Text

Transmittal of Technical Specifications Bases Changes and Technical Requirements Manual Changes
	ML081070072
Person / Time
Site:	Kewaunee
Issue date:	04/08/2008
From:	Wilson M; Dominion, Dominion Energy Kewaunee
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	08-0104
	Download: ML081070072 (57)
	v • d • e

Dominion Energy Kewaunee, Inc.

N490 Highway 42, Kewaunee, WI 54216-9511 Domlnion APR 0 8 2008 ATTN: Document Control Desk Serial No. 08-0104 U. S. Nuclear Regulatory Commission LIC/NW/RO Washington, DC 20555-0001 Docket No.: 50-305 License No.: DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION TECHNICAL SPECIFICATIONS BASES CHANGES AND TECHNICAL REQUIREMENTS MANUAL CHANGES Pursuant to Kewaunee Power Station (KPS) Technical Specification 6.21, "Technical Specifications (TS) Bases Control Program," Dominion Energy Kewaunee, Inc. (DEK) hereby submits changes to the TS Bases.

Additionally, DEK submits changes to the KPS Technical Requirements Manual (TRM).

10 CFR 50.71(e)(4) states the requirements for submittal of the KPS Updated Safety Analysis Report (USAR).

As the KPS TRM is considered a part of the USAR by reference, it is also required to be submithed tc, i-le-Nuclear Regulatory Commission (NRC).

The attachment provides copies of the KPS TS Bases and TRM pages reflecting the changes implemented since April 2007.

The changes to the TS Bases and TRM were made in accordance with the provisions of 10 CFR 50.59 and approved by the Facility Safety Review Committee.

If you have questions or require additional information, please feel free to contact Mr.

Gerald Riste at 920-388-8424.

Very truly yours, Mca J. Wilso Director Safety and Licensing

Attachment:

Kewaunee Power Station Technical Specifications Bases Changes and Technical Requirements Manual Changes Commitments made by this letter: NONE

Serial No. 08-0104 Page 2 of 2 cc:

Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Ms. M. H. Chernoff Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-H4A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station

Serial No. 08-0104 ATTACHMENT TECHNICAL SPECIFICATIONS BASES CHANGES AND TECHNICAL REQUIREMENTS MANUAL CHANGES KEWAUNEE POWER STATION TECHNICAL SPECIFICATIONS BASES CHANGES AND TECHNICAL REQUIREMENTS MANUAL CHANGES TS BASES PAGES:

TS B3.1-8 TS B3.6-4 TS B4.6-2 TS B4.6-3 TS B3.3-2 TS B2.1-2 TS B2.2-1 TRM PAGES:

TRM 3.16.1-1 REVISION 0 TRM 3.16.1-2 REVISION 0 TRM 3.7.2-1 REVISION 1 TRM 3.7.2-2 REVISION 1 TRM 3.7.2-3 REVISION 1 TRM 3.7.2-4 REVISION 1 TRM 3.3.1-1 REVISION 0 TRM 3.3.1-2 REVISION 0 TRM 3.7.1-1 REVISION 3 TRM 3.7.1-2 REVISION 3 TRM 3.7.1-3 REVISION 3 TRM 3.5.4-1 REVISION 0 TRM 3.5.4-2 REVISION 0 TRM 3.5.4-3 REVISION 0 TRM 3.5.4-4 REVISION 0 TRM 2.1 -COLR CYCLE 28 REVISION 1 TRM 3.0.9-1 REVISION 0 TRM 3.0.9-2 REVISION 0 TRM 3.0.9-3 REVISION 0 TRM 3.0.9-4 REVISION 0 TRM 3.0.9-5 REVISION 0 TRM 3.0.9-6 REVISION 0 TRM 3.5.1-1 REVISION 2 TRM 3.5.1-2 REVISION 2 KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Maximum Coolant Activity (TS 3.1.c)

The maximum dose that an individual may receive following an accident is specified in GDC 19 and 10 CFR 50.67. The limits on maximum coolant activity ensure that the calculated doses are held to the limits specified in GDC 19 and to a fraction of the 10 CFR 50.67 limits.

The Reactor Coolant Specific Activity is limited to < 1.0 MCi/gram DOSE EQUIVALENT 1-131 to ensure the dose does not exceed the GDC-19 and 10 CFR 50.67 guidelines. The applicable accidents identified in the USAR(15 ) are analyzed assuming an RCS activity of 1.0 pCi/gram DOSE EQUIVALENT 1-131 incorporating an accident initiated iodine spike when required. To ensure the conditions allowed are taken into account, the applicable accidents are also analyzed considering a pre-existing iodine spike of 20 pCi/gram DOSE EQUIVALENT 1-131. The results obtained from these analyses indicate that the control room and off-site doses are within the acceptance criteria of GDC-19 and a fraction of 10 CFR 50.67 limits.

91 PuC The Reactor Coolant Specific Activity is also limited to a gross activity of < 9 Again the E cc accidents under consideration are analyzed assuming a gross activity of 9] c The results E cc obtained from these analyses indicate the control room and off-site dose are within the acceptance criteria of GDC-1 9 and a small fraction of 10 CFR 50.67 limits.

The action of reducing average reactor coolant temperature to < 500OF prevents the release of activity should a steam generator tube rupture occur since the saturation pressure of the reactor coolant is below the lift pressure of the main steam safety valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

(15) USAR Section 14.0 Amendment No. 190 03/08/2007 TS B3.1-8

Ventilation Systems (TS 3.6.c)

Proper functioning of the Shield Building Ventilation System is essential to the performance of the Containment System. Therefore, except for reasonable periods of maintenance outage for one redundant train of equipment, the complete system should be in readiness whenever CONTAINMENT SYSTEM INTEGRITY is required. Proper functioning of the Auxiliary Building Special Ventilation System is similarly necessary to preclude possible unfiltered leakage through penetrations that enter the Special Ventilation Zone (Zone SV).

Both the Shield Building Ventilation System and the Auxiliary Building Special Ventilation System are designed to automatically start following a safety injection signal. Each of the two trains of both systems has 100% capacity. If one train of either system is found to be inoperable, there is not an immediate threat to the containment system performance and reactor operation may continue while repairs are being made. If both trains of either system are inoperable, the plant will be brought to a condition where the air purification system would not be required.

High efficiency particulate air (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential radioiodine release to the atmosphere. Bypass leakage for the charcoal adsorbers and particulate removal efficiency for HEPA filters are determined by halogenated hydrocarbon and DOP respectively. The laboratory carbon sample test results indicate a radioactive methyl iodine removal efficiency under test conditions which are more severe than accident conditions.

Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers. The performance criteria for the safeguard ventilation fans are stated in Section 5.5 and 9.6 of the USAR. If the performances are as specified, the calculated doses would be less than the guidelines stated in 10 CFR Part 100 for the accidents analyzed.

In-place testing procedures will be established utilizing applicable sections of ANSI N510 - 1975 standard as a procedural guideline only.

Accident analysis assumes a charcoal adsorber efficiency of 95%.(2) To ensure the charcoal adsorbers maintain that efficiency throughout the operating cycle, a safety factor of 2 is used.

Therefore, if accident analysis assumes a charcoal adsorber efficiency of 95%, this equates to a methyl iodide penetration of 5%. If a safety factor of 2 is assumed, the methyl iodide penetration is reduced to 2.5%. Thus, the acceptance criteria of 97.5% efficient will be used for the charcoal adsorbers.

(2) USAR TABLE 14.3-8, "Major Assumptions for Design Basis LOCA Analysis" Amendment No. 190 TS B3.6-4 03/08/2007

REFUELING Interval Diesel Generator Inspection, TS 4.6.a.3 Inspections are performed at REFUELING outage intervals in order to maintain the diesel generators in accordance with the manufacturers' recommendations. The inspection procedure is periodically updated to reflect experience gained from past inspections and new information as it is available from the manufacturer.

18-Month Load Rejection Test TS 4.6.a.4 The load rejection test demonstrates the capability of rejecting the maximum rated load without overspeeding or attaining voltages which would cause the diesel generator to trip, mechanical damage, or harmful overstresses.

Operating Cycle Short-Term Load Test TS 4.6.a.5 Loading the diesel generators to their short-term rating will demonstrate their capability to provide a continuous source of emergency AC power during a load perturbation of up to 110% of the diesel generator's continuous rating.

IEEE 387-1977 paragraph 3.7.2, defines a diesel generators short time rating. Paragraph 6.4.3 defines the rated load test for diesel generators, item 2 states to load the diesel generator to the short time rating for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Paragraph 6.6.2 describes the operational testing to be performed for the diesel generators. Although the rated load test is not listed in paragraph 6.6.2, item 2 of paragraph 6.4.3 has been determined to be necessary to be performed on the emergency diesel generators.

NRC Regulatory Guide 1.9, Revision 2, Regulatory Position 14, describes the method in which this test should be performed. This test follows Position 14 except that instead of the continuous rating load being applied for 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months the KPS emergency diesel generators shall be loaded to 2700 kW for 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months . Loading the emergency diesel generators to 2700 kW is acceptable because it will bound the post-accident emergency diesel generator loads without increasing the frequency of the 18-month diesel inspection surveillance. The diesel generator starts for this Surveillance can be performed either from standby or hot conditions. The provisions for prelubricating, warmup, and for gradual loading are applicable to this surveillance requirement.

