Text

Amend 39 to License NPF-5,amending Tech Specs Re Fuel Type, Operating Limit Min Critical Power Ratio & Allowing Use of Hybrid I Control Rod Assemblies
	ML20090G560
Person / Time
Site:	Hatch
Issue date:	07/13/1984
From:	Lainas G; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20090G562	List: ML20090G560; ML20090G564; ML20090G568;
References
	TAC-54169, TAC-54607, TAC-62029, NUDOCS 8407240599
	Download: ML20090G560 (46)
	v • d • e

{{#Wiki_filter:1 ,emrug'o l UNITED STATES ~g E NUCLEAR REGULATORY COMMISSION o-e WASHINGTON, D. C. 20555 %,....../ GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTPJC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET N0.J 0-366 EDWIN 1. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 39 License No. NPF-5 1. The Nuclear Regulatory Commission (the Comission) has found that: A. The applications for amendment by Georgia Power Company, et al., (thelicensee)datedJanuary 23, 1984, as supplemented April 3, 1984, June 7, 14 and 15, 1984; February 6,1984, as supplemented April 3,1984, June 20 and 27,1984; and April 3,1984, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the applications, 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 sithout endangering the health and safety'of the public, and (ii) that such activities will be conducted ~ in compliancs. with the Comission's regulations; 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.

2.. Accordingly, the licen$e '~is amended by changes-to the Technical Specifications 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:

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2-Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 39, 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. FOR THE NUCLEAR REGULATORY COMMISSION ($? l / . (dow. Gus C. Lainas, Assistant Director for Operating Reactors Division of Licensing

Attachment:

Chances to the Technical Specifications Date of Issuance: July 13,1984 e ,e 4 a

0 ATTACHMENT TO LICENSE AMENDMENT N0. 39 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of th Appendix "A" Technical Specifications with _ the enclosed pages. The revised pages are identified by Amendment number and contain verti'al lines indicating the area of change. The corresponding overleaf pages are provided to maintain document completeness. ~.. - Remove Insert ' 2-4 2-4 B-2-10 B 2-10 B 2-12 B 2-12 B 2-13 B 2-13 3/4 1-17 3/4 1-17 3/4 2-1 3/4 2-1 3/4 2-4g 3/4 2-4h 3/4 2-41_ 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3-3/4 2-7a 3/4 247a 3/4 2-7b 3/4 2-7b 3/4 2-7c 3/4 2-7c 3/4 2-7d 3/4 3-2 3/4 3-2 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 4 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3 3/.4 3-13 , 3/4 3-14 3/4 3-14 3/4 3-15' 3/4 3-15 4 3/4 3-16 3/4 3-16 m 3/4 3-17 .3/4 3-17 3/4:3-18 3/4 3-18 3/4 3-19 '3/4 3-19 3/4 3-20 3/4 3-20 3/4 3-21 13/4 3-21 3/4 3-22 3/4 3 3/4 3-23 3/4 3-23 3/4.3-26 3/4:3-26. 1 3/4 3-27; f 3/4~3-27 3/4 3-28 3/4 3-28 3/4 3-29 _3/4.3-29' 4 6 ~ t e v w m. ww e y

_ Remove Insert 3/4 3-31 3/4 3-31 3/4 3-32 3/4 3-32 3/4 3-34 3/4 3-34 3/4 3-35 3/4 3-35 3/4 3-36 3/4 3-36 3/4-3-39 3/4 3-39 3/4 3-40 3/4 3-40 ? 3/4 3-40a 3/4 3-42 3/4 3-42 3/4 3-54 3/4 3-54 3/4 4-18 3/4 4-18 3/4 6-41 3/4 6-41 3/4 9 3/4 9-4 B 3/4 1-4 B 3/4 1-4 B 3/4 1-4a B 3/4 2-1 B 3/4 2-1 B 3/4 2-3 B 3/4 2-3 B 3/4 2-4, B 3/4 2-4 B 3/4 2-5 B 3/4 2-5 B 3/4 2-6 B 3/4 2-6 B 3/4 3-6 B 3/4 3-6 B 3/4 9-1 B 3/4 9-1 5-3 5-3 1 ' _ * ' ~ G y

. z. -- _ i- ~, i i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1. Y APPLICABillTY: As shown for each channel in Table 3.3.1-1. ACTION: With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the apolicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. e ammense .; T e ] .] .j =: ~ I I fI KATCH - UNIT 2 2-3 LL ~

.o... .wi TABLE 2.2.1-1 I ic j REACTOR PROTECTION SYSTEM INSTRilHENTATION SETPOINTS t s 4 FilNCTIONAL llNIT TRIPSETPOINI ALLOWABLE VALUES 1. lutermediate Range Honitor, Neutron Flux-liigh < 120/125 divisions < 120/125 divisions ~ (2C5n-K601 A,B,C,D,E,F,G,II) of full scale of full scale I: 2. Average Power Range Monitor: (2C51-K605 A,B,C,D,E,F) l 3. y Neutron Flux-Upscale, 15% < 15/125 divisions < 20/125 divisions 'l' a. of full scale _ of full scale h. Flow Referenced Simulated Thermal < (0.68 W + 59 %), < (0.58 W + 62 %), l Power-Upscale _ with a maximum with a maximum t < 113.5% of RATED < 115.5% of RATED j TilERNAL POWER TilERNAL POWER Y c. Fixed Neutron Flux-Upscale, 118% -< 118% of RATEi) < 120% of RATED TilERNAL POWER-TilERHAL POWER ~ i I 3. Reactor Vessel Steam Dome Pressure - IIIsh $ 1054 psig <ql054 psis I -i (2B21-N678 A,B,C,D) 4. Reactor Vessel Water Level - Low (Level 3) 1 8.5 inches above 1 8.5 inches above l (2B21-H680 A,B,C,D) instrument zero*- instrument zero* 5. Main Steam Line Isolation Valve - Closure 3 (NA) ~< 10% closed < 10% closed, ~ e. g. 6. Hain Steam Line Radiation - liigh -< 3 x full p6wer < 3 x full power O (2ill i-K603 A, B,C,D) background background, ce ..y 7. Drywell' Pressure - liigh < l.85 psig~ < 1.85 psig l (2C71-N650A,B,C,D) 3. I$ 3 E See Bases Figure IF3/4 3-1. 1

~ u __. u _ _ _ a.__ j' 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 6 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection Syst6m Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each i parameter. The Trip Setpoints have been selected to ensure that the reactor i core and reactor ca.ol, ant 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. Intemediate Range 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 accomodate the increase in po'wer 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 subcritical 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 byrassed. The results of this analysis show that the reactor is shutdown ar ' eak power is limited to 1% of RATED THERMAL POWER, thus maintaining MCPR above 1.07. 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. Averaoe Power Range Monitor t .For operation at low pressure and low flow during STARTUP, the APRM i scram setting of 15/125 divisions of full scale neutron flux provides adequate thermal margin between the setpoint and the Safety Limits. The-ma'rgin accomodates the anticipated maneuvers associated with power ] plant startup. Effects of. increasing pressure at zero or low void - i content are minor and cold water from sources available during startup 1s not 1 much colder than that already in the system. Temperature coefficients 1 are small and control rod patterns are constrained by the RSCS and RWM. 4 i 1 Amendment No.,W,21-HATCH - UNIT 2' 'B29 ~ _

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2.2 LIMITING SAFETY STSTE." SETTINGSS

{ BASES (Continued)- I c REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) Average Pewer Range-Menitor (Continued) Of all the possible sources of reactivity input, uniform control rod with-drawal is the cost probable cause of significant power increase. 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 amount, the rate of power rise is very slow. Gen-e: ally.the heat flux is in near equilibrium with the fission rate. In an p-assumed uniform rod withdrawal approach to,the trip level, the rate of i power rise is not more than 5% of RATED THERMAL. POkT.R per minute and the .APRM system would be more than adequate.to assure shutdown before the i power could exceed the Safety, Limit. The 15% neutron flux trip remains l' active until the mode switch is placed in the Run position. The APRM flux scram trip in the Run mode consists of a flow referenced simulated thermal power scram setpoint and a fixcd neutron flux scram set-4 point. The APRM flow referenced neutron flux signti is passed through a i filtering network with a time constant which is representative of the fuel dynamics. This provides a flow referenced signal that approximates tha average heat flux or thermal power that is developed in the core during transient or steady-state conditions. i The APRM flow referenced simulated thermal power scram trip setting at full recirculation flow is adjustable up to 113.5% of RATED THERMAL PO'iER. i,___ This reduced flow referenced trip setpoint will result in an earlier scram during slow thermal transients, such as the loss of 100*F fdedwater heating ~ event, than would result with the 118% fixed neutron flux scram trip. The lower flow referenced ' scram setpoint therefore decreases the severity, ACPR, of a slow thermal transient and allows lower operating limits if j. such a transient is the limiting abnormal operational transient during a i certain exposure interval in the fuel cycle. !~ l The APRM fixed neutron flux signti does not incorporate the time constant, q but responds directly to instantaneous neutron flux. This scram setpoint H scrams the reactor during fast power increase transients if credit is not. taken for a direct (position) scram, and also serves to scram the1:eactor e if credit is not taken for the flow referenced simulated thermal power j scram. = The APRM setpoints were selected to provide adequate margin'for the Safety 4 Limits and yet allow operating margin that; reduces the possibility of L unnecessary shutdown. 1+ H' i J !!il y,A7;g. 2 - 3 2-10 . Anendnent !!o.~ J4, 39 , $Y ta# 3 "od. t

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Lw M.m. c._:_... u _ _ .r 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continue' ) d REACTOR PROTECTIO'l SYSTEM INSTRUMENTATION SETPOINTS (Co' .. j - 3. Reactor Vessel Steam Dome Pressure-Hioh n High pressure in the nuclear system could cause a rupture to the b nuclear systen process barrier resulting in the release of fission f A pressure increase while. operating will also tend to in ' products. crease the pcwer of the reactor by compressing voids thus adding j !j The trip will quickly reduce the neutron flux, counter-reactivity. The trip 'f acting the pressure increase by decreasing heat generation. setting is slightly higher than the operating pressure t'o permit normal The setting provides for a wide margin operation without spu'rious trips. to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that. occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure trip is typassed. For a turbine trip under these conditions. the transient analysis indicated a considerable margin to the thermal hydraulic limit. j l 4 Reactor Vessel Water Level-Lou The reactor vessel / water level trip setpoint was chosen far enough {~~ below the nonna1 operating level to avoid spurious trips but high enough .l above the fuel to assure that there is adequate protection _for the.. fuel 3-1 and pressure barriers. I Main Steam Line Isolation valve-Closure l,7 5. ? The main steam line isolation valv'e closure trip was provided to limit ' ' The the amount of fission product release for certain postulated events. MSIVs are closed automatically from measured parameters such as high 4 steam flow, high steam line radiation, low reactor water level,- high steam tunnel temperature and low steam line pressure. The MSIV closure a scram anticipates the pressure and flux transients which could follow y MSIV closure, and thereby protects reactor vessel pressure and fuel t thermal / hydraulic Safety Limits. 6. ~ Main Steam Line Radiation-Hioh The main steam ^line radiation detectors are provided to det'ect.a gross failure of the fuel cladding. When the high radiation is' detected q' a trip is initiated ;to reduce the continued failure 'of fuel cladding. ~At the same. time the ma'in steam line isolation valves are closed to. limit the release of fission products. The trip setting is high enough 'l above background radiation levels to prevent spurious trips yet low (7 enough to promptly detect gross failures in the fuel cladding, q-r t HATCH - UNIT 2. .B 2-11 Amendment lo. 14 w

y i,55th$httrrr'szstr.:rs. AnNr" "~ ]: J l: DASES'(CAntinued) ~ C REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 7. Drywell Pressure-High 'High pressure in the drywell could indicate a break in the nuclear process systems. The reactor is': ripped in order to minimize the possi-bility of fuel damage and reduce the amount of energy being added to the coolant. The trip setting was selected as low as possible without causing spurious trips. + ky 8. Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered. The reactor is therefore tripped when the water level has reached a peint high enough to indicate that it is indeed filling up, bu 'the volume is still great enough to accommodate the water from the movement of the rods when they are tripped. 9. Turbine Stoo Valve-Clesure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves. With a ::1p setting of 10% of valve closure from full 'open, the resultant increase in heat flux-is such that-adequate thermal =argins are maintained even.during the worst case transient' that assumes the turbine bypass valves remain closed. This scram is. bypassed when the-turbine steam flow is below that corresponding to 30*. of RATED THERMAL POWER, as measured by turbine first stage.pressurt. 1 10. ~urbine Centrol Valve Tast Closure. Trio Oil Pressure-Low The. turbine control valve fast closure trip anticipates the ~ pressure, neutron flux, and heat flux increase that could result ait from fast closure of the turbine con:rol valves.due to load rejection (( coincident with failures of the turbine bypass valves. The Reactor Protectien System initiates a trip when fast. closure of the1 control b valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of-control valve fast closure. This is [l ' achieved by the action of the fast acting solenoid valves in rapidly d reducing hydraulic trip oil pressure at the main turbine -control ki valve actuator-disc' dump valves. This loss of pressure is. sensed by b t 1~

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lS .. w ~ LIMITING SAFITY' SYSTEM SETTING 1 -BASES (Continued) 4 REACTOR PROTECTION SYSTEM INSTRUMENTATION SET?OIh7S (Continued) T Turbine Control Valve Fast Closure, Trin Oil Pressure-Lew (Continued) + 4 pressure switches whose contacts form the one-out-of-two-twica logic input to the Reactor. Protection System. This trip setting, a nominally 'L 30*,; greater closure time and a different valve characteristic from that /! of,the, turbine stop valve, combine to produce transients very similar to .J., that for the stop valve. No significant change in MCPR occurs. Relevant

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transient analyses are discussed in Section 13,of the Final Safety Jj Analysis Report. This scram is bypassed when turbine steam flou is fj 'below that corresponding to 30% of RATED THER'!AL POWER, as measured by l f turbine first stage pressure. 4 j-11. Reactor Mode Switch In Shutdown Position The reactor mode switch Shutdown position trip is a redundant channel to the automatic. protective instrumentation channels and provides additional manual reactor trip capability. '8 12. Manual Scram The hanual Scram is a redundant channel-to the automatic protective instrumentation channels and provides manual reactor trip capability. o bi .1 1 1 !J ti [ 1 a 3 haj l! ' a a , [.

