ML20087M870

Proposed Changes to Tech Spec Re Grid Undervoltage Protection,Replacing 840209 Submittal

ML20087M870
Person / Time
Site:	Oyster Creek
Issue date:	03/23/1984
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20087M862	List: ML20087M861 ML20087M870
References
NUDOCS 8404020096
Download: ML20087M870 (10)

Text

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FUNCTION LIMITING SAFETY SYSTEM SETTINGS

7) Low Pressure Main Steam Line, h 82S psig (initiated in IRM

.MSIV: Closure range 10)

.8) Main Steam Line-Isolation Valve 610% Valve Closure from full Closure,-Scram open

.9) Reactor Low Water Level, Scram '

a 11',5" above the top of the active fuel as indicated under normal operating conditions.

10) - Reactor Low-Low Water Level, R 78,2" above the top of the Main Steam Line Isolation Valve active fuel as indicated Closure, under normal operating conditions. -

11) Reactor Low-Low Water Level, R 7'2" above the top of the Core Spray Initiation _

active fuel l

12) Reactor Low-Low Water Level, t 7'2" above the top of the Isolation Condenser Initiation- active fuel with time delay 8 3 seconds.

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. 13) Turbine Trip Scram .

10 percent' turbine stop valve (s) closure from full open.

- 14) . Generator Load- Rejection Scram Initiation upon loss of oiI pressure from turbine acceleration relay.

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15) Loss of Power

, n. -4.16'KV Emergency Bus 0 volts with 3 seconds +-

-Undervoltage (Loss of Voltage) 0.S seconds time delay.

b .- 4.16 KV-Emergency Bus 3671 11% (36. 7) volts Undervoltage (Degraded Voltage) 10110%(1.0)second time delay, i:

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2.3-3a BASES: Safety Iimits have been estabIIshed in Spect fIcatIons 2.1 and 2.2 to protect the integrity of the fuel cladding and reactor coolant system barriers.

_]' Automatic protective devices have been provided in the pl ant design to .take corrective action to prevent the safety limits.from being exceeded in normal operation or operational transients caused by reasonable expected single -

operator ' error or' equipment mal f unction. This Specification establishes the tr1p settings for these automatic protection devices.

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The Average Power Range Monitor, APRM"'," trip setting has been estabilshed to assuro never. reaching the fuel cladding integrity safety limit. The APRM system

. responds to changes in neutron flux. However, near rated thermal power the APRM is calibrated, using a plant heat balance, so that the neutron flux that is sensed is read out as percent of rated thermal power. For slow maneuvers, those where core thermal power, surf ace heat flux, ' and the power transferred to the '

-water f ollow the ' neutron flux, the APRM will read reactor thermal power. For -

tast transients, the neutron flux wll! ' lead the power transferred from the

' cl add ing to the water due to the ef f ect of the f uel time constant. Therefore when the neutron flux increases to the scram setting, .the percent increase in heat flux and power transferred to the water will be l ess than the percent increase in-neutron flux.

'Tho ' APRM trip setting will be varied autcmatically with recirculation flow with

- the trip setting at rated flow 61.0 x 10 6 lb/hr or greater being 115.7% of rated neutron flux. Based on a certiplete

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2.3-7 The . low water level trip setting of 11 5" 8

ebove the top of the active f uel has f T bcon' established to assure that the reactor is not operated at a water level below that for which the fuel cladding Integrity safety limit is applicable.

With the scram . set. at this point, the generation of steam, and thus the loss of *

' inventory, is stopped. .For example, for,a loss of feedwater flow a reactor scram at-the'value indicated and Isolation valve closure at the low-low water level: set point results In more then 4 feet of water remaining above the core ofter Isolation. - (11 ) .

During periods when the reactor is shut down, decay heat is present and adequate water level must be maintained to provide core cooling. Thus, the low-low level trip point of 7*2" above .the core is provided to actuate the core spray system to pravide cooling water should the Ievel drop to this point. In addition, the

. normal reactor feedwator system and control rod drive hydraulic system provide -

_ protection f or the water . level safety limit both when the reactor is operating at power or in the shutdown condition.

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~The - turb ine stop valve ( s) scram anticipates the pressure, neutron flux, heat Jflux increase caused by the rapid closure of the turbine stop valve (s) and f ail ure of the turbine bypass system. With a scram setting of 10% of valve

. closure from full open and _ wlth,a failure of the turbine bypass system at 1930 MWt,_the peak pressure will remain well below the first safety valve setting and no thermal limits are approached (7,10).

'The generator: load rejection scram is provided to ailticipate the rapid increase

-l_n p'ressure and neutron flux resulting from fast closure of the turbine control valves to a' load rejection and f ailure of the turbine bypass system. This scram c ,e

is _' initiated by the loss of turbine acceleration relay oil pressure. The timing for -this ~ scram is almost identical to the turbine trip and the resultant peak pressure -and MCHFR are essentially the same.

