ML20244A629

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Amends 790322 Submittal of Tech Spec Changes in Support of Reload 3 Licensing Application.Class IV Amend Fee Encl
ML20244A629
Person / Time
Site: Hatch Southern Nuclear icon.png
Issue date: 05/11/1979
From: Whitmer C
GEORGIA POWER CO.
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 7905190213
Download: ML20244A629 (32)
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Text

{{#Wiki_filter:l c ( Gw;.a Raer CompaN 230 Fervee Street .g NU C"ce Box 4545 i A:I1c;3. Gecrg<a 30303

  • rectone 404 522 6060 04 May 11, 1979

, Chas. F. Whitmer Georoia Power UCe PfeSd6Fi En;meeeng me L v e e:'cs,vem s Director of Nuclear Reactor Regulation U. S. Euclear Regulatory Commission f, Washington, D. C.. 20555 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 EDWIN I., HATCH NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATIONS CHANGES IF SUPPORT i 0F RELOAD-3 LICENSING APPLICJf;3N l 1 Gentlemen: ) i Pursuant to 10 CFR 50.90, as required by 10 CFR 50.59(c)(1), Georgia Power Company hereby amends our submittal of March 22, 1979, which pro-posed Technical Specifications to incorporate changes as a result of the i' reload transient and accident analyses performed for cycle-4 operation. These analyses took credit for incorporating modifications in -design of recirculation pump trip logic and neutron monitoring instrumentation into the plant. The attached proposed changes to the Technical Specifications y are required to support the analyses performed for cycle-4 operation. These changes will be described briefly herein and in more detail in anclo-sures to this letter. ..w The first of these changes to the Technical Specifications -(Enclosure 3) j is necessitated by the installation of the modification to the Recirculation Fump Trip (RPT) system. These RPT modifications are designed to improve fuel thermal margin by tripping both recirculation pumps upon sensing turbine j stop valve closure or fast turbine control valve closures. A description of the modification and its impact on the plant safety analysis is provided in t Enclosur.e 2. This new design feature has been credited in the cycle-4 transient analysis and w111 also be applicable to subsequent reload submittals. The second proposed change (Enclosure 5) is associated with a modifica-tion to the Average Power Range Monitor (APRM) high-high flux scram trip logic. A description of the modification and its effect on the plant safety analysis is provided in Enclosure 4, As discussed therein, the flow referenced logic s design will be modified with a new logic scheme that will minimize inadvertent scrami caused by neutron flux spikes without reducing plant safety. This new design feature has been credited in the cycle-4 transient analysis and vill also ba applicable to subsequent reload submittals. The last of the propcsed changes (Enclosure 6) provides for spiral unloading and reloading the core under spect.a1 conditions without having 3 + el 5 cP ~ s w t. o e m p( p^ 7905190 A/p

(- ( a V Georgia Powrd 1 - Director. of Nuclear Reactor Regulation U.-S. Nuclear Regulatory Commission May 11, 1979 Page Two counts per second registering on the source range monitors. Specifically, the proposed changes to the Technical Specifications would allow spiral j unloading of the core to continue after the source range monitors readings ~ decreased below 3 counts per second. The proposed changes would also allow two bundles to be reinserted to their previous positions immediately adjacent to each source range monitor prior to requiring 3 counts per second on the source range monitors. Spiral reloading would then proceed. Operability j of the source range monitors electronics would be verified by injection of a simulated signal into each channel prior.to the beginning of fuel reloading. The proposed changes to.the Technical Specifications enable the er-fre core to be unloaded and reloaded without the need for inserting portable " dunking chamber" monitors or a source and thus reducing the possibility of dropping-items into the reactor with an ensuing impact on the outage schedule. The Plant Review Board and Safety, Review Board have reviewed and approved these proposed changes 'to the Plant Hatch Unit 1 Technical Specifications contained in Enclosures 3,' 5, and 6 'and have determined that they do not Involve an unreviewed safety.qtestion. In the case.of the RPT the prompt tripping of both recirculation pumps and tht ~resultsiit increase in the core ] void fraction act to decrease the MCPR for pressurization transient and j adequate margins are maintained. The modification to the APRM results in j the maintenance of adequate thermal margins for fuel cladding integrity and the reduction of the cyclic duty of the reactor vessel and fuel by minimization of the number of spurious scrams. The proposed changes to Section 3.10.C of 1 the Tcchnical Specifications does not. affect the existing scrams which would act ] to terminate en inadvertent critical. The proposed changes provide further j assurance of subcriticality by initially loading eight bundlec into their previous location in a demonstrated 1y suberitical configuration in order to achieve an on-scale reading for each source range monitor prior to continuation { t of fuel reloading. Thus,~it can be concluded that the probability of occurrence i or the consequences of an accident or malfunction of equipment important to i saf ety is not increased, nor is the possibility of a new accident or malfunction of equipment important to safety created. 1 contains an evaluation,of submittal fee for these changes associated with the Reload-3 Licensing Application of March 22, 1979. 1 I l 1 k 1 l

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) . Director of' Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission May 11, 1979 Page Three These modifications will be completed prior to'startup following' the' current refueling outage. Therefore, your review of this submittal in a timely manner to support the. cycle-4 reload application will be appr6ciated. Yours very truly, } c A Ch9. F. Whitmer RDB/ WEB /TMM/mb Enclosures xc: Mr. Ruble A. Thomas George F. Trowbridge, Esquire E, worn to and subscribed before me this 11th ' day of May,1979. Lk. s ~ . V ' Notary Public Notorf PubPc, Georgia, State at Large My Commission Expires Sept. 29,1981 \\ i-( t

> j 77-r r ll ;o lHQ(, Q1\\i?A [(}'- +. m . 3 3; .c. 4 'W w c ,q g ~i) ; e i! - l ENCLOSURE-1 .t. 4 d i ,NRC DOCKET =50'-321 OPERATING LICENSE:DPR. 1 . EDWIN I'. HATCH NUCLEAR PLANT-UNIT 1 1 PROPOSED 4 CHANGES TO TECHNICAL' SPECIFICATIONS: 1 . Pursuant to110 CFR 170.12(c), Georgia-Power Company.has evaluated the attached j' proposed-amendments ~to Operating License DPR-57 and haverdetermined that.in-s s conjunction with the. Reload-3 Licensing Application 'of March 22,'19.79,~these e T changes;. constitute 'a Class IV! amendment. -These' changes with those previously~ q submitted:in the March-22.' 1979; letter form a. cycle-4 licensing package. " g~ d ^which. involves a complex issue of moreithan one safety issue.' Enclosed isu, a check for?$8,300.00. as.l payment of;the. balance. due for a Class 1IV amendment.. ' The: March 22,:1979,yletter transmitted a $4,000.00 initialspayment as a 1 Class.III amendment to. allow review during the. interim'between' initial'and complete submittals of. the cycle-4 licensing package. 1 i 'l q a e I l } l H 4 h lt; J ___ _1-:-