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~ Au-.=L:M h'Mb . z =.a.a. = =..-a a -.. - - ~.- -- -- ~~--. i ' REACTIVITY CONTROL SYS'IEMS ROD BLOCK MONITOR S LIMITING CONDITION FOR OPERATION ) Both Rod Block Monitor (RBM) channels shall be OPERABLE., 3.1.4.3 APPLICABILITY: CONDITION 1, when TERMAL POWER is greater than the preset power level of _ the RWM and RSCS and when: ~ THERMAL POWER is < 90% of RATID THERMAL POWER and the MCPR is a. less than 1.70, or -f;;. j -b. THERMAL POWER is 2 90% of RATED THERMAL POWER and the MCPR is i less than 1.40. 3 ACTION: With one RBM channel inoperable, POWER OPERATION may continue a. provided that the inoperable RBM channel is restored to 0FERABLE status within 24 hours; otherwise, trip at least one rod block monitor channel within the next hour. j ..] b. With both REM channels inoperable, trip at least one rod block 1i monitor channel within one hour. ,y .s .I SURVEILLANCE REQUIREMEhTS t 4.1.4.3 a. With both RBM channels OPERABLE, surveillance requirements are ,J, given in Specification 4.3.5. l b. With one RBM' channel INOPERABLE.'the other channel shall be. 2 f; demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST ~ i prior to withdrawal of control rods. -1 1 7* d:l- .j 4_ 4 . Ifli HATCH-2 3/t. 1 Amendment No. 39 i a. l

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g, e ' REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM s LIMITING CONDITION FOR OPERATION 4 a l' 3.1.5 "The standby liquid control system shall be OPERABLE with: I . 5. An OPERABLE flow path from the storage tank to the reactor core containing two pumps and two inline explosive injection valves, and b. The contained solution concentration and the solution temperature are within the Operating Range of Figure 3.1.5-1. APPLICABILITY: CONDITIONS 1, 2, and 5*. ACTION: ,a. In CONDITION 1 or 2: j 1. With one pump and/or one explosive valve inoperable, restore.the inoperable pump and/or explosive valve to OPERABLE. status within 7 days or be in at least HOT ' SHUTDOWN within the next 12 hours. 3 L 2.~ With the standby liquid control system inoperable, restore the system to OPERABLE status within 8 hours or ' 1 be in at least HOT SHUTDOWN within the next 12 hours. i i b. In CONDITION 5*:- .i 1 1. With one pump and/or one explosive valve. inoperable, restore the inoperable pump and/or explosive valve to 2 OPERABLE status within 30 days or fully insert all insertable-9 control rods within the next hour. ~. - 2. With the standby liquid control system inoperable, fully ~ insert all insertable control rods within one hour. .h .3. The provisions of Specification 3.0.3 and 3.0.4 are not J applicable. u b-

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.w. 3/4.2 POGR DISSIlr3Im LIMITS 3/4.2.1 AVI!'RM2 PIANAR LINEAR FEAT GE2GRATim RATE LIMITING C:liDITIm KR OPERAT!m t 3.2.1 All AVERAGE: PIANAR LINEAR HEAT G!!NERATICN RATS (APGGRs)' shall be qual to or less than the applicable APEGR limit, which is a function of fuel type and AVERAGE PIANAR EXPOSURE. Se APIEGR limit is given by the applicable rated-power, rated-flew limit taken frca Figures 3.2.1-1 through 3.2.1-9, multiplied by the amaner of either: l a. S e factor given by Figure 3.2.1-10, or 1^ b. D e factor given by Figure 3.2.1-n. l APPLICABILITY: CCNDITION 1, when 'If!!! MAL PCEER 2 25% of RAIED "SE!NAL POWER. l WCN: l l With an APLIER exceeding the limits of Figures 3.2.1-1 through 3.2.1-9, as 1 adjusted per Figures 3.2.1-10 and 3.2.1-11, initiate corrective action within L 15 minutes and continue corrective action so that the API 2GR meets 3.2.1 within 2 hours or reduce THI! MAL POWER to less than 25% of RATED MEMAL PCWER within the next 4 hours. l SURVEILIANCE REQUIRl!M1!NIS 4.2.1 All APIJERs shan be verified to be qual to or less than the applicable limit determined fran Figures 3.2.1-1 through 3.2.1-9, as adjusted per Figu:e 3.2.1-10 and 3.2.1-n 4 s. At least nnce per 24 hours, !j b. Whenever MEINAL PONER has been increased by at least 15% of RATED il THilmAL PONER and steady state operating conditions have been 'l established, and q d c. Initially and at least once per 12 hours when the reactor is 'i operating with a LDiITING CCtmCL RCD PATIDN for APIBGR. 3 l ~

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... z_ _....__ _._awx. POWER DISTRIBUTION LIMITS 3/4.2.2 ~ APRM SETPOIhTS s This'section deleted. t. ) I f t 1., t a 3 1 1 f;. ,e - a t . l r[= 1 ? v 4 ~ .). e

- e

( 'l I ^ P. 4.- '/> _ 7 / ' HATCH-2 3 /t. 2-3 Amendment No. H, H, 39 n. w._

u.x _ w.,. -.... ..~.~-..~.~....a . P0k*ER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 All MINIMUM CRITICAL POWER RATIOS (MCPRs), shall be equal to or greater than the MCPR operating limit (OLMCPR), which is a function of average scram time, core flow, and-core power. For 25% 5 Power < 30%, the OU!CPR is given i in Figure 3.2.3-5. For Power 2 30%, the OU!CPR is the greater of eithert

a..The applicable limit determined from Tigure 3.2.3-4, or f

b. The appropriate X given by Tigure 3.2.3-3, multiplied by the appropriate j p limit from Tigure 3.2.3-1, 3.2.3-2, or 3.2.3-3, where: ~ r = 0 or , whichever is greater, rA _ r g I rA = 1.096 see (Specification 3.1.3.3 scram time limit te notch 36), r3 = 0.834 + 1.65 '_,N,,,, " (0.059), n Ng -I i=1 r n I N = i=1 i 1 i fave n 1 j i I i=1 s n= number of surveillance tests performed to date in cycle, th Ng = number of active control rods measured in the i surveillar.ce j

test, g = average scram time to notch 36 of all rods measured in the i

th i surveillance test, and { Ng = total number of active rods measured in 4.1.3.2.a. p i fl i APPLICABILITY: CONDITION 1, when THERMAL POWER 1 25% RATED THERMAL POWER . ACTION: I Vith MCPR less than the applicable limit determined from Specifica-ion p 3.2.3.a. cr 3.2.3.b, initiate corrective action within 15 minutes and continue corrective action so that MCPR is equal to or greater than the applicable lia.it within 2 hours er reduce THERMAL POWER to less than or equal to 25% of RATED THERMAL POWER within the next 4 hours. t i HATCH-2' 3/ 26 Amendment No. 21, 33, 39 1

t m. A w.c_ w.. -=;a ...., u. u. _-, _ _...:. w _.....a c d wun : 1 3/4.2.3 MINIMUM CRITICAL' POWER RATIC M INUED)' SURVEILLANCE REQUIREMENTS

t ll 4.2.3 The MCPR limit at rated flow and rated power shall be determined for

.each type of fuel (8X8R, PSXSR, and 7X7) from Figures 3.2.3-1, 3.2.3-2, and 3.2.3-3 using: a. t = 1.0 prior to the init,ial scram time measurements for the cycle performed in accordance with Specification 4.1.3.2.a. or i '{- 1 b. t as defined in Specificatica 2.2.3; the determination of the limit i must be completed within 72 hours 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 ' l'imit: 4 b a. At least once per 24 hcurs, a ,gb b. Whenever THERNA ~'IR has been increased by at least 15*. of a RATED THERMAL Pt .. and steady state operating conditions have* been established, and t c., Initially and at least once per 12 hours when the reactor is operating with a LIMITING C0hTROL ROD PA'ITERN for MCPR.* e Is k; l} i f i i f t 7 c t 1 l t 1

j 4

t MATCE-2 3/4 2 7' Amendment No. U, 39~ ? c.- L

g ~ .. 7~.e -. -. - w am.a.2 a ,a m._.. a,, m _n. ,,, 3,gmg.m.; 1 ' 1**so ..n... .a.m....._.._. m...a c.m..a i+-

=::.......

s -- - - : = u ;.=_n..

u:..m..

... l u n. v. e .~..~ 4=:. : :.::. a=;:M O..li t y .; U _ _.j;.. :::;;..:.&.C;% *r

- ""_".: M* a. =* : T *...

. '.. "' : ;r. *.:: : *:: 1 - ' : " J "_M". U

M::.::!.:

t --::M. n 1 k-n'd[N511 Mi!:* i: 5:E! 5E** 5 fliiE3:'.d b.fi$.3'*"Ei[ E ~diI* =... I .:.n = r.n= = n.. :.= := n = a ::. n1:-- ..r.n

=:- :.:== = nr /:;::

: =. r. - : = :--- a= =- : n = - : =:=== = = = ?=== =.- ;:* : = : 2.

3g J .........;-*=50

".:"O.:.. ;u*.;; ;=..... ? Z7.=~ '" ' ~* :.=C; : U ". =.;;.T.:.=.

22.= :== =.p= = =r = =:.= :: = = =- u====_ = -.. : -- - - g i n = m =;==- - _.... _ :== n = u :. :-,- - = r. ;<r

u-n

- - : - u = = r = =_==: :.. _... _ =.u = := - _:-.- -- : -. _=.. = ;,

...:=..:.:_=.:.=_..=..=_..:=._n._.==..=_:.=.:_===__.._==.....-f.!._......

.=._=_a...__:._-..-* .=_= ;;.y_=...i. ;_3;=.;.:=..=..=_. _=.=.. =.=.; =__; ;. g n ; =.rg 3_g ;:.g :___. =__=.:3 :;. .__...../....,

- y 3

~ E.i.i..:E.. i. 2.Eid_. !. :.:_E.E_.. i.E_E =_=... _=:..r.r. " = " - - - - =. _ _. - E i_ E%a-= ;___. _ _3=i T HE__.==== u_ _ _ u E-i*33 3d = : = : :- :-. u:= --- ::: _-/e . m....:.u u =.._:.=.;_ _. : n ........... :.=. a= n un = :. - a i

".. : =: -. =: -. =.. : = u== n 4--

= = = ===n. r_==a -- r a r -- n== :y.. r = =..u :

n. : :_: -, =:=

?' =...a.

==

.1

3'.i

._ i.si.=_=.=.E..d_.i__1..z; 2_:P_.:..n - i.=_i._E!.=__= 5..:.E.jr. =._=."_= - 1 __." = =: _....._.EE =__;=.. #..._.._.._.__ _..__.._. _... _.._.-._._ _.. M- _L__t_._n

- - = -

_=. _ _ ___,__==._.. _ ; _--===_= =.__ j-1,31 = =.L..o __

....._.___.......__.__..._.:=:=t_.

4 = : = =::=y. ::.= = = i== : ===: = =v = nc r.= ___ __ _. _._. _. =_=.:._:.::7'--_.=_.._'"--:~-"_":~=;.._.._- L 30 r, ; [= "::r---

= -

..:======-:===.g.._____.__.m.-__

==u.--.=_::=--.:-~-.---..-..--"----------- = ---- i; 1.29 e --_:--__=_=_._

==--:__.._... w g.. g.._._ 1.2S 0.0 0.2 0.4 0.6 0.8 1.0 ow 7 H N SXSR FUEL

4 m

FIGURE 3.2.3-1 I' ai! l. 3 8,==. ;.=_ =..:._ __. _ _. u: =, = n a.r = = n!=- a ::IT n= =: =.=u.=:a = =:

==._.r_ ._.._ _.. = : n - =_: =_.n - ---.____;__..._._;=......._..._==_=__ = : d =,;., _ _... _m...___. 1,37 = - -.._. - -- = = = : r== - __-==_.._._====-:- cs .=....v..=..-._=..=__=.v._=.___=_=.._...._.3.=_.._g....-_._._......2==.=u==-._

=.

m .1 ___=.._.=.=::.._..______===._..::==u===-_-f_=..=_. z.; ...f. gg 1, 6 cc- ,, :: =_; .,4 - W

===*.:..___- p egC = _ _.. -.f'I 1.35 1 t- = = = = =. =. - j I".- = = - ' - - - -+- _ _ - -. _ _. _. __.._:=.=.g__--_ _.. -.. _.. 1,34 .a .. _.....f w_._._.. _53. i.. 5 D.C..._.7..._ilY.... _'E -"-' -* b!-[-Y 3 . I.. g = .71 g*g 9 fj =_._=_.:===.:---- _ _f; ... = = - - - - = - - - 1 g l=..=.p.__. \\ e " =. J n := j=== =.a._ _: l:n m o ..._ _==_ _ - __ =:= c . h r- - - --- st 1.31 _# r-g_"-'_"=..= s .==_ ,,. =. =. = - ~ _.. _ =;.g.g i m gg.g_. _.. _-.;=.,g__._ =.. A 1,29 7 l ~

0. 0 -

0. 2 -

0.4 0.6 -0.8 .0 T, c1 P8X8R FUEL -FIGURE 3.2.3-2 7 .t s 4 - HATQi > UND 2. 3/4 2-7a A:::endment-No. 27,. gg Y 39 r . k!

i 1

6 L-i y-

~ ..o.. __.....m ,s , -._ m. .- - - - - - ~i -' ~A* A^UM s 4 g !p kj 1.4 r ', '( 1.33 i

3 1.32 1 31 ACCEPTABLE OPERATION 1.30

. f 1.29 ., ). i

.{

e. i + gyg -,2 ~~ 127 i j.i 126 J" NACCEPTABLE OPERATION l i-! t 'j 125 l l .s t3 '.3 124 1 *);

't -

123 'l om o.2 oA es os im 7 FIGURE 3.2.3 3 MCPR LIMIT FOR 7X7 FUEL ] AT RATED FLOW AND RATED POWER !] 1 s. I 4 1 e i t HATCP.-2 3/4 2 7b Amendment No. 33.~39-

?

I

.'i5['r.; iIil ebLt _1 i! ,i t ,j:! (f!lI e k 'i f-f 0 2 1 0 lJ. 1 C. '-: 'F l i 0 1 0 0 I 1 C W 7 1 W O 1 p L 37 7 5 = F 04 g 5 A, y E E B

3. 7 s. 8 T

0 R 1 1 1 1 2 A O R 1 C 0 ( W i s X D 2268 O A E F 3 08 7 L M T A 8 8.5 5 F A 0000 = R MU p W%%% R = W 0 0.0. 5 P X O 7 27 2 M 0 0 5 C A 0 1 1 00 A 2 7 2 e L M W MF M1 0 0 3 1 1 1 1 1 1 1 ID p Jl E T H 1. A P R C M 0 ( i 7 4g W 4 O -3 L F 2 E R 3 O N C 6 R e e U G I F 0 s 'h t 4_ a 2 i A w i C E 0 2 8 7 5 3 2 1 1 I 1 1 1 1 1. ,Eb2 E ~ b 1O::IeJ 9N M4n yo Oo3 +r - Ow up ()a* -c t J l.

