ML19331C205

Proposed Tech Specs 2.3,3.1,3.7,4.1,4.7 & Table 3.1.1 Re Undervoltage Protection Sys

ML19331C205
Person / Time
Site:	Oyster Creek
Issue date:	08/11/1980
From:	JERSEY CENTRAL POWER & LIGHT CO.
To:
Shared Package
ML19331C204	List: ML19331C203 ML19331C205
References
TASK-08-01.A, TASK-RR NUDOCS 8008140337
Download: ML19331C205 (17)

Text

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JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Applicant hereby requests the Commission to' change Appendix A to the' License as follows:

1. Sections to be changed:

Sections 2.3, 3.1 (Tab l e 3.1.1 ) , 3.7, 4.1, ar.J 4.7.

2. Extent of changes:

.New specifications for undervoltage protection systems and station batteries.

3. Changes requested:

Add / Replace Page With Attached Page 2.3-3 2.3-3 2.3-Ja -

2.3-7 2.3-7 2.3-8 -

2.3-8 3.1 -11 b -

3.1-12a 3.1-12a 3.1 -12 b -

3.7-1 3.7-1 3.7-la 3.7-2 3.7-2 3.7-3 3.7-3 '

4.1-6a 4.1-6a 4.7-1

. 4.7-1 4.7-la 4.7-2 4.7-2

4. Discussion:

The changes requested with regard to undervoltage are proposed in order to inco; srate Technical Specifications pertaining to undervoltage protection systems w..ich were installed during the 1980 ref ueling outage. These protection systems are as described in our lei Mrs of September 25, 1979, and November 1, 1979, which were in response to your letters of June 2,1977 and August 11, 1979.

A description of a modi fication to the 125V DC distribution system at Oyster Creek Nuclear Generating Station was submitted by letter dated April 14, 1978 from Ivan R. Fintrock, Jr. to the Director of Nuclear Reactor Regulation.

The safety related lodds which were previously powered f rom 125V DC l d i str ib ution center "A", are now powered from 125V DC distribution center "C".

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The Safety related loads that were previously powered from panet E are now powered f rom P anel 00-F . An additional motor control center, OC-2, has been

. added. The enanges to Section 3.7 are editorial changes to reflect these new power supplies for saf ety related loads.

The change to Section 4.7 updates the battery survelllance requirements to reflect today's standards. This specification also requires that the battery be tested only when the reactor is shutdown. At all times when the plant is operating, both batteries will be at' full capacity; thereby, increasing the safety margin due to increased availability of the station battery.

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FUNCTION i LIMITING SAFETY SYSTEM SETTINGS 7)_ Low Pressure Main Steam Line, h825 psig MSIV Closure

8) Main Steam Line Isolation Valve A 10% Valve Closure from full Closure, Scram open

9) Reactor. Low Water Level, Scram E 11',5" above the top of the active fuel as indicated under normal operating conditions.

10) Reactor low-Low Water Level, R 7',2" above the' top c" the Main Steam Line Isolation "alve active fuel as indicated Closure. under normal operating conditions.

11) Reactor Low-Low Water Level, E 7'2" above the top of the Core Spray Initiation active fuel'

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

13) Turbine Trip Scram .

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

14) Generator bad Rejection Scram Initiation upon loss of oil pressure from turbine acceleration relay.

15) Loss of Power *

a. 4.16 KV Emergency Bus 0 volts with 3 seconds +

Undervoltage (Loss of Voltage) 0.5 seconds time delay.

b. 4.16 KV Emergency Bus 3671 + lt. (36. 7) volts Undervoltage (Degraded Voltage) 10 1 1 C(.1) second time delay.

2.3-3

2.3-3a BASES: Safety limits have been established in Specifications 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 plant design to take corrective action to prevent the safety limits from being exceeded in normal operation or operational. transients caused by reasonable expected single operatcr . error or equipment mal f unction. This Specification establishes the trip settings for. these automatic protection devices.

The Average Power Range Monitor, APRM , trip setting has been established to assure 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 !. eat flux, and the power transf erred to the water f ollow the neutron- fl ux, the APRM will read reactor thermal power. For fast transients, the neutron flux will lead the power transferred from the cl adding 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 less than the percent increase in neutron flux.

The APRM trip setting will be varied autonatically with recirculation flow with the trip setting at rated flow 61.0 x 10 lb/hr or greater being 115.7:f, of rated neutron flux. Based on a complete a

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2.3-7 The low water level trip setting of 11*5" above the top of the active f uel has been 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 than 4 f eet of water remaining above the core after 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 l evel trip point of 7'2" above the core is provided to actuate the core spray system to prov ide cooling water should the level drop to this point. In addition, the normal reactor feedwater system and control rod drive hydraulic system provide protection for the water level safety limit both when the reactor is operating at power or in the shutdown condition.

