ML20069J310
| ML20069J310 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/07/1994 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20069J293 | List: |
| References | |
| NUDOCS 9406140308 | |
| Download: ML20069J310 (22) | |
Text
{{#Wiki_filter: (2) Pursuant to the Act and 10 CFR Part 70, Georgia Power Company to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the final Safety Analysis Report, as suppicmented and amendeo; (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70 Georgia Power Company to receive, possess and use at any time any byproduct, , source and special nuclear material as sealed neutron sources for reactor startuo, sealed sources for reactor instrumentation and radiation monitoring equipment calibratien, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CPR Parts 30, 40 and 70, Georgia Power Cemeany to receive, possess and use in a.ounts as recuired any byproduct, source or special nuclear material without restriction to cnemical or physical e fom, for sarple analysis or inst ument t calibration or associated with radioactive ' 2 apparatus or components; g (5) Pursuant to the Act and 10 CFR Parts 30 and 70, e Georgia Power Company to possess, but not separate, such byprocuct and special nuclear i inaterials as may be procuced by the operation [ of the facility. sy a C. This license shall be deemed to contain and is subject to the s} conditions specified in the follov.ing Comission regulations in J 10 CPR Chapter 1: Part 20, Secticn 30.34 of Part 30, Section 40.41 f Part 40, Sections 50-54 and 50-59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, ano orcers of the Comission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1) Maximum Pcwer t.evel The Georgia Pcwer Company is authorized to operate the facility at steady state reactor core power levels not in excess of 2435 megawatts themal, a^c e pt -the SciU+ 4 m a.1 b* *PerateA at3+* 4 , s een y ,-ermine swa. eisLn Jetr+
* < i~ dew o f ::.s?is meys wstU +%er m s.a o s et go Jsty$ fu puer oper+L a.,
(2) T chnica Scec cations te2t purgis e s. , I The Technical Specifications contained in Appendices A ano B, as revised through Amenoment No. /55 are hereby incorporated in the license. The licensee shall operate the facility in accordance witn the Technical Specifications. i 9406140300 940607 1 PDR ADOCK 0500 P l 1
l I BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.1 FUEL CLADDING INTEGRITY The abnormal cperational transients applicable to operation of the HNP-1 Unit have been analyzed throughout the spectrum of planned operating conditions. The analy:,es were based upon plant operation in accordance with the operating ; map given in Figure 3-1 of Ref. 3. In addition, 2436 MWt* is the licensed l l maximum power level of HNP-1, and this represents the maximum steady-state i 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 reference. As discussed in Reference 2, the core-wide transient analyses for single-loop operation are 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 perfaltted, except during startup testing. The analysis to support operation at various
- Except the unit may be operated at steady-state power levels not in excess of 2558 MWt for a cumulative duration of 30 days for power uprate test purposes.
HAICH - UNIT 1 1.1-10 techsp\h\94-06Ulf. pro \l41
SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 1.2. REACTOR COOLANT SYSTEM INTEGRITY 2.2. REACTOR COOLANT SYSTEM INTEGRITY Aoolicability Acolicability The Safety Limit, established to pre- The Limiting Safety System Settings apply serve the reactor coolant system to trip settings of the instruments and integrity, applies to the limit on the devices which are provided to prevent the reactor vessel steam dome pressure, reactor vessel steam dome pressure Safety Limit from being exceeded. Ob.iective Obiective TheobjectiveoftheSafetyLimit(asso- The objective of the Limiting Safety Sys-ciated with preserving the reactor cool- tem Settings is to define the level of ant system integrity) is to establish a the process variables at which automatic pressure limit below which the integrity protective action is initiated to prevent of the reactor coolant system is not the reactor vessel steam dome pressure threatened due to any overpressure con- Safety Limit from being exceeded. dition. Specifications Specifications A. Reactor Vessel Steam Dome Pressure A. Nuclear System Pressure
- 1. When Irradiated fuel is in the 1. When Irradiated Fuel is in the Reactor Reactor The reactor vessel steam dome pres- When irradiated fuel is present in sure shall not exceed 1325 psig at the reactor vessel, and the head any time when irradiated fuel is is bolted to the vessel, the present in the reactor vessel. limiting safety system settings shall be as specified below:
Limiting Safety Protective System Settings Action (osio) ,
- a. Scram on high s 1054* l l reactor pres-sure(reactor vessel steam domepressure)
- b. Nuclear system 4 valves at relief valves 1080 open on 4 valves at nuclear system 1090 pressure 3 valves at 1100 cThe unit may operate with a setpoint s 1065 psig for a cumulative duration of 35 days for power uprate test purposes.
HATCH - UNIT 1 1.2-1 techsp\h\94-060lG. pro \l03 l l 1
i I: Table 3.1-1 20 -4 Q REACTOR PROTECTION SYSTEM (RPS)(NSTRUMENTATION REQUIREMENTS Q When the reactor is subcritical and the reactor water temperature is less than 212*F, only the following - sources of scram trip signals need to be operable: H ~ Mode Switch in SHUTDOWN Manual Scram IRM High High Flux Scram Discharge Volume High High Level Scram Operable Number Source of Scram Tnp Signal Channels Scram Trip Setting Source of Scram Signalis (a) Required Per Required to be Operable Trip System Except as indicated Below (b) 1 Mode Switch in SHUTDOWN 1 Mode Switch in Automatically bypassed two SHUTDOWN seconds after the Mode Switch is placed in the w SHUTDOWN position. w y 2 Manual Scram 1 Depression of Manual Scram Button 3 1RM High High Flux 3 5120/12S of full IRMs are automaticatty bypassed scale Tech Spec when APRMs are on scale and the 2.1. A.1.a. Mode Switch is in the Run position. Inoperative 3 Not Applicable IRMs are automaticatty bypassed when APRMs are on scale and the rp Mode Switch is in the RUN Sr Position. en V y 4 Reactor Vessel Steem Dome 2 51054 psig' Not required when reactor head l $ Pressure - High Tech Spec 2.2.A.1. is not bolted to the vessel. ? = m h
- The unit may operate with a setpoint s 1065 psig for a cumulative duration of 35 days for power upreta test purposes.
