ML20043J001
| ML20043J001 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/21/1990 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20043H999 | List: |
| References | |
| GL-88-16, NUDOCS 9006270182 | |
| Download: ML20043J001 (35) | |
Text
_ . . - _ . . _ , _ _ _ _ - _ . _ _ _ _ _. _ _ _ _ . _ - - _ _ _ -_.
. . l l
i t
ATTACHMENTI TO IPNW034 PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING REMOVAL OF CYCLE SPECIFIC PARAMETER UMITS IN ACCORDANCE WITH GENERIC LETTER 8816 i
NEWYORK POWER AUTHORITY :
INDIAN POINT 3 NUCLEAR POWER PLANT DOCKET NO,50 286 DPR 64 I bII ,
P 4
- ,e---. w
4 9
Section Title Egge 6,9 Reporting Requirements 6 13a i Routine Reports 6 13a Special Reports 6 17 6.10 Record Retention 6 17 6.11 Radiation and Respiratory Protection Program 6 19 6.12 High Radiation Area 6 20 6.13 Environmental Qualification 6 21 i
t l
l 4
t l
- i iv .
a h
Amendment No. 77, 77, 77, l
7
o *
[
LIST OF FIGURES f Title Pirure No.
Core Limits Four Loop Operation 2.1 1 ;
i Maximum Permissible T,ta for First RCP Start 3.1.A 1 ;
(OPS Operable, llottest SG Temp. > T,,ta)
Maximum Permissible RCS Pressure for RCP Start (s) with 3.1.A 2 OPS Inoperable (SG Temp. > T,.ta for additional pump .;
starts, SG Temp < Teoid for all. pump starts) >
RCS Pressure Limits for Low Temperature Operation 3.1.A 3 ;
Maximum Pressure Limits for OPS Inoperable and 3.1.A.4 .
First RCP Start (SG Temp. >'Teoid) I Pressurizer Level for OPS Inoperable and 3.1.A 5 !
Ono (1) Charging Pump Energized ,
Pressurizer Level for OPS Inoperable and 3.1.A 6 {
One (1) Safety Injection Pump and/or Three (3) !
Charging Pumps Energized ,
Reactor Coolant System lleatup' Limitations 3.1 1 p Reactor Coolant System Cooldown Limitations 3.1 2 Pricary coolant Specific Activity Limit vs. Percent 3.1 3 l of Rated Thermal Power l Gross Electrical Output .1" 110 Backpressure 3.4 1 Gross Electrical Output 1.5" IIG Backpressure. 3.4 2 i Limiting Fuel Burnup vs. Initial Enrichment 3.8 1 Minimum Burnup for Storage of Fuel in Max Density 3~.8 2 Spent Fuel Pit Racks Maximum Density Spent Fuel Pit (SFP) Racks, 3.8 3 Regions and Indexing ,
Required Shutdown Margin 3.10 1-Steam Generator. Primary Side Ultrasonic Test Sectors 4.2 1 .l l Pressure / Temperature Limitations for flydrostatic Leak' 4.3 1' Test 1
Amendment No. J#, 77,'77, 99,
i 1.16 REPORTABLE EVENT .
A REPORTABLE EVENT shall be any of those-conditions specified in Section 50.73 to'10 CFR 50.
1.17 CORE OPERATING LIMITS REPORT The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific !
, document that provides . core operating limits for the current l operating reload cycle. These cycle-specific core operating limits j shall be determined for each reload cycle in accordance with Specification 6.9.1. 6. Plant operation within these operating' i
! limits is addressed in individual specifications.
l .
l l
l 1-6 i Amendment No. 57,
e In meeting this design basis, uncertainties in plant operating parameters, !
nuclear and thermal parameters, and fuel fabrication parameters are ,
considered statistically such that there is at least a 954 probability with .
95% confidence level that the minimum DNDR for the limiting rod is greater than or equal to the applicable DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR limit, establishes a 1 design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. In addition, margin is maintained by performing DNB design evaluations t9 a higher DNBR valuo, called the Safety Limit DNBR.
The curves of Figure 2.1 1 show the loci of points of thermal power, Reactor Coolant System pressure and vessel inlet temperature for which the calculated DNBR is no less than the Safety Limit DNBR value or the average
- enthalpy at the vessel exit is less than the enthalpy of saturated liquid.
The calculation of these limits includes:
RTP N >
- 2. an equivalent steam generator tube plugging level of up to 30% in any steam generator provided the equivalent average plugging level in all i steam generators is less than or equal to 244, (4
- 3. a reactor coolaet mjstem total flow rate of greater than or equal to 332,240 gpm an n ured at the plant,
- 4. a reference cosine with a peak of 1.55 for axial power shape.
Figure 2.1 1 includes an allowance for an increase in the enthalpy rise hot channel factor at reduced power based on the expression:
N RTP Fa s Fa (1 + PFa (1 P))
Where P is the fraction of Rated Thermal Power,
, RTP N Fin is the Fin limit at Rated Thermal Power specified in the COLR, and PFa is the Power Factor Multiptior specified in the COIR, When flow or Fin is measured, no additional allowances are necessary prior to comparison with the limits presented. A 2.6% measurement uncertainty. on .
Flow and a 44 measurement uncertainty of Fin have already been included in the above limits.
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the riaximum allowable control rod insertion limit (Figure 3.10 4) assuming the axial power imbalance is within the limits of the f(AI) function of the Overtemperature AT trip.
When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trips will reduce the setpoints to provide protection consistent with core safety limits.
2.1-2 Amendment No. 99, #7, 97, FF, ,
i
o .
4 t
References
- l. FSAR Section 3.2.2 t
r P
i l
l l
l l
2.1 3 Amendment No. 99, FS, fy, pp,
3.10 CONTROL ROD AND P0gR DISTRIBUTION LIMITS Applienbility:
Applies to the liinits on core fission power distribution and to liinits on control rod operations.
