ML20211H081
| ML20211H081 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/13/1987 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML19292G819 | List: |
| References | |
| TAC-64779, TAC-64780, NUDOCS 8702260060 | |
| Download: ML20211H081 (16) | |
Text
LIST OF FIGURES Figure Title 1.1-1 Core Thermal Power Safety Limit Versus Core Flow Rate 2.1 -1 Reactor Vessel Water Levels 4.1-1 Graphical Aid for the Selection of an Adequate Interval Between Tests 4.2-1 System Unavailability 3.4-1 Sodium Pentaborate Solution Volume Versus Concentration Requirements 3.4-2 Sodium Pentaborate Solution Tercperature Versus Concentration Requirements 3.6-1 Pressure versus Minimum Temperature for Pressure Tests, Such as Required by ASME Section XI 3.6-2 Pressure versus Minimum Temperature for Non-nuclear Heatup/Cooldown and Low Power Physics Test 3.6-3 Pressure versus Minimum Temperature for Core Critical Operation other than Low Power Physics Test (Includes 40*F Margin Required by 10CFR50 Appendix G) 3.6-4 Deleted 3.6-5 Thermal Power Limitations During Operation with Less Than Two Reactor Coolant System Recirculation Loops in Operation 3.11-1 (Sheet 1) Limiting Value for APLHGR (Fuel Type IC Types 1, 2, and 3) 3.11-1 (Sheet 2) Limiting Value for APLHGR (Fuel Types 80250, 8DRB265H, P80RB265H, and BP80RB265H) 3.11-1 (Sheet 3) Limiting Value for APLHGR (Fuel Types P8DRB284H, BP80RB284H, and 80R183) I 3.11-1 (Sheet 4) Limiting Value for APLHGR (Fuel Types 8DR233, P80RB284LA, and BP80RB284LA) 3.11-1 (Sheet 5) Limiting Value for APLHGR (Fuel Types P80RB283 and BP80RB283) 3.11-1 (Sheet 6) Limiting talus for APLHGR (Fuel Types BP80RB299 and Hatch 1 1987 LTAs)
- 3.11-2 Limiting Value for LHGR (Fuel Type 7 x 7) j 3.11-3 MCPRp (Flow Dependent Adjustment Factors for MCPRs) 3.11-4 MCPR Limit for Ali 8 x 8 Fuel Types for Rated Power and Rated Flow 8702260060 870213 PDR ADOCK 05000321 P PDr; x Proposed TS/0105q/365
C. Core Alteration - Core alteration shall be the addition, removal, relocation, or movement of fuel, sources, incore instruments, or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of core alterations shall not preclude completion of the movement of a component to a safe conservative position.
D. Desian Power - Design power refers to the power level at which the reactor is producing 105 percent of reactor vessel rated steam flow.
Design power does not necessarily correspond to 105 percent of rated .
reactor power. The stated design power in megawatts thermal (MWt) is the result of a heat balance for a particular plant design. For Hatch Nuclear Plant Unit 1 the design power is approximately 2537 MWt. l E. Engineered Safety Features - Engineered safety features are those features provided for mitigating the consequences of postulated accidents, including for example containment, emergency core cooling, and standby gas treatment system.
F. Hot Shutdown Condition - Hot shutdown condition means reactor operation
, with the Mode Switch in the SHUT 00WN position, coolant temperature greater than 212*F, and no core alterations are permitted.
I G. Hot Standby Condition - Hot standby condition means reactor operation with the Mode Switch in the START & HOT STANDBY position, coolant ,
temperature greater than 212*F, reactor pressure less than 1045 psig, critical.
H. Immediate - Immediate means that the required action shall be initiated as soon as practicable, considering the safe operation of the Unit and the importance of the required action.
I. Instrument Calibration - An instrument calibration means the adjustment of an instrument output signal so that it corresponds, within acceptable range and accuracy, to a known value(s) of the parameter which the instrument monitors.