l s w
Z Tabie 3.2-7 (Continued) 3D H n Z Ref. Instrument Trip Required Trip Setting Remarks
' No. Operabie Condition
@ tal Nomenclature Channels
- per Tnp System w
- 3. APRM Downscale 2(b)(e) 23/125 of full scale Not required while performing low power physics test at atmospheric pressure during or after refueimg at power levels not to exceed 5 MWt.
12% Flux 2(b)(e) s12/125 of full scale Th;s function is bypassed when the Mode Switch is placed in the RUN position. Upscale 2(b)(e) 50.58 W + 50% - 0.58 AW* See Specification 2.1.A.1.c(1) for l definitions of W and AW. Trip level setting is in percent of rated power. Not required whde f performing low power physics tests y w at atmospheric pressure during or after refueling at power levels not
- to exceed 5 MWt.
4 RBM Inoperative 1(e)(f) Not applicable Inoperative trip produced by switch not in operate, circuit boards not in circuit, fails to null, less than required number of LPRM inputs for rod selected. r+ ro n Downscale 1(el(f) 294/125 of full scale w to V / 7 / f O
- The urst may operate with a setpoint of s 0.58W + 53% -0.58AW for a cumulative duration of 35 days for power uprate test purposes.
l m C c w e 9 O / m.e M w
Z M n Table 3.2-14 m INSTRUMENTATION WHICH ARMS LOW LOW SET S/RV SYSTEM C 2
- Required Operable
~ Trip Channets Ref Condition per Trip No.I*I instrument Nomenclature System Trio Settina Remarks
- 1. Reactor Vessel Steam Dome High 2" s1054 psig' l Pressure
- 2. Relief / Safety Valve High 2/ valve 85. + 15. -5 The I miting condition Tailpipe Pressure psig of operation of these switches is provided in Specification 3.6.H.1.
w N I N w o. Notes for Table 3.2-14 l The unit may operate with a setpoint of s 1065 psig for a cumulative duration of 35 days for power uprate test purposes. l rv n a. The column entitled *Ref. No." is only for convenience so that a one-to-one relationship can be established between y in items in table 3.2-14 and items in table 4.2-14. t
/
,7 . b.1. With the requirements for the minimum nurnber of OPERABLE channels not satisfied for one trip system, place the to inoperable channelin the tripped condition or declare the associated system inoperable within one hour. With i
C the requirements for the minimum number of OPERABLE channels not satisfied for both trip systems, declare the
, associated system inoperable within one hour.
C [ b.2. One instrument channel may be inoperable for up to 6 hours to perform required surveillances prior to entering g other applicable actions. i O
/
, w l u, I l l t _ . _ _ _ _ _
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS l 4.6.H.l. Relief / Safety Valves (Continued)
- e. Operability of Tail Pioe Pressure Switches
- 1. Functional Test:
- b. At each ,cheduled out-age greater than 72 hours during which en-try is made into the primary containment, if not performed with-in the previous 92 days.
- 2. Calibration and verifying the setpoint to be 85, +15
-5 psig at least once per ,
18 nunths. 3.6.H.2. Relief / Safety Valves low Low Sfttfunction 4.6.H.2. Relief / Safety Valves low Low Set Function During power operation startup, and hot standby, the relief The low low set relief valve func-valve function and the low low tion and the low low set function set function of the following pressure actuation instrm enta-reactor coolant system safety / tion shall be demonstrated OPERABLE *** by performance of a: relief valves shall be OPERABLE with the following low low set function lift settings: a. CHW1EL FLETIONAL TEST, including calibration of tre Low Low Set AllowableValue(psig)* trip unit and the dedicated Valve Function Open high steam dome pressure [Jgg channels **, at least once Low s 1005 per quarter. Medim 5 857 5 1020 $ 872 Media High b. CH/#1EL CALIBRATION, logic s 1035 5 887 High $ 1045 System Function Test, and 5 897 simulated automatic operation
- a. of the entirt system at least With the relief valve function and/or once per 18 months.
the low low set function of one of the above required rTactor coolant system safety / relief valves inoper-able, restore the relief valve func-tion and the low low set function to OPERABLE status within 14 days or be in at least ICT SHLTTDOWN within the next 12 hours and COLD SWTDOWN within the following 24 hours.
~ *The lift setting pressure shall correspond to arbient conditions of tre valves at nominal operating terrperatures and pressures. **The setpoint for dedicated high steam dome pressure channels is s 1054 psig, duration of 35 days for power uprate test purposes.except that the unit may ope ***0ne instrment channel may be inoperable for up to 6 hours to perform required surveillances prior to entering other applicable actions.
IMTCH - LNIT I 3.6-9a techsp\h\94-060]J. pro \185 _.__ ___.---- --- -_--_----- - - - - - ~
. 4 .