. _... _.. ~... ^ ' .a i = : = w.- 2^ 5 N

=="LO! 1 .. - ~ ~ - - - -..- -.- 8 3 2 n 8:- E e w t.* a-a g g .t = = = I Z g 3 E-e b 55 E_ 8 o I sa a -- o ww _e x

=

~ g 2 g g oo 6 s w. o uu A g a .O W .a w a s I g ee o o x _I m w d sS g *e~ oo x = I,; g .x .i "h g' g' a-6 j. : 8 + g gE in e e ww = = n ~ ~s s -a g, = 8. n a t eo e n. W $a S e s-2 22 LLI ~ y a. 8 E A t. cc s x x v x v x x e W 6 6 6 3 s z e vi v' vi 'v _cp x 5[ E l l 'l S

e e

e e e .o o o o o 8 o w w u. m J j

g g

~ 4} g g. ~ a

f'.

5 6 4 e e i' A VI N' j- -ec f. [)I g_g - l / q . t. 4e _ _ ff_ _. _ _ _ _ _ _ _ _ _ _ _ _ n i I I I l I r' f I = r I I I JJ c' m t. n n 9 = n-m o. 4 n n n n n n n e e %0CRd WOd (a ) WEl'adII"IfM W43W 031YW X _ %CC>d5%5E WOd WDdW10 ... c...n-e. nai

3/4'2 7d Amendment flo. 39

7; g g -- -+ -- a-. la 3/4.3 INSTRUMENTATION 3/4.3.1 - REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentatien channels shown in Table 3.3.1-1 shall be OPERAELE with the REACTOR PROTECTION SYSTEM y RESPONSE TIME as shown in Table 3.3.1-2. Set points and interlocks. are given in Table 2.2.1-1. APPLICASILITYi As shown in Table 3.3.1-1. ACTION: With the requirements for the minimum number of OPERASLE channels not a. satisfied for one trip system, place at least one inoperable channel in the tripped condition within one hour. b. Uith the requirenents for the minimum n'mber of OPERABLE channels not u ti satisfied for both trip systems, place at least one inoperable channel i in at least one trip system

in the tripped condition within one hour

}

and take the ACTION required by Table 3.3.1-1. The provisions of Specification 3.0.3 are not applicable in OPERA-c. TiONAL CONDITION 5. 4 y SURVEILLANCE REOU!REMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be a demer.strated OPERABLE by the performar.ce of the CHANNEL' CHECK,' CHANNEL FUNCTION TEST and CHANNEL CALIBRATION operations dur n; the OPERATIONAL d COND:TIONS and.at the frequencies shown in Table 4.3.1-1. ~ .ti 4.3.1.2 LOGIC SYSTEM FUNCTIONAL' TESTS and simulated autor.atic operation of all channels shall be performedlat least once per 18 months and'shall 1 .? include calibration of time delay relays and timers necessary for proper ~l functioning of the trip system. 1 4.3.1.3.The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip function of Table 3.3.1-2 shall be demonstrated to be within its limit at

i least once.per 18 months.

Each test shall include at-least'one logic train such that both logic trains ~ are tested at least once per 36 months and one , char.nel cer. function such that all channels are tested at least once every N times 18 months where N is the-total number of redundar.t channels ~ in a specific reactor trip function. 'if botn channels are inoperable in one trip system, select:ac least one increrable -channel in that trip system to place'in the tripped conditions, i exct:t when this could cause'the Tric -Function to occur. ~ HtiCH - UNIT 2 '3/4 3-1 A.endment'No. E J i

I TABLE 3.3.I-1 REACTOR PROTECTION SYSTEH INSTRilHENTATION APPLICABLE HINIHUN NUMBER OPERATIONAL OPERABLE CilANNELS h FilNCTIONAL UNIT CONDITIONS PER TRIP SYSTEN(a) ACTION + .n .m' l. Intermediate Range Honitors: 'h (2C51-K601, A, B, C, D, E, F, G, H) i .-i IC), 5 (I') a. Neutron Flux'.High 2 3 1 3, g) 2 2 h. Inopera t.ive 2,5 3 1 3,4 2 2

2. ' Average Power Range Monitor:

I (2C51-K605 A, B, C, D, E, F) i a. Neutron Flux - Upscale, 15% 2, 5 2 1 h. Flow Referenced Simulated y Thermal Power - Upscale 1 2 3 c. Fixed Neutron Flux - [ Upscale, 118% 1 2 3 -d. Inoperative 1, 2, 5 2 4 e. Downscale-1 2 3 L h f. LPRH_ 1, 2, 5 (d) NA 3. Reactor Vessel Steam Dome Pressure - High (2B21-N678 A, B, C, D) 1, 2(e) 2(j' 2B21-NC45 5 A, B, C, D) $} l; g (ii a 4. Reactor Vessel Water Levhl - g. Low (Level-3) (2B21-N680 A, B, C, D) 1, 2 2(j, 2 1 5 9 5 g 5. Hain Steam Line Isolation Valve - 3' ' Closure (NA) 1(g) 4 3 m 6. Hain Steam Line Radiation - High 1, 2(e) 6'- 2 6 U (2Dil-K603 A, B, C, D) 7. .Drywell Pressure - High 1, 2(8) 2 5 (2C71-N650 A, B, C, D) l

.n-.. l f c TABLE 4.3.1 F ~ h.. REACTOR PROTECTION SYSTEH INSTRUMENTATION SilRVEILLANCE REQUIREMENTS I a .t g CHANNEL OPERATIONAL

i CllANNEL FUNCTIONAL CilANNEL CONDITIONS IN WilICil N

FUNCTIONAL UNIT CllECK TEST CALIBRATION (* SURVEILLANCE REQUIRED i 1. Intermediate Range Monitors: I-II'

Neutron Flux - High

.D-S/U R 2 a. D W R 3,4,5 b. Inoperative NA W NA 2,3,4,5 } '2. Average Power Range Monitor. l S/U(I'} (C), W(d) gjg(h) g(d) Neutron Flux - Upscale, 15% S a. y N S W W 5 ) W ")II) f

b. -Flow Referenced Simulated S

S/U ,W , SA 1 8 Thermal Power -' Upscale f S/U(b) y g(c) Fixed Neutron Flux - Upscale, S c. , SA 1 ,f 118% d. Inoperative NA W NA 1, 2, 5 c. Downscale NA W NA 1 's f. LPRH D NA (g) 1, 2,'S [. L 3. Reactor' Vessel Steam Dome S H R 3 Pressure - liigh 1, 2 l h 's R R 4. Reactor Vessel Water Level - S H R g: 'R L Low (Level 3). 1, 2 E W E 2 5. Main Steam Line' Isolation-Valve - ? Closure' NA H R(h) 1 p, 6. Main Steam Line Radiation - High D W RU) 1, 2 h it 7. -Drywell Pressure - High S H R 1, 2 l h t 8. Scram Discharge Volume Water NA N R(h) L Level - High 1, 2, 5 i b i I L

.. _... ~.... ~ ' " ' ~ ~ ~ mw ?; g. TARI.E 4.3.1-1 (Continued) H p' S REACTOR PROTECTION SYSTEH INSTRilHENTATION SIJRVEII,f.ANCE REQllIREHENTS e',, CHANNEI., OPERATIONAI. CIIANNEL FilNCTIONAI. CII3NNEI. CONDITIONS IN. WilICil r FIICTIONAI IINIT CIIECK TEST CAI.IBRATION SilPVEII.T.ANCE REQlIIRED i. 9 Turbine Stop Valve - Closure NA H R(h) g 10 -Turbine Control Valve Fast ~ 4 Closure, Trip'011 Pressure - i.ow ifA H R I 11. Reactor Hode Switch in Shutdown NA R NA 1,2,3,4,5-Position 12. Manual Scram NA H NA 1,2,3,4,5 y. a. Neutron detectors may be excluded f rom CIIANNEL CA1.IMPATION. h. . Within 24 hours prior to startup, if not performed within the previous 7 days. i l: -The APRM, IRH and SRH channels shall be compared for overlap durir.g each startup, if not [ R-c. { performed within the previous 7. days. i Cp E d. .When changinE from CONDITION 1 to CONDITION 2, perform the required surveillance within 12 hours p after entering CONDITION 2. f. l This calibration shall consist of the adjustment of the APRH channel to conform to the power values O c. calculated by a heat balance 'dt ring CONDITION 1 when THERHAL POWER 2 25% of RATED TilERHAL POWER. b Adjust the-APRH channel If the absolute difference 2 2%.. 1 f. This calibration shall consist of the adjustment of the APRH flow referenced simulated thermal I g

power channel to conform to a calibrated flow signal.

S' .g g. The I.PRit's shall he. calibrated at Icast.once per 1000 effective full power hours (EFPil) using the

(

g TIP system. t

. g h.

Physical inspection and actuation of switches for instruments 2Cll-N013A, B, C, D. M.

l..

Instrument alignment using a standard current source. J. Calthration using a standard radiation sou'rce. I. L ?

t TABLE 3.3.2-1 ISOLATION ACTUATION INSTRIRIENTATION .. [ :a i, ,p VALVE GROUPS HINIHIRI NUMBER APPLICABLE' H OPERATED BY OPERABLE CHANNELS OPERATIONAL y TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEN(b)(c) CONDITION AC','I ON c: N. I. PRIHARY CONTAINHENT ISOLATION

a.. Reactor Vessel Water Level

~ 1. Low (Level 3) 2, 6, 10, 2 1,2,3 20 (2B21-N680 A, B, C, D) 11, 12-2. Low-Low.(Level 2) 5, f,

2 1,2,3 20 (2B21-N682 A, B, C, D) 3.

Low-Low-Low (Level 1) 1 2 1,2,3 20 -- [l; (2B21-N681 A, B,'C, D) f: b. .Drywell Pressure - High 2, 6, 7, 10, 2 1, 2, 3, 20 l '(2C71-N650 A, B,-C, D) 12, f,

H 1

w c. Main Steam Line b 3 1. Radiation - Nigh 1, 12, f, (d) 2 1, 2, 3 21 !?j (2D11-K603 A, B, C, D) g u [;, 1 2 1 22 2.* Pressure - Low s (2B21-N015 A, B, C, D) 3. Flow - High 1, # 2/11ne 1, 2, 3 21 r (2B21-N686 A, B, C, D) h;: (2B21-N687 A, B, C, D) [' (2B21-N688 A, B, C, D) d (2B21-N689 A,.B. C, D), d. Main Steam Line Tunnel 2/line(*} 1, 2, 3 21 k-p' Temperature - High 1 (2B21-N623 A, B, C, D) h o lh (2B21-N624 A, B, C, D) (; (2B21-N625 A, B, C, D)

3 (2821-N626 A, B, C, D) 1, 2,(I) 3(O 23 e.

Condenser Vacuum - Low: 1 2 .g (2B21-N056 A, B,'C, D) p 3. q ' g.

f.

Turbine Building Area g Temperature.- Nigh 1 2 1,2,3 21 L (2061-R001, 2U61-R002,.2U61-R003, e N -2U61-R004) it V

..p g4., .e- . * +. l p TABLE 3.3.2-1 (Continued) l- ? ISOLATION ACTUATION INSTRUNENTATION ~ VALVE GROUPS HINIHUN NUMBER APPLICAllLE N OPERATED llY OPERABLE CllANNELS OPERATIONAL I TRIP FilNCTION SIGNAL (a) PER TRIP SYSTEH(b)(c) CONDITION ACTION U 2. SECONDARY CONTAINHENT ISOLATION u a. Reactor Building Exhaust [ Radiation - liigh 6, 10, 12,

2 1,2,3,5 and**

24 J (2Dil-K609 A, B, C, D) b. Drywell Pressure - liigh 2, 6, 7, 10, 2 1, 2, 3 24 [ (2C71-N650 A, B, C, D) 12, #,

c.

Reactor Vessel Water i, Level - Low Low (Level 2) 5, #,

2 1, 2, 3 24 If (2B21-N682 A, B, C, D)

[; P s" d. Refueling Floor Exhaust I$ [ Radiation - liigh 6, 10, 12, #,

2 1,2,'),5 and**

24 (2Dil-K611 A, B, C, D) 4:, 9 3. REACTOR WATER CLEANUP SYSTEM ISOLATION ~ a. A Flow - liigh (2G31-N603 A, B) 5 l 1, , 3 25 b. Area Temperature - liigh 5 1 1,2,3 25 I -(2G31-N662 A, D, E,li, J, H) l, 3 c. Area Ventilation A Temp. - liigli 5 1 1, 2, 3 25

R (2G31-N663 A, D, E,11, J, H; R-2G31-N661 A, D, E,11, J, H; 2G31-N662 A, D, E, II, J, H) a*

? ~ 5(g) 2 d. SLCS' Initiation (NA) NA 1, 2, 3 25 ~ w-c. Reactor Vessel Water Level - Low Low 5, #,

2-1, 2, 3 25 (f.evel 2);(2B21-N682 A, B, C, D) 5

TABLE 3.3.2-1 (Centinued) ISOLATION ACTUATION INSTRlRIENTATION .l F .-i .4 9 VALVE GROUPS HINIHilH NIJHBER APPLICABLE h OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEH(h)(c) CONDITION ACTION- +g 4. IIIGil PRESSURE COOLANT INJECTION SYSTEM ISOLATION I a. IIPCI Steam Line Flow - High 3 1 1, 2, 3 26 (2E41-N657-A,B) l b. IIPCI Steam Supply Pressure - Low (2E41-N658 A,B,C,D) 3,8 2-1,2,3 26 l c. IIPCI Turbine Exhaust Diaphragm I' w, Pressure - High (2E41-N655 A,B,C,D) 3 2 1, 2, 3 26 l i E d. IIPCI Pipe Penetration Room .k Temperature - High (2E41-N671 A, B) 3 1 1,2,3 26. l 0 u e. Suppression. Pool' Area Ambient { Temperature-High (2E51-N666 C, D ) 3 1 1, 2, 3 26 l i

f. -Suppression Pool Area J.

