ML19350F162
| ML19350F162 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 06/16/1981 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19350F163 | List: |
| References | |
| NUDOCS 8106240299 | |
| Download: ML19350F162 (27) | |
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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. OPR-32 1.
The Nuclear Regulatory Commission (the Cormission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated April 28, 1981, as supplemented May 15, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issua'nce of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-32 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through AmendmGnt No. 70, are-hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Spacifications.
3.
This license amendment is effective as of the date of its issuance.
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FOR THE FUCLEAR REGULATORY COMMISSION
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' Operating Reacto Branch No. 1 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
' JUN 16 1931 O
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3, gr VIRGINIA ELECTRIC AND POWER COMPANY 00CKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. DPR-37 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated April 28, 1981, as supplemented May 15,1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the' Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; L.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part
~
51 of the Commission's regulations and all applicable requirements have been satisfied.
2-2.- Accordingly, the license is amended by changes tc the Technical Specifications as indicated in the attachment to this license
' amendment, and paragraph 3.8 of Facility Operating License No. DPR-37 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contnined in Appendix A, as revised through knendment No. 70, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This ifcense amendment'is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION b!
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Changes to the Technical Specifications Date of Issuance:
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V ATTACHMENT TC LICE"SE AMENDMENTS AMENDMENT N0'. 70 TO FACILITY OPERATING LICENSE NO. DPR-32 A?'ENDMENT NO. 70 TO FACILITY OPERATING LICENSE NO. OPR-37 DOCKET NOS. 50-280 AhD 50-281 Revise Appendix A as follows:
Remove Pages Insert Pages 3.12-1 3.12-1 3.12-2 3.12-2 3.12-3 3.12-3 3.12 4 3.12-4 3.12-4a 3.12-4b L 12A 3.12-5 i
3.12-6 3.12-6 3.12-7 3.12-7 3.12-8 3.12-8 3.12-9 3.12-9 3.12-10 3.72-10 3.12-11 3.12-11 3.12-12 3.12-12 3.12-13 3.12-13 3.12-14 3.12-14 3.12-15 3.12-15 3.12-15a 3.12-16 3.12-16 3.12-16a 3.12-17 3.12-17 3.12-18 3.12-18 3.12-19 3.12-19 3.1240 3.12-21 3.12-22 6.6-9 6.6-9 TS Table 3.12-1 TS Table 3.12-1 A TS Table. 3.12-18 TS Table 3.12-2 TS Figure 3.12-8(Unit 1)
TS Figure 3.12-8(Units 1 and 2)
TS Figure 3.12-8a(Unit 2)
TS Figure 3.12-8b(Unit 2)
TS Figure 3.12-10 TS Figure 3.12-10 i
1 l
T~ 3.12-1 3.12 CONTROL RCD ASSEMBLIES AND POWER DISTRIBUTION LIMITS Applicability Applies to the operation of the control rod assemblies and powtr distri-bution Ifmits.
Objective To ensure core subcriticality after a reactor trip, a limit on potential reactivity insertions from hypothetical control rod assembly ejection, and an acceptable core power distribution during power operation.
Specification A.
Control Bank Insertion Limits 1.
Whenever the reactor is critical, except for physics tests and 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, 3.12-1E, 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 brsis of physics calculations and physics data obtained during unit startup and subsequent operation, in accordance with the following:
a.
The sequence of withdrawal of the centrolling banks, when going from zero to 100% power, is A, B, C, D.
b.
An overlap of control banks, consistent with physics cal-Amendment Nos. 70 & 70
- u TS 3.12-2 l
enlations 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 pcwer, including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be suberitical at
- hot shutdown conditions (T,yg 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.
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 exercise of individual rods. However, the shutdown margin 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 length control rod, expected to have the highest worth, inserted.
Amendment Nos. 70 & 70 I
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7-S.12-3 B.
Power Distribution Limits 1.