The "once per operating cycle" frequency is consistent with the recommendations of IEEE Std. 387-1977, paragraph 6.6.2, and takes into consideration unit conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle lengths.

Three notes modify this Surveillance. Note 1 states that momentary transients due to changing busloads do not invalidate this test. Similarly, momentary power factor transients above the power factor operation will not invalidate the test. The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in the OPERATING or HOT STANDBY MODE is further amplified to allow the Surveillance to be performed for reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is Amendment No. 191 TS B4.6-2 05/01/2007

maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoid 'ed risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when the Surveillance is performed in the OPERATING or HOT STANDBY MODE. Risk insights or deterministic methods may be used for this assessment. Credit may be taken for unplanned events that satisfy this surveillance requirement.

Note 3 ensures that the DG is tested under load conditions that are as close to design basis conditions as possible. When synchronized with offsite power, testing should be performed at a power factor of :5 0.89. This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions. Under certain conditions, however, Note 3 allows the Surveillance to be conducted as a power factor other than 5 0.89. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to :5 0.89 results in voltages on the emergency busses that are too high. Under these conditions, the power factor should be maintained as close as practicable to 0.89 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the DG excitation levels needed to obtain a power factor of < 0.89 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the diesel generator. In such cases, the power factor shall be maintained close as practicable to 0.89 without exceeding the diesel generator excitation limits. When conditions exist where the testing is performed at a power factor greater than 0.89, the circumstances surrounding the conditions need be documented.

Station Batteries-TS 4.6.b Station batteries will deteriorate with time, but precipitous failure is extremely unlikely.

The surveillance specified is that which has been demonstrated over the years to provide indication of a cell becoming unserviceable long before it fails.

If a battery cell has' deteriorated, or if a connection is loose, the voltage under load will drop excessively, indicating need for replacement or maintenance.

Amendment No. 191 TS 84.6-3 05/01/2007

When the inoperable component is part of the Residual Heat Removal (RHR), Component Cooling Water (CCW) or Service Water (SW) Systems, the average Reactor Coolant System temperature (Tavg) will be maintained below 350°F through an alternate heat removal method.

The various alternate heat removal methods include the redundant RHR train and the steam generators.

Assuming the reactor has been OPERATING at full-rated power for at least 100 days, the magnitude of the decay heat decreases as follows after initiating HOT SHUTDOWN.

Time After Shutdown Decay Heat, % of Rated Power 1 minute 4.5 30 minutes 2.0 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1.62 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 0.96 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 0.62 Thus the requirement for core cooling in case of a postulated loss-of-coolant accident while in the HOT SHUTDOWN condition is significantly reduced below the requirements for a postulated loss-of-coolant accident during power operation. Putting the reactor in the HOT SHUTDOWN condition significantly reduces the potential consequences of a loss-of-coolant accident, and also allows more free access to some of the engineered safety features in order to effect repairs. Failure to complete repairs after placing the reactor in the HOT SHUTDOWN condition may be indicative of need for major maintenance, and in such cases the reactor should therefore be placed in the COLD SHUTDOWN condition.

TS 3.3.a.2.B provides a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time to restore an accumulator that is inoperable for a reason other than boron concentration.

The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed to restore an inoperable accumulator to operable status is justified in WCAP-1 5049, Revision 1. (2)

TS 3.3.b.5 allows an SI train to be considered operable while being used for accumulator fill during power operation. Analysis has shown that SI pump runout would not occur during an accumulator fill concurrent with a design basis LOCA. With both trains of SI and both EDGs operable, the SI system will meet accident analysis.

(2) WCAP-15049-A, Rev. 1, "Risk-Informed Evaluation of an Extension to Accumulator Completion Times," April 1999.

TS B3.3-2 07/25/2007

These limiting hot channel factors are higher than those calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. The control rod insertion limits are given in TS 3.10.d. Slightly higher hot channel factors could occur at lower power levels because additional control rods are in the core. However, the control rod insertion limits as specified in the COLR ensure that the increase in peaking factor is more than offset by the decrease in power level.

The Reactor Control and PROTECTION SYSTEM is designed to prevent any anticipated combination of transient conditions that would result in a DNBR less than the DNBR limit.

Two departure from nucleate boiling ratio (DNBR) correlations are used in the generation and validation of the safety limit curves:

the WRB-1 DNBR correlation and the high thermal performance (HTP) DNBR correlation. The WRB-1 correlation applies to the Westinghouse 422 V+ fuel.

The HTP correlation applies to FRA-ANP fuel with HTP spacers.

The DNBR correlations have been qualified and approved for application to Kewaunee.

The DNB correlation limits are 1.14 for the HTP DNBR correlation, and 1.17 for the WRB-1 DNBR correlation.

Safety Limit (SL) Violations The following SL violation responses are applicable to the reactor core SLs. If TS 2.1.a, 2.1.b, or 2.1.c is violated, the requirement to go to HOT SHUTDOWN places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

Amendment No. 193 TS B2.1-2 10/31/2007

BASIS - Safety Limit - Reactor Coolant System Prpssgure (TS 2-2)

The Reactor Coolant System(l) serves as a barrier preventing radionuclides contained in the reactor coolant from reaching the atmosphere. In the event of a fuel cladding failure, the Reactor Coolant System is the primary barrier against the release of fission products. By establishing a system pressure limit, the continued integrity of the Reactor Coolant System is ensured.

The maximum transient pressure allowable in the reactor pressure vessel under the ASME Code,Section III, is 110% of design pressure.

The maximum transient pressure allowable in the Reactor Coolant System piping, valves and fittings under USASI B.31.1.0 is 120% of design pressure. Thus, the SAFETY LIMIT of 2735 psig (110% of desigrh pressure, 2485 psig) has been established. (2)

The settings of the power-operated relief valves, the reactor high pressure trip and the safety valves have been established to prevent exceeding the SAFETY LIMIT of 2735 psig for all transients except the hypothetical RCCA Ejection accident, for which the faulted condition stress limit acceptance criterion of 3105 psig (3120 psia) is applied.

The initial hydrostatic test was conducted at 3107 psig to ensure the integrity of the Reactor Coolant System.

Safety Limit (SL) Violation If the RCS pressure SL is violated when the reactor is in the OPERATING or HOT STANDBY MODE, the requirement is to restore compliance and be in HOT SHUTDOWN within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 67, "Accident Source Term," limits. The allowable Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized. If the RCS pressure SL is exceeded in INTERMEDIATE SHUTDOWN, COLD SHUTDOWN or REFUELING MODES, RCS pressure must be restored to within the SL value within 5 minutes.

Exceeding the RCS pressure SL in INTERMEDIATE SHUTDOWN, COLD SHUTDOWN or REFUELING MODES, is more severe than exceeding this SL in OPERATING or HOT STANDBY MODES, since the reactor vessel temperature may be lower and-the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

(1) USAR Section 4 (2) USAR Section 4.3 Amendment No. 193 TS B2.2-1 10/31/2007

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.16.1 Revision 0 June 27, 2007 3.16.1 EXPLOSIVE GAS MONITORING SYSTEM APPLICABILITY Whenever the in service Waste Gas Decay Tank (WGDT) concentration has the potential to exceed 4% by volume.

hydrogen OBJECTIVE The Waste Gas System is not designed to withstand an explosion. Therefore, conditions are maintained to prevent an explosive gas mixture from existing.

TECHNICAL REQUIREMENTS Administrative Limiting Conditions for Operation (ALCOs) a.

b.

The waste gas analyzer (WGA) shall be Functional.

If the WGA is made or found non-functional, addition of waste gas to the waste gas holdup system may continue provided:

1. Grab samples are taken from the in service WGDT and analyzed once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during normal power operation, or

2. Grab samples are taken from the in service WGDT and analyzed once every four hours when the primary system is being degassed (other than normal gas stripping of the letdown flow).

c. If the concentration of the oxygen in the in service WGDT is greater than 4%

by volume, then

1. Immediately suspend additions of waste gas to the affected WGDT, and

2. Initiate action to reduce the oxygen concentration in the affected WGDT to less than 4% by volume.

Administrative Surveillance Requirement (ASRs)

CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS Waste Gas Analyzer Not Applicable Quarterly Monthly Test consists of an analysis of a known Qas standard 3.16.1-1

KEWAUNEE POWER STATION TRM 3.16.1 TECHNICAL REQUIREMENTS MANUAL Revision 0 June 27, 2007 BASES The Explosive Gas Monitoring System utilizes an inline Waste Gas Analyzer (WGA) to monitor the in service Waste Gas Decay Tank (WGDT) on a continuous basis. The WGA can be temporarily aligned to monitor various points in the waste gas holdup system provided it is subsequently realigned to the normal configuration. Grab sample analysis of the in service WGDT can be accomplished by obtaining a sample locally from the in service gas decay tank or from a sample point located on the WGA. Grab samples are analyzed with chemistry laboratory analytical equipment. If inline or grab sample analysis indicates an explosive mixture, actions will be taken to reduce the oxygen concentration as soon as possible.

The WGA provides a method for monitoring the concentrations of potentially explosive gas mixtures in the waste gas holdup system. An explosive gas mixture consists of a hydrogen gas concentration above the lower flammability limit of 4% AND an oxygen gas concentration above 4%.

