ML20003G913
| ML20003G913 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/28/1981 |
| From: | VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML18139B283 | List: |
| References | |
| NUDOCS 8105040195 | |
| Download: ML20003G913 (71) | |
Text
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l PROPOSED TECHNICAL SPECIFICATION CHANGE
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, TS 3.12-1 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS Applicability ,
Applies to the operation of the control rod assemblies and power distri-bution limits. Obj ective To ensure core suberiticality after a reactor trip, a limit on potential reactivity insertions from hypathetical control rod assembly ejection, and an acceptable core power distribution during power operation. 3 Specification A. Control Bank Insertion Limits
- 1. Whenever the reactor is critical, except for physics tests and l
control rod assembly exercises, the shutdown control rods shall be fully withdrawn.
- 2. Whenever the reactor is critical, except for physics tests and control rod assembly exercises, the full length control rod banks shall be inserted no further than the appropriate limit determined by core burnup shown.on.TS Figures 3.12-1A, 3.12-1B, 3.12-2, or 3.12-3 for three-loop operation and TS Figures 3.12-4A, 3.12-4B, 3.12-5 or 3.12-6 for two-loop operation.
- 3. The limits shown on TS Figures 3.12-1A through 3.12-6 may be revised on the basis of physics calculations and physics data obtained during unit startup and subsequent operation, in I
accordance with the following:
- a. The sequence of withdrawal of the controlling banks, when going from zero to 100% power, is A, B, C, D.
- b. An overlap of control banks, consistent with physics tal-
TS 3.12-2 , 1 i . , culations and physics data obtained during Unit Startup and subsequent operation, will be permitted.
- c. The shutdown margin with allowance for a stuck control rod assembly shall be greater than or equal to 1.77% reactivity under all steady-state operation conditions, except for physics tests, from zero to full power, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at hot shutdown conditions (T,y 2547'F) if all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron.
l 4. Whenever the reactor is suberitical, except for physics tests, the critical rod position, i.e., the rod position at which criticality would be achieved if the control rod assemblies were withdrawn in normal sequence with no other reactivity changes; shall not be lower than the insertion limit for zero power.
- 5. Insertion limits do not apply during physics tests or during periodic e.xercise of' individual rods. However, the shutdown ma'rgin indicated '
, above must be maintained except for the low power physics test to measure control rod worth and shutdown margin. For this test the reactor may be critical with all but one full control rod, expected to have the highest worth, inserted.
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. TS 3.12-3 B. Power Distribution Limits
- 1. At all times except during low power physics tests, the hot channel factc.rs defined in the basis must meet the following limits:
F9 (Z) 5 2.18/P x K(2) for P > 0.5 F9 (Z) 5 4.36 x K(Z) for P $ 0.5 F $ 1.55 (1+0.2(1-P)) where P is the fraction of rated power at which the core is operating, K(Z) is the function given in TS Figure 3.12-8, and Z is the core height location of F . q
- 2. Prior to exceeding 757, power'following each core loading and during each effective full power month of operation thereaf ter, power distri-bution maps using the movable detector system shall be made to confirm that the hot channel factor limits of this specification are satis-fled. For the purpose of this confirmation:
s
- a. The measurement of total peaking factor shall be increased by eight percent to account for manufacturing tolerances, measure-ment error and the effects of rod bow. The measurement of enthalpy rise hot channt.1 factor F g shril be increased by four percent to account for measurement error. If any measured hot channel factor exceeds its limit specified under Specification 3.12.B.1, the reactor power and high neutron flux trip setpoint shall be reduced until the limits under Specification 3.12.B.1 are met. If the hot channel factors cannot be brought to within the limits of F (Z) 9
$ 2.18 x K(Z) and F 5 1.55 within 24 hours, the Overpower AT and Overtemperature AT trip setpoints shall be similarly reduced.
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- TS 3.12-4
- 3. The reference equilibrium indicated axial flux difference (called the target flux difference) at a given power level P, is that indicated axial flux difference with the core in equilibrium xenon conditions (small or no oscillation) and the control rods more than 190 steps withdrawn. The target flux difference at any other power level P is equal to the target value at P, multiplied by the ratio P/P,. The target flux difference shall be measured at least once per equivalent full power quarter. The target flux difference must be updated during each effective full power month of operation either by actual measurements or by linear interpolation using the most recent value and the value predicted for the end of the cycle life.
l l 4. Except as modified by Specifications 3.12.B.4.a, b, c, or d below, the indicated axial flux difference shall be maintained within a
+5% band about the target flux difference (defines the target band on axial flux difference).
- a. At a power level greater than 90 percent of rated power, if the indicated axial flux difference deviates from its target band, within 15 minutes either restore the indicated axial flux difference to within the target band or reduce the reactor power to less than 90 percent of rated power.
- b. At a power level no greater than 90 percent of rated power, (1) The indicated axial flux difference may deviate 1
from its target band for a maximum of one hour (cumulative) in any 24-hour period provided the flux difference is within the limits shown on TS Figure 3.12-10.
a . . TS 3.12-5 One minute penalty is accumulated for each one minute of operation outside of the target band at power levels equal to or above 50% of rated power. (2) If Specification 3.12.B.4.b(1) is violated, then the reactor power shall be reduced to less than 50% power within 30 minutes and the high neutron flux setpoint shall be reduced to no greater than 55% power within the next four hours. (3) A power increase to a level greater than 90 percent of rated
- power is contingent upon the indicated axial flux difference being within its target band.
(4) Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to TS Table 4.1-1 provided the indicated axial flux difference is maintained within the limits of TS Figure 3.12-10. A total of 16 hours of operation may be accumulated with the a'xial flux difference outside of the target band during this testing without penalty deviation.
- c. 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 power increase to a level greater than 50 percent of rated power is contingent upon the indicated axial flux difference not being outside its target band for more than one hour accumulated penalty during the preceding 24-hour period. One half minute penalty is accumulated for each one minute of operation outside of the target band at power levels between 15% and 50% of rated power.
l a . I 1 TS 3.12-6 j l
- d. The axial flux difference limits for Specifications 3.12.B.4.a, . !
b, and c may be suspended during the performance of physics tests provided: (1) The power level is maintained at or below 85% of rated power, and (2) The limits of Specification 3.12.B.1 are maintained. The power level shall be determined to be less than or equal to 85% of rated power at least once per hour during physics . tests. Verification that the limits of Specification 3.12.B.1 are being met shall be demonstrated through in-core flux mapping at least once per 12 hours. Alarms shall normally be used to indicate the deviations from the axial flux difference requirements in Specification 3.12.B.4.a and the flux difference time limits in Specifications 3.12.B.4.b and c. If the alarms are out of service temporarily, the axial flux difference shall be logged and conformance to the limits assessed every hour for the first 24 hours and half-hourly thereafter. The indicated axial flux difference for each excore channel shall be monicored at least once per 7 days when the alarm is operable and at least once per hour for the first 24 hours after restoring the alarm to operable status. ( The allowable quadrant to average power tilt is 2.0%. 5. 6. If, except for physics and rod exercise testing, the quadrant l to average power tilt exceeds 2%, then:
. . \
. TS 3.12-7
- a. The hot channel factors shall be determined within 2 hours l and the power level adjusted to meet the requirement of Specifi- l cation 3.12.B.1, or
- b. If the hot channel factors are not determined within two hours, the power level and high neutron flux trip setpoint shall be reduced from rated power 2% for each percent of quadrant tilt.
- c. If the quadrant to average power tilt exceeum +10%, the power level and high neutron flux trip setpoint will be reduced from rated power 2% for each percent of quadrant tilt.
- 7. If, except for physics and rod exercise testing, after a further period of 24 hours, the power tilt in Specification 3.12.B.5 above is not corrected to less than 2%:
- a. If design hot channel factors for rated power are not exceeded, an evaluation as to the cause of the discrepancy shall be made and reported as a reportable occurrence to the Nuclear Regulatory Commission.