(1) P.arinum fo er Level Georgio Fower Conpany 15 author 12eo to operate the f acility et steady state reactor core power levels not in excess of 2436 regawatts thermal in accordance with the conditions specified Attachment herein 2 is anand in Attachment integral part of this2license. to this licens{ -_
, en e p t-a g +-e Ly the O hpew s+a.+e. h "texels
% 'To *t'YInO'Y '
ente 5s c+ .2558 meg o urts 4Mml (2) Technical Specif1Cationi esn of 10 du.s 4.< S*' powJ*~ /l e,% Mu.,dv .f u rere (W*s
- vy wre re The Tecnnical Specifications containef in 4penMs A and B, as revised through Amendment No. d%p, are herecy incer-porated in the license. The licensee snall operate the f acility in accordance witt, the Tetr.nical Specifications.
(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisf action of the tcmission within the stated tine periods following the issuance of the license or within the operational restrictions indicated. The removal of these conditions shall be r.ade by ar. amencment to the license supported by a f avorable evaluation by the Commission. (g - ? e g.<,.__.-_,. 500'0 42 "^T" E"""; i'2'l, p 12- "; - ^ t . ,- f y , ,,m_, 4r. 4. m,t x.. ; ; ., ;_ . . r
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p y TABLE 2.2.1-1
-4 l $ REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES i --4
- 1. Intermediate Range Monitor, Neutron Flux-High s 120/125 Jivisions s 120/125 divisions (2C51-K601 A.8,C.D.E.F.G.H) of full scale of full scale l
l 2. Average Power Range Monitor: (2C51-K605 A.B.C.D.E.F)
- e. Neutron Flux Upscate,15% s 15/125 divisions s 20/125 divisions
' of full scale of full scale
- b. Row Referenced Simulated Thermal s (0.58 W + 59% - 0.58aW)' s (0.58 W + 62% - 0.58aW)'
Power-Upscale with a maximum with a maximum s 113.5% of RATED s 115.5% of RATED THERMAL POWER THERMAL POWER
- c. Fixed Neutron Flux-Upscale.118% s 118% of RATED s 120% of RATED THERMAL POWER THERMAL POWER
- 3. Reactor Vessel Steam Dome Pressure - High s 1054 psig"* s 1054 psig* "
l (2B21-N678 A.B.C.D) ' to 4 Reactor Vessel Water Level - Low (Level 3) a o inches above a O inches above 4 E (2821-N680 A.B.C.D) instrument zero* instrument zero*
- 5. Main Steam Une Isolatior Velve - Closure s 10% closed s 10% closed (NA)
- 6. (Deleted)
- 7. Drywell Pressure - High 51.92 psig s 1.92 psig (2C71-N650A,B.C.D)
,
- See Bases Figure B 3/4 3-1, to
*
- W = Totalloop recirculation flow rate in percent of rated. Roted loop recirculation flow is equel to 34.2 MLB!hr.
" AW = Maximum measured difference between two-loop and single-loop drive flow for the some core flow ia percent j
7 of rated recirculation flow for single 4oop operation. The value is zero for two-lom operation. to f *" The unit may operate with a setpoint of s 1065 psig for a cumulative duration of 35 days for power uprate test purposes. l 0 CD C to a p V
*1 O
bro. N
- l
l r TABLE 3.3.3-2 (Continued) 2> ~4 Q EMERGENCY CORE COOLING SYSTEM ACTUAT50N INSTRUMENTATION SETPOINTS I Q ALLOWABLE - TRIP FUNCTION TRIP SETPOINT VALUE
-t N 3. HIGH PRESSURE COOLANT INJECTION SYSTEM
- a. Reactor Vessel Water Level - Low Low (Level 21 1 -47 inches
- 1 -47 inches *
- b. Drywell Pressure-High 11.92 psig S 1.92 psig
- c. Condensate Storage Tenk Level- Low 10 inches" 10 inches"
- d. Suppression Chamber Water Level - High < 154.2 inches' * < 154.2 inches * "
- e. Logic Power Monitor NA NA
- f. Reactor Vessel Water Level-High (Level 8)* < 56.5 inches < 56.5 inches
- 4. AUTOMATIC DEPRESSURfZATION SYSTEM
- a. Drywell Pressure-High 11.S2 psig i 1.92 psig
- b. Reactor Vessel Water Level- Low Low Low (Level 1) 1-113 inches
- 1-113 inches *
- c. ADS Timer i 120 seconds i120 seconds w d. ADS Low Water Level Actuation Timer i 13 minutes i 13 rninutes
) e. f. Reactor Vessel Water Level - Low (Level 31 Core Spray Pump Discharge Pressure - High 10 inches
- 1137 psig 10 inches
- 1137 psig i g. RHR (LPCI MODE) Pump Discharge Presrure - High 1112 psig 1112 psig y h. Control Power Monitor NA NA
- 5. LOW LOW SET S/RV SYSTEM
- a. Reactor Steem Dorne Pressure - High 11054 psig S 1054 psig'"
l c+ (D n r en V /
=r
/ 4D s b Cn
- See Bases Figure B 3/4 3-1.
*
- Equivalent to 10.000 gallons of water in the CST.
E * *
- Measured above torus invert.
