ML20214T604
| ML20214T604 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/01/1987 |
| From: | Youngblood B Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214T605 | List: |
| References | |
| TAC-64779, TAC-64780, NUDOCS 8706100342 | |
| Download: ML20214T604 (21) | |
Text
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g3 Ef 3 UNITED STATES E
g NUCLEAR REGULATORY COMMISSION 5
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WASHINGTON, D. C. 20555
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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA 4
CITY OF DALTON, GEORGIA l
DOCKET NO. 50-321 i
EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 i
AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 139 License No. DPR-57 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 1 (the facility) Facility Operating License No. DPR-57 filed by Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensee) dated February 13, 1987, and supple-
'l mented May 18, 1987, complies with the standards and requirements i
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;-
i B.
The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; i
C.
There is reasonable assurance (i) that the activities authorized by
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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 Comission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and j
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have i
been satisfied.
I m
8706100342 870601 ADOCK0500g1 PDR e
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.' 2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.139. are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
\\Y B. J. Youngblood, Director Project Directorate II-3 Division of Reactor Projects-I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: June 1, 1987 t
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ATTACHMENT TO LICENSE AMENDMENT NO.139 i
L FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the followir.g pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert Page Page X
X 1.0-2 1.0-2 1.1-10 1.1-10 1.1-11 1.1-11 1.2-6 1.2-6 3.3-2 3.3-2 3.3-3 3.3-3 3.3-10 3.3-10 3.11-2 3.11-2 3.11-2a, 3.11-2a 3.11-4a 3.11-4a Figure 3.11-1 (Sheet 5) Figure 3.11-1(Sheet 5)
Figure 3.11-2 (Sheet 6) Figure 3.11-2 (Sheet 6)
Figure 3.11-4 Figure 3.11-4
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LIST OF FIGURES Finore Title 1.1-1 Core Thermal Power Safety Limit Versus Core Flow Rate 2.1 -i Reactor Vessel Water Levels 4.1-1 Graphical Aid for the Selection of an Adequate Interval ietween Tests a.2-1 System Unavailability 3.4-1 Sodium Pentaborate Solution Volume Versus Concentration Requirements 3.4-2 Sodium Pentaborate Solution Temperature Versus Concentration Requirements 3.6-1 Pressure versus Minimum Temperature for Pressure Tests, Such as Required by ASHE 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 otr.er than Low Power Physics Test (Includes 40*F Margin Required by 10CFR50 Appendix G) 3.b-4
. Deleted 3.6-5 Thermal Power Limitat. ions During Operation with Less Than Two Reactor Cw iant System Recirculation Loops in Operation 3.11-1 (Sheet 1) Limiting Value for APLHGR (Fuel Type IC lypes 1, 2, and 3) 3.11-1 (Sheet 2) Limiting Value for APLHGR (Fuel Types 80250, 80R8265H, P80RB265H, and 8P80RB265H) 3.11-1 (Sheet 3) Limiting Value for APLHGR (Fuel Types P8DRB284H, 8P80RB284H, and 80R183' l
3.11-1 (Sheet 4) Limiting Value for APLHGR (Fuel Types 80R233, P80RB284LA, and BP8DRB284LA) 3.11-1 (Sheet 5) Limiting Value for APLHGR (Fuel Types P80RB283 and BP80RB283) 3.11-1 (Sheet 6) Limiting Value for APLHGR (Fuel Types BP80RB299 and Hatch 1 1987 LTAs) 3.11-1 (Sheet 7) MAPFACp (Power Dependent Adjustment Factors to MAPLHGRs) -
3.11-1 (Sheet 8) MAPFACp (Flow Dependent Adjustment Factors to MAPLHGRs) 3.11-2 Limiting Value for LHGR (Fuel Type 7 x 7) 3.11-3 MCPRg (Flow Dependent Adjustment Factors for MCPRs) 3.11-4 MCPR Limit for All 8 x 8 Fuel Types for Rated Power and Rated Flow MATCH - UNIT 1 Amendment No. 139 x
.e C.
Core Alteration - Core alteration shall be the addition, removal, r.elocation, or movement of fuel, sources, incore instruments, or reactivity controls within the reactor pressure vessel with the vessetl 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.
Desien Power - Design power refers to the power level at which the D.
reactor is producing 105 percent of reactor vessel rated steam flow.
Design power does not necessarily correspond to 105 percent of rated The stated design power in megawatts thermal (MWt) is reactor power.
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 Engineered Safety Features - Engineered safety features are those E.
features provided 'or mitigating the consequences of postulated accidents, includi.g for example containment, emergency core cooling, and standby gas treatment system.
I Hot Shutdown Condition - Hot shutdown condition means reactor operation lh with the Mode Switch in the SHUT 00WN position, coolant temperature F.