Obiectives: ,
To ensure:
- 1. Core subcriticality after reactor trip.
- 2. Acceptab1', core power distribution during power operation in order to ,
maintair. fuel integrity in normal operation and transients ossociated with raults of moderate frequency, supplemented by automatic t rotection and by administrative procedures, and to maintain the design basis initial conditions for limiting faults, and
- 3. Liinit potential reactivity insertions caused by hypothetical control rod ejection.
Soccifica tions :
3.10.1 Shutdown Reactiviry The shutdown margin shall be at least as great as shown in Figure 3.10 1, 3.10.2 Power Distribution Limits 3.10.2.1 At all times, except during low power physics tests, the hot channel factors defined in the basis must ince t the following limits:
Fo (Z) s (FoRTP/p) x K(Z) for P > 0.5 l Fo (Z) s (FnRTP/0.5) x K(Z) for P s 0.5 l
N Fn A s FAn"TP (14 PFan (l*P))
Where P is the fraction of full power at which the core is operating, K(Z) is the fraction specified in the COLR, Z is the core height location of Fn, FnRTP is the Fg limit at Rated Thermal Power (RTP) specified in the COLR, FanRTP is the Fan limit at Rated Thermal Power specified in the COLR, and PFan is the Power Factor Multiplier specified in the COLR.
3.10 1 Amendment No. 77, #F, #F, 77, 77, FF,
3.10.2.2 Following initial core loading, subsequent reloading and I at regular effective full power monthly intervals ;
thereafter, power distribution maps, using the movable detector system, shall be made to confirm that the hot channel factor limits of this specification are satisfied. !
For the purpose of this comparison, 3.10.2.2.1 The measurement of total peakin8 factoroF *", shall be I increased by three percent to account for manufacturing tolerances and furcher increased by five percent to account ,
for measurement error. l 3.10.2.2.2 When F"a is measured, no additional allowances are necessary prior to comparison with the limits of section 3.10.2. An error allowance of 44 has been included in the limits of section 3.10.2. If either measured hot channel factor exceeds its limit specified under Item 3.10.2.1, ,
the reactor power and high neutron flux trip setpoint shall be reduced so as not to exceed a fraction of rated power '
equal to the ratio of the To or F"a limit to measured value, whichever is less. If subsequent incore mapping cannot, within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, demonstrate that the hot channel factors are met, the reactor shall-be brought to a hot shutdown condition with return to power authorized only for the purpose of physics testing.
3.10.2.3 The reference equilibrium indicated axial flux difference for each excore channel as a function of power level (called the target flux difference) shall be measurt.d at least once per equivalent full power quarter. The target flux differences must be updated each effective full power l month by linear interpolation using the most recent
- measured value and a value of 0 percent at the end of tne' cycle life.
3.10.2.4 Except during physics tests, during excore calibration procedures and except as modified by Items 3.10.2.5 through 3.10.2.7 below, the indicated axial flux difference of all but one operable excore channel shall be maintained within the band specified in the COLR about the target flux .
difference.
3.10 2 .
Amendment No. U , F#,
l 3.10.2.5 At a power level greater than 90% of rated power, 3.10.2.5.1 If the indicated axial flux difference of more than one ,
operabic excore channel deviates from its target band, either such deviation shall be immediately eliminated or the reactor power shall be reduced to a level no greater than 90 percent of rated power.
3.10.2.6 At a power level no greater than 90 percent of rated power, 3.10.2.6.1 The indicated axial flux difference (AFD) may deviate from its, target band specified in the com for a maximum of one .
. hour (cumulative) in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provided the flux difference does net exceed an envelope bounded by that specified in the COLR at 904 power and increasin6 by the value specified in the COM for each 2 percent of rated power below 9 fit power. A two hour deviation is permissible ,
during tests performed as part of the augmented startup !
program, m 3.10.2.6.2 If Item 3.10.2.6.1 is violated by more than one operable excore channel, then the reactor power shall be reduced to no greater than 50% power and the high neutron flux setpoint reduced to no greater than 55 percent of rated values.
3.10.2.6.3 A power increase to a level greater than 90 percent of rated power is contingent upon the indicated axial flux difference of all but one operable excore channel being within their target band.
3.10.2.7 At a power level no greater than 50 percent of rated power, 3.10.2.7.1 The indicated axial flux difference may deviate from its l target band.
3.10.2.7.2 A power increas.e to a level greater than 50 percent of teted power is contingent upon the indicated axial flux dif t rance of all but one operabic excore channel not being outside their target bands for more than two hours-(cumulative) out of the preceding 24. hour period. One half the time the indicated axial flux difference is out of its target band up to 50% of rated power is to be counted as contributing to the one hour cumulative (two hour cumulative during augmented startup tests) m maximum the flux difference may deviate from its target band of a power '
level s 90% of rated power.
3.10 3 >
l Amendment No.
o 3.10.2.8 Alarms are provided to indica *,e non conformance with the flux difference requirements of 3.10.2.5.1 and the flux difference time requirements of 3.10.2.6.1. If the alarms i are temporarily out of service, conformance with the applicable limit shall be demonstrated by logging the flux !
difference at hourly intervals 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 thereaf ter. ,
3.10.2.9 If the core is operating above 75% power with one excore f nuclear channel out of service, then core quadrant power i balance shall be determined once a day using movable incore detectors (at least two thimbles per quadrant).
t 3.10.3 Ouadrant Power Tilt Limits 3.10.3.1 When ever the indicated quadrant power tilt ratio exceeds 1.02, except for physics tests, within two hours the tilt condition shall be eliminated or the following actions shall be taken:
a) Restrict core power level and reset the power range high flux setpoint three percent of rated value for every percent of indicated power tilt ratio exceeding 1.0, ,
and '
b) If the tilt condition is not eliminated af ter 24 ,
hours, the power range nuclear instrumentation setpoint shall be reset to 55% of allowed power.
i Subsequent reactor operation is permitted up to 50%
l Inr the purpose of measurement, testing and cortwetive action, t 3.10.3.2 Except for physics tests, if the indicated quadrant power tilt ratio exceeds 1.09 and there is simultaneous indication of a misaligned control rod, restrict core power level 3% of rated value for every percent of indicated power tilt ratio exceeding 1.0 and realign the rod within two hours. If the rod-is not realigned within two hours or if there is no simultaneous indication of a misaligned rod, the reactor shall- be brought to the hot shutdown l
condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the reactor is shut down, subsequent testing up to 50% of rated power shall be permittod to determine the cause of the tilt. ,
3.10 4 Amendment No. J/t,
3.10.3.3 The rod position indicators shall be monitored and logged
- once each shift to verify rod _ position within each bank assignment.