J. Instrument Channel - An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.
HATCH - UNIT 1 1.0-2 Proposed TS/0099q/364
Y
_,_ BASESiFOR LIN11ING SAFETY SYSTEM SETTINGS
]'2.1 , FUEL CLAD 11NG_JNTEGRITY 1.s lhe abnormal operational transients applicable to operation of the HNP-1 Unit M have been analyzedqthroughout the s spectrum of planned operating conditions.
1 The analyses were based upon plantjperation in accordance with the operating l map given in Figure 3-1 of Ref. 8. . In addition, 2436 MWt is the licensed
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T maximum power lev (I of HNP-1, and this represents the maximum steady-state power which shall not knowingly be exceeded.
7 s Conservatism is incorporated in the transient analyses in estimating the con-trolling factors, such as void reactivity coefficient, control rod scram worth, cram delayt time, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model. This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic performance. Results obtained from a General Electric boiling water reactor have been compared with predictions made by the model. The comparisons and results are summarized in Reference 1.
The absolute value of the void reactivity coefficient used in the analysis is conservatively; estimated to be about 25% greater than the nominal maximum value expected to occur'buring the core lifetime. The scram worth used has been derated> to be equivalent to approximately 80% 6f :the total ! scram worth of the 4 control rods. The scram delay time and rate of r6d insertion allowed by lthe i , analyses are conservatively set equal to the longest delay and slowest inser-
-tion rate acceptable by Technical Specifications. Active coolant flow is equal a
4088%oftotalcoreflow. lhe effect of scram worth, scraidelay time and rod
> n insertion rate, all conservatively applied, are of greatest rignificance in the fearly' portion of the negative reattivity insertion. The rapid insertion of megative reactivity is assured b9. the time requirements for 5% and 25% inser-ftion. By the time the rods'a 0 6Kinserted, approximately four dollars of
' negative reactivity.have baen; inserted (see F hure 7-1, NE00-21124-7) which strongly turns the transient ' and accomplishes the desired ef fect. The times for 50% and 905 insertion are give'n to assure proper completion of the 6xpected performance in the earlier portion of the transient, and to establish the x ' ultimate fully shutdown steady-state condition.
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4 .
l ',1. For dnalyses of the thermal consequences of ,the transients, a MCPR equal to or greater thu the actual op9 rating limit MCPR is ' conservatively assumed to exist
(((1priorto/ipitiationofthetransients.
g es I
~~j~ Steady-state operation without forced recirculation will not be permitted, except during qtartup testing. The analysis to support operation at various l .
I a % %
i ,
.' .\
l 1 (
HATCH - UNIT 1 1.1-10 Proposed TS/0104q/364
._ BASES FOR LIMITING SAFELY SYSTEM SETlINGS 2.1 FUEL CLADDING INTEGRITY (Continued) power and flow relationships has considered operation with either one or two recirculation pumps.
In summary:
- 1. The licensed maximum power level is 2436 MWt.
l ii. Analyses of transients employ adequately conservative values of the controlling reactor parameters. l iii. The analytical procedures now used result in a more logical l answer than the alternative method of assuming a higher start-ing power in conjunction with the expected values for the parameters .
- 3. Trio Settings The bases for individual trip settings are discussed in the following para-paphs.
- 1. Neutron Flux Trio Settings
- a. IRM Flux Scram Trio Setting The IRM system consists of 8 chambers, 4 in each of the reactor protec-tion system logic channels. The IRM is a 5-decade instrument which cov-ers the range of power level between that covered by the SRM and the APRM. The 5 decades are covered by the IRM by means of a range switch and the 5 decades are broken down into 10 ranges, each being one-half of a decade in size. The IRM scram trip setting of 120 divisions is active in each range of the IRM. For example, if the instrument were on range 1, the scram setting would be a 120 divisions for that range; likewise, if the instrument were on range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate the increase in power level, the scram trip setting is also ranged up.
The most significant sources of reactivity change during the power in-crease are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the phys-ical limitation of withdrawing control rods, that heat flux is in equi-librium with the neutron flux and an IRM scram would result in a reac-tor shutdown well before any Safety Limit is exceeded.
