Text

Amends 170 & 108 to Licenses DPR-57 & NPF-5,respectively, Making Miscellaneous Changes to Tech Specs
	ML20059G774
Person / Time
Site:	Hatch ; (DPR-57-A-170, NPF-05-A-108, NPF-5-A-108)
Issue date:	08/30/1990
From:	Matthews D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20059G777	List: ML20059G774; ML20059G779;
References
	NUDOCS 9009130142
	Download: ML20059G774 (48)
	v • d • e

{{#Wiki_filter:. g f ( g+f \\j UNITED sT ATEs )er g NUCLEAR REGULATORY COMMISSION in 1 WASHINGTON, D. C. 20f M %e,,,e/ i GEORGIA POWER COMPANY l l OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA i CITY OF DALTON. GEORGIA l DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT. UNIT 1 -l AMENDMENT TO FACILITY OPERATING LICENSE j l' Amendment No. 170 l License No. DPR-57 1. .The Nuclear Regulatory Comission (the Comission) has found that: i l A. The application for amendment to the Edwin I.' Hatch Nuclear Plant, Unit 1. (the facility) Facility Operating License No. DPR-57 filed by ) l l Georgia Power Cort ;.y, acting for itself, Oglethorpe Power i Corporation, Munici pal Electric Authority of Georgia, and City of i Dalton, Georgia,(tielicensee)datedMarch2,1990,complieswith the standards and requirenents of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; l B. The facility will operate in conformity with the application, the l' provisions of the Act, and the rules and regulations of the j Comission; C. There is reasonable assurance d ) that the activities authorized by this amendment can be conducted without endangering.the health and safety of:the public, and (ii) that such activities will be ~ ,l' conducted in compliance with the Comission's regulations set forth i in 10 CFR Chapter I; D. .The issuance of this amendn.ent will net be inimical to the ccmon defense and security or to the health and safety of the public; and \\ E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulhtions and all epplicable requirements have been satisfied. J $0?I&CK05o99373 A2 900630 P PDC

l /I.; 4,

. O.

2 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amerdment, and paragraph 2.C.(2) cf facility Operating License No. DPR-57 is hereby anended to read as follovs: Technical Specifications The Technical Specifications contained in Appendices A ano B, as revised through Amendment No. 170, ere hereby incorporated in the j license..The licensee shell operate the facility in accordance with the Technical Specifications. ) 3. This license amendment is effective as of its date of issuance and shall Le implemented within 60 days of issuance. l FOR THE NUCLEAR REGULATORY COMMISSION L ) h David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-1/11 Office of Nuclear Reactor Regulation Attachment

i Changes to the Technical Specifications

( Date of issuance: August 30, 1990 1: l 4-L 1 ? A l

1 a t s + ATTACHMENT TO LICENSE AMENDMENT NO. 170 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 ? Replace the following pages of Appendices "A" and "B" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indic6 ting the areas of change. Appendix A Rcrove Page Insert Page 1.0-6 1.0-6 3.1-1 3.1-1 3.1-8 3.1-8 3.2-7 3.2-7 3.2-9a 3.2-9a 3.2-10 3.2-10 3.2-13 3.2-13 3.2-14 3.2-14 3.2-18 3.2-18 3.2-43* 3.2-43 3.2-44 3.2-44 3.4-2 3.4-2 3.5-1 3.5-1 3.5-2 3.5-2 3.5-3 3.5-3 3.5-4 3.5-4 ) 3.5-5 3.5-5 3.5-6 3.5-6 3.5-7 3.5-7 3.5-8 3.5-8 3.5-9 3.5-9 3.5-10 3.5-10 3.5-10a l 3.5-11 3.5-11 I 3.5-12 3.5-12 l 3.5-13 3.5-13 3.5-14 3.5-14 3.5-15 3.5-15 3.5-17 3.5-17 3.5-18 3.5-18 3.5-19 3.5-19 3.5-21 3.5-21 3.6-9c 3.6-9c 6-0 6-6 Appendix B 5-3 5-3

0verleaf page provided to raintain docunent completeness.

1 i a i GG. ijmulated Automotic Actuation - Simulated automatic actuation means applying a simu'ated signal to the sensor to actuate the circuit in question. HH. Start & Hot _ Standby Mode - The reactor is in the Start & Hot Standby Mode when the Mode Switch is in the START & HOT STAND 8Y position. In this mode the reactor protection system is energized with IRM and APRM (Start & Hot Standby Mode) neutron monitoring system trips and control rod withdrawal inter-locks in service. l Surveillance Frecuency - Periodic surveillance tests, checks, calibrations, and examinations shall be performed within the specified surveillance intervals. These intervals may be adjusted plus or minus 25%..The operating cycle interval is defined as 18 months. In the case where the elapsed interval has exceeded 100% of the specified interval, the next surveillance interval shall commence at the end of the original specified interval. JJ.- Surveillance Reauirements - The surveillance requirements are requirements established to ensure that the LCO stated in Section 3 of these Technical Specifications are met. Performance of a surveillance requirement within the specified surveillance interval shall constitute compliance with the operability requirement for an LCO. Surveillance requirements are not required on systems or parts of systems that are not required to be operable or are tripped. If tests are missed on parts not required to be operable or are tripped, then they shall be performed prior to returning the i system to an operable status, KK. Total Peakina Factor (TPF) - The total peaking f actor is the highest product of radial, axial, and local peaking factors siraltaneously operative at any segment of fuel rod. LL. Transition Boilina - Transition boiling is the boiling that occurs between nucleate and film bolling. Transition boiling is manifested by an unstable fuel cladding surface temperature, rising suddenly as steam blanketing of the heat transfer surface occurs, then dropping as the steam blanket is swept away by the coolant flow, then rising again. L 1 1 e i HATCH - UNIT 1 1.0-6 Amendment No. 170 t 'Y --vw-r

I~'L&, l - o y S > ) J f LIMITING CONDITIONS FOR OPERATION $URV[lLLANCE REQVIR[M[NTS L 3.1. REACTOR PROTECTION SYSTEM (RPS) 4.1. REACTOR PROTECT 10W SYSftM (RP$) i p Aeolicability ADelicability The Limiting Conditions for Operation The Surve111cnce Requirements asso-i j associated with the Reactor Protection cisted with the Reactor Protectic.n System apply to the instrumentation and System apply to the instruments; ion. associated devices which initiate a and associated devices which initiate j reactor scram, a reactor scram. Objective Objective The objective of the Limitira Condi-The objective of the Surveillance tions for Operation is to assure Requirements is to specify the type [ the operability of the Reactor and frequency of surveillance to be Protection System. applied to the protection instrumen-i tation to assure operability, i- 'i Specifications $cecifications t A. Sources of a Trio $1onal Whith A. "est and_ Calibration.Recuirements Initiate a Reactor Scram for the RPS The instrumentation requirements RPS instrumentation systems and associated with each source of a associated systems shall be func-scram signal shall be as giv-a in tionally tested and calibrated as Table 3.1-1. indicated in Table 4.1-1. l The action to be taken if the number The trip system containing the of operable channels is not met for unsafe failure may be placed in both trip systems is also given in the untripped condition during the l Table 3.1-1. period in which surveillance testing is being performed on the other RPS channels. B. [ ore Maximum Frattion of B.

ore Maximpm Fraction of Limitina Power Density (CMFl*D)

.imitina Power Density (CMFLPD) This section deleted. This section deleted. l N L HATCH - UNIT 1 3.1-1 Amendment No. 170 l-L 4 }{

g' l~ ~' Table b.1-1 (Cont.) j 5 .-4 Q Sc ram Instrument Check-Instrument Functional Test Instrument Calibretion - Number Source of Scram Trip signa t -- Group.- Minimum Frequency Minimum frequency Minimum Fn n_.,j e fal (b) ic) h 9 Main Steam 1.ine High Radiation 8 D Every 3 months (e) ~ Every 3 months (I} 10 Main Steam t.ine Isolation valve A NA Every 3 months (h) w Closure 11 Turbine Control valve Fast A NA Every 3 months (J) once/ Opera ting Closure . Cycle (k) 12 Turbine stop valve Closure A NA Every 3 months th) RPS ChanneI Switch A NA Once/Operaeing CycIe stot App 8IcebIe J Turbine First Stage Pressure A NA Every 3 months Every 6 months i Permissive I a. The column entitled " Scram Mumber" is for convenience so that a one-to-one relationship can be estabtished i between items in Table 4.1-1 and items in Table 3.1-1. b. The definition for each or the four groups is as rollows: 7 Group A. On-off sensors that provide a scree trip signal. m Group B. Analog devices Coupled with bi-stable trips that provide a scree trip signal. Group C. Devices which only serve a userut function during some restricted mode or operation, such as startup or shutdown. Or for which the only practical test is one that can be performed at shutdown. Group D. Analog transmitters and trip units that provide a scram trip function. c. Functional tests a re not required when the systems are not required to be operable ce are tripped. However, if runctional tests are missed, they shall be perf ormed prior to returneng the systems to an operable status. d. Calibrations are not required when the systems are not required to be operable or are tripped. However, if' calibrations are missed, they shall be performed prior to returning the system to an operable status, y e. This instrumentation is exempted from the instrument functional test definition. This instrument s functional test wi t s consist of injecting a simulated electrical signal into the measurement g channeas. ct 9 f. Deleted c+ g. The wa te r leve t in the reactor will be perturbed and the corresponding levet indicator changes wilt be mon i to red. This perturbation test will be performed every 3 months af ter completion of the zo functional test program. h. Physical inspection and actuation of these position switches wili be performed once per operating cycle. ~

i. Standa rd current source used which provides an instrumetet channet alignment. Calibration using a radiation source shall be made once per operating cycle.

J. Measure time interval rrom E,4C pressure switch actuation to arS relay scib de-energization: w er-_,-_\\ v e-x w__a*"-=..__-__.--___-_--_--_____._w,,m-_. _ _ _ _ _ _ wm

-( .cLe -~ leotes for Tebte 3.2-2 (Cont.) ~ ~ =_ ., w = b. When any CCCSl subsystem is required to be operable by Section 3.5, there siwt I be two operable tr;p systems. t r the required nossber or operable charinet s estenct be met for one of the trip systeme, place the inoperable channet in the tripped cone t tion or declare the associated CCCS inopersbte C -- within 1 hour. If the requirsd nuat:er of operable chonnets connot be met for both trip systests; 2~ declare the associated CCCS istwrable within 1 hour. --s-s-

N 8

N + o 3a ~ 3o 3 + f+ 4 Z o. / W N O

.g<.- s, NOTES FOR TABLE 3.2-3 a. The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be established betweer, items in Table 3.2-3 and items in' Table 4.2-3. b. When any CCCS subsystem is required to be operable oy Section 3.5, there shall be two operable trip systems. If the required number of operable channels cannot be met for one of the trip systems, place the inoperable channel in the tripped condition or declare the associated CCCS inoperable within 1 hour. If the required number of operable channels cannot be met for both trip systems, declare the associated CCCS inoperable within 1 hour. 7 IIATCH - UNIT 1 3.2-9a Amendment No. 170

s. 6 d ~ z . r; t.x. ~-

Table 3.2-1s W

-L INSTRt#tENTATIOlt WHICH INITI ATES OR C088TROLS ADS . -4 $: Re f. Instruments . 7 rip ' Required Trip Setting Reserks No. Con (il ion - Operable ' '. (a) . Nomenclate',re - Channels-per Trip _ c System ib) q.- Wa te r Leve l. Low (Level 3) 1 -110.0 inches; ' Confirms low level,305 permissive 1. Reactor Vessel w .x Reactor VessCI % dater Lawl Low Low Low 2 2-113 inches Permissive signal;to AOS. timer - 4 N ilevel 1) x 2. Drywell IIretsJm High 2 ~$1.92 psig Permissive signal to ADS timer.. RN( u. ~ -G 3. RHR Pump.Di scha rge High'

2' 2112 psig Permissive signe t to ADS timer.