L The undervoltage protection system is a 2 out of 3 coincident logic relay system l- designed to shif t emergency buses C and D to on site power. should normal power bb ' lost or degraded to an unacceptable level . The trip points and time delay settings have been selected to assure an adequate power source to emergency sa f eg uards systems in the event of a total loss of normal power or degraded conditions which would adversely af fect the functioning of engineered safety features connected to the plant emergency power distribution system.

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Rwferences

('1 ) ; FOSAR. Volume 1, Section Vll-4.2.4 ,

(2) FDSAR, Volume 1, Section - l-5.6 (3)- ' L icensing Appi IcatIon Amendment 28, Item Il.1.A-12 -

(4 )' Licensing Application Amendment 32, Que'stion 13

' (5) Letters, Peter A. Morris, Of rector, Division of Reactor Licensing, USAEC to John E. Logan, Vice President, Jersey Central Power & Light Company, dated November 22,

'1967 and January 9, 1968.

- (6) . Licensing Application Amendment 11, Question V-9.

(7) ' License- Application Amendment 76, Supplement No.1

. (8) License Application Amendment 65, Section' B.XI .

' (9) LIconse Application Amendment!69, Section iIl-D-5 (10) License Application Amendment 65, Section B.IV.-

(11) License Application Amendment 65, Section B.lX. '

- (12) License Application Amendment 76,' Supplement No. 3, Section 2.0.

. (13) License Application Amendment 76, Supplement No. 4.

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3.1-11b' -

TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIRBENTS (CONTD) .

Min.No.of -

Min. No. of ' Operable .

Reactor Modes Operable or Instrinsent

, _ ,in which Function Operating Channels Per

Must be Operable (Tripped) Trip Operable

~ Function Trip Setting Shutdown Refuel Startup Run Systems Trip Systents Action Ra=Jired*

N. Loss of Power -

m. 4.16KV Emergency **

X (ee) X(ee) X (ee) X (ee) ~2 1 .

Bus Undervoltage -

(Loss of Voltage) ./. ,

b. 4.16 KV' Emergency ** '

X (ee) X(ee) X (ee) X (e9 2 3' See Note dd

Bus undervoltage "

(Degraded Voltage) 4 e

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3.1-13 TABLE 3.1.1 (CON'D) i.

The interlock is.not requi$ed during the start-up test program and demonstration of plant electrical?

- output but shall be provided following these actions. -

j. Not required below 40% of turbine rated steam flow.

k. - All four (4) drywell pressure instrument channels may be. made inoperable during,the integrated primary containment leakage' rate test (See Specification 4.5), provided that primary containment integrity is not required and that no work is performed on the reactor or its connected systems which could result in lowering the reactor water level to less than 4'8" above the top of the active fuel.

1. Bypassed in IRM Ranges 8, 9, S'10. .

j m. There ~is one time delay relay associated with each of two pumps.

n. One time. delay replay per pump must be operable. . - "-

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'o. There are two time delhy relays associated with each of two s. One timer pei pump is for se starting (SK1A, SK2A) and one timer per pump is for trippin epumpcircuitbreaker(SK7A,~SKkuence A) .

l p. Two time delay relays per pump must be operable.

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q. Manual initiation.of affected component can be accomplished after the automatic load

. sequencing is completed. ~ .

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r. Time delay starts after closing of containment spray pump circuit breaker.

s. These functions not required to be operable with the reactor temperature less than 212*F i and the vessel head removed or vented.

1 f t. These functions may be inoperable or bypassed when corresponding portions in the same core spray system logic train are inoperable per Specification 3.4.A.

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These functions not required to be operable when primary containment integrity is not required to be maintained.

t Amendmerft No. 14, 44, 60 .

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' TABLE 3.1.1 (Cont'd) l V

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Those functions not required to be operable.when the ADS.is not required to be" operable. ,

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I These functions must be operable only when irradiated fuel.is in the fuel pool..or reactor vessel and. secondary containment integrity is required per specification 3.5.8.

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I' The number of operable channels may be reduced to 2 per Specification 3.9.E and F.

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z. The bypass function to permit scram reset'in the shutdown or refuel mode with control rod' block must be operable in this mode.

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Pump circuit breakers will be tripped in 10 seconds 15% during a LOCA by relays SK7A and SK8A.

bb. Pump circuit breakers will trip instantaneously during a LOCA.

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Only applicable during STARTUP Mode while operating in IRM Range 10. .

dd. WiYh the number of operable channels one less, than the Min. No. of Operable

-Instrument Channels per Operable Trip. Systems;ssperation may proceed:until s performance of the next required Channel Functional' Test provided th'.e'in- .

operable channel is placed in the tripped condition within I hour.

ee. This function is not required to be operable when the associated safety bus is not required to be energized or fully operable as per applicable sections of these

tecimical specifications. ,

Amendment No.,4(,Ji(ji( 7] l

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f,p AUXILI ArsY ELECTkCAL POhER_ , . ,

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/.:elicability: Applios to 'th's o;ierafing status, of tt3e eux t lary el ectrical power supply. -

Objective: To assure the operabfilty of the auxillary

. a s w:7.-Ical power supply.*

-!-ac t f ication: A. Tr.e reactor shall 'not bu mode critical unless -

all of the following requirements are satisfied

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. 1. The following buses or panels energized.

a. 4163 velt buses 1C and ID in the turbine building switchgear room.