The WGA has alarm capability that will alert operations personnel of oxygen concentrations approaching an explosive mixture and is set to 2% oxygen by volume.

The 2% alarm setpoint on the WGA is based on the 4% oxygen concentration required for flammability. This allows for a 100% safety margin in the setpoint. The WGA alarms in the control room and is locally monitored at least daily. Laboratory analysis of a grab sample is performed periodically to confirm instrument accuracy.

This process ensures that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammable limits for hydrogen and oxygen. This will minimize the probability of a WGDT rupture, thus, minimizing the probability of an accidental radioactive gas release.

Increased compensatory sampling is required when the WGA is not functional during primary system degassing operations.. Primary system degassing refers to the intentional removal of hydrogen from the reactor coolant system by either mechanical or chemical means and displacement with an alternate gas (e.g. nitrogen).

References

1. Comtrak Commitment Number 85-052, RETS-Explosive Gas Mixtures (DCR 1638).

2. Letter from Carl W. Giesler (WPSC) to Harold Denton (NRC), "Proposed Amendment No. 66 to the KNPP Technical Specifications," dated March 29, 1985.

3. Comtrak Commitment Number 96-122, Item D, Analyze for Explosive Gas Mixtures with Waste Gas Holdup System.

4. Letter from Morton B. Fairtile (NRC) to D.C. Hintz (WPSC), "Amendment 64," dated July 29, 1985.

3.16.1-2

KEWAUNEE POWER STATION TRM 3.7.2 TECHNICAL REQUIREMENTS MANUAL Revision 1 July 18, 2007 3.7.2 COMMON CAUSE TESTING OF EMERGENCY DIESEL GENERATORS APPLICABILITY Whenever the plant is in the HOT STANDBY or OPERATING Modes and one emergency diesel generator is made or determined to be inoperable.

OBJECTIVE To define the requirements necessary to ensure OPERABILITY of the other emergency diesel generator as required by TS 3.7.b.2.

TECHNICAL REQUIREMENTS Administrative Limiting Conditions for Operation (ALCOs)

Note 1:

All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.

Note 2:

A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this TRM as recommended by the manufacturer.

Note 3:

Performance of the load test required by ALCO 3.7.2.a.3 renders both DGs inoperable. This requires entry into TS LCO 3.0.c, "Standard Shutdown Sequence".

a.

When one emergency diesel generator is made or found inoperable:

1. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, verify correct breaker alignment and indicated power availability for each required offsite circuit, and

2. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , verify the requirement of TS 3.7.c is met, and

3. Daily, in accordance with TS 3.7.b.2, perform the surveillance requirement specified in TS 4.6.a.1.

Administrative Surveillance Requirement (ASRs) 3.7.2-1

KEWAUNEE POWER STATION TRM 3.7.2 TECHNICAL REQUIREMENTS MANUAL Revision 1 July 18, 2007 BASES

Background

Technical Specification (TS) 3.7.b.2 states, in part, that one diesel generator (DG) may be inoperable for a period not exceeding 7 days provided the other DG is tested daily to ensure OPERABILITY. However, the TS does not prescribe the specific daily testing required. The purpose of this TRM section is to prescribe the specific DG tests for satisfying the DG testing requirement in TS 3.7.b.2.

TS 4.0 requires that surveillance requirements be met during the MODES specified for individual LCOs. The purpose of TS 4.0 is to ensure that surveillances are performed to verify the OPERABILITY of systems and components. TS 4.6 provides the surveillance requirements to verify that DGs are OPERABLE.

The primary purpose for the requirement for daily testing is to determine that the OPERABLE DG has not become inoperable due to common cause failure. Performing TS 4.6.a.1 satisfies the testing requirement of TS 3.7.b.2 for verifying that the DG is OPERABLE.

Specification Verifying correct breaker alignment and indicated power availability ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that distribution buses and loads are connected to their preferred power source, and that appropriate independence of offsite circuits is maintained. If an offsite circuit is not available, it is inoperable. Upon offsite circuit inoperability, the associated Technical Specification must be reviewed and entered as applicable.

Verifying the requirements of TS 3.7.c is met is intended to provide assurance that a loss of offsite power, during the period that a DG is inoperable, does not result in a complete loss of safety function of critical systems. These safety functions are designed with redundant safety related trains. Although the requirement of TS 3.7.c is applicable at all times, TRM 3.7.2.a.2 and the associated completion time provides assurance that TS 3.7.c is reviewed.

Performing the surveillance requirement specified in TS 4.6.a.1 satisfies the requirement of TS 3.7.b.2 for ensuring DG OPERABILITY. The OPERABLE emergency diesel generator is rendered inoperable because it will not be able to perform its safety function in this condition.

For purposes of this TRM section, the OPERABLE DG refers to the DG that is not the subject of the first inoperability condition per TS 3.7.b.2 (i.e., the OPERABLE DG is the "other diesel generator" referenced in TS 3.7.b.2). Since Kewaunee Technical 3.7.2-2

KEWAUNEE POWER STATION TRM 3.7.2 TECHNICAL REQUIREMENTS MANUAL Revision 1 July 18, 2007 Specifications do not contain a specific allowance for both emergency diesel generators to be inoperable, the requirements of TS 3.0.c, Standard Shutdown Sequence, would apply due to TS 3.7 not being met while one emergency, diesel generator is inoperable and the second is rendered inoperable for testing. However, as stated in NRC Memorandum dated December 11, 1992, "Use of Shutdown Times for Corrective Maintenance (TIA 92-08)" (Reference 9), if action can be accomplished so that an unnecessary plant transient can be avoided, such a decision (to delay a shutdown action) is permitted as long as the shutdown times specified by the TS are observed.

Therefore, inoperability of both EDGs during the one hour load test would normally not require reducing reactor power to satisfy the requirements of TS 3.0.c, provided that the mode change times of 3.0.c are met. Orderly completion of the load test and restoration of the other EDG.to OPERABLE status satisfies the requirement of TS 3.0.c for initiating action within one hour to meet the specification.

This condition is not reportable per 10 CER 50.72 or 50.73 because it meets the guidance in NUREG-1 022 (Reference 10) of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS. NUREG 1022 states:

The following types of events or conditions generally are not reportable under these criteria:

removal of a system or part of a system from service as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that could have prevented the system from performing its function)

Therefore, this condition is not reportable unless:

1) the redundant EDG fails its surveillance test or

2) a TS required unit shutdown is initiated.

In the event the inoperable emergency diesel generator is restored to OPERABLE status before manually starting the redundant OPERABLE emergency diesel generator and the inoperable condition was caused by a diesel generator component, the plant corrective action program will continue to evaluate the common cause possibility. This continued evaluation follows the requirements of the corrective action program.

According to Generic Letter 84-15 (Reference 1, Appendix A, page 2), 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is reasonable to confirm that the OPERABLE emergency diesel generator is not affected by the same problem as the inoperable emergency diesel generator. This is demonstrated by testing the redundant emergency diesel generator once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the inoperable emergency diesel generator is returned to an operable status.

In order to reduce stress and wear on diesel engines, the KPS emergency diesel generator manufacturer recommended a modified start in which the starting speed of DGs is limited, warmup is limited to this lower speed, and the DGs are gradually accelerated to synchronous speed (References 2 and 3).

3.7.2-3

KEWAUNEE POWER STATION TRM 3.7.2 TECHNICAL REQUIREMENTS MANUAL Revision 1 July 18, 2007 References

1. Generic Letter 84-15, "Proposed Staff Actions To Improve And Maintain Diesel Generator Reliability"

2. Engineering Support Request (ESR) ESR 90-170

3. DCR 2571, "Evaluate and Correct Diesel Generator Governor Switch Settings, Add Slow Start," Design Description

4. Letter from R.C. Knop (NRC Rill) to K.H. Evers (WPSC), dated November 5, 1990, Inspection Report 50-305/90017 (DRP) (K-90-230)

5. Letter from R.C. Knop (NRC Rill) to C.A. Schrock (WPSC), dated December 13, 1991, Inspection Report 50-305/91021 (DRP) (K-91-258)

6. Letter from R.C. Knop (NRC Rill) to C.A. Schrock (WPSC), dated January 31, 1992, Inspection Report 50-305/91024 (DRP) (K-92-016)

7. Letter from D.C. Hintz (WPSC) to D.G. Eisenhut (NRC), "Diesel Generator Reliability (Generic Letter 84-15)," dated October 22, 1984 (NRC-84-172)

8. Calculation C-10915, "Safeguard Diesel Generator Loading Adjustments for Operation at Frequencies Other Than 60 Hertz," Revision 4, and Addendums A and B

9. Memorandum from F. J. Hebdon (NRC) to C.I. Grimes (NRC), "Use of Shutdown Times for Corrective Maintenance (TIA 92-08)", dated December 11, 1992

10. NUREG-1 022, Revision 2, "Event Reporting Guidelines 10 CFR 50.72 and 50.73", dated October 2000 3.7.2-4

KEWAUNEE POWER STATION TRM 3.3.1 TECHNICAL REQUIREMENTS MANUAL Revision 0 August 31, 2007 3.3.1 SERVICE WATER SYSTEM APPLICABILITY During all MODES.

