCS Loop Isolation Valves satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO ensures that the loop isolation valves are open and power to the valve operators is removed. Loop isolation valves are used for performing maintenance in MODES 5 and 6. The safety analyses assume that the loop isolation valves are open in any RCS loops required to be OPERABLE by LCO 3.4.4, "RCS Loops - MODES 1 and 2,"

LCO 3.4.5, "RCS Loops - MODE 3," or LCO 3.4.6, "RCS Loops -

MODE 4."

WOG STS B 3.4.17-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 455 of 478

Attachment 1, Volume 9, Rev. 1, Page 456 of 478 RCS Loop Isolation Valves B 3.4.17 BASES APPLICABILITY In MODES 1 through 4, this LCO ensures that the loop isolation valves are open and power to the valve operators is removed. The safety analyses assume that the loop isolation valves are open in any RCS loops required to be OPERABLE.

In MODES 5 and 6, the loop isolation valves may be closed. Controlled startup of an isolated loop is governed by the requirements of LCO 3.4.18, "RCS Isolated Loop Startup."

ACTIONS The Actions have been provided with a Note to clarify that all RCS loop isolation valves for this LCO are treated as separate entities, each with separate Completion Times, i.e., the Completion Time is on a component basis.

A.1 If power is inadvertently restored to one or more loop isolation valve operators, the potential exists for accidental isolation of a loop. The loop isolation valves have motor operators. Therefore, these valves will maintain their last position when power is removed from the valve operator. With power applied to the valve operators, only the interlocks prevent the valve from being operated. Although operating procedures 1 and interlocks make the occurrence of this event unlikely, the prudent action is to remove power from the loop isolation valve operators. The Completion Time of 30 minutes to remove power from the loop isolation valve operators is sufficient considering the complexity of the task.

B.1, B.2, and B.3 Should a loop isolation valve be closed in MODES 1 through 4, the affected loop must be fully isolated immediately and the plant placed in MODE 5. Once in MODE 5, the isolated loop may be started in a controlled manner in accordance with LCO 3.4.18, "RCS Isolated Loop Startup." Opening the closed isolation valve in MODES 1 through 4 could result in colder water or water at a lower boron concentration being mixed with the operating RCS loops resulting in positive reactivity insertion. The Completion Time of Required Action B.1 allows time for borating the operating loops to a shutdown boration level such that the plant can be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

WOG STS B 3.4.17-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 456 of 478

Attachment 1, Volume 9, Rev. 1, Page 457 of 478 RCS Loop Isolation Valves B 3.4.17 BASES SURVEILLANCE SR 3.4.17.1 REQUIREMENTS The Surveillance is performed at least once per 31 days to ensure that the RCS loop isolation valves are open, with power removed from the loop isolation valve operators. The primary function of this Surveillance is to ensure that power is removed from the valve operators, since SR 3.4.4.1 of LCO 3.4.4, "RCS Loops - MODES 1 and 2," ensures that the loop isolation valves are open by verifying every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that all loops are operating and circulating reactor coolant. The Frequency of 31 days ensures that the required flow can be made available, is based on engineering judgment, and has proven to be acceptable. Operating experience has shown that the failure rate is so low that the 31 day Frequency is justified.

REFERENCES 1. FSAR, Section [15.2.6].

1 WOG STS B 3.4.17-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 457 of 478

Attachment 1, Volume 9, Rev. 1, Page 458 of 478 JUSTIFICATION FOR DEVIATIONS ISTS 3.4.17 BASES, REACTOR COOLANT SYSTEM (RCS) LOOP ISOLATION VALVES

1. ISTS 3.4.17 Bases, "Reactor Coolant System (RCS) Loop Isolation Valves," is not included in the Kewaunee Power Station (KPS) ITS since the Specification has not been included in the KPS ITS.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 458 of 478