3.10.3.4 The tilt deviation alarm shall be set to annunciate whenever the excore tilt ratio exceeds 1.02. If one or both of the quadrant power tilt monitors is inoperable, individus1 upper and lower excore detector calibrated outputs shall be logged once per shift and af ter a load change greater than 10 percent of rated power.
3.10.4 Rod Insertion Limits ;
3.10.4.1 The shutdown rods shall be fully withdrawn as specified in the Cola when the reactor is critical or approaching criticality (i.e. , the reactor is no longer suberitical by an amount equal to or greater than the shutdown margin in Figure 3.10 1).
3.10.4.2 When the reactor is critical, the control banks shall be limited in physical insertion to the insertion limits i specified in the C01R. l 3.10.4.3 control bank insertion shall be further restricted if:
- a) The measured control rod worth of all rods, less the worth of the most reactive rod (worst case stuck rod), is less than the reactivity required to provide the design value of available shutdown, b) A rod is inoperable (Specification 3.10.7).
3.10.4.4 Control rod insertion limii.s do not apply during physics tests or during perbdic exercise of itedividual rods.
I lloweve r , the shutdc.m margin indicated in Figure 3.10 1 I
must be maintainer' 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 control l rod inserted.
3.10 5 Amendment No. Jf, FF,
e
- 3.10.5 Rod Misaliennent Limitations i 3.10.5.1 If a control rod is misaligned from its bank demand position by more than 12 steps (indicated position), then ;
realign the rod or determine the core peaking factors within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and apply Specification 3.10.2. '
3.10.5.2 If the requirements of Specification 3.10.3 are determined not to apply and the core peaking factors have not been determinen within two hours and the rod remains misaligned,
.the high reactor flux setpoint shall be reduced to 85% of its rated value. j 3.10.5.3 If the misaligned control rod is not reali6 ned within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> the rod shall be declared inoperable.
3.10.6 Inocerable Rod Position Indicator Channels ,
3.10.6.1 If a rod position indicator channel is out of service then: .
- a. For operation between 50 percent and 100 percent of rating, the position of the control rod shall be checked indirectly by core instrumentation (excore l
detectors and/or movable incore detectors) every shift, or subsequent to rod motion exceeding 24 steps, whichever occurs first,
- b. During operation below 50 percent of rating, no special monitoring is required.
3.10.6.2 Not more than one rod position indicator channel per group j nor two rod position indicator channels per bank shall be permitted to be inoperable at any time.
3.10.6.3 If a control rod having a rod position indicator channel out of service, is found to be misaligned from 3.10.6.la above, then Specification 3.10.5 will be applied. )
3.10 6 Amendment No. 77, l
3.10.7 Jnonerable Rod Limitations
- 3.10.7.1 An inoperable rod is a rod which does not trip or which is ;
declared inoperable under Specification 3.10.5 or fatis to meet the requirements of 3.10.8. ,
3.10.7.2 Not more than one inoperable control rod shall be allowed any time the reactor _is critical except during physics tests requiring intentional rod misalignment. Otherwise, !
the plant shall be brought to the hut shutdown condition.
3.10.7.3 If any rod has been declared inoperable, then the potential ejected rod worth, associated transient power distribution ;
peaking factors and the accident listed in Table 3.10 1 shall be analyzed within 5 days, or the reactor brought to
- the hot shutdown condition using normal operating procedures. 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 plant power level shall be reduced to :
an analytically determined part power level which is consistent with the safety analysis. [
3.10.8 Rod Dron Time At operating temperature and full flow, the drop time to each control rod shall be no greater than 2.4 seconds from loss of stationary gripper coil voltage to dashpot entry.
3.'07 Amendment No. 7#, Jf, FJ, 1
?
9 3.10.9 Rod Position Monitor If the rod position deviation monitor is inoperable, individual rod positions shall be logged once per shift and after a load change greater than 10 percent of rated power.
3.10.10 Reactivity Balance i
The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1% Ak/k at least once per 31 Effective Fuel Power Days (EFPD).
This comparison shall, at least consider reactor coolant system boron concentration, control rod position, reactor coolant system average temperature, fuel burnup based on gross thermal energy generation, xenon concentration, and samarium concentration. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core condition prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.
3.10.11- Notification Any event requiring plant shutdown on trip setpoint reduction because of Specification 3.10 shall be reported to the Nuclear Regulatory Commission within 30 days.
Basis Design criteria have been chosen for normal operations, operational transients and those events analyzed in FSAR Section 14.1 which are consistent with the fuel integrity analysis. These relate to fission gas release, pollet temperature and cladding mechanical properties. Also, the minimum DNBR in the core must not be less than the applicable safety limit DNBR in normal operation or in short term transients.
In addition to the above conditions, the peak linear power density must not exceed the limiting Kw/f t values which result from the large break loss of coolant L
3.10 8 Amendment No. 7J, JJ, JJ, t
.. - - - ~ - . . _ , -- ..
g I
i accident analysis based on the ECCS acceptance criteria limit of 2200*F.
This is required to meet the initial conditions assumed for loss of coolant accident analyses. To aid in specifying the limits on power distribution,
~
the following hot chatinel factors are defined.
