ML19256G015

Revised Pages 15.1-4 Through 15.3.10-16 for Proposed Amend to Tech Specs Re Control Rod & Power Distribution Limits

ML19256G015
Person / Time
Site:	Point Beach
Issue date:	12/19/1979
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML19256G008	List: ML19256G007 ML19256G015
References
TAC-12663, TAC-12664, NUDOCS 7912270220
Download: ML19256G015 (20)

Text

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2) Cold Shutdown The reactor is in the cold shutdown condition when the reactor has a shutdown margin of at least 1% Ak/k and reactor coolant temperature is 3,200*F.

3) Refueling Shutdown The reactor is in the refueling shutdown condition when the reactor is suberitical by at least 10% ak/k and T,yg is 3,140*F.

A refueling shutdown refers to a shutdown to move fuel to and from the reactor core.

4) Shutdown Margin Shutdown margin is the amount by which the rea-tor core would be suberitical if all withdrawn control rods were tripped into the core but the highest worth withdrawn RCCA remains fully withdrawn.

If the reactor is shut down from a power condition, the hot shutdown temperature should be assumed. In other cases, no change in temperature should be assumed.

_~

h. Power Operation The reactor is in power operating condition when the reactor is critical and the average neutron flux of the power range instrumentation indicates greater than 2% of rated power.

i. Refueling Operation Refueling operation is any operation involving movement of core components (those that could affect the reactivity of the core) within the containment when the vessel head is unbolted or removed.

1647 131 7912270 #

15.1-4 November 1979 22O

j. Rated Power Rated power is here defined as a steady state reactor core output of 1518.5 MWT.

k. Thermal Power Thermal power is defined as the total core heat transferred from the fuel to the coolant.

1. Degree of Redundancy Degree of redundancy is defined as the difference between the number of operable channels and the minimum number of channels which when tripped will cause an automatic shutdown.

m. Reactor Critical The reactor is said to be critical when the neutron chain reaction is self-sustaining and keff = 1.0.

n. Low Power Operation The reactor is in the low power operating condition when the reactor is critical and the average neutron flux of the power range instrumentation indicates less than or equal to 2% of rated power,

o. Fire Suppression Water System A FIRE SUPPRESSION WATER SYSTEM shall consist of: a water source; pump (s) ; and distribution piping with associated sectionalizing control or isolation valves. Such valves shall include yard post indicating valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser.

1647 132 15.1-5

Additional peaking factors to account for local peaking due to fuel rod axial gaps and reduction in fuel pellet stack length have been included in the calculation of the curves shown in Figure 15.2.1-1. These curves are based on an F N f 1.58, cosine axial flux H

shape, and a DNB analysis as described in Section 4.3 of WCAP-8050, " Fuel Densification, Point Beach Nuclear Plant Unit 1 Cycle 2" (including the effects of fuel densification and flattened cladding) .

Figure 15.2.1-1 also includes an allowance for an increase in the enthalpy rise hot channel factor at reduced power based on the expression:

FN = 1.58 {l + 0.2 (1-P) } where P is a fraction of rated power AH when P < 1.0. FN = 1.58 when P > 1.0.

The hot channel factors are also sufficiently large to account for the degree of malpositioning of full-length rods that is allowed before the reactor trip setpoints are reduced and rod withdrawal block and load runback may be required.

Rod withdrawal block and load runback occur before reactor trip setpoints are reached. The Reactor Control and Protective System is designed to prevent any anticipated combination of transient conditions that would result in a DNB ratio of less than 1.30.

1647 133 15.2.1-3 November 1979

15.3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability Applies to the operation of the control rods and to core power distribution limits.

Objeccive To insure (1) core suberiticality after a reactor trip, (2) a limit on potential reactivity insertions from a hypothetical rod cluster control assembly (RCCA) ejection, and (3) an acceptable core power distribution during power operation.

Specification A. Bank Insertion Limits

1. When the reactor is critical, except for physics tests and control rod exercises, the shutdown banks shall be fully withdrawn.

2. When the reactor is critical, the control banks shall be inserted no further than the limits shown by the lines on Figure 15.3.10-1.

Exceptions .to the insertion limit are permitted for physics tests and control rod exercises.

3. The shutdown margin shall exceed the applicable value as shown in Figure 15.3.10-2 under all steady-state operating conditions from 350*F to full power. An exception to the stuck RCCA component of the shutdown margin requirement is permitted for physics tests.