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j greater than 212*F, and no core alterations are permitted.
Hot 'tandby Condition - Hot standby condition means reactor operation
'f G.
with the Mode Switch in the START & HOT STANDBY position, coolant qi temperature greater than 212*F, rea: tor pressure less than 1045 psig, critical.
Immediate - Immediate means that the required action shall be initiated
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as soon as practicable, considering the safe' operation,of the Unit and y
- the importance of the required action.
7 l
Instrument Calibration - An instrument calibration means the adjustment
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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.
Instrument Channel - An instrument channel means an arrangement of a 1'
J.
sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter I
monitored by that instrument channel.
HATCH - UNIT 1 1.0-2 Amendment No. 139
I BASES FOR LIMITING SAFETY SYSTEM SElTINGS
' 2.1 FUEL CLAD 0!Nj_ INTEGRITY The abnernal operational transients applicable to operation of the HNP-1 Unit have been analyzed throughout the spectrum of planned operating conditions.
The analyses were based upon plant operation in accordance with the operating l
cap given in Figure 3-1 of Ref. 8.
In addition, 2436 MWt is the licensed maximum power level of HNP-1, and this represents the maximum steady-state power which shall not knowingly be exceeded.
Conservatism is incorporated in the transient analyses in estimating the con-trolling factors, such as void reactivity coefficient, control rod scram worth, scram delay 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, cvolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic perforrance. Results obtained from a 62neral Electric boiling water reactor have been compared with predictions made by the model. The comparisons and results are sunnarized in Reference 1.
The absolute value of the void reactivity coefficient used in the analysis is ccnservatively estimated to be about 25% greater than the nominal maximum value expected to occur during the core lifetime. The scram worth used has been dorated to be equivalent to approximately 80% of the total scram worth of the control rods. The scram delay time and rate of rod insertion allowed by the cnalyses are conservatively set equal to the longest delay and slowest inser-
'L;wc.
.:.t. acceptable by Technical Specifications. Active coolant flow is equal Lo 88% vi total core flow. The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of negative reactivity is' assured by the time requirements for 5% and 25% inser-tion. By the time the rods are 60% inserted, approximately four dollars of negative reactivity have been inserted (see Figure 7-1, NE00-21124-7) which strongly turns the transient, and accomplishes the desired ef fect. The times for 50% and 90% insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the sitimate fully shutdown steady-state condition.
For analyses of the thermal consequences of the transients, a MCPR equal to or greater than the actual operating limit MCPR is conservatively assumed to exist prior to initiation of the transients.
Steady-state operation without forced recirculation will not be permitted, except during startup testing. The analysis to support operation at various l
1 NATCH --UNIT 1 1.1-10 Amendment No. 139 1
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a BASES FOR LIMITING SAFE 1Y SYSTEM SET 1INGS I
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2.1 FUEL CLADDING INTEGRTT.y,(Continued) power and flow relationships has considered operation with either one or two recirculation pumps.
In sumary:
1.
The licensed maximum power level is 2436 MWt.
I l
11.
Analyses of transients employ adequately conservative values i
of the controlling reactor parameters.
l iii.
The analytical procedures now used result in a more logical
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answer than the alternative method of assuming a higher start-l ing power in conjunction with the expected values for the i
parameters.
l A.
Trio Settinas i
The bases for individual trip settings are discussed in the following para-graphs.
I' j
1.
Neutron Flux Trio Settinas a.
IRM Flux Scram Trio Settino 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
]5 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.
active in each range of the IRM.The IRM scram trip setting of 120 divisions is 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 re tor shutdown well before any Safety Limit is exceeded.
2 In order to ensure that the IRM the single rod withdrawal error, pros ided adequate protection against 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 suberitical and the IRM system is not yet on scale.
This condition exists at quarter rod density. Quarter rod den-l sity is illustrated in Figure 7.5-8 of the FSAR.
P Additional conserva-HATCH - UNIl 1 1.1-11 A"
8ASES 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 lished in accordance with the following requirements of Section III the ASME Code:
vessel design pressure and the highest relief /safetT a.
be set to open at or below 105% of design pressure.y valve must b.
The valves must design pressure. limit the reactor pressure to no more than 110% of system pressure, including transients, to the limit ASME Boiler and Pressure Vessel Code Section III, Nuclear Vessels No credit is taken from a scram initiated directly from th power operated pressure relieving devices. event, or for powe i
or other valve closure SCP.AM is conservatively assumed taken for subsequent indirect protection system action such as neutro Credit is flux SCRAM and reactor high pressure SCRAM, as allowed by the A Code.
Credit is also taken for the dual relief / safety valves in their
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ASME Code qualified mode of safety operation.