F0 (2)
- Hef aht ' Decendent Heat Flux Hot channel Factor, is defined as the
. f maximum local heat flux on the surface of a fuel rod at core elevation Z l divided by the average fuel rod heat flux, allowing for manufacturing l tolerances on. fuel pellets and rods.- :
1 I
e i
r i
-i
)
)
i 1
I i t
l
- l. i i -
l j-p t
l.
_3 .' 10 8 a _ 1 Amendment No. 7Ji, i
's
-b
t f
Fnt Fngineering flent Flux flot channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances. The engineering factor -
allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.
Fa" Nucicar Enthalov Rise llot 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.
It should be noted that Fa" is based on an integral and is used as such in the DNB calculations. Local heat fluxos are obtained by using hot channel '
and adj acent channel explicit power shapes which take into account variations in horizontal (x y) power shapes throughout the core. Thus the horizontal power shape at the point of maximum heat flux is not necessarily directly related to Fa". i An upper bound envelope of FnRTP specified in the C01R times the normalized peaking factor axial dependence of K(2) specified in the COLR has been determined consistent with Appendix K criteria and is satisfied for OFA transition mixed cores m by all operating maneuvers consistent with the technical specifications on power distribution control as given in Section 3.10. The results of the loss of coolant accident analysis based on this upper bound normalized envelope of Figure 3.10 2 demonstrates a peak clad temperature not greater m than 2049"F, which is below peak clad temperature limit of 2200'F.
When an Fn measurement is taken, both experimental error and manufacturing tolerance must be allowed for. Five percent is the appropriate allowance for a full core map taken with the movabic incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.
In the specified limit of Fa" there is an 8 percent allowance for uncertainties which means that normal operation of the core is expected to result in Fa" s FaRTP/1.04, where Fa RTP is the Fa" limit at Rated Thermal Power specified in the C01R. The logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power shape r 3.10 9 Amendment No. 15, 99, 99, 97, 7), Ff.
l l
(e.g. rod misalignment) affect Fan", in most cases without necessarily affecting F ,n (b) the operator has a direct influence on Fo through movement of rods. and can limit it to the desired value, he has no direct control over Pui" and'(c) an error in the predictions for radial _ power shape, which may be detected during startup physics tests can be compensated for in Fo by tighter axial control, but compensation for is less readily availabic. When a measurement of Fai" is taken, no additional allowances are necessary prior to comparison with the limit of section 3.10.2. A measurement uncertainty of 4% has been allowed for in determination of the design DNBR value.
Measurements of the hot channel factors are required as part of startup physics tests, at least each effective full power month of operation, and whcnever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors. The incore map '
taken following initial loading provides confirmation of the basic nuclear design basis including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would, otherwise, :
affect these bases, i For normal operation, it is not necessary to measure these quantities.
Instead it has been determined that, providwd certain conditions are observed, the hot channel factor limits will be met; these conditions are as follows:
- 1. Control rods in a single bank move together with no individual rod insertion differing by more than 15 inches from the bank demand ,
position. An indicated misalignment limit of 12 steps precludes a rod '
misalignment no greater than 15 inches with consideration of maximum instrumentation error.
l
- 2. Control Rod banks are sequenced with overlapping banks as described in
' Technical Specification 3.10.4.
- 3. The control rod bank insertion limits are not violated.
1 -
l e l
l l
l L
l I
3.10 10 Amendment No. 77, F#,
- 4. Axial Power Distribution Control Procedures, which are given in terms of flux difference control and control bank insertion limits are observed. Flux difference refers to the difference in signals between ,
the top and bottom halves of two section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.
The permitted relaxation in F n A allows radial power shape changes with rod 8
insertion to t.he insertion limits. It has been determined that provided the '
above conditions 1 through 4 are observed, these hot channel factors limits '
are met. In Specification 3.10.2, Fo is arbitrarily limited for P s 0.5 (except for low power physics tests).
The procedures-for axial power distribution control referred to above are designed to minimize the effects of xenon redistribution on the axial powt;r distribution during load follow maneuvers. Basically, control of flux difference is required to limit the difference between the current value of ,
Flux Difference (AI) and a reference value which corresponds to the full power equilibrium value of Axial Offset (Axial Offset ' - A1/ fractional power). The referenced value of flux difference varies with power level and burnup but expressed as axial offset it varies only with burnup.
The technical specification
- on power distribution control assure that Fn upper bound envelope of t'nRTP times K(Z) (specified in the COLR) is not exceeded and xenon distributions are not developed which at a later time, would cause greater local power peaking even though the flux difference is i then within the limits specified by the procedure.
The target (or reference) value of flux difference is determined as follows, At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with the control rod bank more than 190 steps withdrawn (i.e. normal full power operating position appropriate for j the time in life, usually withdrawn farther as burnup ,
l 3.10-11 Amendment No. 77, 99, 97, 97, /J, 77,
proceeds). Thie 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 indicated equilibrium value 1 was noted, no allowances for excore detector error are necessary and the AFD deviation specified in the COLR is permitted from the indicated reference value. During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring the :
target flux difference every month. For this reason, the specification ,
provides two methods for updating the target flux difference. >
Strict control of the flux difference (and rod position) is not as necessary 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 i control at part power. Strict control of the flux difference is not possible during certain physics tests or during required, periodic, excore calibrations which require larger flux differences .than permitted. >
Therefore, the specifications on power distribution control ar not applied during physics tests or excore calibrations; this is acceptabic due to the low probability of a significant accident occurring during these operations.
In some instances of rapid plant power reduction, automatic rod motion will cause the flux difference to deviate from the target band when the reduced :
power 1cvel is reached. This does not necessarily affect the xenon distribution sufficiently to change the envelope of peaking factors which t p can be reached on a subsequent return to full power within the target band.
I Ilowever, to simplify the specification, a limitation of one hour in any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is placed 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 ,
flux difference in the AFD range specified in the COLA. l l
l l
I 3.10-12 Amendment No.
