Text

License Amendment Request 203 Re Technical Specifications on Rod Position Indication
	ML041620503
Person / Time
Site:	Kewaunee
Issue date:	05/25/2004
From:	Coutu T; Nuclear Management Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NRC-04-062
	Download: ML041620503 (39)
	v • d • e

ComrtOedpord by Kewaunee Nuclear Power Plant Operated by Nuclear Management Company, LLC May 25, 2004 NRC-04-062 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Kewaunee Nuclear Power Plant Docket 50-305 License No. DPR-43 License Amendment Request 203 To The Kewaunee Nuclear Power Plant Technical Specifications. "Rod Position Indication."

References:

1) Letter from NMC to NRC, "Proposed Amendment 181 to the Kewaunee Nuclear Power Plant Technical Specifications," dated January 14, 2002 (Adams Accession No. ML020360432)

2) Letter from NRC to NMC," Kewaunee - Request for Additional Information related to Request for Proposed Amendment to Revise KNPP TS Section 3.10.f," dated May 3, 2002. (Adams Accession No. ML021230038)

3) Letter from NMC to NRC, "Response to Nuclear Regulatory Commission request for additional information regarding Kewaunee Nuclear Power Plant proposed amendment 181," dated July 12, 2002.

(Adams Accession No. ML022040708)

4) Letter from NMC to NRC, "Kewaunee Proposed Amendment 181a Application for Technical Specification Change Regarding Allowed Outage Time for the Individual Rod Position Indicator System," dated September 18, 2002. (Adams Accession No. ML022700500)

5) Email from NRC to NMC, "Kewaunee - Review Regarding Proposed Amendment Request TS 3.10.f, "Inoperable Rod Position Indicator Channels," Changes," dated December 10, 2002. (Adams Accession No. ML023440290 and No. ML023440312)

N490 Highway 42

Kewaunee, Wisconsin 54216-9511 Telephone: 920.388.2560

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Docket 50-305 NRC-04-062 Page 2

6) Letter from NMC to NRC, 'Withdrawal of Proposed Amendment Regarding Allowed Outage Time for Individual Rod Position Indicator System," dated January 8, 2003. (Adams Accession No. ML030150585)

7) Letter from NRC to NMC, "Kewaunee Nuclear Power Plant -

Withdrawal of an Amendment Request Regarding Allowed Outage Time for the Individual Rod Position Indicator System, dated January 10, 2003. (Adams Accession No. ML030100237)

The Nuclear Management Company, LLC, (NMC) is submitting a license amendment request (LAR), number 203, to the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications (TS) to revise Section 3.10.f, "Inoperable Rod Position Indicator Channels." This amendment request was previously submitted in reference 1 and later withdrawn in reference 6. The proposed amendment requested an allowed outage time (AOT) for the Individual Rod Position Indicator (IRPI) system of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with more than one IRPI per group inoperable and incorporated the applicable requirements contained in NUREG 1431, "Standard Technical Specifications for Westinghouse Plants (NUREG 1431). NMC withdrew this proposed amendment based on a review of the internal resources needed at the time. However, NMC also indicated we may decide to pursue this change in the future. NMC has decided to pursue the LAR and is hereby submitting a request to modify the KNPP TS.

The previous submittal formatted the KNPP TS for TS 3.10.f in the tabular format of NUREG 1431, Improved Standard Technical Specifications (ISTS). In reference 5, the NRC requested NMC to provide plant specific analysis that the proposed TS would not result in unacceptable application of ISTS concepts such as those of the completion time clocks and logical connectors. NMC decided the analysis required resources beyond the advantages of adopting the ISTS format and withdrew the request. This amendment request (LAR 203) provides the proposed change in KNPP custom TS (CTS) format and includes the additional information previously requested by the NRC staff, except for the question associated with the change to the ISTS format, which NMC is not requesting.

This LAR requests changes to KNPP TS to add an allowed outage time (AOT) for the IRPI system of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with more than one IRPI per group inoperable. The current TS do not have an AOT for this condition. Additional changes are requested to add the demand step counters to the KNPP TS and to add a note to allow for a soak time subsequent to substantial rod motion for the rods that exceed their position limits before invoking the TS requirements. Additionally this amendment request for TS section 3.10.f is an administrative change to define "immediately" in TS section 1.0. These

Docket 50-305 NRC-04-062 Page 3 changes generally follow the specifications found in NUREG 1431, "Standard Technical Specifications for Westinghouse Plants," with the exceptions explained in enclosure 1.

This LAR incorporates, as much as practical, the Nuclear Energy Institute (NEI) technical specification task force (TSTF)-234 traveler.

Recently NMC has noticed an increase in the number of rod position indication problems and a review of the IRPI system was conducted. The rod positions were reviewed against the Plant Information (PI) system trends, which revealed that the indicated positions had increased approximately six steps since plant startup from the forced outage in January 2004. Trends before 2000 showed that IRPI position drift followed seasonal changes in containment temperature. Since 2000, IRPI indications have drifted upward during the power cycle, independent of seasonal temperature changes. The most probable cause of the position indication drift is the alternating current (AC) power supply to the IRPI system. This places the reliability of the IRPI power supply in question. Under current KNPP TS if the IRPI power supply fails, the plant power level must be reduced. As this may not be the least risk action, when NMC notices a trend towards unacceptable degradation, NMC may request this amendment request be processed expeditiously, to allow for on-line repair.

The NMC has concluded that proposed change will not adversely affect the health and safety of the public. The rod position indication instrumentation is not an assumed accident initiator. This proposed change does not alter the design, function, or operation of any plant component and does not install any new or different equipment.

The rod position indication system is an instrumentation system that provides indication to the operators that a control rod may be misaligned. The rod position indication system also provides a backup indication in the event axial or radial power tilts were to be developed. Inoperable individual rod position indication instrumentation does not by itself in any way harm or impact reactor operation.

This proposed amendment is not currently needed for continued plant operation and does not contain proprietary information. Enclosure 1 to this letter contains a description, a safety analysis, a no significant hazards determination and environmental considerations for the proposed changes. Enclosure 2 contains the marked up Technical Specification pages. Enclosure 3 contains the affected Technical Specification pages as revised. Enclosures 4 and 5 contain the marked up and revised TS Basis pages, respectively. In Enclosure 2, Strikeout TS pages, additions to the specifications are double underlined while deletions are 6trikeeut.

A complete copy of this submittal has been transmitted to the State of Wisconsin as required by 10 CFR 50.91 (b)(1).

Docket 50-305 NRC-04-062 Page 4 NMC requests approval of the proposed amendment by May 13, 2005. Once approved, NMC requests a 60-day implementation period.

This letter makes the no new commitments or revisions to previous commitments.

If you have any questions or require additional information, please contact Mr. Gerald Riste at (920) 388-8424.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on May 25, 2004.

Thomas Coutu Site Vice President Kewaunee Nuclear Power Plant Nuclear Management Company, LLC Enclosures 1.

2.

3.

4.

5.

Evaluation of License Amendment Request Marked Up TS Pages Affected TS Pages Marked Up TS Basis Pages Affected TS Basis Pages cc:

Administrator, Region Ill, USNRC Senior Resident Inspector, Kewaunee, USNRC Project Manager, Kewaunee, USNRC Public Service Commission of Wisconsin

ENCLOSURE 1 NUCLEAR MANAGEMENT COMPANY, LLC EVALUATION OF LICENSE AMENDMENT REQUEST 203 TO KEWAUNEE NUCLEAR POWER PLANT, OPERATING LICENSE NO. DPR-43 DOCKET NO. 50-305

1.0 DESCRIPTION

This letter is a request to amend Operating License DPR-43 for the Kewaunee Nuclear Power Plant (KNPP).

The proposed changes would revise the Operating License to add requirements to the Technical Specification (TS) associated with the axial position indication systems of shutdown rods and control rods. The rod position indication systems are described in KNPP Updated Safety Analysis Report (USAR) Section 7.3, "Regulating Systems."

2.0 PROPOSED CHANGE

Current KNPP TS allows for one Individual Rod Position Indicator (IRPI) per group or two per bank, to be out of service indefinitely. NMC is requesting to add the following new requirements to the KNPP TS:

1. Add a definition to TS 1.0 to define the term "Immediately."

2. Add a note to TS 3.10.e and TS 3.10.f to allow for a soak time of up to one hour before the position requirements take affect after significant rod motion.

3. Reword TS 3.10.f.1 to follow the format of the other additions to the TS.

4. Modify the current method of determining rod position to require verification of rod position by movable incore detectors.