"" The unit may operate with a setpoint s 1065 osig for a cumulative duration of 35 days for power uprate test purposes. '
,P l 7 O / w b-d (A)
I TABLE 3.3.5-2 3= -4 CONTROL ROD WITHDR AWAL BLOCK INSTRUMENTATION SETPOtNTS Q TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE
- 1. APRM N
- e. Flow Referenced Simulated Thermal Power - Upscale s (0.58 W + 50% - 0.58 AWII *3' s (0.58 W + 50% - 0.58 AWII *I' l
- b. Inoperative NA NA
- c. Downscale 2 3/125 of futt scale 2 3/125 of full scale
- d. Neutron Flux - High,12% s 12/125 of fuit scale s 12/125 of full scale
- 2. ROD BLOCK MONITOR
- a. UpscaleM
- 1) Low Trip Setpoint (LTSP) s 115.1/125 of full scale s 115.5/125 of full scale
- 2) Intermediate Tnp Setpoint (ITSP) s 109.3/125 cf full scate s 109.7/125 of full scale W 3) High Trip Setpoint (HTSP) s 105.5!125 of fuit scate s 105.91125 of fuit scale a b. Inoperative NA NA w c. Downscele 2 94/125 of full scale 2 93/125 of full scale g d. Power Range SetpointI#3 o 1) Low Power Setpoint (LPSP) s 27% of RATED THERMAL POWER s 29% of RATED THERMAL POWER
- 2) Intermediate Power Setpoint (IPSP) s 62% of RATED THERMAL POWER s 64% of RATED THERMAL POWER
- 3) High Power Setpoint (HPSP) s 82% of RATED THERMAL POWER s 84% of RATED THERMAL POWER
- e. RBM Bypass Time Delay 5 2.0 sec s 2.0 sec (td,) WI
- 3. SOURCE RANGE MONITORS c+ e. Detector not futt in NA NA 8 b. Upscale s 1 x 10' cps s 1 x 10' cps d
- c. Inoperative NA ilA p d. Downscale 2 3 cps 3 cps
? 6 8 y
- The unit may operate with a setpoint s 0.58W + 53% -0.58AW for a cumulative duration of 35 days for power uprate test purposes. l 6
O <N s
REACTOR COOLANT SYSTEM SAFETY / RELIEF VALVES LOW-LOW SET FUNCTION LUilllNG CONDITION FOR OPERATION 3.4.2.2 The relief valve function and the low-low set function of the following reactor coolant system safety / relief valves shall be OPERABLE with the following low-low set function lift settings: Low Low Set Allowable Value (osia)* Valve Function Opf_q Close . Low $ 1010 $ 860 Medium Low s 1025 s 875 Medium High s 1040 s 890 High s 1050 s 900 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3 ACTION:
- a. With the relief valve function and/or the low-low set function of one of the above required reactor coolant system safety / relief valves inoperable, ,
restore the inoperable relief valve function and low-low set function ' to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within i the next 12 hours and in COLD SHUTDOWN within the following 24 hours. j
- b. With the relief valve function and/or the low-low set function of more ,
than one of the above required reactor coolant system safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUIDOWN within the next 24 hours.
- c. One instrument channel may be inoperable for up to 6 hours to perform -
required surveillances prior to entering other applicable ACTIONS. SURVEILLANCE RE0VIREMENTS 4.4.2.2 The low-low set relief valve function and the low-low set function pressure actuation instrumentation shall be demonstrated OPERABLE by performance of a: j
- a. CHANNEL FUNCTIONAL TEST, including calibration of the trip unit and the dedicated high steam dome pressure channels **, at least once per quarter,
- b. CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simulated automatic operation of the entire system at least once per refueling outage.
*The lift setting pressure of the valves is defined in subsection 3/4 3.4.2.1.
The accuracy of the low-low set setpoints is defined to be the accuracy of the instrumentation controlling the setpoints of the low-low set valves.
**The setpoint for dedicated high steam dome pressure channels is less than or equal to 1054 psig, except that the unit may operate with a setpoint
- s 1065 psig for a cumulative duration of 35 days for power uprate test purposes.
HATCH - UNIT 2 3/4 4-4a techsp\h\94-06U2E. pro \125 I
(2) Pursuant to the Act and 10 CFR Part 70. Georgia Power Company to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70 Georgia Power Company to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation nonitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CPR Parts 30, 40 and 70 Georgia Power Ccmpany to receive, possess and use in amounts as recuired any byproduct, source or special nuclear material without restriction to chemical or physical e fom, for sample analysis or inst ument t calibration or associated with radioactive $ apparatus or components; g (5) Pursuant to the Act and 10 CFR Parts 30 and 70, o Georgia Power Company to possess, but not separate, such byproduct and special nuclear
- d materials as may be procuted by the operation ,,
of the facility, g a C. This license shall be deemed to contain and is subject to the conditions specified in the follov.ing Commission regulations in
}.t l 10 CFR Chapter 1: Part 20, Secticn 30.34 of Part 30, Section 40.41 of Part 40, Sections 50-54 and 50-59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, anc orders of the Com.ission now or hereafter in effect; and is subject to the additional conditions specified or incorporated belo,<:
l (1) Maximum Pcwer level The Georgia Power Company is authorized to operate the facility at steady state reactor core power levels not in excess of 2436 megawatts themal ex ce pt %e kil;+ .j m g 6
- OP e% fr A af s taal3 s hin y.j tr Ites}s n e 4 i ~ am o f 35% o sey .sstts %cemLI (2) The"hnicaNc'e'ci Nc'afi t![ s$
The Technical Specifications contained in Appendices l A and B, as revised through Amencment No. /% are l hereby incorporated in the license. The licensee shall operate the facility in accordance with the . Technical Specifications.