If, for any reason, flux dif ference is not controlled within the AFD limit specified in the C01.R for as long a period as one hour, then xenon distributions may be significantly changed and operation at 50 percent is required to protect against potentially more severe consequences of somo
- accidents.
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 ta position the control rods to produce the required indicated flux difference.
For FSAR Section 14.1 events, the core is protected from overpower and a minimum DNBR of _the applicable safety limit DNBR by an automatic protection -
system. Compliance with operating procedures is assumed as a precondition
- for FSAR Section 14.1 events. However, operator error and equipment malfunctions are separately assumed to lead to the cause of the transients considered.
- Quadrant power tilt limits are based on the following considerations.
Frequent power tilts are not anticipated during normal operation, as this ,
phenomenon is caused by some asymmetric perturbation, e.g., rod misalignment, or inlet temperature mismatch. A dropped or misaligned rod :
will easily be detected by the Rod position Indication System or core instrumentation per Specification 3.10.6, and core limits are protected per Specification 3.10.5. A quadrant tilt by some other means would not appear instantaneously, but would build up over several hours and the quadrant tilt limits arn met to protect against this situation.- They also serve as a backup protection against' the dropped or misaligned rod. Opecstional experience shows that normal power tilts are less than 1.01. Thus, sufficient time is available to recognir.e the presence of a tilt and correct the cause before a severe tilt could build up. During startup and power escalation, howeve , a large tilt could be initiated. Therefore, the Technical Specification has been written so as to prevent escalation above 50 percent power if a large ti1~ is present. The numerical limits are set i to be commensurate with design and safety limits for DNB protection and linear heat generation rate as l
l
\'
l 3.10 13
! Amendment No. J#, #f, ##,
l t
w
.. t .
1 described below.
The radial power distribution within the core must satisfy the design values '
assumed for calculation of power capability. Radial power distributions are measured as part of the startup ph'ysics testing and are periodically ,
measured at a monthly or-greater frequency. These measurements are taken
, _to assure that the radial power distribution with any quarter core radial power asymmetry conditions are consistent with the assumptions used in power capability analyses. It is not intended that reactor operation would continue with- a power tilt condition. which exceeds the radial power asymmetry considered in the power capability. analysis.
l The quadrant tilt power deviation alarm is used to indicate a sudden or ,
l unexpected change from the radial power distribution mentioned above. The two percent tile alarm setpoint represents a minimum practical value ce:nsistent *d t h instrumentation errors and operating procedures. This ,
asymmetry 1me) is sufficient to detect significant misalignment of control rods. Misalignment of cons d ads is considered . to be the - most likely cause of radial power asynusetty. The requirement for verifying rod position ,
once each shift is imposed & freclude rod misalignment which would cause a tilt condition less than the 24 alarm level.
2-
%e two hour time intervr] .in this specification- is considered ample to identify a dropped or misaligned rod and complete realignment procedures to 1 eliminate the tilt. In the event that the tilt condition cannot be
- eliminated within the two hour time allowance, additional time would be
- needed to investigate the cause of the tile condition. The measurements would include a full core physics map _ utilizing the moveable detector
>- system. For a tilt condition s 1.09, an additional 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> time interval is authorized to accomplish these measurements. However, to assure that the l
peak core power is maintained below limiting values, . a reduction of reactor '
power of three percent for each one percent of indicated. tilt is required.
Physics measurements have indicated that the core radial power peaking would not exceed a two to one relationship with the indicated tilt from the excore nuclear detector system for the worst rod misalignment.
l- In the event a tilt condition of s 1.09 cannot be eliminated af ter 24. hours, the reactor power level will be: reduced to the range required for low power physics testing. To avcid reset of a large number of protection setpoints,_
the power range nuclear instrumentation would be reset to cause an automatic
- reactor trip at $5% of allowed. power.
i' l
3.10 14 Amendment No.
3
. . . , . . . . . - . , . - , . . - - ~ - _ . - .. . . ,
t
- i
. i A reactor trip at this power has. been selected to prevent, with margin, !
exceeding core safety limits even with a nine percent tilt condition.
If tilt ratio greater than 1.09 occurs which is not due to a misaligned rod, the reactor shall be brought to a hot shutdown condition for investigation.
However, if the tilt condition can be identified as due to rod misali 6nment, operation can continue at a reduced power (3% for each one percent the tilt.
ratio' exceeds 1.0) for two hours to correct the rod misalignment.
Trip shutdown reactivity is provided consistent with plant safety analysis assumptions. One percent shu';down is adequate except for steam break analysis, which requires more shutdown if the boron concentration is low. ,
Figure 3.10 1 is drawn accordingly.
Rod insertion limits are used to assure adequate' trip reactivity, to assure meeting power distribution liinits , and to limit the consequence of a hypothetical rod ejection accident. The available control rod reactivity, or excess beyond needs, decreases with decreasing boron- concentration because the negative reactivity required to reduce the . core power level f rom full power to zero is largest when the boron concentration is low.
The intent of the test to measure control rod worth and shutdown' margin ,
(Specification 3.10.4) is to measure the worth of all rods less'the worth of the worst case for an assumed stu:k rod, that is, the most reactive rod.
The measurement would be anticipated a= part of the initial startup program and infrequency over the life of the plant, to be associated primarily with determinations of special interest such as end of life cooldown, nr startup of fuel cycles which deviate from normal equilibrium conditions in terms of '
fuel loading patterns and anticipated fcontrol . bank worth. These measurements will augment the normal fuel cycle design calculations and place the knowledge of shutdown capability on a firm experimental as well ,
as analytical basis.
l The rod position indicator channel is sufficiently accurate to detect a rod i
17 inches away from its demand position. An indicated misalignment less than 12 steps does not exceed the power peaking factor limits. If the rod position indicator channel is not operable, the operator will be fully aware i
c f the inoperability of the . channel, and special surveillance of core power l tilt indications, using established procedures and relying on excore nuclear detectors, and/or moveable incore. detectors, will be used to verify power d.!stribution symmetry. These indirect measurements do not have the same resolution if the bank is near either end of the core,:because a 12 step misalignment would have no effect on power distribution. .Therefore, it is necessary to apply the indirect checks following significant rod motion.