5. Add a note to TS 3.1O.f.1 and TS 3.10.f.4 to allow for separate entry conditions for each IRPI or demand position indicator inoperable.

6. Add TS 3.1 0.f.2 to allow for more than one rod per group to be out of service for a maximum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

7. Add TS 3.1 O.f.3 to add requirements if a rod with an inoperable IRPI is moved in excess of 24 steps in one direction since its position was last determined.

8. Add TS 3.1 0.f.4 requiring the demand position indicators to be operable and stating action requirements if the indicators are inoperable.

The main purpose for this license amendment request is to allow the condition where more than one IRPI per group may be inoperable. If more than one IRPI per group were found to be out of service this change will prevent unnecessary power changes due to minor equipment problems that can be corrected and pose no safety concern.

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The 24-hour allowed outage time (AOT) will include a required action that the rod control system immediately is placed under manual control and the reactor coolant temperature is logged hourly during the inoperable condition. Additional changes, as listed above, are requested to incorporate the requirements contained in NUREG 1431, "Standard Technical Specifications Westinghouse Plants," revision 2 (ISTS), as practical. Deviations from ISTS are explained in a later section of this document.

KNPP TS Basis for TS 3.10.e and TS 3.10.f are modified to describe the changes and provide clarification to the operators for their use. The TS basis pages are enclosed.

3.0 BACKGROUND

KNPP has 29 rod control cluster assemblies (RCCAs)I1] separated into shutdown rods and control rods. Of the 29 RCCAs, 8 are shutdown rods and 21 are control rods.

Each RCCA has 16 rodlets that are inserted into the fuel assembly at distinct locations to uniformly control the reactivity of the fuel assembly the RCCA is inserted into or removed from. The rodlets contain a neutron absorbing material that, when inserted into the fuel assembly, remove neutrons from the fission process, shutting down the reactor. The RCCAs (called rods) are further separated into banks and groups as shown below.

KNPP RCCA Designations [2]

Shutdown Rods Bank A Bank B Group 1 Group 2 Group 1 2 Rods J_2 Rods 4 Rods Control Rods Bank A Bank B Bank C Bank D Group 1 Group 2 Group 1 Group 1 Group 2 Group 1 4 Rods 4 Rods 4 Rods 2 Rods 3 Rods 4 Rods KNPP USAR Figure 3.2-1 shows the location of the rod assemblies in the core. The four control banks (A, B, C, and D) are the only rods that can be operated under automatic control. All RCCAs in a group are electrically paralleled to step simultaneously.

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Reactor startup is accomplished by first manually withdrawing the shutdown rods to the full out position. This action requires that the operator select one of the shutdown banks on a control board mounted selector switch and then position the IN-HOLD-OUT lever (which is spring returned to the HOLD position) to the OUT position. The operator then selects the other shutdown bank and repeats the process.

The control banks are then withdrawn manually and sequentially by the operator by first selecting the MANUAL position on the control board mounted selector switch and then positioning the IN-HOLD-OUT lever to the OUT position. In the MANUAL selector switch position, the rods are withdrawn (or inserted) in a predetermined programmed sequence by the programming equipment.

As a plant startup progresses, when reactor power reaches approximately 15%, the operator may select the AUTOMATIC position, where the IN-HOLD-OUT switch is electrically removed from the rod control circuit, and rod motion is controlled by the Rod Control System. An interlock limits automatic rod withdrawal to reactor power levels above 15%. In the AUTOMATIC position, the rods are again withdrawn (or inserted) in a predetermined programmed sequence by the automatic programming equipment, maintaining a programmed reactor coolant system (RCS) average temperature (Tavg).

Control rod programming is sequenced such that as the first bank out reaches a preset position, the second bank begins to move out simultaneously with the first bank. This staggered withdrawal sequence continues until control rods either reach their fully withdrawn position, or reach the desired position to control axial flux, normally all rods are fully withdrawn at full power. The programmed insertion sequence, manual or automatic, is the opposite of the withdrawal sequence, i.e., the last control bank out is the first control bank in.

The shutdown groups of rods together with the control groups are capable of shutting the reactor down under all conditions. They are used in conjunction with the adjustment of chemical shim to provide shutdown margin of at least 1% Ak/k following reactor trip with the most reactive rod in the fully withdrawn position.

During normal power operation, it is desirable to maintain the rods in alignment with their respective banks. This provides consistency with the assumptions of the safety analyses, maintains symmetric neutron flux and power distribution profiles, provides assurance that peaking factors are within acceptable limits, and assures adequate shutdown margin. The Bank D Rod Withdrawal Limit at 220 steps ensures that the control rods do not automatically withdraw beyond the fully withdrawn position. At 220 steps, the operator must manually withdraw the control rods to the fully withdrawn position.

Two separate systems are provided to sense and display control rod position as described below:

a) Analog System (IRPI) - a linear position transmitter produces an analog signal of actual position for each RCCA.

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An electrical coil stack linear variable differential transformer is placed above the stepping mechanisms of the control rod magnetic jacks external to the rod/reactor coolant system pressure housing. When the associated control rod is at the bottom of the core, the magnetic coupling between primary and secondary windings is small and there is a small voltage induced in the secondary. As the magnetic jacks raise the control rod, the relatively high permeability of the lift rod causes an increase in magnetic coupling. Thus, an analog signal proportional to rod position is derived.

Direct, continuous readout of every RCCA position is presented to the operator by individual control board meter indications, without need for operator selection or switching to determine rod position. The rod position is also displayed on the Plant Process Computer System (PPCS).

Another means of detecting individual rod position is by directly reading the voltage produced by the detection circuits conditioning module. Rod position in steps withdrawn is determined by use of a table correlating the conditioning module output voltage to rod steps.

Lights are provided for rod bottom positions for each rod. Bistable devices operate the lights in the analog system.

b) Digital System (Demand Position) - The digital system counts pulses generated in the rod drive control system. One counter is associated with each group of RCCAs. Readout of the digital system is in the form of add-subtract counters reading the number of steps of rod withdrawal with one display for each group. These readouts are mounted on the control panel.

The digital and analog systems are separate systems; each serves as backup for the other. Operating procedures require the reactor operator to compare the digital and analog readings upon receiving a rod deviation alarm. Therefore, a single failure in rod position indication does not in itself lead the operator to take erroneous action in the operation of the reactor.

4.0 TECHNICAL ANALYSIS

1. NMC is proposing eight changes to the KNPP TS.

The first change is to add a definition to TS Section 1.0 to define the term "Immediately."

This change is necessary to clarify the action required by the new TS item, TS 3.10.f.2.A. This item requires the operators to "immediately" place the control rods in manual. Without this definition, confusion may ensue as to the time required for completion of the action. This definition is consistent with NUREG 1431, Standard Technical Specifications for Westinghouse Plant." This definition states that the action shall be pursued without delay in a controlled manner.

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Thus, the urgency of completing the action is communicated and the emphasis on safe operation maintained.

The second change is to add a note to TS 3.1 O.e and TS 3.1 O.f to allow for a soak time of up to one hour before the position requirements take affect after significant rod motion.

When the control rods have been moved significantly (2 10 steps in one direction in < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ), IRPIs will occasionally differ from the demand position indication by more than the TS allowed limits. With current TS, the operators recognize that rod alignment requirements may not be met and that IRPI operability requirements are not met.

An inoperable IRPI currently requires verifying the position of the rod indirectly by core instrumentation (excore detector and/or thermocouples and/or movable incore detectors) at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or reduce power to less than 50%

rated power (RP). When the operators observe an indication of rod misalignment, the operators first verify rod alignment.

Checking that reactor flux distribution is within normal steady-state values is the method used by the operator to perform rod alignment verification. To check reactor flux distribution the operators monitor the power range nuclear instruments, power range delta flux monitors, core exit thermocouples, and available incore flux map results. If this check determines the rod is misaligned, the requirements of TS 3.10.e, "Rod Misalignment Limitations," are entered. If the check determines the rod is properly aligned the IPRI is inoperable.

Subsequent to the approval of this amendment request, the operators will declare the IRPI inoperable and notify the Reactor Engineering Group to request a determination of the rod(s) position using the moveable incore detectors. They will also monitor the behavior of the IRPI(s) to determine if the rod position indication returns to within limits. Typically, within 30 minutes, the IRPI difference from the demand position returns to within limits. When the IRPI returns to within limits, the limiting condition for operation (LCO) is met before expiration of the completion time and completion of the required actions is not required. The operators then declare the IRPI operable, exit the TS item, and notify the Reactor Engineering Group that rod position verification will not be required. Since this scenario repeats each time there is substantial control rod movement and no substantive remedial actions are required, this is considered a "nuisance" LCO condition entry.