- b. if the design hot channel factors for rated power are exceeded and the power is greater than 10%, the Nuclear Regulatory Commission shall be notified and the Nuclear Overpower, Nuclear overpower AT, and Overtemperature AT trips shall be reduced one percent for each percent the hot channel factor exceeds the rated power design values.
- c. If the hot channel factors are not determined the Nuclear Regulatory Commission shall be notified and the Overpower
O O TS 3.12-8 AT and Overtemperature AT trip settings shall be reduced by the equivalent of 2% power for every 1% quadrant to average ! power tilt. C. Inoperable Control Rods
- 1. A control rod assembly shall be considered inoperable if the assembly cannot be moved by the drive mechanism or the assembly remains misaligned from its bank by more than 15 inches. A full-length control rod shall be considered inoperable if its rod drop time is greater than 1.8 seconds to dashpot entry. .
- 2. No more than one inoperable control rod assembly shall be per-mitted when the reactor is critical.
- 3. If more than one control rod assembly in a given bank is out of service because of a single failure external to the individual rod drive mechanism, i.e. programming circuitry, the provisions of Specifications 3.12.C.1 and 3.12.C.2 shall not apply and the reactor may remain critical for a period not to exceed two hours provided immediate attention is directed toward making the necessary repairs. In the event the affected assemblies cannot be ret'urned to service within this specified period the reactor will be brought to hot shutdown conditions.
- 4. The provisions of Specifications 3.12.C.1 and 3.12.C.2 shall not apply during physics tests in which the assemblies are intentionally misaligned.
- 5. The insertion limits in TS Figure 3.12-2 apply:
l
- a. If an inoperable full-length rod is located below the 200 step level and is capable of being tripped, or
. TS 3.12-9
- b. If the full-length rod is located below the 30 step level, whether or not it is capable of being tripped.
- 6. If an inoperable full-length rod cannot be located or if the inoperable full-length rod is located above the 30 step level and cannot be tripped, then the insertion limits in TS Figure 3.12-3 apply.
- 7. If a full-length rod becomes inoperable and reactor operation is continued, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days. The analysis shall include due l allowance for non-uniform fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the xafety analysis, the unit power level shall be re'duced to an analytically determined part power level which is consistent with the safety analysis.
D. Core Quadrant Power Balance:
- 1. If the reactor is operating above 75% of rated power with one excore nuclear channel out of service, the core quadrant power balance shall be determined:
- a. Once per day, and i
- b. After a change in power level greater than 10% or more than 30
~
\ inches of control rod motion. e
, TS 3.12-10
- 2. The core quadrant power balance shall be determined by one of the following methods:
- a. Movable detectors (at least two per quadrant)
- b. Core exit thermocouples (at least four per quadrant)
E. Inoperable Rod Position Indicator Channels
- 1. If a rod position indicator channel is out of service, then:
- a. For operation between 50% and 100% of rated power, the position of the RCC shall be checked indirectly by cnre instrumentaton (excore detector and/or thermocouples
, and/or movable incore detectors) every shift or subsequent to motion of the non-indicating rod exceeding 24 steps, whichever occurs first.
- b. During operation below 50% of rated power, no special moni-toring is required.
- 2. Not more than one rod position indicator (RPI) channel per group nor two RPI channels per bank shall be permitted to be inoperable at any time.
F. Misalianed or Dropped Control Rod
- 1. If the Rod Position Indicator Channel is functional and the associated full length control rod is more than 15 inches out of alignment with its bank and cannot be realigned, then unless the hot channel factors are shown to be within design limits as l
l specified in Specification 3.12.B.1 within 8 hours, power shall be reduced so as not to exceed 75% of permitted power. l t I
TS 3.12-11
- 2. To increase power above 75% of rated power with a full-length control rod more than 15 inches out of alignment with its bank, an analysis shall first be made to determine the hot channel factors and the resulting allowable power level based on Section 3.12-B.
Basis The reactivity control concept assumed for operation is that reactivity changes accompanying changes in reactor power are compensated by control rod assembly motion. Reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated for by changes in the soluble boron concen-tration. During power operation, the shutdown groups are fully withdrawn and control of power is by the control groups. A reactor trip occurring during power operation will place the reactor into the hot shutdown condition. The control rod assembly insertion limits provide for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod assembly remains fully withdrawn, with sufficient margins to meet the assumptions used in the accident analysis. In addition, they provide a limit on the maximum inserted rod worth in the unlikely event of a hypothetical assembly ejection and provide for acceptable nuclear peaking factors. The limit may be deter-mined on the basis of unit startup and operating data to provide a more realistic limit which will allow for more flexibility in unit operation and
i 1 TS 3.12-12 l I still assure compliance with the shutdown requirement. The maximum shut-down margin requirement occurs at end of core life and is based on the value used in the analysis of the hypothetical steam break accident. The rod insertion limits are based on end of core life conditions. The shut-down margin for the entire cycle length is established at 1.77% reactivity. All other accident analysis with the exception of the chemical and volume control system malfunction analysis are based on 1% reactivity shutdown margin. Relative positions of control rod banks are determined by a specified control rod bank overlap. This overlap is based on the consideration of axial power shape control. The specified control rod insertion limits have been revised to limit the potential ejected rod worth in order to account for the effects of fuel densification. The various control rod assemblies (shutdown banks, control banks ~A, B, C, and D) are each to be moved as a bank; that is, with all assemblies in the bank within one step (5/8 inch) of the bank position. 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, Linear Variable Differential Transformer, which indicaces the actual assembly position. The position indication accuracy of the Linear Differential Transformer is approximately 15% of span (! 7.5 inches) under steady state conditions. The relttive accuracy of the linear position indicator is such that, with the most adverse errors, I an alarm is actuated if any two assemblies within a bank deviate by more than 14 inches. In the event that the linear poisition indicator is not
. TS 3.12-13 in service, the effects of malpositioned control rod assemblies are obser-able from nuclear and process information displayed in the Main Control Room and by core thermocouples and in-core movable detectors. Below 50% power, no special monitoring is required for malpositioned control rod assemblies with inoperable rod position indicators because, even with an unnoticed complete assembly misalignment (full length control rod assenbly 12 feet out of align-ment with its bank), operation at 50% steady state power does not result in exceeding core limits.
The specified control rod assembly drop time is consistent with safety analyses that have been performed. An inoperable control rod assembly imposes additional demands on the operators. The permissible number of inoperable control rod assemblies is limited to one in order to limit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable control rod assemblies upon reactor trip. Tvo criteria have been chosen as a design basis for fuel performance related to fission gas release, pellet temperature, and cladding mechanical properties. First, the peak value of fuel centerline temperature must not exceed 4700*F. Second, the minimum DNBR in the core must not be less than 1.30 in normal operation or in short term transients. l l l
TS 3.12-14 ! l
~
In addition to the above, the peak linear power density and the nuclear enthalpy j l rise hot channel factor must not exceed their limiting values which result from the large break loss of coolant accident analysis based on the ECCS acceptance criteria limit of 2200*F on peak clad temperature. This is required to meet the initial conditions assumed for the loss of coolant accident. To aid in specifying the limits of power distribution, the following hot channel factors are defined: F (Z), Height Dependent Heat Flux Hot Channel Factor, is defined as the maximum 9 local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerance on fuel pellets and rods.
, Engineering Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances. The engineering factor allows for loca' 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. Fg, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power for both LOCA and non-LOCA considerations. l
1 , TS 3.12-15 j .' It should be noted that the enthalpy rise factors are based on intergrals and are used as such in the DNB and LOCA calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in radial (x-y) power shapes throughout the core. Thus, the radial power shape at the point of maximum heat flux is not necessarily directly related to the enthalpy rise factors. The results of the loss of coolant accident analyses are conservative with respect to the ECCS acceptance criteria as specified in 10 CFR 50.46 using an upper bound envelope of 2.18 times the hot channel f&ctor normalized operating envelope given by TS Figure 3.12-8.
