NUREG/CR-3052, Amends 141 & 77 to Licenses DPR-57 & NPF-5,respectively Modifying Tech Specs to Permit Only One Recirculation Loop in Operation & to Implement Jet Pump Suveillance Recommendations of NUREG/CR-3052
| ML20215C127 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/10/1987 |
| From: | Youngblood B Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215C132 | List: |
| References | |
| RTR-NUREG-CR-3052 TAC-61900, TAC-61901, NUDOCS 8706170591 | |
| Download: ML20215C127 (82) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSIO'N n
' WASHINGTON, D. C. 20555 i
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321
.EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 141 License No. DPR-57 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 1 (the facility) Facility Doerating License No. DPR-57 filed by Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, ar.d City of Dalton, Georgia, (the licensee) dated June 20, 1986, as supplemented by letters dated July 22, 1986, and January 2,1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; i
i B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; J
'C.
There is reasonable assurance (1) that the activities authorized by
)
this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted j
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 common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied, s
8706170591 870610 ADOCK 050g1 PDR P
~2-2.
Accordingly, the license is' amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendmeht, 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.141, 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 COMMISS10h pg 006 t
/
5.
oungbl od, Director Pr je t Directorate II-3 Division of Reactor Projects-I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: June 10, 1987
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ATTACHMENTTOLICENSE'AMENDMEN'TNd.ht.1 FACILITY OPERATING LICENSE NO. DPR-57
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DOCKET NO. 50-321 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 area of change.
Remove Insert iv iv 1.1-1 1.1-l' 1.1-2 1.1 1.1-6 1.1-6 1.1-7 1.1-7 1.1-8 1.1-8 1.1-10 1.1 1.1-11 1.1-11 1.1-12 1.1-12 1.1-13 1.1-13 1.1-14 1.1-14 1.1-17 1.1-17 3.1 3.1-4 3.2-16 3.2-16 3.2-63 3.2-63 3.2-64 3.2-64
'3.6-9b 3.6-9b 3.6-9c 3.6-21 3.6-21 r
3.6-22 3.6-22 3.6-32 3.6-32 Figure 3.6-5 Figure 3.6-5 3.11-1 3.11-1 3.11-la 3.11-2 3.11-2 3.11-2a 3.11-2a 3.11-3 3.11-3 3.11-4 3.11-4 3.11-4a 3.11-4a 3.11-6 3.11-6 Figure 3.11-1 Figure 3.11-1 (Sheet 6)
(Sheet 6) l l
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I Section Section Page
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LIMITING CONDITIONS _,[gR 0PERATION SURVEILL Ahrt REQUIREMENTS 3.6.
PRIMARY SYST'.M 80VN3ARY-4.6.
PRIMARY SYSTEM BOUNDARY 3.6-1 A.
Reactor Coolant Heatq)
A.
Reactor Coolant Heatup 3.6-1 and Cooldown and Cooldown B.
Reatter Vessel Tempera ture B,
Reactor Vessel Temperature 3.6-1 and Pressure and Pressure C.
Reactor Vessel Head Stud C.
Reactor Vessel Head Stud 3.t-2 Tensioning Tensioning D.
Idle Recirculation Loop D.
Idle Recirculation Loop J. i -2 Sta rtup Sta rtup E.
Recirculation Pump Sta -t E.
Recirculation Pump Start 3.6-3 F.
Reactor Coolant Chemist.ry F.
Reactor Coolant Chemistry 3.6-4 6.
Retetor Coolant Leakagt 6.
Reactor Coolant Leakage 3.6-7 H.
Safety and Relief Valves H.
Safety and Relief Valves 3.6-9 Jet Pumps 1.
Jet Pumps 3.6-9b i
J.
Recirculation System J.
Recirculation System 3.6-9c K.
Structural Integrity K.
Structural Integrity 3.6-10 L.
Snubbers L.
Snubbers 3.6-10a 3.7.
CONTA!NMENT SYSTEMS 4.7.
CONTAINMENT SYSTEMS 3.7-1 A.
Primary Containment 3.7-1 8.
Standby Eas Treatment Sy: item B.
Standby Eas Treatment System 3.1-10 C.
Secondary Containment 3.7-12 0.
Primary Containment 3.1-13 Isolation Valves Isolation Valves 3.8.
RAD 10 ACTIVE MATERIALS 4.8.
RADIDACTIVE MATERIALS
- 3. 8 -1 A.
Miscellaneous Radioactive A.
Miscellaneous Radioactive 3.8-1 Materials Sources Materials Sources 3.9.
AUXILIARY ELELeRICAL SYSTEMS 4.9.
AUXILIARY E.LECTRICAL SYSTEMS 3.9-1 A.
Requirements for Reactor A.
Auxiliary Electrical 3.9-1 5tartup Systems Equipment NATCH - UNIT 1 iv Amendment No. 141
SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 1.1.
FUEL CLADDING INTEGRITY 2.1.
FUEL CLADDING INTEGRITY Applicability ADD 1icability I
The Safety Limits established to pre-The Limiting Safety System Settings serve the fuel cladding integrity apply apply to trip settings of the instru-to those variables which monitor the ments and devices which are provided to fuel thernal behavior.
prevent the fuel cladding integrity Safety Limits from being exceeded.
Objective Objective The objective of the Safety Limits is The objective of the Limiting Safety to establish limits below which the System Settings is to define the level integrity of the fuel cladding is-of the process variables at which auto-preserved.
natic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.
Specifications Specifications A.
Reactor Pressure > 800 osia and Core A.
Trio settinas Flow > 10% of Rated The limiting safety system trip set-The existence of a minimum critical tings shall be as specified below:
power ratio (MCPR) less than 1.07 for two-1 cop operation or 1.08 for 1.
Neutron Flux Trio Settinas single-1 cop operation shall constitute violation of the fuel cladding integrity safety limit.
a.
IRM High High Flux Scram Trio Settina B.
Core Thermal Power limit (Reactor The IRM flux scram trip setting Pressure s 800 osia) shall be 5 120/125 of full scale.
When the reactor pressure is 5 800 b.
APRM Flux Scram Trip Settino psia or core flow is less than 10% of (Ref uel or Start & Hot Standby rated, the core thermal power shall Mode) not exceed 25% of rated thernal power.
When the Mode Switch is in the REFUEL or START & HOT STANDBY position, the APRM flux scram j
C.
Power Transient trip setting shall be 5 15/125 of full scale (i.e.
5 15% of rated To ensure that the Safety Limit estab-thernal power).
lished in Specification 1.1.A and 1.1.E is not exceeded, each required c.
APRM Flux Scram Trio scram shall be' initiated by its Settings (Run Mode) expected scram signal. The Safety Limit shall be assumed to be exceeded (1) Flow Referenced Simulated-when scram is accomplished by a means Thermal Power Monitor Scram other than the expected scram signal.
Trio Settina When the Mode Switch is in the RUN position the APRM flow referenced simulated thermal power scram trip setting shall be:
HATCH - UNIT 1 1.1 -1 Amendment No. 141
a e
F.
SAFETY t1M?TS LIMITING SAFETY SYSYEM SETTINGS 1.1.D.
Reactor Water tevel (Hot or Cold 2.1.A.1.c.(1)
Flow Referenced simulated Shutdown Condition)
Thermal Power Monitor Trio -
Settina (Run Mode) (Continued)
Whenever the reactor is in the Hot or Cold Shutdown Condition with 5 1 0.58W + 62% - 0.58 AW irradiated fuel in the reactor vessel.
(Not to exceed 117%):
-the. water level shall be > 378 inches above vessel invert when fuel is where:
seated in the core.
S = Setting in percent of' rated thermal power (2436 MWt)
W = Total loop recirculation flow rate in percent of-rated (rated loop recircu-lation flow rate equals 34.2 x 106 lb/hr)
AW = Maximum measured difference between two-loop and single-loop drive flow for the same.
core flow in percent of rated recirculation flow for single-loop operation. The value is zero for two-loop operation, j
(2) Fixed APRM Hich Hich l
Flux Scram Trio Settino IRun Mode)
The.APRM fixed flux scram trip setting shall not be allowed to exceed 120% of rated thermal I
power.
l HATCH - UNIT 1 1.1-2 Amendment No. 141
s..
i BASES FOR SAFETY LIMITS 1.1 FUEL CLADDING INTEGRITY
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A.
Fuel Claddino intecrity Limit at Reactor Pressure >800 osia and' Core t
Flow >10% of Rated The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation the thern21 and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region
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where fuel damage could occur. Although it.is recognized that a departure from nucleate boiling would not necessarily result 'in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been' adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.
The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB (1), which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to i
calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X)
Boiling Length (L), GEXL, correlation. This Safety. Limit MCPR is increased by 0.01 for single-loop operation over the comparable two-loop value. (*)
The GEXL correlation is valid over the range of conditions used in the a
tests of the data used to develop the-correlation.
j The required inputs to the statistical model are the uncertainties listed in Table 5.2-1 of Reference 2 and the nominal valres of the core partmeters listed in Table S.2-2 of Reference 2.
1 HATCH - UNIT 1 1.1-6 Amendment No. 141 m
__._....._..s
m,,
1.1.A.
Reactor Pressure > 800 osia and Core Flow > 10% of Rated (Cont'd)
Th' ' basis for the uncertainties in the core parameters is gOen in '
e NEDD-20340 8 f and the basis for the' uncertainty in the GEXL correlation
- is given in NEDD-10958' 4 The power distribution is based on ) typical.-
-764 assembly core in which the rod pattern was arbitrarily chosen to produce p
a skewed power distribution having the greatest number of assemblies at;
.the highest power, levels. The worst distribution in Hatch Unit No.1 -
during any fuel cycle would not be as severe as the distribution used in the analysis. 'The method used to handle the uncertainty in the' statistical-a analysis to determine' the MCPR Cladding integrity Safety Limit for single-loop operation-.is described in Reference 4'.
8.
Core ~ Therinal Power Limit ' f Reactor Pressure s 800 osia)
At pressures below 800 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 psi. At'10w powers and flows this pressure-differential is maintained in the bypass region'of the core. Since the, pressure droptin the bypass region is essentially all elevation head.
the' core l pressure drop.at-low powers and flows will always be greater than 4'.56 psi.. Analyses'show that with a flow of 28x108 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle. power and-has a value'of.3.5 psi. Thus,! the: bundle flow with a' 4.56 psi. driving head will be' greater than 28x108 lbs/hr. Full scale ATLAS test data taken
' at pressures f rom 14.7 psia'to 800 psia indicate that the f uel assembly-critical' power at this flow is approximately 3.35 MWt. 'With the design -
peaking' factors this corresponds to a core thermal power of more than 50%. Thus, a core thermal power limit of 25% for reactor pressures below 800 psia is conservative.
J., Power Transient Plant safety analyses have shown that the scrams caused by exceeding any safety system setting will assure that the Safety Limit of 1,1.A.orJ 1.1.8 will not be exceeded. Scram times are thecked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves)' does not necessarily cause fuel damage. However, for this s
specification a Safety Limit violation-will be assumed when a scram is l
only accomplished by means of a backup feature of the plant design. The concept of not approaching a' Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.
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l NATCH - UNIT 1 1.1-7 Amendmenc No. 141
1 D.
Rebeter Water level (Hot or Cold Shutdown Condition)
For the fuel in the core during periods when the reactor i' *.hutdown,' consi-l deration must be given to water level requirements due to the effect of decay heat. If the water. level should drop below the top of:the fuel during i
this time, the ability to remove decay heat is reduced. This reduction i
in cooling capability could lead to elevated cladding temper 4tures and clad perforation in the event that the water level became less than two-
. thirds of the core height. The Safety limit has been established at 378 inches above vessel invert to provide a point which can be monitored and also
. provide adequate nargin.
E.
References 1.
" General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and Design Application," NEDO-10958-P-A and NE00-10958-A, January 1977.
2.
" General Electric Standard Application for Reactor Fuel (Supplement for.
United States)," NEDE-240ll-P-A.
l 3.