This LAR proposes to add a note which will allow one hour soak time during which individual IRPIs are not required to be within limits. This one-hour time period is based on the time deemed necessary to allow the control rod drive shaft to reach thermal equilibrium. With this note, unnecessary entry into an LCO condition will be avoided following substantial movement of control rods.

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Operators are required to be aware of plant conditions and to meet the applicable TS. With these notes, the operators will still be required to monitor the IRPIs and take the appropriate actions after one hour if the IRPI and demand position indication continue to differ by more than the TS allowed limits. Thus, the practical effect of the proposed notes is to extend by one hour (from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months ) the time that individual IRPI instrumentation could be inoperable or a rod could be misaligned before TS required remedial actions would be implemented.

The likelihood of having a misaligned control rod and an event sensitive to a significantly misaligned rod is small. Since IRPI instrumentation is seldom inoperable, the possibility of inoperable IRPI instrumentation coincident with a misaligned control rod and an event sensitive to a significantly misaligned rod is even more remote. Thus, a one-hour time extension does not introduce significant safety issues with respect to IRPI operability. There may also be other IRPI instrumentation indications available to the operators that an IRPI may be inoperable, such as erratic indications, which may be more meaningful than a difference between the IRPI and demand position indication. These other indications could appear independent of rod motion.

There is a small possibility that an IRPI difference from the demand position indication is a legitimate indication of a misaligned rod. It is the expectation in the use of the KNPP TS that differences between IRPI and demand position indication will be evaluated first as IRPI instrumentation inoperability, therefore, a rod could be misaligned for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to identification as a misaligned rod.

This proposed change would extend the time by an additional hour. The likelihood of an event sensitive to a significantly misaligned rod during this one-hour time period is low.

The plant is analyzed for control rod misalignment. The minimum misalignment for individual rods assumed in the safety analyses is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to full insertion is assumed.

Plant safety analyses consider two types of rod misalignment events, static misalignment and a dropped rod. The analyses show that a single dropped rod event, without any operator intervention, does not result in any fuel pin failure, therefore the rod drop event is not time dependent and an additional hour with the misalignment undetected and unmitigated does not adversely impact plant safety. Multiple rod drop events cause the reactor to trip and therefore, an additional hour would not have any impact on this event.

In the static misalignment event, one or more control rods are assumed statically misplaced from the allowed position. This situation might occur if a rod were left behind when inserting or withdrawing banks, or if a single rod were to be withdrawn. Modeling the most limiting configuration bound the analysis of this event, which are the control banks at the full power insertion limit except for a single control rod fully withdrawn. The analyses show that, without any operator intervention, a single fully withdrawn rod event does not result in any fuel pin Page 6 of 16

failure, therefore the static rod misalignment event is not time dependent and an additional hour with the misalignment undetected and unmitigated does not adversely impact plant safety. Multiple rod misalignment events are bounded by the single rod misalignment event analyses and therefore an additional hour would not have any impact on this event.

The proposed one-hour extension does not significantly affect the safety of the plant. Allowing the operator and engineers to focus on monitoring the reactor without unnecessary entry into an LCO condition and required actions may enhance plant safety and reliability of plant equipment. With these proposed changes, the KNPP TS will continue to protect the health and safety of the public.

The third change is to reword TS 3.1O.f.1 to follow the changes of the other additions to the TS.

The current KNPP TS 3.10.f.1 contains requirements indicating that the TS is applicable during operation between 50% and 100% rated power, states the frequency of verifying the rods position with inoperable IRPI, and states that NMC must verify the position of the rod if it is moved by 24 steps or more. These requirements have been changed to be consistent with ISTS and moved to other locations within TS 3.10.f.

The fourth change is to modify the current method of determining rod position to require verification of rod position by movable incore detectors.

This change modifies the current licensing basis of the KNPP. Current KNPP TS states that if an IRPI is inoperable the position of the rod cluster control shall be checked indirectly by core instrumentation (excore detector and/or thermocouples and/or movable incore detectors). NMC is requesting to modify the position verification such that the determination of rod position will be performed using the moveable incore detectors. If the rod(s) are moved by more than 24 steps, TS 3.10.f.3 is applied and the rod(s) position is once again verified by movable incore detectors. This provides more restrictive requirements than the current KNPP TS.

The fifth change is to add a note to TS 3.10.f.1 and TS 3.1O.f.4 to allow for separate entry conditions for each IRPI or demand position indicator found to be inoperable This is acceptable because the required actions for each condition provide appropriate compensatory actions for an inoperable position indicator. Also, this requirement is consistent with ISTS.

The sixth change is to add TS 3.1 O.f.2 to allow for more than one rod per group to be out of service for a maximum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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This change is consistent with NEI TSTF 234, which allows verification of core peaking factors and shutdown margin (SDM) to satisfy the action requirements, providing the non-indicating rods have not been moved. The additional time to restore an inoperable IRPI is appropriate because the proposed Action would require that the control rods be under manual control, that RCS Tavg be monitored and recorded hourly, and that rod position be verified indirectly every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, thereby assuring that the rod alignment and rod insertion LCOs are met. Therefore, the required shutdown margin will be maintained. Given the alternate position monitoring requirement, and other indirect means of monitoring changes in rod position (e.g., alarms on Tavg - Tref deviation), a 24-hour completion time to restore all but one IRPI per group provides sufficient time to restore operability while minimizing shutdown transients during the time that the position indication system is degraded.

The seventh change is to add TS 3.1O.f.3 for requirements if a rod with an inoperable IRPI is moved in excess of 24 steps in one direction since its position was last determined.

These required actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the required actions of the inoperable position indicators, as applicable, are still appropriate but must be initiated to begin verifying that these rods are still properly positioned, relative to their group positions. If, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the rod positions have not been determined, thermal power must be reduced to < 50% rated power (RP) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that could result from continued operation at

> 50% RP, if one or more rods are misaligned by more than 24 steps. The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period to verify the rod positions.

The eighth change is to add TS 3.10.f.4 requiring the demand position indictors to be operable and stating action if the indicators are inoperable.

This addition places requirements that are more restrictive on KNPP operation than those required by the current KNPP TS. With one demand position indicator per bank inoperable, the IRPI System can determine the rod positions.

Since normal power operation does not require excessive movement of rods, verification by administrative means that the individual rod position indicators are operable and the most withdrawn rod and the least withdrawn rod are < 12 steps apart within the allowed completion time of once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is adequate.

Reduction of thermal power to < 50% RP puts the core into a condition where rod position is not significantly affecting core peaking factor limits. The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period of time to verify the rod positions or reduce power to < 50% RP.

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In summary, the requirements that are modified for the rod position indication system ensure that during normal power operation and abnormal anticipated occurrences the position of the rods is known. Knowing the rod's position ensures rod's alignment with their respective banks to provide consistency with the assumption of the safety analyses, maintaining symmetric neutron flux and power distribution profiles, providing assurance that peaking factors are within acceptable limits and assuring adequate shutdown margin.

5.0 VARIATIONS FROM NUREG 1431 NMC has made the following deviations from the guidance found in NUREG 1431:

1. Addition of a note allowing a 1-hour soak following substantial movement of control rods prior to invoking the rod position limit requirements (previously explained).

2. NUREG 1431, Revision 2, LCO 3.1.7, required action C.1, has a completion time of [4] hours. NMC changed that completion time for KNPP to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . (See below for explanation)

3. A second note was added to TS 3.1 0.f that allows a substitution for verification of rod position by movable incore detectors. This note allows substitution of verification of rod position by ensuring that Fa satisfies TS 3.10.b.1.A (FoN(Z)),

TS 3.10.b.5 (FOEo), FAHN satisfies TS 3.10.b.1.B, and SHUTDOWN MARGIN satisfies TS 3.1 0.a. (See below for explanation)

The second variation is to change an ISTS bracketed completion time and insert a KNPP specific completion time. NMC requests to require the verification of the rods after movement of > 24 steps be performed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . ISTS LCO 3.1.7, required action C.1, defines the requirements for when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction. This requirement is contained in proposed TS item 3.10.f.3. These required actions clarify that when one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the required actions of verifying rod position, as applicable are still appropriate but must be initiated promptly to begin verifying that these rods are still properly positioned, relative to their group positions. If, within [4] hours, the rod positions have not been determined, thermal power must be reduced to

50% RP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that could result from continued operation at > 50% RP, if one or more rods are misaligned by more than 24 steps. The allowed completion time of [4] hours provides an acceptable period of time to verify the rod positions.