)N> j u k ENCLOSURE 2 -i 4 l-b:L EDWIN I. HATCH NUCLEAR PLANT UNIT 1 N . RECIRCULATION PUMP TRIP. l i l ~ - The. Recirculation Pump Trip (RPT) System is to be installed at the l Edwin I. Hatch Nuclear Plant, Unit 1, during the Cycle 4 fuel reload. The RPT system is designed to improve fuel thermal margin by tripping both recirculation pumps upon sensing stop valve closure or fast control ] valve closure. The reduced core flow reduces the void collapse in the I core during the pressurization ever.ts analyzed for each reload. Tripping of the recirculation pumps results in a smaller net positive void reactivity addition to the system during these pressurization events..This results in a lower power increase and consequently a smaller decre'ase in MCPR. Although the reduction in core flow in itself may cause a slight decrease in thermal margins, the effect of reduced flow on the power increase is a considerably more dominant effect and the net result is to red. ace the thermal severity of turbine trip, generator load rejection, and feedweter controller failure events. 1. SYSTEM DESCRIPTION The RPT system includes all the equipment that trips the recirculation pump motors from their power supplies in response to a turbine stoa valve or fast control valve closure. The RPT system is connected I to the reactor protection system (RPF) funct' ions which provide fuel barrier integrity protection for the same events. The system consists of turbine control and stop valva closure sensors, seprate division logic relays, bypass switches and two circuit brea%r; for both pump motors. The RPT system is designed to be opere'a whenever the turbine generator trip scram is operable. Existing turbine first-stage pressure sensors will prevent RPT initia*.ico for turbine generator trips occurring below the existing 30% pewr ~i bypass of turbine and generator trip scram signals. A logic diagram is shown in Figure 1. The RPS inputs which sense turbine stop valve closure or turbire control valve fast closure utilize four RPS logics as noted in Subsection 7.2.3 of the FSAR. These inputs are received by the two separate RPT trip divisions. Each RPT division has two se m te trip channels, sensors and associated equipment for each mev a -d i A variable. The primary trip channels and division logic eier mt-are f 6st-response, high-reliability type relays which are cm ble dith those relays used in the reactor orotection system. Sen c"" and associated equipment are designed to be highly reliable, arc the components are of a quality that is consistent dth minie.:- l maintenance requirements and low failure rate. Channel inputs from each variable are combined in two-out-of-two configurations. i + f ( C______._,__.___.__

.c y 1,. . O. O - .g. ,a Each recirculation purp trip' system logic energizes a trip coil which, in turn, opens the asscciated circuit breaker. The circuit breakers include current-interrupting devices, instruments, isolatin; devices, electrical power control components, and incoming and outgoing connections. This equipment provides electrical tripping 7., functions for the recirculation pump rrctors. There are two identical recirculation system loops, each of which has two circuit breakers for redundancy. Each RPT system can trip both recirculation punps. The breakers are required to open and trip a variable-frequency recirculation pump motor fed from a variable-frequency motor generator set. The_breaxers are required to interrupt load current within t 135 mi'lliseconds of the initiating event. The system is designed so that it may be tested during plant operation from sensor device to final actuator logic. One stop valve or control valve can be closed and system status tested.by observing / relay contact status and logic test light status without causing RPT operation. The entire logic of one division and input sensors ( of the RPT system may be tested without tripping the pumps by placing that system bypass switch briefly in the "inop" position for the duration of'the test; the test is initiated by closing the T two stop valves which initiate that system to the 10% closed (90% open) position. The control valve fast closure pressure switches may also be tripped. Successful completion of the test is indicated by annunication of "RPT initiate" as the annunciation relays are 72 L energized. During this brief interval, the redundant RPT syste, is, of course, continuously available to perform its safety function. . L '.,. Circuit breakers shall be tested a.5.per the p.lant technical specifi-

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The RPT system utilizes the RPS power supplies for the norma'l', energized logic and essential ESF 125-Vdc for the breaker trip coils. The 125-Vdc is supplied by two separate divisions of stat e-batteries which are also utilized by the RPS and ECCS systems. 2. CONFORMANCE TO IEEE STANDARDS AND REGULATORY REQUIREMENTS . f. The RPT system was designed to comply with IEEE Standards 279. 373 338 and 379 and General Design Criteria 13,19 through 24, and 2's of 10CFR50, Appendix A. In addition, the design complies with Regulatory Guides 1.22, 1.47 and 1.53. The system is designed to meet the single-failure criterion such 9 that any single trip channel'(sensor and associated ecui est) c-Jr system component or power supp'y failure shall not prevent the D, ; system from performing its intended safety function. Electr:mec anical .,,i relays used as the logic elements within the system an1 the.syste-logic are failsafe '(i.e., trip on loss of electrical PSer). Iba 1 RPT system is designed to acccm31ish the desired functier with a I 1 minimum effect on plant availability. The system logic is cesi n d I ' ? such that it will not cause the inadvertent trip of mere tha9 cw l 'y pumo given a single component failure in the system. Each '. rip l 'a l .i I y i r.;t 4 t e. J

i. -n .s' "A i 4 division shall be cle.u'lv identified to reduce the possibility of inadvertent-trir 'ot~ the 5 circulation pump during routine maintenance and test operatiens. Redundant circuits in each system (sensors; wiring, relays, p *0F tupplies, circuit breakers, etc.) are electrically, mechanically, aN?. physically independent so that they cannot be ~ disabled by a si gle eg ent, 1 C.o"ntrol Room annunciators are provided to identify the tripped portions of RPT in add' tion to the instrument channel annu9ciators associated with the. R.M described in the FSAR. These same functions 'i the system status.'he P"ecess computer to provide a' typed record of 3 are connected t*a + Annunciation will also be provided to indicate that part of a system is not operab'e, The system has annunciators, lighting and sounding whenever one t"ip system is bypassed. Indicater lights. l are provided fur logi

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All' bypass and inope"atility indicators both at the division level and the component leW will be grouped for ' operational convenience. i As a result of cesi 9', creoperational testing, and sta tuo testing, J no erroneous bypass 4 7:# cation is possible. These inci:stion provisions serve te 5-201ement administrative controls ar:: aid the .i operator in assessi's t'.e availability of component anc sjstem 1 t level protective at; 2 3, The annunciator i-< t' 3t 'on signals are provided throu;

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- n preventprotectfW actions. 's!P ng wi'i devices and can d' I 'I alth te 02'.c basis when equipoent associa.i: be included on 3-o-- indication is tes:<; 3. EFFECT OF RPT Cs - :EPJORM N E 'I An inherent des 4 g-C'2 acteristic of the boiling wate- -uctor -) (DWR) is the re? ati zvip of the core average moderat:- ce95ity to .a is represented by a negative.:id. reactivity ? neutron moderati:, .-e lead following w:Nnive void reactivity coef ficier' :ermits-?' contro coefficient. T.mi $ control rbd movr *- To increase power, core flow is #

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tion l which decreases t C 'C c f raction and increases the ns t*:n " ode-a and reactor powe" 2' Oressuri:ation transients, the :: e pressure increases rapic' ' - CA. sing a reduction in the cora away vcid ,raction, i The basic phen 0't'?' cssociated with veid feedback is ; i decrease in neutron mnce s t :-' esulting from an increase in k: : f ra tt j e.9. 1

-'-s neutron fli x cccurs wherein t's eerma i i

A spectral shift jecreases and the Efc'e'N5I flux, and hencE t'i ec.icn rate, flux, and hence ; e Oih9anceCa;;tJrerate,increasc!. Mnverse'y, j he i C h. cases an 6ncrease in re:-:;.' y.Jnction of t7 m adabie] a decrease in v c ~ --.:or void coefficie t - inart'y the # for any fixed b. [ etry: '(1) tre average voi:j