] ~~ "

~ ~ _ _ Pressu re 4. CS Pump Discharge ( Hioh 2 2137 psig Permissive-signal to ADS timer Pressu re 5. Auto Depressurization 2 513 minutes Bypasses high drywell pressure-Low Water Level Timer - permissive upon sustained Level 1 6. Auto Deprnssa.1 ration 1 120 1 12 seconds with Level 3 and Level 1 and higte y d rywel I oressure and CS or RMR pump Timer N N at pressure, timing sequence begins. If the ADS timer is not-i reset it will Initiate A05. o 7. Automatic Blowdown Control 1 Ilot applicable Monitors avellability of power to logic system Powe r Fa i lu re Moni to r The column entitled "Ref. No." -is only for convenieece so that e one-to-one relationship can be established between a. items in Table 3.2-4 and items in Table 4.2-4 D. When any CCCS subsystem is required to be operabic by Section 3.5, there shall be two operable trip systems. If the required number of operab!e channels cannot be met for one of the trip systems, g place the inoperable channel in the tripped condition or declare the associated CCCS inoperable l within 1 hour. If the required number of operable channels cannot be met for both trip systems, 3 ci declare the associated CCCS inoperable within 1 hour. Eo 3 et Z .O W er. _,,_.?--' * ' _ e-r+- W 'w

...A L,'L ',' a -.- o i C ,.y. .y; .. b;; ~ ~' vg @^;. ' - w v. y 3 ngcil w p ;' ~ ~ 5 i ^ sootes ror Taste 's.2 I; _ ' ~ 2? - + NiD ~ - ? --~ 1-2. a, - - - q.- s - The column - entitled "Ref. fto." is only for convenience ~ so that a one-to-one. relationship can be,- .. n. . c;.a. established between items in Table 3.2-5 and item'in Table as. 2-5. - C b. When any CCCS subsystem is ' required to be operable by Section 3.5,- there shal a be two operable 3-

trip systems. If the required number of operable channels cannot be met for one of the trip systems,,

place the inoperable channel-in the tripped condition or declare the associated CCCS inoperable -4 within l' hour. If-the required number of operable channels cannot be met for both trip systems, decla re the associated CCCS : inoperable-within 1 50er., .c.J N IW () N B r- <3 3 O. 3 cv 3r+ Z .O W M O

=--a-~'

g .,.c gs, ~ 1 = ~ 3- ~ - t .y. ~

Table 3.2-6 llISTRUMENTATIOle WHICH INITIATES OR CON 1ROLS CORE SPRAY

-4O Ref. Instrument-Trip - Operable Required Trip Setting-Rema rks -5, ~ No. Condition 8 (a) NomencIature-ChanneIs c-- per Trip 2: SVstem Ib) U 1. Reactor Vessel Water Level Low Low Low 2-113 inches initiates CS. ~~Si (Level 1)- 2 2. Drywell Pressure High 2 11.92 psig initiates'CS. Also' Initiates HPCI and LPCI mode of RHR and provides a permissive signet to ADS. 3. Reactor Vessel Steam-Dome Low 2 2b22 psig* . Permissive to open CS Pressu re injection valves. l 4. Core Spray Sparger 1**8 5 3.1 psid Monitors. integrity of CS Di f fe rent i a l Pressure greater (less piping inside vessel (between negative) than the nozzle and core shroud). the normal g indicated AP at rated core power N. and flow. l 5 5. CS Pump Discharge Flow Low 1 2610 gpa Minimum flow bypass line is (2 4.13 inches) closed w"Hrn low flow signal is not present. 6. Core Spray Logic Power 1 Not Applicable Monitors avails ~ llity or o Failure Monitor power to logic system. ~ t I

This trip fur.; tion shall be %500 psig.

a. The column entitled "Ref. No." is only for convenience so that a one-to-one relat ship can be established g between items in Table 3.2-6 and stems in Table 4.2-6. cr b. When any CCCS subsystem is required to be operable by Section 3.5, there shall be two operable trip systems. If the required number of operable channels cannot be met for one of the trip systems, + 7 place the inoperable channel in the tripped condition or declare the associated CCCS inoperable c+ within 1 hour. if the required number of operable channels cannot be met for both trip systems, f z declare the associated CCCS inoperable within 1 hour. o c. Alarm only. When inoperable, verify that the Core spray differential pressure is within limits at least once per 12 hours or, declare the associated core spray loop' inoperable. O { l; .,=c__ c = = = ---a ~ ~ ~ - a -- -- l -

~ = ~;3 7-g.'- ^

Q _ [:f

-9.

s;

..o ,8 + ?.. . Table 3.2-8 RADI ATIOft MoselTORIIIG SYSTEMS WHICH LIMIT RADIOACTIVITV RELEASE h, cZ Ref. Instrument Trip Required Trip Setting Action to be taken if Remarks a No. Condition Ope rable there are not two operable (a) .Nomencts-Channels or tripped trip systees e ture per Trip 2 System tb1 g 1. orr-gas upscate/ 1 At a value not (c) (.') 2 upscases, or 1 w Post Treatment Downscale to exceed the downscale and 1 Radiation equivalent of vescale,' or 2 down- . scales will isolate-the stack re-Monitors the SJAE off-gas lease aimit indicated in Environmentai Tech Specs 2. Rerueling Floor Upscale 2 At a value not Cease refueling opera-2 uescale will to exceed the tions, it in progress. isolate the secondary Exhaust vent Rad ia tion equivalent of Isolate the secondary containment and - Monitors the stack re-containment and start initiate the standby lease limit the standby gas treat-gas treatment system indicated in ment system. Enviconsentai Tech Specs-3. Reactor Stdg. Upscale 2 S20 er/hr Isolate the secondary 2 urscale will isolate- - 2 g containment, start. stand-the secondary con-- Exhaust vent by gas treatment system, tainment and initiate. N Radiation close primary contain-the standby gas ,L Moni to rs .eent and vent valves. treatment system. c) Es. Control Room Downscale 1 20.015 er/hr Refer to Specifications. 1 upscale or 2 down-u- 3.12.C.'and 3.12.D. scales will actuate Intake the MCRECS la the Radiation Hi 51.0 ar/hr control room pres. Monitors surIzation mode.- b= 0S (D= c+ Z .O O

l -, .,... ; y: e 4 ' n S,.! J I ~- 4 j t t =~ x

; i-==:: ~ := c ^

m-

^,.. -..

~.'p,' ~. yf -T - j; 3 7 _..= ~= 3 = 3 = a f ::a . y;. -;p:: n,3 - Notes for Table 4.2-8 ' (Cont'd)'- r' ~ -"2 _ _ -"^ Instrument functional tests are-not~ required when the instruments are not required - to be operable or c. are tripped. However, if instrument functional tests are missed, they shall be performed prior to returning the instrument to an operable status. Instrument calibrations are not required when the instruments are not required to be operable or are d. tripped. However, if instrument calibrations are misse6, they shall b-arformed prior.to return-ing the instrument to an operable status, val of not less than one Initially once per month or according to Figure 4.1-1 with an inte e. month nor more than three months. The compilation of instrume.F failere rate data may include data obtained from other BWR's for which the same design instrument operates in an environment 7 The failure rate date must be reviewed and approved by the AEC prior similar to that of HNP-1. to any change in the once-a-month frequency. This instrument f. This instrumentation is exempted from the. instrument functional test definition. functional test will con'sist of injecting a simulated electrical signal into the measurement - channels. Standard current eurce used which provides 'an instrument channel alignment. Calibration using a 3 radiation source shall be made once per operating cycle. logic system functional tests and simulated automatic actuation shall be performed once each operating cycle for the following: 1. Secondary Containment Actuation

',.O f. VN, t 4y [ i ,.-e. > '1; 1, 'F.' ,h-sI c he p e 9 9- .k-enL W i 7 -e ah .e. E h ..e .m ,D 2- _8 'd. p.

c.

. & ! (.O. O . 3 ea 'e-e-

-e i

N,

.O., t 3 p

a

'U'.'{' 2 k 1

, ;t

.t t .~ ae ~.3 ,n,,.-.- m-gp- ~o -. W i >,. ' e e. U Le-.3: W; C, O 2 -d' C'- -C e w '. w g, y -.Oe-M-ri = O 9 p. {.. - e c' .g ,p..3 64 W -- 3 E C.= w = u:'g..o.: g.L U .M g _7 p

== g= M l D D .C. Q g< g.g-e p e ~* = M $ [L-g_ g .= m o = o 6. e. C o 5 x ,,~yv y. g ;, L en. CL E' u,. ; C ', ' C g O E M ~ "b y a m. g y en M m L.4 3 9 C. U M 9 L;>a O e 4 p 6 -.W W- ( C W 7J 4 e o- - O C

-Q -Q O m. e ~p o b 9 ,e U % .U. '- g '.S g s< 6 e u C U C - g - S.. e. O e e .C -C 4' s C - .e.a. w = o % O e . - Ie

o -..

L .: o. u m-L.- ..oC M. 3 N

9 # n @

b h b-w -U .a 2 _?.. L ~C O n >= b q

HATCH - UNIT 1 3.2-44 Amendment No. 170 o

u t .!G ii; j

- ~ - ...[. 4,g.; \\? ':i;! o f+ LIMITING CONDlilDNS FOR OPERATIDW SURVElLLANCE REQUIREMENTS-M L R -1 4.4. A 2. : Each Goeratina Cvtig (Continued) ,'; W 's c.- vessel; This-test checks the ~4"' explosive chariei proper opera-- 'Si" g tion of the associated valves - and selected pump operability. 'z The replacement. charge to be

EU q installed will be selected from-a manuf actured batch which has-e o

-been tested. l il " ' > <d.: Both: loops including both explo-1 sive valves should be tested in i the. course of two operating cycles. e. Prior to startup, verify (by analysis) that the sodium. pentaborate enrichment is within prescribed limits.- 3.4.B. goeratino with inocerable B. Surveillance with inoperable .Comoonents-Components If one Standby liquid Control (Deleted) - redundant component is inoperable the reactor may remain in operation for a period not to exceed seven-(1) days provided the redundant 4

component-is operatle.