The turb ine stop valve ( s) scram anticipates the pressure, neutron flux, heat flux increase caused by the rapid closure of the turbine stop valve (s) and failure of 1he turbine bypass system. With a scram setting of 10f, of valve closure from full open and with 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 apprcached (7,10).

The generator load rejection scram is provided to anticipate the rapid increase in pressure 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 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 essential ly the same.

The undervoltage protection system is a 2 out of 3 coincident logic relay system designed to shif t emergency buses C and 0 to on site power shoul d normal power be' lost or degraded to an unacceptable level . The trip points and time delay settings have been selectad to assure an adequate power source to emergency safeguards systems in the event of a total loss of normal power or degraded conditions which would adversely affect the functioning of engineered safety f eatures connected to the plant emergency power distribution system.

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2.3-8 Ref e. ences (1 ) FDSAR. Vol ume I, Section Vll-4.2.4 L (2)- 'FDSAR, Volume I, Section I-5.6 (3) Licensing Application Amendment 28, item til.A-12 (4 ) Licensing Application Amendment 32, Questler. 13 (5) ' Letters, Peter A. Morris, Director, Division of Reactor '

t.icensing, 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) License Application Amendment 69, Section 111-D-5 (10) License Application Amendment 65, Section B.IV.

(11) License Application Amencment 65, Section B.lX.

(12) License Application Amenoment 76, Supplement No. 3, Section 2.0.

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

3.1-llb t

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TABI.E 3.1.1 PROTECTIVE INSTRUENTATION REQUIREMENTS (CONTD)

' Min.No.of Min. No. of . Operable Reactor Modes Operable or Instrument in which Function Operating Channels Per l- Must be Operable (Tripped) Trip Operahle l' Systems Action Required

• Function Trip Setting Shutdown Refuel Startup Run- Trip Systems N. loss of Power

• • X (aa) X(aa) X (aa) X (aa)

a. 4.16KV Emergency 2 1 Bus Undervoltage  :

(toss of Voltage)

! b. 4.16 KV Emergency ** X (aa) X(aa) X (aa) X (aa) 2 -3 See Note ~

Bus undervoltage (Degraded Voltage) t l

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3.1-12a TABLE 3.1.1 -(CON'D)

1. The interlock is not required 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 integcated 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, G 10.

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

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

o. There are two time delay relays associated with each of two pumps.

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

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

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 and the vessel head removed or vented,

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.

l-3.1-12b

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

l l v. These functions not required to be operable when the ADS is not required to be operable.

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! w. 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 l 3.5.B.

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

z. With the number of operable channels one less than the Min. No. of Operable Instrument Channels per Operable Trip Systems, operation may proceed until performance of the next required Channel Functional Test provided the in-operable channel is placed in the tripped condition within I hour.

aa. This function is not required to be operable when th'e associated safety bus is not required to be energized or fully operable as per applicable sections of these technical specifications.

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3.7-1 3.7 AUXILIARY ELECTRICAL POWER A oo l _i, cab i l i ty: Applies. to the operating status of the auxilary elettrical power supply.

Ob jective: To assure the operability of the auxiliary electrical power supply.

Specification: A. The reactor shall not be made critical unless all of the following requirements are satisfied:

1. The following buses or panels energized. .

a. 4160 volt buses 1C and 10 in the turbine building switchgear room.

b. 460 volt buses 1 A2,182, I A21,1821 vital MCC 1A2 and 182 in the reactor building switchgear room: 1A3 and 183 at the intake structure; 1 A21 A,1B21 A,1 A218, and 1821B and vital NCC 1AB2 on 23*6" elevation in the reactor building; 1A24 and 1824 at the stack.

c. 208/120 volt panel s 3, 4, 4A, 48, 4C and V ACP-1 in the reactor building swi chgear room.

d.120 volt protection panel 1 and 2 [en the cable room.

e. 125 voit DC distribution centers C and B, and panel D, Panel DC-F, isolation valve motor control center DC-1 and 125V DC motor control center DC-2.

f. 24 volt D.C. power panels A and B in the cable room.

2. One 230 KV line is fully operational and switch gear and both startup transformers are energized to carry power to the station 4160 volt AC buses and carry power to or away from the plant.

3. An additional source of power consisting of one of the following is in service connected to feed the appropriate plant 4160 V bus or buses:

a. A second 230 KV line fully operational.

b. One 34.5 KV line fully operational.

4. The station batteries B and C are available for normal service and a battery charger is in service for each battery.

5. Bus tie breaker EG or EC is in the open position.

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'N- 8. The reactor shalI be placed in the cold shut-.

oown position if'the availability of power talIs below that required by SpectfIcation A above, except that the reactor may remain in operation'for a period not to exceed 7 days in any 30 day period if a startup transformer is out of service.