General Electric " Process Computer.Perf ormance Evaluation Accurar,',
NED0-20340, and Amendment 1, NEDO-20340-1, dated June, 1974 and December,1974, respectively.
4
'Edwin 1. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,'
NE00-24205, August 1979.
HATCH - UNIT 1 1.1-8 Amendment No. 141 d
-s.
ly BASES FOR tlMil.',dG SAFETY SYSTEM SETTINGS j
i 2.1 FUEL CLADDING INTEGRITY j
The abnormal operational transients applicable to operation of the HNP-1 Unit have been analyzed throughout the spectrum of planned operating conditions up to the thermal power condition of 2537 MWt. The analyses were based upon plant operation in accordance with the operating sep given in Figure 3-1 of Ref. 3.
l 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.
Transient analyses performed for each reload are given in Reference 1.
Models and model conservatism are also described in this referer.ce. As discussed in Reference 2,.the core-wide transient analyses for single-loop operation are 1
conservatively bounded by two-loop analyses. The flow dependent rod block and scram setpoint equations are adjusted for one-pump operation.
Steady-state operation without forced recirculation will not be permitted, except during startup testing. The analysis to support operation at various I
i HATCH - UNIT 1 1.1-10 Amendment No. 141 is.
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'q BASES FOR LIMITlHG SAFETY SYSTEM SETTINGS 5
' 2.1 FUEL CLA001NG INTEGRITY (Continued) l i
l power and flow relationships has considered operation with either one orltwo j
recirculation pumps, References 1 and 2.
1
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1 A.
Trip Settinos j
The bases for individual trip settings are discussed in the following para-g raphs.
1.
Neutron Flux Trio Settinos a.
IR4 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 APRM. The 5 decades are covered by the IRM by means of a range switch and the. $ 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 acconrnodate the increase in power level, the scram trip setting is also ranged up.
The most significant sources of reactivity change during the power in-I 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-ital limitation.of withdrawing control. rods, that heat flux is in equi-libriWn with the neutron flux and an IRM scram would result in a reat-tor shotdown well before any Safety Limit is exceeded.
In order to thsure 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 subtritical and the IRM system is not yet on-scale. This condition exists at quarter rod density. Quarter rod den-I sity-is illustrated in Figure 7.5-8 of the FSAR. Additional conserva-l HATCH - UNIT 1 1.1-11 Aw ndment No. 141
r e
BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.1.A.1.a.
IRM Flux Scram Trio Settina (Continued) tism was taken in this analysis by assuming that the IRM thannel closest l
to the withdrawn rod is bypassed. The results of this analysis show that I
.the reactor is scrammed and peak power limited to one percent of rated power, thus maintaining MCPR above the fuel cladding integrity Safety Limit.
l Based on the above analysis, the IRM provides protection against local control rod withdrawal errors and continues withdrawal of control rods.in sequence and provides backup protection for the APRM.
1 b.
APRM Flux Scram Trio Settino (Refuel or Start & Hot Standbv Mode),
l For operation in the startup mode while the reactor is at low pressure, the APRM scram setting of 15 percent of rated power provicles adequate thermal margin between the setpoint and the safety limit, 25 percent of rated. The margin is adequate to accommodate anticipated maneuvers asso-i ciated with power plant startup. Effects of increasing pressure at zero l
or low void content are minor, cold water f rom sources available during l
startup is not much colder than that already in the system, temperature coef ficients are small, and control rod patterns are constrained to be i
uniform by operating procedures backed up by the rod worth minimizer and the Rod Sequence Control System. Worth of individual rods is very low l
in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of'sig-nificant' power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage l
of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform fod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when l
_ reactor pressure is greater than 825 psig.
l c.
APRM Flux Scram Trio Settings (Run Mode)
The APRM Flux scram trips in the run mode consist of the flow referenced simulated thermal power monitor scram setpoint and a fixed high-high neutron flux scram setpoint. In the simulated thermal power monitor, the APRM flow referenced neutron flux signal is passed through a ft?ter-ing network with a time constant which is representative of the fuel dy-l namics. This provides a flow referenced signal that approximates the average heat flux or thermal power that is developed in the core during transient or steady-state conditions. This prevents spurious scrams, which have an adverse ef fect on reactor safety beer.use of the resulting thermal stresses. Examples of events which can result in momentary neutron flux spikes are momentary flow changes in the recirculation system flow, and small pressure disturbances during turbine stop valve and turbine control valve testing. These flux spikes represent no hazard to the fuel since they are only of a few seconds duration and less than 120% of rated thermal power. The flow independent portion of this scram setpoint must be adjusted downward during single-loop oosra-tion to account for lower core flow with respect to two-loop operation with the same drive flow.
1 HATCH - UNIT 1 1.1-12 Amendment No. 141
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i BASES FOR tlMITING SAFETY SYSTEM SETTINGS l
2.1.A.1.c.
APRM Flur Scram Trio Settinos (Run Mode) ( Cont'i nu ed )
The APRM flow referenced simulated thermal power monitor scram trip i
setting at full recirculation flow is adjustable up to 117% of rated power for two-recirculation loop and single-recirculation loop operations.
l' L
This reduced flow referenced trip setpoint will result in an earlier scram during slow thermal transients, such as the loss of 100*F l
feedwater heating event, than would result with the 1201, fixed high I
neutron flux scram trip. The lower flow referenced scram setpoint l
therefore decreases the severity (ACPR) of a slow thermal transient and allows lower Operating Limits if such a transient is the limiting abnormal operational transient during a certain exposure interval in the cycle.
The APRM fixed high-high neutron flux scram trip, adjustable up to 120%
of rated power for two-recirculation loop and single-recirculation loop operations, does not incorporate the time constant, but responds directly to instantaneous neutron flux. This scram setpoint scrams I
the reactor during f ast power increase transients if credit is not taken for a direct (position) scram, and also serves to scram the reactor if l
credit is not taken for the flow referenced scram.
2.
Reactor Vessel Water low level Scram Trio Settino (Level 3)
The trip setting for low level scram is above the bottom of the separator skirt, Figure 2.1-1.
This level is > 14 feet above the top of the active fuel. l This level has bon used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR Section 14.3 show that a scram at this level adequately protects the fuel and the pressure barrier. The designtied scram trip setting is at least 22 inches below the bottom of the homal operating range and is thus adequate to avoid spurious scrams.
HATCH - UNIT 1 1.1-13 Amendment No. 141
a.
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BASES FCR LIMITlHG SAFETY SYSTEP r,ETTINGS l
2.1. A.3. ' Turbine Stoe Valve Closure Scram Trio settinas J
The turbine stop valve closure scram trip anticipates the pressure, neutron j
flux and heat flux increase that could result from rapid closure of the 1
turbine stop valves. Nith a scram trip setting of 510 percent of valve I
closure from full open, the resulti.nt increase in surface heat flux is.
i limited such that MCPR remains above the fuel cladding integrity Safety Limit during the worst case transient that assumes the turbine bypass is closed. This scram is bypassed when turbine steam flow is below that' corresponding to 30% of rated thermal power, as taeasured by turbine first stage pressure.
4.
Turbine Control Valve Fast Closure Scram Trio settina This turbine control valve fast closure scram anticipates the pressure, neutron flux, and heat. flux increase that could result f rom fast closure of J
-the turbine control valves due to load rejection exceeding the capability
- i of the turbine bypass. The Reactor Protection System initiates a scram when fast closure of the control valves is initiated by the fast' acting solenoid valves. This is achieved by the action of the fast acting solenoid 1
valves in rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is-sensed by pressure switches whose contacts form the one-out-of-two-twice logic 1
input to the. reactor protection system. This trip setting, a nominally 50% greater closure time and a dif ferent valve characteristic from that of the. turbine stop valve,: combine to produce transients.very similar and-no more severe than for the stop valve. This scram is bypassed when turbine steam flow is below that. corresponding to 30% of rated thermal power, as measured by turbine first stage pressure.
5.
Main Steam Line isolation Valve Closure Scram Trio Settino The main steam line isolation valve closure scram occurs within 10% of valve movement from the fully open position and thus anticipates the neutron flux and pressure scrams which remain as available backup pro-tection. ' This scram function is bypassed automatically when the Mode l
Switch is not in the RUN position.
6.
Main Steam Isolation Valve Closure on low Pressure The low pressure isolation of the main steam lines at 825 psig was provided
]
to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel, which might' result from a pressure regulator failure causing inadvertent opening of the control and/or bypass valves.
1 i
HATCH - UNIT 1 1.1-14 Amendment No. 141
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i E'..5 FOR tlMlilNG SAFETY SYSTEM SETTINGS l
2.1.C.
References-
" General Electric Standard Application for Reactor Fuel (Supplement.for 1.
'J' United States)," WEDE-24011-P-A.
l 2.
'Edwin 2 Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,'
NE00-24205, August 1979.
3.
' Average 4 Power Range Monitor, Rod Block Monitor and Technical Specifications improvement (ARTS) program for Edwin I. Hatch Nuclear Plant Units 1 and 2,"
NEDC-30474-P, December 1983.
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HATCH - Oh!T 1 1.1 17 Amendment No. 141 i-L
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BASES FOR LIMITING CONDITIONS FOR OPERATION R
13.2.F.5. Core 'SDray Pumo Disth'arce Flow A differential pressure transmitter is provided downstream of each core spray
. pismp~ to indicate the condition 'of each pump.. To protect the pumps from over-
' heating at low flow rates a' minimum flow bypass line..which routes water from -
- the ' pump' discharge to the suppression chamber,. is-j,rovided.. A single motor-operated valve-controls the condition of each bypass line. The minimum flow.
bypass valve. automatically. opens upon sensing. low flow'in.the discharge line, tThe, valve automatically closes whenever the flow is above the low flow setting.
i 1
~
- 6. Core' Sprav Loaic Power Failure Monitor The Core Spray Logic Power Failure Monitor monitors the availability of power.to the logic system..In the event of loss of availabflity of power
~
~to theLlogic system, an alarm is annunciated in the control room.
l h
G.
Neutron Monitorino instrumentation Which initiates Control Rod Blocks i
- g (Table 3.2-7)^
These control red block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to the fuel cladding integrity.
Safety Limit. The trip logic for this. function is 1 out of n:.
e.g., any trip.
on one'of'six APRM's, eight IRM's or four SRH's will result in a rod block.
t.
j The minimum instrument channel requirements assure. sufficient instrumentation l
to assure that the single failure criteria is met.
ra i
a.
InoDerative 1
This rod block assures that no control rod is withdrawn during low neutron flux level operations unless proper neutron monitoring capa--
bility is available, in that all SRM channels are in service or properly bypassed.
b.
Not Fully inserted
~Any source range monitor not fully inserted' into the core when the SRM count rate level is below the retract permit level will,cause a rod block. This assures that no control rod is withdrawn unless all SRM 1
detectors are' properly inserted when they must be relied upon to pro-l vide the operator with a knowledge of the neutron flux level.
c.
Downscale d
This rod block assures that no control rod is withdrawn unless the SRM count rate is above the minimum prescribed for low neutron flux level monitoring.
a HATCH - UNIT 1 3.2-63 Amendment No. 141
- e m.
j.
J' BASES FOR LIMITINti CONDITIONS FOR OPERATION 3.2.G.1.d.
Voscale This rod block assures that no control is withdrawn unless the SRM detectors are properly retracted during reactor startup. This setting is selected at the upper end of the range over which the SRM is designed to detect and.measurs neutron flux.
2.
,IEd The trip 'fogic for this function is 1 out of 8; any trip on one of the eight IRM's will result in a rod block. The IRM rod block function provides local as well as gross core protection.
4.
InoDerative This rod block assures that no control rod is withdrawn unless the IRM's are in service, b.
Not Fully inserted (Refuel and Start & Hot Standby Model This rod block asiures that no control rod is withdrawn during low neutron flux level operations unless proper neutron monitoring capability is available in that all IRM detectors are properly located.
3 c.
