Text

Attachment 1, Volume 14, Kewaunee Power Station Improved Technical Specifications Conversion, ITS Section 3.9 Refueling Operations, Revision 1
	ML102371178
Person / Time
Site:	Kewaunee
Issue date:	08/18/2010
From:	; Dominion Energy Kewaunee
To:	; Office of Nuclear Reactor Regulation
References
	10-457, TAC ME2139
	Download: ML102371178 (182)
	v • d • e

Attachment 1, Volume 14, Rev. 1, Page i of i Summary of Changes ITS Section 3.9 Change Description Affected Pages The changes described in the KPS response to Pages 5, 6, 27, 28, 50, 51, 55, 60, 61, 66, 67, 76, question RPG-010 have been made. This change 82, 83, 84, 89, 91, 96, 97, 110 through 135, 139, adds ITS 3.9.6, Containment Penetrations into the 146, 152, 159, and 167 KPS ITS submittal. Note that minor editorial changes were made to DOC L02 that are not shown in the version loaded on the website. Specifically, in the second paragraph, next to last sentence, the word "additional" has been deleted and the word "available" has been changed to "specified."

The changes described in the KPS response to Pages 7, 8, 9, 10, 11, 14, 15, 17, 19, 21, and 22 question MEH-004 have been made. This change adds the spent fuel pool boron concentration requirements in MODE 6 to ITS 3.9.1.

Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page i of i

Attachment 1, Volume 14, Rev. 1, Page 1 of 181 ATTACHMENT 1 VOLUME 14 KEWAUNEE POWER STATION IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.9 REFUELING OPERATIONS Revision 1 Attachment 1, Volume 14, Rev. 1, Page 1 of 181

Attachment 1, Volume 14, Rev. 1, Page 2 of 181 LIST OF ATTACHMENTS

1. ITS 3.9.1

2. ITS 3.9.2

3. ITS 3.9.3

4. ITS 3.9.4

5. ITS 3.9.5

6. ITS 3.9.6

7. Relocated/Deleted Current Technical Specifications (CTS)

8. ISTS Not Adopted Attachment 1, Volume 14, Rev. 1, Page 2 of 181

, Volume 14, Rev. 1, Page 3 of 181 ATTACHMENT 1 ITS 3.9.1, BORON CONCENTRATION , Volume 14, Rev. 1, Page 3 of 181

, Volume 14, Rev. 1, Page 4 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 4 of 181

Attachment 1, Volume 14, Rev. 1, Page 5 of 181 A01 ITS ITS 3.9.1 3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS.

OBJECTIVE To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION Applicability a. During REFUELING OPERATIONS: M01 Add proposed Applicability Note

1. Containment Closure M02

a. The equipment hatch shall be closed and at least one door in each personnel air lock shall be capable of being closed (1) in 30 minutes or less. In addition, at See ITS least one door in each personnel air lock shall be closed when the reactor vessel 3.9.6 head or upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valve or an operable automatic isolation valve.

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage See CTS 3.8.a.2 pool shall be monitored continuously.

See CTS

3. The reactor will be subcritical for 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months prior to movement of its irradiated fuel 3.8.a.3 assemblies. Core subcritical neutron flux shall be continuously monitored by at least See ITS two neutron monitors, each with continuous visual indication in the control room and 3.9.2 one with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux monitor shall be in service. See ITS 3.9.3

4. At least one residual heat removal pump shall be OPERABLE.

M01 Applicability

5. When there is fuel in the reactor, a minimum boron concentration as specified in the M02 LCO 3.9.1 COLR shall be maintained in the Reactor Coolant System during reactor vessel Applicability head removal or while loading and unloading fuel from the reactor. The required M01 SR 3.9.1.1 boron concentration shall be verified by chemical analysis daily. every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months L01 LA01 (1)

Administrative controls ensure that:

Appropriate personnel are aware that both personnel air lock doors are open,

A specified individual(s) is designated and available to close the air lock following a required See ITS evacuation of containment, and 3.9.6

Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed.

Amendment No. 165 TS 3.8-1 03/11/2003 Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 5 of 181

Attachment 1, Volume 14, Rev. 1, Page 6 of 181 ITS ITS 3.9.1 A01

6. Direct communication between the control room and the operating floor of the See CTS containment shall be available whenever changes in core geometry are taking 3.8.a.6 place.

7. Deleted.

8. The containment ventilation and purge system, including the capability to initiate See ITS automatic containment ventilation isolation, shall be tested and verified to be 3.9.6 operable immediately prior to and daily during REFUELING OPERATIONS.

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be operating during fuel handling and when any load is carried over the pool if irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool See CTS 3.8.a.9 sweep system, including the charcoal adsorber, is not operating when required, fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

See ITS

10. The minimum water level above the vessel flange shall be maintained at 23 feet. 3.9.5

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes See CTS for this test must be equal to or greater than the maximum load to be assumed by 3.8.a.11 the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

See CTS

12. A licensed senior reactor operator will be on-site and designated in charge of the 3.8.a.12 REFUELING OPERATIONS.

b. If any of the specified limiting conditions for REFUELING OPERATIONS are not met, ACTION A refueling of the reactor shall cease. Work shall be initiated to correct the violated conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be performed.

Amendment No. 200 TS 3.8-2 11/20/2008 Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 6 of 181

ITS A01 ITS 3.9.1 See CTS TABLE TS 4.1-2 3.1.e See ITS MINIMUM FREQUENCIES FOR SAMPLING TESTS 3.4.16 SAMPLING TESTS TEST FREQUENCY (1)

1. Reactor Coolant a. Gross Radioactivity Determination (excluding tritium) 5/week Samples b. DOSE EQUIVALENT I-131 Concentration 1/14 days(2)

c. Tritium activity Monthly

d. Chemistry (Cl, F, O2)(3) 3/week(4)

e. Determination 1/6 months(5)

f. RCS isotopic analysis for Iodine Once per 4 hours4.62963e-5 days 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 L01 See ITS SR 3.9.1.1 3.4.16 , Volume 14, Rev. 1, Page 7 of 181 Attachment 1, Volume 14, Rev. 1, Page 7 of 181 (1)

Maximum time between tests is 3 days. See ITS (2) 3.4.16 Sample required only when in the OPERATING MODE.

(3) A02 See CTS See ITS Test required in all plant modes. 3.1.e 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.

A02 Amendment No. 119 Page 1 of 2 04/18/95 Page 3 of 4

Attachment 1, Volume 14, Rev. 1, Page 8 of 181 ITS A01 ITS 3.9.1 5.4 FUEL STORAGE APPLICABILITY Applies to the capacity and storage arrays of new and spent fuel.

OBJECTIVE To define those aspects of fuel storage relating to prevention of criticality in fuel storage areas.

SPECIFICATION

a. Criticality

1. The spent fuel storage racks are designed and shall be maintained with the following:

a. Fuel assemblies having a maximum enrichment of 56.067 grams Uranium-235 per axial centimeter

b. keff < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties See ITS 4.0

2. The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum enrichment of 56.067 grams Uranium-235 per axial centimeter

b. keff < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties

c. keff < 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties LCO 3.9.1

3. The spent fuel pool is filled with borated water at a concentration to match that used A03 in the reactor REFUELING cavity and REFUELING canal during REFUELING OPERATIONS or whenever there is fuel in the pool.

Applicability See ITS 3.7.14

b. Capacity See ITS The spent fuel storage pool is designed with a storage capacity of 1205 assemblies and 4.0 shall be limited to no more than 1205 fuel assemblies.

c. Canal Rack Storage Fuel assemblies stored in the canal racks shall meet the minimum required fuel See ITS 3.7.15 assembly burnup as a function of nominal initial enrichment as shown in Figure TS 5.4-1. These assemblies shall also have been discharged prior to or during the 1984 REFUELING outage.

Add proposed ACTION A M03 Amendment No. 162 TS 5.4-1 09/19/2002 Page 4 of 4 Attachment 1, Volume 14, Rev. 1, Page 8 of 181

Attachment 1, Volume 14, Rev. 1, Page 9 of 181 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION 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 Table TS 4.1-2 Sample Test 2 footnote (3) states that the boron concentration test is required in all plant modes. CTS Table TS 4.1-2 Sample Test 2 footnote (6) states that a reactor coolant boron concentration sample does not have to be taken when the core is completely unloaded. ITS 3.9.1 is applicable in MODE 6. This changes the CTS by specifically stating that the Applicability is in MODE 6. In addition, ITS 3.1.1, "SHUTDOWN MARGIN (SDM)" discusses the remaining plant modes.

This change is acceptable because the Applicability has not changed. ITS 3.9.1 covers the MODE 6 requirements only. Other MODES are covered by ITS 3.1.1.

This change results in a format change only to comply with the ISTS presentation of the Applicability. Therefore, the change is acceptable because the boron concentration requirements have not changed. This change is designated as administrative since it does not result in any technical changes to the CTS.

A03 CTS 5.4.a.3, in part, requires the spent fuel pool boron concentration to match that used in the reactor refueling cavity and refueling canal during REFUELING OPERATIONS. CTS 3.8.a.5 provides the boron concentration limit for the RCS during fuel movement, and states that the limit is specified in the COLR.

ITS 3.9.1, in part, requires the boron concentration of the spent fuel pool to be within the limit specified in the COLR during MODE 6 operations (which encompass refueling operations), but a NOTE to the Applicability limits this requirement to only when the spent fuel pool is connected to the RCS. This changes the CTS by clearly stating when the spent fuel pool limit is to be maintained consistent with the RCS limit.

The purpose of CTS 5.4.a.3, in part, is to ensure the spent fuel pool boron is within the limits required by the RCS during REFUELING OPERATIONS. Since REFUELING OPERATIONS can only be achieved when the spent fuel pool is connected to the RCS, the added clarification in the Applicability is not changing the current requirements, with respect to REFUELING OPERATIONS.

Therefore, this change is acceptable and is designated as administrative since the technical requirements are not changing. Note that the spent fuel pool boron concentration limit when not connected to the RCS is governed by ITS 3.7.14.

MORE RESTRICTIVE CHANGES M01 CTS 3.8.a.5 is applicable during REFUELING OPERATIONS. CTS 3.8.a.5 also states, in part, that a minimum boron concentration shall be maintained when Kewaunee Power Station Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 9 of 181

Attachment 1, Volume 14, Rev. 1, Page 10 of 181 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION there is fuel in the reactor and during reactor vessel head removal or while loading and unloading fuel from the reactor. ITS 3.9.1 is applicable at all times while in MODE 6. This changes the CTS by requiring the boron concentration be maintained at all times while in MODE 6 and not just during those refueling activities/conditions contained within the CTS.

The purpose of CTS 3.8.a.5 is to ensure the boron concentration of the Reactor Coolant System (RCS) is sufficient to maintain the reactor subcritical during REFUELING OPERATIONS. As defined in CTS Section 1.0, REFUELING OPERATIONS is movement of reactor vessel internal components that could affect the reactivity of the core within the containment when the vessel head is unbolted or removed. CTS 3.8.a.5 also requires the boron concentration be maintained during reactor vessel head removal and when there is fuel in the reactor. ITS 3.9.1 requires the boron concentration of the RCS, fuel transfer canal, and refueling cavity be maintained at all times while in MODE 6, with the exception of the conditions listed in the Applicability NOTE. MODE 6 is defined as when one or more reactor vessel head closure bolts are less than fully tensioned. As a result, MODE 6 encompasses all the aforementioned CTS conditions and, in addition, those times when there is no movement of reactor internal components. This change is acceptable because the requirements continue to ensure that process variables are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis.

This change is designated more restrictive because the boron concentration requirements for ITS are more restrictive than the conditions stated in the CTS.

M02 CTS 3.8.a.5 states, in part, that a minimum boron concentration as specified in the COLR shall be maintained in the Reactor Coolant System. ITS LCO 3.9.1 states, in part, that the boron concentrations of the Reactor Coolant System (RCS), the fuel transfer canal, and the refueling cavity shall be maintained within the limit specified in the COLR. ITS 3.9.1 Applicability contains a NOTE that states the LCO is only applicable to the fuel transfer canal and the refueling cavity when connected to the RCS. This changes the CTS by including the boron concentration of the fuel transfer canal and the refueling cavity in the LCO and adds a NOTE to the Applicability stating that the LCO is only applicable to the fuel transfer canal and the refueling cavity when connected to the RCS.

The purpose of CTS 3.8.a.5 is to ensure the boron concentration of the water surrounding the reactor fuel is sufficient to maintain the reactor subcritical during refueling. When the reactor head is removed for refueling of the reactor, the refueling cavity is open to the reactor itself. The refueling cavity is flooded with borated water from the Refueling Water Storage Tank and when the cavity is filled, the fuel transfer canal, refueling cavity, and the reactor vessel all share the same volume of water. As a result, the soluble boron concentration is relatively the same in each of these volumes. The NOTE to ITS 3.9.1 Applicability states that the limits on boron concentration are only applicable to the fuel transfer canal and the refueling cavity when those volumes are connected to the RCS.

When the fuel transfer canal and the refueling cavity are isolated from the RCS (i.e., the reactor vessel head is on the vessel), no potential path for boron dilution exists. In addition, prior to connecting the refueling cavity and the fuel transfer canal to the RCS, a boron concentration verification will be performed (as required by SR 3.0.4) to ensure the newly connected portions cannot decrease Kewaunee Power Station Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 10 of 181

Attachment 1, Volume 14, Rev. 1, Page 11 of 181 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION the boron concentration below the limit. These changes are acceptable because the requirements continue to ensure that process variables are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. This change is designated more restrictive because the LCO requirements are applicable in more operating conditions than in the CTS.

M03 CTS 5.4.a.3 does not provide any ACTIONS to take when the spent fuel pool boron concentration is not within limit during REFUELING OPERATIONS. When the spent fuel pool boron concentration is not within limit, ITS 3.9.1 ACTION A requires the immediate suspension of positive reactivity additions and immediate action to restore the boron concentration to within limit. This changes the CTS by providing specific ACTIONS when the spent fuel pool boron concentration is not within limit during refueling operations (i.e., in MODE 6 when connected to the RCS).

The purpose of CTS 5.4.a.3 is to ensure adequate dissolved boron is in the spent fuel pool water to maintain the required subcriticality margin in the RCS, when connected to the RCS during MODE 6 (i.e., refueling operations). ITS 3.7.14 ACTION A effectively places the unit outside of the Applicability by requiring the plant to immediately suspend any operations that would further decrease the subcriticality margin and to initiate actions to restore the boron concentration to within limits. The proposed Required Actions reflect the importance of maintaining the boron concentration within the required limit. This change is designated more restrictive because a new proposed ACTION has been added.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.8.a.5 requires the boron concentration of the Reactor Coolant System be determined "by chemical analysis" daily. ITS SR 3.9.1.1 requires verification that boron concentration is within the limit specified in the COLR. ITS SR 3.9.1.1 does not specify that the boron concentration be determined by chemical analysis. This changes the CTS by moving details of how the boron concentration is determined from the CTS to the Bases. The discussion of the change from daily to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is provided in DOC L01.

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. The ITS still retains the requirement that the boron concentration be verified within its 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 Kewaunee Power Station Page 3 of 4 Attachment 1, Volume 14, Rev. 1, Page 11 of 181

Attachment 1, Volume 14, Rev. 1, Page 12 of 181 DISCUSSION OF CHANGES ITS 3.9.1, BORON CONCENTRATION 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 7 - Relaxation of Surveillance Frequency) CTS 3.8.a.5 requires the boron concentration be verified by chemical analysis daily. CTS Table TS 4.1-2 Sample Test 2 requires a test of the boron concentration twice per week. ITS SR 3.9.1.1 requires verification that the boron concentration is within the limit specified in the COLR every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This changes the CTS by changing the Surveillance Frequency from daily and twice per week (with no specific time between performances specified) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The purpose of CTS 3.8.a.5 and CTS Table TS 4.1-2 Sample Test 2 is to provide assurance that the coolant boron concentration in the RCS and connected portions of the fuel transfer canal and the refueling cavity is within limits. The boron concentration limit specified in the COLR ensures that a core keff of 0.95 is maintained during operations in MODE 6. This change is acceptable since the new Surveillance Frequency is a reasonable amount of time to verify the boron concentration of representative samples. This change is designated as less restrictive because a Surveillance will be performed less frequently under the ITS than under the CTS.

Kewaunee Power Station Page 4 of 4 Attachment 1, Volume 14, Rev. 1, Page 12 of 181

Attachment 1, Volume 14, Rev. 1, Page 13 of 181 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 13 of 181

Attachment 1, Volume 14, Rev. 1, Page 14 of 181 CTS Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration (RCS) fuel transfer 1 2 3.8.a.5, LCO 3.9.1 Boron concentrations of the Reactor Coolant System, the refueling canal, 5.4.a.3 and the refueling cavity shall be maintained within the limit specified in the 3 COLR. , and the spent fuel pool 3.8.a, 3.8.a.5, APPLICABILITY: MODE 6.

5.4.a.3

NOTE-------------------------------------------- , and the spent DOC M02, fuel pool 5.4.a.3 Only applicable to the refueling canal and refueling cavity when connected to the RCS. fuel transfer , 2 3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.8.b, A. Boron concentration not A.1 Suspend CORE Immediately DOC M03 within limit. ALTERATIONS.

AND TSTF-A.2 1 Suspend positive reactivity Immediately 471 additions.

AND A.3 2 Initiate action to restore Immediately boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 3.8.a.5 SR 3.9.1.1 Verify boron concentration is within the limit 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months specified in the COLR.

WOG STS 3.9.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 14 of 181

Attachment 1, Volume 14, Rev. 1, Page 15 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.1, BORON CONCENTRATION

1. Editorial change made for consistency.

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

3. The spent fuel pool boron concentration requirement when the spent fuel pool is connected to the RCS has been added to the LCO 3.9.1 statement for clarity and ease of use. CTS 5.4.a.3 requires the boron concentration to be consistent with the RCS during refueling operations. Since refueling operations can only take place when the spent fuel pool is connected, this allowance has also been added into the Applicability Note.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 15 of 181

Attachment 1, Volume 14, Rev. 1, Page 16 of 181 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 16 of 181

Attachment 1, Volume 14, Rev. 1, Page 17 of 181 All changes are 1 Boron Concentration B 3.9.1 unless otherwise noted B 3.9 REFUELING OPERATIONS B 3.9.1 Boron Concentration BASES BACKGROUND The limit on the boron concentrations of the Reactor Coolant System , and the spent 6

fuel pool fuel transfer (RCS), the refueling canal, and the refueling cavity during refueling ensures that the reactor remains subcritical during MODE 6. Refueling boron concentration is the soluble boron concentration in the coolant in each of these volumes having direct access to the reactor core during refueling.

The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of the volumes. The refueling boron concentration limit is specified in the COLR. Plant procedures ensure the specified boron concentration in order to maintain an overall core reactivity of keff 0.95 during fuel handling, with control rods and fuel assemblies assumed to be in the most adverse configuration (least negative reactivity) allowed by plant procedures. Updated Safety Analysis Report (USAR), General Design Criteria (GDC) 27, "Redundancy of Reactivity Control" preferably GDC 26 of 10 CFR 50, Appendix A, requires that two independent shall reactivity control systems of different design principles be provided 2

(Ref. 1). One of these systems must be capable of holding the reactor core subcritical under cold conditions. The Chemical and Volume Control INSERT 1 System (CVCS) is the system capable of maintaining the reactor subcritical in cold conditions by maintaining the boron concentration.

The reactor is brought to shutdown conditions before beginning operations to open the reactor vessel for refueling. After the RCS is and cooled and depressurized and the vessel head is unbolted, the head is is slowly removed to form the refueling cavity. The refueling canal and the refueling cavity are then flooded with borated water from the refueling water storage tank through the open reactor vessel by gravity feeding or by the use of the Residual Heat Removal (RHR) System pumps.

The pumping action of the RHR System in the RCS and the natural circulation due to thermal driving heads in the reactor vessel and refueling fuel transfer cavity mix the added concentrated boric acid with the water in the refueling canal. The RHR System is in operation during refueling (see 4

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High 3 Water Level," and LCO 3.9.6, "Residual Heat Removal (RHR) and 5

Coolant Circulation - Low Water Level") to provide forced circulation in the RCS and assist in maintaining the boron concentrations in the RCS, the refueling canal, and the refueling cavity above the COLR limit. 6 fuel transfer , and the spent fuel pool WOG STS B 3.9.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 17 of 181

Attachment 1, Volume 14, Rev. 1, Page 18 of 181 B 3.9.1 2

INSERT 1 USAR GDC 30, "Reactivity Holddown Capability," requires that the reactivity control system provided shall be capable of making the core subcritical under credible accident conditions with appropriate margins for contingencies and limiting any subsequent return to power such that there will be no undue risk to the health and safety of the public (Ref.