- When an qF measurement is taken, measurement error, manufacturing tolerances, and the effects of rod bow must be allowed for. Five percent is the appropriate allowance for measurement error for a full core map (2:40 thimbles monitored) taken with the movable incore detector flux mapping system, three percent is the appropriate allowance for manufacturing tolirances, and five per-cent is the appropriate allowance for rod bow. These. uncertainties are statistically combined and result in a net increase of 1.08 that is applied ta the measured value of Fq.
In the specified limit of [g there is an eight percent allowance for uncer-tainties, which means that normal operation of the core is expected to result in [g 5 1.55 (1+0.2 (1-P))/1.08. The logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power shape (e.g., rod misalignment) affect [g , in most cases without necessarily affecting F q, (b) the operator has a direct influence on F through q movenient of rods and can limit it to the desired value; he has no direct control over [g, and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests and which may influence F , can l q
l TS 3.12-17 ' between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial offset 1 which is defined as the difference in normalized power between the top and the bottom halves of the core. The permitted relaxation in F with decreasing power level allows radial AH power shape changes with rod insertion to the insertion limits. It has been determined that provided the above conditions 1 through 4 are observed, this hot channel factor limit is met. A recent evaluation of DNB test data obtained from experiments of fuel rod bowing in thimble cells has identified that the reduction in DNBR due to rod bowing in thimble cells is more than completely accommodated by existing thermal margins in the core design. Therefore, it is not nec-essary to continue to apply a rod bow penalty to F H' The procedures for axial power distribution control are designed to mini-mize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers. Basically, control of flux difference is required to limit the difference between the current value of flux dif-
- l ference (AI) and a reference value which corresponds to the full power equilibrium value of axial offset (axial offset = AI/ fractional power).
The reference value of flux difference varies with power level and burnup, but expressed as axial offset it varies only with burnup. The technical specifications on power distribution control given in Specification 3.12.B.4 together with the surveillance requirements given in Specification 3.12.B.2 assure that the Limiting Condition for Operation for the heat flux hot channel factor is set. 1 l
I
. TS 3.12-16 . be compensated for by tighter axial control. Four percent is the appropriate allowance for measurement uncertainty for F AH btained from a full core map (240 thimbles monitored) taken with the movable incore detector flux mapping system.
Measurement of the hot channel factors are required as part of startup physics tests, during each effective 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 core loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic incore mapping provides additional - assurance that the nuclear design bases remain inviolate and identify opera-tional anomalies which would, otherwise, affect these bases. For normal operation, it has been determined that, provided certain condi-tions are observed, the enthalpy rise hot channel factor F 1i'it "III AH 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 15 inches from the bank demand position. An indicated misalignment limit of 13 steps precludes a rod misalignment no greater than 15 inches with consideration of maximum instrumentation error.
- 2. Control rod banks are sequenced with overlapping banks as shown in TS Figures 3.12-1A, 3.12-1B, and 3.12-2.
- 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 l limits are observed. Flux difference refers to the difference i
t i l l l l l
l TS 3.12-18 The target (or reference) value of flux difference is determined as follows. At any time that equilibrium xenon conditions have been estab-l lished, the indicated flux difference is noted with the full length rod j control bank more than 190 steps withdrawn (i.e., normal full power opera-ting position appropriate for the time in life, usually withdrawn farther . as burnup proceeds). This value, divided by the fraction of full power at which the core was operating, is the full power value of the target flux difference. Values for all other core power levels are obtained by multiplying the full power value by the fractional power. Since the indi-cated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviation of +5% AI are permitted from the indicated reference value. During periods where extensive load following is required, it may be Lepractical to establish the required core conditions for measuring the target flux difference every month. For this reason, the specification provides two methods for updating the target flux difference. Strict. control of the flux difference (and rod position) is not as neces-sary during part power operation. This is because xenon distribution control at part 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 part power. Strict control of the flux difference is not always possible during certain physics tests or during excore detector calibrations. Therefore, the specifications on power distribution control are less restrictive during physics tests and excore detector calibrations; this is acceptable due to the low probability of a significant accident occurring during these operations.
l j
- TS 3.12-19 i
In some instances of rapid unit 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 dis-tribution sufficently 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, a limitati'on of one hour in any period of 24 hours is placed on operation outside the band. This ensures that the resulting xenon distributions are not significantly different from those resulting free operation within the target band. The instantaneous consequences of being outside the band, provided rod . insertion limits are observed, is not worse than a 10 percent increment in peaking factor for the allowable flux difference at 90% power, in the range + 13.8 percent (+10.8 percent indicated) where for every 2 percent below rated power, the permissible flux difference boundary is extended by 1 percent. As discussed above, the essence'of the procedure is to maintain the xenon distribution in the core as close to the equilibrium full power condition as possible. This is accomplished, by using the boron system to position the full length control rods to produce the required indicated flux dif-fer-nce. A 2% quadrant tilt allows that a 5% tilt might actually be present in the core because of insensitivity of the excore detectors for disturbances near the core center such as misaligned inner control rod and an error allowance. No increase in F occurs with tilts up to 5% because-misaligned q control rods producing such tilts do not extend to the unrodded plane, where the maximum qF occurs. l
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Unit No.1 Amendment No. 42 , Unit No. 2 Amendment No. 41 P00R D M K,
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CHANGE NO. 19
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. TS FIGL7E 1.12-5 1 CHANGE NO. 9 8-9-73 .
I CONTROL BANK INSERTION LIMITS FOR 2140P OPERATION WITH ONE BOTTO'CD ROD o.o - I I I l 1
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II I i I I I 1.o o.o o.2 c.4 o.s FRACTIC.1 FULL PO",2R l CHANGE NO. 9 (
~
CHN:GE No.9v
,- 8-9-73 I
CONTROL BA:!K II!SERTION LI!ITS FOR 2 LOOP CPERATIO!! WITH ONE I!! OPERABLE ROD VI i I i l l I i l i l I/l l I
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O . l 3 TS FICURE 3.12-7 12-2-77 DF.LETE 47 ( Unit 1 Amer _daent No. 31 Unit 2 Amend ==at No. 34 ( .
TS FIGURE 3.12-8 l l l l i HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE SURRY POWER STATION
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TS TIGURE 3.12-9 l 4
. 7-27-79
' 4, i
t 4 i DELETED l 1 ( 59 f 4 f Amendment No. 51. Unit I ( Amendment No. 50. Unit 2 ; l f i l l
TS FIGURE 3.12-10 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED POWER SURRY POWER STATION 120 _u; g g; _ ,g p. g c ._ __ s. _. _ __ s_ { ..__ _-=._ .: __:
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-50 -40 -30 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE (AI) %
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TS 6.6-9 The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event. (1) Reactor protection system or engineering safety feature instrument settings which are found to be less conserv-ative than those established by the technical specifica-tions but which do not prevent the fulfillment of the functional requirements of atfected systems. (2) Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation. Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurations as described in items 2.b(1) and 2.b(2) need not be reported except where test results themselves reveal a degraded mode as described above. (3) Observed inadequacies in the implementation of administra-tive or procedural controls which threaten to cause reduc-L tion of degree of redundancy provided in reactor protec-tion systems or engineered safety feature systems. l l (4) Abnormal degradation of systems other than those specified 1 ! in item 2.a(3) above designed to contain l
TS 3.12-1
! 25 70 I
3.12 CONDt0L RCD ASSDiBLIES AND POWER DISTRIBUTION LIMITS Applicability App as to the operation of the control rod assemblias and power di .- bution itaits. Objective To ensure cor subcriticality after a reactor trip, a e on potential reactivity inser ons from hypothetical control rod a embly ejection, and an acceptable co a power distribution during p er operation. , Specifiu tion A. Control Bank Insertion L es ,
- 1. Whenever the reactor is tic , except for physics costs and
( control rod assembly exerci s, the shutdown control rods shall be fully withdrawn. ,
- 2. Whenever the reactor critical, cape for physics tests and control rod assemb y exarcises, the langth control rod banks shall be acted no further than he appropriate limit determined b core burnup shown on TS Fi s 3.12-1A, 3.12-13, 3.12-2, 3.12-3 for three-loop operation an TS Figures 3.12-4A, 3.12-4 , 3.12-5, or 3.12-6 for two-loop operati
- 3. Th limits shown on TS Figures 3.12-11 through 3.12- may be l
evised on the basis of physics calculations and physic data obtained during unit startup and subsequent operation, in accordance with the following: l
- a. The sequence of withdrawal of the controlling banks, when j going from zero to 100% power, is A, 3, C, D.