At all times except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:
F (Z) $ 2.18/P x K(Z) for P > 0.5 q
F (Z) 5 4.36 x K(Z) for P 5 0.5 9
F' S 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 75% power following each core loading and during each effective full power month of operation thereafter, power distri-bution maps using the movable detector system shall be made to confirm that the hot channel factor limits of this specification are natis-fled. For the purpose of this confirmation:
eas a.
The measurement of total peaking factor T shall be increased by eight percent to account for ma'n'ufacturing tolerances, measure-ment error and the effects of rod bow. The measurement of enthalpy rise hot channel factor F shall'be increased by four percent to g
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)q 5 2.18 x K(Z) and T
$ 1.55 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Overpower AT and Overtemperature AT trip setpoints shall be similarly reduced.
Amendment N6s. 70 & 70
IS 3.11-4 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 Pg 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 l
recent value and the value predicted for the end of the cycle life.
4.
Except as modified by Specifications 3.12.B.4.a, b, c, or d below, the indicated axial flux difference shallJbe 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 b,and, 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 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.
Amendment No. 70 & 70
T: 0.12 ~
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 Lucrease 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 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of operation may be accumulated with the' aIxial fldx 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.
Amendment Nos. 70 & 70
IS 3.12-o d.
The axial flux difference limits for Specificatiens 3.12.B.4.a, b, and c may be suspended during the perfoesance 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 shsil 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 *t least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Alarms shall normally be used to indicate the deviations from tne axial flux difference requirements in Specification 3.12.B.4.a and the flux difference time limits in Specifications 3.12.B.4.b ano 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and half-hourly thereafter.
The indicated axial flux difference for each excore channel shall be monitored at least once per 7 days when the alarm is operable and at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the alarm to operable status.
The allowable quadrant to average power tilt is 2.0%.
6.
If, except for physics and rod exercise testing, the quadrant to average power tilt exceeds 2%, then:
l l
l Amendment Nos. 70 & 70
70 0.10 7 The hot channel f actors shall be determined wi, thin 2 nours a.
andthepowerleveladjustedtomeettherequirementofSpecifi-l cation 3.12.3.1, or b.
If the hot channel factors are not determined within two hours, the p'ower 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 exceeds +10%, the power level and high neutron flux trip setpoint will be reduced from rated power 2% for ea~h percett of quadrant tilt.
7.
If, except for physics and rod exercise testing, af ter a further period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 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 !setors are not determined the Nuclear Regulatory Commission shall be notified and the Overpower Amendment Nos. 70 & 70
TS 3.12-5 aT and Overtemperature,aT trip settings shall be reduced by the equivalent of 2% power for every 1% quadrant to average power tilt.
C.
- 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 cut 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 no,t 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 returned to ser'vice 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:
a.
If an inoperable full-length rod is located below the 200 step level and is capable of being tripped, or Amendment Nos. 70 & 70
I; 3.11-B b.
If the full-lengt' 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 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 safety 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 b.
After a change in power level greater than 10% or more than 30 inches of control rod motion.
Amendment No. 70 & 70
TS 3.12-10 i
)
i 2.
Tne core quadrant power balance shall be determined by one of the following methods:
Movable detectors (at least two per quadrant) a.
b.
Core exit thermocouples (at least four per quadrant)
E.
Inopera,ble 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 ahall be checked indirectly by core instrumentation (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 p'er bank shall be permitted to be inoperable at any time.
F.
Misaligned 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 specified in Specification 3.12.B.1 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, power shall be l
reduced so as not to exceed 75% of permitted power.
j Amendment No. 70 & 70 l
i T4 3 17-11 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 reactivit'y 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 tuto.the hot shutdown condition.
The control rod assembly insertion limr_a 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 Amendment No. 70 & 70
os a.te-A4 still assure ccepliance with the snutoown 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 analyscs 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 bankt; 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 indicates the actual assembly position. The position indication accuracy
,of the Linear Differential Transformer is approximately +5% of span
(+ 7.5 inches) under steady state conditions. The relative accuracy of the linear position indicator is such that, with the most adverse errors, an alarm is actuated if any. two assemblies within a bank deviate by more than 14 inches.