2). Refueling boron concentration is sufficient to maintain the clean, cold fully loaded core subcritical with all rod cluster control assemblies withdrawn.

Insert Page B 3.9.1-1 Attachment 1, Volume 14, Rev. 1, Page 18 of 181

Attachment 1, Volume 14, Rev. 1, Page 19 of 181 All changes are 1 Boron Concentration unless otherwise noted B 3.9.1 BASES APPLICABLE During refueling operations, the reactivity condition of the core is SAFETY consistent with the initial conditions assumed for the boron dilution ANALYSES accident in the accident analysis and is conservative for MODE 6. The boron concentration limit specified in the COLR is based on the core reactivity at the beginning of each fuel cycle (the end of refueling) and includes an uncertainty allowance.

The required boron concentration and the plant refueling procedures that verify the correct fuel loading plan (including full core mapping) ensure that the keff of the core will remain 0.95 during the refueling operation.

Hence, at least a 5% k/k margin of safety is established during refueling.

fuel During refueling, the water volume in the spent fuel pool, the transfer canal, the refueling canal, the refueling cavity, and the reactor vessel form a single mass. As a result, the soluble boron concentration is relatively the same in each of these volumes.

The limiting boron dilution accident analyzed occurs in MODE 5 (Ref. 2).

4 A detailed discussion of this event is provided in Bases B 3.1.1, "SHUTDOWN MARGIN (SDM)."

The RCS boron concentration satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that a minimum boron concentration be maintained in , and the fuel transfer 6 the RCS, the refueling canal, and the refueling cavity while in MODE 6. spent fuel pool The boron concentration limit specified in the COLR ensures that a core keff of 0.95 is maintained during fuel handling operations. Violation of the LCO could lead to an inadvertent criticality during MODE 6.

APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a keff 0.95. Above MODE 6, LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," ensures that an adequate amount of negative reactivity is available to shut down the reactor and maintain it subcritical.

fuel transfer The Applicability is modified by a Note. The Note states that the limits on boron concentration are only applicable to the refueling canal and the refueling cavity when those volumes are connected to the RCS. When spent , and the fuel pool 6 the refueling canal and the refueling cavity are isolated from the RCS, no potential path for boron dilution exists.

WOG STS B 3.9.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 19 of 181

Attachment 1, Volume 14, Rev. 1, Page 20 of 181 Boron Concentration B 3.9.1 BASES ACTIONS A.1 and A.2 TSTF-471 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon fuel transfer maintaining the unit in compliance with the LCO. If the boron concentration of any coolant volume in the RCS, the refueling canal, or 1 the refueling cavity is less than its limit, all operations involving CORE ALTERATIONS or positive reactivity additions must be suspended TSTF-immediately. 471 Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position. Operations that individually add limited positive reactivity (e.g., temperature fluctuations from inventory addition or temperature control fluctuations),

but when combined with all other operations affecting core reactivity (e.g.,

intentional boration) result in overall net negative reactivity addition, are not precluded by this action.

2 A.3 TSTF-471 In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately.

In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.

has it 5

Once actions have been initiated, they must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.

WOG STS B 3.9.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 20 of 181

Attachment 1, Volume 14, Rev. 1, Page 21 of 181 Boron Concentration B 3.9.1 BASES SURVEILLANCE SR 3.9.1.1 REQUIREMENTS and the spent 6 fuel pool This SR ensures that the coolant boron concentration in the RCS, and 1

connected portions of the refueling canal and the refueling cavity, is within fuel transfer the COLR limits. The boron concentration of the coolant in each required volume is determined periodically by chemical analysis. Prior to re- , 1 connecting portions of the refueling canal or the refueling cavity to the RCS, this SR must be met per SR 3.0.4. If any dilution activity has , or the spent 6 fuel pool occurred while the cavity or canal were disconnected from the RCS, this SR ensures the correct boron concentration prior to communication with the RCS.

A minimum Frequency of once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable amount of time to verify the boron concentration of representative samples. The Frequency is based on operating experience, which has shown 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to be adequate. USAR, Section 3.1.2.3, General Design Criteria (GDC) 27, "Redundancy of Reactivity Control."

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

2. USAR, Section 3.1.2.6, GDC 30, "Reactivity Holddown Capability." 2

2. FSAR, Chapter [15]. 4 WOG STS B 3.9.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 21 of 181

Attachment 1, Volume 14, Rev. 1, Page 22 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.1 BASES, BORON CONCENTRATION

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

2. The ISTS lists GDC 26 of Appendix A to 10 CFR 50 as the reference document for the requirement that there be two independent reactivity control systems of different design principles. Per the information contained in USAR Section 1.8, Kewaunee Power Station (KPS) was designed, constructed, and is being operated to comply with the Atomic Energy Commission (AEC) General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, as proposed on July 10, 1967. Since the plant was approximately 50% complete prior to the February 20, 1971 issuance of 10 CFR 50 Appendix A General Design Criteria, KPS was not required to be reanalyzed and the Final Safety Analysis Report (FSAR) was not required to be revised to reflect these later criteria. However, the AEC Safety Evaluation Report (SER), issued July 24, 1972, acknowledged that the AEC staff assessed the plant, as described in the FSAR (Amendment No. 7), against the Appendix A design criteria and determined that the plant design generally conforms to the intent of the Appendix A criteria. As a result, KPS utilizes AEC GDC 27, Redundancy of Reactivity Control and GDC 30, Reactivity Holddown Capability, as the licensing reference documents for the requirement that there be two independent reactivity control systems of different design principles.

3. ISTS 3.9.5 and ISTS 3.9.6 has been renumbered to ITS 3.9.4 and ITS 3.9.5, respectively, since ISTS 3.9.2 has not been included in the KPS ITS.

4. The paragraph in the Applicable Safety Analyses states the limiting boron dilution accident occurs in MODE 5. The Applicability of the Specification for Boron Concentration is MODE 6. Therefore, the paragraph and associated reference in the Applicable Safety Analyses have been deleted since the Specification for boron concentration is applicable to MODE 6 only.

5. Changes have been made to be consistent with similar wording in the Specification.

6. Changes have been made to be consistent with changes made to the Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 22 of 181

Attachment 1, Volume 14, Rev. 1, Page 23 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 23 of 181

Attachment 1, Volume 14, Rev. 1, Page 24 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.1, BORON CONCENTRATION There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 24 of 181

, Volume 14, Rev. 1, Page 25 of 181 ATTACHMENT 2 ITS 3.9.2, NUCLEAR INSTRUMENTATION , Volume 14, Rev. 1, Page 25 of 181

, Volume 14, Rev. 1, Page 26 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 26 of 181

Attachment 1, Volume 14, Rev. 1, Page 27 of 181 A01 ITS ITS 3.9.2 3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS.

OBJECTIVE To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION Applicability a. During REFUELING OPERATIONS: M01

1. Containment Closure

a. The equipment hatch shall be closed and at least one door in each personnel air lock shall be capable of being closed (1) in 30 minutes or less. In addition, at See ITS 3.9.6 least one door in each personnel air lock shall be closed when the reactor vessel head or upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from See CTS containment atmosphere to the outside atmosphere shall have a closed isolation 3.8.a.2 valve or an operable automatic isolation valve.

A02

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage pool shall be monitored continuously. OPERABLE See CTS 3.8.a.3

3. The reactor will be subcritical for 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months prior to movement of its irradiated fuel assemblies. Core subcritical neutron flux shall be continuously monitored by at least LCO 3.9.2 two neutron monitors, each with continuous visual indication in the control room and LA01 Applicability one with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux M01 LCO 3.9.2 monitor shall be in service. A02 OPERABLE See ITS

4. At least one residual heat removal pump shall be OPERABLE. 3.9.3

5. When there is fuel in the reactor, a minimum boron concentration as specified in the COLR shall be maintained in the Reactor Coolant System during reactor vessel See ITS head removal or while loading and unloading fuel from the reactor. The required 3.9.1 boron concentration shall be verified by chemical analysis daily.

(1)

Administrative controls ensure that:

Appropriate personnel are aware that both personnel air lock doors are open, See ITS

A specified individual(s) is designated and available to close the air lock following a required 3.9.6 evacuation of containment, and

Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed.

Amendment No. 165 TS 3.8-1 03/11/2003 Page 1 of 3 Attachment 1, Volume 14, Rev. 1, Page 27 of 181

Attachment 1, Volume 14, Rev. 1, Page 28 of 181 ITS ITS 3.9.2 A01

6. Direct communication between the control room and the operating floor of the See CTS containment shall be available whenever changes in core geometry are taking 3.8.a.6 place.

7. Deleted.

8. The containment ventilation and purge system, including the capability to initiate See ITS automatic containment ventilation isolation, shall be tested and verified to be 3.9.6 operable immediately prior to and daily during REFUELING OPERATIONS.

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be operating during fuel handling and when any load is carried over the pool if See CTS irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool 3.8.a.9 sweep system, including the charcoal adsorber, is not operating when required, fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

See ITS

10. The minimum water level above the vessel flange shall be maintained at 23 feet. 3.9.5

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes for this test must be equal to or greater than the maximum load to be assumed by See CTS 3.8.a.11 the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

12. A licensed senior reactor operator will be on-site and designated in charge of the See CTS REFUELING OPERATIONS. 3.8.a.12

b. If any of the specified limiting conditions for REFUELING OPERATIONS are not met, ACTION A refueling of the reactor shall cease. Work shall be initiated to correct the violated A03 conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be performed.

Add proposed Required Action A.2 M02 Add proposed ACTIONS B and C Amendment No. 200 TS 3.8-2 11/20/2008 Page 2 of 3 Attachment 1, Volume 14, Rev. 1, Page 28 of 181

ITS A01 ITS 3.9.2 TABLE TS 4.1-1 See ITS 3.3.1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

1. Nuclear Power Range Each shift(a) Daily(a) Monthly(b) (a) Heat balance (b) Signal to T; bistable action (permissive, rod stop, trips)

Effective Full Effective Full Power Quarterly(d) (c) Upper and lower chambers for axial off-set Power Month(c) Quarter(c) using incore detectors.

The check and calibration for axial offset shall also be performed prior to > 75% power following any core alteration.

(d) Permissives P8 and P10 and the 25% reactor trip are tested quarterly.

2. Nuclear Intermediate Each shift(a,c) Not applicable Prior to each (a) Once/shift when in service Range startup if not (b) Log level; bistable action A04 done previous (permissive, rod stop, trips) week(b) (c) Channel check required in all plant modes

3. Nuclear Source Range Each shift(a,c) Not applicable Prior to each (a) Once/shift when in service startup if not (b) Bistable action (alarm, trips)

A04 M03 18 months done previous (c) Channel check required in all plant modes A04 week(b)

4. Reactor Coolant Each shift (c) Each refueling cycle Monthly(a) (a) Overtemperature T Temperature (b) Overpower T Monthly(b) (c) Channel check not required below HOT SHUTDOWN

5. Reactor Coolant Flow Each shift Each refueling cycle Monthly , Volume 14, Rev. 1, Page 29 of 181 Attachment 1, Volume 14, Rev. 1, Page 29 of 181 See ITS See ITS See ITS 3.3.1 3.3.1 3.3.1 Add proposed SR 3.9.2.2 M03 SR 3.9.2.1 Amendment No. 151 Page 1 of 7 2/12/2001 Page 3 of 3

Attachment 1, Volume 14, Rev. 1, Page 30 of 181 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR 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 3.8.a.3, in part, states core subcritical neutron flux shall be "continuously monitored" by at least two neutron monitors and at least one neutron flux monitor shall be "in service". ITS LCO 3.9.2 requires two source range neutron flux monitors to be OPERABLE and one source range audible count rate circuit to be OPERABLE. This changes the CTS by requiring the source range neutron flux monitors to be OPERABLE, instead of only "continuously monitored" or "in service".

This change is acceptable because it is consistent with the current use and understanding of the LCO. It is not sufficient for a monitor to be "continuously monitoring" or "in service" 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.

A03 When the neutron monitors are not in the required condition specified in CTS 3.8.a.3, CTS 3.8.b requires refueling of the reactor to cease, initiation of action to restore the neutron monitors to the required conditions, and no operations be performed that could increase the reactivity of the core. Under similar conditions, ITS 3.9.2 ACTION A only requires the suspension of positive reactivity additions.

This changes the CTS by deleting the requirements to initiate action to restore the neutron monitors to the required conditions.

The purpose of CTS 3.8.b is to ensure proper compensatory actions are taken to exit the Applicability of the LCO. CTS 3.8.a.3 is only applicable during REFUELING OPERATIONS, which is defined in CTS Section 1.0 as the movement of reactor vessel internals that could affect the reactivity of the core within the containment when the vessel head is unbolted or removed. Thus, after the first requirement of CTS 3.8.b is met (i.e., suspend refueling the reactor), the Applicability has been exited and thus, continuation of the requirements of CTS 3.8.b are not required. Therefore, this change is acceptable and is designated as administrative since the technical requirements have not been changed.

A04 Remark (c) to CTS Table TS 4.1-1, Channel Description 3, requires the Channel Check of the nuclear source range instrumentation to be performed in all plant modes. ITS 3.9.2 is only Applicable in MODE 6. This changes the CTS by only including the MODE 6 requirement in this Specification. ITS 3.3.1 will describe changes to this requirement in all other plant modes.

This change is acceptable since ITS 3.9.2 is only covering the MODE 6 requirements. Any changes related the other plant modes will be described in Kewaunee Power Station Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 30 of 181

Attachment 1, Volume 14, Rev. 1, Page 31 of 181 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION ITS 3.3.1. This change is designated as administrative because no technical changes are being made by this change.

MORE RESTRICTIVE CHANGES M01 CTS 3.8.a is applicable during REFUELING OPERATIONS. CTS 3.8.a.3 also states, in part, that core subcritical neutron flux shall be monitored by two neutron monitors when core geometry is being changed. When core geometry is not being changed, only one is required. ITS 3.9.2 is applicable at all times while in MODE 6. This changes the CTS by requiring the core subcritical neutron flux to be monitored by two neutron monitors at all times while in MODE 6 and not just during those times when core geometry is being changed.

The purpose of CTS 3.8.a.3 is to ensure core subcritical neutron flux is monitored while core geometry changes may be occurring during REFUELING OPERATIONS. As defined in CTS Section 1.0, REFUELING OPERATIONS is movement of reactor vessel internal components that could affect the reactivity of the core within the containment when the vessel head is unbolted or removed.

ITS 3.9.2 requires two source range neutron flux monitors to be OPERABLE and one source range audible count rate circuit to be OPERABLE at all times while in MODE 6. MODE 6 is defined as when one or more reactor vessel head closure bolts are less than fully tensioned. As a result, MODE 6 encompasses the aforementioned CTS condition and, in addition, those times when there is no movement of reactor internal components. This change is acceptable because the requirements continue to ensure that process variables are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. This change is designated more restrictive because the core subcritical neutron flux monitoring requirements for ITS are more restrictive than the conditions stated in the CTS.

M02 When the required neutron monitors are not capable of continuously monitoring the core subcritical neutron flux during REFUELING OPERATIONS (which is defined as movement of reactor vessel internal components that could affect the reactivity of the core within the containment when the reactor head is unbolted or removed), CTS 3.8.b requires refueling of the reactor to cease. When one source range neutron flux monitor is inoperable, ITS 3.9.2 ACTION A requires an action similar to the first action of CTS 3.8.b, (Required Action A.1), but also includes an additional action to be taken. ITS 3.9.2 Required Action A.2 also requires immediate suspension of operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the boron concentration of LCO 3.9.1. When both source range neutron flux monitors are inoperable, in addition to ITS 3.9.2 ACTION A, ITS 3.9.2 ACTION B requires immediately initiating action to restore one source range neutron flux monitor to OPERABLE status (Required Action B.1) and performing SR 3.9.1.1 once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (Required Action B.2). When the audible count rate circuit is inoperable, ITS 3.9.2 ACTION C requires immediately initiating action to isolate unborated water sources (Required Action C.1). This changes the CTS by adding new Required Actions when the required nuclear instrumentation is inoperable.

Kewaunee Power Station Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 31 of 181

Attachment 1, Volume 14, Rev. 1, Page 32 of 181 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION These added Required Actions are acceptable since they assist in minimizing the consequences of the required source range neutron flux monitors being inoperable. Note that while CTS 3.8.b requires operations not be performed that could increase the reactivity of the core, this action is actually not required to be taken since once movement of fuel has been suspended, REFUELING OPERATIONS has ceased (thus, the LCO requirements of CTS 3.8.a are not required to be met). Thus, the addition of a similar Required Action (Required Action A.2), as well as the other above described Required Actions, is designated as more restrictive.

M03 CTS 3.8.a.3, in part, requires the core subcritical neutron flux to be monitored under certain conditions, but does not provide a Surveillance Requirement to periodically verify the calibration of the source range neutron flux monitors. ITS SR 3.9.2.2 requires a CHANNEL CALIBRATION of the Nuclear Instrumentation (source range neutron flux monitors) every 18 months. SR 3.9.2.2 also contains a NOTE excluding the neutron detectors from the CHANNEL CALIBRATION.

This changes the CTS by adding a new Surveillance Requirement to periodically verify the calibration of the source range neutron flux monitors.

The purpose of ITS SR 3.9.2.2 is to provide additional assurance that the source range neutron flux monitors are capable of providing a reliable and accurate indication of core subcritical neutron flux. The 18 month Frequency for the proposed Surveillance is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. This change is acceptable because the Surveillance Requirement ensures the system can provide a reliable and accurate indication of core subcritical neutron flux while in MODE 6. This change is designated as more restrictive because a new SR has been added to periodically verify the calibration of the source range neutron flux monitors.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.8.a.3, in part, states that core subcritical neutron flux shall be monitored by two neutron monitors, "each with continuous visual indication in the control room" and one with audible indication "in the containment." ITS 3.9.2 LCO states that two source range neutron flux monitors shall be OPERABLE and one source range audible count rate circuit shall be OPERABLE. This changes the CTS by moving the requirement that each monitor have a "continuous visual indication in the control room" from the CTS to the Bases. This also changes the CTS by changing the location of the audible indication from the containment to the control room and moving this requirement from the CTS to the Bases.

Kewaunee Power Station Page 3 of 4 Attachment 1, Volume 14, Rev. 1, Page 32 of 181

Attachment 1, Volume 14, Rev. 1, Page 33 of 181 DISCUSSION OF CHANGES ITS 3.9.2, NUCLEAR INSTRUMENTATION The removal of these details, which is 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. The ITS retains the requirement that two channels be OPERABLE and one audible count rate circuit be OPERABLE and continues to require the associated Surveillances to verify OPERABILITY. 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 change in the location is acceptable since it is the control room as the location assumed in the accident analysis for this audible indication. The boron dilution accident assumes the audible count rate provides prompt and definite indication of any boron dilution. The audible indicator in the containment cannot provide this indication since the operators in the control room cannot hear the indicator in the containment. Furthermore, the containment is not occupied at all times in MODE 6. 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 14, Rev. 1, Page 33 of 181

Attachment 1, Volume 14, Rev. 1, Page 34 of 181 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 34 of 181

Attachment 1, Volume 14, Rev. 1, Page 35 of 181 CTS All changes are 1 Nuclear Instrumentation unless otherwise noted 3.9.3 2 2

3.9 REFUELING OPERATIONS 2

3.9.3 2 Nuclear Instrumentation 2

3.8.a.3 LCO 3.9.3 2 Two source range neutron flux monitors shall be OPERABLE.

AND

[ One source range audible [alarm] [count rate] circuit shall be OPERABLE. ]

3.8.a, APPLICABILITY: MODE 6.

3.8.a.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 4

3.8.b A. One [required] source A.1 Suspend CORE Immediately range neutron flux ALTERATIONS.

TSTF-monitor inoperable. positive reactivity 471 additions.

AND A.2 Suspend operations that Immediately would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

3.8.b B. Two [required] source B.1 Initiate action to restore one Immediately 4 range neutron flux source range neutron flux monitors inoperable. monitor to OPERABLE status.