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n 'Q - 4 '- U i l and (3) exposure. As each of these three parameters increases, the e absolute magnitude of the void coefficient increases and becomes more negative. For pressurization transients, the rate of flux rise is dependent l on the magnitude of the void coefficient. The more negative the The rate at vo,id coefficient, the greater the flux' rise rate. which the negative reactivity can be added to the core by the scram j determines the severity of the transient. The scram reactivity 1 on the ability of the control r,ds to be in the high flux, depends The minimum scram reactivity occurs at end of regions of the core. In this cycle when control rods are fully withe awn for the core. situation, it takes a longer time for the control rod to travel to For this reason, the pw'suri-a high importance region in the core. zation transients.are most severe near the end of the cycle. l The degree to which the pressure and thermal margins are reduced during pressurization events depends on the tradeoff between.the' negative scram and positive void reactivities. Typically, at beginning of cycle (BOC), control rods are partially inserted; this permits a prompt shutdown of the. system without a'significant !l As the fuel cycle proceeds toward end of decrease in margins. cycle (EOC), the cor. trol rods are withdrawn ur cil, ideally, they I Hence, the effectiveness of' scram reactivity I are all withtsrawn. for shutdown of certain pressurization transients is decrea. sed.as the core approaches EOC conditions. Hence, toward the end of cycle, tha. void r.eactivity feedback can momentarily add positive reactivity to the system faster than thr. 1 control rods add negative scran reactivity. d Prompt tripping of both recirculation pumps results in a rap d-reduction in fl Ore flow. Tnis increases the core void 'raction during pressurization transients.and consequently minimizes the l J Thus, the decrease in f4CPR for pressurization power rise experienced. In addition, installation of RPT will' . f transients is reduced. increase the margin to the vessel pressure limit experienced during Because the RPT system is not initiated pressurization transients. loss of coolant, or vessel over- - j during the rod withdrawal error, pressure protection events, ttere will be no effect on these ar1alyses. J \\ l I 1 I I 1 JSC:gmm/549-552 l 3/20/79 { i I r, b b a l ~ _ _ _ _, h

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] ag _ + a; - ,s. n. ,w. .i I,. L x . 3w. ,1 -s . ) o p' ' ', ..'U; i .EMCLOSURE 3 - t u-NRC DOCKET 50-321 EDWIN I. HATCH NUCLEAR PLANT UNIT 1 OPERATING LICENSE DPR-57 i, PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS The proposed changes to the Technical Specifications ( Appendix A 'to Operating . License DPR-57) would, be incorporated as follows: Remove Page Insert Page 3.2-20 3.2-20 '3.2-45 3.2-45 3.2-67 3.2-67 4 e 4

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" ".1 C m 7- -s 1 ~ ~ ~ - ~ ~ ~ ~ ~ - -~R~ ~ 'f' a 1 BASES FOR LIMITING CONDIT10N5 FOR OPERATION r t. jp,E3.21H.5D ; Main ~ Steam Line Radiation' Monitors -(Continued)' ' the react'or throughithe main steamilines toL the: condenser. Two'instrumen't-channels with two: radiation. detectors in each channel are: arranged in a one-upscale per channel trip ' logic'. (one-out-of-two-taken-twice). The trip settings are based'on: limiting the release of radioactivity via the normal ventilation path and rerouting this1 activity to'be processed through the standby gas. a . treatment: Sys tem. I. Instrumentation Which Initiates Recirculation Pump Trip:(Table 3'.2-9) h, cThe ATWS recirculation' pump trip.has been added 'at the suggestion of ACRS as.a uleans! v' of ' limiting: the consequences of the unlikely occurrence of a failure :to. scram during an anticipated transient. The response of the plant-to this postulated ~ event falls within'the envelope of study events given in' General Electric l 'l Cori1pany. Topical Report NED0-10349, dated March, 1971 and~ Appendix L of the FSAR.- A prompt recirculation pump trip (RPT) has been installed which trips both.recircula ~ tion pumps rupon sensing turbine stop valve or fast control valve closure. :The prompt. J .RPT is' comprised of two separate systems,.each system capable of tripping both recir - culation~ pumps. J.- Instrumentation Which Monitors Leakage Into The Drywell (Table '3.2-10) -l'. Drywell Equipment Drain Sump Flow Integrator ' The equipment drain sump is provided with twoisump pumps. A flow integrato is j .provided on the' discharge header. The starting of each sumpipump and high sumpi h level 1are annunciated in the control room.. The restarting frequency. of a pump ' q .) motor, in conjunction with the predetermined volume'of liquid pumped out during. g .each period, provides an alarm in the main control room indicating when the - i identified leakage rate limit is reached. 2, Drywell Floor Drain: Sumo Flow Integrator 1 The floor drain sump'is provided.with two sump pumps. A. flow integrator is 'provided on the discharge header..The starting of each sump pump;and high sump 4 level are annunciated in the control room. The restarting frequency of a pump . motor, in conjunction with the predetermined volume of liquid pumped out during i each period, provides an alarm in the main control room indicating when the unidentified leakage rate limit is reached. ~ 3. Scintillation Detector For Monitoring Air Particulate . A sodium-iodine scintillation detector contained in an instrument rack is used l to monitor the release.of airborne radioactive particulate in the arywell and '3 torus. A high radiation level reading is indicative of a' leak in the nuclear system process barrier in the primary containment. A sample that is continu-ously drawn from'the primary containment is collected on traveling filter paper and' monitored.by a gamma sensitive scintillation detector. Radiation levels are read out by a log rate meter and recorded on a strip chart located in the control room. A high radiation level alarm and a failure alarm are provided and are annunciated in the control room. Also, a high-low flow alarm is pro-vided which' annunciated in the control room. 4 L$ scintillation Detector For Monitorino Radioiodine 7 N A sodium-iodine scintillation detector contained in an instrument rack is used to l monitor the release of radiciodine in the drywell and torus. A high radiation 3.2-67

