ML042160302

Bases Revision(S) to the Kewaunee Nuclear Power Plant Technical Specifications

ML042160302
Person / Time
Site:	Kewaunee
Issue date:	07/19/2004
From:	Coutu T Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-04-082
Download: ML042160302 (9)

Text

Committed to Nuclear Excellenc Kewaunee Nuclear Power Plant Operated by Nuclear Management Company, LLC July 19, 2004 NRC-04-082 TS 6.21 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Kewaunee Nuclear Power Plant Docket 50-305 License No. DPR-43 Bases Revision(s) to the Kewaunee Nuclear Power Plant Technical Specifications Nuclear Management Company (NMC), licensee for the Kewaunee Nuclear Power Plant (KNPP), hereby submits a revision to the Basis for the Technical Specifications (TS) 3.10, "Control Rod and Power Distribution Limits". The basis for TS 3.1 0.b.6 was expanded to explain the options available when the FQEQ margin to the limit decreases with successive flux maps.

Additionally, this issue caused a change in the KNPP Core Operating Limits Report (COLR), which was submitted on July 12, 2004. This caused a change in the KNPP Technical Requirements Manual (TRM) Table of Contents, which is hereby submitted.

These changes have been screened for evaluation pursuant to the requirements of 10 CFR 50.59 in accordance with approved KNPP procedures and were determined to be acceptable.

Attached is a copy of the revised'Technical Specification Bases page(s) for your controlled TS.

Thomas Coutu Site Vice-President, Kewaunee Nuclear Power Plant Nuclear Management Company, LLC Enclosures cc: US NRC, Region IlIl US NRC, Senior Resident Inspector Electric Division, PSCW N490 Highway 42

• Kewaunee, Wisconsin 54216-9510 Telephone: 920.388.2560

KEWAUNEE NUCLEAR POWER PLANT TRM INDEX TECHNICAL REQUIREMENTS MANUAL Revision 4 l July 8, 2004 Table of Contents KEWAUNEE NUCLEAR POWER PLANT TECHNICAL REQUIREMENTS MANUAL Table of Contents SECTION TITLE Revision TRM 1.0 Technical Requirements Manual Content and Organization . 1 TRM 2.0 Reports TRM 2.1 Core Operating Limits Report (COLR) Cycle 26. 4 1 TRM 3.0 Use, Application, Limiting Condition for Operation (TLCO)

Applicability and Surveillance Requirement (TSR) Applicability TRM 4.0 Programs i

Fq!(Z). Height Dependent Nuclear Flux Hot Channel Factor F0 N(Z), Height Dependent Nuclear Flux Hot Channel Factor, is defined as the maximum local linear power density in the core at core elevation Z divided by the core average linear power density, assuming nominal fuel rod dimensions.

An upper bound envelope for F N(Z) as specified in the COLR has been determined from extensive analyses considering all OPERATING maneuvers consistent with the Technical Specifications on power distribution control as given in TS 3.10. The results of the loss-of-coolant accident analyses based on this upper bound envelope indicate the peak clad temperatures, with a high probability, remain less than the 22000 F limit.

The FQN(Z) limits as specified in the COLR are derived from the LOCA analyses. The LOCA analyses are performed for Westinghouse 422 V+ fuel, FRA-ANP heavy fuel and for FRA-ANP standard fuel.

When a FoN(Z) measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate allowance for a full core map taken with the movable incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.

F0N(Z) is arbitrarily limited for P *0.5 (except for low power physics tests).

FoEQ(Z) is the measured FoN(Z) obtained at equilibrium conditions during the target flux determination. FoEO(Z) must satisfy the relationship that is in the COLR.

Because the value of FoN(Z) represents an equilibrium condition, it does not include the variations of FQN(Z) which are present during non-equilibrium situations such as load following or power ascension. To account for these possible variations, the equilibrium value of FoN(Z) is adjusted by an elevation dependent factor, W(z), that accounts for the calculated worst case transient conditions. Core power distribution is controlled under non-equilibrium conditions by operating the core within the core operating limits on axial flux distribution, quadrant power tilt, and control rod insertion.