- b. An overlap of control banks, consistent with physics cal-l 1
~ '
TS 3.12-2 7257';- culations and physics data obtained durine, Unit Startup and subsequent a operation, will be permitted. The shutdown cargin with allowance for a stuck control rod assembly hall be greater than or equal to 1.77% reactivity under all stea y-st.te operation ::ndicians, except for physi:2 to ts, fr;m :er to full ver, including effects of axial power distribution. . e shut-down mar, n as used here is defined as the amount by whi che reactor a core would suberitical at hot shutdown conditices ,y h547 F) if all control od assemblies were tripped, assumi g that the highest worth control rod ssembly remained fully withd sun, and assuming - no changes in xenon, baron.
~
54 4 Whenever the re?ctor is sube 'tical, except to physics tests, the critical rod position, i.e., the rod post-ion at whi criticality would be achieved if the control rod assemblies were ichd awn in normal sequence with no
! other reactivity changes, shall not b lower than the insertion limit f::
zero power. ,
- 5. Deleted 11r
- 6. Insertion limits do ne apply during physics tests o during periodic exercise of indivi al rods. However, the shutdown mar 'n indicated above must be maintal. d except for the low power physics test t ceasure control rod worth an shutdown margin. For this test the reactor may a critical with all b c one full length control rod, expected to have the h hest worth, nserted. hNh
/
.".... ad.. u i. Tie.50 Unit 1-l ?;;.. f :nt 90. 'O "ni t 2 -
1 i
l TS. 3.12-3 11 25-75
\ 7, DEIaT.
B. Pove Distribution Limits 1 At all times except during low power physics tests an implacen-cation of 3.12.B.2.b. (2), the hot channel factors define in the basis must meet the following limics: 1
.-- >____ u. sr g.
f
TS 3.11-4
' ~
Unit 1 Unit 2 3 15 33 Fq(Z) 12.05/P x K(Z) for P > 0.5 Fq(Z) 12.19/P x K(Z) for ? > o,3 h i Fq (Z) 14.10 x K(Z) for P 1 5' Fq .(Z) 14.38 x KCZ) for P 10.5 [h11.55(1+0.2(1-P))xT(3U) T H 11.55 (1+0.2(1-P)) x T(SU) .g N LOG ' < l.476/* Assa.
< l.38/P [AE Assm. .
T U F 1 1* #/P l 1 l'.45/P H d vhere P the fraction of rated power at wh'^ the re is operating, l K(Z) is the ' unction given in TS Figure 3.12-Ba f Unit 1 and Figure 3.12-8b for Un 2, Z is the core height loca n of Fq, and T(3U) is the interim thimb cell red bow penalty oE :{E siven 1: TS Figure i 3.12-9. I
- 2. Prior to exceedihg 75% ver folio - each core leading, and during i
each effective full power uth o operation thereaf ter, power distribu-l tion maps using the movable d cror system, shall be nade to confir= i l that the hot chant.al factor 4-4 of this specification are sacisfied. For the purpose of this afir=atio *
- a. The ceasurement a total peaking f tor, F(eas, shall be ir,'reased
\
by eight perce to' account for manuf turing tolerances, neasure-ment error, and the.* effects of red bow. e measurement of enthalpy rise hot channeT. factor, the hot assembly e halpy rise factor, shall be FfE ss., and the hot rod enthalpy rise facto 7{E . reased by four percent to account for measurem c' error. If any measured hot channel factor exceeds its limit speci ed under 3.12.3.1, the reactor power and high neutron flux trip etpoint , shall be reduced until the limits under 3.1.2.3.1 are met., If the hot channel factors cannot be brought to within the li=1ts 1 ted below within 24 hours, the overpower AT and Overtemperature AT t trip setpoints shall be si_ilarly reduced. innt:nt 2. SS , 'A f: ? a "'
. . . - , , . = . . . . . __ e
. . e .: . 4 %,
e 1e , , .. e. n.
, I rai: 1 Unit 2 FQ < 2.05 v.
K(I) FQ < 2.19 x K(Z)
~'
U_ u-
< 1.55 E < 1.55 s JT " LOCA
< 1.38 < 1.47
- !_'LOCA
= su. - e.a- Ass =.
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l TS 3.12-5 f: :? (Unit 1)
- 3. The reference equilibrium indicated axial flux difference (called the carget flux difference) at a given power level Po, is that /
/
indicated axial flux difference with the core in equilibrium xenon conditions (small or no oscillation) and theore control than rods m/ 1 0 steps withdrawn. The carget flux difference at an other power leve li d by the ratio, P, is equal to the carget value of P =ultip/ P/Pa . .e target flux difference shall be meas d at least onca per equiva ne full power quarter. The targe flux difference must be updated du each effective full powe nonth of operation either by actual easurement, or by lin ar interpolation using the most recent value the value pred cred for the and of the cycle life.
- 4. Except as modified by 3. 3.4 a, b, c., or d.below, the. indicated i
axial flux difference shal maintained within a 25% band about the target flux differe. e (deft es the carget band on asdal flux difference),
- a. At a power lav greater than 88 petcent of rated power, if M the indicat axial flux difference de ates from its target band, vi ir. 15 minute.4 either restore th indicated axial flu:c diffe nce to within the target band, or red a the reactor
--5+
p r to less than 88 percent of rated power. j b. At a power level no greater than 88 percent of rat power, (1) The indicated axial flux difference may deviate from its target band for a ==v4 m m of one hour (cumulative) in any 24-hour period provided the flux difference is within the limits shown on Figure 3.12-10. -P
' t '-
..t ":. "' , "si c 1
1 TS 3.12-5 i 5-15 SO I (Unic 2)
- 3. The reference equilibrium indicated axial flux difference (called the target flux difference) at a given power level Po, is that .
/
indicated axial flux difference with the core in equilibriu= xenon conditions (small or no oscillation) and the control rods = ore chan 19 steps withdrawn. The target flux difference at any.other power
/
- 1evel, ?, is aqual to the target value at ?c =ultip11.e by the ratio,
?/?o. . target flux difference shall be aasured at least once per aq=ivalast . pcver quarter. The target fl* difference sust be
=pda:ad d.::1:3 ch affective f.:11 power =c th /of operation either by ac:nal =aas=ra c, or by lisaar inte clacion using the most ranant value a:d the alue predicted or tha and of the cycle life.
- 4. I: cape as modified by 3.12.3.4.a, , c, or d beIov, the indicated crial flux differenca sha be incained withis a is: band'about the cargat flux differecca ines the target band on axial flux difference). -
9
- a. At a power 1sval eacar than 9 percent of rated. pcver, if a the indicated flux differene , deviates from its target band, with 15 minutes either re: tore the indicated axial flux l
t differen to within the carnet band, or educe the reaccer power to la chan 90 percent of rated pover.