9 9 L, - EHCLOSURE a 1 [ ' HEW APRM SCRAM TRIP LOGIC P: -FOR THE .EDWIN I. HATCH NUCLEAR PLANT UNIT 1 i 1 1. PURPOSE 'This report describes the new APRM scram trip logic and discusses' ' the impact of this logic on plant safety analyses. This APRM logic is being installed in the Edwin I. Hatch Nuclear Plant Unit.1 for operation during_ Cycl.e 4 and all subsequent cycles. This report shows that the new APPdi scram trip logic will reduce fuel and reactor cycle duty without compromising the safety of the plant. 2. BACKGROUNO 4 ) Scrams have been reported at operating'8WRs at a result of,momen-tary anomalous neutron flux spikes which exceeded the high-high APRM flow referenced trip setting. Frequent causes of these flux 1 l spikes are momentary flow changes in the ' recirculation system flow and small pressure disturbances during turbine stop valve and turbine control valve testing. Although many of these scrams occurred during operation with less than rated core flow, the neutron flux did not exceed the 100% flow flux senm trip value (120%) assumed in the-transient safety analysis. These small neutron flux spikes represent no danger to the fuel because their duration is less than the fuel thermal time constant. Therefore, the fuel surface heat flux does not i,n, crease Lu.fficiently to ) challenge the fuel cladding integrity safety limit. The new APRM. l scram logic will reduce the number of spurious scrams occurring along the power-flew line without reducing the fuel safety margins ] for any accidents or abnormal operational transients for which the plant is licensed. 3. DESCRIPTION The APRM flow referenced scram feature was designed and installed on Hatch 1 as noted in subsection 7.5.7.3.4 of the Final Safety Analysis Report. The trip setpoint is varied as a function of reactor recirculation driving loop flow relative to a value of 12C% of nuclear-boiler rated power at full flow. However, the transient j analysis provided to the NRC for Hatch I does not take credit for i this flow referenced scram and the 12C% high flux scram trip was l assumed to be independent of the recirculation. flow. t There are six APRM channels, three for each reactor protection trip system. The trip unit for one of these three channels can supoly the trip signal to the associated reactor protection trip system. At least one APRM cnannel in each trip system must trip to cause a Presently, each APP.M channel cerives its trip signal from l scram. l l LMZ:at/57J l i l 1

e--.

~ - ~ ~ ~ - ~ ~ - ~ - ~ ~

l.PRM neutron flux asurements. The APRM scram ?.r setpoints (given in technical specification 2.1.A 1.c) are varied as a func-To obtain the tion of reactor recirculation driving loop. flow. proper reference signal, each APRM channel is staplied with two redundant'and isolated flow signals associated with the trip system, p Although the present APRM flow reference scram system accuratel'v predicts the thermal' power level for steady-state operation,-it overpredicts the fuel heat powerlevel during power' increase events. During such events, the neutron flux leads the reactor thermal power because of the fuel time constant. Therefore, neutron flux trip levels are reached before the reactor thermal power has signifi-cantly increased. While this anticipatory response in the APRM ~ 1 scram provides additional protection to the core during abnormal operational transients or accidents, it results in many unnecessary f scrams for small neutron flux disturbances along the flow control j line (F,igure 1). Many of.these unnecessary scrams will be avoided by replacing'the present APRM trip logic with a Thermal Power Simulator and an APRM Simulated Thermal Power (STP) trip unit. The APRM signal.for the scram trip will be processed through a Thermal Power Simulator. consisting of a time constant delay circuit. This circuit repre-sents the fuel dynamics which will approximate the reactor thermal A faster power during a transient or steady state condition. response trip unit on APRM neutron flux utilizing a non-flow l referenced 120% neutron flux scram trip setpoint has also been added. Figure 3 illustrates the new APRM scram trip logic. Figure 2 shews the response of the new APRM scram trip logic to the y ~ i same flux spike as shown in Figure 1. In this case, a scram does not occur, since the transient peak of the simulated thermal power is below the flow referenedd scram setpoint. The total recirculation drive flow signal to the APRM STP trip unit Therefore, an APRM channel trip could be initiated i remains'the same. from either a non-flow referenced 'APRM neutron flux trip unit or i the flow referenced APRM STP trip unit. 4. CONFORMANCE TO GUIDES AND STANDARDS The electrical components used in the new APRM scram trip logic l circuitry are in conformance with all applicable IEEE Standards, with all applicable NRC Regulatory Guides, and with the Code of i Federal Regulations, Title 10, Chapter 1, Part 50, Appendix A. The new APRM scram trip logic was qualified to the following codes by which Hatch 1 was ifcensed: Criteria for Protection Systems for Nuclear a) IEEE 279-1971 - j Power Generating Stations. General Guida for Qualifying Class 1 Electrical b) IEEE 323-1971 - Equipment for Nuciear Pcwer Generatirg stations. ) V i L"Z: at/57J 2 ~ ~-- - --

2EEE 344-1971 h Recommended Practices for the E ismic \\ c) ' Qualifications) Class 1 Electric Equipe nt f H uclear Power Generating Stations. Both AC and DC APRM power supplies remain the.same. 0 The Average Power Range Monitor (APRM) system is one subsystem of the' neutron monitoring system. The APRM setsystem is augmented to include the Simulated Thermal Power Trip (STPT). The APRM sub-system has 6 APRM channels, each using input signals from a number of LPRM channels. Three APRM channels are associated with eacn of ~ the trip systems of the Reactor Protection System. The APRM subsystem is designed to meet the requirements of IEEE-279 as documented in Subsection 7.5.7.4 of the FSAR. f The STPT augments each of the 6 APRM channels such that each APRM channel has a 120% neutron flux trip whose setpoint is not recircu-lation flow biased. The new thermal power upscale trip has a setpoint that is flow biased and is obtained by filtering the APRM signal to obtain a signal which represents the thermal flux of the fuel. This time delay is accomplished by conditioning the APRM neutron flux through a first order low pass filter that has a 6 second RC time constant. Since each of the 6 APRM channels was identically modified to add the STPT, and' the independence between the 6 APRM channels was not altered, the redundancy requirements of IEEE-279 are still maintained. 5. EFFECTS ON SAFETY ANALYSIS Cumulative fatigue damage analyses are performed for the fuel assembly, the reactor and reactor internals. The cyclic loads considered in these analyses ' include ctclant p~ressure and thermal gradients. Details of the methodology used for the fuel analysis are given in Reference 1. Reactor and reactor internal analyses are addressed in Acpendix C of the FSAR. Avoidance of spurious scrams will reduce the plant cyclic duty and will, therefore, Because provide additienal margin to the fatigue damage limits. the present abnormal operational transient analyses do not account for the present flow-reference APRM scram setpoints, the only l transient which is affected by the new APRM scram trip logic is the loss-of-feedwater heating transient. Because the flow referenced scram with the new APRM scram trip logic is set at 117% neutron flux at 100% ficw, a scram will occur earlier for slow transien.5, before the fixed APRM scram setpoint of 120% neutron flux is reached. l Therefore the transient ACPR will be less for the most limiting ' slow transient, a loss-of-feedwater heating. At less than rated power conditions, the new APRM scram trip logic provides greater thermal margins to the fuel cladding integrity safety limit than at rated power. As reactor power is reduced, I In the loss-of-feedwater heating total steam generation decreases. transient, the reduced steam flow at low power results in a decrease in both feedwater flow and the maximum temperature rise across a I LMl:at/57J 3 1 V N --~--~.~...._.._.,,....,;.... j