A Temp.-High (2E51-N665 C, D; 3 1 1, 2, 3 26 i -2E51-N663 C,:D; { 2E51-N664 C, D) g F- ~ g. Suppression Pool Ares Temperature } g)

s Timer Relays (2E41-N603 A, B) 3 1,

1, 2, 3 26 l [ h. Emergency Area Cooler Temperature-High (2E41-N670 A, B) 3 1 1,2,3 26 l z .i. Drywell Pressure-High 3 -(2 Ell-N694 C, D) 8 1 1,2,3 26 w J. Logic Power Monitor (2E41-K1) NA 1 1, 2, 3 27 i i t t ['

' ~ ~ ... ~ - " T.~;.' :. '~ ^ - - ~ ~ ~ ' - ~ ,w-J j TABLE 3.3.2-1 (Conticutd) ISOLATION ACTUATION INSTRlRlRNTATION '{ VALVE GROllPS HINIHilH NilHRER APPLICAHLE v-OPERATED BY OPERABLE CHANNELS OPERATIONAL (c TRIP FilNCTION SIGNAL (a) PER TRIP SYSTEN(b)(c) ' CONDITION _ ACTION H' 5. REACTOR CORE ISOLATION i h COOLING SYSTEH ISOLATION { 9 I i y RCIC Steam Line Flow-High 4 1 1, 2, 3 26 1, i. a. (2E51-N657 A,B) 3 y h. RCIC Steam Supply Pressure - .l Low (2E51-N658 A, B, C, D) 4, 9 2 1, 2, 3 26 l [ 't c. RCIC Turbine Exhaust . Diaphragm Pressure - High 4' 2 1, 2, 3 26 (2E51-N685 A, B,-C, D) l

d. -Emergency Area Cooler Temperature -

t u D High (2E51-N661 A, B) 4 I 1, 2, 3 26 _ta e s [ e. Suppression Pool Area Ambient c-Temperature-High 4 1 1,2,3 26 {. (2E51-N666 A, B) I t f. Suppression Pool Area a T-High. 4 1 1, 2, 3 26 [ (2E51-N665 A, B; 2E51-N663 A,B; 2E51-N664 A,B) g. Suppression Pool Area Temperature '[- 4,g) C Timer Relays (2E51-H602 A, B) 1 1,2,3 26 h. Drywell Pressure - High f g (2E11-N694 A, B) 9 1 1,2,3 26 l l Sy i. Logic Power Monitor (2E51-K1) NA(h) 1 1,2,3 27 3 6. SHUTDOWN COOLING SYSTEM ISOLATION 2 a. Reactor Vessel Vater Level-Low (Leveli 6,.10, 11, 2 2 3, 4, 5 26 7 3)(2R21-N683 A, B, C, D) ~ 12 b. Reactor Steam Dome Pressure-High 11 1 1,2,3 28 (2831-N679 A, D).. 1

w-n uw w. ~~. c = ua. w ..w. 3lI TABLE' 3.3.2-1 (Continusd) .ISOI.ATION ACTUATION INSTRUENTATION l~ ACTION ACTION 20 Be in at least HOT SHUTDOWN within 6 hours and in COLD SHUTDOWN within the next 30 hours. -ACTICH 21 Be in at least STARTUP with the main steam line isolation valves closed within 2 hours or be in at least HOT SHUTDOWN within 6 hours and in COLD SEUTDOWN within the next 30 hours. ACTION 22 Be in at least STARTUP within 2 hours. 1 i ACTION 23 Be in at least STARTUP with the Group 1 isolation valves closed within 2 hours or in at least HOT SHUTDOWN within 6 hours. ACTION 24 Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within one hour. ACTION 25 Isolata the reactor water cleanup system. l ACTION 26 Close the affected system isolation valves and declare _the affected system inoperable. ACTION 27 Verify power availability to the bus at least once per 12 hours or close the affected system isolation valves and declare the affected system inoperable. ACTION 28 Close the shutdown cooling supply and reactor. vessel head spray i isolation valves unless reactor steam dome pressure i 145 psis. I NOTES i Actuates operation'of the main control room en'ironmental control v system in the pressurization mode of operation. Actuates the standby gas treatment system. When handling irradiated fuel in the secondary containment. See Specification 3.6.3, Table 3.6.3-1 for valves in each valve group. a. b. A channel may be placed in an inoperable status'for up to 2 hours for L;j condition provided at least one other OPERABLE channel in the same trip required surveillance without placing the trip system in the tripped system is monitoring that parameter. ~. i' With a design providing only one channel per trip system, an inoperable c. channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable-channel.shall be restored to OPERABLE status within.2 hours or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken. } -d. Trips the mechanical vacuum pumps. e. A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE. f. May be bypassed with all turbine stop valves closed. g. Closes only RWCU outlet isolatica valve-2G31-F004. h. Alarm only. 1. Adjustable up to 60 minutes. KATCM'- UNIT 2 3/4 3-15 Amendment No. P, 39 m

. ~;. : ^ ~... :..

' ~ " ^ -

^ ~ ' ~ ^ - ~ -^ ~ ~ 'l t TAELE 3.3.2-2 ISOLATION ACTilATION INSTRiitIENTATION SETPOINTS g ~ a ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALilE i, 1. PRIHARY CONTAINHENT ISOLATION {~ a. Reactor Vessel Water Level 1. Low (i.evel 3) > 8.5 inches * > 8.5 inches

2.

Low Low (Level 2) 5 inches

5 -55 inches
3.

Low Low Low (Level 1) [-121.5 inches * [-121.5 inches

b.

Drywell Pressure - High < 1.85 psig 5 1.85 psig l c. Main Steam Line 3 -< 3 x full power background 1. Radiation - High < 3 x full -backgroun0ower y o~ 2. Pressure - Low 1 825 psig 1 825 psig, t 3. Flow - High < 138% rated flow < 138% rated flow l m E d. Main Steam Line Tunnel Temperature - High $ 194*F < 194*F l i 1 e. Condenser Vacuum - Low 1 7" Hg vacuus 1 7" Hg vacuum l f. Turbine Building Area Temp.-High < 200*F $ 200*F 2. SECONDARY CONTAINHENT ISOLATION a. -Reactor Building Exhaust y Radiation - High < 60 mr/hr 5 60 mr/hr l $g b. Drywell Pressure _- High 5 1.85 psig i 1.85 psig I* E' c. Reactor Vessel Water g --Level - Low Low (Level 2) 1 -55 inches

1 -55 inches
l g

d. Refueling Floor Exhaust i w -< 20 mr/hr -< 20 mr/hr Radiation - Nigh -u

See Bases. Figure B 3/4 3-1.

r -- -- - ... a a..,- u ., c. w h(.; p TABLE 3.3.2-2 (Continued) L, y. ISOLATION ACTilATION INSTRUNENTATION SETPOINTS

c y

ALLOWABLE [] TRIP FUNCTION TRIP SETPOINT VALUE g 3. REACTOR WATER CLEANUP SYSTEM ISOLATION a. A Flow - High i 79 gpm < 79 gpm .[hj-b. Area. Tempera ture-liigh < 124*F $ 124*F l [ c. Area Ventilation A Temperature - High < 67*F $ 67'F l [ f-d. SLCS Initiation NA NA [, u e. Reactor Vessel Water Level-Low Low 3 (Level 2) ~> -55 inches * > -55 inch'es* I- ~ i .L w' 4. HICil PRESSURE COOLANT INJECTION SYSTEM ISOLATION O -c a.. IIPCI Steam Line Flow-liigh < 307% of rated flow ,$ 307% of rated flow l b. IIPCI Steam Supply Pressure - Low > 100 psig > 100 psig c. IIPCI Turbine Exhaust Diaphragm Pressure-High $ 20 psig < 20 psig l d. HPCI Pipe Penetration Poom Temperature - High' $ 169*F $ 169*F l e. Suppression Pool Area Ambient ' Temperature-High $.169"F $ 169"F l-f. Suppression Pool Area AT - High < 42.5'F < 42.5"F 1 g. Suppression Pool Area Tmsperature 'i l Timer. Relays NA NA L. h. Emergency Area Cooler Temperature - p - High $ 169"F < '59*F h 1. Drywell-Pressure.- High < 1.85 psig $ 1.85 psig j. g. J. Logic Power Bus Monitors NA NA

~

i j; 1

Sce Bases Figure B 3/4 3-1.

2 [ii 'k

k Y.

,,_--_=---___--.--__...u-- - _--.w : . s ~ 'h. t b e 'f TABLE 3.3.2'-2 (Continued) n n J. T ISOLATION ACTt1ATION INSTRllHENTATION SETPOINTS M. ' 9. b o ]p. y ALLOWABLE TRIP FilNCTION TRIP SETPOINT VALUE t, ' r. 5. REACTOR CORE ISOLATION h COOLING SYSTEM ISOLATION xll a. RCIC Steam Line Flow - High $ 312% of rated flow $ 312% of rated flow l f'li

b..RCIC Steam Supply Pressure - Low 1 60 psig 1 60 psig l

l c. RCIC Turbine Exhaust Diaphragm j.' Pressure - High $ 20 psig 5 20 psig l [, t d. Emergency Area Cooler Temperature-High < 169*F < 169"F l yx I _ Suppression Pool Area Ambient Temperature 5 169"F $ 169'F l e. m1 liigh ll = f." Suppression Pool Area AT - High $ 42.5'F $ 42.5*F l i r g. Suppression Pool Area Temperature Timer-Relays NA NA h. Drywell Pressure - High $ 1.85 psig i 1.85 psig l 1

i. ' Logic Power Honitor NA NA ic l1 6.

SHilTDOWN COOLING SYSTEM ISOLATION, 7 a. Reactor Vessel Water Level - Low 1 8.5 inches *- 1 8.5 inches * -(Level 3) ~. .b. Reactor Steam Dome Pressure - High. $ 145 psig i 145 psig [I iJj. s

See_ Bases Figure B 3/4 3-1.

h. 0;. s g

3 ^F?.'1 G;1 _ _w a & L-A a -- " " N"- -C L"---- 4 *- AU o l TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME'(Seconds)# 1. PRIMARY CONTAINMENT ISOLATION i a. Reactor Vessel Water Level P-. 1. Low (Level 3) i 13* 2. Low Low (Level 2) < 13* 3. Low Low Lov (Level 1), except MSIVs 513* b. Drywell Pressure - High 1 13* c. Main Steam Line 1. Radiation - High*** 1 1.0** 2. Pressure - Low < 13* 3. Flow - High 51.0** 4. Reactor Vessel Water Level - Low Low Low < 1.0** (Level 1) ~ i d. Main Steam Line Tunnel Temperature - High 1 l'3* e. Condenser Vacuum - Low NA a f. Turbine Building Area Temperature - High NA l '2. SECONDARY CONTAINMENT ISOLATION a. Reactor Building Exhaust Radiation - High*** 1 13* b. Drywell. Pressure - High 1 13* Reactor Vessel Water Level - Low Low (Level 2) 1 13* l q c. 1 j d. Refueling Floor Exhaust l Radiation - High*** 1 13*

The isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE

-TIME. Response time specified'is diesel generator start delay time 'I assumed in accident analysis.

- Isolation actuation instrumentation response time.
  - Radiation detectors are exempt from response time testing. Response i

time shall be measured from detector obtput or the input of the first . electronic component in the channel.

Times to be added to valve movement times shown in Tables 3.6.3-1, 3.6.5.2-1 and 3.9.5.2-1 to obtain ISOLATION SYSTEM RESPONSE TIML for each valve.

HATCH - UNIT 2 3/4 3-19 Amendment No. 33,39

.. :.... L TAELE 3.3.2-3 (Continued) ISCLAYICN SYSTEM INSTRUENTATION RES:0NSE TIE TRIP FtJCTION RESPONSE TIE (Seconcs)# 3. REACTOR WATER CLEANLP SY.du:.M ISOLATION a. 4 Flow - High $ 13+ b. Area Tem::erature - High. $13+ Area Ventilation Temperature 4T High 6 D+ c. d. A R Initiation NA e. Reacter Vessel Water Level-Low Low (Level 2) 6 13* l 4. HICH PRESSURE COOLANT INatt TION SYSTEM ISCLATION a. HPCI Steam Line Flow-High 3 6 Isolation Timen D* b. HPCI Steam Su:: ply Pressure - Low 613+ c. HPCI Turbine Exhaust Diaphragm Pressure - Hich NA d. HPCI Pipe Penetration Room Temperature - High NA l Suppression Pool Area Amcient Temp. - High NA e. f. Sup;:ressicn Pool Area 4 T - High NA g. Suppression Pool Area Temp. Timer Relays NA h. Emergency Area Cooler Temperature - High NA. 1. Drywell Pressure - High 513+ j. Logic Power Monitor NA 5. REACTOR CORE ISCLATICN COOLItC SYSTEM ISCLATION a. RCIC Steam Line Flow - High 3 $ Isolation TimedD* b. RCIC Steam Supply Pressure - Low NA RCIC Turbine Exhaust Diaphragm c. P essure - High NA d. Emergency Area Cooler Temperature - High NA Suppression Pool Area Amoient Temp. - High NA e. f. Sucpression Pool Area 4' - High NA g. Suppressicn Pool Area Tecocrature Timer Relays NA h. Drywell Pressure - High n D+ 1. Logic Powe'r Monitcr NA 6. SHUTDOWN CDOLItG SYSTEM ISCLATION a. Reactor Vessel Water Level - Low (Level 3) NA l b. Reactor Steam Dcme Pressure ,High NA HATCH - UNIT 2 3/4 3-20 Amendment No. 33, 39

... -.... ~. l .k t:e TABLE 4.3.2-1 . ISOLATION ACTUATION INSTRUNENTATION SilRVEILLANCE REQllIREHENTS y. N CilANNElf OPERATIONAL _y CilANNEL FUNCTIONAL CliANNEL CONDITIONS IN WilICil t.e Q. TRIP FilNCTION CIIECK TEST CALIDRATION SURVEILLANCE REQtlIRED ( M L 1. PRIMARY CONTAINHENT ISOLATION l . i;; a. Reactor Vessel Water Level .I 1. Low (Level 3) S H R 1, 2, 3 { '. 2. Low Low (Level 2). S X R 1, 2, 3 3. Low Low Low'(Level 1) S H R 1, 2, 3 b. Drywell Pressure - liigh S, if R 1, 2, 3 l c. Main Steam Line ] g 1. Radiation - High D W R 1, 2, '3 { 2. Pressure - Low NA H Q l 3. Flow - High S H R 1, ' 2, 3 l p' u L d. Main Steam Line Tunnel g Temperature - Nigh S H R 1, 2, 3 l u t.1, e.. Condenser Vacusim - Low NA H. Q 1, 2#, 3# I. f. Turbine Building Area Temp. - liigh NA H R 1, 2, 3 2. SECONDARY CONTAINNENT ISOLATION p a. Reactor Building Exhaust r l-Radiation - Nigh D M(*) R 1, 2, 3, 5 and

1 E

I g h. Drywell Pressure - Nigh S H R 1, 2, 3 r* +. y c. Reactor Vessel Water Level - { ~: -Low Low (Level 2) S H R U 1, 2, 3 l-d. Refueling Floor Exhaust t Radiati~n - High D H(*} Q 1, 2, 3, 5 and * '( to o

When handling irradiated fuel in the secondary containment, y

y,

Nay he bypassed with all turbine stop valves closed.

l alnstrument alignment using a standard current source. [.