If a power distribution measurement indicates that the FaEo(Z) transient relationship's margin to the limit has decreased since the previous evaluation then TS 3.1 0.b.6.C provides two options of either increasing the FoEO(Z) transient relationship by the appropriate penalty factor or increasing the power distribution surveillance to once every 7 EFPD until two successive flux maps indicate that the FoEQ(Z) transient relationship's margin to the limit has not decreased. IF FoEo(Z) with the penalty factor applied is greater than the limit, then TS 3.1 0.b.6 is not satisfied and TS 3.1 0.b.7 should be applied to maintain the normal surveillance interval. Based on TS 3.10.b.7.A, the axial flux distribution (AFD) limits are reduced by 1% for each 1% that the FoEo(Z) transient relationship exceeds its limit within the allowed time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The contingency actions of TS 3.1 0.b.6 and TS 3.1 0.b.7 are to ensure that FaN(Z) does not exceed its limit for any significant period of time without detection. Satisfying limits on FQN(Z) ensures that the safety analyses remain bounding and valid.

TS B3.10-2 07/06/2004

F6HN Nuclear EnthalpV Rise Hot Channel Factor F&N, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the maximum integral of linear power along a fuel rod to the core average integral fuel rod power.

It should be noted that F&N is based on an integral and is used as such in DNBR calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizontal (x-y) power shapes throughout the core. Thus, the horizontal power shape at the point of maximum heat flux is not necessarily directly related to FAHN.

The FAHN limit is determined from safety analyses of the limiting DNBR transient events. The safety analyses are performed for FRA-ANP heavy fuel, FRA-ANP standard fuel, and Westinghouse 422 V+ fuel. In these analyses, the important operational parameters are selected to minimize DNBR.

The results of the safety analyses must demonstrate that minimum DNBR is greater than the DNBR limit for a fuel rod operating at the F&HN limit.

The use of F&HN in TS 3.1 O.b.5.C is to monitor wupburn" which is defined as an increase in FAHN with exposure. Since this is not to be confused with observed changes in peak power resulting from such phenomena as xenon redistribution, control rod movement, power level changes, or changes in the number of instrumented thimbles recorded, an allowance of 2% is used to account for such changes.

TS B3.10-3 07/0612004

Rod Bow Effects No penalty for rod bow effects rideds to be included in TS 3.10.b.1 for FRA-ANP fuel.(t)

Surveillance Measurements of the hot channel factors are required as part of startup physics tests, at least each full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map taken following initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identifies operational anomalies which would otherwise affect these bases.

For normal operation, it is not necessary to measure these quantities. Instead it has been determined that, provided certain conditions are observed, the hot channel factor limits will be met.

These conditions are as follows:

1. Control rods in a single bank move together with no individual rod insertion differing by more than an indicated 12 steps from the bank demand position where reactor power is 285%, or an indicated 24 steps when reactor power is < 85%.

2. Control rod banks are sequenced with overlapping banks as specified in the COLR.

3. The control bank insertion limits as specified in the COLR are not violated, except as allowed by TS 3.1 O.d.2.

4. The axial power distribution, expressed in terms of axial flux difference, is maintained within the limits.

The limits on axial flux difference (AFD) assure that the axial power distribution is maintained such that the FQ(Z) upper bound envelope of FQLIMIT times the normalized axial peaking factor [K(Z)] is not exceeded during either normal operation or in the event of xenon redistribution following power changes. This ensures that the power distributions assumed in the large and small break LOCA analyses will bound those that occur during plant operation.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD monitor program. The computer determines the AFD for each of the operable excore channels and provides a computer alarm if the AFD for at least 2 of 4 or 2 of 3 operable excore channels are outside the AFD limits and reactor power is greaterthan 50 percent or RATED POWER.

For Condition II events the core is protected from overpower and a minimum DNBR less than the DNBR limit by an automatic Protection System. Compliance with the specification is assumed as a precondition for Condition II transients; however, operator error and equipment malfunctions are separately assumed to lead to the cause of the transients considered.

(1)N. E. Hoppe, 'Mechanical Design Report Supplement for Kewaunee High Bumup (49 GWD/MTU)

Fuel Assemblies," XN-NF-84-28(P), Exxon Nuclear Company, July 1984.