BASES FOR LIMITING SAFETY SYSTEM SETilWGS 2.1 FREL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the HNP-1 Unit have been analyzed throughout the spectrum of planned operating conditions. The analyses were based upon plant operation in accordance with the operating map given in Figure 3-1 of Ref. 3. In addition, 24368MWt is the licensed l ' maximum power level of HNP-1, and this represents the maximum steady-state power which shall not knowingly be exceeded. Transient analyses perforned for each reload are given in Reference 1. Models and model conservatism are also described in this reference. As discussed in Ref erence 2, the core-wide transient analyses for single-loop operation are 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 6 f l 1 (
/ h ert h e on,i m43 6e oper,, w d ,t . 3+e ,1 9 3444c p.- v lee b oo+ ih e<cas oE ZSSP wL for a cents M< dved u .t J o days f - Ac Jt' A kt kes k pur, soc), l HATCH - UNIT 1 1.1-10 Amendirent No. 21, 38, #2, 185, 139, 141 l I I
SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS j 1.2. REACTOR COOLANT SYSTEM INTEGRITY 2.2. REACTOR COOLANT SYSTEM INTEGRITY I i ADDlicability ADolicability The Safety Limit, established to pre- The Limiting Safety System Settings apply serve the reactor coolant system to trip settings of the instruments and integrity, applies to the limit on the devices which are provided to prevent the j reactor vessel steam dome pressure. reactor vessel steam done pressure Saf ety Liinit f rom being exceeded. Objective Objective r The objective of the Safety Limit (asso- The objective of the Limiting Safety Sys-ciated with preserving the reactor cool- tem Settings is to defint,the level of ant system integrity) is to establish a the process variables at which automatic pressure limit below which the integrity protective action is initiated to prevent of the reactor coolant system is not the reacto'r vessel steam dome pressure threatened due to any overpressure con- Safety Limit from being exceeded, dition. Specifications Specifications A. Reactor Vessel Steam Dome Pressure A. Nuclear System pressure
- 1. When Irradiated Fuel is in the 1. When Irradiated Fuel is in the Reactor Reactor The reactor vessel steam dome pres- When irradiated fuel is present in sure shall not exceed 1325 psig at the reactor vessel, and the head any tine when irradiated fuel is is bolted to the vessel, the present in the reactor vessel. limiting safety system settings shall be as specified below:
Limiting Safety protective System Settings Action (psic) as '
- a. Scram on high < 1054 I '
reactor pres- 4 sure (reactor vessel steam l
- done press,ure) ,
- b. Nuclear system 4 valves at relief valves 1080 open on 4 valves at nuclear system 1090 pressure 3 valves at 1100 i
# Tk 06,+ ma opersle w oYk a selps'on+ b h psiy 4*v s eml%n dorak of Ydags Gr ywe < uper+e +gs + pap ose.s .
HAlCH - UNIT I 1.2-1 Amendment No. 27, 37,'103
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Table 3.1-1 REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION REQUIREMENTS g.,,,,; ; , w_1 [When:the reactor is subcritical and the reactor water temperature is-less than 212*F,'only;the'rollowing ' sin } sources.or.. scram trip signals need . to be operable: - x < Mode Switch in' SHUTDOWN 1 ' l
%.E^E m O. N j ' &',Q[M f
- Manual Sc ra m .
~
O + '
*- % %g u g i @ .KIRM High High Flux: ~ , .. e . . , . .
". ,- ; g .,
e v.yt:hM+:s ey &rScram . + y. Discharge Volume High High Level . 1 we. y
- e d,, a
;+ C . . : :. .- '
Cpe rab le R Screa s . ' . - - - . zs 3- --
' , ' VF illumbe r' . L Source or _ Scram Trip Signal '. n -Channels ' Scram Trip Setting" '? lSou.ce or- Scram signal is
, m h(a)46.,4
^ ~ '- Requi red Per .
...... l Required to .be Operobie Mof;N wAN * . . Trip system #,2 Except:ss Indicated Below
. + n#w n r:
s et,,. y. (b) ,i. g=y . v N g g p Mode Switch in' SHUTDOWN :.. 1 . Mode' Switch:in , : 6: W$Au'tomatIc'e'l1y bypeased two
, 7@{@D e
jgjg @ + ' 14
~
. SHUTDOWN: , seconds arter the Mode s
, ?
?33 - N- ,
*=. '
'83'/ Switch is pieced in the
.g sfnu e ..%
~+3ySHUTDOWNposItion.
i ,321:] %(' Manuel 4-- Scram 1 Depression of' Manual Scram Button
. , g' %
;;t3 . 43 ' *- t .. .IRM High High Flux ,
3 '5120/125 or rull. '
. IRMs a re automatically bypassed fen =T scale Tech Spec , when APRMs are on scale and the
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in.wp _. Mg.< a',
. .e n,g.,,-....
e
._ 2.1.A,1.m. ,
- ,4n ,c.. n, ,. .c 4 Mode Switch'Is in the Run N f9 O -
# q - QN .
v1 "-position.w * - ' CChE:h!.MMI"
. r
', Inopera t ive (p , 3'
't Not Applicable - '
Q.p l RMs i e re automa t i ca l ly bypa ssed
'fQ mAMQ%.
s
%.M. f-g:dcu-
- Wp T +:-
6.x C g m
<# 1vhen. APRMs are on ses se end. the tMode Switch is in the RUN
, n -e p .%;.sa
- ;+mn p wgem,p
.~ .
y~ *m ,.m :,
, ,' c 'po,sitioni '-
. s.
- - - ~
.: m v wen - . ,
- 4,. . .,
.m. - .>
, ~
51054 psIg *. . , s ", Ph Not' required'when reector head
. . - % J.. ;g 7 7;3s Reactor vesee1 Steam. Dome c(e ,~. . 23 M f/ M jd W TdC Pressure - Highy _
c
~g+; - y-3 Q; ,
-= Tech Spec,2.2.A.1..a, g j g ls not' bolted to the vessel.
93qmgpp ym y y n Q. 3 - a > ,-n .yz..,.uppy L mg W pi g3 q - pm . , . . 3 H,- a e_ n-
~r,_ . .
r,-g,yz_ g y
. , . g;-
.P-
;, y
.-' jem.