Downscale A downscale indication of 5 5/125 full scale on an IRM is an indication that the instrument has f ailed or the instrument is not sensitive enough.
j in either case, the instrument will not respond to changes in control rod motion end thus, control rod motion is prevented. The downscale trip is set at > 5/125 f ull scale. This rod block trip is bypassed when the IRM f
is on the range 1, q
l d.
Hich Flux If the IRM channels are in the worst condition of allowed bypass, the scaling arrangement is such that for unbypassed IRM channels a rod block signal is generated before the detected neutron flux has increased by more than a factor of 10, 3.
APRM The trip logic for this function is I cut of 6; any trip on one of the six APRH's will result in a rod block. The APRM rod block function provides gross core protection; 1.e., limits the gross core power increase f rom withdrawal of control rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater than the fuel cladding integrity Safety Limit under normal operating conditions, a.
Inocerative This rod block assures that no control rod is withdrawn unless the APRM's are in service.
HATCH - UNIT 1 3.2-64 Amendment No. 141 l
~
3 3
i i
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS j
i b.
With the relief valve function and/or the low low set function of more than one of the above required reactor coolant system 1
relief / safety valves inoperable, be in at least HOT SHUTOOWN with-in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, j
3.6.1.
Jet Pumos 4.6.!.
Jet Pumps All jet pumps corresponding to the Each of the jet pumps shall be operating loop (s) shall be operable demonstrated operable prior to
{
during STARTUP and RUN modes by thermal power exceeding 25% of
(
meeting at least one of the following rated power: following recircu-
]
requirements:
lation pump restarts; following 4
any unexpected or unexplained 1
For any specific core flow change in core flow, jet pump loop condition, each individual jet pemp flow, recirculation pump flow, or 1
flow shall not differ by more than core plate differential pressure; 10% of the everage loop jet pump and at least on'ce per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by flow from the normal range
- of recording jet pump loop flows, average loop jet pump flows recirculation pump flows, recircu-experienced for these flow lation pump speeds, and individual conditions, or jet pump flows (D/P); and verifying that neither of the folicwing 2.
For any specific core flow conditions occur:
)
condition, each individual jet
?
pump diffuser to lower plenum 1.
The recirculation pump flow /
differential pressure (D/F} shall speed ratio deviates more than not dif fer by rnore than 20% of 5% from the normal range,* or the average loop D/P from the normal range
- of average loop jet 2.
The jet pump loop flow / speed pump D/Ps experienced for these ratio deviates more than 5%
flow conditions, from the normal range.*
With one or more jet pumps exceeding If any required jet pump fails to the above requirements, evaluate the meet either or both of the above reason for the deviation, and be in 4.6.I.1 or 4.6.1.2 Sur veilience HOT SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the Requiremer.ts, review the jet pump circumstance that one or more jet operability as defined in the LCO pumps are verified to be inoperable.
Section 3.6.1 and in BASES Section 3.6.1.
- Normal expected operating range based on data obtained from operating experience.
HATCH - UNIT 1 3.6-$b Amendment No. 141
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS j
3.6.J.
Recirculation System-4.6.J.
Recirculation System
)
1.
Core-thermal power shall not exceed 1.
Recirculation pump speeds shall.be 1% of rated thermal power uithout recorded atileast once per day.
I forced recirculation.
2.
With only one recirculation <1oop i
2.
Whenever the reactor is in the in operation, verify that the 57ARTUP or RUN modes, at least one reactor operating conditions are i
recirculation loop shall be in outside the Operation Not Allowed operation.
Region in Figure 3.6-5:
3.
The requirements applicable to (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 />, single-loop operation as identified in Sections 1.1. A. 2.1. A. 3.1.1 (b) Whenever thermal power has 3.2.G. 3.11.A. and 3.11.C shall be been changed by at least 5% of in effect following the removal of rated thermal power and steady-i one recirculation loop from service, state conditions have been or the unit shall be placed in.the reached.
HOT SHUT 00WN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUT 00WN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.
With only one recirculation loop in operation and the unit in the
-)
Operation Not Allowed Region, il specified in Figure 3.6-5, initiate action within 15 minutes ~ to place the unit in the Operation Allowed Region, identified in Figure 3.6-5, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Otherwise, place the reactor in the HOT SHLn30WN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
5.
Following one pump operation the i
discharge valve of the low speed pump may not be opened unless the speed of the faster pump is less than 50% of its rated speed.
}
I i
HATCH - UNIT 1 3.6-9c Amendment No. 141
i I
m 1
. _[ i j
i
.i.
' 8ASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.6.H..: Relief / Safety Valves (Continued)
Experience in relief / safety valve cperation shows that a testing of 50 per-cent of.the valves per year is adequate to detect f ailure.or deteriorations.
l
'The relief / safety valves are benchtested every second operating cycle to ensure that their set points are within the tolerance given in Specification 2.2. A'..The relief / safety valves are tested in place at low reactor presture once per operating cycle to establish that they will open and pass steam.
The requirements established above apply when the nuclear system can be pres-surized above ambient conditions. These requirements are applicable at nu-clear system pressures below nomal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed. However, these transients are much less severe in tems of pressure, than those. starting at rated condi-tions. 'The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
i, The low. low. set-(LLS) system lowers the opening and closing setpoints on four
' preselected relief / safety valves. The LLS system lowers the setpoints af ter any relief / safety' valve has opened at its normal steam pilot setpoint when a concurrent high reactor vessel steam dome pressure scram signal is present.
The purpose of the LLS.is to mitigate the induced high frequency loads on the containment and thrust loads on the SRV discharge line. The LLS system increases the amount of reactor depressurization dur'eng a relief / safety valve blowdown because the lowered LLS setpoints keep the four selected LLS relief / safety valv,es open for a longer time. The high reactor vessel. steam dome pressure signal for the LLS logic is provided by the exclusive analog trip channels. The purpose 'of installing special dedicated steam dome pres-sure channels is to maintain separation f rom the RP.S high pressure scram functions.
l 1.
Jet Pumps Failure of a jet pump nczzle assembly hold down mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown fol-
{
lowing the design basis double, ended line break. Therefore, if a failure occurred, repairs must be made, 1
HATCH - UNIT 1
$.6-21 Amendment No. 141
1
~
8ASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEllLANCE REOUIREMENTS 3.N I. Jet Pumos (Continued).
A nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible without an initial norrie riser system failure.
One of the acceptable procedures for jet pump surveillance, identified in NURE6/CR-3025, Reference 2, was adopted for Hatch Unit 1.
The surveillance '
is performed to verify that neither of the following conditions occur:
(a) The Recirculation Pump Flow / Speed Ratio deviates by more than 55 from the normal range, or (b) The Jet Pump Loop Flow /$ peed Ratio deviates by more than 55 from the nonnel range.
If either criterion is failed, then the procedure calls for comparing I
either the individual jet pump flaw or individual jet pomp diffuser to lower plenum differential pressures to the criteria of the Limiting.
Conditions for Operation (LCO). If the LCD criteria are not satisfied and pump speed.is less than 605 of rated, it may be necessary to increase pump speed to above 605 of rated and repeat the measurements-before declaring a jet pump inoperable. In this case, it is-reconsnended that close monitoring and increased recirculation pump speed should be performed only if the criteria are exceeded by an amount to-be deternbed f rom previous plant operating experience.
3.6.J.
Recirculation System Operation with a single reactor coolant system recircciation pump is allowed, provided that adjustments to the flow referenced scram and APRM rod block setpoints, MCPR cladding integrity Safety Limit. MCPR Operating Limit, and MAPLHG4 limit are made. An evaluation of the performance of the ECC$
with single-loop operation has been perfonned and determined to be acceptable Reference 4.
Based on this Reference, a MAPLH6R factor of 0.75 is applied to the specifications Figure 3.11-1 (Sheet 6). To account for increased uncertainties in the total core flow and TIP readings when operating with a single recirculation loop, a 0.01 increase is applied to the MCPR cladding integrity Safety Limit and MCPR Operating Limit over the comparable two-loop values. The flow referenced simulated thermal power scram and rod block setpoints for single-recirculation-loop operation is reduced by the amount of a&W, where a is the flow reference slope for the rod block monitor and 4W is the largest difforence between two-loop and single-loop offective drive flow when the active loop indicated flow is the same. This adjustment is necessary to preserve the original relationship between the rod block and actual effective drive flow.
l When restarting an idle pump, the discharge valve of the idle loop is required to remain closed until the speed of the faster pump is below 505 of its rated speed to provide assurance that when going from one-to two-loop operations, excessive vibration of the jet pump risers will not occur.
The postibility of experiencing limit cycle oscillations during single-loop operation is precluded by restricting the core flow to greater than or equal to 45% of rated core flow when core power is greater than the 805 rod line.
This requirement is bised on General Electric's reconenendations contained in
$1L 380, Revision 1, which defines the region where the limit cycle oscillations are more itkely to occur.
HATCH - UNIT 1 3.5-22 Amendment No. 141
i
. BASES
,j I
3 /4. 6. L. SNUBBERS (Continued)z f
' The. service lif e of a snubber is evaluated via ma. uf acturer input and.
information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation. area, in high temperature a rea, etc... ). The requirement to monitor the snubber service life is
. included to ensure that the snubbers periodically undergo a performance
. evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of. snubber service life. The requirements for the maintenance of records and the snubber service life review are not intended to af fect plant operation.
References:
(1) Report, H. ' R. Erickson, Bergen Paterson to. K. R.' Goller, NRC, October 7, 1974.
Subject:
' Hydraulic Shock Sway Arrestors. -
'(2) NUREG/CR - 3052, 'Closcout of IE Bulletin 80-07: BWR Jet Pump Assembly Failure,' Published November.1984.
(3) " General Electric BWR Licensing Report:. Average Power Range Monitor, Rod Block Monitor, and Technical Specifications Improvement (ARTS)
Program for Edwin I. Hatch Nuclear Plant Units 1 and 2 "
NEDC-30474-P, December 1983.
i
'(4) "Edwin 1. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"
NE00-24205, August 1979.
~
i l
)
1 d
HATCH - UNIT 1 3.6-32 Amendment No. 141
1 1
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(01.tvW %) WIM04 *tvWWBH13W03 HATCH UNIT 1 Amendment No. 141
)
1 l
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(.-
LIMITING CONDlil0NS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.11.. FUEL RODS.
4.11.
FUEL RODS 1
1 ADD 11cability ADD 11cability
)
a The Limiting Conditions for. Operation' The Surveillance Requirements apply U
associated with the fuel-rods apply to to the parameters which' monitor.the those parameters which monitor the
' fuel. rod operating conditions.
- fuel rod operating conditions.
^
pp.iective; Ob.iective The Objective of the Limiting'Condi-The Objective of the Surveillance.
tions for. Operation is to assure the Requirements is to specify the> type:
performance of the fuel rods.
.and frequency of surveillance to be applied to the fuel rods..
j Specifications Specifications
- l A.
Averace P'lanar Linear Heat Genera-A.
Averace Planar linear Heat' Genera-l tion Rate (APLHGR) tion Rate (APLHGR)
]
1 During power operation, the APLHGR The APLHdR' for each type of fuel as
~
for all core locations shall not.
'a f unction of average planar.
exceed'the appropriate APLHGR limit exposure shall be. determined. daily.
for those core locations. The APLHGR during reactor operation at,> 25%
limit, which is a function of average rated thermal power.
planar exposure and fuel ~ type, is the appropriate value f rom Figure 3.11-'1, sheets 1 through 5, multiplied by the.
smaller of the two MAPFAC. factors de :
termined f rom Figure 3.11-1, sheets 6 and 7.
For single-loop operation, the MAPFAC is a constant value of p
0.75 when power is greater than 52%
of rated thermal power..For power less than 52% of rated thermal power, the MAPFAC is the same as the.
3 p
l comparable two-loop value (Figure
~
3.11.1, sheet 6).
If at any time during operation it is determined l
by. normal surveillance that the I
limiting value for APLHGR is being
't exceeded, action shall be initiated l
within 15 minutes to restore operation e
to within the prescribed limits.
l If the APLHGR 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 l
progression to less than 25% of rated thermal power is not required.