Amendment No. 167 TS B3.10-4 04/04/2003

Quadrant Power Tilt Limits (TS 3.1Q.c The radial power distribution within the core must satisfy the design values assumed for calculation of power capability. Radial power distributions are measured as part of the startup physics testing and are periodically measured at a monthly or greater frequency. These measurements are taken to assure that the radial power distribution with any quarter core radial power asymmetry conditions are consistent with the assumptions used in power capability analyses.

The quadrant tilt power deviation alarm is used to indicate a sudden or unexpected change from the radial power distribution mentioned above. The 2% tilt alarm setpoint represents a minimum practical value consistent with instrumentation errors and operating procedures. This symmetry level is sufficient to detect significant misalignment of control rods. Misalignment of control rods is considered to be the most likely cause of radial power asymmetry. The requirement for verifying rod position once each shift is imposed to preclude rod misalignment which would cause a tilt condition less than the 2% alarm level. This monitoring is required by TS 4.1.

The two hour time interval in TS 3.1 0.c is considered ample to identify a dropped or misaligned rod.

If the tilt condition cannot be eliminated within the two hour time allowance, additional time would be needed to investigate the cause of the tilt condition. The measurements would include a full core power distribution map using the movable detector system. For a tilt ratio > 1.02 but

• 1.09, an additional 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> time interval is authorized to accomplish these measurements. However, to assure that the peak core power is maintained below limiting values, a reduction of reactor power of 2% for each 1% of indicated tilt is required. Power distribution measurements have indicated that the core radial power peaking would not exceed a two-to-one relationship with the indicated tilt from the excore nuclear detector system for the worst rod misalignment. If a tilt ratio of > 1.02 but

• 1.09 cannot be eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then the reactor power level will be reduced to *50%.

If a misaligned rod has caused a tilt ratio > 1.09, then the core power shall be reduced by 2% of rated value for every 1% of indicated power tilt ratio > 1.0. If after eight hours the rod has not been realigned, then the rod shall be declared inoperable in accordance with TS 3.1 0.e, and action shall be taken in accordance with TS 3.1 0.g. If the tilt condition cannot be eliminated after 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, then the reactor shall be brought to a minimum load condition; i.e., electric power 5 30 MW. If the cause of the tilt condition has been identified and is in the process of being corrected, then the generator may remain connected to the grid.

If the tilt ratio is > 1.09, and it is not due to a misaligned rod, then the reactor shall be brought to a no load condition (i.e., reactor power*5 5%) for investigation by flux mapping. Although the reactor may be maintained critical for flux mapping, the generator must be disconnected from the grid since the cause of the tilt condition is not known, or it cannot be readily corrected.

Rod Insertion Limits (TS 3.10.d)

The allowed completion time of two hours for restoring the control banks to within the insertion limits provides an acceptable time for evaluation and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

Operation beyond the rod insertion limits is allowed for a short-time period in order to take conservative action because the simultaneous occurrence of either a LOCA, loss-of-flow accident, ejected rod accident, or other accident during this short time period, together with an inadequate power distribution or reactivity capability, has an acceptably low probability.

Amendment No. 167 TS B3.10-5 04/04/2003

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 O.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, orthrough 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 priorto initial operations forthatcycle. Upon reaching full powerconditions 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.

Amendment No. 167 TS B3.10-6 04/04/2003

Inoperable Rod Position Indicator Channels (TS 3.10.f)

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, then the operator will be fully aware of the inoperability of the channel, and special surveillance of core power tilt indications, using established procedures and relying on excore nuclear detectors, and/or movable incore detectors, will be used to verify power distribution symmetry.

Inoperable Rod Limitations (TS 3.10.g)

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 Temrerature (TS 3.1 0.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 3.10.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. 167 TS 83.10-7 04/04/2003

DNBR Parameters (TS 3.10.nA The DNBR related safety analyses make assumptions on reactor temperature, pressure, and flow.

In the event one of these parameters does not meet the TS 3.1 O.k, TS 3.10.1 or TS 3.1 O.m limits, an analysis can be performed to determine a power level at which the MDNBR limit is satisfied.

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