A[ o TABLE 4.3.2-1 (Continued) -y ISOLATION ACTUATION INSTRUMENTATION SilRVEILLANCE REQUIREllENTS os i 'E ~ CHANNEL OPERATIONAL i h CilANNEL FUNCTIONAL CIIANNEL CONDITIONS IN WilICll ll< 4 TRIP FUNCTION CllECK TEST CAL 1hRATION SURVEILLANCE REQUIRED ca L '3. REACTOR WATER CLEANUP SYSTEM ISOLATION ~ ,I a. A Flow - High D H R 1,2,3 b. Area Temperature - liigh S H R 1, 2, 3 l c. Arca Ventilation A [ Temperature - liigh S H R 1, 2, 3 l d. SLCS Initiation NA R NA 1, 2, 3 t R e. Reactor _ Vessel Water Level - S H R 1; 2, 3 Low Low (Level 2) u y 4. HIGil PRESSilRE COOLANT INJECTION ) F SYSTEM ISOLATION l a. HPCI Steam Line Flow-High S H R 1, 2, 3

h. ;HPc1 Steam Supply Pressure-Low S

H R 1, 2, 3 c. HPCI Turbine Exhaust S H R 1, 2, 3 Diaphragm Pr2ssure - High d. HPCI Pipe Penetration Room ~ f Temperature - High ' S H R 1, 2, 3 l ,E g- .e.. Suppression Pool Area Ambient g ' Temp. --High. S H R 1, 2, 3. l 3 f. Suppression Pool Area AT - High S H R 1, 2, 3 l [ 2 o g. Suppression Pool Area Temp. Timer Relays NA SA R 1, 2,. 3 u* h. Emergency Area Cooler Temp. - i High S H R 1, 2, 3

i. Drywell Pressure - High S

H R 1,2,3 .J. Logic Powcc Honitor NA R NA 1, 2, 3 [ ; E t hj

" ~ ~ ' ^ . :. d. [ m. -__a- . - S a s _... / -m j ~ . l.. TABLE 4.3.2-1 (Continued) V ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREllENTS I ca

g

-A i d CilANNEL OPERATIONAL CilANNEL FUNCTIONAL CllANNEL CONDITIONS IN WiiICil TRIP FUNCTION CllECK TEST CALIBRATION SURVEILLANCE REQUIRED 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION t a. RCIC Steam Line Flow-High S H R 1, 2, 3 b. RCIC Steam Supply Pressure-S H R 1, 2, 3 Low r c. RCIC Turbine Exhaust S H R 1, 2, 3 l Diaphragm Pressure-Nigh d. Emergency Area Cooler S H R 1, 2,.3 l g-Temperature - High R e.. Suppression Pool. Area S H. R 1, 2, 3 l [ Ambient Temperature-Nigh

Y f.

Suppression Pool Area AT - S H R 1, 2, 3 l f Nigh [ g. Suppression Pool Area Temp. Timer itelays NA SA R 1, 2, 3 h. Drywell Pressure - High S H R 1, 2, 3 l } 3 i. Logic Power Monitor NA R NA 1, 2, 3 6. SNUTDOWN COOLING SYSTEM ISOLATION a. Reactor Vessel Water Level - S H H 3, 4, 5 Low (Level 3) k. .b. Reactor Steam Dome S H R 1, 2, 3 l R Pressure - High I-a [

E l..$

l l I t 1

a. n

~.a.

w _.. _ - w: ,=... -..

1 INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 The emergency core cooling system (ECCS) actuation instrumentation shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3. APPLICABILITY: As shown in Table 3.3.3-1. ACTION: a. With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperable and place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value. ~ b. With the requirements for the minimum number of OPERABLE channels not satisfied for one trip system, place the inoperable channel in the tripped condition or declare the associated ECCS inoperable within one hour. c. With the requirements for the minimum number of OPERABLE channels not satisfied for both trip systems, declare the associated ECCS j inoperable within one hour. 2 j d. The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION 5. i SURVEILLAN'CE REQUIREMENTS ':a 1 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated DPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.3-1. 4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least cnce per 18 months and shall include calibration of time delay relays and ti:ners necessary for proper functioning of the trip system. 4ATCH - UNIT 2 3/4 3-24

- ~ ' - c-- 'i INSTRUMENTATION SURVEILLANCE REOUIREMENTS (Continued) 4.3.3.3 The ECCS RESPONSE TIME of each ECCS function shown in Table 3.3.3-3 shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one logic train such,that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total. number of redundant channels in a specific ECCS function. 4 .i i 1 I j a HATCH - UNIT 2 3/4 3-25

p _ _ =. _. j, h' TABLE 3.3.3-1 .f . EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUNENTATION N ? -Q HININUM NUMBER APPLICABLE 8j OPERABLE CilANNELS OPERATIONAL TRIP FUNCTION PER TRIP SYSTEH CONDITIONS u 1. CORE SPRAY SYSTEM Reactor vessel Water Level - Low Low Low (Level 1) 2 1,2,3,4,5 a. (2P21-N691A,B,C,D) b. Drywell' Pressure - High (2 Ell-N694 A,B,C,D) 2 1,2,3 c. Reactor Steam Dome Pressure - Low (Injection Permissive) l (2B21-N690A,B C.D) 2 1,2,3,4,5 l d. Logic Power Monitor (2E21-K1A,B) 1/ bus,) 1,2,3,4,5 g 2. LOW PRESSURE COOLANT INJECTION HODE OF RNR SYSTEM { Drywell Pressure - High (2 Ell-N694A,B,C,D) 2 1,2,3 a. b. Reactor Vessel Water Level - Low Low Low (Level 1) 2 1,2,3,4*,5* l 'i' '(2821-N691A,B,C,D) S* Reactor Vessel Shroud Level (Level 0) - High (Drywell Spray !2 c. l Permissive)~(2821-N685A, B) 1 1,2,3,4*,5* d. Reactor Steam Dome Pressure - Low (Injection Permissive) (2821-N690A,B,C,D) 2 1,2,3,4*,5* . Reactor Steam Dome Pressure - Low (Recirc. Discharge Valve i._ e. Permissive) (2B21-N6413,C and 2B21-N690E,F) 2 1,2,3,4*,5* l f. RNR Pump Start - Time Delay Relay 1/ pump 1,2,3,4*,5* e i

1) Pump A (2E11-K70A, 2 Ell-K1258)

2) Pump B_(2EE1-K708, 2El1-K125A)

3) Pump C (2E11-K75B) l t 7f

4) Pump D (2E11-K75A, 2 Ell-X126) g g.

Logic Power Monitor (2 Ell-K1A,5) 1/ bus 1,2,3',4*,5* a" 5 Not applicable winen two core spray system subsystems are OPERABLE per Specification 3.5.3.1. g t (a) Alarm only. lAnen inoperable, verify power availability to the bus at. least once per 12 hours or declare the system inoperable. \\ ~

i.: ~ TABLE 3.3.3-1 (Continued) I.- I !.g. EfERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION L M 't. ,= A MININUM NiflBER APPLICABI.E M OPERABLE CHANNELS OPERATIONAL M TRIP FINICTION PER TRIP SYSTE*1 CONDITIONS # !n ~ -3. NIGN PRESSIRE COOLANT INJECTICII SYSTEM [, I a.. Reacter Vessel Water Level - Low Low (Level 2) 2 1, 2, 3 (2321-N692 A,B,C,5) c, b. Drywell Pressure - Eigh (2 Ell-N694 A,B,C,D) 2( ) I, 2, 3 c. Condensate Storage Tank Level-Low (2E41-Il002, 2E41-II003) 2 1, 2, 3 d. Suppression Chamber Water Level-Nigh (2E41-M6628,D) 2(b)(c) 1, 2, 3 F' l e. Logic Power Hemiter (2E41-EI) 1 1, 2, 3 f. Reacter Vessc1 Water Level-Nigh (Level B) (2321-N693 B,D) 2 1, 2, 3 - l j.'! ! i 4. AUTonATIC DEPRESSIRIZATICII SYSTEN us a a. Drywell Pressere - Eigh-(Fermissive) (2 Ell-N694A,B,C,D) 2 1, 2, 3 f. y b. Reacter Vessel Unter Level - Low Law Law (Level 1)

i

, y (2B21-II691 A,3,C,D) 2 1, 2, 3 W ! b c. ADS Timer (2B2.1-E5A,5) 1 1, 2, 3

d. _ Reacter Vessel Unter Level-Law (Level 3) (Permissive) 1 1, 2, 3 (2B21-II695A,5) e.

Core Spray Pump Discharge Pressure.- Nigh (Perunissive) l (2E21-II6554,8; 2E21-If652A,B) 2 1,2,3 f. RWt (LPCI IIODE) Pump Discharge Pressure - Nigh (Permissive) (2 Ell-II655A,3,C,D; 2 Ell-N6561,B,C,D) 2/leeg*) I, 2, 3 l g. Centrol Power 80emiter (2521-ERA,5) 1/ bus 1, 2, 3 5. LOW LOW SET S/RV SYSTEN k a. Reactor Steam n==a Pressure - Nigh (Permissive) 3 (2B21-II6204,5,C,D) 2 1, 2, 3 5 (a) Alarm only. ttnem inoperable, verify power availability to the bus at least esce per 12 i w beers er declare the system inoperable. f (b) Provides signal to IIPCI pump section valves only. w* - (c) nihen either cheesel of the matematic transfer logic is inoperable, aliga NPCI pump suction to the suppression peel. L,i IIPCI and ADS are met regstred to be OPERABLE with reacter steam done pressure < 150 psig. 1

^ - - ' - -

w..

I . ~ - [, L ? t TARLE 3.3.3-2 f . g EpracwT CORE COOLING STSTEh ACTHATIOlt IIISTRl4ENTAT10N SETPOINTS' ( fj l ? c., y ALLOWABLE TRIP FINICTICII TRIP SETPolNT VALilE I. CORE SPRAY SYSTEN a. Reacter Wessel ideter Level - Law Low Law (Level 1) > -121.5 inches * > -121.5 inches * 'I b. Drywell Pressure - Nigh 31.85psis 31.85psag ~h< c. Reacter Steam Dome Pressure - Law > 422 psig** > 422 psig** l_ d. Logic Pomer Nomiter NA NA 1 2. Im iNiw C00lANT IIEJECTICII IIDGE OF RIIR STSTEN [ { a. Brywell Pressure - Nish $ 1.85 psig i 1.85 Psig b. Reacter vessel lanter 14wel - Law Low Imv (Level 1) > -121.5 inches * > -121.5 inches

Y c.

Reacter Wessel Shrood Level (Level 0) - Eigh I -207 inches

i -207 inches
E

- d. Reacter Steam Dame Pressure-Lew [422psig** [422psig** [l] e. Reacter Steam Emme Pressure-Lew ' ) 325 psig > 325 psig f. RIIR Pump Start - Time Belay Relay )

1) Pumps A, B and D 10 i I seconds

,10 i i seconds [-

2) Pump C 0.5 1 0.5 seceeds 0.5 1 0.5 seconds 3

Imgic Power IIseiter NA NA ? I

See Bases Figere B 3/4 3-1. -

.g " This trip inacties shall be less then er e p t to 500 psig. l 3 t Z cU e a v, g-y _,,,__y. f _, _, _.

2. _ -.

(

6 i TABII 3.3.3-2 (Continued) (- .N t ig E3erarsurY CORE C005.ING SYSTEM ACTtAATION INSTRt9ENTATION SETPolNTS a L. U ALIMdASIE p TRIP F150CTION TRIP SETPOINT VALIE 3. EIGIE Fettser enntmarT IIUECTICII SYSTEM

a. - Beacter Vessel hter Level - Inw Imv (Level 2) 1 -55 inches

3 -55 inches
j b.

Lryw11 Pressure-Nigh i I.85 psig i 1.85 psig t c. r W sate Storage Teatt Level - Law 1 e inches ** 3 0 inch-s** g d. Suppressies (L ar hier Level - Ifish i 33.2 inchea $ 33.2 inches I e. Iogic Power Monfter NA NA g j f. Meacter Vessel hter Level-Bigh (Level 4)* < 56.5 inches < 56.5 inches I i l 4. AlfTOIIATIC BEPMSSWIZATICII SYSTEM t", a. Brywell Iressure-Eigh 1 1.85 psig i 1.85 psig b. Reacter. Vessel hter Level - Law Law Lew (Level 1) 1 -121.5 inches

1 -121.5 inches
l c.

ABS Timer i 120 seconds i 120 seconds y

d.. Reacter Yessel hier Level-Low (Level 3) 1 S.5 inches

1 4.5 inches
i e.

Core Spray Pump Discharge Pressure - Nigh > 130 psig 3 130 psig g. Centre! Power Heatter ~ 1 105 psig 1 105 psig f. EM (LPCI IIDGE) Pump Discharge Pressure - Eigh NA NA 5. LOW LN SET S/RV SYSTEM 1 a. Reacter Steam Dame Pressure - Nigh 1 1054 psig i 1054 psig i N 3 5 .I. F ww C

See Bases Figure 3 3/4 3-1.

~

- Eysavalent to 38,000 gallees of teater in the CST.

\\

.:w:.. ~ -.. ) i TA&LE 3.3.3 3 I ESRCENCY CORE CCOLING $YSTEM RISPONSE TIMIS l ECCS RESPONSE TIS (Seconds) t. COM SPRAY SYSTEM < 27 2. LOW PMSSURE COOLANT INJECTION .100E CP RM SYSTEN 1 40 3. HIGN PMSSURF. COOLANT INJECT!0N SYSTEM i 30 i l 4 AUTOMATIC DEPRESSURIZATION SYSTEN NA 5. ARM LOW LOW SET SYSTEN 11 l l 1 lt l n a l l-f. l l. 1 i 't MATCH UNIT 2 3/!. J 30 Mendment fic. 33 r w

t s.:.L. w a.kx& l. :. -. .. ~,,, :. ... : ^ ,a TABLE 4.3.3-1 f' ENERGENCY CORE COOLING SYSTEH ACTUATION INSTRilHENTATION SURVEILLANCE REQUIREHENTS ,s h. CliANNEL .0PERATIONAL CIIANNEL FUNCTIONAL CIIANNEL CONDITIONS IN WilICll [: y. TRIP FUNCTION CllECK TEST CALIBRATION hURVEILLANCE REQUIRED A ra 1. CORE SPRAY SYSTEM a. Reactor Vessel Water Level - i' Low Low Low (Level 1) S H R 1,2,3,4,5 [ b. Drywell Pressure - liigh S H R 1, 2, 3 [ c. Reactor' Steam Dome i, Pressure - Low S H R 1,2,3,4,5 d. Logic Power Honitor NA R NA 1,2,3,4,5 I u. 2. LOW PRESSURE COOLANT INJECTION HODE OF RilR SYSTEH

)

. w a. Drywell' Pressure - High S H R I', 2, 3 l Jd, b. Reactor Vessel Water Level - l . Low Low Low (Level 1) S H R 1, 2, 3, 4*, 5* i c. Reactor Vessel Shroud Level (Level 0) - liigh S H R 1, 2, 3, 4*, 5* l d. Reactor Steam Dome Pressure'. Low S. H R 1, 2, 3, 4*, 5* e. Reactor Steam Dome Pressure - Low S H R 1, 2, 3, 4*, 5* f. RHP Pump Start-Time Delay Relay NA' NA R 1, 2, 3, 4*, 5* g. Logic Power Honitor ' NA R NA 1, 2, 3, 4*, 5* 1 ot applicable when two core spray system subsystems are OPERABLE per Specification 3.5.3.1. N t-s' 3. v o 0 ti e E L1 t; hi I ti =

~r

a. a u _..