- b. At power level no greater than 90 percent of aced power,
- 1) The indicated me st flux difference =ay devia a from its target band for a max 1=us of one hour (cumulative) in any 24-hour period provided the flux difference is within the li=1ts shown on Figure
-10.
l ( . l f end= tnt N;. 50, ' Jai; 2-r
TS 3.12-6 5 70 (Unit 1) One minute penalty is accumulated for each one minute of operation ' outside of the target band at power levels equal to or above _9 50% of raced power. (2) If 3.12.B.4.b(1) is violated, then the reactor power shall be reduced to less chan 50% power within 30 minutes a che high neutron flux setpoint shall be reduced to o -4 greater than 55% power within the next four hour (3) A power increase to a level greater than 88 pe enc of rated power is contingent upon the indicated axial lux difference being within its carget band. (4) rveillance testing of the Power Range autron Flux Cha als may be performed pursuant to able 4.1-1 provided the icated AFD is maintained wit in the limits of Figure 3.12,10. A total of 16 hours of peration may be accumulatad with the outside of the ta et band during this casting without p deviation. 8
- c. At a power level no reater th 50 percent of rated power, (1) The indicated 4 fl difference say deviate from its target band.
I (2) A power increase to a evel greater than 50 percent of rated power is e tingen upon the indicated axial flux difference not eing outsi its target band for more than one hou accu =ulated pa cy during the preceding 24-ho.t p .od. One half minu penalty is accumulated for'mac one minute of operation eside of the target band at p r icvels between 15% and 50% f raced power.
- d. The axi flux difference limits of Specif tions 3.12.B.4.a, b, an c may be suspended during the perfor- ce of physics casts pr ed:
The power level is maintained at or below 85. of raced power, and_ p I (2) The limits of Specification 3.12.3.1 are maint ed. The power level shall be determined to be < 85% f rated power j . at least once per hour during physics tests. Verif ation l l that the limits of Specification 3.12.3.1 are -being e shall be demonstrated through in-core flux mapping at least on e per 12 hours. i
"'m;L.10, M;1 l l
~ ,- TS 3.12-6 M 30 (Unit 2)
One =1=uce pecal'.y is accu =ulated for each one =i=ute of operation outside of the carget band at power levels equal to or above 50% of rated power. , (2) If 3.12.3.4.b(1) is violated, then the reac:or pcuec shill be. reduced to less than 50% power within 30 =1 utespa'nc the high neucron flux seepoint shall be reduced evno
/ -
greater chan 55: powerwithisthecextfourhou7s.
"O
- 3) A power increase :o a level greater than 90 perces: of rated power is con:1= gent upon the i= dica:ed axi . flux difference ei=g vi his its carget ba:d.
(4) 5- .=d*'= ce testing of the Power ?.a: Neutron Flux f" mix =ay be perfor=ed pursuan: - Table 4.1-1 provided ' 6.
.he d- 'cated AFD is safstained v :hin the li=its of Figure 3.12-10. A total of 16 hours operacion may be accumulated vi h the outside of the arget band duri=g this casting
-ithout penal deviacion.
- c. At a power level no eater as 50 percent of rated power.
(1) ne 1sdicated axia f' - differe=ce may deviace from its 1 s carget band. (2) A power increase oa vel greater chan 50 percent of raced power is entingen upon the indicacea axial flux
. difference e bei=g outs its target band for more than one ur accuculated pe :7 during the preceding 24-hour eriod. One half minut penalty is accu =n.tlated for ch one minute of operacion o eside of the target band at over levels be:veen 15% and 50%
- rated power.
- d. The al flux differe=ce 1** :s of Specifi tions 3.12.3.4.a, b, d c may be suspended duri:s the performa: e of physics casts p ovided.:
(1) The power level is mais:ai=ed at or belov 85% of rated power, and l (2) na 11=1:s of Specification 3.12.3.1 are maintai=ed. I
\
l The power level shall be deter =ined to be < 85% of raced - g power at least once per hour duri=g physics tests. Verifi-
\
cation that the 14 d es of Specification 3.12.3.1 are bein'
- st shall be de=custraced chrcush in-core flux =appi g at
( l 1 east once per 12 hours.
^ren&cnt 5. 53, " nit 2 1 1
l - . . l I TS g l Alatas shall normally be used to indicate the deviations from the swint flux difference requirements in 3.12.B.4.a and the flux difference time limits in 3.12.3.4.b and c. If the ala are ut of service temporarily, the axial flux difference shall e log ed, and conformance to the limits assessed, every ho for the f at 24 hours, and half-hourly thereafter. The indi tod axial flux difference for each excor channel shall be no tored at least once per 7 days wh the alarm is operable and at east once per hour for the rst 24 hours after restoring the al to operable status.
- 5. The allowable quadran to average powe tilt is 2.0%.
- 6. If, except for physics and arcise testing, the quadrant f
- to average power tilt exceeds , then:
- a. The hot channel factor shall e determined within 2 hours and the power level adjusted to et the specification of 3.12.B.1, or
- b. If the hot ch nnel factors are not det ed within two E
hours, th power level and high neutron f trip setpoint shall reduced from rated power, 2% for es percent of l 1 q anc tilt.
- c. the quadrant to average power tilt exceeds 10%, the -
power level and high neutron flux trip setpoint will b reduced from rated parer, 2% for each percent of quadrant tilt. 1
'--.O UO. 17, U it 1-
, ; "-- "- - t U: . -
':S, Unit 2 l
- - - - - -- ---"'N" ' ' ' ' - " '
. o TS 3.12-8 11 % 7; I
- 7. If, except for physics and rod exarcise casting, after a further
.G-period of 24 hours, the power tilt in 3.17.3.5 above is not cor rected to less than 2%:
. If design hot channel factors for rated power are not exceeded, an evaluation as to the cause of the dis .apancy a be made and reported as a reportabla occu ience to the . mar Regulatory Cc:mnission.
- b. If the d igu hot channel factors for rat power are exceeded and the pov is greater than 10%, the accisar Regulatory Commission s be notified and th Nuclear Overpower, Over-power AT and Ove.. eratura AT ..ps shall be reduced one percent for each per at the e channel factor exceeds the rated power design valu .
- c. If the hot channel fact s re not deternised tha Nucles:
Regulatory Commissic shall be notified and the Overpower AT and Overcamper ure AT trip ce tings shall be reduced by the equivalent of 2 power for every ' qu.adrant to average power tilt. C. Inoperable Contr Rods
- 1. A control od assembly shall be considered inoper le if the assemb cannot be moved by the drive mechanism, or e asse=bly re s misaligned from its bank by more than 15 inches. A
.211-length control red shall be considered inoparable if . s l
rod drop c1=a is greacar than 1.8 seconds to dashpot entry.
- 2. No more than one inoperabia control rod asse=bly shall be per-mitted when the reactor is critical.
- 3. If more than one centrol rod assembly in a given bank is out of service because of a single failure external to the individual rod driva mer har 4 ems, i.e. progra==ing circuitry, the provisions
<...u... ... e
. . j
. \
\
TS 3.12-9
,_g5_,a i
of 3.12.C.1 and 3.12.C.2 shall not apply and the reactor nay renain critical for a period not to exceed two hours provided insediata attentian Ls dir.atted troard nn.cing the nacessary epairs. In the event the affected assemblies cannot be eturned to arvice within this specified period the reactor w 1 be broug to hot shutdown conditions. 4 The prov. ions of 3.12.C.1 and 3.12.C.2 shall e apply during physics test I in which the assemblies are in neionally misaligned.
- 5. If .'n inoperab full-length red is loca d below the 200 stap level and is capa e of being tripped or if the full-length red is located below the O step level hether or not it is capable of being tripped, then e inse . ion li=its in TS Figure 3.12 ~
apply.
- 6. If an inoperable full-len h d carnot be locatad, or if the inoperable full-length ed is loc ted above the 3O step level and cannot be tripp .- then the inse tion li:- d es in TS Figura 3.12-3 apply.