AND B.2 Perform SR 3.9.1.1. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months WOG STS 3.9.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 35 of 181

Attachment 1, Volume 14, Rev. 1, Page 36 of 181 All changes are 1 Nuclear Instrumentation CTS unless otherwise noted 3.9.3 2

2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

REVIEWER'S NOTE----- C.1 Initiate action to isolate Immediately ]

Condition C is included only unborated water sources.

for plants that assume a boron dilution event is 3 mitigated by operator response to an audible source range indication.

3.8.b C. [ Required source range audible [alarm] [count rate] circuit inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Table TS 4.1- 2 1, Channel SR 3.9.3.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 2 Description 3 2

2 DOC M03 SR 3.9.3.2 -------------------------------NOTE------------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. [18] months WOG STS 3.9.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 36 of 181

Attachment 1, Volume 14, Rev. 1, Page 37 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.2, NUCLEAR INSTRUMENTATION

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 3.9.3 has been renumbered to ITS 3.9.2 since ISTS 3.9.2 has not been included in the Kewaunee Power Station (KPS) ITS.

3. The Reviewer's Note has been deleted. The 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. KPS assumes the boron dilution event in MODE 6 is mitigated by operator response to an audible source range, so ACTION C is included in the KPS ITS.

4. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. The KPS design includes only two source range neutron flux monitors, both of which are required by this Specification. Therefore, the term "required" is not needed and has been deleted. 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 14, Rev. 1, Page 37 of 181

Attachment 1, Volume 14, Rev. 1, Page 38 of 181 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 38 of 181

Attachment 1, Volume 14, Rev. 1, Page 39 of 181 All changes are 1 Nuclear Instrumentation unless otherwise noted B 3.9.3 6 2

B 3.9 REFUELING OPERATIONS B 3.9.3 Nuclear Instrumentation 6 2

BASES BACKGROUND -----------------------------------REVIEWERS NOTE-----------------------------------

Bracketed options are provided for source range OPERABILITY requirements to include audible alarm or count rate function. These options apply to plants that assume a boron dilution event that is 2 mitigated by operator response to an audible indication. For plants that isolate all boron dilution paths (per LCO 3.9.2), the source range OPERABILITY includes only a visual monitoring function.

The source range neutron flux monitors are used during refueling operations to monitor the core reactivity condition. The installed source range neutron flux monitors are part of the Nuclear Instrumentation System (NIS). These detectors are located external to the reactor vessel and detect neutrons leaking from the core.

fission chamber The installed source range neutron flux monitors are BF3 detectors 3 operating in the proportional region of the gas filled detector characteristic curve. The detectors monitor the neutron flux in counts per second. The instrument range covers six decades of neutron flux (1E+6 cps) with a 4

[5]% instrument accuracy. The detectors also provide continuous visual indication in the control room [and an audible [alarm] [count rate] to alert operators to a possible dilution accident]. The NIS is designed in accordance with the criteria presented in Reference 1.

APPLICABLE Two OPERABLE source range neutron flux monitors are required to SAFETY provide a signal to alert the operator to unexpected changes in core ANALYSES reactivity such as with a boron dilution accident (Ref. 2) or an improperly loaded fuel assembly. [The audible count rate from the source range neutron flux monitors provides prompt and definite indication of any boron dilution. The count rate increase is proportional to the subcritical multiplication factor and allows operators to promptly recognize the initiation of a boron dilution event. Prompt recognition of the initiation of a boron dilution event is consistent with the assumptions of the safety analysis and is necessary to assure sufficient time is available for isolation of the primary water makeup source before SHUTDOWN MARGIN is lost (Ref. 2).]

WOG STS B 3.9.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 39 of 181

Attachment 1, Volume 14, Rev. 1, Page 40 of 181 All changes are 1 Nuclear Instrumentation unless otherwise noted B 3.9.3 6

2 BASES APPLICABLE SAFETY ANALYSES (continued)

REVIEWERS NOTE-----------------------------------

The need for a safety analysis for an uncontrolled boron dilution accident 2 is eliminated by isolating all unborated water sources as required by LCO 3.9.2, "Unborated Water Source Isolation Valves."

The source range neutron flux monitors satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO requires that two source range neutron flux monitors be OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity. To be OPERABLE, each monitor must provide visual indication [in the control room]. [In addition, at least one of in the control room the two monitors must provide an OPERABLE audible [alarm] [count rate]

function to alert the operators to the initiation of a boron dilution event.] 3 APPLICABILITY In MODE 6, the source range neutron flux monitors must be OPERABLE to determine changes in core reactivity. There are no other direct means available to check core reactivity levels. In MODES 2, 3, 4, and 5, these same installed source range detectors and circuitry are also required to be OPERABLE by LCO 3.3.1, "Reactor Trip System (RTS) 3 Instrumentation [and LCO 3.3.9, "BDPS"]. Protection P ACTIONS A.1 and A.2 positive reactivity With only one source range neutron flux monitor OPERABLE, additions TSTF-471 redundancy has been lost. Since these instruments are the only direct means of monitoring core reactivity conditions, CORE ALTERATIONS and introduction of coolant into the RCS with boron concentration less than required to meet the minimum boron concentration of LCO 3.9.1 must be suspended immediately. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position.

WOG STS B 3.9.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 40 of 181

Attachment 1, Volume 14, Rev. 1, Page 41 of 181 All changes are 1 Nuclear Instrumentation unless otherwise noted B 3.9.3 6

2 BASES ACTIONS (continued)

B.1 With no source range neutron flux monitor OPERABLE, action to restore a monitor to OPERABLE status shall be initiated immediately. Once initiated, action shall be continued until a source range neutron flux monitor is restored to OPERABLE status.

B.2 With no source range neutron flux monitor OPERABLE, there are no direct means of detecting changes in core reactivity. However, since TSTF-CORE ALTERATIONS and positive reactivity additions are not to be 471 made, the core reactivity condition is stabilized until the source range neutron flux monitors are OPERABLE. This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists.

The Completion Time of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.

[ C.1 6 With no audible [alarm] [count rate] OPERABLE, prompt and definite indication of a boron dilution event, consistent with the assumptions of the safety analysis, is lost. In this situation, the boron dilution event may not be detected quickly enough to assure sufficient time is available for operators to manually isolate the unborated water source and stop the dilution prior to the loss of SHUTDOWN MARGIN. Therefore, action must be taken to prevent an inadvertent boron dilution event from occurring.

This is accomplished by isolating all the unborated water flow paths to the Reactor Coolant System. Isolating these flow paths ensures that an inadvertent dilution of the reactor coolant boron concentration is prevented. The Completion Time of "Immediately" assures a prompt response by operations and requires an operator to initiate actions to isolate an affected flow path immediately. Once actions are initiated, they must be continued until all the necessary flow paths are isolated or the circuit is restored to OPERABLE status. ]

WOG STS B 3.9.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 41 of 181

Attachment 1, Volume 14, Rev. 1, Page 42 of 181 All changes are 1 Nuclear Instrumentation unless otherwise noted B 3.9.3 2 6 BASES 2

SURVEILLANCE SR 3.9.3.1 6 REQUIREMENTS 2 SR 3.9.3.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LCO 3.3.1.

2 SR 3.9.3.2 6 2

SR 3.9.3.2 is the performance of a CHANNEL CALIBRATION every 18 months. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the source range neutron flux monitors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. [The CHANNEL CALIBRATION also includes verification of the audible [alarm]

[count rate] function.] The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 13, GDC 26, GDC 28, and GDC 29.

U 3

2. FSAR, Section [15.2.4]. 14.1.4.2.1 USAR, Section 7.1, General Design Criteria (GDC) 12, "Instrumentation and Control Systems," and Section 5 7.4.1.1, GDC 13, "Fission Process Monitors and Controls."

WOG STS B 3.9.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 42 of 181

Attachment 1, Volume 14, Rev. 1, Page 43 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.2 BASES, NUCLEAR INSTRUMENTATION

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 Reviewer's Note has been deleted. The 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. In addition, KPS ITS does not include LCO 3.9.2. Therefore, there is no need to include the Reviewer's Note.

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

4. The specific accuracy of the source range neutron flux monitors is not part of the licensing basis of Kewaunee Power Station and has been deleted.

5. The ISTS lists GDC 13, GDC 26, GDC 28, and GDC 29 of Appendix A to 10 CFR 50 as the reference document for the criteria of Nuclear Instrumentation System (NIS) is designed. 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 12, Instrumentation and Control Systems and GDC 13, Fission Process Monitors and Controls, as the licensing reference document for the design criteria of the NIS.

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

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 43 of 181

Attachment 1, Volume 14, Rev. 1, Page 44 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 44 of 181

Attachment 1, Volume 14, Rev. 1, Page 45 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.2, NUCLEAR INSTRUMENTATION There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 45 of 181

Attachment 1, Volume 14, Rev. 1, Page 46 of 181 ATTACHMENT 3 ITS 3.9.3, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL Attachment 1, Volume 14, Rev. 1, Page 46 of 181

, Volume 14, Rev. 1, Page 47 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 47 of 181

Attachment 1, Volume 14, Rev. 1, Page 48 of 181 ITS A01 ITS 3.9.3 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 3.9.3 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.

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 3.4.7 and temperatures 200°F unless the secondary water temperature of each steam 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 4 Attachment 1, Volume 14, Rev. 1, Page 48 of 181

Attachment 1, Volume 14, Rev. 1, Page 49 of 181 ITS A01 ITS 3.9.3 A03 LCO 3.9.3 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, LCO 3.9.3 except when in the REFUELING MODE with the minimum water level above the A03 and top of the vessel flange 23 feet, one train may be inoperable for maintenance.

Applicability

1. Each residual heat removal train shall be comprised of:

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 A04 Add proposed Required Actions A.1, A.4, A.5, and A.6 M02 Add proposed ACTION A for loop not in M03 operation Add proposed SR 3.9.3.1 M04 Amendment No. 165 TS 3.1-2 03/11/2003 Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 49 of 181

Attachment 1, Volume 14, Rev. 1, Page 50 of 181 A01 ITS ITS 3.9.3 3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS.

OBJECTIVE To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION Applicability A05

a. During REFUELING OPERATIONS:

1. Containment Closure

a. The equipment hatch shall be closed and at least one door in each personnel air lock shall be capable of being closed (1) in 30 minutes or less. In addition, at See ITS 3.9.6 least one door in each personnel air lock shall be closed when the reactor vessel head or upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valve or an operable automatic isolation valve.

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage See CTS 3.8.a.2 pool shall be monitored continuously.

See CTS

3. The reactor will be subcritical for 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months prior to movement of its irradiated fuel 3.8.a.3 assemblies. Core subcritical neutron flux shall be continuously monitored by at least See ITS two neutron monitors, each with continuous visual indication in the control room and 3.9.2 one with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux monitor shall be in service.

LCO 3.9.3 4. At least one residual heat removal pump shall be OPERABLE.

5. When there is fuel in the reactor, a minimum boron concentration as specified in the COLR shall be maintained in the Reactor Coolant System during reactor vessel See ITS head removal or while loading and unloading fuel from the reactor. The required 3.9.1 boron concentration shall be verified by chemical analysis daily.

(1)

Administrative controls ensure that:

Appropriate personnel are aware that both personnel air lock doors are open,

A specified individual(s) is designated and available to close the air lock following a required See ITS 3.9.6 evacuation of containment, and

Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed.

Amendment No. 165 TS 3.8-1 03/11/2003 Page 3 of 4 Attachment 1, Volume 14, Rev. 1, Page 50 of 181

Attachment 1, Volume 14, Rev. 1, Page 51 of 181 ITS ITS 3.9.3 A01

6. Direct communication between the control room and the operating floor of the See CTS containment shall be available whenever changes in core geometry are taking 3.8.a.6 place.

7. Deleted.

8. The containment ventilation and purge system, including the capability to initiate See ITS automatic containment ventilation isolation, shall be tested and verified to be 3.9.6 operable immediately prior to and daily during REFUELING OPERATIONS.

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be operating during fuel handling and when any load is carried over the pool if irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool sweep system, including the charcoal adsorber, is not operating when required, fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements See CTS 3.8.a.9

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

See ITS

10. The minimum water level above the vessel flange shall be maintained at 23 feet. 3.9.5

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes See CTS for this test must be equal to or greater than the maximum load to be assumed by 3.8.a.11 the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

See CTS

12. A licensed senior reactor operator will be on-site and designated in charge of the 3.8.a.12 REFUELING OPERATIONS.

ACTION A b. If any of the specified limiting conditions for REFUELING OPERATIONS are not met, refueling of the reactor shall cease. Work shall be initiated to correct the violated conditions so that the specified limits are met, and no operations which may increase the M02 reactivity of the core shall be performed.

Add proposed Required Actions A.1, A.4, A.5, and A.6 M02 Add proposed Required Action A.3 A04 Amendment No. 200 TS 3.8-2 11/20/2008 Page 4 of 4 Attachment 1, Volume 14, Rev. 1, Page 51 of 181

Attachment 1, Volume 14, Rev. 1, Page 52 of 181 DISCUSSION OF CHANGES ITS 3.9.3, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL 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 one RHR train to be OPERABLE (see DOC A03),

is applicable whenever the average reactor coolant temperature is 200ºF and irradiated fuel is in the reactor. ITS 3.9.3, which also includes similar RHR train OPERABILITY requirements, is applicable in MODE 6 with the water level 23 ft above the top of the reactor vessel flange. RHR requirements in MODE 5 are provided in LCO 3.4.7 and LCO 3.4.8 and the MODE 6 RHR requirements with the water level < 23 ft above the top of the reactor vessel flange are provided in LCO 3.9.4. 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.

A03 CTS 3.1.a.2.B requires two RHR trains to be OPERABLE. However, it provides an allowance that one RHR train may be inoperable for maintenance when in the REFUELING MODE (ITS MODE 6) with the water level 23 ft above the top of the reactor vessel flange. This essentially means that when in the REFUELING MODE with the water level 23 ft above the top of the reactor vessel flange, only one RHR train is required to be OPERABLE. ITS 3.9.3, in part, specifies that one RHR loop shall be OPERABLE in MODE 6 with the water level 23 ft above the top of the reactor vessel flange. This changes the CTS by clearly stating the LCO requirement when in MODE 6 with the water level 23 ft above the top of the reactor vessel flange.

The change is acceptable since the requirements in the CTS and ITS are the same; only one RHR loop is required to be OPERABLE when in MODE 6 with the water level 23 ft above the top of the reactor vessel flange. This change is designated as administrative because it does not result in any technical changes.

A04 When the required RHR train is inoperable, CTS 3.1.a.2.B.2 essentially requires immediate action to be taken to restore the train to OPERABLE status.

Furthermore, when a required RHR pump is inoperable, CTS 3.8.b, in part, essentially requires immediate suspension of loading fuel in the reactor. Under similar conditions, ITS 3.9.3 ACTION A requires an action similar to CTS 3.1.a.2.B.2 (Required Action A.3), and an action similar to CTS 3.8.b (Required Action A.2). This changes the CTS by including specific Required Actions applicable to the related conditions of CTS 3.1.a.2.B.2 and CTS 3.8.b. The addition of other Required Actions (A.1, A.4, A.5, and A.6) is discussed in DOC M02.

Kewaunee Power Station Page 1 of 6 Attachment 1, Volume 14, Rev. 1, Page 52 of 181

Attachment 1, Volume 14, Rev. 1, Page 53 of 181 DISCUSSION OF CHANGES ITS 3.9.3, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL The addition of the specific Required Actions is acceptable since they are already required by another Specification. CTS 3.8.a requires an RHR pump to be OPERABLE during REFUELING OPERATIONS, which is defined in CTS Section 1.0 as the movement of reactor vessel internal components that could affect the reactivity of the core within the containment when the vessel head is unbolted or removed. CTS 3.1.a.2.B requires one RHR train to be OPERABLE during REFUELING (ITS MODE 6) with the water level 23 ft above the top of the reactor vessel flange. Thus, CTS 3.8.a is applicable under some of the same conditions as when CTS 3.1.a.2.B is applicable. When the required pump is inoperable, CTS 3.8.b, in part, requires refueling of the reactor to cease and CTS 3.1.a.2.B.2 requires immediate action to be taken to restore the inoperable RHR train (i.e., both actions have to be taken). Thus, this change is designated as administrative since the technical requirements (as they relate to this specific change) have not been changed.

A05 CTS 3.8.a.4 requires one RHR pump to be OPERABLE. As stated in CTS 3.8.a, this requirement is applicable during REFUELING OPERATIONS. ITS 3.9.3, which also requires an RHR loop to be OPERABLE, is applicable in MODE 6 with the water level 23 ft above the top of the reactor vessel flange. This changes the CTS by requiring an RHR loop to be OPERABLE at all times in MODE 6 with the water level 23 ft above the top of the reactor vessel flange, not just during REFUELING OPERATIONS.

CTS 3.1.a.2.B requires one RHR loop to be OPERABLE at all times in MODE 6 with the water level 23 ft above the top of the reactor vessel flange, not just during REFUELING OPERATIONS. As defined in CTS Section 1.0, REFUELING OPERATIONS is movement of reactor vessel internal components that could affect the reactivity of the core within the containment when the vessel head is unbolted or removed. Since REFUELING OPERATIONS can essentially only occur in MODE 6 with water level 23 ft above the top of the reactor vessel flange, CTS 3.8.a is applicable under some of the same conditions as when CTS 3.1.a.2.B is applicable. Since CTS 3.1.a.2.B already requires one RHR loop to be OPERABLE at all times in MODE 6 with the water level 23 ft above the top of the reactor vessel flange, the addition of the expanded Applicability for CTS 3.8.a is considered acceptable. The change is designated as administrative since the technical requirements have not been changed.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.a.1.A, which (in MODE 6) requires a residual heat removal (RHR) pump to be in operation, is applicable in MODE 6 only when a reduction is made in the boron concentration of the reactor coolant. ITS 3.9.3, in part, requires an RHR loop to be in operation and is applicable at all times when in MODE 6 with the water level 23 ft above the top of the reactor vessel flange, except as allowed in the LCO Note. The Note allows the required RHR pump 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, provided 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 3.9.1. This changes the CTS by specifying that the LCO requirement for an RHR pump to be in operation is applicable in MODE 6 with the water level 23 ft above the top of the reactor Kewaunee Power Station Page 2 of 6 Attachment 1, Volume 14, Rev. 1, Page 53 of 181

Attachment 1, Volume 14, Rev. 1, Page 54 of 181 DISCUSSION OF CHANGES ITS 3.9.3, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL vessel flange at all times except for the condition specified in the ITS 3.9.3 LCO Note.

In MODE 6, 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 REFUELING (equivalent to ITS MODE 6) with the water level 23 ft above the top of the reactor vessel flange. Other MODES are discussed in other ITS discussions. The ITS 3.9.3 Applicability is MODE 6 at all times with the water level 23 ft above the top of the reactor vessel flange, except for the condition in the ITS 3.9.3 LCO Note, and primarily addresses the decay heat removal function, but also covers the boron mixing issue. RHR pumps are utilized in MODE 6 to provide for removal of decay heat from the reactor core to the RHR heat exchangers, and to ensure proper boron mixing within the reactor coolant. The MODE 6 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 or evolutions 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 or evolution. This change is designated more restrictive because one RHR pump is required to be in operation in MODE 6 at all times with the water level 23 ft above the top of the reactor vessel flange, except for the condition specified in the ITS 3.9.3 LCO Note.

M02 When the required RHR train is inoperable, CTS 3.1.a.2.B.2 requires immediate action to be taken to restore the train to OPERABLE status. In addition, CTS 3.8.a.4 requires one RHR pump to be OPERABLE during REFUELING OPERATIONS (which is defined as movement of reactor vessel internal components that could affect the reactivity of the core within the containment when the reactor head is unbolted or removed). If this requirement is not met, CTS 3.8.b requires refueling of the reactor to cease, initiation of action to restore the RHR pump to OPERABLE status, and no operations be performed that could increase the reactivity of the core. Under similar conditions, ITS 3.9.3 ACTION A requires an action similar to CTS 3.1.a.2.B.2 and the second action of CTS 3.8.b, (Required Action A.3) and an action similar to the first action of CTS 3.8.b (Required Action A.2), but also includes additional actions to be taken. ITS 3.9.3 ACTION A also requires immediate suspension of operations that would cause introduction of coolant into the RCS with a boron concentration less than required to meet the boron concentration of LCO 3.9.1 (Required Action A.1), closing the equipment hatch and securing it with four bolts in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Required Action A.4),

closing one door in each air lock in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Required Action A.5), and a Kewaunee Power Station Page 3 of 6 Attachment 1, Volume 14, Rev. 1, Page 54 of 181

Attachment 1, Volume 14, Rev. 1, Page 55 of 181 DISCUSSION OF CHANGES ITS 3.9.3, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL verification that each penetration providing direct access from the containment atmosphere to the outside atmosphere is either closed with a manual or automatic isolation valve, blind flange, or equivalent, or is capable of being closed by an OPERABLE Containment Purge and Vent Isolation System in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Required Action A.6). This changes the CTS by adding new Required Actions when the required RHR loop is inoperable.