~h TABLE 4.3.3-1 (Continued) EllERGENCY CORE COOLING SYSTEM ACTUATION IN3TRUllENTATION SilRVEILLANCE REQllIREffENTS .m h CllANNEL OPERATIONAL i ',-1 CilANNEL FUNCTIONAL CilANNEL CONDITIONS IN WilICil ' TRIP FUNCTION CliECK TEST CALIBRATION SURVEILLANCE REQUIRED # 3. IIIGli PRESSURE COOLANT INJECTION SYSTEM a. Reactor Vessel Water Level - Low Low (Level 2) S fi R 1, 2, 3 i b.' Drywell Pressure-Iligh-S H R 1, 2, 3 c. Condensate Storage Tank Level -

Low NA H

Q 1, 2, 3 d. Suppression Chamber Water Level - Iligh-S H R 1, 2, 3 l e. Logic Power Monitor NA R NA 1, 2, 3 f. Reactor Vessel Water Level-liigh S H R 1, f, 3 R (Level 8) n '4. AUTOHATIC DEPRESSURIZATION SYSTEM w a. Drywell Pressure-liigh S H R 1, 2, 3 l b. Reactor Vessel Water Level - l Low Low Low (Level 1) S H, R 1, 2, 3 l c. ADS Timer NA NA R 1, 2, 3 d. Reactor Vessel Water Level - Low S H R 1, 2, 3 (Level 3) c. Core Spray Pump. Discharge Pressure - liigh E H R 1, 2, 3 l g f. RilR (1.PCI HODE) Pump Discharge g g Pressure - liigh S H R 1, 2, 3 l t g g. ' Control Power Honitor NA' R NA 1, 2, 3 m. E 5. LOW LOW SET S/RV SYSTEM lO a. Reactor Steam Dome Pressure - p fligh S H R 1, 2, 3

llPCI and ADS are not required to be OPERABLE with reactor steam donc pressure < 150 psig.

l i t e

a ..~... _ \\ z1 : f. I INSTRUMENTATION 1 3/4.3.4 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION f _IMITING CONDITION FOR OPERATION 3.3.4 The reactor core isolation cooling (RCIC) system actuation instru-nentation shown in Table 3.3.4-1 shall be OPERABLE with their trip set-l', points set consistent with the values shown in the Trip Setpoint column of Table 3.3.4-2. l APPLICABILITY: CONDITIONS 1,.2 and 3 with reactor steam dome pressure ^ > 150 psig. i a ACTION: a. With a RCIC system actuation instrumentation channel trip set-point less conservative than the value shown in the Allowable Values column of Table 3.3.4-2, declare the channel inoperable and place the inoperable channel in the tripped condition until the channel is restored to OPERABLE status with its trip set-l point adjusted consistent with the Trip Setpoint value. 4 ,3 b. With the requirements for tne minimum number of OPERABLE channels not satisfied for one trip system, place the in-operable channel in the tripped condition or declare the RCIC ]. system' inoperable within one hour. c. With the requirements for the minimum number of OPERABL-E channels not satisfied for both trip systems, declare the RCIC system inoperable within one hour. 4

]

SURVEILLANCE REQUIREMENTS o lh s l 4.3.4.1 Each RCIC systen actuation -instrumentation _ channel shall be. d demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL. FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies. shown in i Table 4.3.4-1. + A 4.3.4.2 ~ LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation; j of all channels shall'be-performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip system. v 1 l HATCH - UNIT 2 3/4 3-33 j i + m +u, y. jy

...,,: a... ~ ~ - N N TAntE 3.3.4-1 td I le'AcitXI ODIE ISOLATION 000LIFG SYSI'EM ACitmTION INS 11GENTATION Ny MINI >tN NLMER OF OPEllAIN.B OINCEIS FUNCTIOW. UN!'IS ETJI '1111P SYSEM a. Iteactor Vessel Water Invel - ImW Low (Level 2) .2 (2B27-N692 A, B, C, D) b.- Cotylensato Storage Ta:A 2(a) Water Level - Inw (2E51-N060, 2E51-H061) y c. Sigx>ression Pool Water 2(a) y Level-liigli (2E51-N062A, I1) 4 (a) I'ruvides Signal to ICIC lup Suction Valves Only N B a rt h f N t

..1:. J r_1.D.. L - _ ; .L a _.., ,._ i. l TMitE 3.3.4-2 . N, i i, Q l4F>CIOil WlE ISOI ATION 000011C SYSflN AClimTI0tl INS 111tM}irATION q ALIDSIlLE l i EUtCPIONAL int'IS TRIP SETITOINP VAll18 so a. Reactor Vessel Water level - tow Low (Level 2) 2. -55 inches

h-55 incien*

l b. Condensate Storage Tank revel - Iow 2 0 inctes** 10 incies*

Suppression Pool Water Level-tiigli

$ 151 inches [151 inches c.

See Bases Figure B 3/4 3-1 I

'a 'this correspoids to a level of 131'-0" above mean sea level. L l> u, .s I .i h 1, .g i [ r 4, l \\ \\ t1 4' 1

...._.m.:.. .j,

l..

I I .b'i TMll.E 4.3.4-1 O '= HEIClut ODIE ISOIATI0li GXX,ltKi SYSI'EH /CIVATION INS 111tM2irATIQ1 SUlWEIIJMCB lilUlllGEITIO t TJ r1

CIIANNET, H

CllMREI, EUtCTIONM,

OlAIREI, FlRCrIONM, UN11S CllHCK 11EP CM.IBl1ATION i

i a. Iteactor vessel Water level-S H R Iow I,ow (Level 2) l l b. Condensate Storage Tank NA M g Imvel-tow v. 2 c. Suppression Pool Water NA M g level-liigli u l, h h 1r a 3-O ~ Ini 4 L "

r- _.as;

.R._ u....a

_.. c.. _ n

_.. _;-.2

. a n~12

~ TABLE'3.3.5-1 (Continu-d) CO.NT40L ROD VI'I"'DRAVAL BLOCK INSTRUMENTATION NOTE a. When THERMAL POWER exceeds' the preset power level of the RWM and RSCS and when the limiting conditi'on defined in section 3.1.4.3 exists. b. "Tnis function is bypassed if detector is reading > 100 cps'or the IRM channels are on range 3 or higher. i c. This function is bypassed when the associated IRM channels are on range 8 or higher. d. A total of 6 IRM instruments must be OPERABLE. This function is bypassed when the IRM channels are on range 1. e. f. With any control rod withdrawn. Not applicable to control rods removed per. Specification 3.9.11.1 or 3.9.11.2. -1 'i .j i 1 4 N ,1

4

'j. i; 5 -S 4 I .4

r I

RATCH-2 3 /t. 3-39 JAmendment No. 39 [" %.a'

,.-u...., d. TABLE 3.3.5-2 ~ nm CONTROL ROD WITIIDRAWAL BLOCK INSTRUHENTATION SETD0ItiTS p. TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 1. APRM-a. Flow Referenced Simulated l Thermal Power - Upscale s (0.58 W + 50%)(a) 5 (0.58 W + 50%)(a)- h. Inoperative NA NA c. Downscale 23/125 of full scale 2 3/125 of full scale d. Neutron Flux - liigh,12% 512/125 of. full scale-s 12/125 of full scale. 2. ROD BLOCK HONITOR Upscale (h) a.

1) Low Trip Sctroint (LTSP) 5115.1/125' of full scale

$ 115.5/125 of full scale i-

2) Intermediate Trip Setpoint (ITSP) 5109.3/125 of full scale 5109.7/125,or full scale U

3) liigh Trip 'Setpoint (llTSP) 5105.5/125 of full scale 5105.9/125 of full scale i..

h. Inoperative NA NA T-I-" 294/125 of full scale ?93/125 of full scale c. Downscale d. Power Range Setpoint(c)

1) Low Power Setpoint (LPSP)

$27%-of RATED TilERHAL POWER s 29% of RATED TilERHAL POWER ~ 2) Intermediate Power Setpoint (IPSP) .s62% of RATED TilERHAL POWER 564% of RATED TIIERHAL POWER f

3).Iltgh Power Setpoint (IIPSP) 582% of RATED THERHAL POWER 584% of RATED TilERHAL POWER c.

RRH nypass Time Delay $2.0.sec S2.0 sec (td )(d) 2 3g 3. SOURCE RANCE HONITORS. ?i m a. Detector not full in NA NA 5 h.' . Upscale-5 1 x 105 cps 5 1 x 105 cpn j' g c. Inoperative NA NA d. Downscale 23 cps 2 3 cps D. i. L. v

.. N

2..

a _x-. ..u- ' ~ " '~ i ig ' TABLE 3.3.5-2 (Continued) M. A CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETDOINTS i

n j:

TRIP FUNCTION, TRIP SETPOINT ALLOWARI.E VALUE 4. INTERHEDIATE RANCE MONITORS a. Detector not full in NA NA b. Upscale s 108/125 of full scale s 108/125 of full scale i c. Inoperative NA NA d. Downscale 2 5/125 of full scale > 5/125 of full scale } 5. SCRAH DISCHARCE VOLUME ',l

s.. - Wa ter-Level-i2igh 536.2 gallons

$36.2 gallons i NOTES : u. ) Y W = Loop recirculation flow in percent of rated flow. u-1 b. There are three upscale trip levels. Only one 'in applicable over a specified operating core thermal power E range. All RBH trips are automatically bypassed below the low power setpoint. The upscale LTSP is applied .between the low power setpoint and the' intermediate power setpoint. The upscale l'ISP is applied between the intermediate power setpoint and the high power cetpoint. The IITSP is app 11 i above the high power setpoint. 't Power range setpoints control enforcement.. of. appropriate upscale trips over the proper core thermal power c. g' ranges. The power signal to the RBM is provided by the APRH. d. RBM bypass time delay is set low enough to assure minimum rod movement while upscale trips are bypassed. S' g-a" O 1 8 e

1 i' f!l iI j,, f i! I!i i !1I D il E CR l I l U i LWQ I A E NNR 0I I E S 1 T5C Afl N R0A 2 5 ))) 555 a 5555 2 E1 L ddd ((I PTL I 1 1 1 2 1 y I 2222 2222 Ol 1 0E NV OR Ci l S S T t f E H C I H ^ l Il f l[ i f 0 E L1 ET l l NA E ! R RARR RgR ARAR ARAR R i l p i~ t l I N C AD i N l l A lCl A I L C I )I E cC'c V MgM (If R I. H, M, H, f WWWW WWWW i l LA o i S_ lf l 0i ) ) ))I )I I) f i O h i I' bh b h ',, h I' I t r I T I g I(I (I III l U 0 ATS 1 U1 UUU 1 l I l U II I ) j 1 l I ! 1 l l CE l/ /// /// //// //// T CNT SSSS SSS SSSS SSSS Q 5 A il I 3 I F 4 Ml] l t l E. l i I S L P. A i EK l T liC l NE K Al l C iC AAAA AAA AAAA AAAA A t 0 C NiNN INN i f ? i l l l l ! l NNll ! N L i l L A I l r A ee 2" R tl D aa S 1 l l c R l us n O n mp h i T i i W i J q l I S a l C i l o d r li S l D 0 N h l 5 R u M u U g ~ ee O f f L i t i cw T E O l a. L no x R I t G t V l O eP e u O e N o e N o e l R r v l T v 0 n v A n v E e T el iei ie 1 i e R e G v I i 2 f i f atl N tl r tl r tl R e t o emaan O eaa E oeaa E oeaa A L C R rrco M l rc G t l. rc T tl rc H L eesr aes N caes A caes C r whpnt K cpn A ecpn ccpn S e I oT owu C sow R t s ow D t s ow l 1 t .r noe 0 pno epno E cpno 0 a 0 l Df l UI D E DUI D M OUI D W l . I F I t, C M _.. B C R T c. l R __ f R E A 0 l T R l l 0 u l'. i t

i...

... 0 C t f A. a hct ah c 5 al cd ah cd S a I w P I R J T 1 2 3 4 5 l9. C.y m . % wk~ rb[M ?* sN i ~

~a 4- .~ a.- -.. ~- +~-s=< ~. TABLE 4.3.5-1 (Continued)~ CONTROL ROD WITEDRAVAL BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS t NOTES: a. Neutron detectors may be excluded from CHANNEL CALIBRATION. b. Within 24 hours prior to startup, if not performed within the previous 7 days. When changing from CONDITION 1 to CONDITION 2,- perform the c. required surveillance within 12 hours after entering CONDITION 2. d. When THERMAL POWER exceeds the preset power level of the RWM and RSCS. The additional surveillance defined in Specification 4.1.4.3 will be required when the Limiting Condition defined in Specification 3.1.4.3 exists. e. With any centrol red withdrawn. Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2.' ? l I p f i. t 4 l l i if' -uATCu ' J/

3. a Amendment No.-39

~ l~

'u=.= ...._..u.-_, INSTRUMENTATION POST-ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATICN i i 1 j 3.3.6.4 The post-accident monitoring instrumentation channels shown in Table 3.3.6.4-1 shall be OPERABLE. 'i ,l l APPLICAB'ILITY : CONDITIOP.S 1 and 2. ACTION: With one or more of the above required post-accident monitoring a. channels inoperable, either restore the inoperable channel (s) to OPERABLE status within 30 days or be in at least HOT Sh1JTDOWN within the next 12 hours. { b. The provisions of Specification 3.0.4 are not applicable.

j

~j SURVEILLANCE REOUIREMENTS s I 4.3.6.4 Each of the above required post-accident monitoring instrumen-tation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies i 4 shown in Table 4.3.6.4-1. l} ~ t .-g .- j + .1 i -] l 4 l HATCH - UNIT 2 3/4 3-53

_..m .,-..,m_-.-- .r.,_..._ l i-p- - p-(,

lC: N' TABLE.3.3.6.4-1

(- . te ],e POST-ACCIDENT HONITORING INSTRilHENTATION H u. HINIHUM CIIANNELS ' INSTRUMENT OPERABLE 1. Reactor Vessel Pressure (2B21-R623 A, B) ~ 2 2. Reactor Vessel Shroud Wat r Level (2B21-R610, 2Bil-R' 15) 2 l 6 I 3. Suppression Chamber Water Level (2T48-R622 A. B) 2 4. Suppression Chamber Water Temperature (2T47-R626, 2T47-R627) 2 R y- .l 5. Suppression Chamber Pressure (2T48-R608, 2T48-R609) 2 g 6. Drywell Pressure (2T48-R608, 2T48-R609) 2 7. Drywell Temperature (2T47-R626, 2T47-R627) 2 8. Post.-LOCA Gamma Radiation (2D11-K622 A, B, C, D) 2 i 9. Drywell H -0 Analyzer (2P33-R601 A, B); 2 i 2 2 10.a) Safety / Relief Valve Position Primary Indicator (2B21-N301 A-H and K-M) I b) Safety / Relief Valve Position Secondary Indicator (2B21-N004 A-H and K-M)

If either the primary or secondary indication is' inoperable, the torus temperature will be monitored R

