Text

Amends 168 & 106 to Licenses DPR-57 & NPF-05,respectively, Replacing Values of cycle-specific Parameter Limits in core- Related Specs W/Ref to Core Operating Limits Rept
	ML20005E572
Person / Time
Site:	Hatch
Issue date:	12/29/1989
From:	Matthews D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20005E573	List: ML20005E572; ML20005E574;
References
	NUDOCS 9001080183
	Download: ML20005E572 (66)
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{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _... _... _......... ousg%, ; . UNITED STATES ?> NUCLEAR REGULATORY COMMISSION E ~[ WASHINGTON, D. C. 20666 %,.....-f GEORGI A POWER COMPANY OGLETHORPE POWER CCRPORATION - MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA-CITY OF DALTON, CEORGIA DOCKET 00, 50-321 EDWIN 1. HATCH NUCLEAR PLANT, UNIT NO.1 - AMEFDMENT TO FACILITY CPERATING LICENSE Amendraent No.168 License No. DPR-57 l '. The Nuclear Regulatory Coma.ission (the Comission) has found that: A. The application for ar.endment to the Edwin I. Match Nuclear Plant,. Unit 1 (the facility) Facility Operating License No. DPR-57 filed - by Georgia Power. Conpany, acting for itself, Oglethorpe Power Corporation, Municipal Electric. Authority of Georgia, and. City of Dalton, Gecrgia (the licensee) dated June 22, 1989, as airended Culy 31, 1989, and October 4, 1989, complies with the standards and requirements of the Atomic Energy Act of'1954, as amended (the -Act), and the Commission's rules und regulations set forth in 10 CFE Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. Thereisreasonableassurance(i)thattheactivitiesauthorizedby this an:enoment can be conducted without endangering the health and safety of the public, and (ii) that such cctivities will be conducted in conpliance with the Commission's regulations set forth in 10 CFR Chapter I; U. The issuance of this aniendment will rot be ininiical to the comon defense ano 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 Conaission's regulations end all applicable requirements beve been satisfico. 9001080183 891229 PDR ADOCK 05000321 p PDC

d 2. ' 2. Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby anended to read-as follows: -Technical Specifications The Technical Specifications contained in A)pendices A and B, as revised.through Amendment No.168, are here)y incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of its date of-issuance and shall be implemented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COMMISSION p u N o.z L > l 4 David B. flatthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: December 29, 1989 I I ,/ i I l. t l i l i ll L le 8

7 y e ATTACHMENT TO LICENSE AMENDMENT NO.168 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 . Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.- The revised pages are identified by amendment number and - contain vertical lines indicating the areas of change. Remove Pages Insert Pages i iii iii x x xi { 1.0-7 1.0-7 1.1-6 1.1-6 1.1-7 1.1-7 1.1-8 1.1-8 3.3-5 3.3-5 3.3-6 3.3-6 3.3-7 3.3-7 3.3-9 3.3-9 3.3-15 3.3-15 3.3-16 3.3-16 3.3-18 3.3-18 3.6-22 3.6-22 3.11-1 3.11-1 3.11-la 3.11-la 3.11-2 3.11-2 3.11-2a l 3.11-3 3.11-3 l 3.11-4 3.11-4 l 3.11-4a 3.11-4a Figure 3.11-1, Sheets 1 through 8 Figure 3.11-3 Figure 3.11-4 Figure 3.11-6 6-15d 6-15d 1

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{ l __ LIMITING CONDITION 5 FOR D'PERATION SURVEILLANCE RE0VIREMENTS 3.3. REACTIVITY CONTROL (CONT') 4.3. REACTIVITY CONTROL (CONT') G. Limiting the Worth of a G. Control Rod Below 105 l Limiting the Worth of a 3.3-5l Control Rod 8elow 105 Rated Thermal Power Rated Thermal Power H. Shutdown Requirements 3.3-7 3.4 STAND 8Y L10U10 CONTROL SYSTEM 4.4 STANDBY L10VID CONTROL SYSTEM 3.4-1 A. Normal System Availability A. Normal Operational Tests 3.4-1 B. Operating with inoperable Components B. Surveillance with InoperaDie Components 3.4-2 C. Sodium Pentaborate Solution C. Sodium Pentaborate Solution 3.4-2 0. Shut down Requirements 3.4-3 3.5. CORE AND CONTAINMENT COOLING SYSTEMS 4.5. CORE AND C'ONTAINMENT COOLING SYSTEMS 3.5-1 A. Core Spray (CS) System A. Core Spray (CS) System

3. 5 -1.

B. Residual Heat Removal (RHR) B. Residual Heat Removal (RNR) System (LPCI and Containment 3.5-2 Cooling Mode) System (LPCI and Containment Cooling Mode) C. RHR Service Water System C. RHR Service Water System 3.5-5 0. High Pressure Coolant Injection D. High Pressure Coolant in-(HPCI) System 3.5-6 jection (HPCI) System E. Reactor Core isolation Cooling E. Reactcr Core Isolation (RCIC) System 3.5-1 Cooling (RCIC) System F. Automatic Depressurization F. Automatic Depressurization 3.5-9 System (ADS) System (ADS) -G. Minimum Core and Containment G. Surveillance of Core and Cooling Systems Availability 3.5-10 Containment Cooling Systems H. Maintenance of Filled Otscharge Pipes H. Maintenance of Filled 3.5-10 Discharge Pipes I. Minimum River Flow 1. Minimum River Flow 3.5-11 .J. Plant Service Water System J. Plant Service Water System 3.5-12 K. Engineered Safety Features K. Engineered Safety Features 3.5-13 ~ Compartment Cooling Compartment Cooling HATCH - UNIT 1 iii i w

( 1 g( * -., LIST OF FIGUR[5 f.iSEt 11.111 1.1 -1. (Deleted) 2.1 -1 Reactor Vessel Water Levels 4.1-1 Graphical Aid for the Selection of an Adequate' Interval -8etween Tests

4. 2 -1 System Unavailability

3. 4 -1 Sodium Pentaborate Solution volume Versus. Concentration Requirements 3.4-2' Sodium penteborate Solution Temperature'Versus Concentration Requirements

3. fr-1 Pressure versus Minimum Temperature for Pressure Tests 1

Based on Surveillance Test Results i 3.6-2 Pressure versus Minimum Temperature for Non-nuclear Heatup/Cooldown .snd Low-Power Physics Test t 3.6-3' Pressure versus Minimum Temperature for Core Critical Operation other than Low-Power Physics' Test (includes 40'F Margin Required by 10 CFR 50 Appendix G) j 3.6-4 Deleted j. 3.6-5 Power-Flow Operating Map with One Reactor Coolant System Recirculation Loop in Operation .j l l 3.1$-1 Unrestricted Area Boundary j 1 l l lj y i l HATCH - UNIT 1 Amendment No.168 x

BASES FOR SAFETY LIMIT 5 1 I' FUEL-CLADDING-INTEGRITY -+ FuelCladbinoIntecritylimitatReactorPressure>800osiaandCore-A. Flow > 10% of Rated The fuel cladding integrity Safety Limit is. set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure I from nucleate boiling would not necessarily result in damage to BWR j fuel rods, the critical power at which boiling transition is calculated - to occur has been adopted as a convenient limit. However, the uncertainties-in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the 1 .P critical power. Therefore the fuel cladding integrity safety limit is ' efined as the critical power ratio in the limiting f uel assembly for d which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. The MCPR Safety Limit is determined using a model that combines all of the uncertainties in operating parameters and the procedures used to. calculate critical power. The probability of the occurrence of boiling transition is determined using an NRC-approved critical power correlation. -This MCPR Safety Limit is increased for single-loop operation over the. -comparable two-loop value (Reference 2). Details of the fuel cladding integrity Safety-Limit calculation are presented in Reference 1. l l -l I l 1 i HATCH - UNIT 1 1.1-6 Amendment No.168

g f -' 3 MM. Minimum critical-Power Ratio (MCPRS - Minimum Critical Power Ratio (MCPR) is the value of the critica power ratio associated with the most limiting assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to cause some point.in the assembly to experience boiling transition, to the actual assembly operation power. NN. - Trio System - A trip system means an arrangement of instrument channel-trip signals and auxiliary equipment required to initiate action to accomplish a protective function. A trip system may require one or-more instrument channel trip signals related to one or more plant parameters.in order to initiate trip system action. Initiation of -protective action may require the-tripping of a single trip system or the coincident tripping of two trip systems. 00. (Deleted) PP. Core Operatino Limits Report - The Core Operating Limits Report is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.11. Plant operation-within these operating limits is addressed in individual specifications. QQ. Channel Calibration - A Channel Calibration is the adjustment, as necessary, of the channel output such that it responds with the-necessary: range and accuracy to known values of -the parameter which the channel monitors. The Channel Calibration shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the Channel Functional Test. The Channel Calibration may be perf ormed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated. RR. Channel Functional Test - A Channel Functional Test shall be: Analog channels - the injection of a simulated signal into the a. channel as close to the primary sensor as practicable to verify operability including alarm and/or trip functions. b. Bistable Channels - the injection of a simulated signal into the 7 channel sensor to verify operability including alarm and/or trip functions. SS. Fraction of limiting Power Density (FLPD) - the ratio of the linear heat generation rate (LHGR) existing at a given location to the design LHGR for the bundle type. TT. Core Maximum Fraction of limiting Power Density (CMFLPD) - the CMFLPD is the highest value existing in the core of the FLPD. HATCH - UNIT 1 1.0-7 Amendment ho.168 l l 1

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1 E l8 BASES FOR SAFETY LIMITS ? l' 1 .l.l.B. Core Thermal Power Limit (Reactor Pressure < B00 osia) At pressures below 800 psia, the core elevation' pressure drop (0 power O flow) is greater than 4.56 psi. LAt low powers and flows this pressur,e differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head. - the core pressure drop at low powers and flows will always be greater than 4.56 psi. Analyses show that with a flow of 2Bx108 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus. the bundle flow with a 4.56 est driving head . will be greater than 28x108'lbs/hr. -Full scale ATLAS test data taken i at pressures from 14.7 psia to.800 psia indicate that the fuel assembly -critical power at this flow is approximately 3.35 MWt. With the design- . peaking f actors this corresponds to a core thermal power of:more than 50%. 'Thus, a core thermal power limit of 25% for reactor pressures-below 800 psia is conservative. C. Power Transient Plant safety analyses have shown that the scrams caused by exceeding any saf ety system setting will assure that the Saf ety Limit of 1.1. A or ~ 1.1.B-will.not be exceeded. Scram times are checked periodically to' assure the insertion times-are adequate. The-thermal power transient re'ulting when a scram is accomplished other than by the~ expected scram signal (e.g.,-scram from neutron flux following closure of the main turbine stop valves)' does not necessarily cause fuel damage. However, for this specification a Safety Limit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant saf ety analysis. l [ t 3-t HAICH UN!! 1 1,1 - ~l Amendment No. 168 .y.