- 7. Deleted
-44
- 8. If a f -length red becones inoperable and reac r operation is e ncinued he potential ejected rod worth and as ciated e ansient power distribution peaking factors shall be eter=ined by analysis within 30 days. The analysis shall include d a ellowance for non-uniforn fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more li=iting hypothetical transient than the cases reported in the i
safety analysis, the unit power level shall be reduced to an A:;;;;;;; N;. 50 , Uni; i - A:;;2:;;; M;. '? , 'J;it 2
TS 3.12-10
, '-25 '?
analytically determined part power level which is consistent with the safety analysis. - D. *f the reatt.,r is opar:cing above 75% of rsted pcwer with one n::ars nu ear channel out of service, the core quadrant power bai nee shall be de rmined.
- 1. Once er day, and
- 2. After a hange in power level greater than 10~ or more than 30 inches of c trol red motion. ,
The core quadrant over balance shall be de r=ined by one of the following methods:
- 1. Movable detectors (a least two p quadrant)
- 2. Core exit thermocouples (at le se four per quadrant)
; E. Inoperable Rod Position Indica e Channels
- 1. If a rod position indica or c unel is out of ser tice then:
- a. For operation het een 50% a 100% of rated power, the position of t. RCC shall be ch ad indirectly by cera instrument ion (excore detector a /or ther= occupies and/or vable incore detectors) eve shift or subsequent to ion, of the non-indicating rod, e. eding 24 steps, w chever occurs first.
j b. During operation below 50% of rated power no s cisi =oni-toring is required.
. Not more than one rod position indicator (RPI) channel p group nor two RPI channels per bank shall be permitted to be inop able at any time.
- 7. ".isalizned or 3roeced Control Red
- 1. If the Rod Position Indicator Channel is functional and the associated full length control red is = ora chan __58
. :.& :.;: .":. : , L',7l: l-l - ' r " zt 5. AQ , U,71; 7 i
l TS 3.12-11 1 7 25 70 15 inches out of alignment with its bank and cannot be realign , I then unless the hot channel fsetors are shown to be within esign ; 11=1ts as specified in 3ection 3.12.3.1 within S hours, over Sall be reduced so as not to exceed 75% of permitted power.
- 2. To crease power above 75% of rated pouer with a
--5&
full 1 geh control rod more than 15 inches ou of alignment with its nk an analysis shall first be ma to determine the hot channel ccors and the resulting al able power level based on Sectio 3.12.B. Basis The reactivity control concept su=ed # r operation is that reactivity thanges accompanying changes in reactor p are compensated by control red assa bly motion. Reactivity changes asso 4 te with xenon, samarium, fuel da;1ation, and large changes in reactor c lant te_ rature ('operacids temperatura to cold shutdown) are compensa ed for by chang _ in the soluble boren ::stan-tration. During power o eration, the shutdown roups are fully withdrawn - and control of power s by the control groups. A eactor trip occurring during power oper ion will place the reactor into th hot shutdown etedition. The control re assembly insertion limits provide for ac eving hot shutdown by reactor rip at any time, assuming the highest worth con el rod assembly remains ully withdrawn, with sufficient margins to meet the a u=ptions used in t accident analysis. In addition, they provide a limit on c . maxi =um acted rod worth in the unlikely event of a hypothetical assembly dection, and provide for acceptable nuclear peaking factors. The limit may be d ar-mined on the basis of unit startup and operating data to provid's a = ora realisti: li=it which will allow for = ore flexibility in unit operstion and A;;;d;;;t NO. 50 , Unit 1 A;;nd;;nt M . '9 , Unit 2
~
TS 3.12-12 7 :: 70 still assure compliance with the shutdown requirement. The =axi=u= shut-8 down =argin requirement occurs at end of core life and is b'ased on the lue used in the analysis of the hypothetical steam break accident. The rod insertion limits are based on end of core life conditions. The shu - devn : egin for the cncire cycle length is estsblished at 1.77 rea civity. All ocher ccident analyses with the exception of the chemical an volume control syst i =alfunction analysis are based on 1~ reactivity shutdown margin. Relative positions f control rod banks are determined a specified control rod bank overlap. Th overlap is based on the cons
- eration of axial ,
power shape control. The specified control rod i ertion limits hav been revised to limit the potential ejected rod worth in eder to acc at for the ef fects of fuel densification. ( The various control rod assemblies ( utdown banks, centrol banks A, 3, C and D) are each te be moved as a ank, that is, with all assemblies in --4HF the bank within one step (5/8 ch) of the ank position. Position indication is provided by tw methodst a dig. al count of actuating pu*ses . which shows the demand p ition of the banks and linesi position indicator, Linear Vard ble Diff erential Transformer, hich indicates the actual assembly po . tion. The position indication accu acy of the Linear D' ferential Transformer is approxi=acely + ~ of span (+7.5 inches under steady state conditions. The relative a uracy of , l the linea position indicator is such that, with the most adver errors, l i l an al is actuated if any two asse=blies within a bank deviate b = ore l th 14 inthes. In the event that the linear position indicator is n t n service, the effects of ( a:::::::t n:. :: , uni: n A : f ::t M:.10 , " it 2
l i l TS 3.12-13 I
- t ,c_-n,
/
calpositioned centrol rod asse=blies cre cbsertable from nuclec: and process'
/
for=c:fon disp 1:yed in the Main Cen::a1 Roc = and by core cher: occupies d in ors =cv:ble da:ac:c:s. Ecicu SC" power, no speci:1 =ecitoring is -inquired for : 1 positioned cent ci rod esse =blics vi:h inoperable red posi:iot indicsters b cam, ev:.: 1:'.: :n unna -tw; ca.--'..e ca ;;;c-hi;. cirelit; =sa:
-SS.
(full leng- control rod asse=bly 12 feet out of ali;ncen: vi:h .s bank) opers-tion at SC" teady state power does no: result in exceeding c e li=1:s. The specified co .ol rod asse_bly drop ti=e is consista. wi:h safety analyses that have been per: r=ed. An inoperable ecntrol re asse=bly i= poses additi .a1 de=sads on the operators. The per=1ssible number of oper:ble control re ' asse=blias is li=1:ed te one in order to li=1t :he =sgnit de of the opera g burden, bu: such a failure would not prevent dropping of he operable out cl red asse:blies upcn rea::c trip. e Two criteria have been chosen :s a ign basis for fuel perfor ance rela:ed to fission gas release, pellet ce:pe cure d cladding =echanical proper:1es.
\
First, the peak value of fuel e terline perature cust =ct e?:ceed i~MC7. Secend, the =1:1=u= DER in e core cust no be less than 1.30 in c: :*.. operation or in short ter= ransients. In addition :c the abo e, the peak linear power duc icy, the nuclear enthalpy rise hot channel factor, d the hot asse=bly enthalpy ris fac:or =ust not at:eed thei- limiting v *'es which result f:c :he largs break toss of cools : accident analysis based n the ECCS acceptance criteria li=it of 2_ CCF on peak clad tecperature. This is required to =eet the i=itial cocdition assu=ed fo: the loss of co anc accident. To aid in specifying the 11=1:s on var distributics the fall .-ing hot channel factors are defined.