These added Required Actions are acceptable since they assist in minimizing the consequences of the required RHR loop being inoperable. Note that while CTS 3.8.b requires operations not be performed that could increase the reactivity of the core, this action is actually not required to be taken since once movement of fuel has been suspended, REFUELING OPERATIONS has ceased (thus, the LCO requirements of CTS 3.8.a are not required to be met). Thus, the addition of a similar Required Action (Required Action A.1), as well as the other above described Required Actions, is designated as more restrictive.

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. 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 6, 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 3.9.3 ACTION A 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 boron concentration of LCO 3.9.1 (Required Action A.1), immediate suspension of loading irradiated fuel assemblies into the core (Required Action A.2), immediate initiation of action to restore one RHR loop to operation (Required Action A.3),

closing the equipment hatch and securing it with four bolts in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Required Action A.4), closing one door in each air lock in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Required Action A.5),

and a verification that each penetration providing direct access from the containment atmosphere to the outside atmosphere is either closed with a manual or automatic isolation valve, blind flange, or equivalent, or is capable of being closed by an OPERABLE Containment Purge and Vent Isolation System in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Required Action A.6). This changes the CTS by adding a new ACTION.

The purpose of the Actions should be to place the unit outside of the Applicability of the Specification. ITS 3.9.3 ACTION A 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 boron concentration of LCO 3.9.1 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.

Kewaunee Power Station Page 4 of 6 Attachment 1, Volume 14, Rev. 1, Page 55 of 181

Attachment 1, Volume 14, Rev. 1, Page 56 of 181 DISCUSSION OF CHANGES ITS 3.9.3, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL 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 3.9.3.1 requires verification that each required RHR loop is in operation and circulating reactor coolant 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 to periodically verify the required pump is in operation.

The purpose of ITS SR 3.9.3.1 is to ensure that the RHR loops are in operation and circulating reactor coolant at a flow rate necessary to provide sufficient decay heat removal and to prevent thermal and boron stratification. 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 acceptable because the Surveillance Requirement ensures the system can remove reactor decay heat and provides proper boron mixing via the forced circulation of the reactor coolant. This change is designated as more restrictive because a new SR has been added to periodically verify the required RHR loop is in operation.

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 one residual heat removal (RHR) train be OPERABLE (see DOC A03) with the 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 3.9.3 requires one RHR loop 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 loop in MODE 6. 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 5 of 6 Attachment 1, Volume 14, Rev. 1, Page 56 of 181

Attachment 1, Volume 14, Rev. 1, Page 57 of 181 DISCUSSION OF CHANGES ITS 3.9.3, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 6 of 6 Attachment 1, Volume 14, Rev. 1, Page 57 of 181

Attachment 1, Volume 14, Rev. 1, Page 58 of 181 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 58 of 181

Attachment 1, Volume 14, Rev. 1, Page 59 of 181 CTS RHR and Coolant Circulation - High Water Level 3.9.5 4 3

3.9 REFUELING OPERATIONS 4

3.9.5 3 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level 3.1.a.1.A, LCO 3.9.5 3 One RHR loop shall be OPERABLE and in operation. 4 3.1.a.2.B, 3.8.a.4

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

The required RHR loop 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 no operations are permitted that would cause introduction of coolant into the Reactor Coolant System with boron concentration less than that required to meet the minimum required boron concentration of LCO 3.9.1. , "Boron Concentration." 1 3.1.a.1.A, APPLICABILITY: MODE 6 with the water level 23 ft above the top of reactor vessel 3.1.a.2.B, flange.

3.8.a ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.2.B.2, A. RHR loop requirements A.1 Suspend operations that Immediately 3.8.b, not met. would cause introduction of DOC M03 coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

AND A.2 Suspend loading irradiated Immediately fuel assemblies in the core.

AND A.3 Initiate action to satisfy Immediately RHR loop requirements.

AND WOG STS 3.9.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 59 of 181

Attachment 1, Volume 14, Rev. 1, Page 60 of 181 RHR and Coolant Circulation - High Water Level CTS 3.9.5 4 3

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.2.B.2, A.4 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 3.8.b, secure with [four] bolts. 2 DOC M03 AND A.5 Close one door in each air 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lock.

AND Verify A.6.1 Close each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from the containment atmosphere to the outside is either closed atmosphere with a manual or automatic isolation valve, blind flange, or equivalent.

3

or OR A.6.2 Verify each penetration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months capable of being closed by an OPERABLE Containment Purge and Vent Exhaust Isolation System.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 3 4 DOC M04 SR 3.9.5.1 Verify one RHR loop is in operation and circulating 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 5

reactor coolant at a flow rate of [2800] gpm.

WOG STS 3.9.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 60 of 181

Attachment 1, Volume 14, Rev. 1, Page 61 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL

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

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 requirements.

3. ISTS 3.9.5 Required Actions A.6.1 and A.6.2 are connected by an "OR" logical connector, such that either one can be performed to meet the requirements of the ACTION. However, the two Required Actions are applicable to all the penetrations; either Required Action A.6.1 or Required Action A.6.2 must be performed for all the penetrations. Thus, this will not allow one penetration to be isolated by use of a manual valve and another penetration to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. This is not the intent of the requirement. The requirement is based on ISTS LCO 3.9.4, which requires each penetration to be either: a) closed by a manual or automatic isolation valve, blind flange, or equivalent; or b) capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. For consistency with the actual LCO requirement, ISTS 3.9.5 Required Actions A.6.1 and A.6.2 have been combined into a single Required Action in ITS 3.9.3 Required Action A.6. Furthermore, the title of the system has been changed to be consistent with the Kewaunee Power Station System name (i.e., Containment Purge and Vent Isolation System).

4. ISTS 3.9.5 has been renumbered to ITS 3.9.3 since ISTS 3.9.2 has not been included in the KPS ITS and ISTS 3.9.4 is being numbered as ITS 3.9.6.

5. The minimum required RHR flow rate requirement has not been included in KPS ITS SR 3.9.3.1. The Bases states that the flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. However, the decay heat is not always the same; it is a function of time after shutdown and the power history of the fuel. Furthermore, SRs in Section 3.4 (ITS SRs 3.4.6.1, 3.4.7.1, and 3.4.8.1) require a similar verification that the RHR loop is in operation, but do not specify a flow rate requirement. This change is also consistent with the KPS current Technical Specifications, which does not specify a flow rate for the RHR loop.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 61 of 181

Attachment 1, Volume 14, Rev. 1, Page 62 of 181 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 62 of 181

Attachment 1, Volume 14, Rev. 1, Page 63 of 181 RHR and Coolant Circulation - High Water Level B 3.9.5 11 3

B 3.9 REFUELING OPERATIONS B 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level 11 3

BASES BACKGROUND The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by 1 GDC 34, to provide mixing of borated coolant and to prevent boron 2

stratification (Ref. 1). Heat is removed from the RCS by circulating Refs. 1 and 2 reactor coolant through the RHR heat exchanger(s), where the heat is transferred to the Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg(s). Operation of the RHR 2 an System for normal cooldown or decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of reactor coolant through the RHR heat exchanger(s) and the bypass. Mixing of the reactor coolant is maintained by this continuous circulation of reactor coolant through the RHR System.

APPLICABLE If the reactor coolant temperature is not maintained below 200°F, boiling SAFETY of the reactor coolant could result. This could lead to a loss of coolant in ANALYSES the reactor vessel. Additionally, boiling of the reactor coolant could lead to a reduction in boron concentration in the coolant due to boron plating out on components near the areas of the boiling activity. The loss of reactor coolant and the reduction of boron concentration in the reactor loop coolant would eventually challenge the integrity of the fuel cladding, which is a fission product barrier. One train of the RHR System is required to be 3

operational in MODE 6, with the water level 23 ft above the top of the OPERABLE and in operation reactor vessel flange, to prevent this challenge. The LCO does permit the RHR pump to be removed from operation for short durations, under the condition that the boron concentration is not diluted. This conditional stopping of the RHR pump does not result in a challenge to the fission product barrier.

and Coolant Circulation - High Water Level 4

The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO Only one RHR loop is required for decay heat removal in MODE 6, with the water level 23 ft above the top of the reactor vessel flange. Only one RHR loop is required to be OPERABLE, because the volume of water above the reactor vessel flange provides backup decay heat removal capability. At least one RHR loop must be OPERABLE and in operation to provide:

5

a. Removal of decay heat, ; and WOG STS B 3.9.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 63 of 181

Attachment 1, Volume 14, Rev. 1, Page 64 of 181 RHR and Coolant Circulation - High Water Level B 3.9.5 11 3

BASES LCO (continued) .

b. Mixing of borated coolant to minimize the possibility of criticality, and 6

c. Indication of reactor coolant temperature.

An OPERABLE RHR loop includes an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in 6 one of the RCS hot legs and is returned to the RCS cold legs. 2 an The LCO is modified by a Note that allows the required operating RHR loop to be removed from operation for up to 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 no operations are permitted that would dilute the RCS boron concentration by introduction of coolant into the RCS with boron concentration less than required to meet the minimum boron

, "Boron concentration of LCO 3.9.1. Boron concentration reduction with coolant 7 Concentration." at boron concentrations less than required to assure the RCS boron concentration is maintained is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to RHR isolation valve testing.

During this 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.

APPLICABILITY One RHR loop must be OPERABLE and in operation in MODE 6, with the water level 23 ft above the top of the reactor vessel flange, to provide decay heat removal. The 23 ft water level was selected because it corresponds to the 23 ft requirement established for fuel movement in 11 5

LCO 3.9.7, "Refueling Cavity Water Level." Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling 8 Systems (ECCS). RHR loop requirements in MODE 6 with the water level < 23 ft are located in LCO 3.9.6, "Residual Heat Removal (RHR) 11 and Coolant Circulation - Low Water Level."

4 WOG STS B 3.9.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 64 of 181

Attachment 1, Volume 14, Rev. 1, Page 65 of 181 RHR and Coolant Circulation - High Water Level B 3.9.5 11 3

BASES ACTIONS RHR loop requirements are met by having one RHR loop OPERABLE and in operation, except as permitted in the Note to the LCO.

A.1 If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in 12 the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.

A.2 If RHR loop requirements are not met, actions shall be taken immediately to suspend loading of irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core.

A minimum refueling water level of 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading a fuel assembly, is a prudent action under this condition.

A.3 If RHR loop requirements are not met, actions shall be initiated and continued in order to satisfy RHR loop requirements. With the unit in MODE 6 and the refueling water level 23 ft above the top of the reactor vessel flange, corrective actions shall be initiated immediately.

9 A.4, A.5, A.6.1, and A.6.2 If no RHR is in operation, the following actions must be taken:

a. The equipment hatch must be closed and secured with [four] bolts, 10 5 WOG STS B 3.9.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 65 of 181

Attachment 1, Volume 14, Rev. 1, Page 66 of 181 RHR and Coolant Circulation - High Water Level B 3.9.5 11 3

BASES ACTIONS (continued) 5

b. One door in each air lock must be closed, and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or Vent verified to be capable of being closed by an OPERABLE 2

Containment Purge and Exhaust Isolation System.

With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.

Performing the actions described above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.

Circulating coolant ensures thermal and The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows fixing of most RHR problems and boron stratification is reasonable, based on the low probability of the coolant boiling in that are minimized time.

SURVEILLANCE SR 3.9.5.1 11 REQUIREMENTS 3 This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate 9 necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator in the control room for monitoring the RHR System.

U 2 10 REFERENCES 1. FSAR, Section [5.5.7].

14.1.4.2 2

2. USAR, Section 9.3.1.2.

WOG STS B 3.9.5-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 66 of 181

Attachment 1, Volume 14, Rev. 1, Page 67 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3 BASES, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL

1. Kewaunee Power Station (KPS) was designed and under construction prior to the promulgation of 10 CFR 50 Appendix A. KPS was designed and constructed to meet the intent of the proposed General Design Criteria, published in 1967. The KPS USAR Section 1.8 provides a description of each of the proposed General Design Criteria and how KPS meets the intent of each one. However, the proposed General Design Criteria did not have a criteria equivalent to 10 CFR 50 Appendix A, GDC 34.

Therefore, reference to this GDC has been deleted.

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 have been made to be consistent with the actual wording in the Specification. ISTS 3.9.5 (ITS 3.9.3) requires an RHR "loop" to be "OPERABLE and in operation."

4. The correct title for the LCO has been provided, since this LCO is what meets Criterion 4, not the entire RHR System.

5. 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.

6. Changes have been made to be consistent with the actual wording in the Specification. ISTS 3.9.5 (ITS 3.9.3) does not require reactor coolant temperature indication to be OPERABLE; it requires an RHR loop to be OPERABLE and in operation.

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

8. The wording has been modified since Section 3.5 does not provide requirements for the RHR decay heat removal function.

9. Changes made to be consistent with changes made to the Specification.

10. 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.

11. ISTS 3.9.5 has been renumbered to ITS 3.9.3 since ISTS 3.9.2 has not been included in the KPS ITS and ISTS 3.9.4 is being numbered as ITS 3.9.6.

12. Typographical error corrected.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 67 of 181

Attachment 1, Volume 14, Rev. 1, Page 68 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 68 of 181

Attachment 1, Volume 14, Rev. 1, Page 69 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.3, RHR AND COOLANT CIRCULATION - HIGH WATER LEVEL There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 69 of 181

Attachment 1, Volume 14, Rev. 1, Page 70 of 181 ATTACHMENT 4 ITS 3.9.4, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL Attachment 1, Volume 14, Rev. 1, Page 70 of 181

, Volume 14, Rev. 1, Page 71 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 71 of 181

Attachment 1, Volume 14, Rev. 1, Page 72 of 181 ITS A01 ITS 3.9.4 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 See ITS 3.4.5 and

1. Reactor Coolant Pumps 3.4.6 LCO 3.9.4 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.

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 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 14, Rev. 1, Page 72 of 181

Attachment 1, Volume 14, Rev. 1, Page 73 of 181 ITS A01 ITS 3.9.4 Add proposed LCO 3.9.4 Note 2 L01 LCO 3.9.4 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.

1. Each residual heat removal train shall be comprised of:

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.

or both M02

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.

A03 Add proposed Required Action A.2 Add proposed ACTION B M03 Add proposed SR 3.9.4.1 and SR 3.9.4.2 M04 Amendment No. 165 TS 3.1-2 03/11/2003 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 73 of 181

Attachment 1, Volume 14, Rev. 1, Page 74 of 181 DISCUSSION OF CHANGES ITS 3.9.4, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL 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 200ºF and irradiated fuel is in the reactor. ITS 3.9.4, which also includes similar RHR train OPERABILITY requirements, is applicable in MODE 6 with the water level < 23 ft above the top of the reactor vessel flange. RHR requirements in MODE 5 are provided in LCO 3.4.7 and LCO 3.4.8 and the MODE 6 RHR requirements with the water level 23 ft above the top of the reactor vessel flange are provided in LCO 3.9.3.

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.

A03 When one required RHR train is inoperable, CTS 3.1.a.2.B.2 requires immediate action to be taken to restore the train to OPERABLE status. Under similar conditions, ITS 3.9.4 ACTION A requires an action similar to CTS 3.1.a.2.B.2 (Required Action A.1), but also includes an additional action that can be taken.

In lieu of performing ITS 3.9.4 Required Action A.1, ITS 3.9.4 Required Action A.2 requires immediate initiation of action to establish 23 ft of water above the top of the reactor vessel flange. This changes the CTS by adding a new Required Action to CTS 3.1.a.2.B.2.

The addition of the new Required Action is acceptable since it is already an option, albeit an unstated option. CTS 3.1.a.2.B only requires one RHR train to be OPERABLE when in the REFUELING MODE (equivalent to ITS MODE 6) if the water level is 23 ft above the top of the reactor vessel flange. Thus, if only one required RHR loop is inoperable, then in lieu of restoring the inoperable RHR train, it is acceptable to just raise water level to 23 ft above the top of the reactor vessel flange in order to meet the CTS 3.1.a.2.B allowance. Thus, this change is designated as administrative since the technical requirements (as they relate to this specific change) have not been changed.

MORE RESTRICTIVE CHANGES M01 CTS 3.1.a.1.A, which (in MODE 6) requires a residual heat removal (RHR) pump to be in operation, is applicable in MODE 6 only when a reduction is made in the boron concentration of the reactor coolant. ITS 3.9.4, in part, requires an RHR loop to be in operation and is applicable at all times when in MODE 6 with the Kewaunee Power Station Page 1 of 5 Attachment 1, Volume 14, Rev. 1, Page 74 of 181

Attachment 1, Volume 14, Rev. 1, Page 75 of 181 DISCUSSION OF CHANGES ITS 3.9.4, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL water level < 23 ft above the top of the reactor vessel flange, except as allowed in ITS 3.9.4 LCO Note 1. Note 1 allows all RHR pumps to not be in operation for 15 minutes when switching from one loop to the other provided a) core outlet temperature is maintained > 10ºF 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 3.9.1; 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 6 with the water level < 23 ft above the top of the reactor vessel flange at all times except for the conditions specified in the ITS 3.9.4 LCO Note 1.

In MODE 6, 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 REFUELING (equivalent to ITS MODE 6) with the water level

< 23 ft above the top of the reactor vessel flange. Other MODES are discussed in other ITS discussions. The ITS 3.9.4 Applicability is MODE 6 at all times with the water level < 23 ft above the top of the reactor vessel flange, except for the conditions in the ITS 3.9.4 LCO Note 1, and primarily addresses the decay heat removal function, but also covers the boron mixing issue. RHR pumps are utilized in MODE 6 to provide for removal of decay heat from the reactor core to the RHR heat exchangers, and to ensure proper boron mixing within the reactor coolant. The MODE 6 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 6 at all times with the water level < 23 ft above the top of the reactor vessel flange, except for the condition specified in the ITS 3.9.4 LCO Note 1.

M02 CTS 3.1.a.2.B does not contain any ACTIONS to take if both required RHR loops are inoperable in MODE 6. 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 unit is already in MODE 6, the CTS 3.0.c requirement does not really provide any relevant compensatory measures. ITS 3.9.4 ACTION A provides the Required Actions when one or both required RHR loops are inoperable. The Required Actions are to immediately initiate action to restore the required RHR loops to Kewaunee Power Station Page 2 of 5 Attachment 1, Volume 14, Rev. 1, Page 75 of 181

Attachment 1, Volume 14, Rev. 1, Page 76 of 181 DISCUSSION OF CHANGES ITS 3.9.4, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL OPERABLE status or to immediately initiate action to establish 23 ft of water above the top of the reactor vessel flange. 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 a DBA occurring during the allowed Completion Times. The immediate initiation of action to restore the required loops to OPERABLE status or to establish 23 ft of water above the top of the reactor vessel flange reflects the importance of maintaining operation for decay heat removal. This change is designated as more restrictive because a new ACTION is being added to the ITS that was not required by the CTS.

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. 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 6, the CTS does not provide any compensatory measures. Therefore, 10 CFR 50.36 (c)(2)(i) would apply, which states to shut down 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 3.9.4 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 boron concentration of LCO 3.9.1 (Required Action B.1), immediate initiation of action to restore one RHR loop to operation (Required Action B.2), closing the equipment hatch and securing it with four bolts in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Required Action B.3), closing one door in each air lock in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Required Action B.4), and a verification that each penetration providing direct access from the containment atmosphere to the outside atmosphere is either closed with a manual or automatic isolation valve, blind flange, or equivalent, or is capable of being closed by an OPERABLE Containment Purge and Vent Isolation System in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (Required Action B.5). This changes the CTS by adding a new ACTION.

The purpose of the Actions should be to place the unit outside of the Applicability of the Specification. ITS 3.9.4 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 boron concentration of LCO 3.9.1 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.