4. Except for physics tests a shutdown margin of at least 1% Ak/k shall be maintained when the reactor coolant temperature is less than 350*F.

5. When the reactor is in the hot shutdown condition or during any approach to criticality, except for physics tests, the critical rod position shall not be lower than the insertion limit for zero power. Tha t is , if the control rods were withdrawn in normal sequence with no other reactivity change, the reactor would not be critical until the control banks were above the insertion limit.

15.3.10-1 November 1979

B. Power Distribution Limits

1. a. Except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:

Fg (Z) < (2.32) x K(Z) for P> .5 P

Fg(Z)<4.64 x K(Z) for P<,.5 FN 1.58 x {l + 0.2 (1-P)}

Where P is the fraction of full power at which the core is operating, K(Z) is the function in Figure 15.3.10-3 and Z is the core height location of F .

Q

b. Following a refueling shutdown prior to exceeding 90% of rated power and at effective full power monthly intervals thereafter, power distribution maps using the moveable incore detector system shall be made to confirm that the hot channel factor limits are satisfied. The measured hot channel factors shall be increased in the following way:

(1) The measurement of total peaking factor, F. eas, shall be increased by three percent to account for manufacturing tolerances and further increased by five percent to account for measurement error.

(2) The measurement of enthalpy rise hot channel factor, FN LH, shall be increased by four percent to account for measure-ment error.

c. If a measured hot channel factor exceeds the full power limit of Specification 15.3.10.B.l.a, the reactor power and power range high setpoints shall be reduced until those limits are met. If subsequent flux mapping cannot, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, demonstrate that the full power hot channel factor limits are met, the overpower 15.3.10-2 November 1979 1647 135

and overtemperature AT trip setpoints shall be similarly reduced and reactor power limited such that Specification 15.3.10.B.l.a above is met.

2. a. The target flux difference as defined in the basis shall be measured at least quarterly. A target flux difference update value shall be determined monthly by measurement, or by linear interpolation between the last measured value and 0% at end of cycle life (that is when the boron concentration in the coolant is zero ppm), or by extrapolation of the last three measured points.

The target flux difference and its associated alarm setpoints need not be updated if the update value for full power target flux difference is within +0.5% of the presently employed full power target flux difference value.

b. Except for physics testing, excore detector calibration (including recovery) , or as modified below, the indicated axial flux difference shall be maintained within a range of +6 and -9 percent of the target flux difference. This is defined as the target band.

c. At a power level greater than 90 percent of rated power, if the indicated axial flux difference deviates from its target band, the flux difference shall be returned to the target band immediately or reactor power shall be reduced to a level no greater than 90 percent of rated power.

d. At a power level no greater than 90 percent of rated power, (1) The indicated axial flyv difference may deviate from its +6 to -9% target band for a maximum of one hour (cumulative) in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provided the flux difference does not exceed an envelope bounded by -11 percent and +11 percent

9% wer and increasing by -1% and +1% for each 2% of 15.3.10-3 November 1979

rated power below 90%. If the cumulative time exceeds one hour in r- r 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, then the reactor power shall be reduced immediately to no greater than 50% power and the high neutron flux setpoint reduced to no greater than 55%

of rated power.

(2) A power increase to a level greater than 90% of rated power is contingent upon the indicated axial flux difference being within its target band.

e. At a power level no greater than 50 percent of rated power, (1) The indicated axial flux difference may deviate from its target band.

(2) A powerincrease to a level greater than 50% of rated power is contingent upon the indicated axial flux difference not being outside ita target band for more than two hours (cumulative) out of the preceding 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. One half of the time the indicated axial flux difference is out of its target band up to 50% of rated power is to be counted as contributing to the one hour cumulative maximum the flux difference may deviate frem its target band at a power level less than or equal to 90% of rated power.

f. Alarms shall normally be used to indicate non-conformance with the flux difference requirement of 15.3.10.B.2.c or the flux difference-time requirement of 15.3.10.B.2.d(l) . If the alarms are temporarily out-of-service, the axial flux difference shall be noted and conformance with the limits assessed every hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and half-hourly thereafter.