'b. 460 volt buses 1 A2, 182,' 1 A21, 1821 yItal '

MOC 1A2 and 182 in the reactor but Idtng switchgear roem: 1A3 and 183 at the intake structure; 1 A21 A,1821 A, IA218, and 1821B and v ital MCC 1 A32 on 23 '6" elevation in the reactor building; 1A24* and 1324 at the stack.

c. 208/120 volt . ponel s 3, it , 4A, 48, 4C and VAC?-1 in .the rucctor bu'llding switchgear rooni.

d.120 voit prote'ction panel 1 and 2 I n the cab l e ,

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e.125 volt D0 distribution centers C and B, and p ar.el, Q, Panel 00-F, Isolation valve motor control center 00-1 and 125V 00 motor esr. trol center D0-2. . -

f'. 24 icit D.C. power panel s A and B in the cahie rocm. .

2. Cna 233 KY IIne is fulIy operational and *

., . switch geir ~end both startup transformers are er.argized to carry power to t'he station 4160 voli AC buses and carry power to or

• m any fecm the pl ant. '

3. An additicnal source of power consistYng of *. ' ' ' '

cna of the following is in service connected to f eed the appropriate plant 416Q V bus cr buses: -

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a. . A second 230 KY line f ully operational .

1: . One 34.5 KV line ful ly operational .

4. The, statica batterles B and C are available f or n:.=al service and a battery charger is in service for~each battery.

!. 5.:s tie breaker. E3 s.nd EC are in the cpan p:sittens.

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. In nt rasmcut / Cha nniYI - ~ t:In?ck ' -(*: I ilar.s t.c - rent f(smarke'(ApptIen'to' rent A t!.e i Ila ra t I ung * .*

In Manua l' Sc ram .11uttons .. NA NAl- 1/3 mo. .,

20. High Temperature Maln NA 10ach itefuel-- Each refuet- Using heat source' box StcamLine Tunnel Ing outage . Ing outage 2L. -SRM * *

• UsLng built-in calibration cegulpment

[ 2'2. 'Isointion Condenner Illnh _ NA I/3 mo. 1/3 sq. By appilcation of' test prendure Flow AP (Steam aiwi linter) ,.

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23. Turbine Trip Scram NA Every',

3 sonntius Gencentor Load lle,]ection NA Kvery Kvery

24.

Scram 3 months 3 monthn~ -

25. Recirculation Loop ilow' NA Each Refuel- NN

• By application.of test pressure Ing Outage .

2 6. - Low Reactor Frensure NA Every Every By appitcation of test pressure -

Core Spray Valve 3 months 3 monthn , E+

Perminsive <

27. Scram Discharge Volume .

(Rod Block) a) Water level high NA Each Refu'el- Every 3 'By varying level in switch column '

ing Outage months .-

b) Scram trLp bypass ~

NA NA Each refucl- -

ling ' outage

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28. Loss of Power c

Dolly' a) 4.16 KV Emergene,- 1/18 mos. 1/mo. .

Bus Undervoltage ,

(Loss of iroltage) .

en b) 4.16 KV Emergency Daily 1/18 mos. 1/mo. >

Bus Undervoltage L (Degraded Voltage) 1 m

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.e,.r check 1/s and test 1/uk until no longer .-

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3 asis: The biweekly tests of the diesel ganarators are primarily to check for failures and deterioration.in the systaa since last use. The manufacturer has recom= ended the two week test interval, based on experience with many of their engines. One factor in -datarmining this test interval (besides ch'ecking whether or not the engina starts and runs) is that the lubricating oil should be circulated .. .

through the angina approxigately every two weeks. The diasals should be loaded to at leas't 20% of rated power until. angina and generator tamparaturas have" stabilized (about one hour). The

-minimum 20% load will prevent soot formation in the cylinders and injection nozzlas. Operation up to an equilibrium tamparatura ensures that there is no over-heat problem. The tests also pro-vida an engina and generator operating' history to be compared with subsequent angine generator test data,to identify and correct any .

s mechanical or electrical deficiency before it can result in a system failure.

The test during refueling outages is more comprehensive, including procedures that are nost effectively conducted at that time. These include auto:atic actuation and functional capability tests, to

. verify that the generators can start and assuma load in less than 20 seconds and testing of the diesel generator load sequence timers

,which provida protection from a possible diasal genera. tor overload during LOCA conditions. Thorough inspections will detect any signs of wear long before failure.

The canufacturer's. instructions for battery care and maintenance with regard to the floating charge, the equalizing charge, and the addition of ater vill be 'followed. In addition, written records will ba =airtained of the battery performance. Station batteries will deteriorate with time, but precipitous failure is unlikely. The station surveillance procedures follow the recommended maintenanca and testing practices of IEEE STD. 450 which have deoonscrated, through experience, the ability to' provide positive indications of l cell deterioration tendencies long before such tende6cies causa cell irregularity or improper cell performance. , , , ,

l l-Amendment No. 60 e

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