1 4 C;~ Sodium Pentaborate Splution C. lodlym Pentaborate Solution-a 'At all times when the Standby The following tests shall be Liquid Control System is re-performed to verify the avail-77 l' quired to be operable the bility,of the liquid control following conditions shall be solution: met .l. Volume 1. Volume The volume of the liquid Check the standby liquid control solution in the control tank volume at least liquid control tank shall once per day. _ JU be maintained as required in Figure 3.4-1. 2. Concentration 2. Concentration The concentration of.the Check the concentration of the_ liquid control tank shall liquid in the standby liquid- ~ ;3) be maintained as required control tank by chemical in Figure'3.4-1. analysis: r Amendment No. 170 HATCH - UHli 1 3.4-2 - 1 = W

4 G p' .6: 11 4 if i 9 LIMITING CONDlil0NS FOR OPERATION SURV[lLLANCE REOUIR[M[NTS '3,$. CORf AND CONTAINMENT COOLING 4.5. CORF AND CONTAINMENT C00l!NG SYSTEMS SYS1[M5 1 Acolicability' Apolicability The Limiting Conditions'for The Surveillance Requirements Oteration apply to the apply to the core and containment operational status of the core. cooling systems when the corres-and containment cooling systems, ponding limiting conditions for operation are in effect. 'Objectivt Objective J .The objective of the Limiting The objective of the Surveillance . Conditions for Operation is to Requirements is to verify the assure the operab111ty of the operability of the core and con-core and containment cooling tainment cooling systems under all ' systems under all conditions conditions for which this cooling

for which this cooling capa-capability is an essential response-bility is'an essential to plant abnormalities.-

response to plant abnor<- malities.. ,$1gtifications Specifications A.. Core Sorav fCS) System A. Core Sorav fCS) System .o

1. ' Normal System Availability 1.

Normal Operational Tests 4. The CS-System shall be operable: CS' system testing-shall be performed as follows: (1)' Pricr to reactor startus, from : cold. condition, or item Freauency (2). When irradiated fuel is in the a. Simulated 'Once/ Operating reactor vessel and the reactor Automatic Cycle. . pressure is' greater than Actuation =* m pheric pressure, except as Test ~1 in Specification 3.5.A.2. t b. System flow Once/3 months. rate: Each 100D: can i' deveTop at l: least 4250' gpm against a system head corresponding 'to a reactor 4 vessel pres-sure of-at least 113 psig. c. Valve lineups: Once/31 days. Verify that each valve in

ci =

the flow path b' that is not locked, sealed, or otherwise secured in post-tion is in its i-. correct position. d. (Deleted) -1; Amerldment NO. 170 HATCH - UNIT 1 3.5-1 M,,.. ' S, i p_

~

t, i

LIM 111d6 CON 01110NS FOR OPERA 110N = $URy[lLLANC[ R[0UIREMENTS

3.6. A.2. - Gnaration with Inonarable -

"4.6.A.2. Surveillance with Inonerable ConDonents COnDonenis If one C$ system loop is inoper- - (Deleted) able, the reactor may remain in operation for a period not to exceed 7 days providing all - l' active components in the._- other C$ system loop, the RHR system LPCI mode and the diesel generators (per Specification 4.9.A.2.4) are operable.- When performing an inservice' hydrostatic or leakage test with the reactor coolant temperature above or below 212'F the CS system is not required to be operable.. 3. Shutdown R,t.quirements If Specification 3.5.A.I.a. or 3.5.A.2. cannot be met the reactor shall be placed in the Cold Shutdown Condition within 24 hours. B. Residual Heat Rgmdo al (RHR) B. Residual Hegt Removal (RHR) System (LPCI and Containment $vstem (LPC,( and Containment Coolina Mode) Go_olino Mode) o 1. (in_rmglsystem Availabilitv-1. Normal Operational Tests RHR systen. testing shall be performed as.follows: Item Frecyenly e a. The RHR System shall be operable: a. Air test on Once/10 years. l drywell head-(1) Prior to reactor startup ers and nozzle, from a cold condition, or and air or water test on (2) When irradiated fuel is in torus headers the reactor vessel-and the. and nozzles reactor pressure is greater-than atmospheric except as stated in Specifichtton 3.5.B.2.

HATCH - UN!i 1 3.5-2 Amendment No. 170

' ' - ~ n, m _ L 1 LMbib ^ f p '.;g s q A i m[ LIMITING QtDITIONS FOR OPERATION SURV[]LtANCE RIOUIRENENTL Ql[,L, ~it Normal Operational Tests W 3.5.8.1.' Normal System Availability (Cont.1. -.4.5.6.1, i v,k

b.

One RHR loop with two pumps or.twoj ligg Frecuency j i s loops with one pumo per loop shall Ai T be operable in the shutdown cool- ' b. Simulated .Once/ Operating. T ,O. .ing mode when irradiated fuel is Automatic -Cycle. ?b ' in the reactor vessel and the Actuation jb reactor pressure is atmospheric Test U/jh ; e3 Cept pr br to a reactor startup as stated in Specification

3. 5. 8.1. a.' During an inservice hydrostatic or leakage test, one-M RHR loop with two pumps or two Pb,

loops with one pump per loop-shall also be operable in the LPCI mode. W c. The reactor shall not be started up. c. System flow Once/3 months, c with the RHR system supplying rate -Cach l~ cooling to the fuel pool. RHR pump <p shall deliver P" d. During reactor power operation, the at least 1700 LPCI system discharge cross-tie gpm against a i valve. Ell-F010, shall be in the system head closed position and the associated corresponding valve motor starter circuit - to a reactor M'. ' breaker shall be locked in the vessel pressure L off position. In addition, an of at~least 20 Ip annunciator which indicates that psig. the cross-tie valve is not in the fully closed position shall be d. Valve lineups: Once/31 days. available in the control room.' Verify that $^" 3 each valve in e. Both recirculation pump discharge the flow path . valves shall be operable prior to that is not reactor startup (or closed if per-locked. sealed, mitted elsewhere in these speci-or otherwise

1 cations).

secured in posi- .T tion is in its correct position. 2. Doeration with Inoperable e. (Deleted) 1 Components s }n -f. Both recirculation pump discharge ~ One LPCI Pumo Inoperable valves shall be tested for oper-a. Eo ability during any outage exceeding If one LPCI pump is inoperable. 48 hours, if operability tests have s < the reactor may remain in opera-not been performed during the Lion for a ptriod not to exceed preceding month. 7'daysprovidedthattheremainingl LPCI pumps, both LPCI subsystem 2. Surveillance with Inocerable flowpaths,theCSsystem,andthel Components associated diesel generators are operable-(per Specification a. (Deleted) s .p. 4.9.A.2.a). yim b. One LP_Cl Subsyslem Inoperable b. (Deleted) A LPCI subsystem is considered to be Inopterable if (1) both of the [yd, LPCI pumps within that system are N inoperable or (2) the active valves in the subsystem flow path g are inoperable. i i HATCH - UNIT 1 3.5-3 Amendment No. 170 9- -ni I. !W

9;. + 8. l (if <

p..

2 ,o 2 t LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS s,'. g '3.$.8.2. @eration with InoDerable 4.6.8.2. (Deleted) 4 i - Comoenents (Continued) = b. If one LPCI subsystem is inoper- ' able. the reactor may remain in ,-s operation for a period not to . exceed 1 days provided that l t L all active components of. the remainingLPCIsubsystem,theC5l-Jystemi and the associated diesel-generators are operable

(per Specification 4.9.A.2.a).

c. - When perf orming an inservice

. hydrostatic or leakage test ' with the reactor coolant temperature above or below 212*Fi~ comply with Specification 3.5.B.1.b. \\; * - t 1 e I k_ t '\\.' p /t J .r 'l

V i.

e W, I J HATCH - UNIT 1 3-4 Amendment No. 170 5 - J h g 'l ?, h

c-i V... l 0-1 ' '~ 3, LIMITING CONDITIONS FOR CIPIRATION SURv[lLLANC[ k(OUIR[MENTS

3.5.B.3.

Shutdown Reautrements If Specification 3.5.8.1.a.' or, j 3.5.8.2. cannot be met, the reactor shall be placed in the Cold Shutdown Condition within 24 hours. -l C. RHR 5ervice Water System 4.5.C. RHR service Water System ] 1. Normal System Availability 1. Normal Doerational Tests The RHR service water system RHR service water system testing shall be operable: shall be performed as follows: ltt'g - Freauency

a.. Prior to reactor startup 4.

Valve lineups: Once/31 days. 'from a Cold Shutdown Verfiy that Condition, or each valve in the flow path that is not locked, sealed, i l-or otherwise secured in posi-tion is'in its correct position. b. When irradiated fuel is in b. Pump Capacity Once/3 the reactor vessel and the Test: months. reactor vessel pressure is -[ach RHR ser-f l~ greater than ~ atmospheric vice water I l-- pressure except as stated in pump shall l' Specification 3.5.C.2 or l deliver at least 4000 gpm at a system head l, of at least u B47 feet. 1: l c. When irradiated fuel is in-

the reactor vessel and the reactor-15 depressurized at least oneiRHR service i

water loop shall be opera ble. l 2. One PumD Inoperable 2. One Pump Inoperable If one RHR service water (Deleted) pump is inoperable the l reactor may remain in operation for a period not to exceed ' 30 days provided all other active components of both subsystems tre operable. l~ When perform)*3 an inservice l L hydrostatic or leakage test, i comply with Specification 3.5.C.1.c. i HATCH - UNIT 1 3.5-5 Amendment No. 170 i 4

,y ps3 q, i jQWfh y 3 /% r, t .n g ) g 4(; g 4 ,4~,< 9J j u-a .,. i g.. an M iU }C - 4 LIMillNG COND1110NS FOR OPERA 110N SURV[lLLANCE RE0VIRENENIS '3.5.C.3. Two'Punos Inoperable 4.5.C.3. Two'Punos Inoperable 3 1 e- ' q,;:

e

. &.a. gin T If two RHR service water pumps are (Deleted) F ' +"" ' ; inoperable, the reactor may retain hg in operation f or a period not to -1 '~ exceed 7 days provided all redun-l dant active components in both of the RHR service water sut' systems s are operable. i 4 Shutdown Recuirements O M If Specifications 3.5.C cannot be met, the reactor shall be placed in the Cold Shutdown Condition within 24 hours.- L,' Y D. High Pressure Coolant injection D. High Pressure Coolant injection .[ (HPCI)$vstem (HPCI) System b 1. Normal System Availability <l. Normal Doerational Tests ,[ L HPCI' system testing shall be performed as follows: l' Item Frecuency l h .a. The HPCI System shall be a. Simulated Once/ Operating- !~ operable: automatic Cycle. .i l' actuation

(1) Prior to reactor startup-l test y

from a cold condition, or b.(1) Flow rate Once/3 5 I l (2) When irradiated fuel.is in l for a system, months, li the reactor vessel and the head corre-ii greater than 150 psig, except l. reattor vessel pressure,is .sponding to a reactor i-as stated in Specification vessel pres-- 3.5.D.2.* sure of 2 1000 psig when steam is being sup-plied to the turbine at 5j 1000 psig, and-(2) Flow ~ rate for Once/ Operating N ' a system head Cycle. l corresponding i to a reactor I, vessel pres-i- sure of 1 L 165 psig when steam is being supplied to the l turbine at 165 l & 15 psig. 1 i

HPCI is not required to be operable for performance of inservice hydrostatic or leak testing with reactor pressure greater than 150 psig and all control rods inserted.