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 3.9.4.1 requires verification that each Kewaunee Power Station Page 3 of 5 Attachment 1, Volume 14, Rev. 1, Page 76 of 181

Attachment 1, Volume 14, Rev. 1, Page 77 of 181 DISCUSSION OF CHANGES ITS 3.9.4, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL required RHR loop is in operation and circulating reactor coolant 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 3.9.4.2 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 3.9.4.1 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 3.9.4.2 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.

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 3.9.4 requires two RHR loops to be OPERABLE, but does not define the components and the associated flow path that comprise each 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 loop in MODE 6. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases.

Kewaunee Power Station Page 4 of 5 Attachment 1, Volume 14, Rev. 1, Page 77 of 181

Attachment 1, Volume 14, Rev. 1, Page 78 of 181 DISCUSSION OF CHANGES ITS 3.9.4, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL 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 3.9.4 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 6 with the water level < 23 ft above the top of the reactor vessel flange. 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 the only time when such testing is safe and possible. 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.

Kewaunee Power Station Page 5 of 5 Attachment 1, Volume 14, Rev. 1, Page 78 of 181

Attachment 1, Volume 14, Rev. 1, Page 79 of 181 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 79 of 181

Attachment 1, Volume 14, Rev. 1, Page 80 of 181 CTS RHR and Coolant Circulation - Low Water Level 3.9.6 8 4

3.9 REFUELING OPERATIONS 8

3.9.6 4 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level 3.1.a.1.A, LCO 3.9.6 4 Two RHR loops shall be OPERABLE, and one RHR loop shall be in 8 3.1.a.2.B operation.

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

DOC M01 1. All RHR pumps may be removed from operation for 15 minutes when switching from one train to another provided: 1

º loop 1

a. The core outlet temperature is maintained > 10 degrees F 2

below saturation temperature, ;

b. No operations are permitted that would cause introduction of coolant into the Reactor Coolant System (RCS) with boron concentration less than that required to meet the minimum 3

required boron concentration of LCO 3.9.1, and

, "Boron Concentration";

c. No draining operations to further reduce RCS water volume are permitted.

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 4

surveillance testing, provided that the other RHR loop is OPERABLE and in operation.

3.1.a.1.A, APPLICABILITY: MODE 6 with the water level < 23 ft above the top of reactor vessel 3.1.a.2.B flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.1.a.2.B.2 A. Less than the required A.1 Initiate action to restore Immediately number of RHR loops required RHR loops to OPERABLE. OPERABLE status.

OR A.2 Initiate action to establish Immediately 23 ft of water above the top of reactor vessel flange.

WOG STS 3.9.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 80 of 181

Attachment 1, Volume 14, Rev. 1, Page 81 of 181 CTS RHR and Coolant Circulation - Low Water Level 3.9.6 8

4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME DOC M03 B. No RHR loop in B.1 Suspend operations that Immediately operation. would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9.1.

AND B.2 Initiate action to restore one Immediately RHR loop to operation.

AND B.3 Close equipment hatch and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months secure with [four] bolts. 5 AND B.4 Close one door in each air 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months lock.

AND Verify B.5.1 Close each penetration 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months providing direct access from the containment atmosphere to the outside is either closed atmosphere with a manual or automatic isolation valve, blind flange, or equivalent.

6 OR ; or Add words from ISTS Required Action B.5.2 on next page to here WOG STS 3.9.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 81 of 181

Attachment 1, Volume 14, Rev. 1, Page 82 of 181 CTS RHR and Coolant Circulation - Low Water Level 3.9.6 8 Move to Required Action 4 B.5 on previous page 6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME DOC M03 B.5.2 Verify each penetration is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months capable of being closed by 6

an OPERABLE Containment Purge and Vent Exhaust Isolation System.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 8

DOC M04 SR 3.9.6.1 Verify one RHR loop is in operation and circulating 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 4 reactor coolant at a flow rate of [2800] gpm. 9 8

DOC M04 SR 3.9.6.2 Verify correct breaker alignment and indicated 7 days 4 power available to the required RHR pump that is not in operation.

NOTE------------------------- 7 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.

WOG STS 3.9.6-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 82 of 181

Attachment 1, Volume 14, Rev. 1, Page 83 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL

1. The term "train" in ISTS 3.9.6 LCO Note 1 has been changed to "loop" to be consistent with the actual LCO statement, which requires loops to be OPERABLE and in operation, not trains. In addition, the term "degrees" has been replaced with the unit designator "º" consistent with its use throughout the ISTS (see ISTS LCO 3.4.6, 3.4.7, and 3.4.8 Note 1).

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

4. Typographical error corrected.

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 requirements.

6. ISTS 3.9.6 Required Actions B.5.1 and B.5.2 are connected by an "OR" logical connector, such that either one can be performed to meet the requirements of the ACTION. However, the two Required Actions are applicable to all the penetrations; either Required Action B.5.1 or Required Action B.5.2 must be performed for all the penetrations. Thus, this will not allow one penetration to be isolated by use of a manual valve and another penetration to be capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. This is not the intent of the requirement. The requirement is based on ISTS LCO 3.9.4, which requires each penetration to be either: a) closed by a manual or automatic isolation valve, blind flange, or equivalent; or b) capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System. For consistency with the actual LCO requirement, ISTS 3.9.6 Required Actions B.5.1 and B.5.2 have been combined into a single Required Action in ITS 3.9.4 Required Action B.5. Furthermore, the title of the system has been changed to be consistent with the Kewaunee Power Station System name (i.e., Containment Purge and Vent Isolation System).

7. TSTF-265 was previously approved and incorporated in NUREG-1431, Rev. 2, in similar SRs (e.g., ISTS SRs 3.4.5.3, 3.4.6.3, 3.4.7.3, and 3.4.8.2). Consistent with TSTF-265, a Note is added to ITS SR 3.9.4.2 that permits the performance of the SR to verify correct breaker alignment and power availability to be delayed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after a required RHR pump is not in operation. This provision is required because when RHR pumps are swapped under the current requirements, the Surveillance is immediately not met on the RHR pump taken out of operation. This change avoids entering an Action for a routine operational occurrence. The change is acceptable because adequate assurance exists that the RHR pump is aligned to the correct breaker with power available because, prior to being removed from operation, the applicable pump had been in operation. Allowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to perform the breaker alignment verification is acceptable because the RHR pump was in operation, which demonstrated OPERABILITY, and because 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed in the ISTS by invoking SR 3.0.3. This is also a new Surveillance Requirement not required in CTS 3.1.a.2.B.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 83 of 181

Attachment 1, Volume 14, Rev. 1, Page 84 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL

8. ISTS 3.9.6 has been renumbered to ITS 3.9.4 since ISTS 3.9.2 has not been included in the KPS ITS and ISTS 3.9.4 is being numbered as ITS 3.9.6.

9. The minimum required RHR flow rate requirement has not been included in KPS ITS SR 3.9.4.1. The Bases states that the flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. However, the decay heat is not always the same; it is a function of time after shutdown and the power history of the fuel. Furthermore, SRs in Section 3.4 (ITS SRs 3.4.6.1, 3.4.7.1, and 3.4.8.1) require a similar verification that the RHR loop is in operation, but do not specify a flow rate requirement. This change is also consistent with the KPS current Technical Specifications, which does not specify a flow rate for the RHR loop.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 84 of 181

Attachment 1, Volume 14, Rev. 1, Page 85 of 181 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 85 of 181

Attachment 1, Volume 14, Rev. 1, Page 86 of 181 RHR and Coolant Circulation - Low Water Level B 3.9.6 14 4

B 3.9 REFUELING OPERATIONS B 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level 14 4

BASES BACKGROUND The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by 1 GDC 34, to provide mixing of borated coolant, and to prevent boron stratification (Ref. 1). Heat is removed from the RCS by circulating Refs. 1 and 2 2 reactor coolant through the RHR heat exchangers where the heat is transferred to the Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg(s). Operation of the RHR 2 an System for normal cooldown decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of reactor coolant through the RHR heat exchanger(s) and the bypass lines. Mixing of the reactor coolant is maintained by this continuous circulation of reactor coolant through the RHR System.

APPLICABLE If the reactor coolant temperature is not maintained below 200°F, boiling SAFETY of the reactor coolant could result. This could lead to a loss of coolant in ANALYSES the reactor vessel. Additionally, boiling of the reactor coolant could lead to a reduction in boron concentration in the coolant due to the boron plating out on components near the areas of the boiling activity. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant will eventually challenge the integrity of the fuel cladding, which is loops a fission product barrier. Two trains of the RHR System are required to 3 be OPERABLE, and one train in operation, in order to prevent this challenge. loop and Coolant Circulation - Low Water Level 4 The RHR System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO In MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, both RHR loops must be OPERABLE. Additionally, one loop of RHR must be in operation in order to provide:

5

a. Removal of decay heat, ; and

b. Mixing of borated coolant to minimize the possibility of criticality, and 6

c. Indication of reactor coolant temperature.

WOG STS B 3.9.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 86 of 181

Attachment 1, Volume 14, Rev. 1, Page 87 of 181 RHR and Coolant Circulation - Low Water Level B 3.9.6 14 4

BASES LCO (continued)

This LCO is modified by two Notes. Note 1 permits the RHR pumps to be loop removed from operation for 15 minutes when switching from one train to 3 another. The circumstances for stopping both RHR pumps are to be limited to situations when the outage time is short [and the core outlet 7 temperature is maintained > 10 degrees F below saturation temperature].

The Note prohibits boron dilution or draining operations when RHR forced 8 flow is stopped. º 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 the other loop is OPERABLE and in operation. Prior to declaring the loop inoperable, consideration should be given to the existing plant configuration. This consideration should include that the core time to boil is short, there is no draining operation to further reduce RCS water level and that the capability exists to inject borated water into the reactor vessel. This permits surveillance tests to be performed on the 9 inoperable loop during a time when these tests are safe and possible.

An OPERABLE RHR loop consists of an RHR pump, a heat exchanger, valves, piping, instruments and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in 6 one of the RCS hot legs and is returned to the RCS cold legs. 2 an Both RHR pumps may be aligned to the Refueling Water Storage Tank to support filling or draining the refueling cavity or for performance of required testing.

APPLICABILITY Two RHR loops are required to be OPERABLE, and one RHR loop must be in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the RHR System in other MODES are covered by LCOs in 10 Section 3.4, Reactor Coolant System (RCS), and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level 23 ft are located in LCO 3.9.5, "Residual Heat Removal 14 (RHR) and Coolant Circulation - High Water Level." 3 WOG STS B 3.9.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 87 of 181

Attachment 1, Volume 14, Rev. 1, Page 88 of 181 RHR and Coolant Circulation - Low Water Level B 3.9.6 14 4

BASES ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE, action shall be immediately initiated and continued until the RHR loop is restored to OPERABLE status and to operation or until 23 ft of water level is established above the reactor vessel flange. When the water level is 3

23 ft above the reactor vessel flange, the Applicability changes to that of 14 LCO 3.9.5, and only one RHR loop is required to be OPERABLE and in 11 operation. An immediate Completion Time is necessary for an operator to initiate corrective actions.

B.1 If no RHR loop is in operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that what would be required in 9 the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.

B.2 If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation. Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.

12 B.3, B.4, B.5.1, and B.5.2 If no RHR is in operation, the following actions must be taken:

a. The equipment hatch must be closed and secured with [four] bolts, 7 5 5

b. One door in each air lock must be closed, and WOG STS B 3.9.6-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 88 of 181

Attachment 1, Volume 14, Rev. 1, Page 89 of 181 RHR and Coolant Circulation - Low Water Level B 3.9.6 14 4

BASES ACTIONS (continued)

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE 2

Vent Containment Purge and Exhaust Isolation System.

With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.

Performing the actions stated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.

SURVEILLANCE SR 3.9.6.1 14 REQUIREMENTS 4 Circulating coolant This Surveillance demonstrates that one RHR loop is in operation and ensures thermal and boron stratification are circulating reactor coolant. The flow rate is determined by the flow rate 12 minimized. necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator for monitoring the RHR System in the control room.

SR 3.9.4.2 9 14 Verification that the required pump is OPERABLE ensures that an 13 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 the required pump. 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.

U 7 2

REFERENCES 1. FSAR, Section [5.5.7].

14.1.4.2

2. USAR, Section 9.3.1.2. 2 This SR is modified by a Note that states the SR is not 12 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.

WOG STS B 3.9.6-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 89 of 181

Attachment 1, Volume 14, Rev. 1, Page 90 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4 BASES, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL

1. Kewaunee Power Station (KPS) was designed and under construction prior to the promulgation of 10 CFR 50 Appendix A. KPS was designed and constructed to meet the intent of the proposed General Design Criteria, published in 1967. The KPS USAR Section 1.8 provides a description of each of the proposed General Design Criteria and how KPS meets the intent of each one. However, the proposed General Design Criteria did not have a criteria equivalent to 10 CFR 50 Appendix A, GDC 34.

Therefore, reference to this GDC has been deleted.

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 have been made to be consistent with the actual wording in the Specification. ISTS 3.9.6 (ITS 3.9.4) requires RHR loops to be OPERABLE and in operation, not trains.

4. The correct title for the LCO has been provided, since this LCO is what meets Criterion 4, not the entire RHR System.

5. 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.

6. Changes have been made to be consistent with the actual wording in the Specification. ISTS 3.9.6 (ITS 3.9.4) does not require reactor coolant temperature indication to be OPERABLE; it requires two RHR loops to be OPERABLE and one in operation.

7. 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.

8. The term "degrees" has been replaced with the unit designator "º" consistent with its use throughout the ISTS (see ISTS LCO 3.4.6, 3.4.7, and 3.4.8 Note 1).

9. Typographical error corrected.

10. The wording has been modified since Section 3.5 does not provide requirements for the RHR decay heat removal function.

11. The discussion in the ACTION is concerning how many RHR loops are required to be OPERABLE in the two LCOs. Both LCOs have the same "in operation" requirement. Thus, the additional LCO phrase "and in operation" is not germane to the discussion and has been deleted.

12. Changes made to be consistent with changes made to the Specification.

13. The words "RCS or" have been deleted since the actual SR is for RHR pumps only, not reactor coolant pumps.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 90 of 181

Attachment 1, Volume 14, Rev. 1, Page 91 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4 BASES, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL

14. ISTS 3.9.6 has been renumbered to ITS 3.9.4 since ISTS 3.9.2 has not been included in the KPS ITS and ISTS 3.9.4 is being numbered as ITS 3.9.6.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 91 of 181

Attachment 1, Volume 14, Rev. 1, Page 92 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 92 of 181

Attachment 1, Volume 14, Rev. 1, Page 93 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.4, RHR AND COOLANT CIRCULATION - LOW WATER LEVEL There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 93 of 181

, Volume 14, Rev. 1, Page 94 of 181 ATTACHMENT 5 ITS 3.9.5, REFUELING CAVITY WATER LEVEL , Volume 14, Rev. 1, Page 94 of 181

, Volume 14, Rev. 1, Page 95 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 95 of 181

Attachment 1, Volume 14, Rev. 1, Page 96 of 181 A01 ITS ITS 3.9.5 3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS.

OBJECTIVE To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION Applicability a. During REFUELING OPERATIONS: L01

1. Containment Closure

a. The equipment hatch shall be closed and at least one door in each personnel air lock shall be capable of being closed (1) in 30 minutes or less. In addition, at See ITS least one door in each personnel air lock shall be closed when the reactor vessel 3.9.6 head or upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valve or an operable automatic isolation valve.

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage See CTS 3.8.a.2 pool shall be monitored continuously.

See CTS

3. The reactor will be subcritical for 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months prior to movement of its irradiated fuel 3.8.a.3 assemblies. Core subcritical neutron flux shall be continuously monitored by at least See ITS two neutron monitors, each with continuous visual indication in the control room and 3.9.2 one with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux monitor shall be in service.

4. At least one residual heat removal pump shall be OPERABLE. See ITS 3.9.3

5. When there is fuel in the reactor, a minimum boron concentration as specified in the COLR shall be maintained in the Reactor Coolant System during reactor vessel See ITS head removal or while loading and unloading fuel from the reactor. The required 3.9.1 boron concentration shall be verified by chemical analysis daily.

(1)

Administrative controls ensure that:

Appropriate personnel are aware that both personnel air lock doors are open,

A specified individual(s) is designated and available to close the air lock following a required See ITS evacuation of containment, and 3.9.6

Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed.

Amendment No. 165 TS 3.8-1 03/11/2003 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 96 of 181

Attachment 1, Volume 14, Rev. 1, Page 97 of 181 ITS ITS 3.9.5 A01

6. Direct communication between the control room and the operating floor of the See CTS containment shall be available whenever changes in core geometry are taking 3.8.a.6 place.

7. Deleted.

8. The containment ventilation and purge system, including the capability to initiate See ITS automatic containment ventilation isolation, shall be tested and verified to be 3.9.6 operable immediately prior to and daily during REFUELING OPERATIONS.

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be operating during fuel handling and when any load is carried over the pool if irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool See CTS 3.8.a.9 sweep system, including the charcoal adsorber, is not operating when required, fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

LCO 3.9.5 10. The minimum water level above the vessel flange shall be maintained at 23 feet.

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes for this test must be equal to or greater than the maximum load to be assumed by See CTS 3.8.a.11 the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

See CTS

12. A licensed senior reactor operator will be on-site and designated in charge of the 3.8.a.12 REFUELING OPERATIONS.

ACTION A b. If any of the specified limiting conditions for REFUELING OPERATIONS are not met, refueling of the reactor shall cease. Work shall be initiated to correct the violated L02 conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be performed.

Add proposed SR 3.9.5.1 M01 Amendment No. 200 TS 3.8-2 11/20/2008 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 97 of 181

Attachment 1, Volume 14, Rev. 1, Page 98 of 181 DISCUSSION OF CHANGES ITS 3.9.5, REFUELING CAVITY WATER LEVEL 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 3.8.A.10 requires the minimum water level above the vessel flange to be maintained at 23 feet during REFUELING OPERATIONS, but does not provide a Surveillance Requirement to periodically verify this requirement is met. ITS SR 3.9.5.1 requires verification that the refueling water cavity level is 23 feet above the top of the reactor vessel flange once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This changes the CTS by adding a specific Surveillance to periodically verify this requirement.

The purpose of SR 3.9.5.1 is to verify the water level of the refueling cavity is sufficient to ensure that the analysis assumptions of the postulated fuel handling accident during refueling operations are met. Maintaining water at the required level above the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment. This change is acceptable because a specific Surveillance has been added to verify that the LCO will be met. This change is designated as more restrictive because a Surveillance has been added to the ITS that is not required by the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L01 (Category 2 - Relaxation of Applicability) CTS 3.8.a.10 is applicable during REFUELING OPERATIONS. ITS 3.9.5 is applicable during movement of irradiated fuel assemblies within containment. This changes the CTS by requiring the minimum water level above the reactor vessel flange to be maintained at 23 feet only during times when irradiated fuel assemblies are being moved within containment, in lieu of during REFUELING OPERATIONS.

Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 98 of 181

Attachment 1, Volume 14, Rev. 1, Page 99 of 181 DISCUSSION OF CHANGES ITS 3.9.5, REFUELING CAVITY WATER LEVEL The purpose of CTS 3.8.a.10 is to ensure that the refueling cavity water level is greater than or equal to that assumed in the fuel handling accident analysis.

CTS 3.8.a.10 requires the minimum water level above the vessel flange to be maintained at 23 feet during REFUELING OPERATIONS. As defined in CTS Section 1.0, REFUELING OPERATIONS is movement of reactor vessel internal components that could affect the reactivity of the core within the containment when the vessel head is unbolted or removed. This would include both moving irradiated fuel assemblies and control rods. Since movement of irradiated fuel assemblies within containment (specifically within the refueling cavity as stated in ITS 3.9.5) can only occur with water level 23 ft above the top of the reactor vessel flange, with respect to moving fuel assemblies, ITS 3.9.5 is applicable under the same conditions as CTS 3.8.a.10. 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 fuel handling accident is based on damaging a single irradiated fuel assembly. Movement of control rods is not assumed to result in a fuel handling accident. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS.