)
HATCH - UNIT 1 3.11-1 Amendment No. 141 i
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9.
3 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 8.
Linear Heat Seneration Rate (LHGR) 8.
Linear Heat Generation Rate RHGR)
During' power operation, the LHGR as The LHGR as function of core a function of core height shall not height shall.be checked daily dur-exceed the limiting value.shown in ing reactor operation at > 25%
Figure 3.11-2 for 7 x 7 fuel or the rated therini power.
limiting value of 13.4 kw/ft for 8 x 8/-
8 x 8R fuel. If at any time during l
HATCH - UNIT 1 3.11-la Amendment No. 141
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s LIMITING C'0TDITIONS FOR OPERATION SURVEILMNJ..REGUIREMENTS J
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3.11.B.
Linear Heat Gere gt_ ion Ratt (LHGR)
(Continued) i I
operation it is detemined by nomal s
surveillance that the limiting value T)'*
for LHGR is being 6xceeded, action shall i
WMtiated within 15 minutes to j
restore operation to withfo x
J
.[;;
- the prescribed limits, if the
)
LNGTI is not returmd to within the i
prescribed limits within two (2)
V houds, then reduce reactor power to leu than 25% of rated themal power witnin the next four (4) hours. If the limiting condition for ope'ation r
is restored prior to expiration of the specified time interval, then further progression to less than 25%'
q of rated thermal power is'not requi red.
C.
Minimum CriM jal Pene,r,, Ratio (MCPR)_
4.11.C.).
Minimum Critical Power Ratio (MCPR)
J The minimum critica'.' power ratio. (MCA) l MCPR shall be determined to be for two-loop Operation shall be equal' equal to or greeter than the 2
to or greater than the ' operating
~
applicable limit, daily during s
limit MCPR (OLMCPR), which is a reactor power corration at 1 25%
function of scram time, core rated thermal power and following power, and core flow. for 25% 5 any change.in 9twer level or dis-power < 30%, the OLMCPR.is givem in tribution that sould cause opera-c Figure 3.11.7.
For power 2 30%.
tion with a limiting control rod the CLHCP9 is the greater of either:
pattern as described in the bases for Specification 3.3.F.
1 1.
The anplicable limit detemined
' f rom Figure 3.11.3, or i
4.11.C.2.
Minimum Critical Power Ratio Limit i
2.
The applicable limit from either Figures 3.1.1.4. 3,11.5 The MCPR limit at rated flow and 4
a or 5.11.6, multirlied by the rated power shall be deterniined for
[;(
K f actor detemined f rom each f uel type, BXBR. P8XOR, 7X7 p
Figure 3;11.7, where:
from figures 3.11.4, 3.11.5, and j
3.11.', respectively using:
)
t - 0 or Tava 58
. whichever is a.
t=1.0 prior to initial scram j
(
~ y. i.y ~
gr ater time measurements for the cycle, perferned in accorcance j
SA = 0.90 sec (Specification 3.3.C.2.a.
with specifications 4.3.C.2.a.
i scram time limit to 20% fnsertica j
from fully withdrawn) or i
i 1/s b.
i as defined in specification
~
tg = 0.710+1.65 H,[
(0.053)(F.ef.7]l 3.11.C.
n INi The determination of the limit i=1 j
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 i' J,11-2 Amendment No. 141
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LIMITINE_,r0NDIT10NS FOR OPERAT!_2N-SURVEILLANCE REOUIREMENTS
'3.11.C.
Minimum Critical Power Ratio (MCPR) n L
I Njtj tave = i=1 n
I Nj
. i=1.
n = number of surveillance tests performed to date in cycle Nj = number of active control rods measured in the ith surveillance test tg = Average scram time to 20% in-sertion f rom f ully withdrawn of all rods measured in the ith surveillance test and, 3
N) = total number of active rods measured in 4.3.C.2.a.
For single-loop operation, the MCPR limit is increased by 0.01 over the comparable two-loop value, If at any time during operation it is determined by nornal 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 thernal power within the next four(4) hours. If ti.e Limiting Condition for Operaticq is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated thermal power is not required.
I D.
Reportino Reovirements i
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 actior, is taken, as described, a thirty-day written report will meet the requirements of this specification.
HATCH - UNIT 1 3.11-2a Amendment No. 141 t
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BASES FOR LIMITING CONDITIONS FOR ODERATION AND SURVEILLANCE RE0VIREMENTS
,3.11.
FUEL RODS A. ' Averace Planar Linear Heat Generation Rate-( APLNGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the.10 CFR 50, Appendix K, even considering the postulated effects of fuel pellet densification.
The peak cladding temperature following a postulated loss-of-coolant acci-dent is primarily a function of the average heat generation rate of.all the rods of a fuel assembly at.any axial location and is only dependent second-arily on the rod to rod power distribution within an assembly. Since ex-pected local variations in power distribution within a fuel assembly af fect the calculated peak clad temperature by less than i 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation ~ rate'is sufficient to assure that calculated temperatures-conform to 10 CFR 50.46. The limiting value for APLHGR at rated conditions is shown in Figures 3.11.1. sheets 1 thru 5.
A flow dependent correction factor incorporated in to Figure 3.11-1 (sheet 7) is applied to the rated conditions APLHGR to assure that the 2200'F PCT limit is complied with during LOCA initiated from less then rated core flow. In addition, other power and flow dependent corrections given in Figure 3.11-1 (sheets 6 and 7) are applied to the rated conditions APLHGR limits to assure, that the f uel thermal-cechanical design criteria are met during abnormal ~
transients initiated from of f-rated-conditions for two-loop and single-loop operations. References 2'and 8.
For single-loop operation, a 0.75 multiplica-tion factor to APLHGR limits for all fuel bundle types conservatively bounds 1 hat required by Reference 2.
l The calculational procedure used to establish the APLHGR shown in Figures-3.11.1, sheets 1 thru 5. is based on a loss-of-coolant ~ accident analysis.
The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.
A list of the significant plant input parameters to the loss-of-coolant ~
accident analysis is presented in Table 3-1 of NE00-24086(s). Further l
discussion of the APLHGR bases is found in NEDC-30474-p(*).
ARTS (*)gle-loop operation (SLO), the most restrictive of the SLO and For sin MAPLNGRS will define the Limiting Conditon for Operation.
,j q
i I
HATCH - UNIT 1 3.11-3 Ameniment No.141
4; BASES FOR LIM] TING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS
- 3.11.B.
Linear Heat Generation Rate (LHGR)
This-specification assures that the linear heat generation rate in any rod
. is less than the design linear heat generation if fuel pellet'densification is' postulated. The power spike penalty specified for 7 x.7 fuel is based on the analysis presented in Section 3.2.1 of Reference 4 and References 5 and 6, and assumes a linearly increasing variation in axial gaps between core bottom and.
top, and assures with a 95% confidence, that no more than one fuel rod exceeds the design linear heat generation rate due to power spiking. The LHGR as a-function of core height shall be checked daily during reactor operation at t 25%.
power to determine if fuel burnup, or control rod movement has caused changes in power distribution. For LHGR to be a limiting value below 25% rated
' therwal power, the ratio of peak LHGR to core average LHGR would have to be greater than 9.6, which is precluded by a considerable margin when employing any permissible control rod pattern.
C.
Minimum Critical Power Ratio (MCPR)
The required operating limit MCPR as specified in Specification 3.11.C. is derived from the established fuel cladding integrity Safety Limit MCPR and an analysis of abnornal operational transients presented in References 1, 2, and 8.
Various transient events will reduce the' MCPR below the operating MCPR.
To assure that the fuel cladding integrity safety limit is not violated during anticipated abnornal operational transients, the most limiting transients have been analyzed to determine which one results.in the largest-reduction in critical power ratio ( A MCPR). Addition of the. largest A MCPR to the safety limit MCPR gives the minimum operating li. nit MCPR to avoid violation of the safety limit should the most limiting transient occur.
The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
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HATCH - UNIT 1 3.11-4 Amendment No. 141 l
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a BASES FOR t.lMITING CONDITIONS FOR OPERATION AND SURVEILLA3 [ REOUIREMENTS 3.11.C.
Minimum Critical Power Ratio (MCPR) (Continued)
The purpose of the MCPR, and the Kp of Figures 3.11.3 and li.11.7, respectively, is f
to define operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power, the required MCPR is the larger value of the MCPRr and MCPRp at the existing core flow and power state. The MCPRys are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.
The MCPR s were calculated such that for the maximum core flow rate and the corres-f ponding THERMAL POWER along the 105% 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 different points along the 105% of rated steam flow control line corresponding to different core flows. The calculated MCPR at a given point of core flow is defined as MCPR.f The core power dependent MCPR operating limit MCPR is the power rated flow MCPR p
operating limit multiplied by the Kp factor given in Figure 3.11.7.
The K s are established to protect the core f rom transients other than core flow p
t increases, including the localized event such as rod withdrawal error. The Kps were detennined based upon the most limiting transient at the given core power level. (For further information on MCPR operating limits for of f-rated conditions, ref erence NEDC-30474-P.(*))
When operating with a single-recirculation pump, the MCPR Safety and Operating Limits are increased by an amount of 0.01 over the comparable values for two-recirculation pump operation.(2) i l
f J
l l
MATCH - UNIT 1 3.11 -4 a Amendment No. 141 l
l._.
4 BASES FOR t!MITING CONDITIONS f0R OPERATION AND SURVEILLANCE RE0VIREMENTS 3.11.E.
References I
l 1.
' General Electric Standard Application for Reactor Fuel (Supplement for United States),' NEDE-24011-P-A.
2.
"Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operaticn,'
NE00-24205 August 1979.
j 3.
" Loss-of-Coolant Analysis for Edwin I. Hatch Nuclear Plant Unit 1,"
NE00-24086,- December 1977.
4.
" Fuel Densification Ef fects on General Electric Boiling Water peactor Fuel *, Supplements 6, 7 and 8, NEDM-10735, August,1973.
Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 16, 1974 (USA Regulatory Staff).
6.
Conrnunication:
V. A. Moore to I. 5. Mitchell, ' Modified GE Model for Fuel Densification" Docket 50-321, March 27, 1974 7.
Letter f rom R. H. Buchholz (G. E.) to P. S. Check (NRC), " Response to NRC request for information on ODYN computer model', September 5,1980.
8.
" Average Power Range Monitor. 'dod Block Monitor and Technical Specification l Improvement ( ARTS) Program fo'. Edwin I. Hatch Nuclear Plant Units 1 and 2,'-
NEDC-30474-P December 1983.
l HATCH - UNIT 1 3.11-6 Amendment No. 141
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.s ucq*%i UNITED STATES :
-[
w NUCLEAR REGULATORY COMMISSION r
.j.
WASHINGTON, D. C. 20$55
%... f
' GEORGIA POWER COMPANY; j
0GLETHORPE POWER CORPORATION
. MUNICIPAL ELECTRIC AUTHORITY OF' GEORGIA CITY 0F DALTON, GEORGIA 1.
DOCKET NO.'50-366 EDWIN 1.~ HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 77 License 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, acting for itself, Oglethorpe Power Corporation.iMunicipal Electric Authority of Georgia, and City of Dalton, Georgia', (the licensee) dated June 20, 1986, as.supolemented by letters dated July 22, 1986 and January 2, 1987, complies with
~ he standards ~and requirements of the' Atomic Energy Act of 1954, as-t amended (the Act) ; and the-Comission'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 safety of the public, and (ii) that such activities will be conducted l
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 common defense and security or to the health and safety of the public; and-1 E.
The issuance of this' amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and allL app 11 cable requirements have
)
been satisfied.
t '
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. NPF-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. 77, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specificatinns.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR TH N CLEAR REGULATORY COMMISSION bN B. Jr Yo ngblood Director Profect Director te II-3 Divisio of Reactor Projects-I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:. June 10, 1987 i-4
.n~
ATTACHMENT TO LICENSE AMENDMENT NO. 77 FACILITY'0PERATING 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 l
contain vertical lines indicating the area of change. corresponding overleaf i
pages are provided to maintain document completeness.