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BASES FOR SAFETY LIMITS k

D.. Reatter Water Level (Hot or Cold Shutdown Condition) ' For the fuel in the core during periods when the reactor is shutdown, consi-deration must be given to water level requirements due to the effect of decay heat.l !f the water level should drop below the top of the fuel during this time, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height. The Safety Limit has been established at 378 inches above vessel invert to provide a point which can be monitored and also provide adequate margin. E. References 1. " General Electric Standard. Application for Reactor Fuel (Supplement for l =' United States)," NEDE-240ll-P-A. 2. "Edwin I. Hatch' Nuclear Plant Units 1 and 2 Single-Loop Operation." l NEDO-24205, August 1979. i I l l i i 4 1 IfA1CH - UNil 1 1.1 8 Amendment No. 168 3 0

yy .m jf -LIMITING CDNDITION5 FOR OPERATION

p_e. 3 SURVEILLANCE REQUIREMEhT5

~ 3.3.F. )neration with a.imitina Control - tod Pattern (for itod withdrawal 4.3.F. Deeration with a.imitina Control Rod Pattern (for. lod Withdrawal ~ 1:rror. RWE) Error. RWE) A Limiting Rod Pattern for RWC exists' when the MCPR is less than the value Ouring operation when a Limiting provided in the Core Operating Limits Control Rod Pattern for RWE exists-

Report, and only one RBM channel is i

operable, an instrument functional ] test of the RBM shall be performed i prior to withdrawal of the control-During operation with a Limiting rod (s). A Limiting Rod Pattern for Control Rod Pattern for RWE and RWE is defined by Specification 'when core thermal power is-1-30%, 3.3.F. either; 1. Both rod' block monitor (RBM) l Channels shall be operable, or 2. If only one R8M channel is oper-able, control rod withdrawal shall l be blocked within 24 hours, or i 3. If neither. RBM channel is oper-j able, control rod withdrawal shall be blocked. G. .imitino the Worth of a Control Rod' 1elow 10% Rated Therma l Power lG, limitino the Worth of a Control Rod Below 10% Rated Thermal Power 'l' j 1. Rod Worth Minimirer (RWM1 1. Rod Worth Minimiter (RWM) i Whenever the reactor is in the Start j L & Hot Standby or Run Mode below 10% s eior to the start of control tated thermal power, the RWM.shall rod withdrawal at startup, and be operable or a second licensed as soon as automatic initiation ]~ operator shall verify that the of the RWM occurs during rod operator at the reactor console is insertion while shutting down, following the control rod program. the capability of the RWM to properly fulfill its function j shall be verified by tne i I following checks. \\ a. The correctness of the i l' Banked Position Withdrawal Sequence input to the RWM i computer shall be verified, i b. The RWM computer on line j diagnostle test shall be successfully performed. i Proper annunciation of the c. I selection error of at least one out-of-sequence control' rod in each fully inserted group shall be verified, l~ d. The rod block function of L the RWM shall be verified by p withdrawing or inserting an o out-of-sequence control rod no more than to the block point. HATCH - UNIT 1 3.3-s Amendment No. 168 l

fe ~ l ' tiMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENT5 3.3.G.2. Rod Secuence Control System (RSCS)

2. Rod Secuence Control System (RSCS) a.

Ooerability a. Operability v When the reactor is in the Start As soon as the group notch and Hot Standby or Run Mode below mode is entered during each-10% rated thermal ~ power and control l reactor startup and es. goon /' rod movement is within the group as automatic initiation of notch-mode after 50% of the the RSCS occurs during rod control rods have been withdrawn, insertion while shutting the Rod Sequence Control System shall be operable except when down, the capability of the~ Rod Sequerte Control System performing the RWM surveillance to properly fulfill its tests. function shall be verified =! by attempting to select and move an inhibited control rod, When the control rod movement is within the group notch' mode and as soon as automatic - initiation of the RSCS occurs during rod insertion while shutting down, the operability of the notching restriction shall be demonstrated by { attempting to move a' control rod more than one notch in the first programped rod group. - ~ .j b. Failed Position Switch b. Failed Position Switch-Control rods with a failed " Full. A second licensed operator in" or " Full-out' position' switch-shall verify.the conformance may be bypassed in the Rod to Specification 3.3.G.2.b Sequence Control System if the ac-before a rod may be bypassed tual rod position is known. These in the Rod Sequence Control rods shall be moved in sequence to

System, their correct positions (full in on insertion or f ull out on withdrawal).

+ u HATCH - UNIT 1 3.3-6 Amendment No.168

7 7 '. LIf.lTItG COND1Tl0h5 FOR OPERATION SURv[ILLANtf RE0VIR[d[NT5 3.3.G 2.c. Shutdown Marcin/stram 4.3.G.2.c. Shutdown Marcia/ scram time 3 time festino 3'c Testing I in order to perf orm the prior to control rod with-required shutdown margin drawal f or startup, verif y demonstrations subsecuent the conformante to speti-to any fuel loading opera-fication 3.3.G.2.b. before { tions, or to perform ton-a rod may be bypassed in trol rod drive scram and/or the R$C$. the requirements f riction testing as specified to allow use of the indi-in Survet11ance Requirement vidual rod position bypass 4.3.C.? and the initial start-switches within rod groups up test program, the relata-A)g, A34, 812, or 834 of j tion of the following R5C$ the R$C5 during shutdown i restraints is permitted. The margin, scram time or fric-secuence restraints imposed tion testing are { on control rod groups A)g, _ RW operable as per $pect. A34, 8)g, or g3a after $ut (1) 'F of the control rods have been fication 3.3.G.1. withdrawn may be removed f or the i test period by means of the (2) After the bypassing of individua) rod position bypass the rcds in the R$C$ groups switches. A g, Age, $13, or $34 for l test purposes, it shall be demonstrated that movement 1 of the rods in the $0% dens-ity to 105 of rated thernal 't power range is blocked or i limited to the single notch mode of withdrawal. (3) A second licensed operator shall verify the conformance to procedures and this Specification. >v H. Shutdown Recuirements If Specifications 3.3.A through 3.3.G are not met, an orderly thutdown shall te initiated and tf.e reactor placed in the Cold shutdown Condition within 24 hours. l 1 l HATCH + UNIT 1 3.3-7 Amendment No. 168 \\>

c 2 PA5E5 FOR tiplTING CONDITIONS FOR OPf eATION AND SUAvtitt AN'Cf GECulsEMENT5 B. Control tods limiting tonditions for Operation 1 i Specification 3,3.6.1 requires that a rod which cannot be movJ with drive pressure be taken out of servite by being disarmed electric 41,y. To disarm the drive electrically, four emphenol type plug connectors are removed from the drive insert and withdrawal solenoids rendering the rod incapable of withdrawal. This procedure is touivalent to valving out the drive and is preferred because, in this condition, drive water cools and minimites crud accumulation in the drive. [lectrical disarming does not eliminate position indication. If the rod 15 fully inserted and disarmed electrically, j it is in a safe position of maailrum contribution to shutdown reactivity. i 'If it is disarmed electrically in a non-fully inserted position, that I polition shall be consistent with the shutdown reactivity limitation stated in Specification 3.3.A. This assures that the core can be shutdown at all s 4 times with the remaining control rods, assuming the highest worth operable l. control rod does not insert. An allowable pattern for control rods disarmed J. electrically, which shall meet this $pecification, will be determined and made available to the operator. Also if damage within the control rod drive { l, mechanism and in particular, cracks in drive internal housing, cannot be ruled out, then a generic problem affecting a number of drives cannot be ruled j l= i out. Circumferential cracks resulting f rom stress $$sisted intergranular t. [- corrosion have occurred in the collet housing of drives at several BWRs. This type of cracking could occur in a number of drives and if the cracks 3 propagated until severance of the collet housing occurred, scram could be prevented in the affected rods. Limiting the period of operation with a potentially severed collet housing will assure that the reactor will not be 1 operated with a large number of rods with failed collet housings. Surveillance Requirements: i The weekly control rod exercise test serves as a periodic check against Jt:terioration of the control rod system and also verifies the ability of the control rod drive to scram, since, if a rod can be moved with drive pressure, it will scram oecause of higher pressure soplied during scram. Tne frequency of exercising the control rods under the conditions of three or more inoperable rods provides even further 455urance of the reliability of the remaining control rods. l J i ? Y f HArcH - tlNIT 1 3.3 9 Amendment No. 168 1

4 U BA>E5 FOR LIMITICG CONDITION 5 FOR OPERATION AND 5URVEILLANCE REOUIREMENTS 3.3.F. Doeration with a Limitino Control Rod Pattern (for Rod withdrawal trror. DWt1 $urveillance Reduirements: A limiting control rod pattern for RWE is a pattern which, due to unrestricted withdrawal of any single control rod. could result in violation of the MCPM Safety Limit. Specification 3.3.F. defines a limiting control rod pattern for RWE. During use of such patterns when both RBM channels are not operable. it is judged that testing of the RBM system prior to withdrawal of control rods to assure its operability will assure that improper withdrawat does not occur. Ref erence Nt0C 30414-P (Ref.17) for more inf ormation. G. Limitina the Worth of a Control Rod Below 101 Rated Thermal Pown l

1. Rod Worth Minimiter (PWM)

Limiting Conditions for Operation: The RWM and the Rod Sequence Control System (R$CS) restrict withdrawals l and insertions of control rods to prospecified seguences. All patterns associated with these seguences have the characteristico that. assuming the worst single deviation from the seguence, the drop of any control rod from the fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in any pellet everage enthalpy in excess of 280 calories per gram. An enthalpy of 280 calories per gram is well below the level at which rapid fuel dispersal could occur (i.e. 425 calories per gram). Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Reference sections 3.6.5.4. 3.6.6. 7.14.5.3 and 14.4.2 and Appendix P of the F$AR, and ht00-24040. HATCH = UNIT 1 3.3-15 Amendment No.168