The NEI guidance for ISTS13] implementation contains information on incorporating specifications with brackets [ ]. Typically, brackets are used in the generic Technical Specifications and Bases to indicate where plant specific input is needed. Various types of material may be bracketed, such as values, system or component names, or Page 9 of 16

descriptive material. Brackets may contain information from either the leadplant or typical vendor information. Plant-specific Improved Technical Specifications contain the appropriate value for the particular plant in lieu of brackets. As the completion time for verifying the rod position is bracketed, NMC requests a completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

This time is consistent with the completion time for the other specifications requiring verification of rod position. Additionally, due to the location of the plant and operation of the equipment, if rod motion were to occur during off-normal hours completion of verification of rod position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months could not occur.

The third deviation from ISTS of allowing substitution of verification of rod position by ensuring that F0 satisfies TS 3.1 0.b.1.A (FQN(Z)), TS 3.1 0.b.5 (FOEO), FAHN satisfies TS 3.10.b.1.B, and SHUTDOWN MARGIN satisfies TS 3.10.a, is contained in the ISTS basis for rod position]. Instead of this allowance residing the TS basis, NMC requests the allowance be placed as a note in the TS. As the purpose of determining rod position is to ensure these parameters are satisfied, verifying these parameters are satisfied meets the intent of verifying rod position.

6.0 REGULATORY SAFETY ANALYSIS 6.1 No Significant Hazards Consideration The Nuclear Management Company, LLC (NMC), is requesting to amend Operating License DPR-43 for the Kewaunee Nuclear Power Plant (KNPP). The proposed changes would revise the Operating License to add requirements to the Technical Specification (TS) associated with the axial position indication systems of shutdown and control rods.

Current KNPP TS allows for one IRPI per group or two per bank, to be out of service indefinitely. NMC is requesting to add new requirements to the KNPP TS including: 1) add a definition to TS 1.0 to define the term "Immediately," 2) add a note to TS 3.1 O.e and TS 3.1 O.f to allow for a soak time of up to one hour before the position requirements take affect after a significant rod motion, 3) reword TS 3.1 0.f.1 to follow the format of the other additions to the TS, 4) modify the current method of determining rod position to require verification of rod position by movable incore detectors, 5) add a note to TS 3.1O.f.1 and TS 3.1 O.f.4 to allow for separate entry conditions for each IRPI or demand position indicator inoperable, 6) add TS 3.1 Q.f.2 to allow for more than one rod per group to be out of service for a maximum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 7) add TS 3.1 0.f.3 to add requirements if a rod with one inoperable IRPI is moved in excess of 24 steps in one direction since its position was last determined, and 8) add TS 3.1 0.f.4 requiring the demand position indicators to be operable and stating action if the indicators are inoperable.

Control and shutdown rod position accuracy is essential during power operation.

Power peaking, ejected rod worth, or shutdown margin (SDM) limits may be violated in the event of a Design Basis Accident (DBA), with control or shutdown Page 10 of 16

rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM. Rod positions are continuously monitored by two separate systems (individual rod position and demand rod position) to provide operators with information that ensures the plant is operating within the bounds of the accident analysis assumptions.

The Nuclear Management Company has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Rod position indication instrumentation is not an assumed accident initiator, providing indication only of the control and shutdown rods position. Normal operation, abnormal occurrences and accident analyses assume the rods are at certain positions within the reactor core. The changes requested herein modify the time the existing two rod position indication systems may be inoperable and provide appropriate actions to compensate for that inoperability and add the second, digital, rod position indication system to the TS. Thus, this change does not involve a significant increase in the probability of an accident.

The condition of concern is the alignment of the rods. Operating with a rod position indicator inoperable does not change the position of the rod; an inoperable rod position indication instrument does not make a rod misaligned.

An increase in the consequences with the rods only comes from a rod being misaligned such that an increase in the heat produced in a localized area causes the fuel to fail either during operation, during a plant transient or post-accident.

An inoperable rod position indicator does not change the position of the rod. Rod position is subsequently verified by other means if the rod is moved by greater than a predetermined amount. Indication of rod position by other means ensures rod position remains within analytical limits. Thus, inoperable rod position indication instrumentation does not involve an increase in the consequences of an accident.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Page 11 of 16

Response: No.

This proposed change does not alter the design, function, or operation of any plant component and does not install any new or different equipment. The malfunction of safety related equipment, assumed operable in the accident analyses, would not be caused because of the proposed technical specification change. No new failure mode has been created and no new equipment performance burdens are imposed. Therefore, the possibility of a new or different kind of accident from those previously analyzed has not been created.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The rod position indication system is an instrumentation system that provides indication to the operators that a control rod may be misaligned. Inoperable individual rod position indication instrumentation does not by itself harm or affect reactor operation, but may impair the ability of the operators to detect a misaligned rod. To compensate for this potential impairment of the operators ability to detect a misaligned rod, requirements to verify the inoperable rod position indicators position are added. The impact of inoperable rod position indication instrumentation is offset by the availability of other indications that a rod is misaligned. Excore and incore nuclear instrumentation provides indication that reactor power, flux density, may have shifted axially or radially. Also, thermocouple indication would show that the core temperatures have increased in one region of the core and/or decreased in another region of the core.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, NMC concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

6.2 Applicable Regulatory Requirements/Criteria The US Atomic Energy Commission (AEC) issued their Safety Evaluation (SE) of the Kewaunee Nuclear Power Plant on July 24, 1972 with supplements dated December 18,1972 and May 10, 1973. In the AEC's SE, section 3.1, "Conformance with AEC General Design Criteria," described the conclusions the Page 12 of 16

AEC reached associated with the General Design Criteria in effect at the time.

The AEC stated:

The Kewaunee plant was designed and constructed to meet the intent of the AEC's General Design Criteria, as originally proposed in July 1967. Construction of the plant was about 50% complete and the Final SafetyAnalysis Report (Amendment No. 7) had been filed with the Commission before publication of the revised General Design Criteria in February 1971 and the present version of the criteria in July 1971. As a result, we did not require the applicant to reanalyze the plant or resubmit the FSAR. However, our technical review did assess the plant against the General Design Criteria now in effect and we are satisfied that the plant design generally conforms to the intent of these criteria.

As such, the appropriate 10 CFR 50 Appendix A General Design Criteria are listed below with the associated criteria KNPP is licensed to from the Final Safety Analysis (Amendment 7), which has been updated and now titled the Updated Safety Analysis Report (USAR). Below are listed the applicable 10 CFR Part 50, Appendix A, General Design Criteria (GDC) with the associated information for the KNPP USAR described afterward.

10 CFR 50 Appendix A, "General Design Criteria," lists the appropriate general design criteria (GDC) as 'Criterion 13-Instrumentation and control." This GDC is worded the same as that when the NRC reviewed KNPP against as stated in the Federal Register, Volume 36, NO. 130, dated July 7, 1971, page 12733. 10 CFR 50, GDC 13 states:

Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

KNPP USAR section 1.8 describes the applicable design criteria, which KNPP was designed to meet. For the rod position indication systems, the applicable design criterion is Criterion 13, "Fission Process Monitors and Controls. USAR section 1.8, criterion 13 states:

Page 13 of 16

Means shall be provided for monitoring and maintaining control over the fission process throughout core life and for all conditions that can reasonably be anticipated to cause variations in reactivity of the core, such as indication of position of control rods and concentration of soluble reactivity control poisons.

Additionally, NMC reviewed NUREG 0800, "Standard Review Plan," (SRP) for other applicable acceptance criteria, although KNPP is a pre-SRP plant. NUREG 0800, Section 7.0. Instrumentation and Controls -

Overview of Review Process contains the criteria associated with the rod position instrumentation systems.

This review revealed three sectionst51 that may apply to this amendment request, section 7.5, "Information Systems Important to Safety," section 7.6, "Interlock Systems Important to Safety," and section 7.7, "Control Systems."

NUREG 0800, Section 7.5, stated that the systems reviewed using this section include the Post-Accident monitoring system, bypassed or inoperable status indications, plant annunciator system, safety parameter display system, and emergency response facilities information systems and the nuclear data link. As this request does not affect these systems, this section was determined to be not applicable.

Section 7.6 stated that the systems applicable to this section are those systems that include interlock systems to prevent overpressurization of low-pressure systems (for example, residual heat removal (RHR)) when these systems are connected to high-pressure systems (for example, primary coolant), interlocks to prevent overpressure of the primary coolant system during low-temperature operation of the reactor vessel, valve interlocks to ensure the availability of emergency core cooling system (ECCS) accumulators, interlocks to isolate safety systems from non-safety systems (for example, seismic and non-seismic portions of auxiliary supporting systems), and interlocks to preclude inadvertent inter-ties between redundant or diverse safety systems where such inter-ties exist for the purposes of testing or maintenance. As the rod position indication systems do not contain any of these interlocks, this section is not applicable.