4; 4,e ~ ~ ~ - j p s v.. m <,, r ..[(} ~ ' ' ' L) i d' given heater. The core'subcooling' change, as well as the positive: 1 r.cactivity insertion,.will.then'be-less severe. Therefore, the 'i change (in critical power ratio'(ACPR) will' decrease with decreasing. power.. Thus,.the; difference.octween the safety limit and the H , I' ' transient MCPR increases with' decreasing. power, irrespective.of.the-H m scram systems logic. 1 In addition, at anyLgiventrecirculation loop flow rete, the STPT l logic is designed to. maintain.a relatively constant margin between 1 g / ..the reactor power and the STP trip setting..-This margin, or power. .i difference, between~the reactor poker and the thermal power;. trip setting'is established by the STPT setpoint specification: j 6 w + 54% ~ 1 -S < minimum of where "W" is the recirculation loop flow rate as a percent of i rated. Thi's specification requires that the STPT setpoint.be. reduced as the r' circulation loop flow rate (and.hence reactor ' I power) is reduced...The characteristic decrease in' ACPR with 'j decreasing power, and the reduction'~in APRM STPT setpoint with' decreased recirculation locp flow (and hence reactor pcwer), both ' act to assure'that the fuel cladding integrity Safety Limit is-not ~ violated during the loss of feedwater heating transient at less y -than rated power. i As an additional point of interest, at less than. rated recircula-1 tion loop flow (W < 100%), the K, factor increases the operating limit. As' stated in Reference l'this is to assure that the Safety Limit is-not' violated during events ihj.ch prod.uce a larger aCPR

b i

from low power states than frem rated power. However, because the i ACPR during a loss ~of feedwater heeting event is less atticw power ~ than at rated power,- the increased cperating' limit (due-to.the K,. factor) at less than' rated recirculation loop flow provides core' margin between the1 Safety Limit and the transient MCPR than exists' 1 h at rated power. This third fad: tor.also assures that the fuel .s cladding integrity safety limit is not-violated during the loss of' F feedwater heating transient at less than rated power. 1 l Analyses for Hatch Unit 1 Cycle 4 have demonstrated that with only the 120% trip setting, none of the abnormal operational transients 1 analyzed violates the fuel cladding integrity safety limit, and-that there is substantial margin from fuel damage. Therefore, the use of the flow referenced trip.setpoint, with the fixed setpoint as backup, provides adequate thermal margins for fuel cladding integrity. 6. MPLICATIONTOCURRENTOPERATINGPLANTS At present, Brunswick Units 1 and 2 and the James A. FitzPatrick Nuclear Power Plant are operating with the new APRM scram trip . logic. This logic was an integral part of the Brunswick Uni'ts 1 and 2 APRM scram trip systea when these plan:s were initially y l LMZ: at/57J 4 L,.'-. i, e__ _=_

,cu., ~ '. y

y ---

th n. ~ vlicensed (see SecticiP7.5.7,: Average P.ower; Range MonLr.. Subsystem,- fin the' Brunswick Units land 2 F5AR).~ The'new APRM scram trip 1 '0

logic was: licensed as'a retrofit margin improvement. option on-the 4

~ I T ~ James A. FitzPatrick. plant. 3 q h Field experience frca these plants has shown that scrams-from . recirculation ' system excursions have 'been reduced by 50 to-75% due to' this. raodi fication. 'A similar reduction on' spurious-scrams is ' expected When the new APRM scram trip logic is' installed in the

Hatch Unit 1 plant.

J 7.

SUMMARY

AtlDiC0!lCLUSI0tl5-i This document has discussed the design and safety aspects of the. W new APRM; scram trip-logic. ~By reducing the: cyclic duty on the 'l plant, greater. margins.to, cumulative fatigue damage.limitsLexist. l 'In addition,. the effects of slow less of coolant transients are -less severe. jTherefore, the safety of the plant will not be reduced by operation of the plant.with the new APRM. scram trip. logic.- a .s. 8. REFERENCES I 1. " Generic Reload Fuel Application",.NEDE-24011-P-A,' August 1 1978. l 1 k d ) l, t ? E I. - / j s,i [.- y n 1 ( l . p, ' i 0l 0 '1 l l-! 1 L ! b. 8 l E -I 1 e l L. m.. L L <gy s utz: at/s7a - 7 \\ L y ~~~ i j . ~.. -"-. ~~.. .~..-. i n

o,., "y }wso: y ngu m,m /, g m.M. neutron flux- "i 7 J,4 / y 4 ,/"

2. tame

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.:' lc / Fiux? . l:. '1 7-spike: .l.:Transientlinduce' tJ ~ d flux spike causing ' 11 reactor scram. ' ~ n /

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l ..a c .g c y e e 'g. j.. [.. n. u 3 .g. ,q s V. ] 100 .f j CORE ROW %. FIGURE 1 ',I - m. n 4.I ~ - q .1 n ' ij. ( APRMSCP1".S d

  • 1 WITH STP W

flon-FicM 120 APPT STP -d freferenced __--------------I (adjustable V 11 H ).. I; j ~~ -- upper; trip 111mit'. ~APRM neutron' flux scra - /,.... d / q ~~ 100 f 1. a ;. ; .,/ a /r, l. i,j j Fle(referenced .. / d l / STP.s ike' .I' Ef fcct of the sa: 4 -] E#.AFRM STP-Scram -~~~~ flux spike as in figure 1: ir l' but modifled by STP. -l flo scram occurs.. g l 7 s. v. 5- ,i o a. -) I I 100 CORE ROW % J i M 1 ] 4 f i' ' .g. a j { e-E r } ~ ~ - - ~ ~ - - .... -. ~. - u 1 i e i

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? ~ } ,l \\ j 3 i q ENCLOSURE 5 fiRC DOCKET 50-321 f OPERAT!flG LICENSE DPR-57 EDWIft I. HATCH NUCLEAR PLANT UNIT'1 i PROPOSED CHANGES T0 TECHillCAL SPECIFICATIONS j -l The proposed changes to the Technical Specifications ( Appendix A to Operating. .i . License DPR-57) would be incorporated as follows: 'j 1 Remove Page Insert Page l . l.1 -1 1.1-1 l 1.1-2 1.1-2 1 1.1-12 1.1-12 1.1-13 1.1-13 rigure 1.1-1 Figure 1.1 { 3.1-4 3.1-4 3.1-7 3.1-7 3.1-12 3.1-12 q 1 b 1 1 ,k j l i 1 i 1 ~1 I P J r l '. B.. i w =____. _

[.e a;' e l} u, 'n - 5%EETY LIMITS LIMITING SAFETY SYSTEM SETTINGS [.1.11FUELCLADDINGINTEGRITY 2.1 FUE(CLAD 01NGINTEGRITY f Applicability Aoplicabili ty The Safety Limits established to pre-The Limiting Safety System Settings f serve the fuel cladding ~ integrity apply apply to trip settings.of the instru-. I to those variables which monitor the ments and devices which are provided to { , fuel thermal behavior, prevent the fuel cladding integrity ( ' Safety L!mits from being exceeded. ) i i Objective Obj ective i ~ ) The. objective of the' Safety Limits is. The objective of the Limiting Safety 1 .to establish limits below which the. System Settings is to define the level l integrity of the fuel cladding is of the process variables at which auto-preserved. matic protective action is initiated. j to prevent the fuel cladding integrity. 1 Safety Limits from being exceeded.. specifications Specifications ) l A. Reactor Pressure > 800 psia and Core A. Trip Settings 1 Flow > 10% of Rated The limiting safety Lsystem trip set-d The existence of a minimum critical tings shall be as specified below: f? power ratio (MCPR) less than 1.07 l ~ 4 shall constitute violation of the

1. Neutron Flux Trip Settings j

fuel cladding integrity safety limit. .{

a. IRM High High Flux Scram Trio

{ Setting i B. Core Thermal Power limit (Reactor The IRM flux scram trip setting Pressure < 800 psia) shall be 1 120/125 of full scale. When the reactor pressure is 1 800

b. APRM Flux Scram Trip Setting psia or core flow is.less than 10% of (Refuel or Start & Hot Standby a

rated, the core thermal power shall Mode) not exceed 25% of rated thermal power. When the Mode Switch is in the i REFUEL or START & HOT STANDBY position, the APRM flux scram C. Power Transient trip setting shall be < 15/125 of full scale (i.e., 1 15T of' rated .To ensure that the Safety Limit estab-thermal power). lished in Specification 1.1.A and 1.1.B is-not exceeded, each required

c. APRM Flux Scram Trip scram shall be initiated by its 3etting (Run Mode) expected scram signal.