3.10-15 Amendment No. J#,
. l F
One. inoperable control rod is acceptable ~ provided that the power distribution limits are met, trip shutdown capability is available, and L provided the potential hypothetical ejection of the inoperable rod is not worse than the cases analyzed in the safety analysis report.. .The rod ejection. accident for an isolated fully inserted rod will be worse if the residence time of the rod is long enough to_cause significant non. uniform i fuel depletion. The 5 day period is short compared with- the time interval required to achieve a significant, non uniform fuel depletion..
The required drop time to dashpot entry is consistent with safety analysis.
REFERENCE
- 1. WCAP 8576, " Augmented Startup and Cycle 1 Physics Program:, August 1975 '
- 2. FSAR Appendix 14C
- 3. IAtter.from J.P. Bayne to S.A. Varga dated April _ 23, 1985, entitled
" Proposed Technical Specifications Regarding the Cycle 4/5 Refueling".
L i
l 4
3.10 16 Amendment No. 79, #)',
4
t TABLE 3.10 1 ACCIDENT ANALYSES REOUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL LENCTH ROD Rod Cluster Control Assembly Insertion Characteristics
- Rod Cluster Control Assembly Misalignment loss of Reactor Coolant From Small Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power Major Reactor Coolant System Pipe Rupeures (Loss of Coolant Accident)
Major. Secondary System Pipe Rupture j Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection) i i
i I
l l
l l
l l
l' 3.10-17 l
Amendment No. 77, Jft, ~l l
i
- n. - -
- i. .
a ANNUAL REPORTS 6.9.1.5 A report of cpecific activity analysis results in which the primary coolant exceeded the limits of Specification 3.1.D. The following information shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> -l prior.to the first sample in which the limit was exceeded; (2) Results of '
the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis-while activity was reduced to less than limit.
Each. result should include date and time of sampling and the radioiodine <
concentrations; (3) Clean up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which.the limit was exceeded; (4) Data providing the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady state level; and (5) The time duration when the specific activity of.the primary coolant exceeded the.radioiodine limit. l 6.9.1.6 CORE OPERATING LIMITS REPORT j 6.9.1.6 a Core operating-limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload -
cycle or any remaining part of a reload cycle for the following:
- 1. Axial Flux Difference limits for Specification 3.10.2. .
- 2. lleat Flux flot Channel . . Factor and K(Z) for Specification 3.10.2.
- 3. Nuclear Enthalpy Rise'llot Channel Factor and Power '
! Factor Multiplier for Specification 3.10.2.
l
!- 4. Shutdown ' Bank Insertion Limit for Specification l 3.10.4.
i l S. Control Bank Insertion Limits for Specification 3.10.4. ,
6.9.1.6.b The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:
- 1. WCAP-9272-P A, "WESTINCil0USE RELOAD SAFETY EVALUATION MET 110D0iDGY," July 1985 (H Proprietary) . .
(Methodology- for -Specification 3.10 4 -
Shutdown Bank Insertion Limit, Control. Bank -
Insertion Limits and 3.10.2 - Nuclear Enthalpy Rise llot Channel Factor.
6 15 Amendment No. JJ, ##, ff, JP, JJ, pp, 77,
, r . '1
'=
.)
l 2a. WCAP 8385, " POWER DISTRIBUTION CONTROL AND IDAD l FOLLOWING PROCEDURES -- TOPICAL REPORT", September
J 1974 (H Proprietary).
(Methodology for Specification 3.10.2 .- Axial l Flux Difference (Constant Axial Offset Control).)
2b. T. M. Anderson to K, Knell (Chief of Core Performance
, Branch, NRC) January 31, 1980 --
Attachment:
l Operation and Safety Analysis Aspects of an Improved l Load Follow Package.
(Methodology for Specification 3.10.2 Axial- ,
! Flux Difference (Constant Axial Offset i Control).)
, 2c. NUREG 0800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981. Branch Technical Position : CPB 4.3-1,-
Westinghouse Constant Axial Offset Control- (CAOC), r Rev. 2, July 1981. _ _
(Methodology for Specification 3.10.2 - Axial Flux Difference -(Constant Axial Offset. ,
control).)
3a. WCAP-9220 P-A, Rev.1, " WESTINGHOUSE ECCS EVALUATION j MODEL 1981 VERSION", February 1982 (H Proprietary) .
(Methodology for Specification 3.10.2 - Heat Flux Hot Channel Factor.)
3b. WCAP 9561-P A ADD. 3, Rev. 1 "BART A-1: A= COMPUTER CODE FOR THE BEST ESTIMATE ANALYSIS OF REFiDOD TRANSIENTS - SPECIAL REPORT: THIMBLE MODELING H ECCS EVALUATION MODEL," July 1986, (H Proprietary).
(Methodology for ' Specification 3.10.2 - Heat Flux Hot Channel Factor.).
3c. WCAP 10266 P-A Rev. 2, "THE 1981 VERSION OF '
WESTINGHOUSE EVALUATION.MODEL USING BASK CODE", March .
1987, (H Proprietary).
(Methodology for Specification 3.10.2 - Heat Flux Hot Channel Factor).
- 6.9.1.6.c The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal mechanical limits, ,
core thermal hydraulic limits, ECCS limits, nuclear limits such as- shutdown margin, and transient and accident analysis limits) of the safety limits are met.
6.9.1.6.d The CORE OPERATINC LIMITS REPORT, including any mid cycle revision 2 or supplements thereto,_shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to- the Regional Administrator and Resident-Inspector.
+
Amendment No. J/, ##, JJ,'79, #7, F#, 77, 6-16 d
i ,;o .