Section 7.7, Table 7.7-1, contains examples of control systems typically included in section 7.7. Rod position instrumentation is one of those listed systems.

Therefore, NUREG 0800, section 7.7 was reviewed for acceptance criteria.

On review of section 7.7 of NUREG 0800, NMC concluded that after the amendment to the KNPP TS associated with the rod position indication system, the rod position indication system still includes the necessary features for manual and automatic control of rods within prescribed normal operating limits. The plant accident analysis in Chapter 14 of the USAR does not rely on the operability of rod position indication system function to assure safety. The safety analysis does riot include consideration of the effects of both rod position Page 14 of 16

indication systems action and inaction in assessing the transient response of the plant for accidents and anticipated operational occurrences. Lastly, the failure of any rod position indication system component or any auxiliary supporting system for rod position indication system does not cause plant conditions more severe than those described in the analysis of anticipated operational occurrences in Chapter 14 of the USAR.

KNPP Updated Safety Analysis Report (USAR) Section 3.1.2, "Principle Design Criteria, "Reactor Core Design," states, "the reactor core, with its related controls and protective systems, shall be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits which have been stipulated and justified." Also, "the core and related auxiliary system designs shall provide this integrity under all expected conditions of normal operation with appropriate margins for uncertainties and for specified transient situations which can be anticipated." KNPP USAR Section 7.2.1 "Design Basis - Reactor Protection System," requires that "core protection systems, together with associated equipment, be designed to prevent or suppress conditions that could result in exceeding acceptable fuel damage limits." KNPP USAR Section 7.4.1, "Design Basis - Fission Process Monitors and Controls," requires that, "means shall be provided for monitoring or otherwise measuring and maintaining control over the fission process throughout core life under all conditions that can reasonably be anticipated to cause variations in reactivity of the core."

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Page 15 of 16

8.0 REFERENCES

['I USAR section 7.3.2

[2] SP-87-125, "Shift Instrument Channel Checks - Operating."

[3]

NEI 01-03, "Writer Guide for the Improved Standard Technical Specifications, November 2001

[4] NUREG 1431, Revision 2 basis, LCO 3.1.7, page 3.1.7-4 and page 3.1.7-5

[5] NUREG 0800, section 7.1, page SRP T7.1-3 Page 16 of 16

ENCLOSURE 2 NUCLEAR MANAGEMENT COMPANY, LLC, MARKED UP TS PAGES FOR LICENSE AMENDMENT REQUEST 203 TO KEWAUNEE NUCLEAR POWER PLANT, OPERATING LICENSE NO. DPR-43, DOCKET NO. 50-305 Marked Up TS Pages:

TS ii TS 1.0-6 TS 3.10-5 TS 3.10-6 4 Pages to Follow

Section Title Paqe 3.2 Chemical and Volume Control System 3.2-1 3.3 Engineered Safety Features and Auxiliary Systems 3.3-1 3.3.a Accumulators.......................

3.3-1 3.3.b Emergency Core Cooling System.......................

3.3-2 3.3.c Containment Cooling Systems.......................

3.3-4 3.3.d Component Cooling System.......................

3.3-6 3.3.e Service Water System.......................

3.3-7 3.4 Steam and Power Conversion System 3.4-1 3.4.a Main Steam Safety Valves............................

3.4-1 3.4.b Auxiliary Feedwater System............................

3.4-1 3.4.c Condensate Storage Tank............................

3.4-3 3.4.d Secondary Activity Limits............................

3.4-3 3.5 Instrumentation System 3.5-1 3.6 Containment System 3.6-1 3.7 Auxiliary Electrical Systems 3.7-1 3.8 Refueling Operations 3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits

............................. 3.10-1 3.1O.a Shutdown Reactivity...............................

3.10-1 3.10.b Power Distribution Limits................................

3.10-1 3.10.c Quadrant Power Tilt Limits................................

3.10-4 3.1O.d Rod Insertion Limits...............................

3.10-4 3.1O.e Rod Misalignment Limitations...............................

3.10-5 3.10.f Inoperable Rod Position Indicator Channels............................ 3.10-5 3.10.g Inoperable Rod Limitations.................................. 3.10-76 3.10.h Rod Drop Time.................................

3.10-Z6 3.10.i Rod Position Deviation Monitor...............................

3.1 O-Z6 3.10.j Quadrant Power Tilt Monitor.................................. 3.10-Z6 3.1 O.k Core Average Temperature................................. 3.1 0-Z6 3.10.1 Reactor Coolant System Pressure...............................

3.1 0-Z6 3.1 0.m Reactor Coolant Flow.................................. 3.1 0-F 3.10.n DNBR Parameters.................................

3.10-F 3.11 Core Surveillance Instrumentation 3.11-1 3.12 Control Room Post-Accident Recirculation System

........................ 3.12-1 3.14 Shock Suppressors (Snubbers) 3.14-1 4.0 Surveillance Requirements 4.0-1 4.1 Operational Safety Review

4.1-1 4.2 ASME Code Class In-service Inspection and Testing 4.2-1 4.2.a ASME Code Class 1, 2, 3, and MC Components and Supports

................ 4.2-1 4.2.b Steam Generator Tubes

................. 4.2-2 4.2.b.1 Steam Generator Sample Selection and Inspection.

4.2-3 4.2.b.2 Steam Generator Tube Sample Selection and Inspection...............

4.2-3 4.2.b.3 Inspection Frequency...............

4.2-4 4.2.b.4 Plugging Limit Criteria...............

4.2-5 4.2.b.5 Deleted 4.2.b.6 Deleted 4.2.b.7 Reports.

4.2-5 4.3 Deleted L202AreRdMentN46 TS ii 04104°2003

p. DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1:131 is that concentration of 1-131 (,uCi/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be as listed and calculated based on dose conversion factors derived from ICRP-30.

l DOSE CONVERSION FACTOR I

ISOTOPE 1.0000 I

1-131 0.0059 1-132 0.1692 1-133 0.0010 1-134 0.0293 1-135

q.

CORE OPERATING LIMITS REPORT (COLR)

The COLR is the unit specific document that provides cycle-specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 6.9.a.4. Plant operation within these limits is addressed in individual Specifications.

r.

SHUTDOWN MARGIN (SDM)

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

1. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.

However, with all RCCAs verified fully inserted by two independent means (TS 3.10.e), it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM, and

2. In the OPERATING and HOT STANDBY MODES, the fuel and moderator temperatures are changed to the nominal zero power design temperature.

S. IMMEDIATELY When "Immediately is used as a completion time in a LCO, the required action should be pursued without delay and in a controlled manner.

Amondmont No. 167LAfl203 04/04/2003 TS 1.0-6

e.

Rod Misalignment Limitations NOTE: Individual RPIs may be outside their limits for < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following substantial rod movement.

This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.1 0.e.1 and TS 3.1 O.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.

Rod misalignment limitations do not apply during physics testing.

1. When reactor power is >85% of rating, the rod cluster control assembly shall be maintained within +12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is 285%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.1O.b applied.

If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating.

2. When reactor power is < 85% but > 50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than

24 steps when reactor power is

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

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

f.

Inoperable Rod Position Indicator Channels NOTE: Individual RPls may be outside their limits for < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following substantial rod movement.

NOTE: Verification of rod position may be satisfied by ensuring that Fa satisfies TS 3.10.b.1.A WFN(ZY).

TS 3.10.b.5 (F

.N satisfies TS 3.10.b.1.B, and SHUTDOWN MARGIN satisfies TS 3.10.a.

1. If aOQne ndividul rod position indicator channelj~

pgr is eifewioperablef one or more groups, then perform either A or B below: (Note: Separate condition entry is allowed for each inoperable individual rod position indicator.)

A. For oporation botwoon 50%o and 100%5 of ratingVerify the position of the rod cluster control abjyshall bo chockod indiroctly! by coro inetrumontation (oxcoro detector andFor thermocouplos and/or movable incore detectors) each 8 hoummanmrst nc per eight hours, or subsequent to rod motion oxcooding a total displacement of 21 stops, whirhovor occur frct.or B. During operation < 60%0 of rating, no special monitoring is roquiredWithin 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months reduce reactor thermal power to < 50% of RATED POWER.