In order to ensure that the IRM provided adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM system is not yet on scale. This condition exists at quarter rod density. Quarter rod den-sity is illustrated in Figure 7.5-8 of the FSAR. Additional conserva-HATCH - UNIl 1 1.1-11 Proposed TS/0104q/364
r BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.2 REACTOR COOLANT SYSTEM INTEGRITY A: Nuclear System Pressure
- 1. When Irradiated Fuel is in the Reactor The 11 relief / safety valves are sized and set point pressures are estab-lished in accordance with the following requirements of Section III of the ASME Code:
- a. The lowest relief / safety valve must be set to open at or below vessel design pressure and the highest relief / safety valve must be set to open at or below 105% of design pressure.
- b. The valves must limit the reactor pressure to no more than 110% of design pressure.
The primary system relief / safety valves are sized to limit the primary system pressure, including transients, to the limits expressed in the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.
No credit is taken from a scram initiated directly from the isolation event, or for power operated relief / safety valves, sprays, or other power operated pressure relieving devices. Thus, the probability of failure of the turbine-generator trip SCRAM or main steam isolation valve closure SCRAM is conservatively assumed to be unity. Credit is taken for subsequent indirect protection system action such as neutron flux SCRAM and reactor high pressure SCRAM, as allowed by the ASME Code. Credit is also taken for the dual relief / safety valves in their ASME Code qualified mode of safety operation. Sizing on this basis was l applied to the most severe pressurization transient, which is the main steam isolation valves closure, starting from operation at 105 percent of the reactor warranted steamflow condif n. The adequacy of this I relief / safety valve sizing is verified is:n cycle by comparing the results of the analysis of the MSIV closure event starting from 102%
of rated thermal power with the ASME limits described above.
Reference 2, Figure 4 shows peak, vessel bottom pressures attained when the main steam isolation valve closure transients are terminated by various modes of reactor scram, other than that which would be initiated directly from the isolation event (trip scram). Relief /
safety valve capacities for this analysis are 84.0 percent, represen-tative of the 11 relief / safety valves.
The relief / safety valve settings satisfy the Code requirements for relief / safety valves that the lowest valve set point be at or below the vessel design pressure of 1250 psig. These settings are also sufficiently above the normal operating pressure range to prevent unnecessary cycling caused by minor transients, lhe results of postu-lated transients where inherent relief / safety valve actuation is required are given in Section 14.3 of the FSAR.
- 2. When Operating the RHR System in the Shutdown Cooling Mode An interlock exists in the logic for the RHR shutdown cooling valves, which are normally closed during power operation, to prevent opening of the valves above a preset pressure setpoint of 145 psig. This setpoint is selected to assure that pressure integrity of the RHR system is main-tained. Administrative operating procedures require the operator to HATCH - UNIT 1 1.2-6 Proposed TS/0103q/364
0 U@H.llG _C ON_0 l0T !_0NS F_0R__0_PJ R A T 10N SURVEILLANCE REQUIRENENTS 3.3.8.2. Excessive Scram Time 4.3.8. Operable Control Rod Exercise Requirements (Cont'd)
Control rods with a scram insertion time to reach notch position 6 which l When it is initially determined exceeds 7.00 seconds shall be con- that a control rod is incapable sidered inoperable, but if they can of normal insertion, an attempt be moved with control rod drive to fully insert the control rod pressure, they need not be fully shall be made. If the control inserted or disarmed electrically. rod cannot be fully inserted the reactor shall be brought to 3.3.B.3. Inoperable Accumulators the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and a shutdown margin test made to demonstrate Control rods with inoperable under this condition that the accumulators or those whose core can be made subcritical position cannot be positively for any reactivity condition determined shall be considered during the remainder of the inoperable. operating cycle with the analytically determined,
- 4. Limitina Number of Inoperable Control highest worth control rod Rods capable of withdrawal, fully withdrawn, and all other control During reactor power operation, no rods capable of insertion fully l more than one control rod in any inserted.