OBJECTIVE To provide additional information during an abnormal condition in the service water system that renders a service water train inoperable.

TECHNICAL REQUIREMENTS Administrative Limiting Conditions for Operation (ALCOs)

a. When a train of service water is declared inoperable:

1. Apply TS 3.3.e for an inoperable service water system

2. Apply TS 3.7.b.2 for an inoperable emergency diesel generator

3. Apply TS 3.3.d for an inoperable component cooling train

4. Apply TS 3.3.b for an inoperable safety injection train

5. Apply TS 3.3.c for an inoperable containment fan coil units train

6. Apply TS 3.3.b for an inoperable residual heat removal train

7. Apply TS 3.4.b for an inoperable auxiliary feedwater train Administrative Surveillance Requirement (ASRs)

CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS Not Applicable Not Not Not Applicable Applicable Applicable 3.3.1-1

KEWAUNEE POWER STATION TRM 3.3.1 TECHNICAL REQUIREMENTS MANUAL Revision 0 August 31, 2007 BASES ALCO 3.3.1.a Kewaunee Power Station Technical Specification (TS) 1.0.e, "Operable - Operability,"

states that a system or component is OPERABLE or has OPERABILITY when it is capable of performing its intended function within the required range. The system or component shall be considered to have this capability when: (1) it satisfies the LIMITING CONDITIONS FOR OPERATION defined in TS 3.0, and (2) it has been tested periodically in accordance with TS 4.0 and has met its performance requirements. Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that is required for the system or component to perform its intended function is also capable of performing' their related support functions.

Because service water provides a support function for various TS required systems or components, when a service water train is inoperable the supported system or component is also inoperable. The supported system inoperability is because it has not been demonstrated that the support function for the other TS related supported systems or components (e.g., emergency diesel generators) can be performed with the service water system inoperable.

A review of the supported system and components allowed outage times was performed. This review concluded that the allowed outage time of the service water system was equal to or less than the allowed outage times of the supported systems rendered inoperable by an inoperable service water train.

References

1. TS 1.0.e, "Operable-Operability"

2. PORC Meeting 97-021 Minutes

3. KPS License Amendment 63 3.3.1-2

KEWAUNEE POWER STATION TRM 3.7.1 TECHNICAL REQUIREMENTS MANUAL Revision 3 August 31, 2007 3.7.1 TECHNICAL SUPPORT CENTER (TSC) I STATION BLACKOUT (SBO)

DIESEL GENERATOR (DG)

APPLICABILITY At ALL Times.

OBJECTIVE To ensure the provision of a reliable emergency power source for the TSC and for a postulated SBO condition.

TECHNICAL REQUIREMENTS Administrative Limiting Condition for Operation (ALCO)

a. The TSC / SBO DG shall be functional, except as specified in section TRM 3.7.1.b below.

b. If the TSC / SBO DG is made or found not to be functional, the following actions shall be initiated immediately:

1. Notify Emergency Preparedness (EP).

2. Initiate the following compensatory / mitigating measures:

a) The work on TSC Diesel will be performed as a priority and the TSC diesel will be retuned to service as soon as possible.

b) In the Operating, Hot Standby, Hot Shutdown, or Intermediate Shutdown Modes, BOTH Emergency Diesel Generators (EDGs) should be protected (except for unplanned necessary emergent circumstances, work and testing related to the EDGs should not be conducted).

c) In the Cold Shutdown, or Refueling Modes, or when Defueled, at least one EDG should be protected (except for unplanned necessary emergent circumstances, work and testing related to the EDG should not be conducted).

d) Except for unplanned necessary emergent circumstances, work and testing relative to the following high voltage electrical distribution equipment should not be undertaken:

3.7.1-1'

KEWAUNEE POWER STATION TRM 3.7.1 TECHNICAL REQUIREMENTS MANUAL Revision 3 August 31, 2007 Tertiary Auxiliary Transformer and the Reserve Auxiliary Transformer Switchyard equipment 0 Transmission and substation equipment / components in the relay house (e.g., protective relaying, fault detection, etc.)

e) Except for planned work, or unplanned necessary emergent circumstances, work and testing relative to Bus 46 should not be undertaken.

f) Direct EP to ensure Dose Assessment can be performed from the Control Room.

c. A change in plant operational MODES or conditions is acceptable with the TSC / SBO DG non-functional.

Administrative Surveillance Requirement (ASR)

a. The TSC / SBO DG shall be tested for functionality once every 31 days.

3.7.1-2

KEWAUNEE POWER STATION TRM 3.7.1 TECHNICAL REQUIREMENTS MANUAL Revision 3 August 31, 2007 BASES The purpose of the TSC / SBO DG is to provide emergency AC power to Bus 1-46 through breaker 14604, and for station blackout to act as the alternate AC (AAC) power supply as specified in NUMARC 87-00.

The TSC / SBO DG starts automatically on loss of voltage to Bus 1-46 and automatically connects to the Bus after attaining voltage and frequency provided that Source Breaker 14601 has tripped.

The TSC / SBO DG design requirement is to provide emergency power for the TSC Building, security lighting system, and other non-ESF plant systems which are required to operate upon loss of the Main Generator and off-site electrical sources.

The TSC / SBO DG does not typically supply power to QA Type 1 equipment.

During an SBO event, the TSC / SBO DG can be manually tied between Bus 1-46 and Bus 1-52 to become the AAC.

From an EP perspective, including probabilistic risk assessment insight, with the TSC / SBO DG out-of-service the intent of the compensatory / mitigation measures are to:

Have adequate measures in-place to minimize the probability of creation of a loss-of-power event.

Have adequate measures in-place to deal with an event requiring activation and manning of the TSC, preceded by or followed by a loss of offsite power, rendering the TSC non-functional.

During the time the TSC Diesel is non-functional, if the emergency response organization is activated and there is a loss of offsite power to the TSC, it may be necessary to relocate the TSC due to loss of the TSC Ventilation and loss of the ability to acquire data. If it becomes necessary to relocate the TSC, the following functions/activities are covered by the Emergency Plan Implementing Procedures:

" Communications / Notifications will be performed from the Control Room until the emergency operations facility (EOF) is activated.

" Classification will be maintained in the Control Room.

Protective Action Recommendations will be maintained in the Control Room until the EOF is activated.

In addition, if it becomes necessary to relocate the TSC, the Emergency Preparedness group will place the equipment necessary to perform Dose Assessment in the control room.

The once-per-31-day surveillance test requirement for the TSC / SBO DG is deemed adequate to demonstrate functionality of the TSC / SBO DG and its auxiliaries.

3.7.1-3

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.5.4 Revision 0 November 20. 2007 3.5.4 ENGINEERED SAFEGUARDS TRAIN LOGIC ALCO 3.5.4 Both trains of engineered safeguards train logic shall be OPERABLE.

APPLICABILITY: OPERATING, HOT STANDBY, HOT SHUTDOWN, INTERMEDIATE SHUTDOWN ACTIONS

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

The Conditions below only apply during performance of maintenance or Technical Specification Surveillance Requirements.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One train inoperable.

A.1 Restore train to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

B.

Required Action and B.1 Be in HOT STANDBY.

7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in HOT SHUTDOWN in an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND B.3 Be in COLD SHUTDOWN in an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY In accordance with Technical Specification Requirements In accordance' with Technical Specification Requirements 3.5.4-1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.5.4 Revision 0 November 20. 2007 BASES

.BACKGROUND Engineered Safety Features Systems are actuated by redundant logic and coincidence networks similar to those used for reactor protection. Each network actuates a device that operates the associated engineered safety features equipment, motor starters and valve operators. The channels are designed to combine redundant sensors, independent channel circuitry, and coincident trip logic. Where possible, different but related parameter measurements are utilized. This ensures a safe and reliable system in which a single failure will not defeat the intended function. The action initiating sensors, bistables and logic are described in USAR Section 7.5. (Reference 1).

TS 3.5, "Instrumentation System" (Reference 2), provides requirements for automatic initiation of the engineered safety features in the event that principal process variable limits are exceeded, and to delineate the conditions of the reactor protection instrumentation and engineered safety features circuits necessary to ensure reactor safety.

TS 3:*5.b allows plant operation to continue at RATED POWER in accordance with Tables TS 3.5-2 through TS 3.5-5. TS 3.5 provides restrictions on continued operation with specific channels inoperable, but does not expressly limit operation with one train of engineered safeguards train logic inoperable.

TS 3.3, Engineered Safety Features and Auxiliary Systems (Reference 3), allows one train of safety injection to be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and allows one containment spray train to be inoperable for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The engineered safeguards train logic supports the safety injection system and the containment spray system.

TS 4.1, Operational Safety Review (Reference 4), requires protective system logic channel testing on a monthly frequency.

ALCO and APPLICABILITY Both trains of engineered safeguards train logic are required to be OPERABLE when the reactor is in the Modes of OPERATING, HOT STANDBY, HOT SHUTDOWN, or INTERMEDIATE SHUTDOWN.

Below these modes, the Engineered Safeguards Train Logic is not required to be operable.