Remove Insert III III XI XI XII XII 2-1 2-1 2-4 2-4 B2-1 B2-1 B2-2 B2-2 B2-3 B2-3 B2-6 B2-6 B2-7 B2-7 B2-9 B2-9 B2 B2-10 B2-13' B2-13 3/4 2-1 3/4 2-1 3/4 2-41 3/4 2-41 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 3-40 3/4 3-40 3/4 3-40a 3/4 3-40a 4
3/4 4-1 3/4 4-1 3/4 4-la 3/4 4-la 3/4 4-lb 3/4 4-2 3/4 4-2 3/4 4-2a 3/4 10 3/4 10-4 B3/4 1-2 B3/4 1-2 B3/4 2-1 B3/4 2-1 B3/4 2-3 B3/4 2-3 B3/4 2-4 B3/4 2-4 B3/4 2-6 B3/4 2-6 B3/4 4-1 B3/4 4-1 B3/4 4-la B3/4 4-lb B3/4 4-7
)
.i INDEX SAFETY LIMITS AND LIMIT 1NG SAFETY SYSTEM SETTINGS-SECTION PAGE l
2.1 SAFETY LIMITS THERMAL POWER ( Low Pressure or Low F1ow)...................... 2-1 THERMAL POWER (High Pres sure and High Flow)...................
2-1 Reactor Coolant' System Pressure...............................
2-1 Reactor Vessel Water Leve1....................................
2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints...........
2-3 BASES-i 2.1 SAFETY' LIMITS l
l THERMAL POWER ( Low Pres sure or Low F1ow)...................... B 2-1 THERMAL POWER (High Pres sure and High Flow)...................
B 2-2 Reactor Coolant System Pressure...............................
B 2-8 Reactor Vessel Water Leve1....................................
B 2-8 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints...........
B 2-9
2.3 REFERENCES
B 2-13 i
)
P 1
HATCH-UNIT 2 III Amendment No. 77 1
INDEX L'IMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
'E SECTION PAGE W
3/4.0 APPLICABILITY.............................................
3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 4
3/4.1.1 SHUTDOWN MARGIN........................................
3/4 1-1 i
g.
3/4.1.2 REACTIVITY AN0MALIES...................................
3/4 1 2 1
^
3/4.1.3 CONTROL RODS f
Con t rol Ro d O pe rab i l i ty................................ 3/41-3 Control Rod Maximum Scram Insertion Times..............
3/4 1-5 Control Rod Average Scram Insertion Times.............. 3/4 1-6 q
',)
Four Control Rod Group Scram Insertion Times........... 3/4 1-7 Con trol Rod Scram Accumul ators......................... 3/4 1-8 1
Control Rod Dri ve Coupl i n g............................. 3/4 1-9 Control Rod Position Indication........................ 3/4 1-11 Control Rod Dri ve Housing Support...................... 3/4 1-13 a E.
3/4.1.4 CONTROL ROD PROGRAM CONTROLS e
Rod Worth Minimizer....................................
3/4 1-14 i
y
[
Rod Sequence Control Sys tem............................ 3/4 1-15 1
Rod Block Monitor......................................
3/4 1-17 1
f, 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM..........................
3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLAN AR LINEAR HEAT GENERATION RATE............. 3/4 2-1 3/4 2.2 APRM SETP0lNTS.........................................
3/4 2-5 I
3/4.2.3 MINIMUM CRITICAL POWER RATI0...........................
3/4 2-6 B
3/4.2.4 LI NEAR HEAT GENERATI ON RATE............................ 3/4 2-8 l
HATCH-UNIT 2 IV f4 O
^
b i
(
INDEX -
BASE 5 CEOTION M
- 3/4.0 APPLICA!!LITY
,. 3 3/4 0-1 3/4.2 REAOTIVITY 00h-"ROL SYSTEMS l
1 3/4.1.1 'SEiDOWN MARCIN 3 3/4 1-1 3/4.1.2 RIACTIVITY ANOMALIES 3 3/4 1-1 l
l 3'/4.1. 3 C0hiROL-RdDS 3 3/4 1-2 3/4.1.4 CohiROL RCD PROGRAM C0hiROLS 3 3/4 1-3
]
3/4.1.5 STANDBY LIQUID CohTROL SYSTEM 3 3/4 1-4 3/4.2 POWER DI53:3" TION LIMITS 3/4.2.!
AVERAGE PLANAR LIhT.AR EEAT CENERATION i
- (..
RATE B 3/4 2-1 3/4.2.2 APRM SETPOINTS 3 3/4 2-3 3)4.2.3 MINIMUM CRITICAL POWER RATIC 3 3/4 2-3 i
^
3 3/4 2-5 3/4.2.4 LINEAR BEAT CENERATION RATE 3/4.3-INsTRtmENTATION 3/4.3.1 REA0 TOR PROTECTON SYSTEM INSTRUMENTATION 3 3/4 3-1 3/4.3.2 ISCLATION ACTUATION INSTRUMENTATION 3 3/4 3-2 1
3/4.3.3 EMERcEN Y CORE COOLING SYSTEM ACTUA'520N INSTRUMENTATION 3 3/4 3-2 3/4.3.4 REACTOR CORE ISOLATION C00LINC SYSTEM ACTUATION INSTRUMEhTATION 3 3/4 3-3 3/4.3.5 CohTROL RCD WITEDRAWAL BLOCK INSTRUMENTATION 3 3/4 3-3 3/4.3.6 MONITORING INSTRUMENTATION
-Radiation Monitoring Instrumentation 3 3/4 3-3 Seismic Monitoring Instrumentation 3 3/4 3-3 a
+
v INDEX
, BASES SECTION PAGE INSTRUMENTATION (Continued)-
Remote Shutdown Monitoring B 3/4 3-3 Instrumentation Post-Accident Monitoring-B'3/4 3-4
. Instrumentation l
Source Range Monitors B'3/4 3-4 Traversing Incore Probe System B'3/4 3-4 Chlorine Detectors B 3/4 3-4 Fire Detection Instrumentation B 3/4 3-4 i
Radioactive Liquid Effluent.
Instrumentation B 3/4 3-5 Radioactive Gaseous Effluent Instrumentation B 3/4 3-5 3/4.3.7 TURBINE OVERSPEED PROTECTION SYSTEM B 3/4 3-5.
3/4.3.8 DEGRADED STATION VOLTAGE PROTECTION i
INSTRUMENTATION B 3/4 3-5a 3/4.4 REACTOR COOLANT SYSTEM 4
3/4.4.1 RECIRCULATION SYSTEM B 3/4 4-1 i
Jet Pumps B 3/4 4-1 j
Idle Recirculation Loop Startup B 3/4 4-la l
3/4.4.2 SAFETY / RELIEF VALVES B 3/4 4-la-Low-Low Set Systems B 3/4 4-Ib i
3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems B 3/4 4-2 Operational Leakage B 3/4 4-2 3/4.4.4 CHEMISTRY B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS B 3/4 4-4 HATCH-UNIT 2 XI Amendment No. 77 I
... T.".% :..
4
I
-- o -
INDEX
[
l J
BASES 4
1 I
PAGE.
SECTION REACTOR COOLANT SYSTEM (Continued) 3/4.4.7
, MAIN STEAM LINE ISOLATION VALVES B 3/4 4 l 3/4.4.8 STRUCTURAL INTEGRITY B 3/4 4-6 i
3/4.5 EMERGENCY CORE COOLING SYSTEMS I
3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM B 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM B 3/4 5-1 3/4.5.3 LOW PRESSURE CORE COOLING SYSTEMS Core Spray System B 3/4 5-2 Low Pressure Coolant Injection System B 3/4 5-3 3/4.5.4 SUPPRESSION CHAMBER B 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT INTEGRITY Primary Containment Integrity B 3/4 6-1 Primary Containment Leakage B 3/4 6-1 LI Primary Containment Air Lock B 3/4 6-1 MSIV Leakage Control System B 3/4 6-2 Primary Containment Structural Integrity B 3/4 6-2 Primary Containment Internal Pressure B 3/4 6-2 Drywell Average Air Temperature B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS B 3/4 6-3 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES B 3/4 6-4 3/4.6.4 VACUUM RELIEF B 3/4 6-5 3/4.6.5 SECONDARY CONTAINMENT B 3/4 6-5 3/4.6.6 CONTAINMENT ATMOSPHERE CONTROL B 3/4 6-5 HATCH-UNIT 2 XII Amendment No. 77
- ,k...
..a
t INDEX E
BASES
.=
SECTION PAGE I
3/4.7' PLANT SYSTEMS 3/4.7.3 SERVICE WATER SYSTEMS B 3/4 7-1 3/4.7.2 MAIN CONTROL ROOM ENVIRONMENTAL CONTROL SYSTEM B 3/4 7-1 3/4.7.3 REACTOR CORE ISOLATION COOLING' SYSTEM B 3/4 7-1 3/4.7.4 SNUBBERS B 3/4 7-2 3/4.7.5-SEALED SOURCE CONTAMINATION B 3/4 7-3 3/4.7.6 (Deleted) 3/4.7.7 (Deleted) 3/4.7.B SETTLEMENT OF CLASS I STRUCTURES B 3/4 7-4 3/4.8 ELECTRICAL POWER SYSTEMS B 3/4 8-1 3 /4. 9 REFUELING OPERATIONS
~
3/4.9.1 REACTOR MODE SWITCH B 3/4 9-1 3/4.9.2 INSTRUMENTATION B 3/4 9-1 3/4.9.3 CONTROL RDD POSITION B 3/4 9-1
{
1 3/4.9.4 DECAY TIME B 3/4 9-1 1
l 3/4.9.5 SECONDARY CONTAINMENT B 3/4 9-1 j
3/4.9.6 COMMUNICATIONS B 3/4 9-2 l
3/4.9.7 CRANE AND HOIST OPERABILITY B 3/4 9-2 q
3/4.9.8 CRANE TRAVEL-SPENT FUEL STORAGE POOL B 3/4 9-2 i
3/4.9.9 WATER LEVEL - REACTOR VESSEL ANO 1
and WATER LEVEL - SPENT FUEL STORAGE 3/4.9.10 POOL B 3/4 9-2 3/4.9.11 CONTROL ROD REMOVAL B 3/4 9-2 3/4.9.12 REACTOR COOLANT CIRCULATION B 3/4 9-3
{
HATCH-UNIT 2 XIII
/McMmmR 9h W I
o 2.0 SAFETY-LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
=-
2.1 SAFETY LIMITS i
THERMAL' POWER (Low Pressure or Low Flow) 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
APPLICABILITY:
CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 26% of RATED THERMAL PDWER and the reactor vessel steam dame pressure less than 785 psig or core flow less thsn 10% of rated flow, be in at least HOT SHUTOOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
THERMAL POWER (Hich Pre =ssure and Hioh Flow) 1 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 for two-loop recirculation or 1.08 for single-loop recirculation operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.
APPLICABILITY: CONDITIONS 1 AND 2.
ACTION:
i With MCPR less than d.07 for two-loop recirculation or 1.08 for single-loop recirculation operation and the reactor vessel steam dome pressure o
greater than 785 psig and core flow greater than 10% of rated flow, be in at least POT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
REACTOR COOLANT SYSTEM PRESStlRE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
i APPLICADILITY: CONDITIONS 1, 2, 3 and 4.
ACTION:
l l'
With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure s 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
l HATCH-UNIT 2 2-1 Amendment No. 77
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p SAFETY LIMITS AND LIMITING' SAFETY SYSTEM SETTINGS 4
N 2.1 -~ SAFETY LIMITS (Continued)
REACTOR VESSEL WATER LEVEL' 2.1.4 The reactor' vessel water level shall be above the top of the active irradiated fuelo, APPLICABILITY:. CONDITIONS 3.4and5 ACTION:
W'ith'the reactor. vessel water level'at or below the top of.the-active.
irradiated fuel,'. manually initiate'the low pressure ECCS.to restore the reactor vessel water-level, after depressurizing the reactor vessel,
~
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. HATCH - UNIT 2, 2-4 Amendment No. 77
2.1 SAFETY LIMITS BASES i
2.0 The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated tran-sients. The fuel cladding integrity Safety Limit is set such that no 1
I fuel damage is calculated to occur if the limit is not violated.
Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07 for two-loop operation and 1.08 for single-loop operation. These limits represent a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladoing barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.
These conditions represent a significant departure from the condition intended by design for planned operation.
The evaluations which justify normal operation, abnormal transient, and accident analyses for two-loop operation are discussed in detail in Reference 3.
Evaluation for single-loop operation demonstrates that two-loop transient
{
and accident analyses are more limiting than single loop, Reference 4.
)
i 2.1.1 THERMAL POWER (Low Pressure or Low Flow) ihe use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10' lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 26 x 10' lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembl'y critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL
' POWER for reactor pressure below 785 psig is conservative.
HATCH - UNIT 2 B 2-1 Amendment No. 77
.L J.
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~ SAFETY ' LIMITS:
- 1 BASES (Continued) l 2.1.2 THERMAL. POWER-(High Pressure and High Flow)
The fuel:eladding integrity Safety Limit is set suchJ hat no fuel l
~
t damage is calculated to occur if the limit is not violated.
Since.the parameters which ras 11t in fuel damage are not directly observable dur-ing reactor: operation, the ~ thermal and hydraulic conditions resulting in 1
a departure from nucleate boiling have been used to mark the beginning of thel region where fuel damage could occur. Although it is recognized
'that a departure from nucleate boiling would not necessarily result in t
damage to BWR fuel rods, the critical power at which boiling transition is calculatedito occur has been adopted as a convenient limit.
- However, the uncertainties in monitoring the core operating state and in the procedure's used to calculate the critical power result in an uncertainty J
in the value of the critical power.
Therefore.-the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for_which more than 99.9% of the fuel rods in the core are expected to avoid boilingLtransition considering the power distribution within the core and all uncertainties.
The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB'", which is a statistical model that combines all of the uncertainties in operating parameters and'the pro-cedures used to calculate critical power.
The probability of the occur-rence'of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L), GEXL correlation.
The GEXL correlation is valid over the range of conditions used in the tests of the data used-to develop the correlation.
l I
l HATCH - UNIT 2 B 2-2 Amendment No. 77 1
mrf j
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SAFETY LIMITS I
BASES (Ccntinued) l I
2.1.2 THERMAL POWER (High Pressure and High Flow') (Continued) j.
The required' input to the statistical model are the uncertainties
. listed in Bases Table _ B 2.1.2-1, the nominal values of the core para-meters listed in Bases Table B 2.1.2-2.
The bases for the uncertainties in the core pa'rameters are given in NEDO-2034088), and the. basis for the uncertainty in the GEXL correlation 1
is given in NEDO-10958'". The power distribution is based on a' typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce.a. skewed power distribution having the greatest number of
)
assemidies at the highest power levels. The worst distribution in i
Hatch - Unit 2 during any fuel cycle would not be as severe as the dis-l tribution.used in the ar;alysis.
The method used to handle the uncertainty in the statistical analysis to determine the MCPR cladding integrity Safety Limit -
i for single-loop operation is based on. Reference 3, as described in Reference 4.
The pressure Safety Limits are arbitrarily selected to be the
. lowest transient.overpressures allowed by the applicable codes, ASME Boiler and Pressure Venel Code,Section III, and USAS Piping Code, Section B31.1.
i l
HATCH - UNIT 2 8 2-3 Amendment No. 77
7
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' UNCERTAINTIES USED IN THE DETERMINATION 4
0 0F THE FUEL CLADDING SAFETY LIMIT
- Standard Deviation Quantity
(% of Point)-
'Feedwater Flow
,1.76 l
.1 Feedwater Temperature 0.76 1
Reactor: Pressure
' O.5
.1 Core Inlet Temperature 0.2 Core Total Flow.
2.5 Channel Flow Area
'J. 0 Friction Factor Multiplier.
10.0 I
Channel Friction factor Multiplier 5.0
'TIP Reading's-8.7 R' Factor, 1.6 Critical Power 3.6
- The uncertainty analysis used to establish the core wide Safety i
Limit MCPR is based on the essenption of quadrant power symmetry for the reactor core.
t i
B 2-4 Amendment No. 21 HATCH - UNIT 2 FEB 1; 93; 1
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'l Bases Table B 2.1.2-2 j
t NOMINAL VALUES OF PARAMETERS USEO IN
'THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY SAFETY LIMIT
-THERMAL POWER 3323 MW s.
n-
. Core Flow 108.5lM1b/hr
'i, Dome Pressure 1010.4 psig P
2
. Channel Flow Area 0.1089 ft R-Factor:
High enrichment 1.043 Medium enrichment
'1.039 3
(-
Low' enrichment 1.030' p
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HATCH - UNIT 2 B 2-5
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J l
l Amendment No. 77 HATCH - UNIT 2 8 2-6
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P HATCH - UNIT 2 B 2-7 Amendment wn. 77
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1 SAFETY LIMITS i
BASEA,1 Continued)
{
2.1. 3 REACTOR COOLANT SYSTEM PRESSURE The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. However, the pressure safety limit is set high enough such that no foreseeable circumstances I
can cause the system pressure to rise to this limit. The pressure safetylimit is also selected to be the lowest transient overpressure allowed by the applicable codes. ASME Boiler and Pressure Vessel Code,Section III' and USAS Piping Code Section B 31.1.
2.1. 4 REAC'iOR VESSEL WATER LEVEL With fuel in the reactor vessel during periods when the reactor is shutdown, considerttion must be given to water level requirements due to the effect of decay heat.
If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead I
to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action.
l 1 4.
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O HATCH - UNIT 2 B 2-8 4
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2.2 LIMITING' SAFETY SYSTEM SETTINGS BASES j.
2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-l'are the values at which the reactor trips are set for each parameter. The Trip' Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits. Operation with a trip set less conservative than its Trip Setpoint, but within its specified Allowable Value, is acceptable on the basis that'each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
1.
Intermediate Rance Monitor. Neutron Flux The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The.IRM is a 5 decade 10 range instrument.. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges.
Thus, as the IRM is ranged up.to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.
The most significant source of reactivity changes during the power increase are due to control rod withdrawal.
In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed, Section 7.5 of the FSAR..The most severe case involves an initial condition in which the reactor is just suberitical and the IRM's are not yet on scale. Additional conservatism was.taken in this analysis by assuming the IRM channel closest to the rod being withdrawn is bypassed.
The results of i
this analysis show that the reactor is. shutdown and peak power is limited to 1%
of RATED THERMAL POWER, thus maintaining MCPR above the fuel cladding integrity
Safety Limit. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal ~ of control rods in sequence and provides backup protection for the APRM.
2.
Average Power Rance Monitor
~
For operation at low pressure and low' flow during STARTUP, the APRM scram setting of 15/125 divisions of full scale neutron flux provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Temperature coefficients are small and control rod patterns are constrained by the RSCS and RWM.
. HATCH - UNIT 2 B 2-9 Amendment No. 77, 1
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BASES (Continued)
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i REACTOR PR07FCTION SYSTEM.INSTRUMENTATIJ.N SETPOINTS (Continued) i Averaoe Power'Rance Monito (Continue @/
Of all the possible sources of reactivg#ficant power increase.
y input, uniform control rod with-drawal is the most probable ca4se offs Because j
the fict. distribution associetid Wit.h uniform rod withdrawals does not power by a significart uw}it[ the R.ite of power rise is very slow. involve high lo i
Gen-erally the heat flux is in Aer/ egoilibrium with the fission rate.
In an assumed eniform rod withdrh ty approhch to the trip level, the rate of 3
power rise is not more thm 5% of RATED THfRMAL POWER per minute and the APRM system would be n< ore than adequate tdassure shutdown before the power could exceed the S'afety Limit.
Thh 15% neutron flux trip remains active until the mody 5% tch is plac;ed in the Run position.
The APRM flux scram (1 inthbRunmodeconsistsnfaflowreferenced simulated ',hermal power scrs setpoint and a fixed neutron flux scram set-point.
Tae APRM flow referenced neutron flux signal is passed through a filtering network with a time constaat which is representative of the fuel dynamics. This prevides a flow,refarm ced signal that approximates the average heat flux or thermal power -thad.is developed in the core during transient or steady-state sonditionr. The flow independent portion of this scram setpoint must be adjustec\\ downward during single-loop operation to account for. lower core fliw with rupect to twonloop operation with the same drive flow.
The APRM flow referenced simuln d' thermal power scram trip setting for two-loop anm single-loop operation is adjustable up to 113.5% of RATED THERMAL p0WER. This reduced flow referenced trip setprint will result in an earlier scram during slow thermal transients, such as the loss of 100*F feedwater heating event, than would result with the 118% fixed neutron flux scram trip. The lower flow referenced scram utpoint therefore decreases the severity, ACPR, of a slow thermal transient and allows lower operating limits,
if such a transient is the limiting abnormal operational transient during a certain exposure interval in the fuel cycle.
The APRM fixed neutron flux signal does net incorporate the time constant',
but responds directly to instantaneous neutron flux.
This scram setpoint l
scrams the reactor during fast power increase transients if credit is not I
taken for a direct (position) scram, and also serves to scram the reactor if credit is not taken for the flow referenced simulated thermal power scram.
The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown.
1 HATCH-UNIT 2 B 2-10 Amendment No. 77
.a
1 LIMITING SAFETY SYSTEM SETTING BASES (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Turbine Control Valve Fast Closure', Trip 011 Pressure-Low (Continued) pressure switches whose contacts form the one out-of-two-twice logic input to the Reactor Protection System. This trip setting, a nominally 50% greater closure time and a different valve characteristic from that of the turbine ~stop valve, combine to produce transients very similar to that for the stop valve.
No significant change in MCPR occurs.
Relevant transient analyses are discussed in Section _15 of the Final Safety Analysis Report. This scram is bypassed when turbine steam flow is below that corresponding to 30% of RATED THERMAL POWER, as measured by turbine first stage pressure.
II.
Reactor Mode Switch In Shutdown Position The reactor mode switch Shutdown position trip is a redundant channel tc the automatic protective instrumentation channels and provides additional manual reactor trip capability.
-12.
Manual Scram The Manual Scram is a redundant cFannel to tLe automatic protective instrumentation channels and provides manual reactor trip capability.
2.3 REFERENCES
1.
" General Electric BWR Thermal Analysis Bases (GETAB) Data, Correlation and Design Application," NEDE-10958-P-A and NE00-10958-A, January 1977, 2.
General Electric " Process Computer Performance Evaluation Accuracy,"
NEDO-20340 and Amendment 1, NE00-20340-1, June 1974 and December 1974',
respectively.
3.
" General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDO-24011-P-A.
4.
"Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"
NE00-24205, August 1979.
i HATCH-UNIT 2 B 2-13 Amendment No'. 77
4-3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION
\\
I 3.2.1 ALL AVERAGE PLANAR LiWEAR HEAT GENERATION RATES (APLHGRs) shall be' equal to or less than the applicable APLHGR limit, which is a function of fuel type and AVERAGE PLANAR EXPOSURE. The APLHGR limit is given by the applicable rated power, rated-flow limit taken from Figures 3.2.1-1 through 3.2.1-9, multiplied by. the. smaller of either:
a.
The factor given by Figure 3.2.1-10, or b.
The factor given by Figure 3.2.1-l'1.
APPLICABILITY: CONDITION 1, when THERMAL POWER 2 25% of RATED THERMAL POWER.