l 5 x 5 array may be inoperable (at least 4 operable control rods must Once per week, check the status separate any 2 inoperable ones). If of the pressure and level alarm this Specification cannot be met the for each accumulator.
reactor shall not be started, or if at power, the reactor shall be 4.3.C. Control Rod Drive System brought to a shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 1. Control Rod Drive Couplina In-tearity C. Control Rod Drive System lhe coupling integrity shall be
- 1. Control Rod Drive Couplina verified for each withdrawn con-Integrity trol rod as follows:
Each control rod shall be cou- a. When the rod is withdrawn the pled to its drive or completely first time after each refuel-inserted and its directional ing outage or af ter mainte-control valves disarmed electri- nance, observe discernible re-cally except during control rod sponse of the nuclear instru-drive maintenance as stated in mentation and rad position in-Specification 3.10.E. dication including where ap-plicable the " full-in" and
" full-out" position. However, for initial rods when respor.se is not discernible, subsequent exercising of these rods after the reactor is above 30% power shall be performed to verify instrumentation response.
HATCH - UNIT 1 3.3-2 Proposed TS/01004/364
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UNIllii[C0NDIT10N'STF6R]PERATION
, SURVEILLANCE REQUIREMENIS 4.3.C.l.b. When the rod is fully withdrawn the first time after each refueling outage or af ter main-tenance, observe that the drive does not go to the overtravel position.
3.3.C.2. Scram Insertion Times 4.3.C.2. Scram Insertion Times
- a. All ODerable Control Rods a. After each refueling outage all control rods capable of The average scram insertion time normal insertion shall be of all operable control rods at scram time tested from the a reactor dome pressure 1950 psig fully withdrawn position based on the de-energization of af ter a reactor dome pressure the scram pilot valve solenoids of 950 psig has been attained.
as time zero, shall be no greater This testing must be complete than: before 40% rated thermal Notch Position Average Scram l From Fully Insertion b. Routine Time Tests Withdrawn Time (Sec)
At 16-week intervals, 10% of 46 0.358 the control rods capable of 36 1.096 movement with control rod 26 1.860 drive pressure shall be scram 6 3.419 timed above 950 psig. When-ever such scram time measure-
- b. Three Out of Four Rods in a ments are made, an evaluation Two-by-Two Array shall be made to provide reasonable assurance that The average of the scram inser- proper control rod drive tion times for the three fast- performance is being maintained.
est control rods of all groups of four control rods in a two-by-two array at a reactor dome pressure 1 950 psig shall be no greater than:
Notch Position Average Scram l From Fully Insertion Withdrawn Time (Sec) 46 0.379 36 1.162 26 1.972 6 3.624 HATCH - UNIl 1 3.3-3 Proposed TS/0100g/364
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BASE _5~FIfR LlMITING COND! LIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.3.C. Control Rod Drive System
- 1. Control Rod Drive Couplina Integrity l
l Limiting Conditions for Operation: l Operability of the control rod drive system requires that the drive be coupled to the control rod. In the analysis of control rod drop accidents it has been assumed that one control rod drive coupling has lost its integrity. To assure that not more than one coupling could be in this con-dition, it is required that either a drive is coupled to the control rod or the drive is fully inserted and disarmed electrically. This requirement serves to maintain operation within the envelope of conditions by the plant safety analyses.
Surveillance Requirements:
Observation of a response from the nuclear instrumentation during an attempt to withdraw a control rod provides an indication that the rod is following the drive. The overtravel position feature provides a positive check on the coupling integrity, for only an uncoupled drive can reach the overtravel position.
l 2. Scram Insertion Times Limiting Conditions for Operation:
The control rod drive system is designed to bring the reactor sub-critical at a rate fast enough to prevent excessive fuel damage. Analysis of the limiting transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specification provide the required protection and MCPR remains greater than 1.07. lhe limit on the number and pattern of rods permitted to have long scram times is specified to assure that the effect of rods of long
, scram times are minimized in regard to reactivity insertion rate. Crouping of long scram time rods is prevented by not permitting more than one slow rod in any four rod array. The minimum amount of reactivity to be inserted during a scram is controlled by permitting no operable control rod to nave a scram insertion time to notch position 06 greater than 7 secondt. l HATCH - UNIl 1 3.3-10 Proposed TS/0100g/364
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.8. Linear Heat Generation Rate (LHGR)
(Continued) operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the 1
LHGR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power within the next four (4) hours. If the limiting condition for operation is restored prior to expiration of the specified time int:rval, then further progression to less than 25%
of rated thermal power is not required.