? .;;i;;t '!0. 50 , "-it '
s~. _=, un aa ' Pit 2 x I . l l l l
TS 3.12-14 5-15 SO Fq(Z), Height Decendent Heat Flux Hot Channel Factor, is defined as the =axi=u= t local heat flux on the surface of a fuel rod at core elevation Z divided by th a rage fuel rod heat flux, alleving for manufacturing tolerances on fuel poll s and rods. . Ff,Eng. eering Heat Flux Hot Channel Factor, is defined as the allev nee on heat flux quired for zanufacturing tolerances. The engineering actor allows for local var. tions in enrichment, pellet density and dia=ete , surface area of the fuel rod a d eccentricity of the gap between pellet d clad. Combined statistically the ne effect is a factor of 1.03 to be a lied to fuel rod surface heat flux. FNg , Nuclear Ee-haley Rise .ot Channel Factor,.is 'ined as the ratio of the integral of linear power alon the rod with the ighest integrated power to the average rod power for both LOCA d non-LOCA onsiderAtions. L A F 4H Assa., Hot Assembly Nuclear Ent aluy inse Factor, is defined as the ratio
; of the integral of linear power along e assembly vich'the highest integrated power to the average assembly power ,
It should be noted that the enth py rise fa ors are based on integrals and are used as such in the DNB a d LOCA calculatio . Local heat fluxes are obtained by using hot ch al and adjacent channel licit power shapes which take into account varia 1ons in radial (x-y) power sha es throughout the core. Tius, the radial pow r shape at the poin. of max 1=um hea flux is not necessarily l directly related a the enthalpy rise factors. The results f the loss of coolant accid c analyses are conservative with respect to the CCS acceptance criteria a specified in 10 CFR 50.46 using an upper bound envelo of 2.05 (Unit 1) or 1.19 (Unit 2) times the hot channel factor nor=alized op ating
--62>
enva pe given by TS Figures 3.12-8a and 3.12-8b.
/
(
?::ndment No. 52, 'Jr.it 2 l
l
; TS 3.12-15
' 27 ??
l When an Fq measurement is taken, measurement error, manufacturing tolerances, ) f and the effects of rod bow must be allowed for. Five percent is the app - priate allowance for measurement error for a full core map (240 thimb s itored) taken with the movable incore detector flux mapping sys m, three l pere c is the appropriate allowance for manufacturing toleranc , and five percent s the appropriate allowance for rod bow. These une tainties are j statistical combined and result in a net increase .of 1. that is applied to the measure value of F q. In the specified mit of T there is an eight per at allowance for uncer-H tainties which means hat normal operation of th core is expected to result in F $ 1.55 (1+ 0.2 ( P))/1.08. The logic ehind the larger uncertainty - , in this case is that (a) a 1 perturbati s in the radial power shape (e.g., rod misalignment) affe F , in se cases without necessarily affecting Fq , (b) the operator ha a direct influence on F through movement q of rods, and can limit it to the red value, he has no direct contre' over(H,and(c)anerrorin a pred tions for radial power shape, which may be detected during sta physics tesgs and which may influence Fq can
. be compensated for by ti ter axial control. Four percent is the appropriate allowance for measur ent uncertainty for * "" '# * * " ** **E H
Q40 thimbles mon ored) taken with the movable inc re detector flux mapping system. The values specified for the limits of and *** * * ** "** H Ro H Ass used the LOCA analysis. It has been determined that four reent is the appr priate allowance to be applied for measurement uncertainty ,r each of ese parameters. Measurement of the hot channel factors are requi d as part of startup physics tests, during each effective full power month l operation, f..; t at M:. 51, "n f t '
"n . t xt M:. 50 , " cit 2
- 1 1
I IS 3, g l,13 1 and whenever abnormal power distribution cor.ditions require a reduction o i core power to a level based on measured hot channel factors. The incor taken following core loading provides confirmation of the basic ear des bases 4aatad4=- proper fuel loading patterns. The perio incore
, aspy provides additional assurance that the nuclear design asas remain inviolate end identify operational ==a==14es which would, hersise, affect these bases.
1 n_,. t .___>- 7; , , rote 1 i-~d-- t Mr . 2 6
TS 3.12-16
' 25 70 For nor=al operation, it has been deter =ined thst, providad certain c di-N tions are observed, the enthalpy rise hot channel factor, F'cg, l' d will be net; these conditions are as follous:
l
- 1. Control rods in a sing *a back nave too e:her :1:h : ' indiriuud j l
red insertion differing by more than 13 inches . rom the bank demand position. An indicated misalign=eaa of 13 steps p ecludes a red misalignment no greater t
/ 15 inches with cons eration of m
- u= instrumentati error.
- 2. Contro.1 ed banks are sequenced wit overlapping banks as shown in TS Fig s 3.12-1A, 3.12-13, d 3.12-2.
- 3. The full long 52 control bank . ercion li=1ts are not violated. 4 s
D
- 4. Axial p er distribution control p cedures, which are given in to.ms of flux difference control and nerol bank insertion 1 ts are observed. Flux difference re rs to the difference etween the top and bottom halves of two-se ion excore neutren detectors. The flux difference is a =sasure o the axial offset which is defined as the difference in nor=ali:ad war between i
the top and bottom halves of the core. N The pernitted relaxation in F~ag with decreasing power level allow radial power shape changes with rod insartion to the inser ion 11=its. It s t ( been deter =ined that provided the above conditions 1 through 4 are obsa. ad, this het chs nel factor limit is net. i f ::f ::t Me. 50 , " it i l ?c f ::t M0. SO . Sit 2 t l
TS 3.12-162 7 27 70 A recent evaluation of DNB test data obtained fros experiments of fuel
/
rod bowing in thimble cells has identified that the reducticn in DNBR due e rod bowing in chimble cells is more than completely .accommo'date'd'by exis ng thermal margins in the core design. Therefore, it is not nec-N essary continue to a'pply a rod bow penalty to FAH* 9 e O e J O e
*= 45 1.45h 51 e g W D
. .;;i ;;t M0. ;; , 'Jnit 2
- T3 3.12-17
+
g f_Tf OA
- =w ww 4
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-l -
/
J j e p 4
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DELETED ! = W e t e 1 j
- 1 4
l J i I f i 4 ? 9 1
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- w- m 4 _ -,.4-.-.;- -,.o.-- mA.F5---- ,-- s_.__h44 -. A A a J ._m .. A.
t. g TS 3.12-18 e , z em e av vv S
).
r
/
/
3
- e k
1 4 6 9 1 9 DELETED . g _. i a O h. 8 l-s \ l (
,__.2__. ._ .. ..m ,
wvg J' 5 4 b b I l r 4 - ,o--O- - - + , ,ww>o.m ----og,ys.e gm., ,-m~nw--.m e-o gem-g
TS 3.12-19 . e 4e on j l I s DU.1TE s
.9 i
1 The procedures fo ar d 21 power distribution couer are designed to nini-nize the effe s of xenon redistribution on the av d a power distributice dur1=g los follow naneuvers. Basically, control of f1 difference is require to 1 *-4 e the difference between the current value of flux dif-fare a (i!) and a reference value which correspends to the . il power e 111brium value df axial offset (axial offset = AI/ fractional over). , 1 l The reference value of flux diff erence varies with power level and ruup, but expressed as axiai. offset it varies only vich bc7up. ( A:cr.i;;r.t "c. . 53, "mit 2
T3 3012-20 5 15 30 9 The technical specifications on power distribution control given in
' 3.12.3.4 cogether with, "the surv'e111ance requirements given i= 3.1:4.3.2
~
assure that the Li=' ting Condition for Operacion for the heat flux he hannel factor is =et. Tha : et (or reference) value of flux difference is decer- ad as fci". us. : any -de : hat equilibrics xenos conditices 5 va been estab-li.shed, the -dica:ed flux differase is noted with the full length red c= - 21 ba=k : . *-=- 190 steps withdrawn (i.e. mal nor/full pcuer opera- .
-'- ;:siti:. ap;.. riate f'or the ti=a in life, /ually withdrawn farther as . f ;;oceeds). value, divided by he fraction of full power a: 9'A the core was ope ating is the f ' power value of the target fit == difference. Values for all other ore power levels are obtained by
=ul:1 plying the full power valu by he fractional power. Since the indi-( ca:ed equilibric= value was note , no allevances for excere detector err.= are necessary and indic. ed dev tion'of j-5* AI are per=1tted from the i=diha:ed reference va e. During p tods where ex:ensive load fo11c4sg is required, may be imprac:1c co establish the required core condi: ions for easuring the carget flux #ference every month.