The Safety Limit shall be assumed to be exceeded (1) Flow Referenced Neutron Flux when scram is accomplished by a means Scram Trip Setting other than the expected scram signal. When the Mode Switch is in the RUN position the APRM flow referenced flux scram trip setting shall be: 1 1.1-1 1

TWw,

.w: + R . P M ' ', N T 5 ,i. l j-i [ IU b .f Jih ,h., 'l. IIC-1- 3 .q, [m -SAFETYoLIMI6o LIMITING 3EEfTY SYSTEM SETTINGS-ghd((R'eact'o N$terLeve1(Hot 4crDCo'l[ 2.1Yd.1'.C. NPRM T1ux Scram' Trip 7 .L s Shutdown - Condi tion). 'Settinast (Run Mode) ;(Continued)- s gs- .,c i 1Whenever:the reactorlis in the Hot " LS ( 0.66 WL+.54% - b b [m, $orJ. Cold Shutdown Condition with; m Sp (irradiated; fuel in theireactorfvessel,- lwhere: a the ;ytater : level shall' be.>' 378 inches' i" O(w' : ' above. vessel. invert when fuel is.- S = Setting.'in percent of-g ?

  • 1seatedlin.the dore.
rated thema1 ' power

.;(2436MWt) i 4 s. n . r - 'i-E*.l T 'j ; ', W:= Lo'o'p? recirculation-flow 1 L. l rate in: percent of rated; D L(rated loop recirculation Es .i- ' flow rate equals j 34.2.x 4106 lb/hr)- 'D-t* w .t;. JInLthe event of operatio'n:with a .. i maximum'totalipeaking factor (MTPF) greater.)than,the design: 1 m' value, the-setting shall be- ~ - modified as: follows: = a x. 0; -S~s (0.66;WJ+ 54%) g -- A-4 -s where: j i G i MTPF = The ~ value Lof the ;...... existing maximum = total. D _y peaking-factor 1 A = 2.60 forJ7x7: fuel 2.42 for 8x8 fuel J 2.48 for 8x8R fuel For no combination of lo'op recir- "1 .cu at on flow rate and cores ( r. l i i thermal power shall the APRM flux scram trip setting;be allowed toi exce'ed 117%'of rated thermal . power. 1 Surveillance requirements:for a MTPF 'are given in ' Specification g 4.1.B. d (2) Fixed High Neutron Flux h Scram Trip Setting i .i .When the Mode Switch is in 'l the RUN position, the APRM i fixed high flux' scram trip .l setting shall be: ( S 5 120% Power !L 1.1-2 ..b l r

m ? L-s l-BASES FOR LIMITING SAFETY SYSTEM SETilNGS { ~ 2.1. A.I.a. IRM Flux Scram Trip Setting (Continued) tism was taken in this anahsis by assuming that the IRM channel closest j l to the withdrawn rod is byoassed. The results of this analysis show that l l the reactor is scrammed and peak power limited to one percent of rated l . power, thus maintaining MCPR sbove 1.07. Based on the above analysis, the IRM provides protection against local control rod withdrawal errors i and continues withdrawal of control rods in sequence and provides backup protection for the APRM.

b. APRM Flux Scram Trip Setting (Refuel or _St_ art & Hot Standby Mode)

For uperation in the startup mode while the reactor is at low pressure, the AFRM scram setting cf 15 percent of rated power provides adequate thermal margin between the setpoint and the safety limit, 25 percent of l rated.. The margin is adequate to accomodate anticipated maneuvers asso-J ciated with power plant startup. Effects of increasing pressure at zero l or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be 1 uniform by operating procedures backed up by the rod worth minimizer and 1 the Rod Sequence Control System. Worth of individual rods is very low in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of sig-nificant power rise. Because the flux distribution associated with i uniform rod withdrawals does not involve high local peaks, and because q several ro'ds must be moved to change power by a significant percentage J of rated power, the rate of power rise is -very slowr-Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no I I more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit. The 15 percent APRM scram remains active until the mode switch is placed in the RUN position. This switch occurs when ] reactor pressure is greater than 825 psig.

c. APRM Flux Scram Trio Settinas (Run Mode)

The APRM Flux scram trip in the run mode consists of a flow referenced scram setpoint and a fixed high neutron' flux scram setpoint. The APRM flow referenced neutron flux signal is passed through a filtering network with a time constant which is representative of the fuel dy-namics. This provides a flow referenced signal that approximates the average heat flux ~or thermal power that is developed in the core during transient or steady-state conditions. This prevents spurious scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses. Examples of events which can result in momentary neutron flux spikes are momentary flow changes in the recirculation system flow, and small pressure disturbances during turbine stop valve and turbine control valve testing. These flux spikes represent no-hazard to the fuel since they are only of a few seconds duration and less than 120*; of rated thermal power. j l 1.1-12 s

mm,. 7-4 J l 1 WMfA@cf y ]PRM FluxlScram Trip Settings"(Run Mode)-(Continued) i y[f The APRM flow,raferenced scram. trip setting at full recirculation'n flow i. ~is' adjustable up ta 117% of rated power.: This reduced flow referenced-1 trip setpoint will result-in an' ea511er scram during: slow thermal R o-transients., such as the loss ofcE F feedwater heating event,.than. i L would result with the 120% fixed.high neutron #1ux scram trip., The - 1 L lower flow ' referenced scram setpoint therefore decreases the ~ severity. 1 l 1 J(4CPR) of a slow thermalitransient'and allows lower Operating 1.imits ' H !Y#, .if such a transient,isithe limiting abnormal operational transient ~ i during:a certain exposure interval in the cycle. Thel APRM' fixedx high. neutron flux signar does not incorporate j the time. constant, but responds directly to instantaneous neutron flux. j . This' scram setpoint scrams ithe. reactor during.' fast power increase D ~ transients if credit is not taken for a direct (position) scram, and