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applied to the most severe pressuritation transient, which is the main l
steam isolation valves closure, starting from operation at 105 percent of the reactor warranted steamflow condition.
i relief / safety valve sizing is verified each cycle by comparing theThe
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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, vessef 6atter pressures attained when the main steam isolation valve clobre transients a by various edes of reactor scram, other than tilat which would be initiated directly from the isolation event safety valve capacities for this analysis are(trip scram). Relief /
tative of the 11 relief / safety valves. _
(14.0 percent, represen-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 settir suf ficiently above the normal operating pressure range,gs are also unnecessary cycling caused by minor transients.
to prevent required are given in Section 14.3 of the FSAR. lated tran 2.
When Coeratina the RHR System in the Shutdown Coolina Mode An interlock exists in the logic for the RHR shutdown cooling valves which are nornelly closed during power operation, to prevent openin the valves above a preset pressure setpoint of 145 psig.
l is selected to assure that pressure integrity of the RHR system is main This setpoint tained.
Administrative operating procedures require the operator to HATCH - UNIT 1
- 1. 2-6 Amendment No 139 r
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.idMITlNG CQNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.B.2.
Excessive Scram Time 4.3.8.
Doerable control Rod Exercise -
Reauirements (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 red pressure, they need not be fully shall-be made. If the control inserted or disamed electrically, rod cannot be fully inserted the reactor shall be brought to i
3.3.8.3.
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 i
margin test made to demonstrate i
Control rods with inoperable under this condition that the I
accumulators or those whose core can be made subcritical position cannot be positively for any reactivity condition detemined shall be considered during the remainder of the inoperable.
operating cycle with the analytically detemined, 4.
Limitina Number of InoDerable Control highest worth control rod
!t.od..s capable of withdrawal, fully withdrawn, and all other control During reactor power operation, no rods capable of insertion fully more than one control rod in any inserted.
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 alam 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 />.
'l.
Control Rod Drive Couplina in-teority C.
Control Rod Drive System The coupling integrity shall be 1.
Control Rod Drive Couplina verified for each withdrawn con-Intearity trol rod as follows:
Ea:h 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 after mainte-control valves disarmed electri-nance, observe discernible re-i cally except during control rod sponse of the nuclear instru-drive maintenance as stated in mentation and rod position in-Specification 3.10.E.
dication including where ap-plicable the ' full-in' and
" full-out' position. However, for initial rods when response is not discernible, subsequent exercising of these rods after the reactor is above 30% power shall be performed to verify instrumentation response.
j HATCH - UNIT 1 3.3-2 Amendment No. 139 4
@ H.[NT_ CONDITIONS 7OR_0PERATION SURVEILLANCE REQUIREMENTS
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4.3.C.1.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 ga to the overtravel position.
~ :.:.:. Scram Insertion Times 4.3.C.2.
Scram Insertion Times a.
All Doerable Control Rods After each refueling outage a.
The average scram insertion time all control rods capable of of all operable control rods at normal inser".lon shall be a reactor dome pressure 1950 psig '
scram time tested from the based on the de-energization of fully withdrawn position the scram pilot valve solenoids af ter a reactor dome pressure as time zero, shall be no greater of 950 psig has been attained.
than:
This testing must be complete before 40% rated thermal Notch Position Averabe Scram l
From Fully Insertion b.
Routine Time Tests Withdrawn Time (See) 46 0.358 At 16-week intervals,105 of j
36 1.096 the control rods capable of 26 1.860 movement with control rod 5
3.419 drive pressure shall be scram timed above 950 psig. When-0 b.
Three Out of Four Rods in a ever such scram time measure-Tva-by-Two Arrav ments are made, an evaluation P
shall be' made to provide The average of the scram inser-reasonable assurance that
..wo times for the three fast-proper control rod drive est control rods of all groups performance is being maintained.
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 Amendment No. 139
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8AS1,$~FOR LIMITING CONDI110NS FOR OPER& TION AND SURVEILLANCE REQUIREMENTS l
3.3.C.
Control Rod Drive System
- 1. Control Rod Drive Couplina intearity Limiting Conditions for Operation:
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 requi.ed 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 3
overtravel position.
]
- 2. Scram insertion Times limiting Conditions for Operation:
The control rod drive system is designed to bring the reactor sub-1 critical at a rate fast enough to prevent excess 1ve fuel damage. Analysis of the limiting transient shows that the negative reactivity rates resulting i
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.
1he 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. Grouping 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 j
have a scram insertion time to notch position 06 greater than 7 seconds.