5
'. > -. M;p' , - , gg !_.W %m ,e , .
C3- '70 f h m-* A
~
..n ;:_G s ,&
4
,1 q_ ( e ; ' , _ ? hI - -
_,-c *y~ k G \q
",7*F]:e~~c b.h%-YNO WWlf :s%
j u V '# d ,I %
- i 'V,w
~i
- KNC),
Lyn ;;, .' :giL N O' h- wer YkAj WhY m e, m.. - Y g- g .s . S & ' , ;.y.O . , , l e [- ' 6 a y
* * ' ',,y u h xU:;}
i
> :* gWQ.f:32 + _
, , s _.
J.,,,,i TVM ew s_ m.
'. g p-} W? n q %s g i % .?4?
- - , n n.
i *. ._ _-__-__ _ __-_ - _ _ _ _ _ - _ _ _ - - _ _ _ _ _ _ _ . - . _ _ _ _ - - -_ - - -~ _ _.
I Table 3.2-7 (Continued) 33
--4 o
I Requ i red e Ope rab l e c Rer. Trip Channels 2 No. Condition per Trip i ns t rument Nomenc l a tu re System Trip Setting Remarks
--4 (a)
Downscale 2(b)(e) 23/125 of full scale Not required while performing low
- 3. APRM power physics test at atmospheric pressure during or af ter refueling a t powe r levels not to exceed 5 MWt.
12% Flux 2(b)(e) 512/125 or rull scale This function is bypassed when the Mode Switch is placed in the RUN position. Upscale 2(b)(e) 50.58 W + 50% - 0.58 AW see specirication 2.1.A.1.c(1) fo r definitions of W and AW. Trip level setting is in percent or ra ted power. Not required while performing low power physics tests at atmospheric pressure during or arter rerueling at power levels not lY to exceed 5 Mwt. N O4 RBM Inoperative 1(e)(r) Not applicable inoperative trip produced by switch not in operate, c i rcta i t boa rds not m in ci rcui t, rails to null, less than requi red number of LPRM inputs for rod selected. Downscale 1(e)(r) 291/125 4 or full scale b' a - 3
@m a
r+
- Tks v., i+ my opu<se. J. +L a re fro m+
.S h 0 C8ts +53 7 - o.Se AD x el m No- o f 1S dqs for pa w or rah. tes 1 pa
e s . ? % r commoiJav =d N t9 tyt *- W sN Sm . .tse N -N I i --t Table 3.214 $ INSTRUMENTATION WHICH ARMS LOW LOW SET S/RV SYSTEM C z Required [ OperaNe Trip Channele ~ Ref Condition per Trip @ No.8d Instrument Nomencleture Sy, tem . Trio Settina Romerke 3
- 1. Reactor Vessel Steem Dome High 2*' s1054 " sig #
7(M.g eres.ur. I 2. Relief /Sofety Velve High 2/volve 85. + 15. -5 The limiting cornstion o Tadpipe Pressure peig of operation of these awitchee le provided fc in Specificadon 3.6.H.1. kS lO W . o )c Tjg on,f mag c/ Ara be- ([t b (L Sch' ) "" '
- J u wh o n eF tr 49 & pa-u ur ~+c *'* P* 'P s es .
I r9 w a $ e. The column entitled "Ref. No." le only for convenience so that e one-to+ne relationship con be established between iteme in table 3.2-14 and iteme in table 4.2-14. c. $ b.1. With the requiremente for the rnirwmum number of OPERABLE channele not satisfied for one trip system, piece the l % inoperable channelin the tripped ' condetion or declare the sesociated system inoperable writhm one hour. With the requiremente for the minimum number of OPERABLE chennele not satisfied for both trip systeme. declare the
- essocisted system inoperable within one hour.
.o ** b.2. One instrument channel may be inoperable for up to 6 hours to perform required surveillances prior to entering {3 other applicable actione. De # D us I LIMITIPC C001T106 Fm OPERATIm SLRYFf ti#CE REQUIRIMNTS ! 4.6.H.l. Relief / Safety Valves (Continued)
- e. Ooerability of Tail Pine Pressure Switches
- 1. Fmettonal Test:
- b. At each scheduled out-age greater than 72 hours during dich on-try is made into the primary containment, if not performed with-in the previous 92 l days.
- 2. Calibration and verifying /
the setpoint to be 85, 415 -5 psig at least once per 18 months. 3.6.H.2. Relief / Safety Valves low Low 4.6.H.2. Relief / Safety Valves low Low Set Function Set Function During power operation startup, The low low set relief valve fac-and hot standby, the relief tion and the low low set faction valve function and the low low pressure actuation instrimenta-set function of the following tion shall be demonstrated OP0 TABLE *** l reactor coolant system safety / by perforinance of a: relief valves shall be OPERABLE with the following low low set a. OWNEL FLMCTIONAL TEST, function lift settings: including calibration of the trip unit and the dedicated Low Low Set AllowableValue(psig)* high steam done pressure i Valve Function QggD GSit channels", at least once Low 5 1005 5 857 Meditan 5 1020 5 872 b. O W NEL CAllBRATION, Logic Meditsn High s 1035 5 887 Systen function Test, and High $ 1045 5 897 simulated automatic operation of the entire system at least
- a. With the relief valve function and/or once per 18 sonths, the low low set function of one of the above required reactor coolant systen safety / relief valves inoper-
- able, restors the relief valve func-tion and the low low set function to OPERABLE status within 14 days or be in at least HDT SHLITDOW within the next 12 hours and COLD SHJTDOW within the following 24 hours. !