L02 (Category 4 - Relaxation of Required Action) CTS 3.8.a.10 requires the minimum water level above the vessel flange to be maintained at 23 feet during REFUELING OPERATIONS (which is defined as movement of reactor vessel internal components that could affect the reactivity of the core within the containment when the reactor head is unbolted or removed). If this requirement is not met, CTS 3.8.b requires refueling of the reactor to cease. In addition, CTS 3.8.b, requires work to be initiated to correct the violated conditions so the specified limits are met and no operations which may increase the reactivity of the core shall be performed. ITS 3.9.5 ACTION A only requires immediate suspension of movement of irradiated fuel assemblies within containment. This changes the CTS by eliminating the requirement for work to be initiated to correct the violated conditions so the specified limits are met and the requirement that no operations which may increase the reactivity of the core be performed.

The purpose of 3.8.b is to place the unit in a condition in which the LCO does not apply by ceasing the refueling of the reactor. ITS 3.9.5 ACTION A accomplishes the equivalent by the immediate suspension of movement of irradiated fuel assemblies within containment. In addition, the CTS requirements to initiate work to correct the violated conditions so the specified limits are met and ceasing operations that may increase the reactivity of the core are not required since the cessation of the movement of irradiated fuel assemblies places the unit in a condition in which the LCO does not apply. By placing the unit in a condition in which the LCO does not apply, the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analyses is minimized.

If irradiated fuel assemblies are not being moved within containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Thus, this change is acceptable and the deletion of the other above described Required Actions is designated as less restrictive.

Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 99 of 181

Attachment 1, Volume 14, Rev. 1, Page 100 of 181 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 100 of 181

Attachment 1, Volume 14, Rev. 1, Page 101 of 181 CTS Refueling Cavity Water Level 3.9.7 1 5

3.9 REFUELING OPERATIONS 5

1 3.9.7 Refueling Cavity Water Level 5

1 3.8.a.10 LCO 3.9.7 Refueling cavity water level shall be maintained 23 ft above the top of reactor vessel flange.

3.8.a APPLICABILITY: During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.8.b A. Refueling cavity water A.1 Suspend movement of Immediately level not within limit. irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M01 SR 3.9.7.1 Verify refueling cavity water level is 23 ft above 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 5 the top of reactor vessel flange.

WOG STS 3.9.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 101 of 181

Attachment 1, Volume 14, Rev. 1, Page 102 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.5, REFUELING CAVITY WATER LEVEL

1. ISTS 3.9.7 has been renumbered to ITS 3.9.5 since ISTS 3.9.2 and ISTS 3.9.4 have not been included in the Kewaunee Power Station (KPS) ITS.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 102 of 181

Attachment 1, Volume 14, Rev. 1, Page 103 of 181 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 103 of 181

Attachment 1, Volume 14, Rev. 1, Page 104 of 181 All changes are 1 Refueling Cavity Water Level B 3.9.7 2 unless otherwise noted 5

B 3.9 REFUELING OPERATIONS 5

2 B 3.9.7 Refueling Cavity Water Level BASES BACKGROUND The movement of irradiated fuel assemblies within containment requires a minimum water level of 23 ft above the top of the reactor vessel flange.

During refueling, this maintains sufficient water level in the containment, reactor refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool.

Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to 1 50.67 < 25% of 10 CFR 100 limits, as provided by the guidance of Reference 3. 3 fuel transfer canal APPLICABLE During movement of irradiated fuel assemblies, the water level in the SAFETY refueling canal and the refueling cavity is an initial condition design ANALYSES parameter in the analysis of a fuel handling accident in containment, as 1.183 postulated by Regulatory Guide 1.25 (Ref. 1). A minimum water level of Appendix B 23 ft (Regulatory Position C.1.c of Ref. 1) allows a decontamination factor 3 200 of 100 (Regulatory Position C.1.g of Ref. 1) to be used in the accident 99.5 analysis for iodine. This relates to the assumption that 99% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 10% of the total fuel rod iodine INSERT 1 inventory (Ref. 1).

The fuel handling accident analysis inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay 100 time of [X] hours prior to fuel handling, the analysis and test programs 4 demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Refs. 4 and 5).

3 4 Refueling cavity water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO A minimum refueling cavity water level of 23 ft above the reactor vessel flange is required to ensure that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits, as provided by the guidance of Reference 3. 1 WOG STS B 3.9.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 104 of 181

Attachment 1, Volume 14, Rev. 1, Page 105 of 181 B 3.9.5 1

INSERT 1 The method of analysis used for evaluating the potential radiological consequences of a fuel handling accident complies with Reference 1, except that 50% of the damaged fuel rods do not comply with footnote 11 of Reference 1. For these fuel rods, the gap activity fractions used are taken from Regulatory Guide 1.25 (Reference 5), as modified by NUREG/CR-5009 (Reference 6).

Insert Page B 3.9.7-1 Attachment 1, Volume 14, Rev. 1, Page 105 of 181

Attachment 1, Volume 14, Rev. 1, Page 106 of 181 Refueling Cavity Water Level B 3.9.7 2 5

BASES 5

APPLICABILITY LCO 3.9.7 is applicable when moving irradiated fuel assemblies within containment. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety being 5 moved analysis. If irradiated fuel assemblies are not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.15, "Fuel Storage Pool Water 7 Level." 13 Spent ACTIONS A.1 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving or movement of irradiated fuel assemblies within the 6 containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.

The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS 5 Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.

Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls of valve positions, which make significant unplanned level changes unlikely.

REFERENCES 1. Regulatory Guide 1.25, March 23, 1972. July 2000 3 1.183 U

2. FSAR, Section [15.4.5]. 14.2.1 1 4

3. NUREG-0800, Section 15.7.4. 1 3 4. 10 CFR 100.10. 50.67 3 4 5. Malinowski, D. D., Bell, M. J., Duhn, E., and Locante, J., WCAP-7828, Radiological Consequences of a Fuel Handling Accident, December 1971.

1

5. Regulatory Guide 1.25, March 23, 1972.

6. NUREG/CR-5009, February 1988.

WOG STS B 3.9.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 106 of 181

Attachment 1, Volume 14, Rev. 1, Page 107 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.5 BASES, REFUELING CAVITY WATER LEVEL

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. ISTS 3.9.7 has been renumbered to ITS 3.9.5 since ISTS 3.9.2 and ISTS 3.9.4 have not been included in the Kewaunee Power Station (KPS) ITS.

3. Changes are made to the ISTS Bases which reflect the KPS design. License Amendment 166, issued March 17, 2003 (ADAMS accession No. ML030210062)

(as modified by Amendment 190, dated March 8, 2007, ADAMS accession No.

ML070430020), revised the radiological consequence analyses for the KPS design basis accidents to implement the alternate source term (AST) as described in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design-Basis Accidents at Nuclear Power Reactors" and pursuant to 10 CFR 50.67, "Accident Source Term."

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

5. Changes are made to be consistent with the ISTS.

6. Typographical error corrected.

7. The correct LCO number and title are provided.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 107 of 181

Attachment 1, Volume 14, Rev. 1, Page 108 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 108 of 181

Attachment 1, Volume 14, Rev. 1, Page 109 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.5, REFUELING CAVITY WATER LEVEL There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 109 of 181

, Volume 14, Rev. 1, Page 110 of 181 ATTACHMENT 6 ITS 3.9.6, CONTAINMENT PENETRATIONS , Volume 14, Rev. 1, Page 110 of 181

, Volume 14, Rev. 1, Page 111 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 111 of 181

Attachment 1, Volume 14, Rev. 1, Page 112 of 181 ITS A01 ITS 3.9.6 3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS.

OBJECTIVE To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION Applicability

a. During REFUELING OPERATIONS: L05

1. Containment Closure LA01 LCO 3.9.6.a

a. The equipment hatch shall be closed and at least one door in each personnel air and b lock shall be capable of being closed (1) in 30 minutes or less. In addition, at L01 least one door in each personnel air lock shall be closed when the reactor vessel head or upper internals are lifted.

L02

b. Each line that penetrates containment and which provides a direct air path from LCO 3.9.6.c.1 containment atmosphere to the outside atmosphere shall have a closed isolation and 2 valve or an operable automatic isolation valve. , blind flange, or equivalent See CTS

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage 3.8.a.2 pool shall be monitored continuously.

See CTS

3. The reactor will be subcritical for 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months prior to movement of its irradiated fuel 3.8.a.3 assemblies. Core subcritical neutron flux shall be continuously monitored by at least two neutron monitors, each with continuous visual indication in the control room and See ITS 3.9.2 one with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux monitor shall be in service.

See ITS

4. At least one residual heat removal pump shall be OPERABLE. 3.9.3

5. When there is fuel in the reactor, a minimum boron concentration as specified in the COLR shall be maintained in the Reactor Coolant System during reactor vessel See ITS 3.9.1 head removal or while loading and unloading fuel from the reactor. The required boron concentration shall be verified by chemical analysis daily.

(1)

Administrative controls ensure that:

Appropriate personnel are aware that both personnel air lock doors are open,

A specified individual(s) is designated and available to close the air lock following a required LA01 evacuation of containment, and

Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed.

Amendment No. 165 TS 3.8-1 03/11/2003 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 112 of 181

Attachment 1, Volume 14, Rev. 1, Page 113 of 181 ITS ITS 3.9.6 A01

6. Direct communication between the control room and the operating floor of the See CTS containment shall be available whenever changes in core geometry are taking 3.8.a.6 place.

7. Deleted. on an actual or simulated actuation signal L03

8. The containment ventilation and purge system, including the capability to initiate LCO 3.9.6.c.2 automatic containment ventilation isolation, shall be tested and verified to be SR 3.9.6.2 operable immediately prior to and daily during REFUELING OPERATIONS. L04 Add proposed SR 3.9.6.2 Note

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be L04 operating during fuel handling and when any load is carried over the pool if irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool See CTS sweep system, including the charcoal adsorber, is not operating when required, 3.8.a.9 fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

See ITS 5.5.9

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

See ITS

10. The minimum water level above the vessel flange shall be maintained at 23 feet. 3.9.5

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes See CTS for this test must be equal to or greater than the maximum load to be assumed by 3.8.a.11 the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

12. A licensed senior reactor operator will be on-site and designated in charge of the See CTS 3.8.a.12 REFUELING OPERATIONS.

ACTION A b. If any of the specified limiting conditions for REFUELING OPERATIONS are not met, refueling of the reactor shall cease. Work shall be initiated to correct the violated A02 conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be performed.

Add proposed SR 3.9.6.1 M01 Amendment No. 200 TS 3.8-2 11/20/2008 Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 113 of 181

Attachment 1, Volume 14, Rev. 1, Page 114 of 181 DISCUSSION OF CHANGES ITS 3.9.6, CONTAINMENT PENETRATIONS 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 When the containment penetrations are not in the required condition specified in CTS 3.8.a.1 or CTS 3.8.a.8, CTS 3.8.b requires refueling of the reactor to cease, initiation of action to restore the containment penetrations to the required conditions, and no operations be performed that could increase the reactivity of the core. Under similar conditions, ITS 3.9.6 ACTION A only requires movement of irradiated fuel assemblies within containment to be suspended. This changes the CTS by deleting the requirements to initiate action to restore the containment penetrations to the required conditions and that no operations be performed that could increase the reactivity of the core.

The purpose of CTS 3.8.a.8 is to ensure proper compensatory actions are taken to exit the Applicability of the LCO. CTS 3.8.a.1 and 3.8.a.8 are only applicable during REFUELING OPERATIONS, which is defined in CTS Section 1.0 as the movement of reactor vessel internals that could affect the reactivity of the core within the containment when the vessel head is unbolted or removed. Thus, after the first requirement of CTS 3.8.b is met (i.e., suspend refueling the reactor), the Applicability has been exited and thus, continuation of the requirements of CTS 3.8.b are not required. Therefore, this change is acceptable and is designated as administrative since the technical requirements have not been changed.

MORE RESTRICTIVE CHANGES M01 CTS 3.8.a does not provide a Surveillance Requirement to verify each required containment penetration is in the required status. The ITS adds a Surveillance Requirement (SR 3.9.6.1) to verify each required containment penetration is in the required status once every 7 days. This changes the CTS by adding a new Surveillance Requirement for the containment penetrations.

This change is acceptable because the added Surveillance Requirement ensures that each required containment penetration is in the required status to support the containment penetration conditions assumed in the Fuel Handling Accident (FHA) analysis. In addition, this change is acceptable because the Surveillance Requirement continues to ensure that the structures, systems, and components are maintained consistent with the safety analyses and licensing basis. This change is designated as more restrictive because a new Surveillance Requirement has been added.

Kewaunee Power Station Page 1 of 4 Attachment 1, Volume 14, Rev. 1, Page 114 of 181

Attachment 1, Volume 14, Rev. 1, Page 115 of 181 DISCUSSION OF CHANGES ITS 3.9.6, CONTAINMENT PENETRATIONS RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.8.a.1.a requires at least one door in each personnel air lock to be capable of being closed "within 30 minutes."

CTS 3.8.a.1.a is modified by a footnote (1) that states "Administrative controls ensure that appropriate personnel are aware that both personnel air lock doors are open; a specified individual(s) is designated and available to close the air lock following a required evacuation of containment; and, any obstruction(s)

(e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed." ITS 3.9.6 does not contain this footnote information or the 30 minute requirement. This changes the CTS by moving the information contained in the footnote and the 30 minute requirement to the Bases.

The removal of these details, which are related to procedural details for meeting Technical Specification 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. The ITS still retains the requirement for at least one door in each personnel air lock to be capable of being closed. 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 procedural details for meeting Technical Specification requirements is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES L01 (Category 1 - Relaxation of LCO Requirements) CTS 3.8.a.1.a, in part, requires at least one door in each personnel air lock to be closed when the reactor vessel head or upper internals are being lifted. ITS 3.9.6 does not include this requirement. This changes the CTS by not requiring one door in each personnel air lock to be closed when the reactor vessel head or upper internals are being lifted.

The purpose of CTS 3.8.a.1 is to ensure that if a fuel handling accident occurs, the release of any subsequent fission products results in doses that are well within the guideline values specified in Regulatory Guide 1.183. A fuel handling accident, as analyzed in USAR Section 14.2.1, is postulated to occur during handling of irradiated fuel assemblies; not handling the vessel head or upper internals. Thus, moving the vessel head or upper internals cannot result in a fuel handling accident. Any additional requirements, above those required to meet the assumptions of the fuel handling accident, are more appropriately controlled Kewaunee Power Station Page 2 of 4 Attachment 1, Volume 14, Rev. 1, Page 115 of 181

Attachment 1, Volume 14, Rev. 1, Page 116 of 181 DISCUSSION OF CHANGES ITS 3.9.6, CONTAINMENT PENETRATIONS by plant procedures. Therefore, this change is acceptable and is designated a less restrictive change since the requirement to maintain containment closure capability during movement of the vessel head or upper internals is being deleted from the CTS.

L02 (Category 1 - Relaxation of LCO Requirements) CTS 3.8.a.1.b, in part, states that each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valve. ITS LCO 3.9.6.c.1 states that each penetration providing direct access from the containment atmosphere to the outside atmosphere is closed by a manual or automatic isolation valve, blind flange, or equivalent. This changes the CTS by specifying the use of a blind flange or an equivalent means of isolating a containment penetration.

The purpose of CTS 3.8.a.1 is to ensure the containment penetrations are in the condition assumed in the Fuel Handling Accident (FHA) analysis. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses and licensing basis. The addition of the option to use a blind flange or some other equivalent means of isolating the containment penetration allows flexibility in the ITS that was not specified in CTS. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS.

L03 (Category 6 - Relaxation Of Surveillance Requirement Acceptance Criteria)

CTS 3.8.a.8 requires verification of the automatic actuation of the Containment Ventilation and Purge valves on a containment ventilation isolation signal. ITS SR 3.9.6.2 specifies that the signal may be from either an actual or simulated signal. This changes the CTS by explicitly allowing the use of either an actual or simulated signal to perform the Surveillance.

The purpose of CTS 3.8.a.8 is to ensure that the containment purge and vent valves operate correctly upon receipt of an actuation signal. This change is acceptable because it has been determined that the relaxed Surveillance Requirement acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its required functions. Equipment cannot discriminate between an "actual," "simulated," or "test" signal and, therefore, the results of the testing are unaffected by the type of signal used to initiate the test. This change allows taking credit for unplanned actuation if sufficient information is collected to satisfy the Surveillance test requirements.

The change also allows a simulated signal to be used, if necessary. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS.

L04 (Category 7 - Relaxation Of Surveillance Frequency) CTS 3.8.a.8 includes a Surveillance Frequency of "immediately prior to and daily during REFUELING OPERATIONS" for performing a Surveillance of the Containment Ventilation and Purge System. The ITS SR 3.9.6.2 Frequency for the same requirement is 18 months. ITS SR 3.9.6.2 is also modified by a Note that states that SR 3.9.6.2 is not required to be met for containment purge and vent valve(s) in penetrations that are closed to comply with LCO 3.9.6.c.1. This changes the CTS by Kewaunee Power Station Page 3 of 4 Attachment 1, Volume 14, Rev. 1, Page 116 of 181

Attachment 1, Volume 14, Rev. 1, Page 117 of 181 DISCUSSION OF CHANGES ITS 3.9.6, CONTAINMENT PENETRATIONS changing the Surveillance Frequency from immediately prior to and daily during REFUELING OPERATIONS to 18 months and adding the Note that the SR is not required to be met for containment purge and vent valve(s) in penetrations that are closed to comply with ITS LCO 3.9.6.c.1.

The purpose of CTS 3.8.a.8 is to verify the equipment required to meet the LCO is OPERABLE. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Containment purge and vent valve testing is still required, but at a Frequency consistent with the testing Frequency for containment isolation valves required in MODES 1, 2, 3, and 4. This Frequency provides an appropriate degree of assurance that the valves are OPERABLE. When containment purge and vent valve(s) in penetrations are closed to comply with ITS LCO 3.9.6.c.1, the penetrations are in the expected condition (isolated) to mitigate the effects of a fuel handling accident inside containment. Therefore, there is no need for the actuation signal to reposition the valves to the closed position. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

L05 (Category 1 - Relaxation of LCO Requirements) CTS 3.8.a specifies that the CTS 3.8.a.1 containment closure requirements are applicable during REFUELING OPERATIONS, which is defined in CTS Section 1.0 as the movement of reactor vessel internals that could affect the reactivity of the core within the containment when the vessel head is unbolted or removed. ITS 3.9.6 is applicable during movement of irradiated fuel assemblies within containment.

This changes the CTS by not requiring the containment closure requirements to be met when moving or handling control rods during MODE 6 operation.

The purpose of CTS 3.8.a.1 is to ensure that if a fuel handling accident occurs, the release of any subsequent fission products results in doses that are well within the guideline values specified in Regulatory Guide 1.183. A fuel handling accident, as analyzed in USAR Section 14.2.1, can only occur during handling of irradiated fuel assemblies; not moving or handling control rods (which are the only other reactor vessel internals currently in use at KPS that could affect reactivity of the core). Thus, moving or handling the control rods cannot result in a fuel handling accident. Any additional requirements, above those required to meet the assumptions of the fuel handling accident, are more appropriately controlled by plant procedures. Therefore, this change is acceptable and is designated a less restrictive change since the requirement to maintain containment closure capability during movement or handling of control rods in MODE 6 is being deleted from the CTS.

Kewaunee Power Station Page 4 of 4 Attachment 1, Volume 14, Rev. 1, Page 117 of 181

Attachment 1, Volume 14, Rev. 1, Page 118 of 181 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 118 of 181

Attachment 1, Volume 14, Rev. 1, Page 119 of 181 CTS Containment Penetrations 6

3.9.4 6

3.9 REFUELING OPERATIONS 6

3.9.4 6 Containment Penetrations 6

LCO 3.9.4 6 The containment penetrations shall be in the following status:

1

a. The equipment is hatch closed and held in place by [four] bolts, 2 3.8.a.1.a 3

b. One door in each air lock is [capable of being] closed, and 1

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere is either:

3.8.a.1.b

1. Closed by a manual or automatic isolation valve, blind flange, or equivalent or ; 3 3.8.1.a.b, 2. Capable of being closed by an OPERABLE Containment 3.8.a.8 Purge and Exhaust Isolation System.

4 Vent

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

7 Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

3.8.a APPLICABILITY: During movement of [recently] irradiated fuel assemblies within 5 containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.8.b A. One or more A.1 Suspend movement of Immediately 5

containment [recently] irradiated fuel penetrations not in assemblies within required status. containment.