An Engineered Safeguards Logic Train is OPERABLE when it is capable of fulfilling its required function. As stated in TS 3.5 Basis, safety injection signals can be blocked during those OPERATING MODES where they are not "required" for safety and where their 3.5.4-2

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.5.4 Revision 0 November 20. 2007 ALCO and APPLICABILITY (continued) presence might inhibit operating flexibility. The Engineered Safeguards Logic Train remains capable of fulfilling its required function, with safety injection blocked, during those OPERATING MODES which require safety injection to be blocked. Allowed bypass conditions are stated in TS Table 3.5-3.

The conditions provided in the ALCO allow for an orderly and timely performance of the Surveillance Requirements in TS 4.1 by limiting the time that one train of engineered safeguards actuation logic may be out of service to that time needed to perform the required surveillance.

ACTIONS The ACTIONS are modified by a Note indicating that the Conditions associated with this ALCO only apply during performance of maintenance or Technical Specification Surveillance Requirements. This restriction allows for using the allowable outage time only for performance of the required surveillance and any reasonable maintenance actions.

A.1 If one train of engineered safeguards logic is determined to be inoperable, action is required within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to restore the train to an OPERABLE status.

The engineered safeguards train logic is designed with two actuation logic trains. With one actuation logic train inoperable, the remaining logic train would continue to be capable of actuating the engineered safeguards system.

B.1, B.2, B.3 If the Required Action and associated Completion Time for Condition A is not met, the reactor must be brought to a COLD SHUTDOWN condition, where the ALCO does not apply. The time provided is consistent with the Standard Shutdown Sequence in TS 3.0.c.

SURVEILLANCE REQUIREMENTS Applicable surveillance requirements for the engineered safeguards train logic are specified in Technical Specifications.

3.5.4-3

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.5.4 Revision 0 November 20, 2007 REFERENCES

1. USAR 7.5, "Engineered Safety Features Instrumentation".

2. TS 3.5, "Instrumentation System".

3. TS 3.3, "Engineered Safety Features and Auxiliary Systems".

4. TS 4.1, "Operational Safety Review".

3.5.4-4

TRM 2.1 Kewaunee Power Station CORE OPERATING LIMITS REPORT (COLR)

CYCLE 28 REVISION 1 U

/K'

//L/

~,.

Approved

/i ý.li,/ 2ý'-

Date 0 7 />7 Mtg.#

PORC Chairman

Table of Contents Section Title Page 1.0 CORE OPERATING LIMITS REPORT........................................................................

1 2.0 OPERATING LIMITS..................

......... 2 2.1 Reactor Core Safety Limits

............................................................................ 2 2.2 Shutdown Margin (SDM)................................................................................

2 2.3 Moderator Temperature Coefficient..................................................................

2 2.4 Shutdown Bank Insertion Limit.........................................................................

2 2.5 Control Bank Insertion Limits..........................................................................

2 2.6 Nuclear Heat Flux Hot Channel Limits (F0 N(Z))...............................................

3 2.7 Nuclear Enthalpy Rise Hot Channel Factor (FAHN)............................................. 4 2.8 Axial Flux Difference (AFD).............................................................................

4 2.9 Overtemperature AT Setpoint...........................................................................

5 2.10 Overpower AT Setpoint....................................................................................

5 2.11 RCS Pressure, Temperature, and Flow Departure From................................

6 Nucleate Boiling (DNB) Limits 2.12 Refueling Boron Concentration........................................................................

6

List of Figures Figure Title Paqe

1.

Reactor Core Safety Limits Curve (1772 MWt)...........................................................

7

2.

Required Shutdown Reactivity vs. Boron Concentration..............................................

8

.3.

Hot Channel Factor Normalized Operating Envelope (K(z))........................................

9 4.

C ontrol Bank Insertion Lim its....................................................................................

10

5.

W(Z) Values (Top and Bottom 9% excluded)...........

11

6.

Penalty Factor, Fp (% ), for F0EQ(Z)..............................................................................

13 7.

A xial Flux D ifference.................................................................................................

.. 14

List of Tables Table Title Page

1.

NRC Approved Methodologies for COLR Parameters 15 iii

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Revision I 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Kewaunee Power Station (KPS) has been prepared in accordance with the requirements of Technical Specification (TS) 6.9.a.4.

A cross-reference between the COLR sections and the KPS Technical Specifications affected by this report is given below:

COLR Section 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 2.10 2.11 2.12 Figure 1 Figure 2 Figure 3 Figure 4 Figure 5 Figure 6 Figure 7 KPS TS 2.1 3.10.a 3.1.f.3 3.10.d.1 3.10.d.2 3.10.b.1.A 3.10.b.5 3.10.b.6 3.10.b.6.C.i 3.10.b.7 3.10.b.1..B 3.10.b.8 2.3.a.3.A 2.3.a.3.B 3.10.k 3.10.1 3.10.m.1 3.8.a.5 Description Reactor Core Safety Limits Shutdown Margin Moderator Temperature Coefficient Shutdown Bank Insertion Limit Control Bank Insertion Limits Heat Flux Hot Channel Factor (FQ(Z))

Nuclear Enthalpy Rise Hot Channel Factor (FAHN)

Axial Flux Difference (AFD)

Overtemperature AT Setpoint Overpower AT Setpoint RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Refueling Boron Concentration Reactor Core Safety Limits (1772 MWt)

Required Shutdown Margin K(Z) Normalized Operating Envelope Control Bank Insertion Limits W(Z) Values (Top and Bottom 9% excluded)

Penalty Factor, Fp, for F 0Q (Z)

Axial Flux Difference Cycle 28 Page 1 of 20 Rev. 1

KEWAUNEE POWER STATION TRM 2.1 TECHNICAL REQUIREMENTS MANUAL Revision 1 CORE OPERATING LIMITS REPORT CYCLE 28 2.0 Operating Limits The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using the NRC approved methodologies specified in Technical Specification 6.9.a.4.

2.1 Reactor Core Safety Limits The combination of rated power level, coolant pressure, and coolant temperature shall not exceed the limits shown in COLR Figure 1 (1772 MWt). The safety limit is exceeded if the point defined by the combination of Reactor Coolant System average temperature and power level is at any time above the appropriate pressure line.

2.2 Shutdown Margin 2.2.1 When the reactor is subcritical prior to reactor startup, the SHUTDOWN margin shall be at least that shown in COLR Figure 2.

2.3 Moderator Temperature Coefficient 2.3.1 When the reactor is critical and _< 60% RATED POWER, the moderator temperature coefficient shall be _ 5.0 pcm/0 F, except during LOW POWER PHYSICS TESTING. When the reactor is > 60% RATED POWER, the moderator temperature coefficient shall be zero or negative.

2.3.2 The reactor will have a moderator temperature coefficient no less negative than -8 pcm/°F for 95% of the cycle time at full power.

2.4 Shutdown Bank Insertion Limit 2.4.1 The shutdown rods shall be fully withdrawn (>_ 225 steps and <_ 230 steps) when the reactor is critical or approaching criticality.

2.5 Control Bank Insertion Limits 2.5.1 The control banks shall be limited in physical insertion; insertion limits are shown in COLR Figure 4.

Cycle 28 Page 2 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 2.1 Revision I CORE OPERATING LIMITS REPORT CYCLE 28 2.6 Nuclear Heat Flux Hot Channel Factor (F

.(Z.

2.6.1 F N(Z) Limits for Fuel FQN(Z) X 1.03 x 1.05*< (2.50)/P x K(Z) for P > 0.5 FQN(Z) x 1.03 x 1.05 < (5.00) x K(Z) for P < 0.5

[422 V+]

where:

P is the fraction of full power at which the core is OPERATING K(Z) is the function given in Figure 3 Z

is the core height location for the Fa of interest 2.6.2 The measured FEQE(Z) hot channel factors under equilibrium conditions shall satisfy the following relationship for the central axial 80% of the core for fuel:

FQEO(Z) X 1.03 x 1.05 x W(Z) x Fp < (2.5)/ P x K(Z)

[422 V +]

where:

P K(Z) z FP W(Z)

F EQ(z) is the fraction of full power at which the core is OPERATING is the function given in Figure 3 is the core height location for the FQ of interest is the FQEQ(Z) penalty factor described in 2.6.3.

Is the function given in Figure 5 is a measured FQ distribution obtained during the target flux determination 2.6.3 The penalty factor of 1.0 shall be used for TS 3.10.b.6.A and TS 3.10.b.6.B.

The penalty factor provided in Figure 6 shall be used for TS 3.10.b.6.C.i.

The penalty factor for all burnups outside the range of Figure 6 shall be 2%.

Cycle 28 Page 3 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 2.1 Revision I CORE OPERATING LIMITS REPORT CYCLE 28 2.7 Nuclear Enthalpy Rise Hot Channel Factor (F,%HN) 2.7.1 FAH N Limits for Fuel FAHt N x1.04*< 1.70 [1 + 0.3(1-P)]

[422 where:

P is the fraction of full power at which the core is OPERATING 2.8 Axial Flux Difference (AFD) v+]

2.8.1 The Axial Flux Difference (AFD) acceptable operation limits are provided in Figure 7.

Cycle 28 Page 4 of 20 Rev. 1

KEWAUNEE POWER STATION TRM 2.1 TECHNICAL REQUIREMENTS MANUAL Revision I CORE OPERATING LIMITS REPORT CYCLE 28 2.9 Overtemperature AT Setpoint Overtemperature AT setpoint parameter values:

AT0

=

Indicated AT at RATED POWER, %

T

=

Average temperature, OF T'

573.0 OF P

=

Pressurizer Pressure, psig P

=

2235 psig K1

=

1.195 K2

=

0.015/0F K3

=

0.00072/psig T 1

=

30 seconds

.2

=

4 seconds f(AI)

=

An even function of the indicated difference between top and bottom detectors of the power range nuclear ion chambers. Selected gains are based on measured instrument response during plant startup tests, where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt + qb is total core power in percent of RATED POWER, such that (a)

For qt - qb within -15, +10 %, f(AI) = 0 (b)

For each percent that the magnitude of qt - qb exceeds +10 % the AT trip setpoint shall be automatically reduced by an equivalent of 1.51 % of RATED POWER.