[ .-HATCH - UNIT 1 3.5-6 Amendment No. 170

~ 4 i 4 4 k, '=, - '7 T LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS '4.5.D.1.b.. Normal 00erational Tests the HPCI pumps shall deliver at least 4250 gpm during each-n flow rate test. 3.5.0.2. Doeration with Ir.cottill.t c. Valve lineups: Once/31 days. 3O/ Ccaponents-Verify that each valve in if the HPCI system is inoperable, the flow path the reactor may remain in opera ' that is not tion for a period not to exceed locked, sealed, fourteen (14) days provided the or otherwise I. ADS, CS system, RHR system LPCI secured in post-mode, and RCIC system are operable.- tion is in its p correct position. With the surveilli.e requirements t of Specification 4.5.0.1. not per-2. Surveillance with Inocerable formed at the required frequencies Components due to low reactor steam pressure, reactor startup is-permitted and.. (Deleted) the appropriate surveillance will be performed within 12 hours after reactor steam pressure is adequate (i.e. reactor pressure is such that the required steam pressure is maintained at the turbine for the duration -of.the test) to perform the tests. 3. Shutdown Reautrements

If-Specification 3.5.0.1. or

~3.5.0.2. cannot be met,'an orderly

thutdown shall be -initiated and

'the reactor vessel pressure shall be reduced to 150 psig or less withir 24 hours. E. Reactor Core Isolation Coolina E. Reactor Core Isolation soolina (RCIC) System (RCIC) System 1. Normal $vstem Availability 1. Normal Operational Tests a. The RCIC system shall be RCIC system testing shall be per-operable with an operable formed as follows: flow path capable of (auto-matically) taking suction item Frecuency from the suppression pool and transferring the water a. Simulated Once/ Operating to the reactor pressure Automatic Cycle. vessel: Actuation (and restart *) (1) Prior to reactor startup

Test, from a cold condition, or i

Automatic Restart on a low Water level which is subsequent to a High Level Trip.

HAiCH - UNIT 1 3.5-7 Amendment No. 170

F-4 h a i E M 2? b.. LIMITING COPDli10N5 FOR OplR&l10N SURV[lLLANCE R100 lit [ MEN 15 my h, D #

3. 5. E.1.

Normal System Availability (Cont.) 4.5 [.1. Normal Doerational Tests (Cont.) ~ b. Verifying that suc. Once/ F a.(2)~ men there is irradiated -tion for.the RCIC .. Operating 4 fuel in the reactor vessel system is automati-Cycle. 'W and the reacter pressure cally transferred is above 150 plig, encept: from the CST to the-Gli? ' as stated in Specification suppression pool on 3.5.t.2.* a simulated low C$T

d..,,

level or high sup-pression pool level signal. c.(1) Flow rate when Once/3 steam is being months. p H supplied to the turbine at nor - mal reactorives- .L sel operating F pressure. 1000 + i 4 20 -80 psig, and . g' (2) Flow rate when .Once/ oJ steam-is being operating h supplied-to the Cycle.- ( turbine.et a pres-sure of 150 + 15 f{' -0 psig. v\\ The RCIC pump shall deliver at least 400 gpm during each flow test. ( d. Yalve lineups: Once/31 days. Verify that each valve in 's the flow path

f 2.

Qoeration with Inoperable that is not components locked, sealed.

4 '

or otherwise' If the RCIC system is inoperable, secured in post-a -o the reactor may remain in oper-- -tion is ir,its-ation for a' period not to exceed correct position. y$ ' I days if the HPCI system is .l 4 operable during such time, e. (Deleted) With the surveillance requirements of Specification 4.5.E.1 not 2. Surveillance with Inocy able performed at the required fre-Components 4 quencies due to low reactor .m Z* steam pressurc, reactor startup (Deleted) 'is permitted and the appropriate surveillance will be performed ~a within 12 hours after reactor t steam pressure is adequate a a. h (i.e., reactor pressure is such ' n $!" that the required steam pressure is maintained at the turbine hlm i for the duration of the test) to perform the test. 3. If Specification 3.5.E.1. or [2-3.5.E.2. is not met, an orderly i shutdown shall be initiated and /!' the reactor shall be depressurited to less than 150 m 7* psig within 24 hours.

RCIC is not required to be operable for performance of inservice hydrostatic or leak

-testing with' reactor pressure greater than 150 psig and all control rods inserted. HATCH - UNIT 1 3.5-0 Amendment No. 170 l 4 9 m i.

Q 1 ~? _. ii 1 4 g LIMlllNG CONDIll0NS FOR OPERAfl0N SURVEILLANCE RE0VIREMEN15 P y-L 3.5.F. Automatic Deoressuritation System : 4.5.F. Automatic Deoressurization system .i X L&O).1 - LA111 j 1. Normal System Availability 1. Normal Doerational Ten,1 ,j 2 The seven valves of the Automatic (

Depressurization System shall be a

operable:

a. -Prior to reactor startup from a a.

A simulated automatic -cold shutdown, orL actuation. test shall be .-i performed on the ADS prior to startup af ter each refueling outage. Survell-1 lance of all relief valves is covered in Specification 4.6.H. + b. When there is irradiated fuel in b. A leak rate test of each the reactor vessel and the ADS valve accumulator, check i reactor.15 above 113 psig except valve, and actuator assembly as stated in Specification shall be performed during 3.5.F.2.* each refueling outage at a 1 pressure of 90 1 18 psig. - [ The leakage rate shall be-verified to be 1 4.5 5 FH. ( - 2. goeration with inoperable 2. Surveillance with Inoperable X LIMlilNG CONDI?l0NS FOR OPERAll0N $URV[lLLANCE REQUIREMENT $ c, 3.5.6. Min' ge Core and Containment. - 4. 5. 6.' Surveillance ~cf Core and contain-P C oo' ine Systems Avai ability ment Coolina Systems sh During any period when one of: . When1it isidetermined that' one M the standby diesel enerators-- of.the standby diesel; generators-is inoperable, cont nued reactor ' operation is limited to 7 l 1s inoperable, all of..the Compo-x i days unless operability of the nents.in the RHR-system LPCI diesel generator is restored mode.and Containment Cooling- .i l boIt omNn !s l mode Connected to the operable in the RHR system LPCI mode diesel generators'shall be veri-l and containment cooling mode fied to be operable, i mode shall be operable. If this l}r requirement cannot be met, an ' orderly shutdown shal/ be initi- . ~ ' -ated and the reactor shall be in the Cold shutdown Condition r' ' within 24 hours. Specification - 3.g. provides further guidance on l-electrical system availability. Any combination of inoperable components in the core and con-tainment cooling systems shall lv not defeat the capability of 4 the remaining operable components L to fulfill the core and contain-ment cooling functions. When irradiated fuel is in the reactor vessel and the reactor is in the Cold Shutdown Condt-tion, both CS systems and the-l - LPCI and containment cooling subsystems of-the RHR system may be inoperable provided that the shutdown cooling subsystem of the RHR system is' operable in accordance with Specification 3.5.8.1.b and that no work-is being done which has the ' potential for draining the. reactor vessel. s a f HATCH - UNIT 1 3.5-10 Amendment No. 170

- 91 3-i "l-' y-i; . jr .St 1 /i 4: h, ... r,h.-3; L1R111NG CONDlil0N5 FOR OPERA 110N - SURVElliANCE REQUIREMENi$ '*- Up s.. 3'. 5. H ;. Maintenance of Filled Discharat 4.5.H. Maintenance of Filled Discharoe f.iltti P1 Del , Whenever the CS system, LPCI, ) The following surveillance re. HPCI, or RCIC are required quirements shall be performed -to be operable, the discharge to assure that the discharge piping-from the pump discharge piping of the CS system, LPCI, l i C Lof these systems to the last HPCI, and RCIC are- -block valve shall be filled, filled when required:. The suction of the HPCI pumps shall be aligned to the conden-1. Every month, the discharge sete storage tank, piping of the LPCI and CS systems shall be vented from the high point and water . t_. flow observed. 2. FcIlowing any period where the LPCI or C$ systems l have not been required to be operable, or have been a E inoperable, the discharge. l. . piping of the system or sys-tems being returned to ser-vice shall be vented from the high point prior to re-turn of the ;ystem to service. 3. Whenever the HPCI or RCIC ,..R system is lined up to take suction from the condensate 'Ki storage tank, the discharge piping'of the HPCI and RCIC i shall be vented-from the high point of the system and water flow observed on a monthly basis. s 4. The level switches which monitor the discharge lines shall be functionally tested F r every month and calibrated l:- ) every 3 months. 1. Minimum River tevel 1. Minimum River Level 1. If the water level, as The water level as, measured ,ii measured in'the pump well, in the pump well, and the is-less-than 61.2 ft MSL, ' level in the river

shall the discharge from each plant be verified with the follow-service water (PSW) pump will ing frequencies be throttled such that each pump does not exceed 7000 gpm.
Level (MSkl Frequency 2. If the water level, as measured l 1. > 61.7 ft 81 weekly, in the pump well, decreases to less than.60.7 ft MSL, or if 2. < 61.7 ft Every 12 hrs, the level in the river
drops to a level equivalent to less

0nly pump veil monitoring is required if a temporary weir is not in place.
HATCH - UNIT 1 3.5-11 Amendment No. 170

.-.----.___.-____-_L_.