*The lift setting pressurt shall correspond to anbient conditions of the valves at nominal operating taveratures and pressures. v **The setpgint for tke s.utu.,a.5 dedicated cyon,tt ~s H- h)gh steam dome cr 1ety.& L. pressure fes en channels.m %is e.5'1054 psib'cm ~0ne instnenent channel may be inoperable for ip to 6 hours required ducdrs% of-surveillances prior to entering other applicable actions. c lo&S~ 2 3rWys9' [b,,, ~ ~ (*~T Sj o x/ pcN. IMTCH - LMIT 1~ 3.6-9a Amen &ent No. M , 85,92,493,185 Y Y I l i 1 ) l . a. (1) P.aninur 10-er Level Georgia Fower Conpany is author 1reo to operate the f acility et steady state reactor core power levels not in excess of 2436 re9aniatts thernal in accordance with the conditions spetified herein and in Atiachnent 21o ihis 1itense1] Attachment 2 is an integral part of this licente-m e. f S I.V 3 MN eau.h4 j , g..pt+t J sMe mues 4 leseltb *40*/'f M (2) Technical Specifications eners,orw arnhow m aj L aatta % es 7.2 f*c *- %vyva.,,,$e in f- W rn u o# y day.s r pe ,er The Technical Specifications containe. 4pendn es A and B. as revised through Amend'nent No. 4(s. are hereby inter-porated in the license. The licensee shall operate the f acility in accordance with the Tec .nical Specifications. (3) Additional Conditions The matters specified in the following conditions shall be completed to the satisf action of the Cornission within the stated time periods f ollowing the issuance of the license or within the operational restrictions indicated. The removal of these conditions shall be r.ade by ar. anendment to the license supported by a f avorable evaluation by the Comnission. ;g e_ _, e g, . , , , .. -<. , - . . e-.. .u,, . ~ _ .. _, . ii9 ig,; ;< a;.,. u.M m,i'4..g;;g;'il,' g,'33; _7,.,--.---,...__,__ _ u v,. ^^^i ;C. r--{2^ 11f 2af' i L' .s- s -- w .- w a a ri- ~ N r re i n , 4 m m scree. **{. P 11_ .- 7 ;i ; _} ;. . .n.h., . d r e, -x .4414,:- sq.,j;_ _ sm i;,-, _ g : _., g -- , + w a p.;, ,77.. g g , , um e n , n-n _m . , . , s, ;.7 ._ _ 1 g el ' --^ -- - ._____ _ ___ q W u ^.2. m n I PEACTOR PROTECTIOW SYSTEM INSTRUMENTATION SETPOfMTS C Z FUNCTIONAL UNIT TRIP SETPO!NT ALLOWABLE VALUf1 -4 m . 1. Intermediate Range Monitor, Neutron Nx-High 5 120/125 divisione s 120ft25 dMehme (2C51 Keot A.B.C.O.E.F.G.H) of fa scale of fue moeie
- 2. Average Power Range Monitor:
(2C51-K605 A.8.C.D.E.F)
- e. Neutron Nu-Upecede.15% 5 15/125 dMeione s 20/125 dMeions of fuG ocale of fuE oce6e
- b. How Referenced SmAsted Thermet 5 (0.58 W + 59% - 0.58AW)*
- s 10.58 W + 82% - 0.58aW1*
- Power-Upecele with a martmum with a medmum s 113.5% of RATED s 115.5% of RATED THERMAL POWER THERMAL POWER
- c. Rxed Neutron Nx-Upecale.118% s 118% of RATED s 120% of RATED THERMAL POWER THERMAL POWER
- 3. Reactor Vessel Steam Dome Pressure - H.gh 51054 peig
- s 1054 peig *
(2821-N678 A.B.C.0) 7 4 Reactor Veseet Water Level - Low (Level 38 2 0 inctwo above = 0 inches above (2B21-N680 A.B.C.D) instrument toro' instrument zero'
- 5. Me n Steam Line teoletion Velve - Closure s 10% closed (NA) s 10% closed
- 6. (Deleted) 7 l
, Drywest Preseure - High s 1.92 poig s :J 1.92 peig c. (2C71-NSECA.B.C.D) B to 3 *See Basee F,gure 8 3/4 3-1. Z o *
- W = Totalloop recircuterion flow rete in percent of rated. Reted loop recirculation flow is squel to 34.2 ML8thr.
a AW = Maximum measured difference between two-toop and single-loop dnve flow for the same core flow in percent , of reted recirculation flow for single loop operation. The value se zero for 1weloop operation. ~ G' x TM &< + m a.3 K ope ~ +e. us% r s&pi , +. of to W w ps a'y C. < a. c .,, m o lda ve. _D du rah oi of 3.c Jay fo< po w uprn+e- tad pu rpos e.s, . 5 ^ 2 k- --e TABLE 3.3.J-2 (Continued) S EMERGENCY CORE COOLING SYSTEM ACTUATIC+# INSTRUMENTATION SETPOINTS E _ . ALLOWABt E Q TRIP FUNCT ON TRIP SETPOINT VALUE
- 3. HtGH PRESSURE COOLANT INJECTION SYSTEM
- e. Reactor Veseed Water Level- Low Low (Level 21 1 -47 inchee
- 1-47 inchee *
- b. Drywell PreesureMgh S 1.92 peig S 1.92 peig
- e. Condeneste Storage Ter* Level - Low 10 inchee' 10 inches"
- d. Suppeession Chamber Water Level- High S 154.2 inchee'* S 154.2 inchee'*
- e. Logic Power Mordtor NA NA
- f. Reector Veseel Water Level High (Level 4)* $ 56.5 inchee S 56.5inchee
- 4. AUTOMATIC DEPRESSURIZATION SYSTEM
- e. Drywen Preeeuve-High S 1.92 poig S 1.92 peig
- b. Reector Vessel Water Level- Low Low Low (Level 1) 1 -13 3 laches
- 1 -113 inches *
- c. ADS Timer S 120 seconde S 120 seconde W d. ADS Low Water level Actuation Twner S 13 minutes S 13 rninutes a e. Reactor Veeeel Water Level- Low (Level 3) 1 0inchee* 1 0inchee* l w f. Core Spray Pump Discharge Pressure - High 1137 peig 1137 poig M g. RHR (LPCI MODE) Purm Discharge Pressure - High 1112 pelo 1112 peig w h. Control Power Monitor NA NA
- 5. LOW LOW SET S/RV SYSTEM ro e. Reactor Steem Dome Pressure - High S 1054 peig M* S 1054 poig N*
3 O. B ro 3 e+ Z .O
- See Basee Figure B 3/4 31.