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3.7-2 l

?.one of the tr.gineered safet:. feature equipment isJ in the remaining transformer may be out of service.

. C. Standby Diesel Generators

1. The reactor shall not be made critical unless both diesel generators are operable and capable of feeding their designated 4160 volt buses.

2. If one diesel generator becomes inoperable during power operation, repairs shall be initiated immediately and the other diesel shall be operated at least one hour every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at greater than 20'. rated power until repairs are co=pleted. The reactor may remain in operation for a period not to exceed 7 days in any 30-day period if a diesel generator is out of service. During the repair period none of the engineered safety features normally fed by the operational di,sel generator may be out of service or the reacter shall be placed in the cold shutdown condition.

3. If both diesel generators become inoperable during power operation, the reactor shall be placed in the cold shutdown condition.

4. For the diesel generators to be considered operable there shall be a minimum of 14,500 gallons of diesel fuel in the standby diesel generator fuel tank.

Bases: The general objective is to assure an adequate supply of power with at least one active and one standby source or' power e.vailable for operation of equipment required for a safe plant shutdown, to maintain the plant in a safe shutdown condition and to operate the required engineered safety feature equipment following an accident.

AC power for shutdown and operation of engineered safety feature equipment can be provided by any of four active (two 230 KV and two 34.5 KV . lines) and either of two standby (two diesel generators) sources of power. Normally all six sources are available.

However, to provide for maintenance and repair of equipment and still have redundancy of power sources the requirement of one active and one standby source of power was established. The plant's main generator is not given credit as a source since it is not available during shutdown. The plant 125V DC power is normally supplied by two batteries, each with two associated full capacity chargers. One charger on each battery is in service at all times with the second charger available in the event of charg>. r failure. These chargers are active sources and supply the normal 125V DC requirements with the batteries as standby sources. (1)

In applying the minimum requirement of one active and one standby scurce of AC power, since both 230 KV lines are on the same set of towers, either one or both 230 KV 13nes are considered as a single active source.

3.7-3 The probability. analysis in Appendix "L" of the FDSAR was based on one diesel and shows that even with.only one diesel the probability of requiring engineered safety features at the same time as the second diesel fails is quite small.

j This anclysis used information on peaking diesels when symchronization was required which is not the case for Oyster Creek. Also the daily test of the second diesel when one is temporarily out of service tends to improve the reliability as does the fact that synchroni:ation is not required.

As_indi.cated in Amendment 18 to the Licensing Application, there are numerous sources of diesel fuel which can _be

obtained within 6 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the heating boiler fuel in a 75,000 gallon tank on the site could also be used. Since the requirements for operation of the required engineered safety features after an accident or for safe shutdown can be supplied by one diesel generator the specification requirement for 14,500 gallons of diesel fuel can operate one diesel at a load of 2640 KW far

• 3 days. As indicated in Amendment 32 of the Licensing l Application, the load requirement for the loss of offsite

! power would require,11,750 gallons for a three day supply.

For the case of loss of offsite power plus loss-of-coolant plus bus failure 11,300 gallons would be' required for a three day supply. In the case of loss of offsite power plus loss-of-coolant with both diesel generators starting the load . requirements (all equipment operating) shown there would not be three days' supply. However, not all of this load is required for three days and, after evaluation of the conditions, loads not required on the diesel will be curt. ailed. It is reasonable to expect that within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> conditions can be evaluated and the following loads curtailed:

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1. One Reactor Building Closed Cooling Water Pump.

2. One Core Spray Fump. _'

3. One Core Spray Booster Pump.

4. One Control Rod Drive Pump.

5. One Service Water Pump.

6. One Containment Spray Pump

7. One Emergency Service Water Pump.

With these pieces of equipment taken off at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the incident it would require a total consumption of 14,230 gallons for a three day supply.

References:

(1) Letter, Ivan R. Finfrock, Jr. to the Director of Nuclear Reactor Regulations dated April 14, 1978.

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Instrument Channel Check Calibrate Test Remarks (Applies to Test G Calibration)

19. Manual Scram Buttons NA NA 1/3 Mo

20. liigh Temperature Main NA Each Refuel- Each refuel- IIsing heat source box Steamline Tunnel ing outage ing outage

21. SRM * *

• lising built-in calibration equipment

22. Isolation Condenser liigh NA 1/3 mo 1/3 mo By application of test pressure Flow P (Steam and Water)

23. Turbine Trip Scram NA Every 3 months

24. Generator Load Rejection NA Every Every ,

Scram 3 months '3 months ,

25. Recirculation Loop Flow NA Each Refuel- NA By application of test pressure ing Outage

26. Low Reactor Pressure NA Every Every By application of test pressure Core Spray Valve 3 months 3 months Permissive

27. Loss 'of Power

a. 4.16 KV Emergency Daily 1/18 mos. 1/mo.