Attachment 1, Volume 9, Rev. 1, Page 459 of 478 ISTS 3.4.18, REACTOR COOLANT SYSTEM (RCS) ISOLATED LOOP STARTUP Attachment 1, Volume 9, Rev. 1, Page 459 of 478

Attachment 1, Volume 9, Rev. 1, Page 460 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 460 of 478

Attachment 1, Volume 9, Rev. 1, Page 461 of 478 RCS Isolated Loop Startup 3.4.18 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 RCS Isolated Loop Startup LCO 3.4.18 Each RCS isolated loop shall remain isolated with:

a. The hot and cold leg isolation valves closed if boron concentration of the isolated loop is less than boron concentration required to meet the SDM of LCO 3.1.1 or boron concentration of LCO 3.9.1 and

b. The cold leg isolation valve closed if the cold leg temperature of the isolated loop is > [20]°F below the highest cold leg temperature of the operating loops.

APPLICABILITY: MODES 5 and 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Isolated loop hot or cold A.1 --------------NOTE--------------

leg isolation valve open Only required if boron with LCO requirements concentration requirement 1

not met. not met.

Close hot and cold leg Immediately isolation valves.

OR A.2 --------------NOTE--------------

Only required if temperature requirement not met.

Close cold leg isolation Immediately valve.

WOG STS 3.4.18-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 461 of 478

Attachment 1, Volume 9, Rev. 1, Page 462 of 478 RCS Isolated Loop Startup 3.4.18 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.18.1 Verify cold leg temperature of isolated loop is Within 30 minutes

[20]°F below the highest cold leg temperature of prior to opening the operating loops. the cold leg isolation valve in isolated loop SR 3.4.18.2 Verify boron concentration of isolated loop is greater Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months than or equal to the boron concentration required to prior to opening meet the SDM of LCO 3.1.1 or boron concentration the hot or cold leg of LCO 3.9.1. isolation valve in isolated loop 1

WOG STS 3.4.18-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 462 of 478

Attachment 1, Volume 9, Rev. 1, Page 463 of 478 JUSTIFICATION FOR DEVIATIONS ISTS 3.4.18, REACTOR COOLANT SYSTEM (RCS) ISOLATED LOOP STARTUP

1. ISTS 3.4.18, "Reactor Coolant System (RCS) Isolated Loop Startup," is not included in the Kewaunee Power Station (KPS) ITS because the Reactor Coolant System hot and cold loops do not include isolation valves.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 463 of 478

Attachment 1, Volume 9, Rev. 1, Page 464 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 464 of 478

Attachment 1, Volume 9, Rev. 1, Page 465 of 478 RCS Isolated Loop Startup B 3.4.18 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.18 RCS Isolated Loop Startup BASES BACKGROUND The RCS may be operated with loops isolated in MODES 5 and 6 in order to perform maintenance. While operating with a loop isolated, there is potential for inadvertently opening the isolation valves in the isolated loop.

In this event, the coolant in the isolated loop would suddenly begin to mix with the coolant in the operating loops. This situation has the potential of causing a positive reactivity addition with a corresponding reduction of SDM if:

a. The temperature in the isolated loop is lower than the temperature in the operating loops (cold water incident) or

b. The boron concentration in the isolated loop is lower than the boron concentration required to meet the SDM of LCO 3.1.1 or boron concentration of LCO 3.9.1 (boron dilution incident).

As discussed in the FSAR (Ref. 1), the startup of an isolated loop is done in a controlled manner that virtually eliminates any sudden reactivity addition from cold water or boron dilution because: 1

a. This LCO and plant operating procedures require that the boron concentration in the isolated loop be maintained higher than the boron concentration of the operating loops, thus eliminating the potential for introducing coolant from the isolated loop that could dilute the boron concentration in the operating loops,

b. The cold leg loop isolation valve cannot be opened unless the temperatures of both the hot leg and cold leg of the isolated loop are within 20°F of the operating loops. Compliance with the temperature requirement is ensured by operating procedures and automatic interlocks, and

c. Other automatic interlocks prevent opening the hot leg loop isolation valve unless the cold leg loop isolation valve is fully closed. All of the interlocks are part of the Reactor Protection System.