~ e -* B4515 FOR LIMITING CONDITlDN5 FOR OPERAfl0N AND SURVEILL ANCE Rf 0VIREMENT5 l 3.3.G.I. Rod Worth Minimiter (RWM) (Continued) In perf orming the f unction described above. the RWM and RSC$ are not re. Quired to impose any restrictions at core power levels in excess of 10% l of rated. Material in the citee references shows that it is imcossible to reach 280 calories per gram in the event of a control rod crop occur-ring at power greater than 105, regardless of the rod pattern. this is l true for all normal and abnormal patterns including those which maximite the individual control rod worth. [ At power levels below 105 of rated. abnormal control rod patterns could l produce rod worths high enough to be of concern relative to the 280 cal-orie per gram rod drop limit. In this range of RWM and the RSC$ con-strain the control rod seguences and patterns to those which involve only acceptable rod worths, The RWM and the RSCS provide autorre ic supervision to assure that out l i of seguence control rods will not 'F withdrawn or inserted; i.e., it limits operator deviations f rom p1pned withdrawal sequences. They serve as a backup to procedural control of control rod seguences, which limit the maximum reactivity worth of control rods. In the event that the RWM is out of service, when required. A second licensed operator or l other qualified technical plant employee whose qualifications have been reviewed by the AEC can manually fulfill the control rod pattern conformance functions of this system. The functions of the RWM and R$C$ make it unnecessary to specify a license limit on rod worth to preclude unacceptable consequences in the event of a control rod drop. At low powers, below 105, these devices force ad-herence to acceptable rod patterns. Above 105 of rated power. the con-sequences of a rod drop event without RWM or R$CS are acceptable. Power level for automatic cutout of the RSCS function is sensed by first stage turbine prellure. Power level for automatic cutout of the RWM function is sensed by feedwater and steam flow and is set to be consistent with the RSCS setting. Surveillance Requirements: Funcitonal testing of the RWM prior to the start of control rod withdrawal at startup, and prior to attaining lot of rated thermal power during rod in-l sertion while shutting down, will ensure reliable operation and minimite the probability of the rod drop accident. 2. Red Secuence Control System (R$CS) 8. Operability limiting Conditions for Operation: l See bases for Technical Specification 3.3.G.I. Rod Worth Minin. iter. I' I 1 l HATCH - UNIT 1 3,3 16 Amendment No. 168 L I

i BASE 5 FOR LIMITING CONDITION 5 50R OPEeATION AND SURv[lLL ANCE 8E0VIREMENi 3.3.G 2.c. $hutdognMarcin/ScramTimeTestino(Continued) [ tion simulation switches provided in the RSCS for such purposes. During the scram time testing. reactor concitions will be such that the reactor rod pattern will be in RSC$ 8 group. All A12 and A 4 rods will be f ully withdrawn, alternatively the rod pattern S w111 be in R$C$ group A and all Big and B34 rods will be fully I withdrawn. To test A3a rods, it will be necessary to simulate i all withdrawn B rods as being at the full-in position, and for testing A 3 i the full 12 rods, all A a and all withdrawn B rods 45 being at in position. The simulation of already withdrawn control rods in the 100s to 50s rod censity range (Al2 and A34 or alternatively B 2 and 834) as being full-in to perform the 1 individual rod test does not violate the intent of the R$CS since; (a) the single notch mode of rod withdrawal for rods in the 50s density to los of rated thermal power range will remain in effect i until that power level has been achieved and the test procedure will require that this be verifiedt (b) no B group rods can be selected either for withdrawal or insertion during the time that an A12 or A a rod is fully inserted or 15 simulated as being 3 + in the fully inserted position (similarly for the A group rods when the B sequence is chosen for startup and (c) all rod polition simulation switch operations will be verified by a second independent check. H. $hutdown Recuirements Should cire.umstances be such that the Limiting Conditions for Operation as stated in Specifications 3.3.A. through 3.3.G. cannot be met, an orderly shutdown shall be initiated and the reactor placed in the Cold Shutdown Condition within 24 hours. 1. Scram Discharae Volume Vent and Drain valves The scram discharge volume vent and drain valves are required to be OPERABIE. So that the scram discharge volume will be available when needed to accept discharge water f rom the control rods durinil a reactor scram and will isolate the reactor coolant system from the containment when required. .1. References 1. F3AR Section 3.4. Reactivity Control Hechanical Design 2. FSAR Section 3.5.2. Safety Design Bases 3. FSAR Section 3.5.4. Safety Evaluation 4. FSAR Section 3.5. Control Rod Drive Housing Supports L HATCH - UNIT 1 3.3-18 Amendment No.168 L a

Frn 3 1 q. D. 1 4 6A l BASE 5 FOR LIMI'INE CONDITIONS FOR OPERATION AND SURVEILLANCE HEDUIRE j 3.6.1. Jet Puent (Continued) 'I i A nottle-riser system f ailure could also generate the coincident f ailure of a jet pump body; however, the converse is not true. The latt of any i substantial stress in the jet pump body makes failure impossible without en initial not21e riser system failure. One of the acceptable orocedures for jet pump surveillance, identified in NUREG/CR-3025. Reference 2 was adopted for Hatch Unit 1. The surveillance is performed to verify that neither of the following conditions occur: (a) The Recirculation Pump Flow /$ peed Ratio deviates by more than $1 from the normal range, or 1 i (b) The Jet Pump Loop Flow / Speed Ratio deviates by more than $5 from the normal range, t If either criterion is f ailed, then the procedure calls for comparing either the individual jet pump flow or individual jet pump dif fuser to lower plenuta of f ferential pressures to the criteria of the Limiting Conditions for Operation (LCO). If the LCO criteria are not satisfied and pump speed is less than 60% of rated, it may be necessary to increase pump speed to above 60% of rated and repeat the measurements bef ore declaring a jet pump inoperable, in this case, it is recommended that close monitoring and increased recirculation pump spted should be performed only if the criteria are exceeded by an ar.rount to be deter 1nined f rom previous plant operating experience. 3.6.J. Recirculation System Operation with a single reactor coolant system recirculation pump is allowes, provided that adjustments to the flow referenced scram and APRM rod block setpoints. MCPR cladding integrity Safety Limit. MCPR Operating Limit, and MAPLHGR litait are made. An evaluation of the performance of the (CCS with single-1 cop operation has been performed and determined to be acceptable, Reference 4 Based on this Reference, a factor is applied to reduce the APLHGR limits during single-loop operation. To account for increased uncertainties in the total core flow and TIP readings when cperatitig with a single recirculation loop, an increase is applied to the l MCPR cladding integrity Safety Limit and MCPR Operating Limit over the - comparable two-loop values. The flow ref erenctsd simulated thermal power scram and rod block setpoints for single-recirculation-loop operation is reduced by the amount of maw, where m is the flow reference slope for the rod block monitor anc AW is the largest dif ference between two-loop and single-loop effective drive flow when the active 1000 indicated flow is the same. This adjustment is necessary to preserve the original relationship tetween the rod block and actual effective drive flow. When restarting an idle pump, the discharge valve of the idle loop is required to ter4 sin closed until the speed of the faster pump is below 50% of its rated speed to provide assurance that when going f rom one-to two-loco l_ operations, excessive vibration of the jet pump risers will not occur. l l the possibility of experiencing limit cycle oscillations during single-loc; oceration is precluded by restricting the core flow to greater that, or ecul i to 45s of rated core flow when core power is greater than the 80s rod line, This requirement is based on General Electric's recommendations contained in 5tl 380. Revision 1. which defines the region where the limit cycle oscillations are more itkely to occur. HAICH ON!) l 3 + 22 Amendment NO. 168

@p ,o LIMITINE CONDIT10N5 FDR OPERATION SURVEILLANCE Alou1REMENT5 L e 3.11. FUtt.ROR$ a.11. FUtl It005 Atalitability Antlitability The Limiting Conditiens for Operation The Surveillance Recuirements acc1r allDciated with the f uel t* Des apply to to the parameters which monitor the those paraesters which monitor the fuel rod operating conditions, f uel ro6 speetting ten 64titms. Obiettive Obiettive F The Objective of the Limitin0 Condt-The Objective of the Surveillance tions f cr Operation is to assure the Requirements is to specif y the type perf orinance of the f uel rods, and frecuency of surveillance to be applied to the fuel rods, r. $betiftrations $aecification,1 A. Avertoe elanet tinear +4 eat Genera-A. avfrane elanar tinter Heat Genere-11pn Rate (M & Don aate (aPL4R) Daring power operation, the APLNH The APLNH for each type of fuel as f or all core locations shall not a function of everage planar exceed the appropriate APLNGR limit exposure shall be determined daily provided in the Core Operating Limits during reactor eperation at 3 258 Report, if at any time during oper-rated thermal power, ation it is determined by nor1nal surveillance that the limiting value for APLHGR 15 being eacteded, action shall be init'ated within 15 minutes to restore operation to within the prescribed limits. if the ALPNGR is not returred to within the prescribed limits within two (2) hours, then reduce reactor power to less than 255 of rated tf.ermal power within the nest four (4) hours, if the limiting condi-tion for operation is restored prior to expiration of the specified time interval, then further progression to less than 255 of rated thermal power is not recuired. I B. tierar heat Generation eate (tHGR) The LNGR as function of core height shall be checkea daily cur-ing reactor operation at 7, 2',5 rat'd thermal power. ) %TCH uN1I i L !' d Amendment No.168 l 1

~ ~... - -. .J LIEITIM LeulTIEN5 FM DPERATIM MEVEILLANEI EEnuIntagggT5 3.11.0. Linear Ilmat Generatian kata (L.IAA) 4.11.8. Linear heat annaratian tate f Lh&ti I l ri,,, -,.p.r.u . t t .n u . n . a.. a. q r.t... r n., u..t 3 ri . n t... u n.4 u,,g. .ri v.lue Dr. vie in the C.re Oper. thefusi power, . ting Lietts he..rt, if.t.ny i tira.oring j l i i 1 4 I l b i i r f f I l t k [ l i 5 l l-MATCH + UN11.1 3.11-14 Amendment No. gg

I .c i r t!NITING CONDIT1045 FOR OP[RATIDh 509v[IttAhCE #foultfMEhT5 3.11.B. Linear Meat Eseeratien Rate (LHGR) (Continued) 4 operation it is determined by normal surveillance that the limiting value i for LHGR is being esteeded, action i shall be initiated within 15 minutes to restore operation to within I the prescribed limits. If the LNGR is cet returned to within the prescribed limits within 2 hours, then reduce reactor power to i' less than 255 of rated thermal l ) power within the next a bours. If the limiting condition for operation is restored prior to j expiration of the specified time interval, then further progression to less than 25% of rated thermal l power is not required, C, Minimum Critical Power Ratio (MCPRI a.ll.C.l. Minimum Critical Power Patio (MCPR) s The minimum critical power ratio MCPR shall be determined to be (MCPR) thall be equal to or greater equal to or greater than the i than the operating limit MCPR applicable limit, daily during (OLMCPR) provided in the CORE OPERAt!NG LIMITS R(PORT. reactor power operation at > 255 l rated thermal power and following any change in power level or If at any time during operation it distribution that would cause il determined by cormal surveillance operation.ith a limiting control that the limiting value for MCPR rod pattern as described in the is being exceeded, action shall be bases for Specification 3.3.F. initiated within 15 minutes to restore operation to within the prescribed limits. If the steady a.11.C.2. Minimum Crit ic al Power Ratio limit state MCPR is not returned to within the prescribed limits within two (2) The MCPR limit at rated flow and hours, then reduce reactor power to rated power shall be determined. less than 25% of rated thermal power as provided in the CORT OP(RAT-1 within the next four(a) hours. If ING LIMITS REPORT. using: the Limiting Condition for Operation is restored prior to expiration of a.- 51.0 prior to initial scram the specified time interval, then time measurements for the further progression to less tnen cycle, performed in accordance 25% of rated thermal power is not with specification a.3.C.2.a.