Bus Undervoltage (loss of voltage)

b. 4.16 KV Emergency Daily 1/18 mos. 1/mo.

, Bus Undervoltage (Degraded Voltage) 4 1

• Calibrate prior to startup and normal shutdown and thereafter check 1/s and test '.f wk until no longer required. .-

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4.7-1 4.7 AUXILI ARY ELE";TRICAL PO'nER A nlicability: Applies to surveil lance - rcquirs war * = of the auxiliary electrical supply.

Ob ject ive: To verify the availacility of. tne auxiliary electrical supply.

S peci f icat ion: -A. Diesel Generator

1. Each diesel generator shall be started and loaded to not less than 20% rated power every two weeks.

2. Each diesel generator shall be automatically activated (by simus ating a loss of ~ of fsite power in conjunction with a saf ety injection actuation _ test signal) and functionally tested during each refueling outage by:

a. Verifying de-energization of. the emergency busses and load shedding from the emergency busses.

b. Verif ying the diesel starts from subient conditions on the auto-start signal, energizes .the. energency busses with permanently connected loads, energizes the auto-connected emergency loads through the load sequence timers listed in Table 3.1.1 and operates fort 5 minutes while its generator is loaded with the emergency loads.

c. Verifying that on diesel generator trip, the loads are shed from the emergency busses and the diesel restarts on the auto-start signal, the energency busses are energizec with permanently connected l oacs ,

the auto-connected emergency loads Jee energized through the load se4cences and the diesel operates for E 5 minutes while its generator -

is loaded with the energency loads.

3. Each diesel generator shall be given a thorough inspection at least annually.

4 The diesel generators8 fuel supply shall be checked following the above tests.

5. The diesel generators

• starting batteries shall be tested and monitored the sane as the station batteries, Speci f ication 4.7.S.

4.7-1s B. Station Batteries

1. Weekly surveillance will be performed to verify the following:

a. The active metallic surface of the plates shall be fully covered with electrolyte in all batteries,

b. The designated pilot cell voltage is greater than or equal to 2.0 volts and

c. The overall battery voltage is greater thac or equal to 120 volts (Diesel battery; 112 volts) .

d. Pilot cell speci fic gravity, corrected to 77* F, shall be recordeo for surveillance review.

2. Quarterly Surveillance will be performed to verify the following:

a. The active metallic surf ace of the plates shall be fully covered with electrolyte in all batteries,.

b. The voltage of each connected cell is greater

. than or equal to 2.0 volts under float charge and

c. The overall battery voltage is greater than or equal to 120 volts (Diesel battery; 112 volts)

d. The speci fic grav ity, corrected to 77

• F, for each cell and the electrolyte temperature of every fifth cell (Diesel; every fourth cell) shall be recorded for surveillance review.

3. At least once per 18 months during shutdown, the following tests will be performed to verify battery capacity. .

a. Battery capacity shall be demonstrated to be at lest 80) of the _manuf acturers' rating when subjected to a battery capacity discharge test.

b. Battery low voltage annunicators are verified to pick up at 115 volts t 1 volt and to reset at 125 volts 11 voit (Diesel; 120 volts 1 1 volt).

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- u . The biwcekly test.' of the diesel generators are pri=arily to check for failures and deterioratien in the syste= since last use. The =anufacturer has rece== ended the two week test interval, based on experience with =any of their engines.

One factor in deter =ining this test interval (besides' checking whether or not the engine starts and runs) is that the lubricating oil should be circulated through the engine approximately every two weeks. The diesels should be loaded to .t 1 cast 20'. of rated power until engine and generator te=peratures have stabili:ed (about one hour).

The mini === 20'. Icad will prevent soot for=ation in the cylinders and injection no::les. Operation up to an equilibriu= te=perature ensures that_there is no over-heat proble=. The tests also provide an engine and generator operating history to be co= pared with subsequent engine-generator test data to identify and correct any mechanical or electrical deficiency before it can result in a system failure.

The test during refueling outages is more co=prehensive, including procedures that are most effectively conducted at that time. These include auto =atic actuation and functional capability tests, to verify that the generators can start and assu=e". load in less than 20 seconds and testing of the diesel generator load sequence ti=ers which provide protection fro = a possible diesel generator overload during LOCA conditions. The annual, thorough inspection will detect any signs of wear long before failure.

The manufacturers instructions for battery care and =ainten-ance with regard to the floating charge, the equalizing charge, and the aridition of water will be followed. In addition, written records will be maintained of the battery perfor=ance. Station batteries will deteriorate with ti=e, but pre. . itous failure is unlikely. The station surveillance procedures follow the reco== ended =aintenance and testing practices of IEEE STD. 450 which have de=enstrated, thorough experience, the ability to provide positive indications of .

cell deterioration tendencies long before such tendencies cause cell irregularity or i= proper cell perfor=ance.

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