1647 137 15.3.10-4 November 1979

3. Except for physics tests, whenever the indicated quadrant power tilt exceeds 2% the tilt condition shall be eliminated within two hours or the following actions shall be taken:

a. Reduce core power level and the power range high flux setpoint two percent of rated values for every percent of indicated quadrant power tilt,

b. If the tilt is not corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but the hot channel factors for rated power are not exceeded, an evaluation as to the cause of the discrepancy shall be made and reported to the Nuclear Regulatory Commission. Return to full power is permitted, providing the hot channel factors are not exceeded.

c. If the design hot channel factors for rated power are exceeded or not determined within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Nuclear Regulatory Commission shall be notified and the overpower AT and overtemperature AT trip set-points shall be reduced by the equivalent of 2% power for every percent of quadrant power tilt,

d. The excore nuclear instrumentation system serves as the primary quadrant power tilt alarm. If the alarm is not functional for two hours, backup methods of assuring that the quadrant power tilt is acceptable shall be used. These methods include hand calculations, incore thermocouples using either a computer or manual calculations or incore detectors.

e. bhen one power range channel is inoperable and thermal power is greater than 75% of rated thermal power, the quadrant power tilt shall be confirmed as acceptable by use of the movable incore detectors at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. Inoperable Rod Cluster Control Assembly (RCCA)

1. An RCCA shall be considered inoperable if one or more of the following occurs:

15.3.10-5 November 1979 1647 138

a. The RCCA does not drop upon removal of stationary gripper coil voltage.

b. The RCCA does not step in properly when the proper voltage sequences are applied to the control red drive mechanism coils. It shall then be assumed inoperable until it has been tested to verify that it does drop.

c. If the bank demand position is greater than or equal to 215 steps, or, less than or equal to 30 steps, and the rod position indicator channel shows a misalignment of 15 inches. The RCCA shall be assumed inoper-able atil it has been tested to verify that it d(,0s step properly.

d. If the bank demand position is between 215 steps and 30 steps, and the rod position indicator channel shows a misalignment of 7.5 inches.

The RCCA shall be assumed inoperable until it has been tested to verify that it does step properly.

2. Specification 15.3.10.C.1.b can be modified by the following:

a. If an RCCA does not step in upon demand, up to six hours is allow-ed to determine whether the problem with stepping is an electrical problem. If the problem cannot be resolved within six hours, the RCCA shall be assumed inoperable until it has been verified that it will step in or would drop upon demand.

b. If more than one RCCA does not step in, apparently due to electri-cal problems, the situation shall be rectified or clearly defined that it is an electrical problem and the RCCAs are capable of drop-ping upon demand or an orderly shutdown shall commence within six hours,

3. No more than one inoperable RCCA shall be permitted during sustained power operation.

4. When it has been determined that an RCCA does not drop an removal of stationary gripper coil voltage, the shutdown margin shall be maintained by boration as necessary to compensate for the withdrawn worth of the inoperable RCCA. If sustained power operation is anticipated, the insert 2'n limit shall be adjusted to reflect the worth of the inoperable rod.

1647 139 15 3 1o-e November 1929

D. Misaliened or Droeped RCCA

1. If the rod position indicator channel is functional and the associated RCCA is more than 7.5 inches indicated out of alignment with its bank and cannot be aligned when the bank is between 215 steps and 30 steps, then unless the hot channel factors are shown to be within design limits as specified in Section 15.3.10.B-1 within eight (8) hours, power shall be reduced to less than 75% of rated power. When the bank position is greater than or equal to 215 steps, or, less than or equal to 30 steps, the allowable indicated misalignment is 15 inches.

2. To increase power above 75% with an RCCA more than 7.5 inches indicated out of alignment with its bank when the bank position is between 215 steps and 30 steps, an analysis shall first be r.ade to determine the hot channel factors and the resulting allowable power level based on Section 15.3.10.B. When the bank position is greater than or equal to 215 steps, or, less than or equal to 30 steps, the allowable indicated misalignment is 15 inches.

3. If it is determined that the apparent misalignment er dropped RCCA indication was caused by rod position indicator channel failure, sustained power operation may be continued if the following conditions are met:

a. For operation between 10% power and rated power, the position of the RCCA (s) with the failed rod position indicator channel (s) will be checked indirectly by core instrumentation (excore detectors, and/

or thermocouples, and/or moveable incore detectors) every shift and after associated bank motion exceeding 24 steps in one direction,

b. For operation below 10% of rated power, no special monitoring is required.

E. RCCA Drop Times

1. At operating temperature and full flew, the drop time of each RCCA shall be no greater than 1.8 seconds frons the loss of stationary gripper ceil voltage to dashpot entry.