ACTION:
With an APLHGR exceeding the applicable limits, initiate corrective action within l 15 minutes and continue corrective action so that the APLHGR meets 3.2.1 l
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 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 />.
l l
SURVEILLANCE REQUIREMENTS 4.2.1 All /.PLHGRs shall be verified to be equal to or less than the applicable limit:
l l
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 per 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 CONTROL ROD PATTERN for APLHGR.
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HATCH - UNIT 2 3/4 2-1 Amendment No. 77
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. POWER OISTRIBUTION' LIMITS-3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2;3 ALL MINIMUM CRITICAL POWER RATIOS (MCPRs) for two loop operation,.
l 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 3.2.3-5.
For. Power 2 30%,
l the-OLMCPR is the greater of either:
a.
The applicable limit determined from Figure 3.2.3-4, or b.
The appropriate K given by Figure 3.2.3-5, multiplied by the appropriate limit from Figure 3.2.3-1, 3.2.3-2, or 3.2.3-3, where:
t = 0 or ave - TB, whichever is greater, T
, IA ~ *B.
A = 1.096 sec (Specification'3.1.3.3 scram time Ifmit to t
notch 36),
~N
~ 1/2 g = 0.834 + 1.65 3
(0.059),
t nbNi 1=1 In
{N t -
= i=1 t,y, n
[N 4
.i = 1 n=
number of surveillance tests performed to date in cycle, th Ng = number of active control rods measured in the i surveillance
- test, g = average scram time to notch 36 of all rods measured in the t
th i
surveillance test, and N3 = total number of active rods measured in 4.1.3.2.a.
For single-loop operation, the MCPR limit is increased by 0.01 over the comparable two-loop value.
I' APPLICABILITY: CONDITION 1, when THERMAL POWER 2 25% RATED THERMAL POWER HATCH - UNIT 2 3/4 2-6 Amendment No. 77 l
3/4.2.3 MINIMUM CRITICAL p0WER RATIO (CONTINUED)
SURVEILLANCE REOUIREMENTS ACTION:
I With MCPR less than the applicabie limit determined from Specification 3.2.3.a or 3.2.3.b for two-loop or single-loop operation, initiate corrective l
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 />.
4.2.3 The MCPR limit at rated flow and rated power'shall be determined for each type of fuel-(8X8R, PBX 8R, and 7X7) from Figures 3.2.3-1, 3.2.3-2, and-3.2.3-3 using:
= 1.0 prior to the initial scram time measurements for the cycle a.
t performed in accordance with Specification 4.1.3.2.a. or as defined in Specification 3.2.3; the determination of the limit-b.
t 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 per 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 CONTROL ROD PATTERN for MCPR.
1 i
I l
HATCH - UNIT 2 3/4 2-7 Amendment No. 77 e.
a
. u i;
TABLE -3. 3.3 1 (Centinued)
CONTROL ROD VITCRAk'AL BLOCK INSTRUMENTATION NO I a.
k* hen THIRMAL Pok'IR exceeds' the press: power level of the RWM and RSCS r.nd eben the limiting conditi'en defined in section 3.1.4.3 exists.
b.
This function is bypassed if detecter is reading >.100 eps c the IRM channels are en range 3 er higher.
s
-c.
~his function is bypassed when the associated IPf. channels are en -
range 8 c: higher.
d.
A tetal of 6 IRM inst:tments must be OPER.ABLI.
.~his functics is bypassed when the IRM channels' are en range 1.
e.
f.
k*ith any centrol red withdrawn. Nc: applicable te centrol reds
.re=cved par.Specifica:icn 3.9.11.1 c 3.9.11.2.
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o 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS i
LIMITING CONDITION FOR OPERATION 3.4.1.1 At.least one recirculation loop of the reactor coolant system shall be in operation with its recirculation punp operating and the associated pump discharge valves OPERABLE, and a.
With only one recirculation loop in operation, the Functional Units
~
2.b of Table 2.2.1-1 and 1.a of Table 3.3.5-2, the limits on APLHGR in Section 3/4.2.1 and MCPR in'Section 3/4.2.3 shall be in effect.
b.
With only one recirculation loop in operation, the limit specified in Figure 3.4.1.1-1 shall be in effect.
)
APPLICABILITY:
CONDITIONS 1* and 2*.
ACTION:
a.
With no recirculation loops in operation, place the rector mode switch in the HOT SHUTDOWN position.
b.
With requirements of Specification 3.4.1.1.a not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the removal of one recirculation loop from service, place the unit in the HOT SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and in COLD SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
c.
With only one recirculation loop in operation and the Unit in the Operation Not Allowed. Region specified in Figure 3.4.1.1-1, initiate action within 15 minutes to place the Unit in the Operation Allowed Region in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Otherwise, place the reactor in the HOT SHUTDOWN condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
1 SURVEILLANCE REQUIREMENTS 4.4.1.1.1 Each pump discharge valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel:
a.
Each startup** prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER, and b.
During each COLD SHUTDOWN which exceeds 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.**
- See Special Test Exception 3.10.4.
- If not performed within the previous 31 days.
HATCH - UNIT 2 3/4 4-1 Amendment No. 77 j
7.
l3/4.4' REACTOR COOLANT SYSTEM =
s.
3/4.4;1' RECIRCULATION SYSTEM '
' RECIRCULATION - LOOPS r
o SURVEILLANCE REQUIREMENTS - 4.4.1.1.2 ' With only one recirculation loop in operation,-verify that the :
reactor operating conditions are outside the Operation Not Allowed Region in
-Figure 3.4.1.1-1:
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 changed by at least 5% of RATED
-THERMAL POWER and steady-state conditions-have' been reached.
.)
i i
~ HATCH - UNIT 2 3/4 4-la Amendment No. 77
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REACTOR COOLANT' SYSTEM
d
? JET p0MPS LIMITING CONDITION FOR OPERATION ~
i 3.4.1.2 'All jet pumps corresponding to-the operating loop (s)'shall be
' OPERABLE.
APPLICABILITY:
CONDITIONS.1 and 2.-
ACTION:
All jet pumps corresponding to the operating loop (s) shall be l
.a.
OPERABLE with at least one of the following requirements:
1.
For any specific flow condition, each individual' jet pump flow shall not differ by more than.10% of the average' loop jet pump
)
flow from the normal range" of average loop jet pump flows experienced for those flow conditions, or 2.
For any. specific core flow condition, each individual jet pump diffuser to' lower plenum differential pressure (D/P) shall not.
1 differ by more than 20% of the average loop D/P.from.the normal l
range
- of average loop jet pump D/Ps experienced for those flow 1
conditioris.
b.
With one or more jet pumps exceeding the above requirements, 1
evaluate the reason for the deviation, and be.in HOT SHUTOOWN within i
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the circumstances that one or more jet pumps are verified to be inoperable.
SURVEILLANCE REQUIREMENTS l
4.4.1.2 ' Each of the above required jet pumps shall be demonstrated OPERABLE" prior to THERMAL POWER' exceeding 25% of RATED THERMAL POWER; following racirculation restarts; following any unexpected or unexplained change in core flow, jet pump loop flow, recirculation pump flow, or core plate
-l
- differential pressure; and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by recording jet pump loop flows, recirculation pump flows, recirculation pump speeds, and 1
individual jet pump flows (D/P) and verifying-that neither of-the following
)
-conditions occur:
l q
- Normal expected operating range is based on data obtained from operating experience.
1 HATCH - UNIT 2 3/4 4-2
^**"d"*"t N ' 77 l
m.m
.t o y
SURVEILLANCE ' REQUIREMENTS
)
.a. 'The'recircula' tion pump flow / speed ratio-deviates more than 5% from
.the normal range," or '
b.. The jet-pump loop flow / speed ratio deviates more than 5% from the normal. range.*'
If any require' jet pump fails to meet either or both of the above j
d Surveillance; Requirements, review the ' jet-pump operability as defined in the i
ACTION. statement for Section 3.4.1.2 and in BASES Section 3/4.4.1.2.
l 1
l l
Normal expected operating range is based on data obtained from operating experience.
)
HATCH - UNIT 2 3/4 4-2a Amendment No. 77
s:
/
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(f h
SPECIAL-TEST EXCEPTIONS H
3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS 1
LIMITING CONDITION FOR'0PERATION J
- l 3.1'0.3 The provisions of' Specification 3.9.1,' Specification 3.9.3_ _.
and Table 1.2 may be suspended to pennit the-reactor. mode switch -to be -
locked in the Startup position and to allow more than one control rod to L
l be withdrawn for the shutdown margin demonstration, provided that at least-t, the.following requirements'are satisfied.
j I.
1.-
The source range monitors are OPERABLE with the RPS circuitry shorting links removed per Specification. 3.9.2, 2.-
The rod worth minimizer is OPERABLE per Specification 3.1.4.1 i
and is; programed for the ' shutdown margin demonstration, or -
conformance with the shutdown margin demonstration procedure is verified by a second 1.icensed operator'or other qualified
~
member of the technical staff.
3.
The " rod-out-notch-override" control shall not be used during i
movement of the control rods, and
- 4..
No other CORE ALTERATIONS are in progress.
I APPLICABILITY:
CONDITION 5, during shutdown margin. demonstrations.
c.:
ACTION:
,J With _the requirements of the above specification not satisfied, imediately
~ i [q restore the reactor mode switch to the Refuel position.
SURVEILLANCE REQUIREMENTS
- e
,(
4.10.3 Within'30 minutes prior to the performance of a shutdown margin demonstration, verify that; a
~
a.
The source range monitors are OPERABLE per Specification 3.9.2, b.
The rod worth minimizer is OPERABLE with the required program per Specification 3.1.4.1, or a second licensed operator is present to verify compliance with shutdown demonstration procedures, and c.
No other CORE ALTERATIONS are in progress.
s
~
SPECIAL TEST' EXCEPTION' RECIRCULATION LOOPS 4
' LIMITING CONDITION FOR OPERATION.
3.10.4I The requirements of Specification 3.4.'1.1 that 'a recirculation loop (s) be in operation may be suspended for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during the performance of the Startup Test Program and PHYSICS TESTS.
APPLICABILITYi CONDITIONS I and 2.
ACTION:
With the above~specified time limit exceeded, actuate the manual sc ram.'
SURVEILLANCE REOUIREMENTS 4.10.4 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at least once per hour during the Startup Test Program and PHYSICS TESTS.
f 4
HATCH - UNIT 2 3/4 10 4 Amendment No. 77
^*
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3/4.1 REACTIVITY' CONTROL SYSTEMS BASES-3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that'(1) the reactor can be made suberitical from all operating conditions (2) the reactivity transients associated with postulated accident conditions are controllable within j
i acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
Since core reactivity. values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold xenon-free condition and shall show the core to be subcritical by at least R + 0.28% aK or R + 0.38%AK, as appropriate.
The value of R in units of %AK is the difference between the calcu1 ted A
value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity. The value of R must be i
positi9e or zero and must be d&tarmined for each fuel loading cycle.
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Satisfaction of this limitation can be best demonstrated at the time of j
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fuel loading but the margin must be determined anytime a control rod is j
incapable of insertion. This reactivity characteristic has been a basic assumption in the analysis of plant performance, j
Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTOOWN MARGIN. The highest worth rod may be determined analytically or by test. The SHUTDOWN MARGIN is demonstrated by an insequence control rod withdrawal at the beginning of life fuel cycle and, if necessary, at q
any future time in the cycle if the first demonstration indicates that the j
margin could be reduced as a function of exposure. Observation of subcriticality in this condition assures subcriticality with the most j
reactive control rod fully withdrawn.
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3 /4.1. 2 REACTIVITY ANOMALIES Since the SHUTOOWN MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditions is necessary, and the changes in reactivity can be inferred from these. comparisons of rod patterns. Since the comparisons are easily done, frequent checks are 1
not an impositdon on normal operations.
.A 1% change is larger than is i
expected for normal operation so a change of this magnitude should be thoroughly evaluated. A change as large as 1% would not exceed the design conditions of the reactor and is on the safe side of the postulated 1
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HATCH - UNIT 2 B 3/4 1-1
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 CONTROL R005 k
The specifications of this section ensure that (1) the minimum SHUTOOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) the potential effects of the rod drop accident are limited. The ACTION statements permit variations f rom the basic requirements, but at tt.e same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set j
i such that the resultant effect on total' rod worth and scram shape will be i
kept to a minimum.