(c)

For each percent that the magnitude of qt - qb exceed -1 5 % the AT trip setpoint shall be automatically reduced by an equivalent of 3.78% of RATED POWER.

2.10 Overpower AT Setpoint Overpower AT setpoint parameter values:

AT0

=

Indicated AT at RATED POWER, %

T

=

Average temperature, OF T'

573.0 OF K4 1.095 K15 0.0275/°F for increasing T; 0 for decreasing T K6.

0.00103/°F for T > T' ; 0 for T < T' T

=

10 seconds f(AI)

=

0 for all A]

Cycle 28 Page 5 of 20 Rev. 1

KEWAUNEE POWER STATION TRM 2.1 TECHNICAL REQUIREMENTS MANUAL Revision 1 CORE OPERATING LIMITS REPORT CYCLE 28 2.11 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 2.11.1 During steady state power operation, Tavg shall be < 576.70F for control board indication or < 576.51F for computer indication.

2.11.2 During steady state power operation, Pressurizer Pressure shall be

> 2217 psig for control board indication or > 2219 psig for computer indication 2.11.3 During steady state power operation, reactor coolant total flow rate shall be

> 189,720 gpm.

2.12 Refueling Boron Concentration 2.12.1 When there is fuel in the reactor, a minimum boron concentration of 2500 ppm and a shutdown margin of > 5% A k/k shall be maintained in the Reactor Coolant System during reactor vessel head removal or while loading and unloading fuel from the reactor.

Cycle 28 Page 6 of 20 Rev. 1

KEWAUNEE POWER STATION TFECHNICAL REQUIREMENTS MANUAL TRM 2.1 Revision I CORE OPERATING LIMITS REPORT CYCLE 28 Figure 1 Reactor Core Safety Limits Curve (1772 Mwt)

(Cores Containing 422V+ fuel) 665 L.

I-,

0.

E W 625 I-605 0

0 5.

O U

S585 565 0

20 40 60 80 100 120 Core Power (percent of 1772 MWt)

Cycle 28 Page,7 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Revision I Figure 2 Required Shutdown Reactivity vs. Boron Concentration Z

0)

A-U) 1700 1600 1542 1500 1400 1300 1200 1100 1000 900 800 700 600 500 400 300 200 100 0

~

1-.

- i 0

200 400 600 800 1000 1200 1400 Full Power Equilibrium Boron Concentration (ppm) 1600 1800 2000 Cycle 28 Page 8 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Revision I Figure 3 Hot Channel Factor Normalized Operating Envelope (K(z))

N 1.2 1.1 1

0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0

(12,1.0) 0 1

2 3

4 5

6 7

Core height (ft) 8 9

10 11 12 Cycle 28 Page 9 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Revision I Figure 4 Control Bank Insertion Limits 240 (12.7%, 225)

(64.6%, 225) 220 BANKB (0.0%, 194)

(80.

,1

)__

(100.0%, 185) 160

,=/BANK C/

140 F

(flI I

,I I

C 120

__ 100I IE BANK D 80

, NKi 6 (.00

, 68)

I _

i

___/_

I 40 20/"

(238%,

I0

/

I 0

0 10 20 30 40 50 60 70 80 90 100 Percent of Rated Thermal Power Fully withdrawn shall be the condition where control rods are at a position between the interval

> 225 and < 230 steps withdrawn.

Note: The Rod Bank Insertion Limits are based on a control bank tip-to-tip distance of 126 steps.

Cycle 28 Page 10 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Revision 1 Figure 5 -W(Z) Values (Top and Bottom 9% excluded)

Height BU [MrV'dMTU]

Inft 150 6000 12000 16000 AO = 0.90 AO = -3.20 AO = -3.36 AO = -1.32 (BOTTOMI 1

0.00 1.0000 1.0000 1.0000 10000 2

0.20 1.0000 1.0000 1.0000 1.0000 3

0.40 1.0000 1.0000 1.0000 1.0000 4

0.60 1.0000 1.0000 1.0000 1.0000 5

0.80 1.0000 1.0000 1.0000 1.0000 6

1.00 1.0000 1.0000 1.0000 1.0000 7

1.20 1.3228 1.2245 1.2002 1.2060 8

1.40 1.3076 1.2135 1.1899 1.1971 9

1.60 1.2899 1.2007 1.1782 1.1870 10 1.80 1.2699 1.1864 1.1652 1.1757 11 2.00 1.2507 1.1709 1.1512 1.1635 12 2.20 1.2316 1.1547 1.1367 1.1508 13 2.40 1.2116 1.1382 1.1217 1.1377 14 2.60 1.1916 1.1218 1.1071 1.1247 15 2.80 1.1724 1.1073 1.0985 1.1127 16 3.00 1.1553 1.0997 1.0953 1.1033 17 3.20 1.1434 1.0972 1.0935 1.1013 18 3.40 1.1372 1.0951 1.0912 1.1033 19 3.60 1.1341 1.0929 1.0888 1.1056 20 3.80 1.1313 1.0934 1.0858 1.1078 21 4.00 1.1281 1.0945 1.0858 1.1098 22 4.20 1.1246 1.0950 1.0891 1.1114 23 4.40 1.1207 1.0954 1.0933 1.1126 24 4.60 1.1164 1.0955 1.0970 1.1132 25 4.80 1.1115 1.0954 1.1001 1.1140 26 5.00 1.1073 1.0946 1.1029 1.1151 27 5.20 1.1034 1.0945 1.1050 1.1158 28 5.40 1.0989 1.0960 1.1066 1.1159 29 5.60 1.0940 1.0979 1.1077 1.1165 30 5.80 1,0895 1.1018 1.1084 1.1220 31 6.00 1.0899 1'.1057 1.1130 1.1314 32 6.20 1.0954 1.1104 1.1198 1.1421 33 6.40 1.1021 1.1174 1.1267 1.1529 34 6.60 1.1079 1.1244 1.1376 1.1626 35 6.80 1.1127 1.1307 1.1480 1.1711 36 7.00 1.1167 1.1354 1.1571 1.1783 37 7.20 1.1197 1.1411 1.1663 1.1842 38 7.40 1.1222 1.1501 1.1753 1.1886 39 7.60 1.1246 1.1583 1.1831 1.1915 40 7.80 1.1260 1.1655 1.1896 1.1928 41 8.00 1.1264 1.1718 1.1949 1.1924 42 8.20 1.1256 1.1769 1,1988 1.1903 43 8.40 1.1238 1.1809 1.2013 1.1864 44 8.60 1.1209 1.1838 1,2023 1.1803 45 8.80 1.1180 1.1842 1.2017 1.1757 46 9.00 1.1214 1.1880 1.1993 1.1755 47 9.20 1.1366 1.1975 1.1996 1.1773 48 9.40 1.1498 1.2089 1.2069 1.1769 49 9.60 1.1624 1.2237 1.2161 1.1765 50 9.80 1.1768 1.2413 1.2232 1.1782 Cycle 28 Page 11 of 20 Rev.1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Revision I Figure 5 Cont'd (Top and Bottom 9% excluded)

Height I

BU

~IŽMYI~Ž.

I 150 6000 12000 16000 Ift)

AO = 0.90 AO = -3.20 AO = -3.36 AO = -1.32 51 10.00 1.1913 1.2555 1.2311 1.1834 52 10.20 1.2053 1.2661 1.2410 1.1893 53 10.40 1.2214 1.2764 1.2504 1.1996 54 10.60 1.2398 1,2816 1.2632 1.2128 55 10.80 1.2526 1.2878 1.2732 1.2258 56 11.00 1.0000 1,0000 1.0000 1.0000 57 11.20 1.0000 1.0000 1.0000 1.0000 58 11.40 1.0000 1.0000 1.0000 1.0000 59 11.60 1.0000

, 1.0000 1.0000 1.0000 60 11.80 1.0000 1.0000 1.0000 1.0000

[TOP] 61 12.00 1.0000 1.0000 1.0000 1.0000 Cycle 28 Page 12 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Revision I Figure 6 Penalty Factor, Fp (%), for FQEQ(Z)

Cycle Burnup Penalty (MWD/MTU)

Factor 150 2.00 20,743 2.00 Note: Linear interpolation is adequate for intermediate cycle burnups.

All cycle burnups outside the range of the table shall use a penalty factor, Fp, of 2.0%.

Refer to TS 3. 1 0.b.6.C.

Cycle 28 Page 13 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Rcvision 1 Figure 7 Axial Flux Difference 110 100 90 I.-

0 0

0 a.