15.3.10-7 1647 140 nove=ber 1979

Basis Insertion Limits and shutdown Margin The reactivity control concept is that reactivity changes accompanying changes in reactor power are compensat-d by control rod motion. Reactivity changes associated with xenon, samarium, fuel depletion and large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated by changes in the soluble boron concentration.

During power operation, the shutdown banks are fully withdrawn and control of reactor power is by the control banks. The control rod insertion limits provide for achieving hot shutdown by reactor trip at any time and assume the highest worth control rod remains fully withdrawn. A 10% margin in reactivity worth of the control rods is included to assure meeting the assumptions used in the accident analysis. So a reactor trip occurring during power operatien will put the reactor into the hot shutdown condition. In addition, the insertion 12mits provide a limit on the maximum inserted rod worth in the unlikely event of a hypothetical rod ejection and provide for acceptable nuclear peaking factors. The specified control rod insertion limits take into account the effects of fuel densificatien.

The rods are withdrawn in the sequence of A, B, C, D with overlap between banks.

The overlap between successive control banks is provided to compensate for the low differential rod worth near the top and bottom of the core.

When tha insertion limits are observed and the control rod banks are above thc solid lines shown on Figure 15.3.10-1, the shutdown requirement is met. The maximum shutdown margin requirement occurs at end of core life and is based on the value used in analysis of the hypothetical steam break accident. Figure 15.3.10-2 shows the shutdown margin equivalent to 2.77% reactivity at end-of-life with res-pect to an uncontrolled cooldown. All other accident analyses assume 1% or greater reactivity shutdown margin. Shutdown margin calculations include the effects 15.3.10-8 November 1979 1647 141

of axial power distribution. One may assume no change in core poisoning due to xenon, samarium or soluble boron.

Part length rod insertion is not permitted, thus eliminating certain adverse power shapes which might occur during power operation. The part length rods have been removed from the core.

Power Distribution Design criteria have been chosen which are consistent with the fuel integrity analyses. These relate to fission gas release, pellet temperature and cladding mechanical properties. Also the minimum DNBR in the core must not be less than 1.30 in normal operation or in short-term transients.

In addition to the above, the peak linear power density must not exceed the limiting kw/ft values which result from the large break loss of coolant accident unalysis based upon the ECCS acceptance criteria limit of 2200*F. This is required to meet the initial conditions assumed for loss of coolant accident.

To aid in specifying the limits on power distribution, the following hot channel factors are defined:

F (Z) , Height Dependent Heat Flux Hot Channel Factor, is defined as the local heat flux on the su2 face of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods. Imposed limits pertain to the maximum Fg(Z) in the core.

E Fg , Engineering Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically, the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.

15.3.10-9 November 1979

PfH,NuclearEnthalpyRiseHotChannelFactor, is defined as the ratio of the integral of linear power along a fuel rod to the average fuel red power. ImposedlimitspertaintothemaximumF$H in the core, that is the fuel rod with the highest integrated power. It should be noted thatF$H is based on an integral and is used as such in the DNB calculations. Local heat flux is 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 necessarilydirectlyrelatedtoF$H*

For normal operation, it is not necessary to measure these quantities.

Instead it has been determined that, provided the following conditions are observed, the hot channel factor limits will be met:

1. Control rods in a single bank move together with no individual rod insertion differing by more tnan 15 inches from the bank demand position, when the bank demand position is between 30 steps and 215 steps. 22.5 inches misali;nment is allowed when the bank position is less than or equal to 30 steps, or, when the bank position is greater than or equal to 215 steps, due to the small worth and consequential effects of an individual rod misalignment.

2. Control rod banks are sequenced with overlapping banks as described in Figure 15.3.10-1.

3. The full-length control bank insertion limits are not violated.

4. Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits, are observed. Flux difference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectcrs.

The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

The permitted relaxation of FfH allows radial power shape changes with rod insertion to the insertion limits. It has been determined that provided the above conditions 1 through 4 are observed, these hot channel factor limits are met. In Specification 15.3.10.B.l.a, Fg is arbitrarily limited for p < 0.5 (except for low power physics tests).

1647 143 15.3.10-10 November 1979

An upper bound envelope of 2.32 times the normalized peaking factor axial dependence of Figure 15.3.10-3 consistent with the Technical Specifications or. power distribution control as given in Section 15.3.10 was used in the LOCA analysis. The results of the analyses based on this upper bound envelope indicate a peak clad temperature of less than the 2200*F limit.

When an Fg 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 moveable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.