c also serves to'seram'the reactor Lif credit'ist not taken.forjthe flow i

referenced scrm l s, The. flow referenced scram trip setting must be: adjusted to ensureithat Lthe LHGR transient' peak is not increased for any combination of MTPF and' reactor , y! core thermal power. The scram setting.is adusted in.' accordance with the : formula:in Specification 2.1. A.~1.c., when the maximum. total peaking factor j. is. greater than 2.60 for 7x) fuel, 2.42'for 8x8' fuel: and 2.48 for 8x8R fuel. j Analyses ofitne limiting transients show that no scram adjuctment.is j i required to assure' MCPR > 1.07'when the transient is initiated from thet J operating MCPR. limit. ld. APRM Rod Block Trip Setting. ...m. Reactor power level may be varied by. moving control rods ;or by varying _ ii the recirculation flow rate. The SPRM system provides a control rod block to prevent rod withdrawal-beyond a given foint at constant r: circulation flow rate, and thus to protect against the condition.of a MCPR less than n-1.07.. This rod. block trip' setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod-withdrawal. The: flow. E . variable trip setting provides substantial margin-from fuel damage, assuming a steady-state operation' at: the-trip setting, over the entire recirculation flow "ange. The marg'in-to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relation-o ship; therefore, the worst case MCPR which would occur during a steady-state operation is at 108% of rated thermal power because of the AFRM rod block trip setting. The actual power distribution in.the core is estab-lished by specified control rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram trip setting the APRM- .l rod block trip setting is adjusted downward if the maximum total, peaking. factor exceeds 2.60 for 7x7 fuel, 2.42 for 8x8 fuel and 2.48 for 8x8R fuel, thus preserving the APRM rod block safety margin. 1 2. Reactor' Water low Level Scram Trio Settino (LLl) ] The trip setting for low level scram is above the bottom of the separator skirt. This level is > 14 feet above the top. of the active fuel. This level has been used in transient analyses dealing with coolant inventory { decresse. The results reported in FSAR Section 14.3 show that a scram at j this level adequately protects' the fuel and the pressure barrier. The scram trip setting is approximately 33 inches below the normal operating range and is thus adequate to avoid spurious scrams. 1.1-13 [ b

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t m c c o o o k k k a nu i i k k m m m e e e r k em l l e e e e e e e mi p p e e 3 3 3 W H W p e un p p W W / / / O W ri A A / / y y y e e e / / tM e e r r r c c c e e s t t c c e e e ) i i i c c n o o n n v v v h w w w n n N N O C E E E ( T T T d O I ) l e a ( no t g g p i s n n u t e i i t c T l l r n ef ef a u l y uo uo t F y ac e f f S c nn l es es ,n oe c rr rr ) f t e i u y u u g o su t q C go go ( eq ce nh nh s l e nr) g s i i ) ) ) s r r uF c n h r4 r4 f f f h ) ) ) ) u l F f ( i t u2 u2 ( ( ( t e e e f o a m t n d ) d ) n ( ( ( ( H nm t u a o ne ne h h h c ou nm r m ni( ki( t t t m k k k k 4 m ei e eh eh n n n e e e e 2 i ti mn p 3 et p et p o o o 3 e e e e cn ui O Wiu wiu M M M W W W W n ni rM / y /wt /wt / / / y / ///i uM t e r e r e r e e e r e e e e h F s c e cd a cda c c c e c c c c t n n n v nnt nnt n n n v n n n n i no I O E OaS OaS O O O E O O O O W oi ~ i t t a 1 ar tb 1 ni el p 4 ma u) uC ob A A C C A A }A^ A h B B B C e r r( l td G b sn a na T I ) y ( u ) i S c h P n g R e i l q l L h me l L g er a ( i tF n i s g l i N l ym e e S u S W e e v m x m O ni p D r r e u u e on i T x u u L l l c ii r U u s s o F n tM T H l s s w V e c S F e e e o h e r et m v r r l e g l e e t s. a n h i P P g i b l f x ? a a e u oe r i g t r r I r c R l i a r l e a rT c m P S h a H r o l t h d e s F c r e t e a c e p n w r f t c h p c w W s x o w o i S g o a y i i n o l 5 o o i n e r r D F I D F 1 t w I R D o c e S l H a c a t ml e r e u h h c ae M R u d n M g g a rv R o o a R i i e ce P i S M M I H l R SL A ~ mr ae rb) 1 2 3 4 5 6 7 8 cma S u(N { ) t f

q., ~i n w.., . D35. A(4 h!}iicih Reactor: Pressure t"i nu'ed )" 4 ^ ~ the' core from.ex'ceeding thermal hydraulic limits.as a result.of, pressure increases.for some events that : occur when'the reactor is operating at less ' 1 Lthan rated power and flow. 1 .5. HichiDrywell Pressure q o-( : Presrure switch instrumentation for the drywell is provided 'to detect at ~ loss.of coolant accident and initiate the. core standby cooling equipment. H A high drywell pressure scram is provided at the same setting j,2 psig). 1 as the core standby. cooling systems' initiation to minimize the.ener,gy which must be accommodated during a' loss of coolant accident.'1The instru-mentation is' a backup'to'the eactor vessel water leveljinstrumentation. T y-

6. ' Reactor Water Low Level (LL1);

8 i. I The bases for the Reactor Water Low level ScramLTrip Setting,(LL1) are- -t + discussed in the bases for, Specification 2.1. A.2. e 7. Scram Discharge Volume Hi_qh'High Level ' The controlf rod drive scram ' system is designed 'so that.all of' the water-which is' discharged from.the reactor by~a scram con be accommodated:in the discharge piping. A_.part of this' piping is< an, instrument volume which is~ the low-point in the pising. No credit.was taken for.this-l f volume in the design of, the disc 1 ace piping 'as concerns the amount of - water which must be accommodated during a scram. During normal operation the discharge volume is empty; however, should the' discharge volume fill with water, the water discharged to the piping from the reactor could-not be accommodated which would. result in.a slow scram time or partial or,no i control rod insertion. ;To preclude this ' occurrence,. level switches have been provided in the instrument volume which scram the reactor when'the volume. of' water reaches 71 gallons. As indicated above there is suffi-cient' vo:ume in the piping to accommodate the scram without impairment of the scram times or amount:of insertion of the control rods. This function shuts the reactor'down whileisufficient volume remains to accommodate the discharged F: . water and precludes the si tuation in which a scram would be required but r.ot able to perform its function adequately.

8..

APRM Three APRM instrument channels are provided for each protection trip sys-( tem. APRM's'A and E' operate contacts in one trip logic and APRM's C and E operate contacts inlthe other trip logic. APRM's B, 0 and F are arranged-similarly in the other protection trip system. Each protection trip sys-tem has one more APRM than is necessary to meet the minimum number re-quired per channel. This allows the bypa'ssing of one APRM per protection

41 trip system for maintenance, testing or calibration.

a. Flowf Referenced and Fixed High Neutron Flux The bases for the APRM Flow Referenced cnd Fixed High Neutron Flux Scram Trip Settings are discussed in the bases for Specification 2.1. A.1.c. l L r. 3.1-12 V

l-l ~ (< 'I t V, ) ' ;f - 1 l . L/' ' ENCLOSURE-6 l L' a NRC 00CKET.50-321 'l i L OPERATING LICENSE DPR-57 .i 7 i. EDWIN I.-HATCH NUCLEAR PLANT UNIT l-i I PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS I r, l' The' proposed changes to the Technical Specifications (Appendix A to Operating' i License DPR-57).would be incorporated as follows: ] R i Remove Pace Insert Paae j 3.10-2' 3.10-2 3.10-7 3.10-7 ,l i 4 ) ,j i 9 4 -i j Y } t i .i i I 1 i i )

w ., r 4 i L., 3 f 3-i y mm .- LIMITING CONDITIONS F(W)PERATION SURVE CfANCE REQUIREMENTS e ~r

1 3.10!C' Core Monitoring:During' Core.-

.4.10.C. Core Monitoring During' Core j

Alterations.