L________
p, i m., 'V y 3 1 l OI. [4.n." e {'~, LIMITING CONDlil0NS FOR OPERATION $URVElLLANCE REQUIR[MENi$ s. V) ,than 60.7 ft in the pump well> 'of>the intake structure, an-3 orderly shutdown of_the reactor ~ shall be-initiated, and the d' reactor shall be in the Cold. Shutdown Condition within 24 . hours until the level.in the n . river. is greater than' or equal to 60.7.f t MSL equivalent in- ' the pump'well.. .3.5.J. ~ Plant Service Water Sys'am 4.5.J. Plant Service Water System 1. Normal Availability = 1. The automatic pump start functions and= automatic The reactor shall not be isolation functions shall made' critical from the be tested once per operating Cold Shutdown Condition cycle. unless the PSW System '(including four PSW pumps and the standby-service water pump) - is operable. 2. Lnoperable Components 2. InoDerable Components a. The' standby service water a. With the standby service pump may be-inoperable for water subsystem inoperable a period not to exceed 60 for up to 60 days, provide s: -days provided that an alter-Unit I service water cooling j -nate Unit 1. P5W water cool-l to the 18 diesel generator i ing source to the IB diesel by verifying.0PERABILITY of an generator is OPERABLE. alternate Unit i service water l cooling source within 8 hours. l Otherwise, declare the 18 diesel generator inoperable - and take the action required I by Specification 3.9.B.2. b.- One P5W pump may be in-b. (Deleted) l operable for a period not to exceed 30 days provided l- .all other P5W pumps and the standby service water pump-are operable, il c. One PSW pump and the stand-c. (Deleted) by service water pump may l-. 'be inoperable for a period not to exceed 30 days pro-j vided all other PSW pumps are operable, j I d. Two P5W pumps or one PSW d. (Deleted) division may be inoperable for j a period not to exceed 7 days i provided all other PSW pumps and the standby service water pump are operable. HATCH - UNil 1 3.5-12 Amendment No. 170
Q ,$ Jii: ) w L m p[t [ y ,a,+ 1 b ; A &fa... ' u F MQ ;, W ,1 w W 1MillNG COND1110NS FOR OPERA 110N SURV[lLLANCf REQUIR[M[NIS Ct< 1 m I L3.5.J. Plant' Service Water system 4.5.J. Plant Service Water Systen [' - 3 T l 't 2.'. Inocerable Compone $ (Cont'd)L 2. Inocerable conoonents (Cont'd)- q
c y e...Two P5W pumps or one P5W-e.. When cooling water to '
division, and the standby diesel generator 1B is service water pump may be intertled with the P5W ,~ inoperable for a period divisional piping supply, not to exceed 7 days operability of the div-- provided all other P5W isional interlock-valves .~ f. pumps are operable. shall be demonstrated, y For each condition above in which the standby service water ' pump is inoperable, cooling
water to diesel generator 1B s
shall be intertled with the r P5W divisional piping supply. 3.1 Shutdown Reautrements
p. m If the requirements of Specifi-cations 3.5.J 1, and 3.5.J.2.
cannot be met the reactor shall be placed in the Cold Shutdown Condition within 24 hours. 3.5.K. Ecuipment Area Coolers 4.5.K. [avioment Area Coolers 1. The equipment area coolers 1. Each equipment area cooler serving the Reactor Core Iso-is operated in conjunction t <r yd.." L"g lation Cooling (RCIC), High with the equis. ment served by
fi?
Pressure Coolant Injection that particular cooler;, i ~ 4% (HPCI).. Core Spray or Residual therefore, the equipment area Heat Removal (RHR) pumps must coolers are tested at the be operable at all times when same frequency as the pumps the pump or pumps served by which they serve. that specific cooler is con-L sidered to be' operable. 2. When an equipment area cooler is not operable the pump (s) ,d served by that cooler must be (: iconsidered inoperable for l Technical Specification purposes. I l i5 ,\\ { c HATCH'- UNii 1 3.5-13 Amendment No. 170
+ a d1 7 ',.L
g RA$f 5 IOR LINlilNG CONDIll0NS FOR OPERAil0N AND SUR_yf dL ANCE R[0VIREMENTS f
- S' m 3.5, CORE AND CONTAINNINT COOLING SYSTEMS-l A.= tore Sorav fCS) Syst r '1l=NormalSystemAYaildkhty l 1 Analyses presented in Reference i demonstrated that the CS system provides adequate cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit fuel clad temperature 'to'below 2200*F-which assures that core geometry remains intact and to tmit any clad metal-water reaction to less than one percent. ' CS distribution has been shown in tests of systems similar l In design to HNP-l to exceed the minimum requirements. In addition, cooling effectiveness has been demonstrated at less than half the j rated flow in simulated fuel assemblies with heater rods to duplicate O the decay heat characteristics of irradiated fuel. 4 The intent of the:CS system specifications is to prevent operation above. atmospheric pressure without all associated equipment being J i d operable. However, during operation, certain components may be out of: service for the specified allowsble repair times. The allowable repair times have been selected using engineering judgment based on 'd experiences and supported by availability analysis. Assurance of the availability of the remaining systems is increased.by demonstra-I . ting operability inenediately and by requiring selected testing during the outage period, j When the reactor vessel pressure'is atmospheric, the limiting conditions for operation are less restrictive. At atmospheric pressure, the minimum requirement is for one supply of makeup water "j to the core. ' Requiring two operable RHR pumps and one CS pump provides redundancy to ensure makeup water availability. 1; 2. Operation with Inoperable Components [ Should one CS loop become inoperable. the remaining CS loop ' and the RHR system are required to be operable to ensure their . availebility,should the need for core cooling arise..The i.
surveillance testing required by Specification 4.5.A, 4.5.H. and 4.6 K ensures the availability of the remaining CS loop. The surveillance testing required by Specif' cations 4.5.B.
4.5'.H,~and 4.6.K ensures the availability of the RHR system. 'These provide extensive margin over the operable equipment needed t for adequate core cooiing. With due regard for this margin, the allowable repair time of 7 days wcs chosen. j 8. ' Residual Heat Removal (RHR) System (LPCI and Containment Coolina Mode)- ^ 1. Normal System Availability The RHR system LPCI mode is, designed to provide emergency cooling to ll the core by flooding in the event of a loss-of-coolant accident. -l
This system is completely independent of the CS system; however, it
!does function in combination with the CS system to prevent excessive fuel clad temperature. The LPCI mode of the RHR system and the CS-J
system provide adequate cooling for break areas of approximately 0.2 square feet up to and including the double-ended recirculation line break without assistance from the high-pressure emergency core cooling systems,
. HATCH - UNIT 1 3.5-14 Amendment NO.'170 i 't 1
lill,. y y* g g, 'i f. )liy6 F BA5[5 FDR LIMITING CONDITIONS FDR Op[ RATION AND SURVEILLANCE REQUIREMENTS plT }
i L3.5.9.1. _ Normal System Availability (Continued) k^{y{' ';
Observation of the stated requirements for the' containment cooling mode 1 .m : P' assures that the suppression pool and the drywell will be sufficiently-cooled.: following a loss-of-coolant accident, to prevent primary contain-FI . ment over pressuritation. The containment cooling function of the RHR.
!h' System is permitted only after the core has reflooded to'the two-thirds-core height level This prevents-inadvertently diverting water needed -
. 'I g* ' ' for core flooding to the less urgent task of containment cooling. The+ ) two-thirds core height. level interlock may be manually bypassed by a i keylock switch. The intent of the RHR system specifications is to prevent operation 14: above atmospheric pressure without all associated equipment being oper-4 4p able. However, during operation, certain components may be out=of H service for the specified allownle. repair times. The allowable repair' l,1, times have been selected using engineering judgment based on experiences, 1 4 and supported by availability analysis. Assurante of the availability b of the remaining systems is increased by demonstrating operability immediately and by requiring selected testing during the outage period.' l',-.... When the reactor vessel pressure is atmospheric, the limiting conditions-(#, _for operation are less restrictive. At atmospheric pressure, the minimum f. requirement is for one supply of makeup water to the core, p L 2. Doeration with Inoperable Components ?~,l1 With one LPCI pump inoperable or one LPCI subsystem inoperable, adequate core; flooding is assured by the required operability of the redundant LPCI pumps and LPCI subsystem and the CS system. The surveillance ttesting required by Specifications 4.5.8. 4.5.H. and 4.6 K ensures' ) the availability of the redundant LPCI pump and LPCI subsystem. The' survelliance testing required by Specifications 4.5.A. 4.5.H. and 4.6.K l-ensures the availability of,the CS system. -The reduced I. redundancy justifies the specified 7 day out-of-service period. L i \\. .n l et r HATCH - UNIT I 3.5-15 Amendment No. 170 =
. ~. 7 4 j m L,; s v, & u l,-y M. BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVE]LLANCE REOUIREMENTS
3.5.0.2.
Doeration with inoperable Components - 1 The HPCI system serves as a backup to the RCIC system as a source of f eedwater ,4 makeup during primary system isolation conditions. The ADS serves as a backup' ito the HPCI system for reactor depressurization for postulated transients and F accidents. <The ADS must be operable if the HPCI system is determined to be - E i inoperable.: In addition, the surveillance testing required by the specified' Specifications *nsures the availability of the following: CS (4.5.Ai 4.5.Hi. .P and 4.6.K), LFCI (4.5.B. 4.5.H. and 4.6 K), RCIC (4.5.E. 4.5 H and- ~1 4.6.K), and ADS (4.5.F and 4.6.K). Considering the redundant systems, an allowable repair time of 1 days was selected. E. Reactor Core Isolation Coolina fRCIC) System jg;. 1. Normal System Availability The various conditions under which the RCIC system plays'an essential role in providing makeup water to the reactor vessel have been identified by evaluating the various plant events over the full l range of planned operations.' The specifications ensure that the function for which the RCIC system was designed will be available when needed.. ~ Because the low-pressure cooling systems (LPCI and CS) are capable of provid-I l r ing all the cooling required for Ony gelant event when nuclear system pres-sure is below 150 psig, the PCIC'sye is not required below this pres-sure. RCIC system design = flow (400 gpm) is sufficient to maintain water level above the top of the active fuel for a complete loss of feedwater flow at the design power. 3 'Two sources of water are available to the RCIC system.. Suction is initially taken.from the condensate' storage tank and is automatically transferred to the 'l suppression ~ pool.upon low. CST. level or high suppression pool level. 9 -2...'doerationWithinoperableComponents M Consideration of the availability of the RCIC system reveals that the average risk associated with failure of the RCIC system to cool the core when-required is not increased if-the RCIC system is inoperable for no longer than 7 days, provided that.the HPCI system is operable during this period. The surveil-lance testing required by Specifications 4.5.0, 4.5 Hi and 4.6.K ensures the j . availability of the HPCI system.- g 'F. . Automatic Depressurization System (ADS)
1.. Normal System Availability This specification ensures the operability of the ADS under all conditions for which thefipressurization of the nuclear system is an essential response to Unit abnormalities.
The nuclear system pressure relief system provides automatic nuclear system depressurization for. small breaks in the nuclear system so that the l LPCI and the C5' systems can operate to protect the fission product barrier. l. Note that this Specification applies only to the automatic feature of the t pressure relief system. a t HATCH - UNii 1 3.5-17 Amendment No. 170 + \\