N In the design limit of FaH there is eight percent allowance for uncertainties which means that normal cperation of the core is expected to result in a design F$g<l.58/1.08 The logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power shape (i.e. , red misalignment) affect FfH, ases u ssarily affecting Fg, (b) while the operator has a direct influence on Fg through movement of rods, and can limit it to the desired value, he has no direct control over F N and (c) an error in the 6H predictions for radial power shape which may be detected during startup physics tests can be compensated for in Fg by tighter axial control, but compensation for F II is less readily available. When a measuren.ent of F II H is taken, aH experimental error must be allow d for and four percent is the appropriate allowance for a full core map taken with the moveable incore detector flux t

mapping system. l Measurements of the hot channel factors are required as part of startup physics i

tests, at least each full power month operation, and whenever abnormal power distribution ccnditions require a reduction of core power to a level based upon measured hot channel factors. The incore map taken following initial leading provides confirmation of the basic nuclear design bases including prcper fuel loading patterns. The periodic monthly incere mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational

}hk7 lk4 15.3.10-11 November 1979

anomalies which would, otherwise, affect these bases.

Axial Power Distribution The procedures for axial power distribution control are designed to minimize the effects of xenon redistribution on the axial power distribution during load follow maneuvers. Ensically, control of flux difference is required to limit the difference between the current value of flux difference ( AI) and a reference value which corresponds to the full power equilibrium value of axial offset hxial offset = AI/ fractional power) .

The full power target flux difference is defined as that indicated flux difference of the core in the following condition: equilibrium xenon aittle or no oscillation) and with the full-length rod control rod bank more than 190 steps withdrawn E.e. , the normal full power position) . Values for all other core power levels are obtained by multiplying the full power value by the factienal power.

At zero power the target flux difference is 0%. S ince the indicated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviation of +6 and -9 percent AI are permitted from the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring the target flux difference every month. For this reason, the specification provides three methods for updating the target flux difference.

S trict control of the flux difference bnd rod position) is not as necessary during reduced power operation. This is because xenon distribution control at reduced power is not as significant as the control at full power and allowance has been made in predicting the heat flux peaking factors for less strict control at reduced power. S trict control of the flux difference is not possible during certain physics tests or during required periodic excore calibrations which require larger flux differences than permitted. Therefore, the specifi-15.3.10-12 November 1979 1647 145

cations on power distribution control are not applied during physics tests or excore calibrations. This is acceptable due to the increased core monitoring performed as part of the tests and low probability of a significant accident occurring during these operations.

In some instances of rapid plant power reduction, automatic rod motion will cause the flux difference to deviate from the target band when the reduced power level is reached. This does not necessarily affect the xenon distribution sufficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target band. However, to simplify the specification for operation up to 90% of full power, a limitation of one hour in any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is placed aaoperation outside the band. This insures that the resulting xenon distributions are not significantly different from these resulting from operation within the target band.

For normal operation and anticipated transients, the core is protected from overpower and ednimum DNBR of 1.30 by an automatic protection system. Compliance with operating procedures is assumed as a pre-condition; however, operator error and equipeent malfunctions are separately assumed to lead to the cause of the transients considered.

Quadrant Tilt The excore detectors are somewhat insensitive to disturbances near the core center such as misaligned inner control rods. It is therefore possible that a five percent tilt might actually be present in the core when the excore detectors respond with a two percent indicated quadrant tilt. On the other hand, they are overly responsive to disturbances near the periphery. -

1647 146 15.3.10-13 November 1979

Tilt restrictions are not applicable during the startup and initial testing of a reload core which may have an inherent tilt. During this time sufficient testing is performed at reduced power to verify that the hot channel factor limits are met and the nuclear channels are properly aligned.

The excore detectors are normally aligned indicating no quadrant power tilt because they are used to alarm on a rapidly developing tilt. Tilts which develop slowly are more accurately and readily discerned by incore measurements.

The excore detectors serve as the prime indication of a quadrant power tilt.

If a channel fails, is out-of-service for testing, or is unreliable, two hours is a short time with respect to the probability of an unsafe quadrant power tilt developing. Two hours gives the cperating personnel sufficient time to have the problem investigated and/or put into operation one of several possible alternative methods of determining tilt.

Inoperable RCCA An inoperable rod imposes additional demands on the operators. The permissible number of inoperable control rods is limited to one in order to limit the magnitude of the operating burden.