required, or b.

t is determined from scram time measurements performed in accordance with Specifica. tion 8.3.C.2. The determination of the limit must be completed within 72 hours of the conclusion of each scram time surveillance test required by Specification a.3.C.P. HATCH - UNIT 1 3.11-2 Amendment No. 168

? M5E5 FOR LIMITlhG C'DhD1110h5 (DR OPER ATION ahD SUR4EILL ANCE EEOUIR 3.11. FUtt ROD 5 A. averace. Planar tingar Neat Generation Pete f aptMGR) [ This 5Decification assures that the. peak cladding temperature following the postulated design t> asis loss-of-coolant accident (LOCA) will not tacted the Itmit scetitied in 10 CfR 50,46 even considering the postulated effects of fuel pellet centification. The peak cladding temperature following a postulated loss-of-coolant acci-dent is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent second-arily on the rod to rod power distribution within an assembly. $1hte et-pected local variations in power distribution within a fuel assembly af f ect t L' the calculated peak clad temperature by less than + 2C'F relative to the peak temperature f or a typical f uel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures conform to 10 CFR 50.46. The limiting value for APLH6R is providea in the Core Operating Limits Report. The calculational procedure used to establish the APLHGR limits is based on a LOCA analysis. The analysis was performed using General Electric (6t) calculational models which are consistent with the requirements of Appendia K to 10 CFR 50. The LOCA analysis was performed utilizing the new improved calcu14tional model. 5Af(R/GESTR-LOCA. The analysis demonstrated that loss-of-coolant concerns do not limit the operation of the fuel since margin to the 2200'F limit was demonstrated (Reference g). Therefore, the APLHGR limits are deri ed to assure that the fuel thermal-mechanical l t design criteria are met. A flow dependent correction f actor is applied to the rated conditions APLHGR to assure that the 2200'F PCT limit is complied with during LOCA initiated from less than rated core flow. In addition, other power and flow dependent corrections are applied to the rated conditions APLHGR l limits to assure that the fuel thermal-mechanical design criteria are met during abnormal transients initiated from of f-rated conditions for two-loop and single-loop operations, References 2 and 8. For single-loop operation, a multiplicative factor is applied to the rated conditions APLNGR limit for all fuel bundles when core power exceeds a specified value. The power and flow-dependent correction f actors, and the limiting values for APLHGR for each fuel type used in a particular cycle are specified in the Core Operating Limits Report. 4 i HATCH - UNIT 1 3.11-3 Amendment No. 168 f i 1

R&5E5 FDR LlWUlWG CDNDITIDNS~ Foa OPERATIow AmD 5uRWEILLahCE eIOUIREMENT5 3.11.0. Linear Heat senaration Rate (LNGR) 1 This specification assures that the LNGR in any rod is less than the destgr. j linear heat generation if fuel pellet dentification is postulated, for LMGR I' ' to be a 11miting value below 25-percent rated thermal power. the ratto f of peak LNGR to core average LNGR would have to be greater than g.6. which L is precluded by a considerable mrgin when employing any permissible control rod pattern. C. Minimus Critical Power Ratio (MCea) the reduired operating limit MCPR as specified in Specification 3.11.C is derived f rom the established f uel cladding integrity Safety Limit MCPR and an analysis of abnormal operational transients presented in Reference 1. Various transient events will reduce the MCPR below the operating MCPR. I b. To assure that the fuel cladding integrity Safety Limit is not violated during anticipated abnormal operational transients, the most limiting transients have been analyted to determine which one results in the largest reduction in critical power ratio (a MCPR). Addition of the largest a MCPR to the Safety Limit.MCPR gives the minimum operating limit MCPR to avoid violation of the Safety Limit should the most. limiting transient occur, e The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. Details of how MCPR evaluations are performed, the methods used, and how the MCPR limit is adjusted for operation at less than rated power and flow . conditions used for single-loop operation are given in Reference 1 and in the Core Operating Limits Report. i b HATCH - UNlt i 3.11-4 Amendment No. 168

r '.~ - p I MLE1 FOR LIMITIM CDNDITION5 FDR DPERATIDN AND 5URVEILLANCE R[DUIRER[WT5 3.11.C. Minimum critical Pe=er Ratio (MCPR) (Continued) According to the CORE OPERATING LIMITS REPORT the 1005 power,1005 l flow operating limit MCPR (OLMCPR) eepenas on the average scram time, t, of the control rods, where: t = 0 or ' ave 't, whichever is greater t -t g g 'A = 1.006 set (Specification 3.3.C.2.a. scram time limit where: to notch 36) '$ = y + 1.65 1 (Reference 1) e n !N 1=1 j. where: y = 0.022 set (mean scram time used in the transient analysis) L =.018 set (standard deviation of u) e 4 n t ii ave = i=1 n i N 11 where n = number of surveillance tests performed to date in the cycle Ni = number of active control rods measured in the ith surveillance test ti = average scram time to notch 36 of all rods in the ith surveillance test N = total number of active rods measured in 4.3.C.2.4 g i l l i l l t 1 HATCH - Unit 1 3.11 -d a Amendment No.168 I L

T t aantN11TRATivt 20NTROL$ j O. Type of container, e.g., LSA, type A, type 3, large quantity. e. I f. Solidification agent, e.g., coment. The Radioactive Effluent Release Report shall include (on a svarterly basis) unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and lievid effluents that were in escess of 1 Ct. j entluding dissolved and entrained gases and tritive for lievia ef fluents, or i those in excess of 150 Ci of noble gases or 0.02 Ci of radiciodines for i gaseous releases. l I ' PROCT 55 CONTROL PROGRAM and to the OFF5!Tt 005t CALCULA11 turing the reporting period. D 800NTHLY DPERATING REPORT - 6.g.1.10. Routine reports of operating statistics and shutdown emperience shall be submitted on a monthly basis to the Director Office of management and Program Analysis, U. 5. Nuclear Regulatory Consnission, Washington. ~D. C. 20$55, with a copy to the Regional Office of Inspection and (nf orcement no later than the 15th of each month following the calender month covered by the report. 1 1 CORf OPERATING L1M175 RLP.QRT 6.g.1.11.4. Core operating )imits shall be established and documented in the C0tt OPERAT!NG LIMlf5 REPORT before each reload cycle or any remaining part of a reload cycle for the followingt { (1) Operation with a Limiting Control Rod Pattern (for Rod Withdrawal trror. RWE) for Specification 3.3 F. i (2) The Average Planar Linear Heat Generation Rate (APLHGR) f or Specification 3.11. A, (3) The Linear Heat Generation Rate (LHGR) for Specification 3.11.B. and (4) The Minimum Critical Power Ratio (MCPR) for Specifications 3.3.F and 3.11.C and Surveillance Requirement 4.11.C. b. The analytical Nthods used to determine the core operating limits shall be those previously reviewed and approved by the hRC in the f ollowing cocuMnts. (1) Nt0E-24011-P-A, ' General Electric ' Standard Application f or Reactor Fuel,' (applicable amendment specified in the CORE OPERAT!hG LIMIT 5 REPORT). (2) *$afety [ valuation by the Of fice of Nuclear Reactor Regulation Supporting Amendment ho. 151 to Facility l Operating License DPR 57,* Cated Septemeer 12, 1988. c. The core operating limits shall be determined 50 that all l applicable limits (e.g., f uel thermal-mechanical limits, core l thermal-hydraulic limits, ICCS limits, nuclear limits such as l-Shutdown margin, and transient and arcident analysis limits) of the safety analysis are met. i d. The CORE OPERAllNG LIM 115 REPORT including any mid-cycle I revisions or suppitments thereto, shall te provided ucon issbance, f or each reload cycle, to the NRC Document Control Desk with copies to the Regional Acministrator and Resident t inspector. HATCH - UNIT 1 6-15d Amendment No, M

.ma nt% +f UNITED $TATES .. Y g NUCLEAR REGULATORY COMMISSION l 5 i 8 W A$HINGTON, D. C. 20655 ,/ y GEORGIA POWER COPFANY OGLETHORPE POWER CORP 0kAT10N MUNICIPAL ELECTRIC AUTHORITY CF GEORGIA CITY OF DALTCN, GEORGIA DOCKET NO. E0-366 i EDWill 1. HATCH NUCLEAR PLANT, UNIT NO. 2 AMEl$ MENT TO FACILITY OPERATING LICENSE 7 Amendment flo.106 i License flo. NPF-5 1. The huclear Regulatory Cona.ission (the Comission) has found that: The app (lication for amendment to the Echin I. Hatch Nuclear Plant, A. Unit 2 the facility) Facility Optrating Licetise No. NPF-5 filed Ly Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipi Electric-Authority of Ceorgia, and City of Dalton, Georgia (the licensee) dated June 22, 1989, as amended Ouly 31, 1989, and Octcber 4,1989, complies with the standards and requirements of the Atoniic Energy Act of 1954, as amended (the Act), t and the Coninission's rules and regulations set forth in 10 CFR Chcpter I; B. The facility will operate in confornity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (1) that the activities authori:ed by this omendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will te conducted in compliance with the Comission's reguletions set forth in 10 CFR Chapter I; D. The issuance of this amendment will rot be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amer.dment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable ru;uirements have been satisfied.

{ i 2 2. Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and i paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby amended to read as follows: Technical Specifications i The Technical Specifications contained in Appendices A and B, as revised through Amendment No.106, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. t 3. This license amendment is effective as of its date of issuance and shall be impleraented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COMMISSION kb co/ A David B. Ma thews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Chances to the Technical i Specifications Date of Issuance: December 29, 1989 i

i , s ATTACHMENT TO LICENSE AMENDMENT NO. 106 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET HO. 50-366 i Replace the following pages of the Appendix A Technical Specifications with ^ the enclosed pages. The revised pages are identified b centain vertical lines indicating the areas of change. y amendment number and Corresponding overleaf pages are provided to maintain document completeness. Remove pages insert Pages I XVI XVI 1-2 1-2 02-1 B2-1 B2-2 B2-2 BP-3 B2-3 B2-4 B2-4 B2-5 B2-5 B2-13 B2-13 3/4 1-4 3/4 1-4 3/4 1-14 3/4 1-14 3/4 1-15 3/4 1-15 3/4 1-16 3/4 1-16 3/4 1-17 3/4 1-17 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-4a 3/4 2 Ab 3/4 2-4c 3/4 2-4d 3/4 2-4e - 3/4 2-41 3/4 2-4j 3/4 2-4k 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-7a 3/4 2-7a 3/4 2-7b 3/4 2-7b - 3/4 2-7d 3/4 2-7c 3/4 2-7d 3/4 2-8 3/4 2-8 3/4 3-39 3/4 3-39 3/4 3-42 3/4 3-42 3/4 10-2 3/4 10-2 B 3/4 1-3 B 3/4 1-3 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-3 B 3/4 2-3 B 3/4 2 4 B 3/4 2-4 B 3/4 2-6 B 3/4 2-6 B 3/4 4-1 B 3/4 4-1 6-14d 6-14d

l 3.:.