C. Minimum Critical Power Ratio (MCPR) 4. ll .C.1. Minimum Critical Power Ratio (MCPR)
The minimum critical power ratio (MCPR) MCPR shall be determined to be shall be equal to or greater than the equal to or greater than the operating limit MCPR (OLMCPR), which applicable limit, daily during is a function of scram time, core reactor power operation at 125%
power, and core flow. For 25% $ rated thermal power and following power < 30%, the OLMCPR is given in any change in power level or dis-Figure 3.11.6. For power 130%, tribution that would cause opera-the OLMCPR is the greater of either: tion with a limiting control rod pattern as described in the bases
- 1. The applicable limit determined for Specification 3.3.F.
from Figure 3.11.3, or 4.11.C.2. Minimum Critical Power Ratio Limit
- 2. The applicable limit from either Figures 3.11.4 or 3.11.5 The MCPR limit at rated flow and multiplied by the Kp factor rated power shall be determined for determined from Figure 3.11.6, each fuel type, as appropriate, where t is the relative f rom figure 3.11.4 or 3.11.5 measured scram speed with respect using:
to Option A and Option 8 scram speeds. If x is determined to a. t = 1.0 prior to initial scram be less than zero, then the time measurements for the OLMCPR is evaluated at t = 0. cycle, performed in accordance with specifications 4.3.C.2.a.
or
- b. t is determined from scram time measurements performed in accordance with specifica-tion 4.3.C.2.
The determination of the limit must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance test required by specification 4.3.C.2.
HATCH - UNIT 1 3.11-2 Proposed TS/0101g/364
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.11.C. Minimum Critical Power Ratio (MCPR) l l
i l
If at any time during operation it is cetermined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the steady ,
state MCPR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power within the next four(4) hours. If the Limiting Condition for Operation is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated thermal power is not required.
D. Reporting Reauirements If any of the limiting values iden-tified in Specifications 3.11.A.,
B., or C. are exceeded, a Reportable Occurrence report shall be submitted.
If the corrective action is taken, as described, a thirty-day written report will meet the requirements of this specification.
HATCH - UNIT 1 3.ll-2a Proposed TS/0101g/364
~. -_ - _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ - _ _ _ _
10 I I l PUEL TYPE: PSORS233 100 MIL CHANNELS
,3 I I I
, UNACCEPTABLE OPER ATION
'T I
{&>w 12 11.3 12.1 12.1 g 9,g
- 11.3 103
=
w ACCEPTABLE OPERATION x
I.
X Z I .
0 5 10 15 20 3 30 5 40 45 50 AVERAGE PLANAR EXPOSURE (GWdh) i I I i l l PUEL TYPES: PSORS283 AND SPSORS283 80 MIL CHANNELS g 14
{ UNACCEPTABLE OPERATION a a eg 13 il5 3
1 li u.