'at least once per: shift to observe any unexplained temperature increases which might be indicative 2 of an open SRV. With both the primary and secondary monitoring channels of an SRV inoperable, S' cither verify that the S/RV is closed through monitoring the backup low low set logic position f . indicators ' (2821-N302' A-H and K-M) at least once per shif t or restore sufficient inoperalile 2 channels such that no more than one SRV has both primary and secondary channels inoperahl,e [ g within 7 days or be in at least hot shutdown within the next 12 hours. i. r y j; e L h:

.......e... h.s ~n ra .e _.m am... . n. s..... 7-~ f' s: x-TABLE 4.4.6.1.3-1 . [':. - -~4 nx REACTOR VESSEL MATERIAL IRRADIATION SURVEILLANCE SCHEDULE- .+, u- ,c 2 SPECIMEN REMOVAL INTERVAL -4 i. 1. 10 years l' 2. 30 years 3. Reserve -I I ( l J w. 1 t f 4= ~ .L. i t i y. m p j.' t w b ^ ( '. 51 4 .). .f f h'- 4 't

-__..-_-n = .) .l 1 l;.

t i

(' REACTOR COOLANT SYSTEM l l REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION l j 3_. 4. 6. 2 The pressure in the reactor steam dome shall be less than 1054 l Psig. .APPLICABILIT1?: CONDITION 1* and 2*. -. ~ -. ACTIQN: With the reactor steam dome pressure exceeding 1054 psig, reduce the pressure to less than 1054 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours. SURVEILLANCE REQUIRLENTS. 't 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1054 psig at least once per 12 hours. '. ) e .j .iq- )-t.;

Not applicable during ant.icipated transients.

il ), L ~ 3/4 4-18 HATCH - UNIT 2 Amendment No. 39.

w:_. A a - ML. ~a ',.L.:Ci?.h-..- 'i a 1 CONTAIN*NT SYS E S SI'RVEmiCE REQUIRE.WNTS (Continued) 3. Verifying a system flow rate of 4000 +0, -1000 cfm during system operation when tested in accordance with ANSI N510-1975. c. After every 720 hours of charcoal adsorber operation by -verifying within 31 days after removal that a laboratory c analysis of a representative carbon sample obtained in accord-ance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision I, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976. d. At least once per 18 months by: 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the filter train at a flow rate of 4000 +0, -1000 cfm. 2. Verifying that the filter train starts and isolation 7}. "- dampers open on each of the following test signals: a. Drywell pressure-high, s b. High ' radiation on the;

1) Refueling floor,

,1

2) Reacto'r building.

(A

1 2

l c} c.. Reactor Vessel k.ter Level-Low Low (Level 2). I if. { 3. Verifying that the heaters dissipate.18.5 + 1.5 KW when tested in accordance with ANSI N510-1975. 4 i l KATCH. ' UNIT 2 3/4 6-41 Amendment No. 39

j. 4' CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) e. After each complete.or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 3- +0, -1000 cfm. f. After each complete or partial replacement of a charcoal adsorber bank by verifying that the _ charcoal adsorbers remove > 99% of a halogenated hydrocarbon regrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 + 0, -1000 cfm. 4.6.6.1.2 Each Hatch-Unit 1 standby gas treatment subsystem shall be demonstrated OPERABLE per Hatch-Unit 1 Technical Specifications. 9 e 'j. 1 i 1 m HATCH - UNIT 2 3/4 6-42

a . -.._ _ i .l_m _ _ _ _ f 1_ -,, _.. _ _ ej 1 RE:UELING 0?! RATIONS 2/4.9.2 INST:UMENTATICN l

! IMITING CONOITICN :0R OPETATICN L

4 htt !!3.9.2 at least 2 sedets est;e moniter- (SF.N enannels snall be OPERAELE .anc insarted to tne nc ~.ai c:erating level: a. Each with continucus visual indication in the control room, b. At least one with an audible alar:n in the control rocm, c. One of the SRM detectors located in the cuadrant where CORE ALTERATIONS are being pe'rformed and the other SRM detector located in an adjacent quadrant, and d. The " shorting links" removed from the RPS circuity during CORE ALTERATIONS and shutdown margin cemenstrations. APCLICAEILITY: CONDITION 5. ACTION: With the requirements of the above specification not satisfied, i::rnediately suspend all operations involving CORE ALTERATIONS" or positive reactivity changes and actuate the manual scram.. The provisions of S;ecification 3.0.3 are not applicable. SURVEILLANCE REQUIREMENT 5 4.9.2 Each of the above required SRM channels shall be demonstrated-OPERAELE by: a. At least once per 12 hours; 1. Performance of a CHANNEL CHECX, l 2. Verifying the detectors are inserted to the normal' operating level, 3. During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located,in the core quadrant i where CORE ALTERATIONS are being performed and one.is i located in the adjacent quadrant. i 'Ine use of special movable detectors during CORE ALTERATIONS in place- ) 1 of the normal SRM nuclear detectors is permissible as long as these

l special detectors are connected to the normal SRM circuits.

.i "Except movement of SRM or special movable detectors, l 4 HATCH'- UNIT 2-3/4 9 3 1

- - - ~ - .;-. a : : .au.1~u..d va.2.. a~. ~ = w.. a.a x.- a ~ ~.. t 4 INSTRIITTATICN SL7?.TILL.NCE FIO*dIFEG5 CCCINLTD b. Perfer. ance of a C*-G.KNEL ID CTICFE TEST: 1. Within 24 hcurs prict to the start of CCRE ALTEP.ATICNS, and 2. At least once per 7 days. c. Verify that the channel count rate is at least 3 cps at least once per 12 hours during CCRE ALTERATICt:S, and at least once per 24 hours, except: 1. The 3 eps is -not required during core alterations involving enly fuel unicading provided the S?Ms were confirmed to read at least 3 cps initially and were checked for neutren respense. 2. Che 3 cps is not required initially en a full core reload. Prior to the reload, up to four fuel asser.blies will be loaded l into their previous core positions next to each of the 4 SRv.s i to obtain the re:;uired count rate. d. Verifying that the RPS circuitry " shorting links" have been removed and that the hPS circuitry is in a non-coincidence trip mode within 8 hours prior to starting CCRE AL E TICNS or shutdown margin denonstrations. ? 1 h k 4 HATCH - LUIT 2 ~ 3/4-9-4 Amendment No. 4,39 a.

.a: - uB & ' =.u. = =. -. :. a..: 2._...-. s-

au..

2: i REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued). than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control sP rods even under the most unfavorable depressurization of the reactors. Control rod coupling integrity is required to ensure compliance with-the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that.a rod is properly coupled and therefore this check must be performed prior to achieving criticality after each refueling. The subsequent check is performed as a backup to the initial demonstration. In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE. The control rod housing support restricts the outward movement of a control rod to less than (3) inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is i not required when there is no pressure to act as a driving force to rapidly eject a drive housing. 4 The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the' system components. 3/4.1.4 CONTROL R00 FROGRAM CONTROLS ^ Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to cause the peak fuel enthalpy for any postulated control rod accident to exceed 280 cal /gm..The specified sequences are characteriTed by homogeneous, scattered patterns of control rod withdrawal..When 4 THERMAL POWER is > 20% of RATED THERMAL POWER, there is no possible rod worth which,-if dropped at the design rate of the velocity limiter, .could result in a peak enthalpy of 280 cal /gm. - Thus, requiring the RSCS and RWM to be OPERABLE below 20% of RATED THERMAL POWER provides adequate-control. t . HATCH - UNIT 2 B 3/4 1-3 .~

. _._ w.a g .2 QcMR.l a.a.L-- : - =. a.l K~i~wk I' REACTIVITY CONTROL SYSTEM .t. . BASES CONTROL ROD PROGRAM CONTROLS (Centinued) 5 The RSCS and RVM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted. p . The analysis of the rod drop accident is presented in Section 15.1.38 of'the FSAR and the techniques of the analysis are presented in a topical report,. Reference 1, and two supplements, References 2 and 3. s i The RBM is designed to auto =atically prevent fuel damage in the event n of erroneous red withdrawal from locations of high power density during-high power operation. The RBM is only required to be operable when the Limiting Condition defined in Specification 3.1.4.3 exists. Two channels are provided. Tripping one of the channels will block erroneous rod with-drawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for. withdrawal of control rods. Further dis-cussion of the RBM system and power dependent setpoints may be found in NEDC-30474-P (Ref. 4). 3/4.1.5 STANDBY LIQUID COhTROL SYSTEM The standby liquid control system provides a backup capability' for p maintaining the reactor suberitical in the event that insufficient rods ~ are inserted in the core when a scram is called for. The volume of the poison solution and weight percent of poison ~ material in solution is based on being able to bring the reactor to the suberitical condition as 'the plant cools to ambient conditien.' The temperature requigement is necessary to. keep the sodium pentaborate in solution. Checking the volume and temperature once each 24 hours assures that the solution' is-av'ailable for use. With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant - components inoperable. .j Surveillance requirements are established on a frequency that assures i a high reliability of1the system. Once the solution is ; established, bcron ~. concentration will not vary unless more boron water is added; thus, a 2 check en the temperature and vclume once sach 24 hours assures that the solution is available for use. 4-n EbCk-L B.3/4 1-4 Amendment'No. 39 -p i

.l-- . r Ame -x- ._m_o,1;i _. _ u. m. :,.. ..._2c2..,,n = can nessa_ LREACTIVITY CONTROL SYSTEM ,J -l.

BASES i

STANDBY LIQUID CO.NTROL SYSTEM (Continued) Replacement of the explosive charges in the valves at regular intervals twill assure that these valves will not fail because of deterioratien of the charges. 4 l S I - 1 a .[ 1. C. J. Paone, R. C. Stirn and J. A. k'oodley, " Rod Drop Accident Analysis i for Large Bk'Rs," GE Topical Report NEDO-10527, March 1972. j.a 2. C J. Paone, R. -C. Stirn.and R. M. Yound, ' Supplement 1 to NEDO-10527, l ; July-1972.

- ij '

3. J. A. Haum,=C. J. Paone and R. C. Stirn, Addendum 2. " Exposed Cores," gi Supplement 2 to NEDO-10527, Januar-/ 1973. 4 " verage Power Range Monitor, Rod Block Monitor and Technical Specifi-cation. Improvement (ARTE) Program for Edwin I. Hatch Nuclear Plant,- r Units 1 and 2," NEDC-30474-P,' December 1983. HATCH-2 5 3/4 1-Aa: Amendment No. 39 ' ~ I .. + -

__.m _ 3/4.2 POER DISTRIBUTICN LIMITS BtJE!i te specifications of this section assure that the peak cladding taperature following the postulated design basis loss-of-coolant accident will not exceed the 22000F limit' specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated erreC S of ruel penec densification. Base specifications also assure that fuel design margins are maintained during abnormal transients. i ..i .3/4.2.1 AVERAGE PIAL %R LINEAR HEAT GMERATICN RATE eis specification assures that the peak cladding temperature following the postulata$ design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50, Appendix K. he peak cladding taperature (PCI) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel.asser.bly at any axial location and is dependent only secondarily on the rod-to-red power distribution within an assably. The peak clad teparature is calculated assming an UiGR for the highest powered rod which is agual to or less than the design LBGR corrected for densification. 21s UER times 1.02 is used in the heetup code along with the exposure ,I dependent steady state gap conductance and rod-to-rod local peaking factor. he Technica' specificatica APLHCR is this IRGR of the highest powered rod 4 divided by its local peaking factor. Se limiting value for APIRGR is shown !l in the figures for in Technical specification 3/4.2.1. j Be calculational procedure used to establish the APIRGR shown in the i figures in Technical Specification 3/4.2.1 is based on a loss-of-coolant accident ansysis. Se analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CER 50. A cmplete discussion of each code uployed in the analysis is presented in Reference 1. Differences in this analysis compared to previous analyses perfomed with Reference 1 ares (1) the analysis assmes a fuel d assably planar power consistant with 102% of the MAPIRGR shown in the figures in Technical Specification 3/4.2.1; (2) ~ fission product decay is emputed 1 - assuming an energy release rate of 200 MEV/ fission; (3) pool boiling is 4 assmed after nucleate boiling is lost during the flow stagnation period; and _j (4) the effects of core spray entralment and counter-current flow limitation j, as described in Reference 2, are included in the reflooding calculations, t s: A flow dependent correction factor incorporated - into Figure 3.2.1-9 is j applied to the rated conditions APIRR to assure that the 22000F PCT limit is emplied with during a ICCA initiated fra less than rated core flow. In addition, other power and flow dependent corrections given in Figures 3.2.1-10 a i 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 frem off-rated conditions. A list of the significant plant input permeters to the. less-of-coolant accident analysis is presented in bases Table B 3.2.1-1. Further discussion of the APIRGR limits is given in Reference 4. HATH-2 B 3/4 2-1 Amendment No. 27, R$, 23, 1 39

$l:.1JL..bAYE ~r .- + L_ -. _., i.., n...a. 1. . :.a .a.. -..ill.A.'.$.' $55 d a Bases Table S 3.'2.1-1 SIGIFICANT fjpUT P AUMETUS TO T*:E. LOSSMF-COOLViT ACCICENT ANALYSIS FOR HATCM-UNIT 2 Plant Parameters: Core Thermal Power............... 2531 Mwt which corresponds to 105 of license core power-3 Yessel Staam Out;ut.............. 10.96 x 10 lbm/h wMen l corresponds to 105 of rated steam flow Yessel Staam Ocme Pressure....... 1055 psia i Design Basis Recirculation Line Break Area For: a. Large Breaks............ 4.0, 2.4, 2.0, 2.1 and 1.0 ft i 2 b. Smal l S rea ks............ 1.0, 0.9, 0.4 and 0.07 ft Fuel Parameters: ~ INITIAL ' PEAK TECHNICAL SPECIFICATION DESI@ MINIMUM l' LINEAR HEAT AXIAL CRITICAL FUEL. BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kw/ft) FACTOR RATIO i Initial Core 8x8 13.4 1.4 1.18 l } A more detailed list of input to each model and its source is presented in Section II of Reference 1 and subsection 6.3.3 of the FIAR. l 'This power level meets the Appendix K recuirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification linear heat generation rate limit.' t 4 I: HATCE - UNIT 2 3 3/a 2-2 l

JWfw.OQy': ng-

~ :.= c ;- ' ~ ~=

-"2==' = '"