INDEX 3 OEFIN!TIONS ]Y i i SECTION i 1.0 DEFINITIONS F EASE ACTION 1-1 AVERAGE PLANAR EXPOSURE ? 1-1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1-1 I CHANNEL CALIBRATION i 1-1 i CHANNEL CHECK 1-1 CHANNEL' FUNCTIONAL TEST 12 CORE ALTERATICN i 1-2 CORE MAXIMUM FRACTION OF LIMITING POWER CENSITY 1-2 CCRE OPERATING LIMITS REPORT 1-2 l CRITICAL PCWER RATIO 1-2 l 'I-AVERAGE 0!SINTE3 RATION ENERGY i 1-2 EVERGENCY CORE COOLING SYSTEM (ECCS) RESPCNSE TIME1-2a FRACT:0N OF LIMITING POWER DENSITY 1-3

REOUENCY NOTATION 1-3 ICENTI !EO LEAKAGE 1-3 ISOLATION SYSTEM RESPONSE TIME 1-3 LIMITING CONTROL ROO PATTERN 1-3

+ LINEAR HEAT GENERATION RATE i 1-3 LO3!C SYSTEM FUNCTIONAL TEST 1-3 MINIMUM'CR!T: CAL POWER RATIO 1-4 0 ERAELE - OPERABILITY 1-4 0:ERATICNAL CCNOITION 1-4

'#5;CS TESTS 1-4

-ATCh-.N:7 2 I Amendment No. 106

I l !!CIX OET N:T!0NS i + 3ETT 0N

0 OEF:N T:CNS (0 ntt.nued) g PRES $URE 3CWCARY :.EAXA0E 14 t

PRIMARY C0h"AINMEtC INTICR TY ^ 15 R.A""ED THERMAL PCWER 15 REACTCR PRC'"ECTICN SYSTEM RESPCNSI T:ME 1-5 RCD DENS *Y i 15 IIO !CARY C !CA:.W.I!C !!CICR:""! 1-6 1 ERU" CWN MARC:N 16 i STAGCEPID !!!! IAS:5 1-6 THERMAL PCWER 1-6 WIDDCIT!ED LEAXACE 1-6 SCURCE ONE X 1-7 PRCCESS CONTROL PROGRAM 1-7 SCLIDITICATION 1-7 I OTTSITI OCSE CALOU ATICN MANUAL 1-7 GA$ ECUS RADWAs*E TREATMENT 3Y37EM 1-7 3 MEMBER (5) CF TEE PUBLIC 1-4 SITE BCUNDARY 1-4 VNRESTRICTED AREA 1-0 DCSE EQUIVALENT ICDINE 1-4 PURGE - FURGING 1-8 VENTING 1-9 MILK ANIMAL 1-9 BATCI-UNIT 2 II Amendment IA 48 m

n 1.0 DEFINITIONS The following terms are defined so that uniform interpretation L of these specificaticas may be achieved. The defined terms appear in capitalized type and shall be applicable,throughout these Technical Specifications. OEIISLN ACTIONS ahall be those additional requirements specified as corollary statements to each specification and shall be part of i the specifications. i AVERACE PLANAR EXPOSURE The AVERACE PLANAR EXPOSURE ahall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle. AVERACE PLANAR LI W J. EEAT CENERATION RATE The AVERACE. PLANAR LINEAR IEAT CENERATION RATE (APLBCR) shall be. applicable to a specific planar height and is equal to the sua i l of the LINEAR EEAT CENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle. j CHANNEL CALIBRATION l A CEANNEL CALIBRATION shall be the adjustment, l as necessary, of L the channel output such that it responds with the necessary range and assuracy to known valuea ef the parametar which the channel monitors. The CEANNEL CALIBRATION aball encompass.the l entire channel including the sensor and alarm and/or trip ~, functions, and shall include the CEANNEL FUNCTIONAL TEST. The CEANNEL CALISKATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated. L 1. CRANNEL CHECK A CEANNEL CEECK ahall be the qualitative assessment of channel behavior during operation by observatten. This determination shall include, where possible, comparison of the channel o l indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. EATCE UNIT 2-1-1 Amendment No. M. 48

g WN. r s INDEX L ~ ADMINISTRATIVE CONTROLS i b i ( SECTION PAGE SAFETY REVIEW BOARD (Continued) Audits 6-10 i Authcrity' 6-11 i Recorcs 6-11 L 6.6 REPORTABLE EVENT ACTION i 6-11 6.7 SAFETY LIMIT VIOLATION 6-11 i~ 6.8 PROCEDURES i-6 d 6.9 REDORTING REQUIRE *ENTS i ROUTINE REPORTS 6-13 STARTUP REPORT 6-13 ANNUAL REDORTS 6-13 ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT 6-14 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6-14a MONTHLY OPERATING REPORT' 6-1Ad CCRE OPERATlNG LIMITS. REPORT 6-14d l SPECIAL ;EPORTS 6-17 6.10 RECORD RETENTION 6-17 6.11 RAO:Ai!ON PRCTECTION DRC3 RAM 6-1B 6.12 -IG' RACIATION AREA 6-1E -A*:--UNIT 2 xV! Amendment No. 106 i

i.0 DEFIN1110NS (Continued) CHANNEL FUNCTIONAL TEST i A CHANNEL FUNCTIONAL TEST shall te: i i Analog channels - the injection of a simulated signal a. into the thannel as close to the primary sensor as pra:ticable to verify CPERABILITY includin trip functions and channel failure trips. g alarm anc/or b. Bistable channels - the inje: tion of a simulated signal i into the channel sensor to verify CPERABILITY in:1ucing alarm and/or trip functions. CORE ALTERATION CORE ALTERATION shall be the accition, removal, relocation or movement of fuel, sources, intore instruments or reactivity { controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspensien of CORE ALTERATIONS shall not cre:1uce cempletion of the movement of a component to a safe conservative position. L CCRE waXIMUV RACT!ON OF liv!*!N3 oCWER DENSITY The CORE MAXIMUM :RACTION OF LIMITING POWER CENSITY (CMFLcD) sh(11 te the largest FLPD wnien exista in the core for a given operating concition. CORE OPERATING L!v!TS REeCRT I The CORE OPERATING LIMITS REPORT shall te the unit-specifi: i' Oo:ument tnat provides core operating limits for the current operating reloac ey:le. These cycle-soecific core operating limits snali te eterminee for each reloac cycle in ac:or:an:e l witn Specification 6.9.1.11. Plant operation within these operating limits is a:cressed in incivicual sce:1f t:ations. l CRIT! CAL ?OWER RATIO The CRITICAL POWER RATIO (CPR) shall be the ratio of that cower in the assemoly.hich is calculated by apolication of an NRC-approvec :ritical power correlation to cause some point in the assemely to experience boiling transition, civiced oy the a:tual assembly operating power, h-AVERAGEDISINTEGRATIONENERGY E shall be tne average, weignied in pre;crtion to the concentration of eacn racionu: lice in the reactor coolant at the time of samoling, of the sum of the average beta anc gamma energies per cisintegration, in MeV, for isotopes with half lives greater than 15 minutes. making up at least 95% of the total non-iodine activity in :ne coolant. l l HATCH - UNIT 2 l 1-2 Amendment No. 106 l l 1

i V j 2.1 SAFETY LIMITS 1 BASES 2.0 The fuel claccing, reactor pressure vessel and primary system piping are the principal barriers to the release of racioactive materials to the environs. of these barriers curing normal clant ecerations and anticipated l sients 'The fuel claccing integrity Safety Limit is set such that no fuel damage is calculatec to oc:gr if the limit is not violated. 1 fuel camage is not cirectly etservable, a stee-back accreach is used to Because establish a Safety Limit such that tre MCPR is not less than 1.04 for two-loop oceration and 1.05 for single-loop operation. These limits represent a conservative margin relative to the concitions required to maintain fuel

Ta::ing integrity.

The fuel clat:ing is one of the pnysical barriers which se:arate the radioactive materials from the environs. The integrity of this cla::ing barrier is relate: to its relative free:om from perforations or cracking. the life of the cla ing, fissi0n Or0 uct migration from this source isAltneug incrementally cumulative ar: :erttaucusly measurable. Fuel cladding [ cerforations, he.ever,

an eesult from thermal stresses which occur from Safety System Settings, reactor coeration sigM f*:artly above :esign ::rcitions and t) wnile fission procu: migration from cla::ing perforation is just as measuraele as that from use related cracking, the thermally cause

la: ing :erf: rations signal a threshold beyond whien still

-i greater thermal stresses may cause ge:ss rather than in:remental cla::ing eterioration. Trerefore, ;*e 'uel :ia::ing Safety.imit is cefined with a margir to the con:itices nich wovic :re:vce onset of transition teiling, Y;PR of 1.0. Ta.ese :Or:1tions represent a signifi: ant ceparture from the :entiti:n intence: Dy esign for planne: ::eration. Tne evaluations wnich justify normal c:eration, aenormal transient, a::10ent analyses fo* twc-loco coeration are Oiscusse and in cetail in Reference 1. Evaluation for single-loep :ceration :emonstrates that two-loop transient and acticent analyses are more limitir; tnan sirgle-leep, Referen:e 2. 2.1.1 T-ERVaL DNER (Lo. D* essure or Lew tiew) The use o' the NRC-approved transition Ociling correlation is not valid l for all :ritical power calculations at pressures telow 755 psig or core flows less tnan IC'. of rated flow. Limit is established by etner means.Treref:re, the fuel claccing integrity Safety This is 00ne by establisning a limiting con ition on core THERMAL E0aER with tne (Oilowing easis. Sin:e the pressure crop in the bypass region is essentially all elevation hea, the Ore pressure crop at 10. oower ar: flows will always ce greater than 4.5 051. Analyses t snow inat witn a burdle fl of 2B x 10' lbs/nt, Dundle cressure crop is rearly incecendent of buncle oo.er an: has a value of 3.5 si. Thus, tne uncle fic..itn a 4.5 :si criving nea: will te greater than 2B x 108 10s/nr.

ull s: ale ATLAS test cata taken at o*essures from 14.7 psia to 500 Osia inci: ate tnat the fuel assemoly criti:al po.er at this flow is appr0ximately 3.35 Yat.