I-0 80 70 60 50 40

-40

-30

-20

-10 0

10 20 30 40 Axial Flux Difference (% delta-I)

Note: This figure represents the Relaxed Axial Offset Control (RAOC) band used in safety analyses, it may be administratively tightened depending on in-core flux map results. Refer to Figure RD 11.4.1 of the Reactor Data Manual.

Cycle 28 Page 14 of 20 R~ev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2. 1 Revision I Table 1 NRC Approved Methodologies for COLR Parameters COLR Section Parameter NRC Approved Methodology 2.1 Reactor Core Safety Limits 2.2 Shutdown Margin WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods for Application to Kewaunee" Report, dated August 21, 1979, report date September 29, 1978.

Kewaunee Nuclear Power Plant-Review for Kewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, Revision 3 (TAC NO MB0306) dated September 10, 2001.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods for Application to Kewaunee" Report, dated August 21, 1979, report date September 29, 1978.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods for Application to Kewaunee" Report, dated August 21, 1979, report date September 29, 1978.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods for Application to Kewaunee" Report, dated August 21, 1979, report date September 29, 1978.

2.3 Moderator Temperature Coefficient 2.4 Shutdown Bank Insertion Limit Cycle 28 Page 15 of 20 Rev. 1

KEWAUNEE POWER STATION TRM 2.1 TECHNICAL REQUIREMENTS MANUAL Rcvision I CORE OPERATING LIMITS REPORT CYCLE 28 Table I (cont)

NRC Approved Methodologies for COLR Parameters COLR Section Parameter NRC Approved Methodology Kewaunee Nuclear Power Plant-Review for Kewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, Revision 3 (TAC NO MB0306) dated September 10, 2001.

2.5 Control Bank Insertion Limits WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods for Application to Kewaunee" Report, dated August 21, 1979, report date September 29, 1978.

Kewaunee Nuclear Power Plant-Review for Kewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, Revision 3 (TAC NO MB0306) dated September 10, 2001.

2.6 Heat Flux Hot Channel Factor WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control-FQ Surveillance Technical Specification,"

February 1994.

Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods for Application to Kewaunee" Report, dated August 21, 1979, report date September 29, 1978.

WCAP-12945-P-A (Proprietary),

"Westinghouse Code Qualification Document for Best-Estimate Loss-of-Coolant Accident Analysis," Volume I, Rev.2, and Volumes II-V, Rev.1, and WCAP-14747 (Non-Proprietary), March 1998.

Cycle 28 Page 16 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2. 1 Revision i Table 1 (cont)

NRC Approved Methodologies for COLR Parameters COLR Section Parameter NRC Approved Methodoloqy (FQ(Z))

Model ANF-88-133 (P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation, dated December 1991.

WCAP-14449-P-A, "Application Of Best Estimate Large Break LOCA Methodology To Westinghouse PWRs With Upper Plenum Injection," Revisioni, and WCAP-14450-NP-A, Rev.1 (Non-Proprietary), October 1999.

WCAP-12610-P-A, "Vantage+ Fuel Assembly Reference Core Report," April 1995.

WCAP-1 0054-P-A/WCAP-1 0081 -NP-A, "Westinghouse. Small Break ECCS Evaluation Using the NOTRUMP Code,"

August 1985.

WCAP-1 0054-P-A/WCAP-1 0081 -NP-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the Code: Safety Injection into the Broken Loop and COSI Condensation Model," July 1997.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods for Application to Kewaunee" Report, dated August 21, 1979, report date September 29, 1978.

Kewaunee Nuclear Power Plant-Review forKewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, NOTRUMP 2.7 Nuclear Enthalpy Rise Hot Channel Factor (FNAH)

Cycle 28 Page 17 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Revision I Revision 3 (TAC NO MB0306) dated September 10, 2001.

Table 1 (cont)

NRC Approved Methodologies for COLR Parameters COLR Section Parameter NRC Approved Methodology 2.8 Axial Flux Difference XN-82-06 (P)(A) Revision 1 and Supplements 2, 4, and 5, "Qualifications of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company, dated October 1986.

ANF-88-133 (P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups for 62 GWd/MTU," Advanced Nuclear Fuels Corporation, dated December 1991.

EMF-92-116 (P)(A) Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs," Siemens Power Corporation, dated February 1999.

WCAP-10216-P-A, Revision 1A, "Relaxation of Constant Axial Offset (AFD) Control-FQ Surveillance Technical Specification,"

February 1994.

Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods for Application to Kewaunee" Report, dated August 21, 1979, report date September 29, 1978.

Kewaunee Nuclear Power Plant-Review for Kewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, Revision 3 (TAC NO MB0306) dated September 10, 2001.

WCAP-8745-P-A, "Design Bases For The Thermal Overpower AT and Thermal Overtemperature AT Trip Functions,"

September 1986.

2.9 Reactor Protection System (RPS)

I nstrumentation-Overtemperature AT Cycle 28 Page 18 of 20 Rev. 1

KEWAUNEE POWER STATION TRM 2.1 TECHNICAL REQUIREMENTS MANUAL Revision I CORE OPERATING LIMITS REPORT CYCLE 28 Table 1 (cont)

NRC Approved Methodologies for COLR Parameters COLR Section Parameter NRC Approved Methodology Kewaunee Nuclear Power Plant-Review for Kewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, Revision 3 (TAC NO MB0306) dated September 10, 2001.

CENP-397-P-A, "Improved Flow Measurement Accuracy Using Cross Flow Ultrasonic Flow Measurement Technology,"

Rev. 1, May 2000.

2.10 Reactor Protection System (RPS)

WCAP-8745-P-A, "Design Bases For The Instrumentation-Overpower AT Thermal Overpower AT and Thermal Overtemperature AT Trip Functions,"

September 1986.

Kewaunee Nuclear Power Plant-Review for Kewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, Revision 3 (TAC NO MB0306) dated September 10, 2001.

CENP-397-P-A, "Improved Flow Measurement Accuracy Using Cross Flow Ultrasonic Flow Measurement Technology,"

Rev. 1, May 2000.

Cycle 28 Page 19 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL CORE OPERATING LIMITS REPORT CYCLE 28 TRM 2.1 Revision I Table 1 (cont)

NRC ApDroved Methodoloaies for COLR Parameters COLR Section Parameter NRC Approved Methodoloqv 2.11 RCS Pressure, Temperature, and Flow Departure From Nucleate (DNB) Limits WCAP-1 1397-P-A, "Revised Thermal Design Procedure, "April 1989, for those events analyzed using RTDP Kewaunee Nuclear Power Plant-Review for Kewaunee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, Revision 3 (TAC NO MB0306) dated September 10, 2001.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985 for those events not utilizing RTDP.

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

Safety Evaluation by the Office of Nuclear Reactor Regulation on "Qualifications of Reactor Physics Methods for Application to Kewaunee" Report, dated August 21, 1979, report date September 29, 1978.

2.12 Boron Concentration Cycle 28 Page 20 of 20 Rev. 1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.0.9 Revision 0 Februarv 11. 2008 3.0.9 Steam Exclusion System ALCO 3.0.9 Steam Exclusion System shall be OPERABLE as follows:

a.

Both trains of the steam exclusion actuation logic, including their associated temperature sensors, shall be OPERABLE,

b.

All required actuators to control damper position and the associated steam exclusion dampers shall be OPERABLE,

c.

All required steam exclusion doors shall be OPERABLE,

d.

All required penetrations shall be OPERABLE.

e.

All required steam exclusion boundaries such as walls, hatches, etc.,

shall be OPERABLE.

NOTE A failed temperature sensor may be considered operable if its output is replaced with a simulated high temperature signal.

APPLICABILITY:

Reactor Coolant Temperature > 350'F ACTIONS

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

Performance of an evaluation, within the required Completion Time, that verifies required equipment supported by the Steam Exclusion System is OPERABLE, is an acceptable alternative to the stated Required Actions.

CONDITION REQUIRED ACTION COMPLETION TIME A.

Steam Exclusion System A.1 Declare all equipment Immediately inoperable for reasons supported by the inoperable other than Conditions B, C steam exclusion barrier or D.

inoperable.

OR OR Required Action and A.2 Verify required equipment Immediately associated Completion supported by the inoperable Time for Conditions B, C or steam exclusion barrier is D not met.

OPERABLE.

(continued) 3.0.9-1

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.0.9 Revision 0 Februarv 11, 2008 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

One train of steam B.1 Close one steam 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months exclusion actuation logic exclusion damper in each inoperable, affected duct.

C.

One of two redundant C.1 Close one steam 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months steam exclusion damper exclusion damper in each actuators or their affected duct.

associated dampers inoperable.

D.

One of three channels of D.1 Close one steam 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months instrumentation for each exclusion damper in each area monitored inoperable, affected duct.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ASR 3.0.9.1 Perform CHANNEL CALIBRATION of each 18 months resistance temperature detector temperature loop.

ASR 3.0.9.2 Perform functional test of actuation logic.

18 months ASR 3.0.9.3 Perform functional test of system dampers.