Amendment No. 165LAB-20 TS 3.10-5 03Q1/42003

2. Net-mlf more than oneidvdaod position indicator channel per group are ner orod pitionC i tor o-hanneoe par bank ohall beo pormItted to bo inoperable-at ian A. Immediately place the control rods in manual, and B. Once per 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . monitor and record RCS T V C. Verify the position of the rod by movable incore detectors each 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . and D. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months restore the inoperable individual rod position indicators to OPERABLE status such that a maximum of one IRPI per group is inoperable or place the plant in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3. If one or more rods with inoperable individual rod position indicators have been moved in excess of 24 steps in one direction since the last determination of the rods position then perform A or B below:

A. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months verify the position of the rod by movable incore detectors. or B. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months reduce reactor thermal power to < 50% of RATED POWER.

4. If one demand position indicator per bank for one or more banks is inoperable then perform either A or B below: (Note: Se-arate condition entry is allowed for each inoperable demand position indicator.)

A. Each 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months verify,

1) AII IRPI's for the affected banks are OPERABLE, by administrative means.

2) The most withdrawn rod and the least withdrawn rod of the affected bank(s) are

< 12 steps apart when > 85% RATED POWER or < 24 steps when < 85%

BATED POWER, B. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months reduce reactor thermal power to < 50% of RATED POWER.

l5, If a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned from TS 3.10.f4A, then TS 3.1O.e will be applied.

Amondmont No. 166LAR.203 TS 3.10-6 03/

/12003

ENCLOSURE 3 NUCLEAR MANAGEMENT COMPANY, LLC AFFECTED TS PAGES FOR LICENSE AMENDMENT REQUEST 203 TO KEWAUNEE NUCLEAR POWER PLANT, OPERATING LICENSE NO. DPR-43 DOCKET NO. 50-305 Affected TS Pages:

TS ii TS 1.0-6 TS 3.10-5 TS 3.10-6 4 Pages to Follow

Section Title Page 3.2 Chemical and Volume Control System 3.2-1 3.3 Engineered Safety Features and Auxiliary Systems; 3.3-1 3.3.a Accumulators.......................

3.3-1 3.3.b Emergency Core Cooling System.......................

3.3-2 3.3.c Containment Cooling Systems.......................

3.3-4 3.3.d Component Cooling System.......................

3.3-6 3.3.e Service Water System.......................

3.3-7 3.4 Steam and Power Conversion System 3.4-1 3.4.a Main Steam Safety Valves............................

3.4-1 3.4.b Auxiliary Feedwater System............................

3.4-1 3.4.c Condensate Storage Tank............................

3.4-3 3.4.d Secondary Activity Limits............................

3.4-3 3.5 Instrumentation System 3.5-1 3.6 Containment System 3.6-1 3.7 Auxiliary Electrical Systems 3.7-1 3.8 Refueling Operations 3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits

............................. 3.10-1 3.10.a Shutdown Reactivity................................

3.10-1 3.1 0.b Power Distribution Limits...............................

3.10-1 3.10.c Quadrant Power Tilt Limits...............................

3.10-4 3.10.d Rod Insertion Limits................................

3.10-4 3.1 0.e Rod Misalignment Limitations...............................

3.10-5 3.1O.f Inoperable Rod Position Indicator Channels............................ 3.10-5 3.10.g Inoperable Rod Limitations................................

3.10-7 3.10.h Rod Drop Time................................

3.10-7 3.10.i Rod Position Deviation Monitor...............................

3.10-7 3.10.j Quadrant Power Tilt Monitor................................

3.10-7 3.10.k Core Average Temperature...............................

3.10-7 3.10.1 Reactor Coolant System Pressure...............................

3.10-7 3.10.m Reactor Coolant Flow...............................

3.10-8 3.10.n DNBR Parameters...............................

3.10-8 3.11 Core Surveillance Instrumentation 3.11-1 3.12 Control Room Post-Accident Recirculation System

........................ 3.12-1 3.14 Shock Suppressors (Snubbers) 3.14-1 4.0 Surveillance Requirements 4.0-1 4.1 Operational Safety Review 4.1-1 4.2 ASME Code Class In-service Inspection and Testing 4.2-1 4.2.a ASME Code Class 1,2, 3, and MC Components and Supports

................ 4.2-1 4.2.b Steam Generator Tubes

................. 4.2-2 4.2.b.1 Steam Generator Sample Selection and Inspection.

4.2-3 4.2.b.2 Steam Generator Tube Sample Selection and Inspection...............

4.2-3 4.2.b.3 Inspection Frequency...............

4.2-4 4.2.b.4 Plugging Limit Criteria...............

4.2-5 4.2.b.5 Deleted 4.2.b.6 Deleted 4.2.b.7 Reports.

4.2-5 4.3 Deleted LAR 203 TSii

p. DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 is that concentration of 1-131 (,uCVgram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be as listed and calculated based on dose conversion factors derived from ICRP-30.

E DOSE CONVERSION FACTOR ISOTOPE 1.0000 1-131 0.0059 1-132 0.1692 1-133 0.0010 1-134 0.0293 1-135

q. CORE OPERATING LIMITS REPORT (COLR)

The COLR is the unit specific document that provides cycle-specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 6.9.a.4. Plant operation within these limits is addressed in individual Specifications.

r. SHUTDOWN MARGIN (SDM)

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

1. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn.

However, with all RCCAs verified fully inserted by two independent means (TS 3.10.e), it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM, and

2. In the OPERATING and HOT STANDBY MODES, the fuel and moderator temperatures are changed to the nominal zero power design temperature.

s. IMMEDIATELY When Immediately" is used as a completion time in a LCO, the required action should be pursued without delay and in a controlled manner.

LAR 203 TS 1.0-6l

e.

Rod Misalignment Limitations NOTE: Individual RPIs may be outside their limits for < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following substantial rod movement.

This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.1 O.e.1 and TS 3.1 0.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs. voltage.

Rod misalignment limitations do not apply during physics testing.

1. When reactor power is >85% of rating, the rod cluster control assembly shall be maintained within +/- 12 steps from their respective banks.

If a rod cluster control assembly is misaligned from its bank by more than + 12 steps when reactor power is 285%, then the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied.

If peaking factors are not determined within four hours, the reactor power shall be reduced to < 85% of rating.

2. When reactor power is < 85% but 2 50% of rating, the rod cluster control assemblies shall be maintained within +/-24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is

< 85% but 2 50%, the rod will be realigned or the core power peaking factors shall be determined within four hours, and TS 3.10.b applied. If the peaking factors are not determined within four hours, the reactor power shall be reduced to < 50% of rating.

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

f.

Inoperable Rod Position Indicator Channels NOTE: Individual RPIs may be outside their limits for < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following substantial rod movement.

NOTE: Verification of rod position may be satisfied by ensuring that F0 satisfies TS 3.10.b.1.A (F0N(Z)), TS 3.10.b.5 (FOEO), FAJHN satisfies TS 3.10.b.1.B, and SHUTDOWN MARGIN satisfies TS 3.10.a.

1. If one individual rod position indicator channel per group is inoperable for one or more groups, then perform either A or B below: (Note: Separate condition entry is allowed for each inoperable individual rod position indicator.)

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

LAR 203 TS 3.10-5l

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

A. Immediately place the control rods in manual, and B. Once per 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , monitor and record RCS Tavg, and C. Verify the position of the rod by movable incore detectors each 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and D. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months restore the inoperable individual rod position indicators to OPERABLE status such that a maximum of one IRPI per group is inoperable or place the plant in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3. If one or more rods with inoperable individual rod position indicators have been moved in excess of 24 steps in one direction since the last determination of the rods position then perform A or B below:

A. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months verify the position of the rod by movable incore detectors, or B. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months reduce reactor thermal power to

50% of RATED POWER.

4. If one demand position indicator per bank for one or more banks is inoperable then perform either A or B below: (Note: Separate condition entry is allowed for each inoperable demand position indicator.)

A. Each 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months verify,

1) All IRPI's for the affected banks are OPERABLE, by administrative means.

2) The most withdrawn rod and the least withdrawn rod of the affected bank(s) are

< 12 steps apart when > 85% RATED POWER or < 24 steps when < 85%

RATED POWER.

B. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months reduce reactor thermal power to < 50% of RATED POWER.

5. If a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned from TS 3.1 0.f, then TS 3.1 0.e will be applied.

LAR 203 TS 3.1 0-6

ENCLOSURE 4 NUCLEAR MANAGEMENT COMPANY, LLC MARKED UP TS BASIS PAGES FOR LICENSE AMENDMENT REQUEST 203 TO KEWAUNEE NUCLEAR POWER PLANT, OPERATING LICENSE NO. DPR-43 DOCKET NO. 50-305 Marked Up TS Basis Pages TS B3.10-5 TS 63.10-6 TS B3.10-7 TS B3.10-8 4 Pages to Follow

The time limits of six hours to achieve HOT STANDBY and an additional six hours to achieve HOT SHUTDOWN allow for a safe and orderly shutdown sequence and are consistent with most of the remainder of the Technical Specifications.