SPECIAL REPORTS j 6.9.2 Special reports shall be submitted to the Regional Administrator-Region 1 within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification;
- a. Sealed source leakage on excess of limits (Specification '
3.9) ~
- b. Inoperable Seismic Monitoring Instrumentation (Specification 4.10)
- c. ' Seismic event analysis (Specification 4.10)
- d. Inoperable plant vent sampling, main steam line radiation l monitoring or effluent monitoring capability (Table 3.5. 4, items 5, 6 and 7)
! e. The complete results of the steam generator tube inservice inspection (Specification 4.9 C)
^
- f. Inoperable fire protection and- detection- equipment i
(Specification 3.14) l
- g. Release of radioactive effluents in . excess of limits (Appendix B Specifications 2.3, 2.4, 2.5, 2.6) l h. Inoperable containment high-range radiation monitors (Table
! 3.5-5, Item-24) i
- l. 1, Radioactive - environmental sampling results in excess of-reporting' levels (Appendix.B Specification 2.7, 2.8, 2.9)
- j. Operation of Overpressure' Protection System (Specification 3.1'.A.8.c.)
6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
i a. Records and' logs of facility operation covering time interval at each power level,
- b. Records and logs: of principal maintenance ' activities, inspection, repair and replacements of' principal items of
-j equipment related to nuclear safety,
- c. All REPORTABLE EVENTS submitted to the Commission.
6-17 Amendment No. 77, ##, J7, JP, #7, pp, JJ,
. no ..- '
. ?
- d. Records of surveillance . activities, inspections and calibrations required by these. Technical-Specifications.
- e. Records of changes made to Operating Procedures,
- f. Records of radioactive shipments, r l
- g. Records of scaled source and fission detector leak' tests and results,
- h. Records of annual physical inventory of all source material of record.
3
- 1. Records-of' reactor tests and experiments. .
6.10.2 The following records shall be retained for the duration of tho' .;
Facility Operating License:
- a. Records of any drawing changes reflecting facility design .
modifications made to systems and equipment described in !
the Final Safety Analysis Report. l b.
. Records of 'new- and irradiated fuel' inventory, ffuel transfers and assembly burnup histories, i
- c. Records of facility radiation. and contamination surveys. '
- d. Records of radiation exposure for-a111 individuals entering.
radiation control areas,
- e. Records of gaseous and liquid radioactive material released-to the environs.
- f. Records of transient or operational cycles' for those facility , components designed for a limited number of transient cycles. +
r
- g. Records of training and qualifications for current members ;
of the plant staff.
- h. Records of in-service inspections performed pursuant to these Technical Specifications,
- i. Records of Quality. Assurance activities required bj. the QA manual.
J. Records of reviews performed for changes made to. procedures or equipment or reviews of tests and experiments. pursuant ,
to 10 CFR 50.59.
6 18 Amendment No. 77, 79, 97, #7, 77, I
e
is .
I t -
- 1. -Records for Environmental, Qualification which are covered under the provisions ' of paragraph 6.13.
- m. Records of secondary water sampling and. water quality.
- n. Records- of analyses required- by the radiological environmental monitoring- program- that would permit r evaluation of the accuracy of the analysis at a later date,
, This should include procedures effective at specified times ,
and records showing that these procedures were followed,
- o. Records of service lives of all safety related hydraulic ~
snubbers including the date at which the service life commences and associated installation and maintenance records.
6.11 RACIATION AND RESPIRATORY PROTECTION P M 6.11.1 Procedures for personnel radiation protection shall'be prepared consistent with the requirements of 10 CFR Part 20' and shall be approved maintained and adhered to for all operations involving personnel radiation exposure as to maintain exposures as,far below the limits.specified in 10 ,
CFR Part 20 as reasonable achievable'. Pursuant to 10 CFR 20.103 allowance <
shall be made for the use of respiratory protective equipment in conjunction with activities _ authorized by the operating license for this plant in determining whether individuals- in restricted areas are exposed to con ~ rations in excess of the limits specified in Appendix B, Table I,.
Col-.. of 10 CFR 20.
t l=
1 l
4 e
i.
6-19 Amendment No. 77, 77, 77, 79, 77, 77, I
w . -
. A.. l:
~
6.12 HIGH RADIATION AREA _
6.12.1' In lieu of the " control device" or." alarm signal" required by paragraph 20.'203- (c) (2) of 10 CFR 20, each high radiation area in which the intensity of rraiation? is 1000 = arem/hr or less and 100 mrem /hr or greater shall be tarricaded and conspicuously posted as a hi h Eradiation area and entrar;e thereto shall be controlled by requiring issuance of a-Radiation Wo k Permit *.- Any individual or group of individuals permitted to enter such areas shall be provided or accompanied by one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.
b.- -A radiation monitoring device which continuously integrates the radiation dose rate in .the area and alarms when a preset 4 integrated dose is received. Entry into such areas with 'this -
monitoring device may be made af ter the dose rate level in the area has been established- and -personnel have been made knowledgeable of them.
- c. An individual qualified in radiation protection procedure who is equipped with' a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the. area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit. i a6.12.2 The requirements of 6.12.1, above,. sha11' also apply to each high radiation area in which the intensity of radiation is -greater than 1000 mrem /hr, In addition, locked doors shall ~be provided. to- prevent unauthorized entry into such areas and the keys shell be maintained:under the administrative control of the Shif t Supervisor ca auty and/or the plant Radiological and Environmental Superintendent or_his designee.
- Health Physics Personnel shall be exempt from the RWP issuance requirements for entries intoi high radiation areas during the performances of their assigned radiation protection duties,- provided they comply with approved radiation protection procedures for- entry into high radiation areas.
6-20 Amendment No. JJ, 77, 79, FF,
6.13 ENVIRONMENTAL OUALIFICATION 6.13.1 By no later than June 30,~1982 all safety related electrical equipment in the facility shall be qualified- in accordance with the provisions of: Division of Operating Reactors " Guidelines for Evaluating.