The requirements for the various sceam time measurements l
4 ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
Damage within :the control rod drive mechanism could be a generic problem; therefore, with & control rod immovable because of excessive
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friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.
Control rods that are inoperable for other reasons are permitted to be i
i taken out of service provided that those in the nonfully-inserted position
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are consistent with the SHUTDOWN MARGIN requirements.
j The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods coulci be indicative of a generic problem end the reactor 1
must be shutdown for investigation and resolution of the problem.
The control rod system is analyzed to bring the reactor suberitical at a
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rate f ast enough to prevent the MCPR f rom becoming less than the fuel cladding integrity MCPR Safety Limit during the limiting power transient analyzed in Section 15 of the FSAR.
This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as j
given in the specifications provide the required protection and MCPR remains i
greater than the fuel cladding integrity MCPR Safety Limit. The occurrence of l
scram times longer than,those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of
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time with a potentially serious problem.
l Control rods with inoperable accumulators are declared inoperable and 1
Specification 3.1.3.1 then applies.
This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram HATCH - UNIT 2 B 3/4 1-2 Amendment No. 77 i
i 3/4.2 POWER DISTRIBUTION LIMITS BASES i
The specifications of this section assure that the peak cladding temperature 'following the postulated design basis. loss-of-coolant accident will not exceed the 2200'F limit specified in the Final Acceptance Criteria
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(FAC) issued in. lune 1971 considering the postulated effects of fuel. pellet densification. These specifications also assure that fuel design margins are maintained during abnormal transients.
2/4.2.1 AVERAGE PLA.NAR LINEAR HEAT GENERATION RATE i
This specification. assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10.CFR 50, Appendix K..
The peak claffding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all-the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming an LHGR.for the highest powered rod
.which is equal to or less' than the design LHGR corrected for densification.
This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor.
The Technical Specification APLHGR is this LHGR of the highest powered rod.
divided by its local peaking factor. The limiting value for APLHGR is shown in the figures for in Technical Specification 3/4.2.1. For single-loop operation, Reference 1 requires a 0.75 multiplication factor to 8X8R and P8X8R bundles.
The calculational'orocedure used to establish the APLHGR shown in the figures in Technical Specification 3/4.2.1 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.
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A flow dependent correction factor incorporated ir.to Figure 3.2.1-10 is applied to the rated conditions APLHR to assure that the 2200'F PCT limit is complied with during a LOCA initiated from less than rated core flow.
In addition, other power and flow dependent corrections given in Figures 3.2.1-10 and 3.2.1-11 are applied to the rated conditions to assure that the fuel thermal-mechanical design criteria are preserved during abnormal transients initiated from off-rated conditions.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in bases Table B 3.2.1-1.
Further discussion of the APLHGR limits is given in Reference 2.
l HATCH - UNIT 2 0 3/4 2-1 Amendment No. 77
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POWER DISTRIBUTION LIMITS BASES 3/4.2.2 ' APRM SETPOINTS This section deleted.
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l 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the establi;hed fuel cladding integrity Sefety Limit MCPR of 1.07 for two-loop operation and 1.08 for single-loop operation, and an analysis of abnormal operational transients as described in References 1 and 3.
Fo; any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting as given in Specification 2.2.1.
To assure tnat the fuel cladding integrity Safety Limits are not exceeded during any anticipated abnonnal operational transient, the most limiting transients have been analyzed to determine which results in the largest reduction in CRITICAL POWER RATIO (CPR were loss of flow, increase in pressure). The type of transients evaluated arid power, positive reactivity insertion, and coolant temperature decrease.
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HATCH - UNIT 2 B 3/4 2-3 Amendment No. 77
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1 POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)
.I The purpose of the MCPR, and the Kp of Figures 3.2.3-4 and.3.2.3-5, f
respectively is to define operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power, the required MCPR is the larger value of the MCPRf and MCPRp at the existing core flow and power state.
The MCPRfs are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.
The MCPR s were calculated such that for the maximum core flow rate and f
the corresponding THERMAL POWER along ~the 105% of rated steam flow control line, the limiting bundle's relative power was adjusted until the MCFR was slightly above the Safety Limit. Using this relative bundle power. the MCPRs were calculated at different points along the 105% of rated steam flow control line corresponding to dif ferent core flows. The calculated MCPR at a given point of core flow is defined as MCPR.t The core power dependent MCPR operating limit, MCPR, is the rated power p
and rated flow MCPR operating limit multiplied by the Kp factor given in Figure 3.2.3-5.
The K s are established to protect the core from transients other than core p
flow increases, including the localized event such as rod withdrawal error. The K s were determined beJed upon the most limiting transient at the given core power plevei. For further information on MCPR operating limits for off-rated conditions.
See RLfervnce NEDC-30474-P.(2) l l
I HATCH - UNIT 2 8 3/4 2-4 Amendment No. 77
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EASES MINIM"M CRITICAL 20kTR RATIO (Continued) f At TEERMAL ?OWER levels less than or equal to 25% of RATID TFERMAL
?diTR, the reacter vill be operating at minimum recirculatien pu=p speed a the modera:c: void centent will be very.small.
red patterns which may be e= ployed at For all designated centrol this point, operating plant experience ind.icated tha: the. resulting MCPR value. is in excess of requirements by a censiderable margin. 'Vith this low void' content, increase would only place operation in a mere conserva:inadver ent any core flow MCPR.
Du;ing inirial star p testing of the plant, Ave mode rela:dve to be made a:
The MCPR margin vill thus be.dsmonstrated such thatO!%
speed.
evaluation below this pever level vill be shown to be unnecessa.y.
future MCPR daily.requi:ement The for calculating MCPR above 25% of RATED TEERMAL P0kIR is suffit:d e.ht been significant. power er centrol red changes.since power. distribution The requirement for that MCPR vill be k:::own following a change in THIR".A regardless cf magnitude that could place operatics at a thermal li=it.
cwer shape, 1
3/4.2.4 ' LINEAR REAT GENERAT70N RATE The DiGR specification assures that the linear heat generatica dans..icatica is postulated.any *od is less than the design linear heat generation ev a:e <-
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EATCH-2
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Amendment No. 33,39 l
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POWER DISTRIBUTION LIMITS R&SES o
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References:
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' General Electric Standard Application for Reactor Fuel (Supplement
.for United States)," NEDE-24011-P-A.
2.
- Average Power Range Monitor, Rod Block Monitor and Technical i
Specification Improvement ( ARTS) Program for Edwin I. Hatch Nuclear Plant, Units 1 and 2," NEDC-30474-P December 1983.
i 3.
"Edwin I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,"
i NED0-24205, August 1979.
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HATCH - UNIT 2 8 3/4 2-6 Amendment No. 77
3 /4. 4 REACTOR COOLANT SYSTEM I '
EASES 3 /4. 4.1.1 RECIRCULATION SYSTEM Operation with a reactor coolant recirculation loop inoperable is allowed, provided that adjustments to the flow referenced scram and APRM rod block setpoints, MCPR cladding integrity Safety Limit, MCPR Operating Limit, and MAPlHGR limit are made. An evaluation of the performance of the LCCS with
. single-loop operation has been performed and determined to be acceptable, Reference 1.
The neximum uncovered time results in a reduction factor to the MAPLHGR limit of 0.75.
To account for increased uncertainties in the total core flow and TIP readings when operating with a single recirculation pump, a 0.01 increase is applied to the MCPR cladding integrity Safety Limit and MCPR Operating Limit over the comparable two-loop values. The flow' referenced simulated thermal power setpoint for single-loop operation is reduced by the amount of mow, where m is the flow reference slope for the rod block monitor and oW is the largest difference between two-loop and single-loop effective drive flow when the active loop indicated flow is the same. This adjustment is necessary to preserve the original re'1ationship between the scram trip and c actual drhe flow.
The possibility of experiencing limit cycle oscillations during single-loop operation is precluded by restricting the core flow to greater than or equal to 45% of rated when core thermal power is greater than the 80%
rod line. This-requirement is based on Genersl Electric's recommendations contained in SIL-380, Revision 1, which d6 fines the region where the limit cycle. oscillations are more likely to occur.
3/4.4.1.2 JET PUMPS l
An inoperable jet pump is not, in itst1f, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design basis Loss-of-Coolant Accident by increasing the blowdown l
i area and eliminating the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable is necessary.
l One of the acceptable procedures for jet pump failure surveillance identified in NUREG/CR-3052, Reference 2, was adopted for Hatch Unit 2.
The surveillance is performed to verify that neither of the following conditions
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occur:
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(a) The recirculation pump flow / speed ratio deviates by more than 5%
I f rom the normal range, or I
-(b) The jet pump loop flow / speed ratio deviates by more than 5% f rom the normal range.
If either criterion is failed, then the procedure calls for comparing either the individual jet pump flows or the individual jet pump diffuser to i
lower plenum differential pressures to the criteria of the Limiting Conditions for Operation (LCO).
If the LCO is not satisfied and pump speed is less than HATCH - UNIT 2 83/44-1 Amendment No. 77 1
BASES 60% rated, it may be nececsary to increase pump speed to above 60% of rated
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and to repeat the measurements before declaring a jet pump inoperable.
In this case, it is recommended that close monitoring and increasing recirculation pump speed be performed only,1f the criteria are exceeded by an amount to be determined f rom previous plant operating experience.
3/4.a.1.3 I.q(E RECIRCULATION LOOP STARTUP When restarting an idle pump, the discharge valve of the idle loop is required to remain closed until the speed of the faster pump is below 50% of its rated speed to provide assurance that when going from one-to two-loop operations, excessive vibration of the jet pump risers will not occur.
In order to prevent undue stress on the vessel nozzles and bottom head region the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop. Since the coolant in the l
bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature dif ference were greater than 145'F.
The loop temperature must be within 50'F of the reactor pressure vessel coolant temperature to prevent thernal shock to the recirculation pump and recirculation nozzles.
3/4.4.2.1 SAFETY / RELIEF VALVES The reactor coolant system safety valve function of the saftty-relief valves operate to prevent the system from being pressurized above the Safety Limit of 1325 psig.
The system is designed to meet the requirements of the ASME Boiler and Pressure Vessel Code,Section III, for the pressure vessel, and ANSI B31.1. 1975 Code, for the reactor coolant system piping.
The capacity of the safety-relief valves is based on the full MSIV closure transient with failed trip scram, position switches, as described in Supplement 5.A of the FSAR, Section 5.A.6.
Demonstration of the safety-relief valve lift settings will occur i
only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure vessel Code.
j HATCH - UNIT 2 8 3/4 4-la Amendment No. 77 4
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SASES i
3 /4. 4. 2. 2 LOW-LOW SET SYSTEM l
on four preselected safety / relief valves.(S/RVs).The low-lo setpoints 'after a.ty S/RV has opened at its normal steam pilot setpoint The LLS system lowers the
_ concurrent high reactor vessel steam dome pressure scram signal is presen ment and thrust loads on the SRV discharge line. purpose of t The amount of reactor depressurization during an S/RV blowdown because t LLS setpoints keep the four selected LLS S/RVs open for a longer time.
high reactor vessel steam dome pressure signal for the LLS logic is prov The by the exclusive analog trip channels.
high pressure scram functions.icated steam dome pressure chann l
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i HATCH - UNIT 2 i
B 3/4 4-lb Amendment No. 77 l
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BASES
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3/4.4.9' REFERENCES 1.
'Edwin I. Hatch Nuclear Plant Units 1 and 2_ Single-Lood Operation,'
NE00-24205, August 1979.
i 2.
NUREG/CR-3052, "Closecut of IE BULLETIN B0-07: BWR Jet Pump Assembly Failure,' Published November 1984.
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HATCH - UNIT 2 8 3/4 4-7 Amendment No. 77 i