Rod Misalignment Limitations (TS 3.1 0.e)

During normal power operation it is desirable to maintain the rods in alignment with their respective banks to provide consistency with the assumption of the safety analyses, to maintain symmetric neutron flux and power distribution profiles, to provide assurance that peaking factors are within acceptable limits and to assure adequate shutdown margin.

Analyses have been performed which indicate that the above objectives will be met if the rods are aligned within the limits of TS 3.1 O.e. A relaxation in those limits for power levels < 85% is allowable because of the increased margin in peaking factors and available shutdown margin obtained while OPERATING at lower power levels. This increased flexibility is desirable to account for the nonlinearity inherent in the rod position indication system and for the effects of temperature and power as seen on the rod position indication system.

Rod position measurement is performed through the effects of the rod drive shaft metal on the output voltage of a series of vertically stacked coils located above the head of the reactor pressure vessel. The rod position can be determined by the analog individual rod position indicators (IRPI),

the plant process computer which receives a voltage input from the conditioning module, or through the conditioning module output voltage via a correlation of rod position vs. voltage.

The plant process computer converts the output voltage signal from each IRPI conditioning module to an equivalent position (in steps) through a curve fitting process, which may include the latest actual voltage-to-position rod calibration curve.

The rod position as determined by any of these methods can then be compared to the bank demand position which is indicated on the group step counters to determine the existence and magnitude of a rod misalignment. This comparison is performed automatically by the plant process computer.

The rod deviation monitor on the annunciator panel is activated (or reactivated) if the two position signals for any rod as detected by the process computer deviate by more than a predetermined value. The value of this setpoint is set to warn the operator when the Technical Specification limits are exceeded.

The rod position indicator system is calibrated once per REFUELING cycle and forms the basis of the correlation of rod position vs. voltage.

This calibration is typically performed at HOT SHUTDOWN conditions prior to initial operations for that cycle. Upon reaching full power conditions and verifying that the rods are aligned with their respective banks, the rod position indication may be adjusted to compensate for the effects of the power ascension. After this adjustment is performed, the calibration of the rod position indicator channel is checked at an intermediate and low level to confirm that the calibration is not adversely affected by the adjustment.

A note indicating individual control rod position indications may not be within limits for up to and including one hour following substantial control rod movement modifies this LCO. This allows up to one hour of thermal soak time to allow the control rod drive shaft to reach thermal equilibrium and thus present a consistent position indication. Substantial rod movement is considered to be 10 or more steps in one direction in less than or equal to one hour Amendment No. 167%AR 203 TS B3.10-5 04/04/2003

Inoperable Rod Position Indicator Channels (TS 3.1 0.f)

The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position Indication System (commonly called group step counters) and the Individual Rod Position Indication (IRPM System.

The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each aroup of rods. Individual rods in a group all receive the same signal to move and should. therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise (+/- 1 step or +/- 5/8 inch). If a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the The IRPI System provides an indirect indication of actual control rod position, but at a lower precision than the step counters. The rod position indicator channel is sufficiently accurate to detect a rod +/- 12 steps away from its demand position. If the rod position indicator channel is not OPERABLE, special surveillance of core power tilt indications, using established procedures and relying on movable incore detectors, will be used to verify power distribution symmetry.

A note indicating individual control rod position indications may not be within limits for up to and including one hour following substantial control rod movement modifies this LCO. This allows up to one hour of thermal soak time to allow the control rod drive shaft to reach thermal equilibrium and thus present a consistent position indication. Substantial rod movement is considered to be 10 or more steps in one direction in less than or equal to one hour.

A second note indicates that the required action of verifying rod position by core instruments may also be satisfied by ensuring that Fo satisfies TS 3.10.b.1.A (F 1 (Zfl. TS 3.10.b.5 (FURL FAHN satisfies TS 3.1 0.b.1.B, and SHUTDOWN MARGIN satisfies TS 3.1 0.a. Limits may be violated with control or shutdown rods operating outside their limits. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the COLR limits.

When one IRPI channel per aroup fails, the position of the rod may be determined indirectly by use of the movable incore detectors. The required action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that Fo satisfies TS 3.1 0.b.1.A (FN(Z)), TS 3.1 0.b.5 F

TS 3.1 0.b.1.B, and SHUTDOWN MARGIN satisfies TS 3.10.a, provided the non-indicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks. If a bank has been significantly moved (> 24 steps), the required action of TS 3.1 0.f.3 is required. Therefore, verification of RCCA position within the completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. A reduction of reactor thermal power to < 50% RATED POWER puts the core into a condition where COLR limits are sufficiently relaxed such that rod position will not cause the core to violate COLR limits2. The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable. based on operating experience, for reducing power to < 50% RATED POWER from full power conditions without challenging plant systems and allowing for rod position determination by movable incore detectors.

2 USAR Chanter 14 Amendment No. 167LAR203 TS B3.10-6 04/04/2003

0.101f2 When more than one IRPI per group fail, additional actions are necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion with not occur. Toaether with the indirect position determination available via movable incore detectors will minimize the potential for rod misalignment. The immediate completion time for placing the Rod Control System in manual reflects the uraency with which unplanned rod motion must be prevented while in this condition. Monitoring and recording reactor coolant Tavg helps assure that significant changes in power distribution and SDM are avoided. The once per hour completion time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions. The position of the rods may be determined indirectly by use of the movable incore detectors. The required action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that Fosatisfies TS 3.1 0.b.1.A (FpN(Z)), TS 3.1 0.b.5 (FnEQ). FAN satisfies TS 3.1 0.b.1.B, and SHUTDOWN MARGIN satisfies TS 3.10.a, provided the non-indicating rods have not been moved. Verification of control rod position once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation for a limited, 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. The 24-hour completion time provides sufficient time to troubleshoot and restore the IRPI system to operation while avoiding the plant challenges associated with the shutdown without full rod position indication.

Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly moved. When one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction. since the position was last determined, the required actions of one or more inoperable individual rod position indicators, as applicable, are still appropriate but must be initiated under TS 3.10.f.3 to begin verifying that these rods are still properly positioned, relative to their aroup positions. If, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the rod positions have not been determined, thermal power must be reduced to

<50% RATED POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that could result from continued operation at > 50% RATED POWER, if one or more rods are misaligned by more than 24 steps. The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period of time to verify the rod positions.

With one demand position indicator per bank inoperable, the IRPI System can determine the rod positions. Since normal power operation does not require excessiye movement of rods, verification by administrative means (logging IRPI position and verifying within rod alignment limitations) that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are < 12 steps apart when operating at > 85% RATED POWER or < 24 steps apart when operating at < 85% RATED POWER within the allowed Completion Time of once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate. A reduction of reactor thermal power to <50% RATED POWER puts the core into a condition where COLR limits are sufficiently relaxed such that rod position will not cause the core to violate COLR limits. The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period of time to verify the rod positions or reduce power to < 50%

RATED POWER Amondmont no. 1 67LABP203 TS B3.10-7 04104/2003

The rod position indicator channol is sufficiently accurate to detect a rod +/- 12 stops away from its demand position. If the rod position indicator channel is not OPERABLE, then the operator will be fully awaro of the inoporability of the channel, and special suRvoillanco of core power tilt indications, using established proceduroe and relying on oxcoro nuclear dotoctors, and/or movable incoro dotoctors, will be used to vorify power distribution symmOtry.

Inoperable Rod Limitations (TS 3.10.ci)

One inoperable control rod is acceptable provided the potential consequences of accidents are not worse than the cases analyzed in the safety analysis report. A 30-day period is provided for the reanalysis of all accidents sensitive to the changed initial condition.

Rod Drop Time (TS 3.1 0.h)

The required drop time to dashpot entry is consistent with safety analysis.

Core Average Temperature (TS 3.10.k)

The core average temperature limit is consistent with full power operation within the nominal operational envelope. Either Tavg control board indicator readings or computer indications are averaged to obtain the value for comparison to the limit. The limit is based on the average of either 4 control board indicator readings or 4 computer indications. A higher Tavg will cause the reactor core to approach DNB limits.

Reactor Coolant System Pressure (TS 3.10.1)

The RCS pressure limit is consistent with operation within the nominal operational envelope. Either pressurizer pressure control board indicator readings or computer indications are averaged to obtain the value for comparison to the limit. The limit is based on the average of either 4 control board indicator readings or 4 computer indications. A lower pressure will cause the reactor core to approach DNB limits.