For _his reason, e specification providas eve =e ods for updating the target flux di ference. Strict e neral of the flux difference (and rod position) is ~ot as neces-sary ring part power operation. This is because xenon diser ution ce crol at part power is not as significant as the control at f I l (
, _ _ _ __ ~ m -. - w - --
TS 3.12-21
- : 7?
(Unit 1) e power and allowance has been made in predicting the heat flux peak.1=g actors for less strict control at part power. Strict control of the flux d arence is not always possible during certain physics tests or duri excor detector calibrations. Therefore, the specifications on pov . W distribu don control are less restrictive during physics tests a excore detector c brations; this is acceptable due to the low prob 111:7 of a signi-ficant acciden occurrfag during these operations. In sone instances o rapid unit power reduction auto ic red notion will cause the flux differ ce to deviate fron the targ band when the reduced power level is reached. ds does not necessard y affect the xenon dis-tribution sufficiently to ch ge the envelop ,of peaking factors which can be reached on a subsequent eturn to 8411 power within the target band, however to si=plify the specifica den a li=ication of one hour in any t period of 24 hours is placed on ope d en outside the band. This ensures that the re2ulting xenon distrib ions a e not significantly different fron those resulting fron ope tion within he target band. The instan-taneous consequences of b g outside the band provided red insertion li=its are observed, is oc worse than a 10 perce - incranent in peaking factor for the allow le flux difference at 88% powe in the range +13.5 percent ($10.3 pa cent indicated) where for every 2 per e below raced -O power, the pa sible flux difference boundary is extende by 1 percent. As discus d above, the essence of the procedure is to naistain the xenon distri tion in the core as close to the equilibriun full power co dition _t r 2---t "- '; ? , "-2 1 t
~
'. T3 3.12-21
. 5 :.5 30
. (Unit 2) '
- pover and allevance has been e.ade in predicting the heat flux pe=kd g factors for less strict control at part power. Strict control of the flux diff erence is not always possible during certain physics test or i
d ing excore detector calibrations. Therefore, the specificacd us on l pover distribution centrol are less restrictive during physi tests and a.:nore d tector calibrations; this is acceptable due to e low probabili-cy cf a si #1ca=t accident occurring during these op ations. 2 sena ins m ess f :apid , unit power reduction tocatic rod notion vill ca=se _he flun diffs. ace to deviate from the arget band when the reduced pcvar level is reached. This does not nec sarily affect the xenon dis-
- i'ction
- sufficiently to e ge the en ope of pahing factors which'
.can be reached on a subsequent acu. to full power within the target g band; however, to s1 plify the sp ification, a limitation of one hour in any pe: Lod of 24 hours is plac on eration outside ,the band. This ensures;that the resulting anon distrib tions are not significantly ,
different fres chose res ting from operat. a t.-ithin the target band. The instantanecus con aquences of beirs outsi che band, provided red insertion 11 sits e observed, is not verse than 10 percent increment 1: panki,g fact r for the allowable flux difference 90:: power, in the
-fr5-range + 13.8 percent -Q10.8 percet indicated) where fo every 2 percent l
[ halow rat d power, the permissible flux difference bounda. is az: ended by 1 p reent. , discussed above, the essence of the procedure is to maintain the anon distribution in the core as close to the equilibrium full power conditi n ( l f;; ,t:nt 2. 2, 2M 2 1
. s TS 3.12-22* j 11-25 75
(' as possible. Tht:2isaccomplished,byusingthebaronsystemtopositick f-N<the erence. full length control rods to produce the required indic
-e '
DELETED A2 quadrant tilt allows that a 5% tilt might actually be present ir. the core because of inten civity of the excore detectors for disturbances near the core center s as misaligned er control rods and an error allowance. No increase in occurs th tiles up to 5% because misaligned i control rods producing such s a not extend to the unrodded plane, ( where the ==v4== Fq occurs. 4 G G (
- - m ; ::. 2s l
- ~ . . - - . - ., - ,.
. TS Table 3.12-1A 3 -i ? 7^
\
,l REMOVE -
THIS TAB HAS BEEN DELETED. 37 N l l i k a-^-t l a-t-Me . 4 9, "-
"-- '-- t Me. 4 8, Sit 2
I l TS Tablo 3.12-13
- , , em e aw wv 1
i .i TEMOVE I- . TEI TA3LI HAS 3EIN IIITED. 4
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_ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ . _ _ _ . _-- - - - - - - v
TS Table 3.12-2
.e a , a,
,t.
REMOVE _ + . THIS TABLE BEEN 71ETED.
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- f. A , Th e =. 1-l
, . _ . . - _ - _ , . . _ _ . . - _ _ ., y
TS FIGURE 3.12-8a e ,, m t i HOT OIANNEL FACTOR NORMALIZED OPERATING ENVELOPE SURRY POWER STATION UNIT NO. 1 -
-65 GEE ATTACHED PAGE
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TS FIGURE 3.12-8)I e ,
. .1 .m.
(. HOT CHANNEL FACTOR NCRp LIZED OPERATING ENVELOPE SURE POWER STATION UNIT NO. 2 c,EE ATTACHED PAGE h._-_._._....._.%... g -e __
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HOT CHANNEL FACTOR NORMALIZLD OPERATING ENVELOPE . SURRY POWER STATION l 4
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l AXIAL FLUX DIFFEFINCE LIMI*S 1 AS;A FUNCTION OF RATED POWER gm pv prtere gas =w , l 120 1 _; . g, t N. !;;Eas j c'$$ ,
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UNACCEPTABLE UNACCEPTABLE CPERATION i 'v OPERATION
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FLUX DIFFERENCE (AI) % l ( i--- a-- e ve ze. "-i- 1 hk '];}=?'"n =e
. TS FIGURE 3.12-10 e ,e n
. GEE KTTACHED PAGE iJL'i) t t
AXIAL FI.UX DITTIFINCE I.IP.ITS AS A FUNCION OF RATED POWER SURRY PCL'ER STATION l L* 3 g l .:_- :-~ - --
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FIUI DIFFIRINCE (AI) l ("
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1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED POWER SURRY POWER STATION l 120 =:= ; n=_u -m - - : art -- - _= ir==:;= ;_.2=:_x== ;._x.
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-50 -40 -30 -20 -10 0 10 20 30 40 50 1 FLUX DIFFERENCE (AI) %
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o .. TS 6.6-9 11 Of 7f I The written report shall include, as a cini=u=, a cocpleted copy of a licensee event report form. lafor=ation provided on the licenses event report form shall be supplemented, as needed, by additional narrative caterial to provide complete explanation of the circu= stances surrounding the event. (1) Reactor protection system or engineered safety feature instru=ent settings which are found to be less conserv-ative than those established by the technical specifica-tions but which do not prevent the fulfill:ent of the functional require =ents of affec' ed t systems. (2) Conditions leading to operation in a degraded code permitted by a limiting condition for operation or plant shutdown required by a li=iting condition for ( ' operation. Note: Routine surveillance testing, insert =ent calibration, . or preventative maintenance which require syste: configurations as described in items 2.b(1) and 2.b(2) need not be reported except where test results the=selves reveal a degraded code as described above. S;;;ific-/
-M-th; isp1;c.;r.t;tica ;f 2.12.2.2.b . (2) i; ne; ;;;;;;;bi;.
(3) Observed inadequacies in the i=plecentation of administra-tion or procedural controls which threaten to cause reduc-tion of degree of redundancy provided in reactor protec-tion systems or engineered safety feature syste=s. (4) Abnor=al degradation of systems other than those specified t I' in item 2.a(3) above designed to contain _ -w ew l l l _ ._ -}}