j'!. f fol f-:- d N - BASES FOR LIM 111NG CON 01110NS FOR OPERA 110N AND SURVEILLANCE REQUIREMENTS -3.5.F.1. LNormal'SystemAvailability(continued) 1J Specification 3.6. states the requirements for the pressure relief function of the valves, it is possible for any number of the valves assigned-to the ADS to be incapable of performing their ADS functions 2 because of instrumentation f ailures yet be fully capable of performing .their pressure relief function. 'Because the automatic depressuritation system does not provide makeup to X. the reactor primary vessel. no credit is taken for.the steam cooling of the core caused by the system actuation to provide further conservatism 4. to the Core Standby Cooling Systems. g' '/, the ADS valve accumulators are sired such that, following loss of the pneumatic sup?ly, at least two valve actuations will be possible with ,the drywell at 70% of its design pressure. This drywell pressure results from the largest break which could: lead to the need for rapid -depressurization through:the ADS. valves.. The allowable accumulator ' leakage criterion ensures the above capability for 30 minutes following loss of the pneumatic supply. \\ 2. poeration with Inoperable Components With one ADS valve known.to be incapable of automatic operation six valves remain operable to perform their ADS function. However, since. 'the LCCS Loss of Coolant' Accident analysis for small line. breaks assumed that all seven ADS valves were operable, reactor. operation with one ADS valve inoperable 15'only allowed to continue for 7 days provided that the HPCI system is operable and that the (remaining) six ADS valves .are operable. In addition, surveillance testing required by the-specified iipecifications' ensures the availability of the following: HPCI (4.5.0, 4.5.H. and 4.6 K) and ADS'(4.5.F and 4.6.K). 6; . Minimum Core and Containment Coolina Systems Availability I .The purpose of this Specification.is to assure that adequate core tooling equipment is available at all times. ' If, for example, one CS -{- loop were out of service and the diesel which. powered the opposite CS were out of service, only 2 RHR pumps would=be;available. i Specification 3.9. must also be consulted to determine other' -0 requirements for the diesel generators. This specification establishes conditions for the performance of major maintenance, such as draining of the suppression pool. 'The availability of the shutdown cooling subsystem of the RHR system and the RHR service water system ensure adequate supplies of reactor cooling and emergency makeup water when the reactor is -in the Cold Shutdown Condition, in l addition this specification provides that, should major maintenance be performed, no work will be performed which could lead to draining the water from the reactor vessel. HAICH - UNii 1 3.5-18 Amendment No. 170
Q 'd 1 ~ ~^ , y; g y 14 gj ;
]
$'4 li; (f[y ' ( BASES FOR LIMITING COND1110NS FOR OPERA 110N AND SURv[lgANCE Al0VIREMf M15 - s. i 3.5.H. Maintenance of Filled Discharoe Pines. If the discharge piping of the CS, LPCI ' HPCl2 and RCIC systems are not filled, a water hammer can develop in this piping when the-. I i i M pump and/or pumps are started. To minielte damage to the discharge; / W piping and to ensure added.margih in the operation of these systems, this-w a ' Technical Specification requires the discharge lines to be filled whenever t .; ". ~ .the system is in an operable condition. -If a discharge pipe is not filled, the pumps that supply that line must be assumed to be-inoperable for Specification purposes. 4 m L The CS and LPCI discharge piping high point vents are visually checked. 's - for water flow once a month to ensure that the lines are filled. N. Assurance that the HPCI and RCIC discharge piping remains filled is pro-vided by observing water flow from these systems high points monthly. ' q ;, I. Minimum River Flow W W 4, A very low-flow river stage-discharge relationship was developed at the Plant Hatch intake structure location. USGS rating data were available for flows above 1740 cf s at the Baxley gauge (at.U.S. Highway No. I bridge, o on the plant site). This data, which includes bathymetric surveys o' the l rating cross-section, were used to extend the USGS rating curve by compu-tation. Since the USGS data used in these computations result in the highest flow for a given low-flow stage ever recorded at the location, H i the computed rating curve should give a conservative low stage for a given flow. The river rating curve at the Plant Hatch intake structure f", was developed by subtracting 0.1 ft from the USGS guage evaluation for a h,, given discharge. The 0.1-ft adjustment was determined by level survey (G' when the river level at the USGS guage was approxienately 62 ft MSL. At. E '. the Plant Natch site, the river level would be 61.3 f t MSL for 1200 cf s which is the. low flow of record at Ccarlotte and 60.8 f t MSL for the X<5 hypothetical minimum low flow of 950 cfs. The minimum low flow is important because of its effect on the operation of ~ PSW and RHR service water pumps. The RHR service water pumps at rated-flow conditions require for net positive suction head (NPSH) a river stage of only i 59.0 ft. Thus, no further consideration is required on river stage with regard to submergence of these. pumps. At the rated flow of 8500 gpm each for the PSW pumps, 4-ft of submergense will satisfy the NPSH and vortening requirement. This corresponds to a stage in the pump well of 61.2 ft. Normal operation requires about 7840 gpi for each of three pumps. Shutdown or emergency conditions require only ont pump with a discharge flow of 4428 gpm. This corresponds to a pump well level of 59.9 ft for safe shutdown. For a 0.1 -f t-head loss through the trash rack and traveling screen, the corresponding river level would be 60.0 ft MSL, which corresponds to a flow of 660 cfs. Similarly, l l II V !s p 1. 1 ls l Amendment No. 170 HATCH - UNIT I 3.5-1 l l l (
%y. 9dr W ., y c, m , f( "h, my: ag t s =; BAS [S FOR LIMillNG CONDill0NS FOR OPERA 110N AND SURVEllLANCE REQUIREMtNTS 4 -3.$ J/4.5.J e Plant Service Water System 'The Planti$ervice Water (PSW) system consists of two subsystems (divisions)-. - of two pumps each and a separate standby service water pump system for diesel -i genera tor lB. During normal full power operation the two subsystems function as a 3 out _of 4 pump cross connected system supplying cooling water to the turbine and reactor building cooling systems, in the' event of an accident isignal, nonsafety-related cooling loads are isolated and the PSW pumps in the two subsystems supply cooling water to diesel generators I A and IC, the .] reactor building cooling system and the control room air conditioners, while the standby service water pump is available to automatically supply cooling water to diesel generator 1B should it be needed. Additionally, diesel IB has a manual backup water supply available f rom the Unit 1 i Division =1 or Division 2 PSW subsystems so that during maintenance _on the standby diesel service water pump, either division of the PSW system can-manually be aligned to supply cooling water to the IB diesel. -The two subsystems and the standby service water pump system are split in the accident mode fo greater reliability with one pump in each of the two [ subsystems automatically starting while a start. signal from diesel generator IB initiates standby service water pump operation. Only one of y the Division 1 PSW pumps and one of the Division 2 PSW pumps are reqdred s for cooling diesel generators I A and IC, respectively, while the standby -service water pump provides adequate cooling water to diesel generator 18. In the event that the standby service water pump is inoperable, the HNP-1 Division 1-Division 2 intertie supply piping can be aligned to cool the IB diesel, in this condition, one PSW pump-is capable of supplying the cooling requirements for the reactor building cooling system, the control room air conditioners, and the IA, IB, and IC diesel generators. ~ l . The PSW system can supply all power generation systems at full load and l-the diesel generators with redundancy if one PSW pump and/or the standby l' service water pump are inoperable. Hence, a 60-day outage time is justified if the standby service water pump is inoperable since all four L PSW pumps ere available (divisional inn rtie to IB diesel required). In addition, a 30-day outage is justified if one PSW pump is inoperable,'or. if one' PSW pump and the standby service water pump are inoperable (divisional intertie to 18 diesel required). Should two PSW pumps (or one subsystem) [ become inoperable, or should two PSW pumps (or one subsystem) and the standby ( '. service
8. ter pump become inoperable (division intertie to 18 diesel required)
I plant cc-ration will probably only continue at less than full power. However. l' safety-related loads are still adequately powered for these conditions. l l' Therefore, a T-day outage time is justified for such events. The surveillance testing required by Specifications 4.S.J and 4.6.K ensures availability of the 4 i redundant pumps and subsystem. i K. Enaineerina Safety Features Eculoment Area Coolers The' equipment area cooler in each pump compartment is capable of -providing-adequate ventilation flow and cooling. Engineering analyses indicate that the temperature rise in safeguard compartments without adequate ventilation flow or cooling is such that continued operation of the saf eguard equipment or associated auxiliary equipment cannot be assured. The surveillance and testing of the equipment area coolers in each of their various modes is accomplishec during the testing of the equipment served by these coolers. The testing is adequate to assure the operability of the equipment area coolers. L. References 1. "Edwin 1. Hatch Nuclear Plant Units 1 and 2 SAFJ/GESIR-LOCA Loss-of-Coolant Accident Analysis " NEOC-31376-P, December 1986. HAICH - UNIT 1 3.5-21 Amendment No. 170 v
u. g' 3w D fj- 'Y h * ?i! L ' j' I s 4 K LIMITING CONDITIONS FOR OPERATION-SURV(llLANCE REOUT EMENTS 'l t ~ 3p6.J2 Recirculation System . 4. 6. J.- Reci rc uT3.1),ori ' System 'M q g.y 7 1. Core thermal power shall not exceed 1r Reciiculation pump speeds shall be "~ 4 1% of rated thermal power without ' recorded at least once per! day.. ( forced recirculation. '2. With only one recirculation loop 1 2. Whenever the reactor is in the in operation, verify that the Gm START & HOT STANDBY or RUN reactor Dperating conditions are J
1. ~
modes, at least one
outside the Operation Not Allowed
'a recirculation loop shall be in Region in Figure 3.6-5( m operation. .(a) At least 64cescerf.M' hours, 3 y f 3. The requirements applicable to (b) W;ienever thermal pWer.has u , single-loop operation as identified in Sections 1.1. A, 2.1. A 3.1. A. been~ changed'by at least'5% of-3.2.G, 3.11. A, ' and 3.11.C shall be rated thernal power, and steady- ) i Lt ineffectwithin24hoursfollowing.b state r.onditions have been l + the removal of one recirculation reached. loop from service, or the' unit shall be placed in the HOT SHUTDOWN Condition within 12 hours and in COLD SHUTOOWN within the following 12 hours. ,y?' 4. With only one recirculation loop i -in operation (and the~ unit in the t l h Operation Not Allowed Region N g s -specified'in Figure 3.6-5, initiate. action within 15 minutes to place the unit in.the Operation = Allowed-1, Region,-identified in Figure 3.6-5, li 'if within 2. hours. -Otherwise, place-LM the reactor in the HOT ' SHUTDOWN l Condition within 12 hours. ~ i> I i4 ll 5. Following one. pump operation the ~ 4 discharge valve of'the low speed
j
. pump may not be opened unless the = speed of the' faster pump is less than 50% of its rated speed. /
p Q
.i f 3 L 9 i li,i = e h i~ $ ; HATCH - UNIT 1 3 />-9c Amendment No. 170 p' Jis ,g i. 1 7 ((
3.xf-f~ g Q' i:- 7.- ,' m. 5 F ADj!1lfiliRATIVE CONTROLS ~ p, ' h ' UNIT STAFF OUALIFICATIONS E ach' member of the" unit staff-shall meet or exceed the minimum N6$3.1 E ior comparable posittors, except for E qualifications of ANSI. N18.1-1911-the Health Physics Superintendent who shall meet or ex*.ted the ' qualifications of Regulatory Guide 1.8, September 1975, and the Shift Technical Advisor who shall have a bachelor's degres 'r equivalent in a -scientific or engineering discipline with specific = training in plant design,-and response and analysis of the plant for transients and accidents.- '6.4 TRAININQ A retraining and replacement training program for the_ unit staff
6.4.1 shall be maintained under the direction of the Manager of Training and shall-. meet or exceed the requirements and reconnendations of Section 5.5