'aitn tre cesign eadng fa:::rs, snis corresponcs to HATCH - UNIT 2 2 2-1 Amendment No. 106

( '.c c SAFETY LIMITS o BASES (Continued) r a THERMAL POWER of more than 50?4 of RATED THERMAL POWER. Thus, a THERMAL ' POWER limit of 25?4 of RATED THERMAL POWER for reactor pressure below 7 is conservative. 2.1.2 THERMAL power (Hich Dressure and Hich Flew) The fuel clatiding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable dur-ing reactor operation, the thermal arid hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning i of the region where fuel camage could o::vr. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been a: opted as a convenient limit.

However, the uncertainties in monitoring the core operating state and in the pro:ecures used to calculate the :ritical power result in an uncertainty in the value of the critical power.

Therefore, the fuel claccing integrity Safety Limit is cefined as the CPR in the limiting fuel assembly for wnich more than 99.9', of the fuel rods in the core are expected to avoid coiling transition considering the power distribution within the core and all un:ertainties. The MCPR Safety Limit is cetermined using a model that combines all of l the uncertainties in operating parameters and the orocedures used to calculate critical power. The probabilit,y of the 00:urrence of boiling transition is det, ermined using an NRC-approved critical power correlation. Details of the fuel claccing integrity Safety Limit calculation are presented in Reference 1. l HATCH - LNIT 2 3 2-2 Amendment No. 106

ff. 1] 4 e (This page intentionally elank) k 9 ~ ~ '... 2 3 '-3 Amendment No.106

n j 4 I ... g o 4 8 l 1 i l i r i i t e t !t p ht$ Da9e intentionally blank) t i b I r T l .) 5 l l l l r l l .~ L T :H - ;'ili 2 Amendment No. 106 l I L

I f-h i L 4 ? I P 1 i ~

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,l p (This page intentionally blank) ? ~~' 2 E 2-5 Amendment No. 106

LIMITING SAFETY SYSTEM SETTING BASES (Continued) REACTCR PROTECTION SYSTEM INSTRUvENTATION SETD0!NTS (Con _ Turbine Centrol valve east Closure, Trio 011 Pressure-Low (Continvec) ' pressure switches whose contacts form the one-out-of-two-twice logic input to the Reactor Protection System. 504 greater closure time and a cifferent valve characteristic from thatThis tr of the turbine stop valve, comoine to procuce transients very similar to that for the stop valve. No significant change in MCPR occurs. Relevant transient analyses are ciscussed in $ection 15 of the Final Safety Analysis Report. This scram is bypassed when turbine steam flow is below that c0rresconding to 30% of RATED THERMAL POWER, as measurec by turcine first stage pressure. 11. Reactor v ce Switch !* Shuteewn Desition o The reacter mede swit:h Sn tcowr position tric is a recundant channel a to the aut;matic protective instrumentation channels and provices aceitional manual reactor trip capability.

12. Manual Scram the Manual $ cram 1s a recuncant channel to the automatic protective instrumentatien channels anc provicet manual reactor trip capacility, 2.3 RE ERENCES

1. '" General Electric Stancarc Acclication for Reactor Fuel (Supplement for l Unitec States)," NE00-24012-P-A.

2. "Ecwin 1. Hatch f,'utlear Plant Units 1 and 2 Single-Loop Operation," NE;;-24:05, August 2979. i I i l -a*:-.',:7 2 E -12 Amendment No. 106 l 1 i

l REACTIVITY CONTROL SYSTEMS 3 /4.1. 3 CONTROL RODS CONTROL ROD OPERABILITY LIMITING CONDITION FOR OPERATION i } 3.1.3.1 All withdrawn control rods shall be OPERABLE. APPLICABILITY: CONDITIONS 1 and 2. ACTION: i' With one withdrawn control rod declared inoperable due to being a. immovable as a result of excessive friction or mechanical inter-ference or known to be untrippable, restore the inoperable control rod to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours, b. With no more than 8 withdrawn control rods declared inoperable, s the provisions of Specification 3.0.4 are not applicable and operation may continue, provided that within 2 hours; i i 1. The insertion capability of each inoperable withdrawn I control rod is demonstrated by inserting the control rod at least one notch

by drive water pressure within the normal operating range and then either the directional control valves are electrically disarmed or the withdraw isolation valve is closed, or 2.

The inoperable control rod is fully inserted and either the directional control valves are electrically disarmed or the withdraw isolation valve is closed, and 3 Each inoperable withdrawn control rod is separated from all other inoperable withdrawn control. rods by at least 2 OPERABLE control rods in all directions; Otherwise, be in at least HOT SHUTDOWN within 12 hours. 'The inoperable control rod may then be withdrawn to a position no further withdrawn than its position when found to be inoperable. o L HATCH - UNIT 2 3/4 1-3 l'

i REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS ' 4.1. 3.1 - All withdfawn control rods that do not have their directional contro1, valves electrically disarmed or their withdraw isolation valve -closte shall be demonstrated OPERABLE by moving each control rod at least one notch; least once per 7 days wnen above 30?4 of. RATED THERMAL POWER, a. A b. At least once per 24 hours when above.30?4 of RATED THERMAL POWER and-three or more control rods are immovable, '"#Eh - UNI 2 1/4 '-4 Amendment No.106 1

1. REACTIVITY CONTROL SYSTEMS 3/4.1.4-CONTROL ROD PROGRAM CONTROLS ROD WORTH MINIM!ZER i LIMITING CONDITION FOR OPERATION 3.1.4.1 The Rod Worth Minimizer (RWM) shall be OPERABLE. APPLICABILITY: CONDITIONS 1 and 2*, when THERMAL POWER is less than 10** I of RATED THERMAL POWER. I ACTION: With the RWM inoperable, the provisions of Specification 3.0.4 are not applicaule, operation may continue and control red movement is permitted provided that a second licensec operator or other qualified member of the technical staff is present at the reactor conteel console and verifies compliance witn the prescribed control rod pattern. MRVEILLANCEREQUIREMENTS 4.1.4.1 The RWM shall be cemonstratec OPERABLE: In CONDITION 2 prior to withdrawal of control rods for the purpose a. of making the reactor cr.itical, and in CONDITION 1 wnen the RWM is initiated curing control rod insertion when reducing THERMAL POWER,. by: 1. Verifying proper annunciation of the selection error of at least one out-of-sequence control roc, and 2. Verifying the rod block function of the RWM by moving an out-of-se uence control rod. By verifying trat the Banked Position Withdrawal Sequence in0ut to the b. - xWM comDuter is correct following any loacing of the secuence program >nto tre computer.

Entry into CPERAi!CNAL CONDITICN 2 and withdrawal of selectec control is permitted for tne Our00se of cetermining tre OPERASILITY of tre P005

"*M "ior t: witncrawal of control r005 for tre ;ar:Ose of orir:i "" the react:- to C*it :ality. ~ 4

- a

."a-e Amendment No.106 s ,4

4 REACTIVITY CONTROL SYSTEMS ROD SE0VENCE CONTROL SYSTEM - LIMITING CONDITION FOR OPERATION - 3.1.4.2 The Rod Sequence Control ~ System (RSCS) shall be OPERABLE. APPLICABILITY: CONDITICNS 1* and 2*#, when THERMAL POWER is less than 10%- l of RATED THERMAL POWER and control rod movement is within the group notch mode ofter 50% of the control rods have been withdrawn. . ACTION: With the RSCS inoperable control r:0 movement shall not be Permitted' ex p+- by a scram. SURVEILLANCE REOUIREMENTS 4.1.4.2 The RSCS shall be demonstrated OPERABLE by: Selecting and-attempting to move an inhibited control rod: a. 1. As soon as the group notch mode is entered during each reactor startup, and -2. As sc0n as the roc innibit mode is automatically initiated during control red insertion.

See 5:e:1a1 Test Exceptica, 3,;C,2.
Entry into COCIT!CN 2 an0 aitn0-awal of selected :Ontrol rocs is permittec for tre cer:ose of cetermining ne CPERAEILITY cf tre RSCS Or'or :: *itrerawal of ::-t 01 -::s fer tne purpose of Ori ge ng ne "ea:ter to cr'.icai'ty.

" AIC" ~ O' $ ~ 2

3. ' ? l-15 Amendment No. 106 I

^ 2 2 c :-. - ~' REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b.. Attempting to move a control rod more than one notch as soon as the group notch mode is automatically initiated during control rod:

1. Withdrawal each reactor startup, and ~' Insertion. Performance of the comparator check of the group notchL circuits c prior to control rod; 1. Movetent within the group notch mode-during each reactor startue, and 2. Insertion to reduce THERMAL POWER to less than 10% of RATED. THERMAL PCWER, ~ '# CH ~ '\\ I 2 3/4 '.-16 Amendment No.106

..{ _.. =, l REACTIVITY CONTROL SYSTEMS RCD ELCOE MCN! TOR LIMITING CONDITION FOR OPERATION 3.1.4.3 Both Red Block Monitor (RBM) channels shall be OPERABLE. APPLICABILITY: CONDITION 1, when THERMAL DOWER is greater than or equal to 30% of RATED THERMAL POWER and when the MCPR is less than the value provided in the CORE OPERATING LIMITS REPORT. ACTION: With one RBM channel inoperable POWER OPERATION may continue a, provided that the inoperable RBM channel is restored to OPERABLE I status within 24 hours; otnerwise, trip at least one rod block monitor channel within the next hour, b. With both REM channels incperable, trip at least one rod block monitor channel witnin one hour. SURVEILLANCE REQUIREMENTS 4.1. 4. 3 ' a. Witn both RBM channels OPERABLE, surveillance requirements are given in Specification 4.3.5. b. With one RBM channel INOPERABLE, the other channel shall be cemonstrated OPERABLE by performance of a CHANNEL FUNCTICNAL TEST prior to withdrawal of control rods. HATCH-UNIT 2 3/4 1-17 Amendment No. 106

+ J. - j - _-3/4.2 -POSER OfSTRfBUTf0N LfMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING-CONDITION TOR OPERATION 3.2.1 ALL AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLMGRs) sha to or less'than their applicable APLHGR limits provided in the CORE OPERATING LIMITS REPORT. APPLICABILITY: CONDITION I, when THERMAL POWER 2 25% of RATED THERMAL POWER, ACTION: With an APLHGR exceeding its applicable limit provided in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and continue corrective action so that the APLHGR meets 3.2.1 within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 ho SURVEILLANCE REQUIREMENTS 4 4.2.1 All APLHGRs shall be verified to be eaual to or less than their applicable APLHGR limits proviced in the CORE OPERATING LIMITS REPORT: a -At least once per 24 hours, b- =Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been estaclished, and Initially and at least once per 12 hours when the reactor is c. operating with a LIMITING CONTROL R00 PATTERN for APLHGR. HATCH - UNIT 2 3/4 2-1 Amendment No.106

e '43 ' - ci,;,. t, M '1 I L-i Figure 3.2.1-1 (Deleted) } Figure 3.2.1-2 (Deleted) i Figure 3.2.1-3 (Deleted) j Figure 3.2.1-4 (Deleted)_ p Figure 3.2.1-5 (Deleted) t. i Figure 3.2.1-6 (Deleted)- Figure 3.2.1-7 (Deleted). Figure 3.2.1-8 (Deleted) Figure 3.2.1-9 (Deleted) l Figure 3.2.1-10 (Deleted) Figure 3.2.1-11 (Deleted) - i Figure 3.2.1-12 (Deleted) Figure 3.2.1-13 (Deleted) -{ -A'"-

N 2 2-2 Amendment No. 106

_ _ _ _ - _ = _ - - - - - -.