Environmental Qualification'of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG 0588 " Interim Staff Position ' on-Environmental Qualification of Safety-Related Electrical equipment,"
December 1979. . Copes of f these documents are attached to Order . for -i Modification of License No. DPR-64 dated October 24, 19PC 6.13.2' By no later than December l i 1980, complete and auditible records must be.available and maintained at a central' location which describe the
- environmental qualification method used for all safety related electrical i equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG 0588 Thereafter, such records should be updated and maintained current . as equipment is replaced, further tested, or otherwise further qualified, t
s E
i 4 '
6-21 ,
o Amendment No. JJ, Order dated October 24, 1980, JJ,
]
I j' 'a
< l
'1 I
F l
l l.
l.
I
. N ATTACHMENT ll TO IPN 90-034 SAFETY EVALUATION FOR TECHNICAL SPECIFICATION CHANGES REGARDING REMOVAL OF CYCLE SPECIFIC PARAMETER LIMITS, IN ACCORDANCE WITH GENERIC LETTER 88-16 3
l.
l l
)
l
(
i NEW YORK POWER AUTHORITY INDIAN POINT 3 NUCLEAR POWER PLANT DOCKET NO. 50-286' DPR44' i
+
- w. -we . - +-,s, m--m< w
. .. : o Attachm2nt ll IPN 90-034 Page 1 of 3 Section 1 Description of Changes The proposed changes remove the cycle specific parameter limits from Technical Specifications and reference a Core Operating Umits Report containing those limits. These
- changes are made in accordance with the guidance contained in NRC Generic Letter 88-16,
" Removal of Cycle Specific Parameter Umits From Technical Specifications." The proposed changes are presented in Attachment 1.
Section 11- Evaluation of Changes Several Technical Specifications _(TS) address limits -associated with reactor physics parameters that generally change with each reload core, requiring changes to the TS to update these = lirnits each fuel cycle. Since these limits are = developed using' NRC approved .
methodologies, procesing the changes is an unnecessary burden on the NRC and the Authority.
NRC Generic Letter 8816 provides an altamative consisting of three separate actions to modify the TS:
- 1. The addition r.f the definitic. of a named formal report that includes the values of cycle. . b specific parmeter %ts- that have been established using an NRC . approved methodol~;j .io consistent with all applicable limits of the safety analysis. At Indian ;
Point 3 tbs report shall be known as the " Core Operating Umits Report," and its definitior,is contained in proposed TS 1.17.
- 2. T!.e addition of an administrative reporting requirement to submit the Core Operating Umits Report (COLR) on cycle-specific parameter limits to the NRC for information. The -
addition of this requirement is contained in proposed TS 6.9.1.6.
- 3. The modification of Individual TS (and assoc'ated bases) to note that cycle specific parameters shall be maintained within the lim;ts provided in the COLR. The following TS parameters have been Identified as cycic e specific limits that can be relocated to the COLR:
- a. Heat Flux Hot Channe' rector and K(Z).
1
- b. Nuclear Enthalpy Rise Hot Channel Factor and Power Factor Multiplier,
- c. Axial Flux Difference Umits.
- d. Shutdown Bank Insertion Umits.
- e. Control Bank Insertion Umits.
The proposed altomative is responsive to industry and NRC efforts to improve TS, reduce the administrative burden on the NRC and the New York Power Authority, and permit future reloads to be accomplished without license amendments.
The Authority will be implementing these Generic Letter changes in support of Cycle 0 operation. The Cycle 8 CULR is being prepared and will be provided to the NRC upon issuance.
p ' , . . . ,
' Att chment 11 l i
IPN 90-034 '
! Page 2 of 3 Section lli No Significant Hazards Evaluation i
Consistent with the requirements of 10 CFR 50.92, the enclosed application is judged to ,
involve no significant hazards based on the following information:
(1) Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
PWee: ,
The proposed amendment is administrative in nature, merely relocating cycle-specific parameter limits from the Technical Specifications to the Core Ooerating Umits Report.
NRC approved methodologies will continue to be used as the basis for establishing these limits. The Core Operating Umits Report will be submitted to the NRC for their use in i
trending the values of cycle specific limits. The proposed changes are in accordance ;
l with the guidance provided by NRC Generic Letter 8816 and do not involve a significant . -
Increase in the probability or consequences of an accident previously evaluated. .
(2) Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response
No safety related equipment, function, or plant operation will be altered as a result of the proposed changes. The changes are administrative in nature and do not create any new accident mode. The level of document control and quality assurance applied to the preparation and use of the Core Operating Umits Report will be equivalent to that applied to the Technical Specifications.
(3) Does the proposed amendment involve a significant reduction in a margin of safety?
Response
The proposed changes are administrative in nature and do not impact the operation of the plant in a manner that will reduce the margin of safety. The proposed amendment still requires operation within the limits determined using NRC approved methods, and that appropriate remedial actions be taken if the limits are violated.
Section IV-Impact of Change These changes will not adversely irnpact the following:
ALARA Program Security and Fire Protection Programs l Emergency Plan FSAR or SER Conclusions Overall Plant Operations and the Environment
, r - - - - - - - . . - - - . . . . . . , , , - , . ,. - .- =- --, -r s +r-.,. ,,v-..%, . re.e ~
g c. s.
^
Attachment ll IPN 90-034 j Page 3 of 3 Section V - Conclusions The incorporation of these changes: a) will not increase the probability nor the consequences of an accident or malfunction of equipment important to safety.as previously evaluated in the Sa'ety Analysis Report; b) will not increase the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; c) will not reduce the .' .
margin of safety as defined in the bases for any Technical Specification; d) does not constitute an unreviewed safety question; and e) Involves no significant hazards considerations as defined in 10 CFR 50.92.
Section VI- References i
a) IP 3 FSAR !
b) IP-3 SER 4 1
j I
L i
4l
(
l!
l:
p.
a l
l .
I
'I i
i
. , . - .