Alterations Prior'to making normal alterations w.. 'O f 1, During normal-' core alterations, p. A two SRM's1shall. be operable; one to the core the'SRM's shall be-

intthe core quadrant where fuel or.

functionallyJtested.and checked N T control; rods are being moved and for neutron response.- There-lM one in an; adjacent quadrant, ex-after, while required to be cept as specified in 2.and 3 below.

operable, the SRM's' will'be g
checked daily for response.

3f 4 'For,an.SRM to be' considered operable,- it shall be' inserted.to the normal Use of special moveable,: dusking i ! operating level and.shall have a type detectors'during initial g ?p,

minimum ofi.3 cps with all rods

. fuel loading and majoracore alter s 7' capable of-normal insertion ~ fully ations in place of n'ormal_de-1 inserted. ltectors is permissible as long. 4

q. -

-as the detector:is' connected t j 2.L-Prior to spiral unloading the SRM's- ' to the. nomal SRM. circuit.. i ,f

shall be' proven operable as stated

'l above. however, Eduring spiral unloading Prior to spiral' unloading o'r re-' .t the' count. rate may. drop below 3 cps. . loading the SRM s shall'be func' - tionally. tested..-Prior to: spir'al

3. Prior to'sprial reload, two' diagonally unloading the'SRM's should'also be adjacent fuel assemblies.wilizbe checked for neutron respons'e.
  1. y

+ : loa'ded into ' their previous core posi-Wi .tions next to each ofL'the 4 SRM's-to obtain the required 3 cps. Until these f f/ cight assemblies have. been loaded, the 3 cps requirement is not necessary. j O D., l Spent: Fuel Pool: Water Level D. Soent Fuel-Pool Water Level 9 'Mhe'never irradiated fuel is Whenever-irradiated fuel is ' stored in the spent fuel pool, stored in,the spent. fuel pool, the water level shall be checked ,the pool water level shall be and recorded' daily. maintalnad at or above 8.5 ' feet 'above the top of the ' l r 4 active fuel. i. 'u. E. Control Rod Driva Maintenance .E. Control Rod Drive Maintenance 4

1. Requirements for Uithdrawal
1. Requirements for Withdrawal of 1~or 2 Control Rods of 1 or 2 Control Reds

-A maximum of two control rod's q separated by at least two control cella in all directions may be with-drawn or: removed from the core, for' the purpose of performing control rod drive maintenance provided that:

a. The Mode Switch is lock d in the REFGL
a. This surveillance requirements pcsition. The refueling interlock is the same.as given in 4.10. A.

which prevents more than one control rod from being withdrawn may be bypassed for one of the control rods on which maintenance is being i .p, ( s 3.10-2 v c n, L O _m_ f u

BASES FOR LIM TING CONDITIONS FOR OPERATION' f- ~# 4 ~ .V 3.10. A;2. Fuel Grapple Hoist Load Settine Interlocks _ Fuel handling is normally conducted with the fuel grnpple hoist. The total load on this hoist when the inter 16ck is required consists of the weight of the fuel grapple a'd the fuel assembly. This total is approximately 150Q lbs. n in comparison to the. load setting of 485 i 30 lbs.-

3. Auxiliary Hoists Load Setting Interlock Provisions have also been made to allow fuel handling with either of the three auxiliary hoists and still maintain the refueling interlocks. The 485 + 30 lb.

load setting of these hoists is adequate to trip the interlock when a fuel'i { ' bundle is being handled. i .J B. Fuel Loading To minimize the possibility of loading. tual into a' cell containing no contro'1 rod, it is required that all control rods are fully inserted when fuel is being loaded into'the reactor core..This requirement assures that during refueling the refueling interlocks, as' designed, will prevent inadvertent criticality. C. Core Monitoring During Core Alterations q i The SRM's are provided to monitor the core during periods of Unit shutdown and 'to guide teh operator during refueling operations and Unit startup, Requiring two operable SRM's in or adjacent to any core quadrunt where. fuel or' l control rods nre being raoved assures adequate monitoring of. that quadrant during ] I such alterations. The requirements,of 3 counts per second provides assurance 1 that neutron flux is being monitored. During sprial unloading, it is not necessary to maintain 3 cps because core nitcrations will involve only reactivity removal and will not result in criticality. The -loading of diagnally adjacent bundles around the SRM's before attaining the ' f 3 cps is permissible because these bundles were in a suberitical configuration f s when they were removed and therefore they will remain suberitical when placed - l back in theLr previous positions. D. S c_ en t Fuel Pool Water Level ) i The design of the spent fuel storage pool provides a storage location for j approximately 150 percent of the full core load of *,uel assemblies in the reactor building which ensures adequate shit:1 ding, cooling, and reactivity l control of irradiated fuel. An analysis has been performed which shows j that a water level at or in excess of eight and one-half feet over the top of the active fuel will provide shielding such that the maximum calculateo f, radiological doses do not exceed the limits of 10CFR20. The normal water j level provides 14-1/? feet of additional water shielding. All penetrations of j the fuel pool have been installed at such a height that their presence does not j provide a possible drainage route that could lower the water level to less than 10 feet above the top of the active fuel. Lines extending'below this level are ecuipped "ith two check valves in series to prevent inadvertent pool drainage, f f. E. Control Rod Drive mintenance During certain periods, it is desirable to perform maintenance on tse control g - rod drives at the same time.

3. 10-7 I

i 1

.-m f .g~ e' ' [, .~ 6i ,4 ?;i REGULATORY IN.AMATION DISTRIBUTION SYST d'(RIOS), o ( . AC PES $farf.N64 : 7 905190213 DUC.DATE:~~ 79/05/11' NOTARIZED 3 NO. .'00CKET s' 'FgA C I L : S O-3 21 EDWIN.I.' HATCH NUCLEAR PLANT,' UNIT.1, GEUNGIA' POWER-C ' 05000321' ,.e 'f

A'UTH.NNIE.-

AUTHOR AFFILIATION-flHITNER,C.F. GEORGIA.-POWER CO. ' 1, f j. iRECIP.NAME .. RECIPIENT AFFILIATION i Of-F I C E OF NUCLEAR REACTOR REGULATION ij

SUBJECT:

AMENDS 790.6P2LSUOMITTA'L[RE: TECH SPEC CHANGES IN: SUPPORT OF RELOAD 3 LICENSING APPLICAT10N:DTD'790322. CLASS IV' AMEND FEE ENCL'.. 1 H 4 S2-

DISTRIBUTION CODE:.A001S; COPIES 4 RECEIVED:LTR $9 ENCL 10 SIZE:

T IT.L E : GENERAL DISTRIBUTION;-FORjAFTER. ISSUANCE'0F'0PERATING.LIC - f 0TES:' RECIPIENT

COPIES RECIPIENT COPIES.

l' ID CODE /NAME' LTTR ENCL ID' CODE /NAME' ' ULTTR ENCL :t

ACTION:-

05 BC o8843 - 7 L7. a s t INTERNAL: 01s t t, rit N 1-

1. '

02.NRC.PDR .1 ' 1, if 15E 2 'a 14 TA/EDD-- ,1 1 15 CORE PERF BR 1 1 '161AD SYS/PROJ- 'l '1' 17 ENGR.BR 1-1 18 REAC SFTY BR

1 1

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