Reactor Coolant Flow (TS 3.1 0.m)

The reactor coolant system (RCS) flow limit, as specified in the COLR, is consistent with the minimum RCS flow limit assumed in the safety analysis adjusted by the measurement uncertainty.

The safety analysis assumes initial conditions for plant parameters within the normal steady state envelope. The limits placed on the RCS pressure, temperature, and flow ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the analyzed transients.

The RCS flow normally remains constant during an operational fuel cycle with all reactor coolant pumps running. At least two plant computer readouts from the loop RCS flow instrument channels are averaged per reactor coolant loop and the sum of the reactor coolant loop flows are compared to the limit. Operating within this limit will result in meeting the DNBR criterion in the event of a DNB-limited event.

Amendment No. 167LAR203 TS B3.10-8 04/04/2003

ENCLOSURE 5 NUCLEAR MANAGEMENT COMPANY, LLC AFFECTED TS BASIS PAGES FOR LICENSE AMENDMENT REQUEST 203 TO KEWAUNEE NUCLEAR POWER PLANT OPERATING LICENSE NO. DPR-43 DOCKET NO. 50-305 Affected TS Basis Pages TS B3.10-5 TS B3.10-6 TS B3.10-7 3 Pages to Follow

The time limits of six hours to achieve HOT STANDBY and an additional six hours to achieve HOT SHUTDOWN allow for a safe and orderly shutdown sequence and are consistent with most of the remainder of the Technical Specifications.

Rod Misalignment Limitations (TS 3.10.e)

During normal power operation it is desirable to maintain the rods in alignment with their respective banks to provide consistency with the assumption of the safety analyses, to maintain symmetric neutron flux and power distribution profiles, to provide assurance that peaking factors are within acceptable limits and to assure adequate shutdown margin.

Analyses have been performed which indicate that the above objectives will be met if the rods are aligned within the limits of TS 3.1 O.e. A relaxation in those limits for power levels < 85% is allowable because of the increased margin in peaking factors and available shutdown margin obtained while OPERATING at lower power levels. This increased flexibility is desirable to account for the nonlinearity inherent in the rod position indication system and for the effects of temperature and power as seen on the rod position indication system.

Rod position measurement is performed through the effects of the rod drive shaft metal on the output voltage of a series of vertically stacked coils located above the head of the reactor pressure vessel. The rod position can be determined by the analog individual rod position indicators (IRPI),

the plant process computer which receives a voltage input from the conditioning module, or through the conditioning module output voltage via a correlation of rod position vs. voltage.

The plant process computer converts the output voltage signal from each IRPI conditioning module to an equivalent position (in steps) through a curve fitting process, which may include the latest actual voltage-to-position rod calibration curve.

The rod position as determined by any of these methods can then be compared to the bank demand position which is indicated on the group step counters to determine the existence and magnitude of a rod misalignment. This comparison is performed automatically by the plant process computer.

The rod deviation monitor on the annunciator panel is activated (or reactivated) if the two position signals for any rod as detected by the process computer deviate by more than a predetermined value. The value of this setpoint is set to warn the operator when the Technical Specification limits are exceeded.

The rod position indicator system is calibrated once per REFUELING cycle and forms the basis of the correlation of rod position vs. voltage. This calibration is typically performed at HOT SHUTDOWN conditions prior to initial operations for that cycle. Upon reaching full power conditions and verifying that the rods are aligned with their respective banks, the rod position indication may be adjusted to compensate for the effects of the power ascension. After this adjustment is performed, the calibration of the rod position indicator channel is checked at an intermediate and low level to confirm that the calibration is not adversely affected by the adjustment.

A note indicating individual control rod position indications may not be within limits for up to and including one hour following substantial control rod movement modifies this LCO. This allows up to one hour of thermal soak time to allow the control rod drive shaft to reach thermal equilibrium and thus present a consistent position indication. Substantial rod movement is considered to be 10 or more steps in one direction in less than or equal to one hour LAR 203 l TS 63.10-5

InoDerable Rod Position Indicator Channels (TS 3.10.f)

The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position Indication System (commonly called group step counters) and the Individual Rod Position Indication (IRPI) System.

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

The IRPI System provides an indirect indication of actual control rod position, but at a lower precision than the step counters. The rod position indicator channel is sufficiently accurate to detect a rod +/- 12 steps away from its demand position. If the rod position indicator channel is not OPERABLE, special surveillance of core power tilt indications, using established procedures and relying on movable incore detectors, will be used to verify power distribution symmetry.

A note indicating individual control rod position indications may not be within limits for up to and including one hour following substantial control rod movement modifies this LCO. This allows up to one hour of thermal soak time to allow the control rod drive shaft to reach thermal equilibrium and thus present a consistent position indication. Substantial rod movement is considered to be 10 or more steps in one direction in less than or equal to one hour.

A second note indicates that the required action of verifying rod position by core instruments may also be satisfied by ensuring that Fa satisfies TS 3.1 0.b.1.A (FaN(Z)), TS 3.1 O.b.5 (FQEQ),

FAHNsatisfies TS 3.10.b.1.B, and SHUTDOWN MARGIN satisfies TS 3.10.a. Limits may be violated with control or shutdown rods operating outside their limits. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the COLR limits.

3.1 0.f.1 When one IRPI channel per group fails, the position of the rod may be determined indirectly by use of the movable incore detectors. The required action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that Fosatisfies TS 3.10.b.1.A (FON(Z)), TS 3.10.b.5 (F0 EO), FMN satisfies TS 3.1 0.b.11.B, and SHUTDOWN MARGIN satisfies TS 3.1 0.a, provided the non-indicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks. If a bank has been significantly moved (2 24 steps), the required action of TS 3.1 0.f.3 is required. Therefore, verification of RCCA position within the completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. A reduction of reactor thermal power to

50% RATED POWER puts the core into a condition where COLR limits are sufficiently relaxed such that rod position will not cause the core to violate COLR limits2. The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on operating experience, for reducing power to

50% RATED POWER from full power conditions without challenging plant systems and allowing for rod position determination by movable incore detectors.

2 USAR Chapter 14 LAR 203 TS B3.10-6l

3.1 0.f.2 When more than one IRPI per group fail, additional actions are necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion with not occur. Together with the indirect position determination available via movable incore detectors will minimize the potential for rod misalignment. The immediate completion time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this condition. Monitoring and recording reactor coolant Tavg helps assure that significant changes in power distribution and SDM are avoided. The once per hour completion time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions. The position of the rods may be determined indirectly by use of the movable incore detectors. The required action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that Fo satisfies TS 3.1 0.b.1.A (FoN(Z)), TS 3.1 0.b.5 (F0 EQ), FMN satisfies TS 3.10.b.11.B, and SHUTDOWN MARGIN satisfies TS 3.10.a, provided the non-indicating rods have not been moved. Verification of control rod position once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation for a limited, 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. The 24-hour completion time provides sufficient time to troubleshoot and restore the IRPI system to operation while avoiding the plant challenges associated with the shutdown without full rod position indication.

3.1 0.f.3 Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly moved. When one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction, since the position was last determined, the required actions of one or more inoperable individual rod position indicators, as applicable, are still appropriate but must be initiated under TS 3.1 0.f.3 to begin verifying that these rods are still properly positioned, relative to their group positions. If, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the rod positions have not been determined, thermal power must be reduced to

50% RATED POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that could result from continued operation at > 50% RATED POWER, if one or more rods are misaligned by more than 24 steps. The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period of time to verify the rod positions.

3.1 0.f.4 With one demand position indicator per bank inoperable, the IRPI System can determine the rod positions. Since normal power operation does not require excessive movement of rods, verification by administrative means (logging IRPI position and verifying within rod alignment limitations) that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are S 12 steps apart when operating at > 85% RATED POWER or S 24 steps apart when operating at

85% RATED POWER within the allowed Completion Time of once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate. A reduction of reactor thermal power to S 50% RATED POWER puts the core into a condition where COLR limits are sufficiently relaxed such that rod position will not cause the core to violate COLR limits. The allowed completion time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period of time to verify the rod positions or reduce power to

50%

RATED POWER Inoperable Rod Limitations (TS 3.10.q)

LAR 203 TS 83.1 0-7l

ML041620503

Contents

Text

References:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

7.0 ENVIRONMENTAL CONSIDERATION

8.0 REFERENCES

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ML041620503

Text

References:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

7.0 ENVIRONMENTAL CONSIDERATION

8.0 REFERENCES

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