WOG STS B 3.4.18-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 465 of 478

Attachment 1, Volume 9, Rev. 1, Page 466 of 478 RCS Isolated Loop Startup B 3.4.18 BASES APPLICABLE During startup of an isolated loop, the cold leg loop isolation valve SAFETY interlocks and operating procedures prevent opening the valve until the ANALYSES isolated loop and operating loop boron concentrations and temperatures are equalized. This ensures that any undesirable reactivity effect from the isolated loop does not occur.

The safety analyses assume a minimum SDM as an initial condition for Design Basis Accidents. Violation of this LCO could result in the SDM being reduced in the operating loops to less than that assumed in the safety analyses.

The boron concentration of an isolated loop may affect SDM and therefore RCS isolated loop startup satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO Loop isolation valves are used for performing maintenance when the plant is in MODE 5 or 6. This LCO ensures that the loop isolation valves remain closed until the differentials of temperature and boron concentration between the operating loops and the isolated loops are within acceptable limits.

APPLICABILITY In MODES 5 and 6, the SDM of the operating loops is large enough to permit operation with isolated loops. Controlled startup of isolated loops 1 is possible without significant risk of inadvertent criticality. This LCO is applicable under these conditions.

ACTIONS A.1 and A.2 Required Action A.1 and Required Action A.2 assume that the prerequisites of the LCO are not met and a loop isolation valve has been inadvertently opened. Therefore, the Actions require immediate closure of isolation valves to preclude a boron dilution event or a cold water event. However, each Required Action is preceded by a Note that states that Action is required only when a specific concentration or temperature requirement is not met.

SURVEILLANCE SR 3.4.18.1 REQUIREMENTS This Surveillance is performed to ensure that the temperature differential between the isolated loop and the operating loops is [20]°F. Performing the Surveillance 30 minutes prior to opening the cold leg isolation valve in the isolated loop provides reasonable assurance, based WOG STS B 3.4.18-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 466 of 478

Attachment 1, Volume 9, Rev. 1, Page 467 of 478 RCS Isolated Loop Startup B 3.4.18 BASES SURVEILLANCE REQUIREMENTS (continued) on engineering judgment, that the temperature differential will stay within limits until the cold leg isolation valve is opened. This Frequency has been shown to be acceptable through operating experience.

SR 3.4.18.2 To ensure that the boron concentration of the isolated loop is greater than or equal to the boron concentration required to meet the SDM of LCO 3.1.1 or boron concentration of LCO 3.9.1, a Surveillance is performed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to opening either the hot or cold leg isolation valve. Performing the Surveillance 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to opening either the hot or cold leg isolation valve provides reasonable assurance the boron concentration difference will stay within acceptable limits until the loop is unisolated. This Frequency has been shown to be acceptable through operating experience.

REFERENCES 1. FSAR, Section [15.2.6].

1 WOG STS B 3.4.18-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 467 of 478

Attachment 1, Volume 9, Rev. 1, Page 468 of 478 JUSTIFICATION FOR DEVIATIONS ISTS 3.4.18 BASES, REACTOR COOLANT SYSTEM (RCS) ISOLATED LOOP STARTUP

1. ISTS 3.4.18 Bases, "Reactor Coolant System (RCS) Isolated Loop Startup," is not included in the Kewaunee Power Station (KPS) ITS since the Specification has not been included in the KPS ITS.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 468 of 478

Attachment 1, Volume 9, Rev. 1, Page 469 of 478 ISTS 3.4.19, REACTOR COOLANT SYSTEM (RCS) LOOPS -

TEST EXCEPTIONS Attachment 1, Volume 9, Rev. 1, Page 469 of 478

Attachment 1, Volume 9, Rev. 1, Page 470 of 478 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 470 of 478

Attachment 1, Volume 9, Rev. 1, Page 471 of 478 RCS Loops - Test Exceptions 3.4.19 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.19 RCS Loops - Test Exceptions LCO 3.4.19 The requirements of LCO 3.4.4, "RCS Loops - MODES 1 and 2," may be suspended with THERMAL POWER < P-7.