WOG STS 3.9.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 119 of 181

Attachment 1, Volume 14, Rev. 1, Page 120 of 181 CTS Containment Penetrations 3.9.4 6 6

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 6

DOC M01 SR 3.9.4.1 Verify each required containment penetration is in 7 days 6 the required status.

6 3.8.a.8 SR 3.9.4.2 -------------------------------NOTE------------------------------ 6 Not required to be met for containment purge and 4

vent exhaust valve(s) in penetrations closed to comply with LCO 3.9.4.c.1. 6 6 vent Verify each required containment purge and exhaust [18] months 4 1 valve actuates to the isolation position on an actual or simulated actuation signal.

WOG STS 3.9.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 120 of 181

Attachment 1, Volume 14, Rev. 1, Page 121 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.6, CONTAINMENT PENETRATIONS

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. A typographical error within the ISTS has been corrected. The word "is" has been placed after the word "hatch".

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

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

5. The ISTS contains bracketed information and/or values that are generic to all Westinghouse vintage plants. In ITS 3.9.6 Applicability and Required Action A.1, the brackets and the term "recently" have been deleted since the term "recently" does not apply to KPS when referring to irradiated fuel assemblies.

6. ISTS 3.9.4 has been renumbered to ITS 3.9.6 since it was added back into the ITS as the result of an NRC RAI and KPS chose not to renumber multiple Specifications due to this later addition.

7. The ISTS 3.9.4 LCO Note allowance has not been adopted.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 121 of 181

Attachment 1, Volume 14, Rev. 1, Page 122 of 181 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 122 of 181

Attachment 1, Volume 14, Rev. 1, Page 123 of 181 All changes are 1 Containment Penetrations B 3.9.4 unless otherwise noted 5 6

B 3.9 REFUELING OPERATIONS B 3.9.4 Containment Penetrations 5 6

BASES BACKGROUND During movement of [recently] irradiated fuel assemblies within containment, a release of fission product radioactivity within containment will be restricted from escaping to the environment when the LCO requirements are met. In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1, "Containment." In MODE 6, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements are referred to as "containment closure" rather than "containment OPERABILITY." Containment closure means that all potential escape paths are closed or capable of being closed. Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of guidance 2 Regulatory Guide 1.183 (Ref. 1) 10 CFR 100. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. During movement of [recently]

irradiated fuel assemblies within containment, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.

During movement of [recently] irradiated fuel assemblies within containment, containment closure is required; therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain [capable of being] closed.

WOG STS B 3.9.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 123 of 181

Attachment 1, Volume 14, Rev. 1, Page 124 of 181 All changes are 1 Containment Penetrations B 3.9.4 unless otherwise noted 6 5 BASES BACKGROUND (continued)

Two systems can be used to purge or ventilate the The requirements for containment penetration closure ensure that a containment; the Containment Purge and release of fission product radioactivity within containment will be restricted Vent System and the Post to within regulatory limits. The Post LOCA LOCA Hydrogen Control Vent Hydrogen Control System.

36 The Containment Purge and Exhaust System includes two subsystems.

The normal subsystem includes a 42 inch purge penetration and a vent 42 inch exhaust penetration. The second subsystem, a minipurge 2 system, includes an 8 inch purge penetration and an 8 inch exhaust vent 2

penetration. During MODES 1, 2, 3, and 4, the two valves in each of the 2 normal purge and exhaust penetrations are secured in the closed vent position. The two valves in each of the two minipurge penetrations can INSERT 1 be opened intermittently, but are closed automatically by the Engineered Safety Features Actuation System (ESFAS). Neither of the subsystems are is subject to a Specification in MODE 5.

fresh, tempered air is provided In MODE 6, large air exchangers are necessary to conduct refueling 2 36 operations. The normal 42 inch purge system is used for this purpose, and all four valves are closed by the ESFAS in accordance with LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation."

[ The minipurge system remains operational in MODE 6, and all four valves are also closed by the ESFAS.

[or]

The minipurge system is not used in MODE 6. All four 8 inch valves are secured in the closed position. ]

or capable of being isolated The other containment penetrations that provide direct access from 6 containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent.

Equivalent isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during [recently] irradiated 2

fuel movements (Ref. 1).

WOG STS B 3.9.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 124 of 181

Attachment 1, Volume 14, Rev. 1, Page 125 of 181 B 3.9.6 2

INSERT 1 The Post LOCA Hydrogen Control subsystem contains two trains. The valves in Train A are normally closed. The valves in Train B are also normally closed but are periodically opened to control containment pressure within the required limits.

The Train B valves receive a signal to close via the Engineered Safety Features Actuation System and the Containment Purge and Vent Isolation System.

Insert Page B 3.9.4-2 Attachment 1, Volume 14, Rev. 1, Page 125 of 181

Attachment 1, Volume 14, Rev. 1, Page 126 of 181 Containment Penetrations All changes are 1 B 3.9.4 unless otherwise noted 5 6

BASES TSTF-471-A APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel assemblies SAFETY within containment, the most severe radiological consequences result ANALYSES from a fuel handling accident [involving handling recently irradiated fuel].

The fuel handling accident is a postulated event that involves damage to 2 irradiated fuel (Ref. 2). Fuel handling accidents, analyzed in Reference 3, 2 vertically onto a rigid surface or include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies. The requirements of 5 LCO 3.9.7, "Refueling Cavity Water Level," in conjunction with a minimum 7 decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to [irradiated fuel movement with containment closure capability or a minimum decay time of [x] days Regulatory Guide 1.183 without containment closure capability], ensures that the release of fission (Ref. 1) product radioactivity, subsequent to a fuel handling accident, results in 2

doses that are well within the guideline values specified in 10 CFR 100.

Standard Review Plan, Section 15.7.4, Rev. 1 (Ref. 3), defines "well within" 10 CFR 100 to be 25% or less of the 10 CFR 100 values. The 2

acceptance limits for offsite radiation exposure will be 25% of 10 CFR 100 values or the NRC staff approved licensing basis (e.g., a specified fraction of 10 CFR 100 limits).

Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO -----------------------------------REVIEWERS NOTE-----------------------------------

The allowance to have containment personnel air lock doors open and TSTF-471-A penetration flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated during fuel movement and CORE ALTERATIONS is based on (1) confirmatory dose calculations of a fuel handling accident as approved by the NRC staff which indicate acceptable radiological consequences and 3 (2) commitments from the licensee to implement acceptable administrative procedures that ensure in the event of a refueling accident (even though the containment fission product control function is not required to meet acceptable dose consequences) that the open air lock can and will be promptly closed following containment evacuation and that the open penetration(s) can and will be promptly closed. The time to close such penetrations or combination of penetrations shall be included in the confirmatory dose calculations.

This LCO limits the consequences of a fuel handling accident [involving 2 handling recently irradiated fuel] in containment by limiting the potential escape paths for fission product radioactivity released within containment.

The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge and exhaust penetrations [and the 2 vent containment personnel air locks]. For the OPERABLE WOG STS B 3.9.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 126 of 181

Attachment 1, Volume 14, Rev. 1, Page 127 of 181 Containment Penetrations B 3.9.4 6 5 BASES LCO (continued) vent Vent containment purge and exhaust penetrations, this LCO ensures that these penetrations are isolable by the Containment Purge and Exhaust Isolation System. The OPERABILITY requirements for this LCO ensure 2 that the automatic purge and exhaust valve closure times specified in the FSAR can be achieved and, therefore, meet the assumptions used in the safety analysis to ensure that releases through the valves are terminated, such that radiological doses are within the acceptance limit.

The LCO is modified by a Note allowing penetration flow paths with direct 8 access from the containment atmosphere to the outside atmosphere to be unisolated under administrative controls. Administrative controls ensure that 1) appropriate personnel are aware of the open status of the TSTF-471-A penetration flow path during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, and 2) specified individuals are designated and readily available to isolate the flow path in the event of a fuel handling accident.

may 4

within 30 minutes The containment personnel air lock doors many be open during movement of [recently] irradiated fuel in the containment provided that 1 one door is capable of being closed in the event of a fuel handling 2 INSERT 2 accident. Should a fuel handling accident occur inside containment, one 2 personnel air lock door will be closed following an evacuation of containment.

APPLICABILITY The containment penetration requirements are applicable during 1

movement of [recently] irradiated fuel assemblies within containment because this is when there is a potential for the limiting fuel handling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1. In MODES 5 and 6, when movement of irradiated fuel assemblies within containment is not being conducted, the potential for a fuel handling accident does not exist.

[Additionally, due to radioactive decay, a fuel handling accident involving handling recently irradiated fuel (i.e., fuel that has occupied part of a 1 critical reactor core within the previous [x] days) will result in doses that are well within the guideline values specified in 10 CFR 100 even without containment closure capability.] Therefore, under these conditions no requirements are placed on containment penetration status.

WOG STS B 3.9.4-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 127 of 181

Attachment 1, Volume 14, Rev. 1, Page 128 of 181 B 3.9.6 2 INSERT 2 When both personnel airlock doors are open during the movement of irradiated fuel in the containment, appropriate plant personnel shall be notified of this condition. A specified individual(s) is designated and available to close the airlock following a required evacuation of containment. Any obstruction(s) (e.g.,

cables and hoses) that can prevent closure of an open airlock shall be able to be removed in a timely manner (i.e., within the 30 minutes specified above).

Insert Page B 3.9.4-4 Attachment 1, Volume 14, Rev. 1, Page 128 of 181

Attachment 1, Volume 14, Rev. 1, Page 129 of 181 Containment Penetrations B 3.9.4 6 5 BASES APPLICABILITY (continued)

REVIEWERS NOTE-----------------------------------

The addition of the term "recently" associated with handling irradiated fuel in all of the containment function Technical Specification requirements is only applicable to those licensees who have demonstrated by analysis that after sufficient radioactive decay has occurred, off-site doses resulting from a fuel handling accident remain below the Standard Review Plan limits (well within 10 CFR 100).

Additionally, licensees adding the term "recently" must make the following commitment which is consistent with NUMARC 93-01, Revision 4, Section 11.3.6.5 "Safety Assessment for Removal of Equipment from Service During Shutdown Conditions," subheading "Containment -

Primary (PWR)/Secondary (BWR)."

"The following guidelines are included in the assessment of systems removed from service during movement irradiated fuel:

3

- During fuel handling/core alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the fuel decays away fairly rapidly. The basis of the Technical Specification OPERABILITY amendment is the reduction in doses due to such decay. The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay.

- A single normal or contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure.

The purpose of the "prompt methods" mentioned above are to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored."

WOG STS B 3.9.4-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 129 of 181

Attachment 1, Volume 14, Rev. 1, Page 130 of 181 Containment Penetrations B 3.9.4 6 5 BASES ACTIONS A.1 If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Vent Containment Purge and Exhaust Isolation System not capable of vent 2 automatic actuation when the purge and exhaust valves are open, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending movement of

[recently] irradiated fuel assemblies within containment. Performance of 1 these actions shall not preclude completion of movement of a component to a safe position.

6 5

SURVEILLANCE SR 3.9.4.1 required REQUIREMENTS is in the This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The Surveillance status 2

vent on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing. Also the Surveillance will demonstrate that each valve operator has motive power, which will ensure that each valve is capable of being closed by an OPERABLE automatic containment purge and exhaust isolation signal.

The Surveillance is performed every 7 days during movement of [recently] 1 irradiated fuel assemblies within containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations. A surveillance before the start of refueling operations will provide two or three surveillance verifications during the applicable period for this LCO. As such, this Surveillance ensures that a postulated fuel handling accident [involving handling 1 recently irradiated fuel] that releases fission product radioactivity within the containment will not result in a release of significant fission product radioactivity to the environment in excess of those recommended by Regulatory Guide 1.183 Standard Review Plan Section 15.7.4 (Reference 3). 2 1

6 5

SR 3.9.4.2 required vent This Surveillance demonstrates that each containment purge and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. The 18 month Frequency maintains 2 consistency with other similar ESFAS instrumentation and valve testing requirements. In LCO 3.3.6, the Containment Purge and Exhaust INSERT 3 WOG STS B 3.9.4-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 130 of 181

Attachment 1, Volume 14, Rev. 1, Page 131 of 181 B 3.9.6 0 -INSERT 3 2

LCO 3.3.6, "Containment Purge and Vent Isolation Instrumentation," provides additional Surveillance Requirements for the containment purge and vent valve actuation circuitry.

Insert Page B 3.9.4-6 Attachment 1, Volume 14, Rev. 1, Page 131 of 181

Attachment 1, Volume 14, Rev. 1, Page 132 of 181 Containment Penetrations B 3.9.4 6 5 BASES SURVEILLANCE REQUIREMENTS (continued)

Isolation instrumentation requires a CHANNEL CHECK every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a COT every 92 days to ensure the channel OPERABILITY during refueling operations. Every 18 months a CHANNEL CALIBRATION is 2 performed. The system actuation response time is demonstrated every 18 months, during refueling, on a STAGGERED TEST BASIS.

SR 3.6.3.5 demonstrates that the isolation time of each valve is in accordance with the Inservice Testing Program requirements. These Surveillances performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident [involving 1 handling recently irradiated fuel] to limit a release of fission product 2 radioactivity from the containment.

The SR is modified by a Note stating that this Surveillance is not required to be met for valves in isolated penetrations. The LCO provides the option to close penetrations in lieu of requiring automatic actuation capability.

REFERENCES 1. GPU Nuclear Safety Evaluation SE-0002000-001, Rev. 0, May 20, 1988.

14.2.1 2 2 2. FSAR, Section [15.4.5].

U Regulatory Guide 1.183, July 2000 1 3. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.

WOG STS B 3.9.4-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 132 of 181

Attachment 1, Volume 14, Rev. 1, Page 133 of 181 JUSTIFICATION FOR DEVIATIONS ITS 3.9.6 BASES, CONTAINMENT PENETRATIONS

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. 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.

4. Typographical error corrected.

5. ISTS 3.9.4 has been renumbered to ITS 3.9.6 since it was added back into the ITS as the result of an NRC RAI and KPS chose not to renumber multiple Specifications due to this later addition.

6. Changes made to be consistent with the actual Specification.

7. The correct ITS number has been provided.

8. Changes have been made to be consistent with changes to the Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 133 of 181

Attachment 1, Volume 14, Rev. 1, Page 134 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 134 of 181

Attachment 1, Volume 14, Rev. 1, Page 135 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.6, CONTAINMENT PENETRATIONS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 135 of 181

, Volume 14, Rev. 1, Page 136 of 181 ATTACHMENT 7 RELOCATED/DELETED CURRENT TECHNICAL SPECIFICATIONS , Volume 14, Rev. 1, Page 136 of 181

, Volume 14, Rev. 1, Page 137 of 181 CTS 3.8.a.2, AREA RADIATION MONITORING , Volume 14, Rev. 1, Page 137 of 181

, Volume 14, Rev. 1, Page 138 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 138 of 181

Attachment 1, Volume 14, Rev. 1, Page 139 of 181 CTS 3.8.a.2 3.8 REFUELING OPERATIONS APPLICABILITY See ITS Applies to operating limitations during REFUELING OPERATIONS. 3.9.1, 3.9.2, 3.9.3, OBJECTIVE 3.9.5, and 3.9.6 To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

See ITS SPECIFICATION 3.9.1, 3.9.2, 3.9.3,

a. During REFUELING OPERATIONS: 3.9.5, 3.9.6, and

1. Containment Closure CTS 3.8.a.3

a. The equipment hatch shall be closed and at least one door in each personnel air lock shall be capable of being closed (1) in 30 minutes or less. In addition, at See ITS least one door in each personnel air lock shall be closed when the reactor vessel 3.9.6 head or upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valve or an operable automatic isolation valve.

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage R01 pool shall be monitored continuously.

See CTS

3. The reactor will be subcritical for 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months prior to movement of its irradiated fuel 3.8.a.3 assemblies. Core subcritical neutron flux shall be continuously monitored by at least two neutron monitors, each with continuous visual indication in the control room and See ITS one with audible indication in the containment whenever core geometry is being 3.9.2 changed. When core geometry is not being changed at least one neutron flux monitor shall be in service.

4. At least one residual heat removal pump shall be OPERABLE. See ITS 3.9.3

5. When there is fuel in the reactor, a minimum boron concentration as specified in the COLR shall be maintained in the Reactor Coolant System during reactor vessel See ITS head removal or while loading and unloading fuel from the reactor. The required 3.9.1 boron concentration shall be verified by chemical analysis daily.

(1)

Administrative controls ensure that:

Appropriate personnel are aware that both personnel air lock doors are open,

A specified individual(s) is designated and available to close the air lock following a required See ITS evacuation of containment, and 3.9.6

Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed.

Amendment No. 165 TS 3.8-1 03/11/2003 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 139 of 181

Attachment 1, Volume 14, Rev. 1, Page 140 of 181 DISCUSSION OF CHANGES CTS 3.8.a.2, AREA RADIATION MONITORING ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.8.a.2 provides the requirements for continuously monitoring radiation levels in the fuel handling area, the containment, and the spent fuel pool storage pool. Since these radiation monitors do not initiate any automatic mitigation systems, they are not of prime importance in limiting the likelihood or severity of accident sequences that are commonly found to dominate risk. These monitors only provide indication of the area radiation. 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 3.8.a.2 does not meet the 10 CFR 50.36(c)(2)(ii) criteria for inclusion into the ITS.

10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. The area radiation monitoring 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 area radiation monitoring Specification does not satisfy criterion 1.

2. The area radiation monitoring 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 area radiation monitoring Specification does not satisfy criterion 2.

3. The area radiation monitoring 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 area radiation monitoring Specification does not satisfy criterion 3.

4. The area radiation monitoring 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 (Appendix A, page A-20) and summarized in Table 1 of WCAP-11618, the area radiation monitoring 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 Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 140 of 181

Attachment 1, Volume 14, Rev. 1, Page 141 of 181 DISCUSSION OF CHANGES CTS 3.8.a.2, AREA RADIATION MONITORING assessment. The area radiation monitoring Specification does not satisfy criterion 4.

Since 10 CFR 50.36(c)(2)(ii) criteria have not been met, the area radiation monitoring Specification 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 14, Rev. 1, Page 141 of 181

Attachment 1, Volume 14, Rev. 1, Page 142 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 142 of 181

Attachment 1, Volume 14, Rev. 1, Page 143 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.8.a.2, AREA RADIATION MONITORING There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 143 of 181

, Volume 14, Rev. 1, Page 144 of 181 CTS 3.8.a.3, DECAY TIME , Volume 14, Rev. 1, Page 144 of 181

, Volume 14, Rev. 1, Page 145 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 145 of 181

Attachment 1, Volume 14, Rev. 1, Page 146 of 181 CTS 3.8.a.3 3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS. See ITS 3.9.1, 3.9.2, OBJECTIVE 3.9.3, 3.9.5, and 3.9.6 To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION

a. During REFUELING OPERATIONS: LA01

1. Containment Closure

a. The equipment hatch shall be closed and at least one door in each personnel air lock shall be capable of being closed (1) in 30 minutes or less. In addition, at See ITS 3.9.6 least one door in each personnel air lock shall be closed when the reactor vessel head or upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valve or an operable automatic isolation valve.

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage See CTS 3.8.a.2 pool shall be monitored continuously.

3. The reactor will be subcritical for 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months prior to movement of its irradiated fuel LA01 assemblies. Core subcritical neutron flux shall be continuously monitored by at least See ITS two neutron monitors, each with continuous visual indication in the control room and 3.9.2 one with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least one neutron flux monitor shall be in service.

See ITS

4. At least one residual heat removal pump shall be OPERABLE. 3.9.3

5. When there is fuel in the reactor, a minimum boron concentration as specified in the COLR shall be maintained in the Reactor Coolant System during reactor vessel See ITS head removal or while loading and unloading fuel from the reactor. The required 3.9.1 boron concentration shall be verified by chemical analysis daily.

(1)

Administrative controls ensure that:

Appropriate personnel are aware that both personnel air lock doors are open, See ITS

A specified individual(s) is designated and available to close the air lock following a required 3.9.6 evacuation of containment, and

Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed.

Amendment No. 165 TS 3.8-1 03/11/2003 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 146 of 181

Attachment 1, Volume 14, Rev. 1, Page 147 of 181 DISCUSSION OF CHANGES CTS 3.8.a.3, DECAY TIME ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA01 (Type 4 - Removal of LCO, SR, or other TS Requirement to the TRM, USAR, ODCM, NFQAPD, CLRT Program, IST Program, ISI Program, or Setpoint Control Program) CTS 3.8.a.3 requires the reactor to be subcritical for 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months prior to movement of its irradiated fuel assemblies. The ITS does not include this requirement. This changes the CTS by moving the explicit decay time requirement from the Technical Specifications to the Technical Requirements Manual (TRM).