*
- Equivaient to 10.000 gettons of water in the CST.
~ $ * *
- Measured above torus invert. go 67 k > avr The- un,+ emay e.ve } e. '
- S * +f* * *
- PN] S' < 4. Cum malz./on C. dordrekof3feq .s f= < fe*e' uf " +* f*3 f P"?M c3
_T_ABLE 3.3.5-2 I CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS --4 m I TRIP FUNCTION TRIP SETPOINT M LOWABLE VALUE i
- 1. APRM C
3 a. flow Referenced Simulated
- H T he rma l Power - Upsca le 5 ( 0.58 W + 50% - 0. 58 AW) ' *M*
(0.58 W5 + 50% - 0.58 SWl'* M l m b. I nope ra t i ve NA NA
- c. Downscale 2 3/125 of full scale 2 3/125 of full scale
- d. Neutron flux - High, 12% 5 12/125 of full scale 5 12/125 of full scale
- 2. ROD BLOCK HONITOR
- a. Upscale
- 1) Low Trip Setpoint (LTSP) 5 115.1/125 of full scale 5 115.5/125 of full scale
- 2) Intermed ia te T rip Setpoint (ITSP) 5 109.3/125 of full scale 5 109.7/125 of full scale
- 3) High T rip Setpoint (HTSP) 5 105.5/125 of full scale 5 105.9/125 of full scale
- b. Inoperative NA NA
- c. Downscale 2 94/125 of full scale 2 93/125 of fulI scale
- d. Powe r Mange Setpo i nt '
- 1) Low Power Setpoint (LPSP) 5 27% of RATED THERMAL POWER 5 29% or RATED THERMAL POWER
- 2) Intermedia te Power Setpoint (IPSP) 5 62% of RATED THERMAL POWER 5 64% of RATED THERMAL POWER
- 3) High Powe r Se t po i nt (HPSP) 5 82% of RATED THERMAL POWER 5 84% of RAIFO THERMAL POWER
- e. RBM Bypass Time Delay 5 2.0 sec 5 2.0 sec W ***
\ (tds)
- 3. SOURCE RANGE MONITORS u
1 C)
- a. Detector not fulI in NA 5 1 x 10 5cps NA 5 1 x 10 5 cps
- b. Upscale
- c. I nope ra t i ve NA NA
- d. Downscale 2 3 cps 2 3 cps P-
~ The. un t[ W4g opWN *
- 5# 88'N P d 0.8 B tu -t f.3 % - Q fg A d f La e c u , ,,,vi d i n. 40<< % af L'~ 4% % p -a< ut< h t s+ po rps e s .
CL 2 c+ 2" O N tu N %1 N c-REACTOR COOLANT SYSTEM SAFETY / RELIEF VALVES LOW-LOW SET FUNCTION LIMITING CONDITION FOR OPERATION l 3.4.2.2 The relief valve function and the low-low set function of the following ! reactor coolant system safety / relief valves shall be OPERABLE with the following ) I_ low-low set function lift settings: ! 1 Low Low Set Allowable Value (osia)* j Valve Function Qgen Close l Low s 1010 s 860 f.' Medium Low Medium High $ 1025 s 1040 5 875 s 890 High s 1050 s 900 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3 l ACTION:
- a. With the relief valve function and/or the low-low set function of one of j the above required reactor coolant system safety / relief valves inoperable, I restore the inoperable relief valve function and low-low set function to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
b b. With the relief valve function and/or the low-low set function of more than one of the above required reactor coolant system safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD 4 SHUTDOWN within the next 24 hours.
- c. One instrument channel may be inoperable for up to 6 hours to perform required surveillances prior to entering other applicable ACTIONS.
SURVEILLANCE RE0VIREMENTS 4.4.2.2 The low-low set relief valve function and the low-low set function pressure actuation instrumentation shall be demonstrated OPERABLE by performance of a:
- a. CHANNEL FUNCTIONAL TEST, including calibration of the trip unit and the dedicated high steam dome pressure channels **, at least once per quarter. l
- b. CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simulated automatic
, operation of the entire system at least once per refueling outage. *The lift setting pressure of the valves is defined in subsection 3/4 3.4.2.1. The accuracy of the low-low set setpoints is defined to be the accuracy of the instrumentation controlling the setpoints of the low-low set valves. (7 **The setpoint for dedicated high steam dome pressure channels is less than or V equal to 1054 psig;# e<c4y t AI Re- u m f ~t orectl dA o. S }p $4 g &r cLluwmahftur u sa b " ~y' W $ HATflI7hNI[2 #' / Amendment No. 33, 125 3/4 4-4a CLARIFicATIou o-95-ol, Rev.o F0Lt cu> s PAGer P/4 )-if, L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _}}