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i (This page intentionally blank)

'"~,.,.2 3,'a 2-3 Amendment No. 106 _____________...______.______._________________..________1

g.3 m [, t, 3 '1 E t. e...; t;. s' (This page intentionally blank) I .i l ) ii 1 .i -l s a 'at H UN 2 3,? 2-? Amendment No.106 i

POWERDISTRfBUTIONLXMITS k 3/4.2.3~ MINIMUM CPITICAL POWER' RATIO LIMITING CONDITION FOR OPERATION 3.2.3 -ALL w!NIMUM CRITICAL POWER RATIOS (MCPRs) shall be eoval to or greater-than their applicable MCPR operating limits provided in the CORE OPERATING LIMITS REFORT. APPLICABILITY: ' CONDITION 1, when THERMAL POWER 2 25*; RATED THERMAL POWER k Q t 4 HATCH - UNIT 2 3/4 2-6 Amendment No. 106 3- [

i .. 1; f '- -3/4.2.3 -MINIMUM CRITICAL POWER RATIO (CONTINUED) l ACTION: With MCPR less than the applicable operating limit provided in the-CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and continue corrective action so that MCPR is equal to or greater' than the applicable limit within 2 hours or reduce THERMAL POWER to less.than or equal' to'25% of RATED THERKAL POWER within the.next 4. hours. SURVEILLANCE REQUIREMENTS 4.2.3.1 The MCPR operating limits shall be determined, as provided in the CORE OPERATING LIMITS REPORT, using: = 1.0 prior to the initial scram time measurements for the cycle a, t performed in accordance with Specification 4.1.3.2.a. or b. t. is determinec from scram time measurerents performed in accercance witn Specification 4.1.3.2. The cetermination of the limit must be ccm:leted witnin 72 hours of the conclusion-of each scram time surveillance test recuired by Specification 4.1.1.2. '4.2.3.2 All MCPRs snall be cetermined to be ec.al to or greater than the applica:le limits: .a. At least once per 24 hours, b. W enever THERMAL PCa'ER ha s oeen increasec Dy at least 15*. o f - R'TE0 inERMAL POWER and steady state o;erating conditions have been esta lisbec, and 1 c. Initially and at least ence per 12 hours wne9 the reactor is c;erating with a LIMITIta CONTROL RO PATTERN for MCPR. "A'CH - UNIT 2 3/4 2-7 Amendment No.106

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{ .i i Figure 3.2.3-1 (Deleted) I .) Figure 3.2.3-2 (Deleted) l Figure 3.2.3-3 (Deleted) Figure 3.2,3-4 (Deleted) a i i s v p -l l l " C CH

N: 72 3f:

.7'3 Amendment No. 106

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i j' ', 4<,_t..,j 8 ,t '(*, 1 4 r v: A i. .i-i l~ (These pages intentionally blank) .i I o s l l i i 1 !'l i - o. m U t,. 2 3/4 '.7x 3/4 "3.,- Amendment No. 106 ~ 1

POWER DISTRICUTTON LIMITS-1 3/4.2.4-LINEAR HEAT GENERATION RATE l LIMITING CONDITION FOR OPERATION 3.2.4 ALL LINEAR HEAT GENERATION RATES (LHGRs) shall' not exceed their' applicable LHGR limits provided in the CORE OPERATING LIMITS REPORT. ~ APPLICABILITY: CONDITION 1, when THERMAL POWER 125% of RATED THERMAL POWER. ACTION: With-the LHGR of any fuel rod exceeding its LHGR limit provided in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and continue corrective action-so that the LHGR is within the limit within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the. next 4 hours. SURVEILLANCE RE0VIREMENTS 4.2.4 'All LHGRs shall be determined to be equal to or less than their applicable LHGR limits proviced.in the CORE OPERATING LIMITS-REPORT: a. At least once per 24. hours, When THERMAL POWER-has been' increased by at least 15% of RATED b. THERMAL POWER and steacy state operating conditions have been established, and Initially and at least once per 12 hours when the reactor is c. operating on a LIMITING CONTROL ROD PATTERN FOR LHGR. HATCH - UNIT 2 3/4 2-8 Amendment No. 106 j l l

,-- < s s, t wnu nveg; C_0NTROL ROD WITHDRAWAL BLOCK TNSTRUNENTATf0N' .i-NOTE 1 When the limiting condition defined in section 3.1.4.3 exists. l s.. b. This. function is bypassed if detector is reading > 100 cps or the IRM ~ hannels are on range 3 or higher. c i This function is' bypassed when the associated IRM channels are on c.. 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 V (1 / l l L - 'T:b - UNI' 2 3/4 3-39 Amendment No.106 l l-l l 1

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..:=." . du ~ t IAl:11._4.l.',-I foill_HO.t__noitwill'imailr.1. lii!qpt,suluil rl Ailivl :.intvl !p Aru : nyp!n HI nIS (llAlifil 1 - ~ (H*l NAlllillAt ' CliAtHiI-IlifW. I IINIAl filAi!NII .00NillIlutl5 IN WHICH Illll' lilllC11tW3 Ilit 01: I I *> I Al lititA 110llId I.' Sippyl lli AtKt N(f)lllilts 5, I. A l'.R..H_ a. Ilow Referenced Simulated thermal theer-tipscale ifA 5/il '.H H 'I h. Ipopera t Ive flA $/11 .H . fiA I. 2, T 5/11, I,H - r. Downstale-HA H .R I Ib .l. flentron i lon - tilgh,120 'NA ~ ? /II H 2, 5-2. !!O0 Iti DCK HONIIDH a. lipsca le IIA 5 / 11 ,M H; I* h. Inopera t ive NA 5/Il,,), M t'A g -l(d)' c. Ih>wnsca le NA 5 /11 ,M H I 1. 5Hl.lR_C.f. H.AN.fi[__IN_L.1.1_10.!l_5 a. Detector not full in flA 5 /11II'I.M 'N .- 2. 5 iA 2, 5 ' 's I, ifA 5/l ,W c. i!A 5/11( .l1 . i!A 2, 5 ~' d. IIA 5/ll ,M R

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TABLE 4.3.5-1 (Continued) CONTROL ROD WITHORAWAL BLOCK INSTRUMENTATION SURVEILLANCE REOUIREM . NOTES: Neutron detectors may be excluded from CHANNEL CA'IBRATION. a. L 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 af ter entering CONDITION 2. d. When' THERMAL POWER exceeds 30% of RATED THERMAL POWER. The l additional surveillance defined in Specification =4.1.4.3 will be required when the Limiting Condition defined in Specification 3.l~4.3 exists, e. With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.11.1 or 3.9.11.2. 4 HATCH - UNIT 2 3, : 3 42 Amendment No. 106 I

s SPECIAl. TEST EXCEPTIONS' i 314.10.1 ' PRIMARY CONTAINMENT-INTEGRITY q LIMITING CONDITION FOR OPERATION 3.10.11 The provisions of. Specifications 3.6.1.1 and 3.6.1.3 may be-suspended to permit the reactor pressure vessel closure head and drywell head to be removed and the air lock doors to be open during low power-PHYSICS TESTS with THERMAL POWER < 1% of RATED 1hERMAL POW coolant temperature < 212'F. APPLICABILITY: CONDITION 2, during low power PHYSICS TESTS. { ACTION: With THERMAL POWER exceeding 1% of RATED THERMAL POWER or with the reactor-coolant temperature > 212'f, irmediately actuate the manual scram. s SURVEILLANCE REQUIREMENTS L 4.10.1 The THERMAL POWER and reactor coolant temperature shall be Verified to be.within the 1imits at least once per hour. j i l l l i HATCH - UNIT 2 3/4 10-1

[ f, w s. SPECIAL TEST EXCEPTIONS 4 3/4.10.2 ROD SE0VENCE CONTROL SYSTEM LIMITING CON 0! TION FOR OPERATION -3.10.2 The secuence constraints imposed on control-red groups Ap, A34,- B1E and -B 34 by the Rod Seouence Control System per Specification 3.1.4.2 may be suspended by means:of the individual rod position bypass switches, provided that at least the requirements of Specification 3.1.3.1 and 3.1.4.1 are satisfied, for the following tests: Shutcown margin demonstrations, Specification 4.1.1~, a. b. Control rod' scram, Specification 4.1.3.2a, Control rod friction measurements, and c. Startuo Test Program, with the THERMAL POWER <- 10% of l-c. RATED THERMAL POWER. I i APPLICABILITY: CONDITIONS 1 and 2. ACTION: With the recuirements of the acove specification not satisfied, verify that the RSCS is OPERABLE per Specification 3.1.4.2. j

f SURVEILLANCE'REOUIREYENTS 4.10.2 When the secuence constraints of control cod groups A i

Bj; anc B34 are typassec, verify; l2, A34' That the RWM is OPERABLE per Soecification 3.1.4.1, a. g%ofRATEDTHERMALPOWERisclockedorlimitedtothesin b. at movement of the centrol rods from 50% RCD CENSITY to .w I notch m0de, and c. Conformance with this specification and procedures by a second licensec operator or other cualifiec memoer of tne technical staff. 'ATCH - LN!' 2 3 'a 10-2 Amendment No.106