of ANSI N18.1-1971 and Appendix
'A' of 10 CFR Part 55.
6-4'.2. The Fire Protectidn Program, except training, is maintained under The Fire Protection the direction of the Manager-Engineering Support.
Program meets or exceeds.the guidelines-of NFPA Code 27,.1975. " Fire Protection Training is maintained under the direction'of the Training Fire Protection Training meets or and Emergency Preparedness Manager. 27, 1975, except retraining frequency. exceeds the guidelines of NFPA Code Fire l Brigade and Fire Emergency Support Group (FB/FESG) members are required to attend retraining once per calendar quarter. 6._5 REYlEW AND AUDli .6.5._1 'PLAHT REVIEW BOARD fPRB1 = FUNCTION
6.5.1.1 'The PRB shall function to' advise the Plant Manager on all matters related to nuclear safety.
[0MPdSITION 6.5.1.2 'The PRB shail be composed of,'as a minimum, a supervisor or Lhigher level individual from each of the departments listed below: Operations Maintenance Quality Control (QC) Health Physics Nuclear Safety an' Compliance 3 Engineering Suppo i The Chairman, his alternate, and other members of the PRB shall be designated by the Plant Manager. -The Chairman and his designated alternate shall both be managers of one of the six above listed-departments or a higher level onsite manager. Q RNATE$- All alternate members shall be appointed in writing by the PRB .'6.5.1.3 Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PRB activities at any one. time. 6-6 Amendment No. 170 HAICH UNil 1 l a -'s
4C.. '7; ~ b;N ' y p ,j g n t' l '5.3.2 ~ Audit Rersonsibtitty 5.3.2 2. The General Manager-Quality Assurance is responsible for an audic, 3 :4 ' condw,ed annually, of the activities of the Plant Manager and +he Maneger-Environmental Affairs, related to cotpliance with ETS. SJ.2.2 Audits of facility activities shall be performed annually under the cognizance of the SRB to ensure conformance of facility operation to provisions of the ETS. t 5d-State and Federal Permit and Certificates Section 401 of PL 92-500, the Federa! Water Pollution control Act Amendments of 1972 (FWPCA), requires any applicant for a Federal license or permit to conduct any activity that may result in any discharge into provisions of Sections 301, 302, 306, and 307 of the FWpCA. Section 401 of PL 92-500 further requires that any certification provided under this section shall set any effluent limitations and other limitations and monitoring requirements necessary to assure that any applicant for a Federal license or permit will comply with the i.pplicable limitations. Certifica,tions provided in accordance with Section 401 set forth conditions on the Federal license or permit for which the certification is provided. Accordingly, the licensee shall comply with the requirements se forth in the currently applicable 401 certification and amendments thereto issued to the licensee by.the Georgia Environmental Protection Division. In accordance with the provisions of the Georgia Water Quality Control Act, the'fVPCA and the rules and regulations promulgated pursuant to each of these acts, the Georgia Environmental Protection Division, under authority delegated by the U.S. EPA, issued NPDES permit No. GA 0004120 to the licensee. The NPDES + ' permit authorizes the licensee to discharge from HNP Units 1 and 2 to the Altamaha River in accordance with ef fluent limitations, monitoring requirements, and other conditions stipulated in the pe rmi t'. Subsequent' revisions to the certifications will be accommodated in accordance with the provisions of section 5.6.3. l 5.5 Procedures l Detailed written procedures, including applicable checklists and instructions, shall be prepared and followed for all activities involved in implementing the ETS. All procedures shall be maintained in a manner convenient for review and inspection. Procedures that are the responsibility of the Mant Manager l l shall be kept at the plant. Procedures that are the responsibility of the Manager-Environmental Af f airs shall be kept at the Georgia Power Company General Of fice. l b HATCH - UNIT 1 5-3 Amendment No. 170 l a w
m t. 'e tt0 ~',)"' UNITED sTATis ',g NUCLE AR REGULATORY COMMISSION s wAsHiwoToN. D. C. 20bbb g 8. %,**"*/ n -GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTP.1C AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKFT HO. 50-366 EDWIN 1. HATCH HUCLEAR PLANT. UNIT ? AMENDMENT TO FACILITY OPERATING LICENSE Amendnent No.108 License No. NPF-5 1. The Nuclear Regulatory Consnission (the Commission) has found that: The app (lication for amendment to the Edwin I. Ilatch Nuclear Plant,thefacility)Faci A. Un M 2 C<orgia Power Company, acting for itselt, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and Otty of ' Dalton, Georgia,(thelicensee)datedMarch2,1990,complieswith the standards and requirements of the Atomic Energy A::t of 1954, as amended (the Act), and the Connission's rules and regulations set forth in 10 CFR Chapter I; B. The ftcility will operate in conformity kith the application, the provisions of the Act, and the rules and regulations of the Commission;- o C. Thereisreasonableassurance(1)thattheactivitiesauthorizedby this amen & ent can be conducted without endangering the health and safety.ofthepublic,and(ii)thatsuchactivitieswillbe -conducted in compliancc< with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to-the common-defense and security or to the health ano safety of the public; and E. The issuance of this amendu.ent is in accordance with 10 CFR Part 51 s of the Conunission's regulations anc all applicable requirenents have 'been satistied. y ?
s 2 Accordingly, the license is apended by changes to the Technical 2. Specifications as indicated in the attachsent to this license anendmen',, and paragraph 2.C.(2) of Facility Operating License ho. NPF-5 is hereby 6 pended to reed as fo11Ws: Technical SDecific*' 2 s The Technical $r :cifict ons contained in Appendices A and B, as revised through. ~c ^ sit No.108, 6te hereby incorporated in the license. The lic; t e shall c>perate the 1acility in accordance with the Technical Spec.. teations. 3. This itcense anandpent is effcctive as of its date of issuance.and shall be iniplenented within 60 days of isauance. FOR THE NUCLEAR REGULATORY COMu!SSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: August 30, 1990 I
_.z__ l = AT'ACHMENT TO LICENSE AMENDMENT NO. 108 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366
.4ce the following pages of Appendices "A" and "B" Technical Specifications
.ith the enclosed pages. The revised pages are identified by Amendnent number and contain vertical lines indicating the areas of change. Appendix A Retaove Page lhsert Page 3/4 5-1 3/4 5-1 3/4 6-46 3/4 6-46 3/4 7-9 3/4 7-9 3/4817 3/4 8-17 3/4 8-18* 3/4 8-18 3/4 0-21 3/4 8-21 3/4 8-22* 3/4 8-22 6-5 6-5 Appendix B 5-3 5-3
0verleaf page provided to maintain document completeness.
w
I' n j 3/4.5 EMERGENCY CORE CDOLING SYSTEMS I " ~ 3/4 5.1 HIGH PRESSURE CDOLANT INJECTION SYSTEM 2 LIMITING CONDITION FOR OPERATION o 3.5.1 ' The High Pressure Coolant Injection (HPCI) system shall be OPERABLE with: a. One OPERABLE HPCI pump, and 5. b. An OPERABLE flow path capable of taking suction from the suppression chamber and transferring the water to the reactor. i pressure vessel. 7 APPLICABILITY: CONDITIONS 1*, 2* and 3* with reactor vessel steam dome pressure > 150 psig. ACTION: 4. With the HPCI system inoperable, POWER OPERATION may continue and the provisions of 3.0.4 do not apply *, provided the RCIC i system, ADS, CSS, and LPCI system are OPERABLE; restore the 1 inoperable HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and i reduce reactor steam dome pressure to s 150 psig within the following 24 hours. b. With the surveillance requirements of Specification 4.5.1 not performed at the required frequencies due to low reactor steam pressure, the provisions of Specification 4.0.4 are not applic-able provided the appropriate surveillance is performed-within 12 hours af ter reactor steam pressure is adequate (i.e., a reactor pressure is such that the required steam pressure is maintained at the turbine for the duration of the test) to perform the tests. i SURVEILLANCE REQUIREMENTS i 4.5.1 The HPCI shall be demonstrated OPERABLE: a. At least once per 31 days by: 1. Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water, and I
See Special Test Exception 3.10.5 1
HATCH - UNIT 2 3/4 5-1 Amendment No. 108 m
a. ? CONTAINMENT SYSTEMS PRIMARY CONTAINMENT PURGE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.5.1 The drywell and suppression chamber 18-inch purge supply and exhaust isolation valves shall be OPERABLE with: i a. Each valve closed except for purge system operation for inerting, deinerting, and pressure control. b. A leakage rate such that the provisions of Specification 3.6.1.2 are met. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION: a. With an 18-inch drywell and suppression chamber purge supply and/or exhaust isolation valve (s) inoperable or open for other than inerting, deinerting or pressure control, close the open 18-inch valve (s) or otherwise isolate the penetrations (s) within 4 hours or be in at least HOT SHUTOOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, q SURVEILLANCE REQUIREMENTS 4.6,6.5.1 The primary containment purge system shall be demonstrated OPERABLE: a. .In addition to the requirements of Specification 3.6.3, at least once per 31 days, when not PURGING and VENTING, by verifying that I aach 18-inch drywell and suppression chamber isolation valve is l
closed, b..
At least once per 18 months by replacing the valve seat of each 18-inch drywell and suppression chamber purge supply and exhaust isolation valve having a resilient material seat and verifying that the leakage rate is within its limit.
H - UNIT 2 3/4 6-46 Amendment No. 108
tA l ~* r. PLANT SYSTEM i 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The Reactor Core Isolation Cooling (RCIC) System shall be OPERABLE with an OPERABLE flow path capable of (AUTOMATICALLY) taking suction from the l . suppression pool and transferring the water to the reactor pressure vessel. APPLICABILITY: CONDITIONS 1, 2, and 3 with reactor steam dome pressure > 150 psig. ACTION. a. With the RCYC system inoperable, operation may continue and the provisions of Specification 3.0.4 are not applicable provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days or be in at least HOT SHUTOOWN within the next 12 hours and reduce reactor steam dome pressure to < 150 psig within the following 24 hours. i b. With the surveillance requirements of Specification 4.7.3 not performed at the required intervals due to low reactor steam pressure, the provisions of Specification 4.0.4 are not applicable provided the appropriate surveillance is performed within 12 hours after reactor steam pressure is adequate (i.e... reactor pressure is such that the required steam pressure is maintained at the turbine for the duration of the test) to perform the tests. SURVEILLANCE REQUIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE: a. At least once per 31 days by: l 1. Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water, and 2. Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position. b. At least once per 92 days by verifying that the RCIC pump develops a flow of 400 gpm on recirculation flow when steam is being supplied to the turbine at normal reactor vessel operating oressure, 1000 + 20, l - 80 psig. l' HATCH - UNIT 2 3/4 7-9 Amendment No. 108 L I m