2 *

~ * - I BASES 3/4.2.2 APRM SETPOINTS This=seeticn delete'd. 3 e 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 established fuel cladding integrity Safety Limit MCPR of 1.07, and an' analysis of abnormal . operational transients. For 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 that the fuel cladding integrity Safety Limits are not exceeded during any anticipated abnormal operational transient, the most limiting 4 transients have been analy. sed to determine which results in the largest reduction in CRITICAL POWER RATIO (CPR). The type of. transients evaluated ' were loss of flow, increase in pressure and power, positive reactivity J insertion, and coolant temperature decrease. .)'I l 1 'k i MATCH-2 B 3/4 2-3 Amendment No. 7A, 27,- 39 =

fww a ~ w - ~ a --w. ~; .a m. ;x x. .m a u. POWER ~ DISTRIBUTION LIM 77S .DASES-MINIMUM CRITICAL POWER RATIO (Centinued) The evaluation of a given transient begins with the system initial parameters shewn in FSAR Table 15.1-6 that are input to a GE-core dynamic -behavior transient computer program described in NEDO-10802'88 Also, the void reactivity coefficients that were input to the transient calculational procedure are based on a new method of calculation termed NEV which provides a better agreement between the calculated and plant instrument power distributions. The cutputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic SCAT code described in ~ NEDO-20566 ' ". The principal result of this evaluation is the reduction in.MCPR caused by the transient. The pu. pose of the MCPR, and the K of Figures 3.2.3-4 and 3.2.3-5, re-f p spectively 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 MCPH and MCPR at the existing core flow and power f state. The MCPR s are established to protect the core from inadvertent core f 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 h the corresponding THERMAL POVER along the 105% of rated steam flow control l line, the limiting bundle's :alative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRs l, were calculated at different points along the 105% of rated steam flow control ? line corresponding to different core flows. The calculated hCPR 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 operating limit multiplied by the K factor given in Figure 3.2.3-5. p 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 j K s were determined based upon the most limiting transient at the given core power P -j level. For further information on MCPR operating limits for off-rated conditions, ef erence

NEDC-30474-P. "'

It h h a li HATCH-2 B 3/4 2 4 , Amen'dment No. 33,39-L

N ~ ^~pDVERDIN_5Y5TI'ONL1 HITS ~~ ~ ~' ' l I BASES . MINIMUM CRITICAL POWER RATIO (Continued) \\ i I At THERMAL POWER levels less than or equal to 25% of RATED THERMAL EdkT.R. 'the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial startup testing of the' plant, an MCPR evaluation will be made at 25*. of RATED THERMAL POWER with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR above 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significaut power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, 1,' regardless of magnitude that could place operation at a tnermal limit. 'i 2 3/4.2.4 LINEA't HEAT GENERATION RATE The LHGR specification assures that the linear heat generation rate in any rod is less than the design linear heat generation even if fuel pellet densification is postulated. ') l, a 4 li: i 1

l HATCH-2 3 3/4 2-3 Amendment No, U,39 q.

1

i

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x..:. a. :c m,. x .,. w se...w ~ j* pot.TR DISTRIBUTION LIMITS _.... BASES l t-

References:

1. General Electric Cc:pany Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K NEDO-20566 (Draft), August 1974., [ 2. General Electric Refill Reflood Calculation (Supplement to SATE 1 Code Description) transmitted to USAEC by letter, G. L. Gyorey to V. Stello, Jr., dated December 20, 1974.

p>

3. R. B. Linford, Analytical Methods of Plant Transient Ev.aluations for the GE BWR, February 1973 (NEDO-10802). 4 '.' Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Edwin I. Hetch Nuclear Plant, Units 1 and 2," NEDC-30474-P December 1983. I i j b J f EA70H*' !3/426 Amendment No.39 ? 4 x .s } .u

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7!!N 5'M

..e.. u.r.ya-. y (...... 4.~... e ) ' u.y.r .s,- _ w .s.e...i.u_g.-.__.y ( 4-....) .m . - ~ _. n :.s ev e.-: ma: a :::::i:n :f me f ' -* - ---' n Ms: =entati:n is 1.=ers:le, in::sssing t.s f : uency Of firs :st::is b :ne affe::ad armas is 3-.4

;;=vi, -...

4 n - _-_ =-e t s ty until me ine e.=is a re r_ne..,2,,1:n v is :25:::s= :: 6. i.AE7.1 TY. 4( 3/t. 3.7 TU 5 NE CVE:.::.--m N :.:. iCN Sb H This s===ifi:sti=n is ';=vicec :: ensurs that the ::.:=ine evers=eed ...=: i:n system bsc menta-'-n and me :.==ine scaec c:n:=1 valves a:s C. -.::' an: will ..a=: = a t: :1.e f =m ex sssive evers=eed. N ::m:-' n f=m.:.==ine ex sssive evers=eec is := ->' nc sines ex--eive evers:eec cf t.e tu:: ire =ub generata p=tentially camaging m'"as vni:n - " i:cac: and mage safety-r:1stac '====nents, e- '--=a.: c= st:.:=..=es. I P 3/A.3.3 CEJR CED E 7-N VC! T G 88ui:.u..uN INS-"RUENT: TION The unearvcitats relays sna11 au:=mati= ally -hitita tne ::is. w_:1:n cf effsita ::cwer s :::ss vnenever t. e vcitage se::ci.v anc t'.me a= lay -'t.s have j,

een :::se ec.

This ac:icn sna11 p : vise vcitage p :tec:1:n f:: :.e emergency

i

wer sysx.s by :::= venti. g sustained cegracec vcitage =nciti=:.s cue t= tne Offsitz ::wer sm and int.=:acti:n between :ne effsita anc ensi.a eme:gan=y

==wer sys:s=s. The uncerveltage :alays have a ti:ne celay =a.ac.a:is 1: tha - ::= vices = ta :1:n against t= M a less cf ' voltage and ceg:acac voltags

=:iti n an :.us ain'-'-== the effect. of sne:. cura-' n cistu=an==s witn=ut ex::eci..; :ne =axiram ti:ne calay, i. :.lucing ma -in, t.a is assumed in me

.3;;t ---' w.: analyses. q j>. 4 ' t, T, i t. ,4 MATCH-t.' NIT 7 m ip. 4. N. :: enc =en:.40. u 1 s. i i i 't I i 1 1 )I l w .2 1 .1-

- L-u w-- +' .e +

- T NOTE: SCALE IN INCHES A80VE VESSEL ZERO I

l WATER LEVEL NOMENCLATURE HEIGHT ABOVE VESSEL ZERO 800 - NO. IlNCH E' ' REAOING INSTRUMENT (8) 57 +56.5 BARTON ~ N/MAC 750 - (7) 559 +42 G (4) 549 v32 G E/MAC 722.75 FLANGE (3) 525.5 +8.5 ~ BARTON (2) 462 -55 BARTO.N 700 - - til 395.5 -121.5 BARTON (0) - 310.0 -207 BARTON 650 " MAIN STE AM -- 640 - LINE / 6M - - 577 { K [ 8) 40 - +60 - 563 40 - +40 - - 7 (8) g (7) 42 HI ALARM 550 -= 54 9 44) (l 2 LO ALARM 80TTOM OF STE AM - 525.5(3) 7 RIPS 0 - 0 h 8.5-(3) 8J LOW (LEVEL 3) (, (3) -ORYER SKIRT 517 INSTRUMENT 0-- 0== REACTOR ZERO MMBNO AK N l FEED _4g3,3 SCRAM - -55 LOW LOW (LEVEL 21 WATER _ CORE - 462(2) SPRAY F INITIATE HPCI, RCIC 450 - - TRIP RE. CIR C. PUMPS l 400 - - Q95J(1) - -121.5 LOW LOW LOW (LEVEL 1) a 3g7 - _t$3. .. INITIATE RHR, CA, I 350 -r_352.56 MAM OIESR ANO CONTRIBUTETO A.D1 l . CLOSE MSIV's 2/3 CO RE -2C 7 HEIGHT 310(0) P ERMtSSIVE 300 - (LEVEL 01 ACTIVE FUEL 250 - - t< i -317 - - 200 = 208.56 } RECIRC q - 178.56 DISCHARGE RECIRC ll SUCTION - 151.5 NOZZLE NOZZLE 150 - - o

i i

100 - - ( a. 50 - - 0. r 9380 2 BASES FIGURE B 3/4 31 . REACTOR VESSEL WATER LEVELS 4 Amendment No. 39, 33, 39 33/4 3-6

,~ =- . ~ -...-. -..., ~.. l ).. 'x 3/4.9 REGLD'O CPERATICNS - EASES 3/4.9.1 REAC"CR MCC2 SWITCH i i . Lcc'<ing the CPEFJd5C reacter mcde switch in the refuel pcsition ensures that the restricticns en red withdrawal and refueling pistform =c re ent j during the refueling cperaticns -are pr:perly activa:ed. These ccnditiens reinforce the refueling precedures and raduce the probacility of inadver ant .~ criticality,. damage -the reacter internals or ~ fuel assablies, and expcsure of personnel'to excessive radioactivity. 3/4.9.2 INSWLMENTATION The CPEFaBILITY of at least two source range monitors ensures that redundant monitoring capability is available to detect changes in the reactivity conditien of the core. Curing the unicading, it is not necessary to maintain. 3 cps because core alterations will involve only reactivity ) reoval and will not result in criticality. The. loading of up to fcur bundles around the SFMs before attaining the 3 cps is permissible because these bundles were in subcritical ccnfiguratien when they were removed and therefore will renain suberitical when placed back in the previous pcsitions. + + 3/4.9.3 CCNIROL RCD POSITION ,~ De requirment that all control rods be - inserted during CCRE ALTERATICNS ensures that-fuel will not be -loaded. into a cell without a centrol rod and prevents two positive reactivity changes frcm occurring simultaneously. 3/4.9.4 DECAY TIME 'Ihe minimum requirenent for reactor.suberiticality prior to fuel-i movement ensures that sufficient time has elapsed to allow the radicactive decay of the.short lived fission products. - This decay time is consistent with the assumptions used in the accident analyses.. l 3/4.9.5 SIXENDARY CCITIAIMETI Q'1 ~ ' Secondary contalment is designed to minimize -any ground level releaFae of radioactive material ' which may result from - an accident. The reactor f building provides secondary contaiment during normal' cperation when - the

4 drywell is sealed and in service. When, the - reactor is shutdown or during i

refueling, the drywell may be open and the reactor building then becmes the-1 primary contaiment. 'Ihe refueling floor is _ maintained under _ the secondary i 4j i contaiment integrity of Hatch-Unit -1. [ Establishing and maintaining a vacuum in ~ the' building with 'the standby gas treatment systa -once per 18 months, along with the. surveillance of the i.. . doors, hatches and dampers,) is : adaquate.to ' ensure ~ that - there are no; ~ violatiens of the integrity of the secondary contaiment. Only; one ' closed l . damper in'each penetration line is required to maintain the integrity of-the secondary contaiment.. - HAn - UNIT 2 B 3/4 9-1 Amendment No,; 4,39 ' l g .._,6 .4,

.j ^ -~ ~ __l q j REFUELING CPERATICNS BASES 3/a.9.5 COMMUNICATIONS The recuirement.for cennunications c:cability ensures that refuelin,g station personnel can te promptly informed of significant changes in the facility status or core reactivity condition during movement of fuel within the reactor pressure vessel. CRANEANDHOISTOPERAS!LhiY 3/4.9.7 The OPERASILITY requirements of the cranes and hoists used for movement of fuel assemblies ensures that: -(1) each has sufficient load capacity to lift a fuel element, and (2) the core internals and pressure vessel are protected from excessive lifting force ir. the event they are inadvertently engaged during lifting operations. 3/4.9.8 CRANE TRAVEL-SPENT FUEL STORAGE PCOL The restriction on movement of loads in excess of the nominal weight of a fuel element over irradiated fuel assemblies ensures that no more than the contents of ona fuel assembly will be ruptured in the event of a fuel handling accident. This assumption is consistent with the activity release assumed in the accident analyses. 3/4.9.9 and 3/4.9.10 WATER LEVEL-REACTOR VESSEL AND WATER LEVEL-SPENT FUEL STORAGE FOOL The restrictions on minimum water level ensure that sufficient water ~ depth is available to remove 99% of the assumed 10t fodine gap activity released from the rupture of an irradiated fuel assembly. This minimum water depth is consistent with the assumptions of the accident analysis. -) i 3/4.9.11 CONTROL' ROD REMOVAL i This specification ensures that maintenance or repair of control rods 4 or control rod drives will be performed under conditions that limit the The requirements.for simultane6us 1 probability of inadvertent criticality. removal of more than one control rod are more st1ringent since the SHUTDOWN MARGIN specification provides for the core to remain subcritical with only one control rod fully withdrawn. HATCH - UNIT 2 .B 3/4 9-2 l 3 ~ .-_.---u, .y,

-a . i%. c. - *+: _ i..z '. : t _. y OESIG' ET IS CSTRC' FCD ASSD"ELIES 5.3.2 The reacter ccre shall centain 137 crucifer =-shaped centrol red i . ass eclies. l

5.4 FIACIO

CCCLA2.7 SYS"I4 C :. SIGN ??IS5" I AND TD_:53A'n;FE - 5.4.1 The reactor coolant systen is designed and shall be maintained: a. In accordance with the code requirements specified in Section 5.2 cf the ISAR, with allcwance for nepal degradatien pursuant to the applicable Surveillance Requirements, 1' b. For a pressure of 1250 psig, and c. For a temperature of 5750F LCLG2 5.4.2 The total water and steam volune of the reacter vessel and

j recirculation system is approximately 17,050_ cubic feet at. a ncrninal Tay, of 5400F.

5.5 PSIECRCIDGICAL '1GER IOCATION 5.5.1 The metecrological tower shall be located as shown on Figure 5.1.1-1. 5.6 FUEL S'IORAGE CRITICALITY

. c 5.6.1 - ne new and spent fuel storage racks are designed and shall be i

i maintained with sufficient center-to-center distance between fuel assemblies

t placed in the storage racks to ensure a k,ff aguivalent to $ 0.95 -when l}

fic,oded with unborated water. ne k,ff of 5 0.95. includes conservative

j allowances for uncertainties.

1 i .i P s HATCHI-UtlIT 2 5-3 Amendment-No.39' 4 t 9 n n T N T s- ?M 1 t

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Li s

5.5." Fuel in the 5:en-vel Pool shall have a maximu:- fuel leading of 15.2 crams of Uranium-235 :er axial centimeter. 1 .. a C..... e..... L. N. - e e., e..... C.,. n... ...a..i 1 i . -... ::. i l _.ir.e cocronents icer.:1:1ec in iable o. 7.1-1 are c.es,.gned and sha,ii i:./.1 be cair.:ained wit.in the cyclic or transient lini:s of Table 5.7.1-1. E u. 4 h. O

l

.t 1 4 Amendment No. 15 f Lt., - 2 .e-- .ii iL. ..}}

v t e Letter Sequence Other
TAC:54169, (Approved, Closed) TAC:54607, (Approved, Closed) TAC:62029, (Open)
Initiation Request, Request, Request, Request, Request, Request Acceptance... Administration Questions Answers Results Approval, Approval Other: ML20079P706, ML20087Q011, ML20090G560, ML20090G568, ML20091E083, ML20092A243, ML20092P177, ML20092P588, ML20093C995, ML20095L022, ML20140C660, ML20197H659
MONTHYEAR1984-06-07 [Table View]

ML20090G560