4 REACTIVITY CONTROL SYSTEMS p 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 reds even under the most unfavorable depressurization-of the reactors. Control rod coupling integrity is reouired to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only oositive means of determining that a rod is properly coupled anc therefore this check must be performed prior to achieving criticality after each refueling. The subsequent check is performed as a backup to the initial cemonstration. In order to ensure that the control red patterns can be followed and therefore that otner 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 nousing failure. The amount of roc reactivity which could be adced by this small amount of rod witndrawal is less than a normal withcrawai increment and will not contribute to any camage to the primary coolant system. The succort is not recuired when tnere is no pressure to act as a criving force to rapicly eject a crive housing. The recuirec surveillance intervals are aceauate to determine that the rocs are OPERABLE and not so frequent as to cause excessive wear on the system components. 3/4.1.4 CCNTROL ROD PROGRAM CONTROLS Control red withdrawal and insertion sequences are establisned to assure that the maximum insequence incivicual control rec.or control rod segments wnich are withdrawn at any time curing the fuel cycle could not be worth encugh to cause the peak fuel entnalpy for any postulatec control rod accicea,t to exceed 280 cal /gm. The specifiec secuences are characteri:ec by homogeneous, scattered catterns of control rod withdrawal. When THERMAL POWER is 2 10% of RATED THERMAL POWER, nere is no possible rod l wortn wnich, if crc;;ed at the cesign rate of the velocity limiter, coulc result in a ceak enthalpy of 280 cal /gm. Thus, requiring the RWV to te OPERABLE ceion 10% of RATED THERVAL POWER anc the RSCS to ce 00ERABLE f om 50'. control roc censity to 10% of RATED THERMAL POWER provices aceouate control. HAT;, - UNIT 2 2 3/4 1-3 Amendment No. 106

j 3/412 ' POWER DISTRIBUTION LIMITS BASES The specifications of this section. assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated ef fects of fuel pellet densification. maintained during abnormal transients.These specifications also assure that fuel ) -3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following-the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46. .The peak cladding temperature (PCT) following a postulated loss-of-coolant accident.is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power districution within an assembly. clad temperature is' calculated assuming an LHGR for the highest powered rodThe peak which is equal to the design LHGR for that fuel type. The calculational procedure used to establish the APLHGR limits for -l Technical Specification 3/4,2.1 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the recuirements of Acpendix K to 10 CFR 50. The Loss-of-Coolant Accident (LOCA) analysis was performed utilizing _the.new improved calculational model, SAFER /GESTR-LOCA. The analysis cemonstrated that loss-of-coolant concerns do not limit the operat on of tne fuel since margin to the 2200'F limit was demonstrated for all of nese i fuel types (Reference 4). Therefore, the APLHGR limits for the fuel types snown in the CORE OPERATING LIMITS REPORT are based on the fuel thermal-mechanical design criteria. A flow dependent correction factor is applied to the rated conditions l APLHGR to assure that the 2200'F PCT limit is complied with during a LOCA Initiated from less than rated core flow. In aedition, other power and flow dependent corrections are applied to the rated conditions APLHGR limit to .l assure that the fuel thermal-mechanical design criteria are preserved during aenormal transients initiated from off-rated concitions. For single-loop operation, a multiplicative factor is applied to the rated concitions APLHGR limit for all fuel bundles when core power exceeds a specified value. The power and flow-cepencent correction factors and the limiting values for APLHGR for each fuel type used in a particular cycle are specified in the CORE OPERATING LIMITS REPORT. HATCH - UNIT 2 8 3/4 2-1 Amendment No.106

) '.n-q. .l. - l l i i '{: i s i (This page intentionally blank) i i i I 4 4"> - LN:T2 2 3/4 2-2 Amendment No. 106

> ;,; y e + POWER DISTRIBUTION LIMITS-pg ;,s [' BASES' i .3/4.2I2 i APRM SETPOINTS This-section' deleted. 3/4.2.3 MINIMUM CRITICAL-POWER RATIO The.recuirec operating limit'MCPRs at steady state operating conditions as specified in Scecification 3.2.3 are cerived from the establisned fuel claccing integrity Safety Limit MCPR of 104 for two-loop operation-and h05 for single-loop' operation, and an analysis of abnormal operational transients: (Reference 1). For any abnormal operating transient analysis evaluation with- ! (specified in the CORE OPERATING ~ LIMITS REPCRT), it is. requi I. resulting MCPR coes not cecrease celow the Safety Limit MCPR at any time 'during the transient assuming instrument' trip settir; as given in = Specification 2.2,1, '~ .To assur_e. that the fuel clacding-integrity Safety Limits are not violated l during any anticioated abnormal operational transient, the most limiting' transients'have oeen analy:ec to cetermine which results in the largest reduction in CRITICAL DOWER RATIO (COR). The type of transients evaluated were loss of. flow, increase in pressure and power, positive reactivity insertion, and coolant temperature cecrease. Details of hcw evaluations are performed, the metnods used, and how the MCPR limit is acjusted for operation at less than ratec power and flow conditions are ghen in Reference 1 and in tne CCRE CPERATING LIMITS REPORTS, HATCH - UNIT 2 B 3/4 2-3 Amendment No. 106 - J

g ,3 1, :- y POWER'0!STRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) i j -l ~As depicted in the CORE OPERATING LIMITS REPORT the 100% power,100% flow l operating limit MCPR (OLMCPR) depends on the average scram time, t, of the control rods, where: = 0 or ' ave

B, whichever is greater t

tA

B i

'A = 1.096 see (Specification 3.1.3,3, scram time limit where: to notch.36) B= + 1.65 N) 1/2 t li n EN I i=1 where: u = 0.822 sec (mean scram time used in the transient analysi5) e =.013 sec (standarc ceviation of p) n int$ 4 t,v,, 3,) n IN I i=1 wnere: n = number of surveillance tests performec to cate in the cycle i = number of-active control. rods measured in the ith surveillance test T i = average scram time to notch 36 of all rods in the ith surveillance test N 1 = total numoer of active rods measured in 4.1.3.2.a. HATCH - UNIT 2 B 3/4 2-4 Amendment No. 106

l. POWER DISTRIBUTION LIMITS s-

BASES,

References:

1. " General Electric Standard Application for Reactor Fuel (Supplement for United States)," NEDE-24011-P-A. HATCH - UNIT 2 5 3/4 2-6 Amendment No. 106

yb i f s:.. c. 3/4.4 REACTOR COOLANT SYSTEM BASES-3/4.4.1.1 RECIRCULATION SYSTEM Operation with a reactor coolar.t recirculation loop inoperable is ~ allowed, provided that adjustments to the flow referenced scram and APRM rod l block setpoints, MCPR cladding integrity Safety Limit, MCPR Operating Limit, i and MAPLHGR' limit are made. An evaluation of the performance of the ECCS with i single-loop operation has been performed and determined to be acceptable, l Reference 1, The adjustments to the APLHGR and the MCPR limits that are required for single-loop operation are provided in the CORE OPERATING LIMITS REPORT. The~ flow referenced simulated thermal power setpoint for single-loop operation is reduced by the amount of m4W, where m is the flow reference slope for the rod block monitor and 4W is the largest difference between two-loop and single-loop effective drive flow when the active loop indicated. flow is the same. This adjustment is necessary to preserve the original relationship between the scram trip and actual drive flow. The possibility of experiencing limit cycle oscillations during single-loop operation is precluded by restricting the core flow to greater than or equal to 4S4 of rated when core thermal power is greater than the 80% [ rod line - This reovirement is bascd on General Electric's recommendations . contained'in SIL-380, Revision 1, which defines the region where the limit cycle oscillations are more likely to occur, 3/4.4.1.2 JET PUMPS An inocerable jet pump is r.ot, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazare in case of a design basis Loss-of-Coolant Accident by increasing the blowdown area-and eliminating the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperaole is nece s sa ry, One of the acceptable procedures for jet pump failure su veillance identified in NUREG/CR-3052, Reference 2, was adopted for Haten Unit 2. The surveillance is performed to verify that neither of the following conditions occur: (a) The recirculation pump flow / speed ratio deviates by more than 5% from the normal range, or (b) The jet pump loop flow / speed ratio deviates by more than Si from the normal range. If eitner criterion is failed, then the procedure calls for comparing either the individual jet pump flows or tne indivicual jet pumo ciffuser to lower plenum differential pressures to the criteria of the Limiting Conditions for Operation (LCO). If tne LCO is not satisfied and pump speed is less snan HATCH - UNIT 2 B 3/4 4-1 Amendment No. 106

-ADMINISTRATIVE CONTROL 5

s.

o. e.' Type of container, e.g.,-LSA. type A, type 8, large ewantity f. Soltelfication agent. e.g.. cement. The Raetoactive Eff1went Release Report shall inc1 wee (on a ovatterly t> asis) unplannec releases from the site to unrestricted areas of radicactive materials in gaseous and lievid eff1wents that were in excess of 1 C1,. enc 1weing etssolved aed entraired gases anc tritium for liowie eff1wents, or those in excess of 150 C1 of noole gases or 0.02 Ci of raciciocines for gaseows releases. The Radioactive Effluent Release Report shall include any changes to the P't0CE55 CONTROL PROGRAM and to the CFF51TE 005E CALCULATION MANUAL mac during the reporting period. MONTWLY OPYRAT!NG REPORT 6.9.1.10 'Routtne reports of operating statistics and shutdown experience shall be submitted on a monthly casts to t.no Director, Office of Management . and Program Acalysis, V. 5. Nwelear Regulatery Commission Washington, D. C. 20555, with a cocy to the Regional Office of Inspection anc Erforceeeat no later tman the 15tn of eacn monta following the calencar month covered Dy the report. CCRE CDERATINO LIwit5 rec RT 6.9.1.11.a. ' Core ecerating limits stall to establisbec one documented in the CORE OPERAT!k3 LIMITS REPCRT before each reloac cycle or ary remainteg part of a relcac cycle for the follecteg: (1) Coet ol Roc Prog *am Controls - Rod Blocit Monitor for Scecificat,1cn 3.1-4.3, (2) Tre average Diarar Lirear West Generation Rate for 5:ecification 3.2.1 anc Surve1114 :e Renwirement 4.2.1, (3) The Mirimwm Critical-Power Ratio for Specifications 3.1.4.3 anc 3.2.3 anc Sweveillan:e Requirement 4.2.3, anc (4) Tne Linear weat Geae atien Rate 'or Scecification 3.2.4 are Swrveillan:e Regwirement 4.2.4 Tne analytical metnoes usec te cete + me tne c:re c:erating b. limits small te snose :revie sly reviewed anc accrovec cy ve NRC in the fol'o.te; cc: meats. (1) NEDE-24011-D-A, "0eaeral Eiectric Staaca c Aced:ation for Reactor twel," (acolicaele amerennt scerifie: in the CORE OPERATING LIMITS RE80RT). (2) " safety Eva16atter by the Of fice of Nwelear Reactor Regulation Sw:certing Amercment Nes. 151 anc 89 t,o Fa:tlity 0:e atieg Licenses ODR 57 anc N;:-5." cated Janwary 22, 1988. c. Tne core ecerating limits shall be cetermine: so snat all a:cli:aele limits (e.g., fuel theemal-ee:nani:41 limits, cc'e - trermal-9yerau11c limits. E 5 Imits, nuclear limits sw:n as snute:we margin, anc transiert er,c accicent analysis limits) of the safety aealysis are met. c. Ta e CORE 05ERATING Llw' 5 RE:0RT, in:1 weir.g any mid-cycle ,evistors er su::le-eets treaeto, snall te revicec w:op 'ss.,aa:e. 'er ea:n reica: cy:le, to tne NRC Occueens Control tess itn :::+es t: tae Reg'enal Acminist-ator and Resicent Ias:e:t:. -ATM - N T 2 6 14e Amendment No. 106 e l ..}}

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