APPLICABILITY: MODES 1 and 2 during startup and PHYSICS TESTS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. THERMAL POWER A.1 Open reactor trip breakers. Immediately P-7.

1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.19.1 Verify THERMAL POWER is < P-7. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SR 3.4.19.2 Perform a COT for each power range neutron flux - Prior to initiation low channel, intermediate range neutron flux of startup and channel, P-10, and P-13. PHYSICS TESTS SR 3.4.19.3 Perform an ACTUATION LOGIC TEST on P-7. Prior to initiation of startup and PHYSICS TESTS WOG STS 3.4.19-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 471 of 478

Attachment 1, Volume 9, Rev. 1, Page 472 of 478 JUSTIFICATION FOR DEVIATIONS ISTS 3.4.19, REACTOR COOLANT SYSTEM (RCS) LOOPS - TEST EXCEPTIONS

1. ISTS 3.4.19, "Reactor Coolant System (RCS) Loops - Test Exceptions," is not included in the Kewaunee Power Station (KPS) ITS because the exception is not needed to perform any required startup or PHYSICS TESTS.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 472 of 478

Attachment 1, Volume 9, Rev. 1, Page 473 of 478 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 9, Rev. 1, Page 473 of 478

Attachment 1, Volume 9, Rev. 1, Page 474 of 478 RCS Loops - Test Exceptions B 3.4.19 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.19 RCS Loops - Test Exceptions BASES BACKGROUND The primary purpose of this test exception is to provide an exception to LCO 3.4.4, "RCS Loops - MODES 1 and 2," to permit reactor criticality under no flow conditions during certain PHYSICS TESTS (natural circulation demonstration, station blackout, and loss of offsite power) to be performed while at low THERMAL POWER levels.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 power plant as specified in GDC 1, "Quality Standards and Records" (Ref. 2).

The key objectives of a test program are to provide assurance that the facility has been adequately designed to validate the analytical models used in the design and analysis, to verify the assumptions used to predict plant response, to provide assurance that installation of equipment at the unit has been accomplished in accordance with the design, and to verify 1 that the operating and emergency procedures are adequate. Testing is performed prior to initial criticality, during startup, and following low power operations.

The tests will include verifying the ability to establish and maintain natural circulation following a plant trip between 10% and 20% RTP, performing natural circulation cooldown on emergency power, and during the cooldown, showing that adequate boron mixture occurs and that pressure can be controlled using auxiliary spray and pressurizer heaters powered from the emergency power sources.

APPLICABLE The tests described above require operating the plant without forced SAFETY convection flow and as such are not bounded by any safety analyses.

ANALYSES However, operating experience has demonstrated this exception to be safe under the present applicability.

As describe 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.

WOG STS B 3.4.19-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 474 of 478

Attachment 1, Volume 9, Rev. 1, Page 475 of 478 RCS Loops - Test Exceptions B 3.4.19 BASES LCO This LCO provides an exemption to the requirements of LCO 3.4.4.

The LCO is provided to allow for the performance of PHYSICS TESTS in MODE 2 (after a refueling), where the core cooling requirements are significantly different than after the core has been operating. Without the LCO, plant operations would be held bound to the normal operating LCOs for reactor coolant loops and circulation (MODES 1 and 2), and the appropriate tests could not be performed.

In MODE 2, where core power level is considerably lower and the associated PHYSICS TESTS must be performed, operation is allowed under no flow conditions provided THERMAL POWER is P-7 and the reactor trip setpoints of the OPERABLE power level channels are set 25% RTP. This ensures, if some problem caused the plant to enter MODE 1 and start increasing plant power, the Reactor Trip System (RTS) would automatically shut it down before power became too high, and thereby prevent violation of fuel design limits.