In the event that the linear position indicator is not Amendment Nos. 70 & 70
TS 3.12-13 in service, tne 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 assemb'ier with inoperable rod position indicators because, even with an unnoticed complete assembly misalignment (full length control rod assembly 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 te Ibnit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable control rod assemblies upon reactor trip.
1 Two 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 4
operation or in'short term transients.
Amendment Nos. 70 & 70
TS ? 12-1^
In addition to the above, the peak linear power density and the nuclear enthalpy 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 *.olerance on fuel pellets and rods.
F, Engineering Heat Flux Hot Channel Factor, is defined as the allowance on q
heat flux required for manufacturing tolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area l-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.
ThH, Nuclear Enthalpv 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.
Amendment Nos. 70 & 70
t ie 73 3.12-13 It should oe noteo enat the enthalpy rise factors are b4 sed ou integrals and are useo 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 intio 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 factor normalized operating eevelope given by TS Figure 3.12-8.
1 j
When an F measurement is taken, measurement error,. manufacturing tolerances, q
and the effects of rod bow must be allowed for. Five percent is the y
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 tolerances, and five per-cent is the appropriate allowance for rod bow. These uncertainties ara si.atistically combined and result in a net increase of 1.08 that is applied to a
the measured value of F.
q In the specified limit of F there is an eight percent allowance for uncer-tainties, which means that normal operation of the core is expected to result in F 5 1.55 (1+0.2 (1-P))/1.08. The logic behind the larger uncertainty
!~
i.
in this case is that (a) normal perturbations in the radial power shape (e.g., rod misalignment) affect F
, in most cases without necessarily affecting F, (b) the operator has a direct influence on F through movement i
q q
of rods and can limit it to the desired value; he has no direct control over F 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 4
i l
Amendment Nos. 70 & 70
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I 3.12-!?
be compensated for by tigater. axial cu4Lrol. Tous y.. cent is the appropriate allowance for measurement uncertainty for T obtained from.a. full coreumap.
(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.
I 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 anthalpy rise hot channel factor F li"It "ill AH be met.
These conditions are as follows:
1.
Cont.:ol 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 oss: lapping 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 limits are observed. Flux difference refers to the difference 1
Amendment Nos. 70 & 70
_ _ - ~, _. _ _,. -
TS ?.12-17 between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and the bottom halves of the core.
V The permitted relaxation in FhH with decreasing power level allows radial power shape changes with rod insertion to the insertion limits.
It has been determined that provided the above conditions I 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 thimele 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'N aH*
The procedures for axial power distribution control are designed to mini-1 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-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 met.
Amendment Nos. 70 & 70
TS 3.12-1A The target (or reference) value of flux difference is determined as follows. At any time that equilibrium xenon conditions have been estab-lished, the indicated flux difference is noted with the full length rod 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 deviations of +5% 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 ev:ry 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 Amendment Nos. 70 & 70 1
E5 3. ii-6 in some instances of rapid unit power reduction automatic rod motion vili cause the flux difference to deviate from the target band when the reduced power level is reached. This does not necessarily affect the zenon 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 limitation of one h w e 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 on operation outside the band. This ensures that the resulting xenon distributions are not significantly different from those resulting from 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-ference.
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 9
control rods producing such tilts do not extend to the unrodded plane, where the maximum F occurs.
q Amendment Nos. 70 & 70 i
Tc f f o The written report shall include, as a minimum, a co=pleted 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 affected systems.
(2) Conditions leading to operation in a degraded mode permitted by a limiting condition _for. operation or plant snutdown required by a limiting condition for operation.
Routine surveillance testing, insth$ ment calibration, Note:
or preventative maintenance which require system 1
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-tion of degree of redundancy provided in reactor protec-tion systems or engineered safety feature systems.
(4) Abnormal degradation of systems other than those specified in item 2.a(3) above designed to contain l
l Amendment Nos. 70 & 70 i
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300R BRIGM Amendment Nos. 70 & 70
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AIIAL FLUI LIFFERENCE LIMITS i
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Amendment Nos. 70 & 70