The removal of this detail from the Technical Specifications is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. The purpose of CTS 3.8.a.3 is to ensure sufficient time has elapsed to account for the decay heat load capacity of the spent fuel storage pool. This change is acceptable because the removed information will be adequately controlled in the TRM. Changes to the TRM are controlled by the provisions of 10 CFR 50.59, which ensures changes are properly evaluated.

This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 147 of 181

Attachment 1, Volume 14, Rev. 1, Page 148 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 148 of 181

Attachment 1, Volume 14, Rev. 1, Page 149 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.8.a.3, DECAY TIME There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 149 of 181

, Volume 14, Rev. 1, Page 150 of 181 CTS 3.8.a.6, COMMUNICATIONS , Volume 14, Rev. 1, Page 150 of 181

, Volume 14, Rev. 1, Page 151 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 151 of 181

Attachment 1, Volume 14, Rev. 1, Page 152 of 181 CTS 3.8.a.6

6. Direct communication between the control room and the operating floor of the containment shall be available whenever changes in core geometry are taking R01 place.

7. Deleted.

8. The containment ventilation and purge system, including the capability to initiate See ITS automatic containment ventilation isolation, shall be tested and verified to be 3.9.6 operable immediately prior to and daily during REFUELING OPERATIONS.

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be operating during fuel handling and when any load is carried over the pool if See CTS irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool 3.8.a.9 sweep system, including the charcoal adsorber, is not operating when required, fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

See ITS

10. The minimum water level above the vessel flange shall be maintained at 23 feet. 3.9.5

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes See CTS for this test must be equal to or greater than the maximum load to be assumed by 3.8.a.11 the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

12. A licensed senior reactor operator will be on-site and designated in charge of the See CTS 3.8.a.12 REFUELING OPERATIONS.

b. If any of the specified limiting conditions for REFUELING OPERATIONS are not met, See ITS refueling of the reactor shall cease. Work shall be initiated to correct the violated 3.9.1, conditions so that the specified limits are met, and no operations which may increase the 3.9.2, 3.9.3, reactivity of the core shall be performed. 3.9.5, and 3.9.6 Amendment No. 200 TS 3.8-2 11/20/2008 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 152 of 181

Attachment 1, Volume 14, Rev. 1, Page 153 of 181 DISCUSSION OF CHANGES CTS 3.8.a.6, COMMUNICATIONS ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.8.a.6 states that direct communication between the control room and the operating floor of the containment shall be available whenever changes in core geometry are taking place. This ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during changes in core geometry. The prompt notification of the control room of a fuel handling accident is not an assumption in the fuel handling accident analysis. While notification is necessary to ensure that the control room is isolated to meet the control room operator dose limits in 10 CFR 50.67, the fuel handling analysis does not take credit for direct communications between the refueling station and the control room. The ITS does not include this Specification. This changes the CTS by relocating this Specification to the Updated Safety Analysis Report (USAR).

This change is acceptable because CTS 3.8.a.6 does not meet the 10 CFR 50.36(c)(2)(ii) criteria for inclusion into the ITS.

10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. Communications 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 Communications Specification does not satisfy criterion 1.

2. Communications are not a process variable, design feature, or operating restriction that is an initial condition of a Design Basis Analysis (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The Communications Specification does not satisfy criterion 2.

3. Communications are not a structure, system, or component that is part of a 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 Communications Specification does not satisfy criterion 3.

4. Communications 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 (Table 1, page 11 and Appendix A, page A-67) of WCAP-11618, Communications was found to be a Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 153 of 181

Attachment 1, Volume 14, Rev. 1, Page 154 of 181 DISCUSSION OF CHANGES CTS 3.8.a.6, COMMUNICATIONS 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 Communications Specification does not satisfy criterion 4.

Since 10 CFR 50.36(c)(2)(ii) criteria have not been met, the Communications Specification may be relocated out of the Technical Specifications. The Communications Specification will be relocated to the USAR. Changes to the USAR 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 USAR.

REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Kewaunee Power Station Page 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 154 of 181

Attachment 1, Volume 14, Rev. 1, Page 155 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 155 of 181

Attachment 1, Volume 14, Rev. 1, Page 156 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.8.a.6, COMMUNICATIONS There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 156 of 181

Attachment 1, Volume 14, Rev. 1, Page 157 of 181 CTS 3.8.a.11, FUEL HANDLING AND MANIPULATOR CRANES Attachment 1, Volume 14, Rev. 1, Page 157 of 181

, Volume 14, Rev. 1, Page 158 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 158 of 181

Attachment 1, Volume 14, Rev. 1, Page 159 of 181 CTS 3.8.a.11

6. Direct communication between the control room and the operating floor of the containment shall be available whenever changes in core geometry are taking See CTS 3.8.a.6 place.

7. Deleted.

8. The containment ventilation and purge system, including the capability to initiate See ITS automatic containment ventilation isolation, shall be tested and verified to be 3.9.6 operable immediately prior to and daily during REFUELING OPERATIONS.

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be operating during fuel handling and when any load is carried over the pool if See CTS irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool 3.8.a.9 sweep system, including the charcoal adsorber, is not operating when required, fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

See ITS

10. The minimum water level above the vessel flange shall be maintained at 23 feet. 3.9.5

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes for this test must be equal to or greater than the maximum load to be assumed by R01 the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

12. A licensed senior reactor operator will be on-site and designated in charge of the See CTS 3.8.a.12 REFUELING OPERATIONS.

b. If any of the specified limiting conditions for REFUELING OPERATIONS are not met, See ITS 3.9.1, refueling of the reactor shall cease. Work shall be initiated to correct the violated 3.9.2, 3.9.3, conditions so that the specified limits are met, and no operations which may increase the 3.9.5, and reactivity of the core shall be performed. 3.9.6 Amendment No. 200 TS 3.8-2 11/20/2008 Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 159 of 181

Attachment 1, Volume 14, Rev. 1, Page 160 of 181 CTS 3.8.a.11 TABLE TS 4.1-3 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS R01 (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 Partial movement of all STANDBY See ITS 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 R01 REFUELING outage

6. Deleted See ITS 3.4.15

7. Deleted

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 R01 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 (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 2 of 2 Attachment 1, Volume 14, Rev. 1, Page 160 of 181

Attachment 1, Volume 14, Rev. 1, Page 161 of 181 DISCUSSION OF CHANGES CTS 3.8.a.11, FUEL HANDLING AND MANIPULATOR CRANES ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R01 CTS 3.8.a.11 states that a dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins.

The load assumed by the cranes for the test must be equal to or greater than the maximum load to be assumed by the cranes during the REFUELING OPERATIONS. CTS 3.8.a.11 also requires a thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling. This Specification ensures the lifting device on the Manipulator Crane has adequate capacity to lift the weight of a fuel assembly and a Rod Control Cluster Assembly, and that an automatic load limiting device is available to prevent damage to the fuel assembly during fuel movement. This Specification also ensures the auxiliary hoist on the Manipulator Crane has adequate capacity for latching and unlatching control rod drive shafts. CTS Table TS 4.1-3, Equipment Test 5 requires the Refueling System Interlocks to be tested each Refueling Outage.

This test ensures the other manipulator crane interlocks (e.g., ensuring only one component can be moved at one time) are OPERABLE. In addition, the testing is modified by Table TS 4.1-3 Note 1, which also requires the test to be performed following maintenance that could affect the operation of the refueling interlocks. The ITS does not include these requirements. This changes the CTS by relocating these requirements to the Technical Requirements Manual (TRM).

This change is acceptable because CTS 3.8.a.11 does not meet the 10 CFR 50.36(c)(2)(ii) criteria for inclusion into the ITS.

10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. The Fuel Handling and Manipulator Cranes 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 Fuel Handling and Manipulator Cranes do not satisfy criterion 1.

2. The Fuel Handling and Manipulator Cranes are not a process variable, design feature, or operating restriction that is an initial condition of a Design Basis Analysis (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The Fuel Handling and Manipulator Cranes do not satisfy criterion 2.

3. The Fuel Handling and Manipulator Cranes are not a structure, system, or component that is part of the primary success path and which functions to actuate or mitigate a DBA or transient that either assumes the failure of or Kewaunee Power Station Page 1 of 2 Attachment 1, Volume 14, Rev. 1, Page 161 of 181

Attachment 1, Volume 14, Rev. 1, Page 162 of 181 DISCUSSION OF CHANGES CTS 3.8.a.11, FUEL HANDLING AND MANIPULATOR CRANES presents a challenge to the integrity of a fission product barrier. The Fuel Handling and Manipulator Cranes do not satisfy criterion 3.

4. The Fuel Handling and Manipulator Cranes 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 (Table 1, page 11 and Appendix A, page A-68) of WCAP-11618, the Fuel Handling and Manipulator Cranes 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 Fuel Handling and Manipulator Cranes do not satisfy criterion 4.

Since 10 CFR 50.36(c)(2)(ii) criteria have not been met, the Fuel Handling and Manipulator Cranes Specification and Surveillances may be relocated out of the Technical Specifications. The Fuel Handling and Manipulator Cranes Specification and Surveillances 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 14, Rev. 1, Page 162 of 181

Attachment 1, Volume 14, Rev. 1, Page 163 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 163 of 181

Attachment 1, Volume 14, Rev. 1, Page 164 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.8.a.11, FUEL HANDLING AND MANIPULATOR CRANES There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 164 of 181

Attachment 1, Volume 14, Rev. 1, Page 165 of 181 CTS 3.8.a.12, REFUELING OPERATIONS - FACILITY STAFFING Attachment 1, Volume 14, Rev. 1, Page 165 of 181

, Volume 14, Rev. 1, Page 166 of 181 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) , Volume 14, Rev. 1, Page 166 of 181

Attachment 1, Volume 14, Rev. 1, Page 167 of 181 CTS 3.8.a.12

6. Direct communication between the control room and the operating floor of the containment shall be available whenever changes in core geometry are taking See CTS 3.8.a.6 place.

7. Deleted.

8. The containment ventilation and purge system, including the capability to initiate See ITS automatic containment ventilation isolation, shall be tested and verified to be 3.9.6 operable immediately prior to and daily during REFUELING OPERATIONS.

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be operating during fuel handling and when any load is carried over the pool if See CTS irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool 3.8.a.9 sweep system, including the charcoal adsorber, is not operating when required, fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 30°C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

See ITS

10. The minimum water level above the vessel flange shall be maintained at 23 feet. 3.9.5

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes See CTS for this test must be equal to or greater than the maximum load to be assumed by 3.8.a.11 the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

A01

12. A licensed senior reactor operator will be on-site and designated in charge of the REFUELING OPERATIONS.

b. If any of the specified limiting conditions for REFUELING OPERATIONS are not met, See ITS 3.9.1, refueling of the reactor shall cease. Work shall be initiated to correct the violated 3.9.2, conditions so that the specified limits are met, and no operations which may increase the 3.9.3, 3.9.5, and reactivity of the core shall be performed. 3.9.6 Amendment No. 200 TS 3.8-2 11/20/2008 Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 167 of 181

Attachment 1, Volume 14, Rev. 1, Page 168 of 181 DISCUSSION OF CHANGES CTS 3.8.a.12, REFUELING OPERATIONS - FACILITY STAFFING ADMINISTRATIVE CHANGES A01 CTS 3.8.a.12 states "A licensed senior reactor operator will be on-site and designated in charge of the REFUELING OPERATIONS." The ITS does not include this requirement. This changes the CTS by deleting this requirement.

10 CFR 50.54(m)(2)(iv) states "Each licensee shall have present, during alteration of the core of a nuclear power unit (including fuel loading or transfer), a person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person." This change is acceptable because the requirement deleted from Technical Specifications is already required by 10 CFR 50.54(m)(2)(iv). This change is designated as administrative because it does 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 14, Rev. 1, Page 168 of 181

Attachment 1, Volume 14, Rev. 1, Page 169 of 181 Specific No Significant Hazards Considerations (NSHCs)

Attachment 1, Volume 14, Rev. 1, Page 169 of 181

Attachment 1, Volume 14, Rev. 1, Page 170 of 181 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.8.a.12, REFUELING OPERATIONS - FACILITY STAFFING There are no specific NSHC discussions for this Specification.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 170 of 181

Attachment 1, Volume 14, Rev. 1, Page 171 of 181 ATTACHMENT 8 Improved Standard Technical Specifications (ISTS) not used in the Kewaunee Power Station ITS Attachment 1, Volume 14, Rev. 1, Page 171 of 181

Attachment 1, Volume 14, Rev. 1, Page 172 of 181 ISTS 3.9.2, Unborated Water Source Isolation Valves Attachment 1, Volume 14, Rev. 1, Page 172 of 181

Attachment 1, Volume 14, Rev. 1, Page 173 of 181 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 173 of 181

Attachment 1, Volume 14, Rev. 1, Page 174 of 181

[Unborated Water Source Isolation Valves]

3.9.2 3.9 REFUELING OPERATIONS 3.9.2 [ Unborated Water Source Isolation Valves ]

REVIEWER'S NOTE-------------------------------------------------

This Technical Specification is not required for units that have analyzed a boron dilution event in MODE 6. It is required for those units that have not analyzed a boron dilution event in MODE 6.

For units which have not analyzed a boron dilution event in MODE 6, the isolation of all unborated water sources is required to preclude this event from occurring.

LCO 3.9.2 Each valve used to isolate unborated water sources shall be secured in the closed position.

APPLICABILITY: MODE 6.

ACTIONS

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

Separate Condition entry is allowed for each unborated water source isolation valve. 1 CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ A.1 Suspend CORE Immediately Required Action A.3 ALTERATIONS.

must be completed whenever Condition A is AND entered.

A.2 Initiate actions to secure Immediately valve in closed position.

One or more valves not secured in closed AND position.

A.3 Perform SR 3.9.1.1. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months WOG STS 3.9.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 174 of 181

Attachment 1, Volume 14, Rev. 1, Page 175 of 181

[Unborated Water Source Isolation Valves]

3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Verify each valve that isolates unborated water 31 days sources is secured in the closed position.

1 WOG STS 3.9.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 175 of 181

Attachment 1, Volume 14, Rev. 1, Page 176 of 181 JUSTIFICATION FOR DEVIATIONS ISTS 3.9.2, UNBORATED WATER SOURCE ISOLATION VALVES

1. ISTS 3.9.2, "Unborated Water Source Isolation Valves", is not included in the Kewaunee Power Station (KPS) ITS. KPS has an analyzed boron dilution event in MODE 6 and, per the ISTS 3.9.2 Reviewer's Note, is not required to include ISTS 3.9.2 in the ITS. A description of this analysis is provided in USAR, Section 14.1.4.

Kewaunee Power Station Page 1 of 1 Attachment 1, Volume 14, Rev. 1, Page 176 of 181

Attachment 1, Volume 14, Rev. 1, Page 177 of 181 Improved Standard Technical Specifications (ISTS) Bases Markup and Bases Justification for Deviations (JFDs)

Attachment 1, Volume 14, Rev. 1, Page 177 of 181

Attachment 1, Volume 14, Rev. 1, Page 178 of 181

[Unborated Water Source Isolation Valves]

B 3.9.2 B 3.9 REFUELING OPERATIONS B 3.9.2 [ Unborated Water Source Isolation Valves ]

BASES BACKGROUND During MODE 6 operations, all isolation valves for reactor makeup water sources containing unborated water that are connected to the Reactor Coolant System (RCS) must be closed to prevent unplanned boron dilution of the reactor coolant. The isolation valves must be secured in the closed position.

The Chemical and Volume Control System is capable of supplying borated and unborated water to the RCS through various flow paths.

Since a positive reactivity addition made by reducing the boron concentration is inappropriate during MODE 6, isolation of all unborated water sources prevents an unplanned boron dilution.

APPLICABLE The possibility of an inadvertent boron dilution event (Ref. 1) occurring SAFETY during MODE 6 refueling operations is precluded by adherence to this ANALYSES LCO, which requires that potential dilution sources be isolated. Closing the required valves during refueling operations prevents the flow of unborated water to the filled portion of the RCS. The valves are used to 1 isolate unborated water sources. These valves have the potential to indirectly allow dilution of the RCS boron concentration in MODE 6. By isolating unborated water sources, a safety analysis for an uncontrolled boron dilution accident in accordance with the Standard Review Plan (Ref. 2) is not required for MODE 6.

The RCS boron concentration satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO requires that flow paths to the RCS from unborated water sources be isolated to prevent unplanned boron dilution during MODE 6 and thus avoid a reduction in SDM.

APPLICABILITY In MODE 6, this LCO is applicable to prevent an inadvertent boron dilution event by ensuring isolation of all sources of unborated water to the RCS.

For all other MODES, the boron dilution accident was analyzed and was found to be capable of being mitigated.

WOG STS B 3.9.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 178 of 181

Attachment 1, Volume 14, Rev. 1, Page 179 of 181

[Unborated Water Source Isolation Valves]

B 3.9.2 BASES ACTIONS The ACTIONS Table has been modified by a Note that allows separate Condition entry for each unborated water source isolation valve.

A.1 Continuation of CORE ALTERATIONS is contingent upon maintaining the unit in compliance with this LCO. With any valve used to isolate unborated water sources not secured in the closed position, all operations involving CORE ALTERATIONS must be suspended immediately. The Completion Time of "immediately" for performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position.

Condition A has been modified by a Note to require that Required Action A.3 be completed whenever Condition A is entered.

A.2 Preventing inadvertent dilution of the reactor coolant boron concentration is dependent on maintaining the unborated water isolation valves secured closed. Securing the valves in the closed position ensures that the valves 1 cannot be inadvertently opened. The Completion Time of "immediately" requires an operator to initiate actions to close an open valve and secure the isolation valve in the closed position immediately. Once actions are initiated, they must be continued until the valves are secured in the closed position.

A.3 Due to the potential of having diluted the boron concentration of the reactor coolant, SR 3.9.1.1 (verification of boron concentration) must be performed whenever Condition A is entered to demonstrate that the required boron concentration exists. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is sufficient to obtain and analyze a reactor coolant sample for boron concentration.

WOG STS B 3.9.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 179 of 181

Attachment 1, Volume 14, Rev. 1, Page 180 of 181

[Unborated Water Source Isolation Valves]

B 3.9.2 BASES SURVEILLANCE SR 3.9.2.1 REQUIREMENTS These valves are to be secured closed to isolate possible dilution paths.

The likelihood of a significant reduction in the boron concentration during MODE 6 operations is remote due to the large mass of borated water in the refueling cavity and the fact that all unborated water sources are isolated, precluding a dilution. The boron concentration is checked every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during MODE 6 under SR 3.9.1.1. This Surveillance 1

demonstrates that the valves are closed through a system walkdown.

The 31 day Frequency is based on engineering judgment and is considered reasonable in view of other administrative controls that will ensure that the valve opening is an unlikely possibility.

REFERENCES 1. FSAR, Section [15.2.4].

2. NUREG-0800, Section 15.4.6.

WOG STS B 3.9.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 14, Rev. 1, Page 180 of 181

Attachment 1, Volume 14, Rev. 1, Page 181 of 181 JUSTIFICATION FOR DEVIATIONS ISTS 3.9.2 BASES, UNBORATED WATER SOURCE ISOLATION VALVES

1. ISTS 3.9.2 Bases, "Unborated Water Source 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 14, Rev. 1, Page 181 of 181

v t e Letter Sequence Other
TAC:ME2139, Deletion of E BAR Definition and Revision to RCS Specific Activity Tech Spec (Approved, Closed)
Initiation Request, Request, Request, Request, Request Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance Questions 13 January 2010 11 March 2010 Answers Results Approval, Approval, Approval Other: ML093130368, ML102210288, ML102371088, ML102371098, ML102371109, ML102371140, ML102371178, ML102371203, ML102371205, ML102371210, ML102371257, ML102371260, ML102371266, ML102371269, ML102371282, ML102371283, ML102371290, ML102371292, ML102371294, ML102371297, ML102580704, ML102580706, ML102580709, ML102580710, ML102670754, ML102730398, ML102730399, ML103370691, ML110260316
MONTHYEAR2010-07-01 [Table View]

ML102371178

Text

Navigation menu

Search