The exemption is allowed even though there are no bounding safety analyses. However, these tests are performed under close supervision during the test program and provide valuable information on the plant's capability to cool down without offsite power available to the reactor coolant pumps. 1 APPLICABILITY This LCO is applicable when performing low power PHYSICS TESTS without any forced convection flow. This testing is performed to establish that heat input from nuclear heat does not exceed the natural circulation heat removal capabilities. Therefore, no safety or fuel design limits will be violated as a result of the associated tests.

ACTIONS A.1 When THERMAL POWER is the P-7 interlock setpoint 10%, the only acceptable action is to ensure the reactor trip breakers (RTBs) are opened immediately in accordance with Required Action A.1 to prevent operation of the fuel beyond its design limits. Opening the RTBs will shut down the reactor and prevent operation of the fuel outside of its design limits.

WOG STS B 3.4.19-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 475 of 478

Attachment 1, Volume 9, Rev. 1, Page 476 of 478 RCS Loops - Test Exceptions B 3.4.19 BASES SURVEILLANCE SR 3.4.19.1 REQUIREMENTS Verification that the power level is < the P-7 interlock setpoint (10%) will ensure that the fuel design criteria are not violated during the performance of the PHYSICS TESTS. The Frequency of once per hour is adequate to ensure that the power level does not exceed the limit. Plant operations are conducted slowly during the performance of PHYSICS TESTS and monitoring the power level once per hour is sufficient to ensure that the power level does not exceed the limit.

SR 3.4.19.2 The power range and intermediate range neutron detectors. P-10, and the P-13 interlock setpoint must be verified to be OPERABLE and adjusted to the proper value. The Low Power Reactor Trips Block, P-7 interlock, is actuated from either the Power Range Neutron Flux, P-10, or the Turbine Impulse Chamber Pressure, P-13 interlock. The P-7 interlock is a logic Function with train, not channel identity. A COT is performed 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. A successful test of the required contact(s) of a channel relay may be performed by the 1 verification of the change of state of a single contact of the relay. This clarifies what is an acceptable COT of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The SR 3.3.1.8 Frequency is sufficient for the power range and intermediate range neutron detectors to ensure that the instrumentation is OPERABLE before initiating PHYSICS TESTS.

SR 3.4.19.3 The Low Power Reactor Trips Block, P-7 interlock, must be verified to be OPERABLE in MODE 1 by LCO 3.3.1, Reactor Trip System Instrumentation. The P-7 interlock is actuated from either the Power Range Neutron Flux, P-10, or the Turbine Impulse Chamber Pressure, P-13 interlock. The P-7 interlock is a logic Function. An ACTUATION LOGIC TEST is performed to verify OPERABILITY of the P-7 interlock prior to initiation of startup and PHYSICS TESTS. This will ensure that the RTS is properly functioning to provide the required degree of core protection during the performance of the PHYSICS TESTS.

WOG STS B 3.4.19-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 476 of 478

Attachment 1, Volume 9, Rev. 1, Page 477 of 478 RCS Loops - Test Exceptions B 3.4.19 BASES REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50, Appendix A, GDC 1, 1988.

1 WOG STS B 3.4.19-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 1, Page 477 of 478

Attachment 1, Volume 9, Rev. 1, Page 478 of 478 JUSTIFICATION FOR DEVIATIONS ISTS 3.4.19 BASES, REACTOR COOLANT SYSTEM (RCS) LOOPS - TEST EXCEPTIONS

1. ISTS 3.4.19 Bases, "Reactor Coolant System (RCS) Loops - Test Exceptions,"

is not included in the Kewaunee Power Station (KPS) ITS since the Specification has not been included in the KPS ITS.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 9, Rev. 1, Page 478 of 478

ML101890218

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1.8 REFERENCES

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