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NATCH - UNIl 1 3.3-10 Amendment No. 139
LIMITING CONDITIONS FOR OPERATION SURVEIll.ANCE REQUIREMENTS 3.11. B. 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 prescrit,ed limits. If the l
LPGR 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) hcurs. If the limiting condition for operation is rsstoreri prior to expiration of 1
the specified time interval, then further progression to less than 25%
of r'ated thermal power is not required.
C.
Mir.19um Critical Power Ratio (MCPR' 4.11.C.I.
Minimum Critical Power Ratio (MCPR)
The minimum critical power ratio (MCPP)
MCPR shall be determined to be shall be equal to or greater than the equal to or greater than the operating limit MCPR (OLMCFR), which applicable limit, daily during is a function of scram time, core reactor power operation at 125%
power, and corn 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 red pattern as described in the bases 1.
The applicable limit determined for Specification 3.3.F.
i f rom 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 i
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 from figure 3.11.4 or 3.11.5 measured scram speed with respect using:
to Option A and Option 8 scram i
speeds.
If t is determined to a.
t = 1.0 prior to initial scren be less than zero, then the time measurements for the OLMCPR is evaluated at t = 0, cycle, perforned 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.
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m LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIRENENTS 3.11.C.
Minimum Critical Power Ratio (MCPR) 1 I
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If at any time during operation it is determined 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 poter to less than 255 of rated thermal power within the next four(4) hours. If the Limiting Condition for Operation is restored prior to expiration of i
the specified time interval, then further progression to less than 25% of rated thermal power is not required.
O.
Reportina Recuirements 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.
Amendment No. 139 HATCH - UNIT 1 3.11-2a l
BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.11.C.
Minimum Critical Power Ratio (MCPR) (Coritinued)
According to Figure 3.11.4 or 3.11.5, the 1005 power,100% flow operating limit MCPR (OLMCPR) depends on the average scram time, t, of the control rods, where:
s = 0 or ' ave
'B, whichever is greater t
-1 g
8 where: tA - 1.096 sec (Specification 3.3.C.2.a. scram time limit to notch 36)
V2 N1 e
vg w + 1.65 (Referer.ce 10]
l n
I Ng i
i=1 l
where: y = 0.822 see (mean scram time used in the transient analysis)
.018 sec (standard deviation of y) o=
n I Nggg
' ave
- 1"1 n
I Ng i=1 where: n = number of surveillance tests performed to date
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in the cycle Ng = number of active control rods measured in the ith surveillance test et - average scram time to notch 36 of all rods in the ith surveillance test N] = total number of active rods measured in 4.3.C.2.a The purpose of the MCPRg, and the Kp of Figures 3.11-3 and 3.11-6, respectively, is i
to define operating limits at other than rated core flow and power conditions. At
'ess than 100% of rated flow and power, the required MCPR is the larger value of the MCPRg and MCPR at the existing core flow and power state. The MCPRgs are p
established to protect the core from inadvertent core flow increases such that the 99.95 MCPR limit requirement can be assured.
The MCPRgs were calculated such that for the maximum core flow rate and the corres-ponding THERMAL POWER along the 1055 of rated steam flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRS were calculated at dif ferent points along the 1051 of rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPRg.
The core power dependent MCPR operating Ilmit MCPR is the power rated flow MCPR g
operating limit multiplied by the Kp factor given in Figure 3.11-6.
The K s are established to protect the core from transients other than core flow p
increases, including the localized event such as rod withdrawal error. The E s p
were deterinined tased upon the most limiting transient at the given core power level. (For further information on MCPR operating limits for off-rated conditions, i
reference NEDC 30474-P.(11))
HATCH - UNIT 1 3.11-4a Amendment No. 139
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Amendment No. 139
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UNITED STATES NUCLEAR REGULATORY COMMISSION n
WASHINGTON, D. C. 20555 GEOFGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Anendment No. 76 Licent.e No. NPF-5 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 2 (the facility) Facility Operating License No. NPF-5 filed by Georgia Power Company, ccting for itself. Oglethorpe Power Corporation Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensee) dated February 13, 1987, as supple-mented May 18, 1987, cormlies with the standards ami requirements of the Atomic Energy Act of 1954, as amended (the Aa.t), and the Comissior's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and l
safety of the public, and (ii) that such activities will be conc'ucted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the comon defenst 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 Comission's regulations and all applicable requirements have been satisfied.
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- 2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in.the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. hPF-5 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 76, are hereby incorporated in the license. The licensee shall ' operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION B. J. Youngblood, Director Project Directorate II-3 Division of Reactor Projects-I/II Attachnent:
Changes to the Technical Specifications Date of Issuance: June 1, 1987 I
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ATTACHMENT TO LICENSE AMENDMENT NO. 76 FACILITY OPERATING LICENSE NO. NPF,5 DOCKET NO. 50-366 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
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