From operating experience to date, an RCCA which steps in properly will drop when a trip signal occurs because the only force acting to drive the rod in is gravity. When it has been determined that a rod does not drop, extra shutdown margin is gained by boration or by adjusting the insertion limit to account for the worth of the inoperable control rod.

Further experience indicates thatcontrol rods which do not step are usually affected by electrical problems. That is, normally the problem is in the rod control cabinets. If operability cannot be restored, the RCCA will be declared inoperable and corrective actions can be taken to compensate for the associated reduction in shutdown margin. If there is more than one RCCA affected, an orderly shutdown would be started. Such an evolution would have to be performed j ] 15.3.10-14 November 1979

in a deliberate manner without undue pressure on the operating personnel because of the unusual techniques to be used to accommodate the reactivity c'inges associated with the thutdown.

Misaligned RCCAS The various control rod banks (shutdown banks and control banks, A, B, C, and D) are each to be moved as a bank; that is, with all rods in the bank within one step (5/8 inch) of the bank position. Direct information on rod position indication is provided by two methods: A digital count of actuating pulses which shows the demand position of the banks and a linear position indicator (LVDT) which indi-cates the actual rod position. The rod position indicatoi channel has a demon-strated accuracy of 5% of span (+7.2 inches) . Therefore, an analysis has been performed to show that a misalignment of 15 inches cannot cause design hot channel factors to be exceeded. A single fully misaligned RCCA, that is, an RCCA 12 feet out of alignment with its bank, does not result in exceeding core limits in steady-state operation at power levels less than or equal to rated power. In other words, a single dropped RCCA is allowable from a core power distributien viewpoint. If the misalignment condition cannot be readily corrected, the specified reduction in power to 75% will insure that design margins to core limits will be maintained under both steady-state and anticipated transient conditions.

The eight (8) hour permissible limit on rod misalignment at rated power is short with respect to the probability of an independent accident.

Because the rod position indicator system may have a 7.5 inch error when a misalignment of 15 inches is occurring, the Specification allows only a 7.5 inch indicated misalign=ent. However, when the bank demand position is greater than or equal to 215 steps, or, less than or equal to 30 steps, the consequences of a misalignment are much less severe. The differential worth of an individual RCCA is less, and the resultant purturbation on power distributions is less than when the bank is in its high differential worth region. At the top and bottom of the core, an indicated 15 inch misalignment may be representing an actual misalignment of 22.5 inches.

The failure of an LVDT in itself does not reduce t... shutdown capability of the j 15.3.10-15 November 1979

rods, but it does reduce the operator's capability for determining the position of that rod by direct means. The operator has available to him the excore detector recordings, incore thermoccuple readings and periodic incore flux traces for indirectly determining rod position and flux tilts should the rod with the inoperable LVDT become malpositioned. The excore and incore instrumenta-tion will not necessarily recognize a misalignment of 15 inches because the concommitant increase in power density will normally be less than 1% for a 15 inch misalignment. The excore and incere instrumentation will, however, detect any rod misalignment which is sufficient to cause a significant increase in hot channel factors and/or any significant loss in shutdown capability. The increased surveil-lance of the core if one or more rod position indicator channels is out-of-service serves to guard against any significant loss in shutdown margin or margin to core thermal limits.

The history of malpositioned RCCA's indicates that in nearly all such cases, the malpositioning occurred during bank movement. Checking rod position after bank motion exceeds 24 steps will verify that the RCCA with the inoperable LVDT is moving properly with its bank and the bank step counter. Malpositioning of an RCCA in a stationary bank is very rare, and if it does occur, it is usually gross slippage which will be seen by external detectors. Should it go undetected, the time between the rod position checks performed every shift is short with respect to the probability of occurrence of another independent undetected situation which would further reduce the shutdown capability of the rods.

Any combination of misaligned rods below 10% rated power will not exceed the design limits. For this reason, it is not necessary to check the position cf rods with inoperable LVDT's below 10% power; plus, the incore instrumentation is not effective for determining rod position until the power level is above approximately 5%.

I647 149 15.3.10-16 November 1979

FIGURE 15.3.10-1 CONTROL BANK INSERTI0fl LIMITS

(~ POINT BEACH UNITS 1 AND 2 (22) (77) 10 0 a=w. - - - - - -

228

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~. - ,--_ ..._.- . ._._./____,.__ - . . -

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55-/..=

a-Bank B - .. -u

. =

. _ . _....__ . _ ,._ _ .....,. -. - . . . .. : -/= . - -- t 80 '=-- 7 ----

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===: - - - - - . _

182

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