18 months 3.0.9-2

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.0.9 Revision 0 February 11, 2008 BACKGROUND Steam exclusion zones were defined to designate locations, which are prote cted against steam intrusion. The steam exclusion system aids in the mitigation of a high-energy line break outside of containment. The primary functions of steam exclusion are to provide suitable environmental conditions for needed equipment operation, and a habitable environment for personnel in areas outside of containment which may require access should a high energy line break occur (Reference 1).

Degradation of the steam exclusion system, whether for facilitating performance of maintenance or for other reasons, requires that corrective or compensatory action be taken.

10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" (Maintenance Rule),

requires assessing and managing the increase in risk associated with the performance of maintenance activities. 10 CFR 50.65(a)(4) applies to the removal of hazard barriers, such as the steam exclusion system. NRC Regulatory Issue Summary 2001 -009, "Control of Hazard Barriers", provides guidance on the control of hazard barriers that is consistent with the provisions of the maintenance rule (Reference 2).

ALCO and APPLICABILITY Above a reactor coolant temperature of 350 0F, the Steam Exclusion System is required to be OPERABLE, including:

both trains of the steam exclusion actuation logic (including their associated temperature sensors); all required actuators to control damper position and the associated steam exclusion dampers; all required steam exclusion doors; all requ~ired penetrations; and all required steam exclusion boundaries such as walls, hatches, etc.

At or below a reactor coolant temperature of 3500F, the Steam Exclusion System is not required to be operable.

Depending on whether a steam exclusion barrier is degraded or inoperable determines whether corrective or compensatory measures may be taken within a certain time period or are immediately required.

The ALCO is modified by a NOTE that allows a failed temperature sensor in the steam exclusion system to be considered operable if its output is replaced with a simulated high temperature signal.

3.0.9-3

KEWAUNEE POWER STATION TRM 3.0.9 TECHNICAL REQUIREMENTS MANUAL Revision 0 February 11,2008 BASES ACTIONS Risk-informed analyses have demonstrated the acceptability of the provided conditions affecting the OPERABILITY of the Steam Exclusion System. The associated Required Actions direct specific equipment or systems to be declared inoperable or compensatory measures to be taken.

The ACTIONS are modified by a Note indicating that an acceptable alternative to the stated Required Action is an evaluation that demonstrates the OPERABILITY of the Steam Exclusion System and/or equipment supported by the Steam Exclusion System. Utilization of the allowance provided by this Note will normally be for specific circumstances.

A.1, A.2 If the Steam Exclusion System is inoperable for reasons other than those listed in Conditions B, C or D, a barrier that may be credited with protecting a supported component or system is no longer capable of providing that protection. This condition requires that all equipment supported by the inoperable steam exclusion barrier be immediately declared inoperable unless an evaluation has been performed to determine that required equipment supported by the inoperable steam exclusion barrier is OPERABLE.

If the Required Action and associated Completion Time for Conditions B, C or D are not met, all equipment supported by the associated steam exclusion barrier must be immediately declared inoperable unless an evaluation has been performed to determine that required equipment supported by the inoperable steam exclusion barrier is OPERABLE.

Because of the immediate completion time, use of the allowance provided by the associated NOTE and in Required Action A.2 would necessitate that any evaluation that demonstrates the OPERABILITY of the Steam Exclusion System and/or equipment supported by the Steam Exclusion System, would need to be completed in advance of entering Condition A.

B.1 If one train of steam exclusion actuation logic is determined to be inoperable, action is required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to close one steam exclusion damper in each affected duct. Closing one steam exclusion damper in each affected duct restores the steam exclusion 3.0.9-4

KEWAUNEE POWER STATION TRM 3.0.9 TECHNICAL REQUIREMENTS MANUAL Revision 0 February 11, 2008_

BASES ACTIONS function of that duct. The completion time is from a risk based analyses (continued) for acceptable out of service times.

The steam exclusion system is designed with two actuation logic trains.

With one actuation logic train inoperable, the remaining logic train would continue to be capable of actuating the steam exclusion system.

Therefore, the steam exclusion barrier remains capable of providing its support function of protecting required equipment and maintaining a habitable environment for personnel.

0. 1 If one steam exclusion damper actuator or its associated damper is inoperable in a duct with two redundant steam exclusion dampers, action is required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to close one steam exclusion damper in each affected duct. Closing one steam exclusion damper in each affected duct restores the steam exclusion function of that duct.

The completion time is from a risk based analyses for acceptable out of service times.

With one steam exclusion damper actuator or its associated damper inoperable in a duct with two redundant dampers, the remaining damper would continue to be capable of providing steam exclusion. Therefore, the steam exclusion barrier remains capable of providing its support function of protecting required equipment and maintaining a habitable environment for personnel.

D. 1 If one of three channels of instrumentation for each steam exclusion area monitored is determined to be inoperable, action is required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to close one steam exclusion damper in each affected duct. Closing one steam exclusion damper in each affected duct restores the steam exclusion function of that duct. The completion time is from a risk based analyses for acceptable out of service times.

The steam exclusion system is designed with three channels of instrumentation for each steam exclusion area monitored. With one instrument channel inoperable, the remaining two channels would continue to be capable of actuating the steam exclusion system.

Therefore, the steam exclusion barrier remains capable of providing its support function of protecting required equipment and maintaining a habitable environment for personnel.

3.0.9-5

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.0.9 Revision 0 February 11, 2008 BASES SURVEILLANCE REQUIREMENTS ASR 3.0.9.1 ASR 3.0.9.1 requires the performance of a CHANNEL CALIBRATION of each resistance temperature detector temperature loop in the Steam Exclusion System every 18 months.

ASR 3.0.9.2 ASR 3.0.9.2 requires the performance of a functional test of Steam Exclusion System actuation logic every 18 months. This test is accomplished by verifying that various combinations of simulated high temperature signals cause the Steam Exclusion System actuation logic to generate an actuation signal.

ASR 3.0.9.3 ASR 3.0.9.3 requires the performance of a functional test of Steam Exclusion System dampers every 18 months. This test is accomplished by verifying the ability of the system's dampers to properly close with a simulated system actuation signal.

REFERENCES

1. KPS USAR Section 10A.2.3.8, Steam Exclusion Zones.

2. NRC Regulatory Issue Summary RIS 2001-009, "Control of Hazard Barriers."

3.0.9-6

KEWAUNEE POWER STATION TECHNICAL REQUIREMENTS MANUAL TRM 3.5.1 Revision 2 February 22, 2008 3.5.1 CONTAINMENT HYDROGEN MONITORING SYSTEM APPLICABILITY During OPERATING or HOT STANDBY Modes.

OBJECTIVE To monitor the beyond design-basis accident containment air and provide a continuous indication of hydrogen concentration.

TECHNICAL REQUIREMENTS Administrative Limiting Conditions for Operation (ALCOs)

a. Two trains of the Containment Hydrogen Monitoring System and associated Containment Dome Fans shall be functional except as allowed below:

1. One train may be nonfunctional for 30 days.

2. Two trains may be nonfunctional for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

b. If functionality is not restored in the timeframes above, a Condition Report will be initiated immediately to address why the hydrogen monitors were not restored to functional status within the allotted time.

c. A -change in operational MODES or conditions is acceptable with one or both trains of the Containment Hydrogen Monitoring System and its associated Containment Dome Vent Fan nonfunctional-.

Administrative Surveillance Requirement (ASRs)

CHANNEL

] CHECK CALIBRATEI TEST I REMARKS DESCRIPTION I_______

I__________

Contanmen HydogenEach C ontin eto Hyr oge Daily refueling Monthly Monitors

~~~~~cycle 35A-1

KEWAUNEE POWER STATION TRM 3.5.1 TECHNICAL REQUIREMENTS MANUAL Revision 2 February 22, 2008 BASES The TS requirements for a Containment Hydrogen Monitoring System have been removed from TS as listed in the Federal Register on September 25, 2003. Guidance for the Consolidated Line Item Improvement Process (CLIIP) has been incorporated in the Technical Specification Task Force (TSTF) Change Traveler 447, Rev.1. Part of the requirements for removing Containment Hydrogen Monitoring System from TS was to place any remaining requirements in a Licensee controlled document (Technical Requirements Manual) with the requirements that a hydrogen monitoring system be available for beyond design-basis accident monitoring of containment hydrogen levels.

Even though the requirements for Hydrogen Monitors were taken out of TS, the system still needs to be available for beyond design-basis accident monitoring of containment hydrogen levels. In the event ALCO a.1 or a.2 are not met, a Condition Report will be initiated immediately to address why the hydrogen monitors were not restored to functional status within the allotted time. Actions shall be implemented in a timely manner to place the unit in a safe condition as determined by plant management. The intent of this Condition Report is to utilize the Corrective Action Program to assure prompt attention and adequate management oversight to minimize the additional time the hydrogen monitors are nonfunctional.

The USAR credits the operation of the Containment Dome Vent Fans in section 14.3.7.17. The sample ports are located near the discharge of the Containment Dome Fans, which permit rapid detection of hydrogen escaping from the reactor. The fans draw suction from the upper areas of containment, which prevents the formation of a stratified atmosphere. KPS takes credit for the containment dome vent fans as a support system for the hydrogen monitors. (

Reference:

PORC meeting 97-097, KAP 01 -527, Commitment 97-115, and Inspection Report 97-10 IFI 305/97010-01.)

3.5.1-2

ML081070072

Text

Attachment:

Background

Reference:

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