i 4 ELECTRICAL POWER SYSTEMS l A.C. CIRCUITS INSIDE PRIMARY CON,TAINMENT h-i [: LIMITING CONDITIONS FOR OPERATION I 3.8.2.5 The following A.C. circuits inside primary containment shali be ? de-energized *: -l a. Breaker Numbers 2, 4, 6, 8, 10, 12, 14, 40 and 42 in panel 2T51-5003, j b. Breaker Numbers 2, 4, 6, 8, 10, 12, 40 and 42 in panel 2T51-l
S004, c.
Breaker Numbers 28 and 34 in panel 2R25-5105, and d. Compartment IEL on MCC 2R24-5014. APPLICABILITY: CONDITIONS 1, 2 and 3. ACTION: With any of the above required circuits energized, trip the associated circuit breaker (s) in the specified panel within 1 hour. l SURVEILLANCE REQUIREMENTS 4.8.2.5 Each of the above required A.C. circuits shall be determined to be'de-energized at least once per 24 hours by verifying that the associated circuit breakers in the specified panels are in the tripped condition. i
Except during entry into the drywell.
HATCH - UNIT 2 3/4 8-17 Amendment No. 108
i 1 b w ? j i ELECTRICA1, PGER SYSTDIS ~ PRI?9M CCtrAD?D?r PDTETPATICH OCMtCTOP CUrn'PPEhi PPOITCTIVE DWICTS LUTTI!G CCCITICt? FDP OPSATICt* l 3.8.2.0 All prirary containcant penetration conductor overeurrent pre-tective devices shown in Table 3.9.2.6-1 shall be OPEPAPLE. j APP 13 CAP.ILITY: CC2'DITICt?S 1, 2 and 3. 5 AcrICth With one or acre of the primary containment penetratien cenfuctor over-current protective devices shown in Table 3.8.2.6-1 inoperable a. De-energize the circuit (s) by tripping the associated circuit breaker (s) within 72 hours and the provisions of 3=cification 3.0.4 are not applicable, or b. Be in at least HOT sherd 3N within the next 12 hours and in ccLD smJrD3M within the fo11cwing 24 hours. SURVEILi>mE PTJIPnT!??S 4.8.2.6.1 All primary containment penetration conductor overcurrent pro-tective devices shown in Table 3.9.2.6-1 shall be demonrtratcf CPEPABLE: a. At least once per 18 renths: 1. Ibr at least one 4 KV recetor recirculation purp circuit, such that both recirculation purp circuits are demonstrated OPEPAELE at least once per 36 renths, by perforrance of; (a) A CHAtNFL CALIBRATICE of the associated protective relays, and (b) An integrated system functional test Which includes simulated autcmatic actuation of the system and verifying that each relay and arsociat.e5 circuit , breakers and control circuits function as designed. l 2. Fbr rolded case circuit breakers, by performance of a functional test of at least ene circuit breaker of each t)T>e e such that all circuit breakers of each type are derenstrated OPERACLE at least once per N x 18 renths, where N is the nurber of circuit breakers of each type. The functional test shall conrist of injecting a current input as specified by PD S A92-1980 to the circuit breaker and verifying that the circuit breaker functicns as designed. Erould any circuit breaker fail to functicn as designed, all other circuit breakers of that type shall be tested. B.TCH - UFIT 2 3/48-19 Amendment l'o. 60 .s - ~ - - - u
(, I TABLE 3.8.2.6-1 (Continued) PRIMARY CONTAINMENT PiNETRATION CONOUCTOR OVERCURRU4T PROT ;CTIVfDIVICE5 $YSTEM/ COMPONENT DEVICE NUMBER POWERED AND LOCATION
c, Type 3:
1. 600 VAC, MCB, T.M. RECIRC. PUMP MOTOR HEATER 2R24-$014 COMPT. 5E 2B31-C0018 2. 600 VAC, MCB, T.M. REAC1 C RECIRC. PUMP MOTOR 2R24-5013, COMPT. 5B HEATER 2B31-C001A MCB, T.M. DRYWELL COOLING UNIT 600 VAC,3. COMPT. 38 3, 2R24 501 2T47 8010A MCB, T.M. ORYWELL COOLING UNIT 600VACl4, COMPT.BA 4. 2R24-$0 2T47*B0102 d. Type 4: 1. 140 VAC, MCB, T.M. CABLE BHXB0BC05 2R25-5102, CKT. 10 2, 120 VAC, MCB, T.M. CABLE BGX70BC05 2R25-$101, CKT, 10 e. Type 5: 1. 600 VAC, MCB, M.0. ORYWELL E00lr. DR. $ UMP 2R24-5014, COMPT. 2A 015CH. MOV 2G11-F018 2 600 VAC, MCB, M.D. ORYWELL EQUIP DRAIN $ UMP 2R24-5014, COMPT. 6C RECIRC. MOV 201*-F015 3. 600 VAC, MCB, M.0. RCIC $TEAMLINE INBOARD 2R24-50128, COMPT. 4A !$0. MOV. 2E51-F007 4, 600 VAC, MCB, M.0. RHR HEAD SPkAY l$0LAT10N 2R24-50ll, COMPT. 9A MOV. 2E11-F022 5. 600 VAC, MCB, M.0. HPCI STEAM LINE INBOARD 2R24-5011A, COMPT. 4A !$0LAT10N MOV. 2E41-F002 6. 600 VAC, MCB, M.0. RWCU INBOARD ISOLATION 2R24-5011, COMPT. 14C MOV. 2G31-F001 7. 600 VAC, MCB. M.0. MAIN STEAM LINE DRAIN 2R24-5011, COMPT. ISB MOV. 2B21-F016
M.C.B. - molded case circuit breaker M.0. - magnetic only T.M. - thermal magnetic HATCH - UNIT 2 3/4 B-21 Amendment No. 108 4
1 [- s I fu l T m v 3.8.2.6-1 (Continued) PMMytt CENTADHDR PDETRTIG4 CQctIt'!cP, On.n.WJENr rrut u.ctIVE ptV. ICI:s Sf8TIWCQ90HDR DIV2CE HlHEER PCWr. PPD AMD IcCE20tM a G. Type 6: 1. 600 VAC, PCs, M.0. IctP 'A' PtNP SLET10t* 2R24-ID18A, COFr. 2A 90/ 2831-!C23A 2. 600 VAC, FCB, M.(, IMOP 'A' PWP IESCM. 2R24-EC18A, 00WT. 28 M 2B31-it31A 3. 600 VAC, MCB, M.C. IctP 'B' PUW CLt'rIct! 2R24-IC1BB, COWT. 3A 60/ 2831-7023B 4. 600 VAC, FCB, M.0. lat? 'B' PLEP UIB:H. 2R24-80188, CCHPr. 3B 60/ 2B31-7031B 5. 600 VAC, FCB, M.0. DRMIL IDUIP. TRAIN 2R24-3014, CCtFr. la PUW B 2G11-C006B 6. 600 VAC, PCR, M.0. DRWILL FIA0R DPAIN StW 2P24-5014, CCbPT. 7D PtHP 'B' 2G11-C0018 Dl%221 FicoP. PPAIN SUMP 7.- 600 VAC, ICB, H.C. 2R24-3013, CGFr. 4A PUMP 1A 2G11 c001A 8. 600 VAC, PCB, M.O. DI'AULL FDUIP. ISAIN StHP 2R24-5013, CatFF. 48 PtNP A 3G11-0006A 9. 600 VAC, FCB, M.0. DR& ELL C0 CLING UTP 2R24-5012, 00FFr. 188 2T47-2007Bi
10. 600 VAC, SCB, M.0.
DPWU21 C0320 tNIT 2R24-9012,- COMPT.19A 2T47-c001B
11. 600 VAC, PCB, M.D.
NIP SHt.UTON COCLING 2R24-5011, CCMPT. 6C ISO. FD/ 2E11-It09
12. 600 VAC,tCB, M.0.
DRnUIL CCCLDD M1IT 2R24-S011, CCw T. 18A 2T47-B007A 13, 600 VAC, FCD', M.C. DR&ILL COCLDU PerUPM 2P24-5011, COWT.1BC AIR FAN 2T47-c001A,
M.C.B. - solded case circuit breaker
. M.C. - sagnetic only T.M. - thermal angnetic 3/4 6-22 /cendment No. 6, 60 IUca - uuT 2 2
I ADMIN!$TRATIVE CONTROLS 6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualtftcations of AN51 N16.1*1971 for comparable positions, except for the Health Physics Superintendent who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and the $htft Technical Adytter who shall have a bachelor's degree or equivalent in a scientific or engineering discipilne with specific training in plant design, and response and analysts of the plant for transients and accidents. 6.4 TRAINING-6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Manager of Training and shall meet or esceed the requirements and recommendations of section 5.5 of AN$1 N18.1-1971 and Appendis A of 10 CFR part 55. 6.4.2 The Fire Protection Program, except training, is maintained under the direction of the Manager-Engineering Support. The Fire Protection Program meets or exceeds the guidelines of NFPA Code 27, 1975. Fire Protection Training is maintained under the direction of the Training and Emergency Preparedness Manager. Fire Protection Training meets or exceeds the guidelines of NFPA Code 27, 1975, except retraining frequency. Fire Brigade and Fire Emergency Support Group (FB/FE$G) members are required to attend retraining once per calendar quarter, 6.5 REVIEW AND AUDIT 6.5.1 PLANT REVIEW BOARD (PRB) FUNCTION 6.5.1.1' The PRB shall function to advise the Plant Manager on all matters related to nuclear safety. COMP 051 TION 6.5.1.2 The PRB shall be composed of, as a minimum, a supervisor or higher level individual from each of the departments listed below: Operations Maintenance Quality Control (QC) Health Phystes Nuclear Safety and Compliance Engineering Support The Chairman, his alternate, and other members of the PRB shall be designated by the Plant Manager. The Chairman and his designated alternate shall both be managers of one of the six above listed departments or a higher level onstte manager. ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PRB Chairman to serve-on a temporary basis; however, no more than two alternates shall participate as voting members in PRB activities at any one time. HATCH - UNIT 2 6-5 Amendment No. 108
'c'- s. k 5.3.2 Audit Responsibility 5.3.2.1 The General Manager-Quality Assurance is responsible for an audit, conducted annually, of the activities of the Plant Manager and the Manager-Environmental Affairs, related to compliance with ETS. 5.3.2.2 Audits of facility activities shall be performed annually under the cognizance of the SRB to ensure conformance of facility operation to provisions nf the ETS. $.4 State and Federal Permit and Certificates Section 401 of PL 92-500, the Federal Water Pollution Control Act Amendments of 1972 (FWPCA), requires any applicant for a Federal license or permit to conduct any activity that may result in any discharge into provisions of Sections 301, 302, 306, and 307 of the FWPCA. Section 401 of PL 92-500 further requires that any certification provided under this section shall set any effluent limitations and other limitations and monitoring requirements necessary to assure that any applicarit for a Federal license or permit will comply with the applicable limitations. Certifications provided in accordance with Section 401 set forth conditions on the Federal license or permit for which the certification is provided. Accordingly, the licensee shall comply with the requirements set forth in the currently applicable 401 certification and amendments thereto issued to the licensee by the Georgia Environmental Protection Division. In accordance with the provisions of the Georgia Water Quality -Control Act, the FWPCA and the rules and regulations promulgated pursuant.to each of these acts, the Georgia Environmental Protection Division, under authority delegated by the U.S. EPA, issued NPDES permit No. GA 0004120 to the licensee. The NPOES permit authorizes the licensee to discharge from HNP Units 1 and 2 to the Altamaha River in accordance with effluent limitations, monitoring requirements, and other conditions stipulated in the permit. Subsequent revisions to the certifications will be accommodated in accordance with the provtsions of section 5.6.3. 5.5 Procedures Detailed written procedures, including applicable checklists and instructions, shall be prepared and followed for all activities involved in implementing the ETS. All procedures shall be maintained in a manner convenient for review and inspection. Procedures that are the responsibility of the Plant Manager shall be kept at the plant. Procedures that are the responsibility of t.he Manager-Environmental Affairs shall be kept at the Geurgia Power Company General Office. HATCH-UNIT 2 5-3 Amendment No. 108 __}}

Person / Time
ML20059G774
Site:	Hatch (DPR-57-A-170, NPF-05-A-108, NPF-5-A-108)
Issue date:	08/30/1990
From:	Matthews D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20059G777	List: ML20059G774 ML20059G779
References
NUDOCS 9009130142
Download: ML20059G774 (48)
v • d • e

ML20059G774

Text

Attachment:

Navigation menu

Search