12.0 12.1 N1.s u 56
<a 11 11.2 3 ,
l5> 10 ACCEPTASLE OPER ATION 102 h
. Y.8 si s
O s 30 is 20 m 30 3s 40 4s s0 AVERAGE PLA fAAR EXPOSURE (Gwdh)
HATCH - UNIT 1 FIGURE 3.11-1 (SHEET 5) Proposed TS/XXXq/364
l 18 PUELTYPE: SPOORS 29W So AND too MllCHANNELS 14 UNACCEPTABLE OPERATION
.a 12.1 12.1 12.o 12 k 11.o E to
$ ACCEPTASLE 9.o
$g$ OPERATION
<, e
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4 2
o 9 5 to 15 2o 25 3o 35 de 45 So
< AVERAGE PLANAR EXPOSURE (Gwdh) 12.9 13.o '
12.5 FUEL TYPE: 19e1 H TCH 1 LTA s 12.o ,
11.s 11.s E E 11.o y) UNACCEPTABLE OPERATION lE d I lo.o --
j lf sI ACCEPTABL OPE"4ATION (g.4 5 s.o
\
lg-g l s.o y
a 7.7 7.o -
S s.3 s.o o s lo is ao a ao a ao es so AVERAGE PLANAR EXPOSURE (OWdh)
HATCH - UNIT l. FIGURE 3.11 1 (SHEET 6) Proposed TS/XXXq/364 l
. - _ _ . - . __ _ - __,- - _._ -_ ~ . - __ - ._, -_ - _ -.._ _ - - ._._ ._-_-__ - . _ --_-,
1.35 1.34 1.33 1.32 *
/
i.31 /
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5 3 ' 30 s #
e I u. '
1.28 1.27 l 1.29 l
1.25 0.0 1,0 FIGURE 3.11-4 MCPR LIMIT FOR ALL 8X8 FUEL TYPES FOR RATED POWER AND RATED FLOW HATCH - UNIT I Proposed TS/XXXa/364
16 5
1 a
x 1 e E
U m
14 i E g_ UNACCEPTABLE OPERATION 5E "3
m-
< us 2 >-
8 K 12.0 _
12.1
'! 12 ',
$ gP j j,9 11.8 i e4 11. 11.4 w y 11.2 yy \10.8 I
h hk ACCEPTABLE OPERATION 10.3 5E \
9.6 i
i 8.9 8
!7
-O
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O 5 10 15 20 25 30 35 40 45 50
! & AVERAGE PLANAR EXPOSURE (Gwd/t) lk
{M i
FUEL TYPES P8DRB283 AND BP80RB283 80 MIL CHANNELS MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION
! .s RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE
)Qa FIGURE 3.2.1 10
POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 ALL MINIMUM CRITICAL POWER RATIOS (MCPRs), shall be equal to or greater than the MCPR operating limit (OLMCPR), which is a function of average scram time, core flow, and core power. For 25% s Power < 30%, the OLMCPR is given in Figure 3.2.3-4. For Power 2 30%, the OLMCPR is the greater of either:
- a. The applicable limit determined from Figure 3.2.3-3, or
- b. The appropriate Kp given by Figure 3.2.3-4, multiplied by the appropriate limit from Figure 3.2.3-1 or 3.2.3-2 where t is the relative measured scram speed with respect to Option A and Option B scram speeds. If t is determined to be less than zero, then the OLMCPR is evaluated at t = 0.
APPLICABILITY: CONDITION 1, when THERMAL POWER 2 25% RATED THERMAL POWER AC TION:
With MCPR less than the applicable limit determined from Specification 3.2.3.a. or 3.2.3.b, initiate corrective action within 15 minutes and continue corrective action so that MCPR is equal to or greater than the applicable limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than or equal to 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
HATCH - UNIT 2 3/4 2-6 Proposed TS/0102q/364
3/4.2.3 MINIMUM CRITICAL POWER RATIO (CONTINUED)
SURVEILLANCE REQUIREMENTS 4.2.3 The MCPR limit at rated flow and rated power shall be determined for each type of fuel (8X8R, P8X8R, BP8X8R, and 7X7) from Figures 3.2.3-1 and 3.2.3-2 using
- a. t = 1.0 prior to the initial scram time measurements for the cycle performed in accordance with Specification 4.1.3.2.a, or
- b. t is determined from scram time measurements performed in accordance with Specification 4.1.3; the determination of the limit must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance test required by Specification 4.1.3.2.
MCPR shall be determined to be equal to or greater than the applicable limit:
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and
- c. Initially and at least once pe- 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTR01. R00 PATTERN for MCPR.
HATCH - UNIT 2 3/4 2-7 Proposed TS/0102q/364