ML19331B947
| ML19331B947 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/11/1980 |
| From: | JERSEY CENTRAL POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19331B942 | List: |
| References | |
| NUDOCS 8008130593 | |
| Download: ML19331B947 (1) | |
Text
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A Jersey Central Power & Light Company
.1 ( (f ()j Madison Avenue at Punch Bow! Road Momstown, New Jersey 07960 (201)455-8200 August 11, 1980 The Honorable Henry Von Spreckelsen Mayor of Lacey Township .
P. O. Box 475 Forked River, New Jersey 08731
Dear Mayor Von Spreckel sen:
Enclosed is one copy of Technical Specification Change Rcquest No. 88 for the Oyster Creek Nuclear Generating Station Operating License.
T5is cocument was fiIed wIth the United States NucI ear Regul atory Commission on August 11, 1980.
Very truly yours, J $1 . .)
Ivan R. Fint ck Jr.
Vice President la Enclosure 6
e 8 0 0 813 0 6 3 central Power & Lignt companyis a Member of the ceneral Public Util!!ies Ststem
JERSEY CENTRAL POWER & LIGHT COMPANY
. OYSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING wlCENSE NO. DPR-16 DOCKET-NO. 50-219 /
l Applicant hereby requests the Commission +o change Appendix A to the License.as follows:
'1. Sections to be changed:
Sections 2.3, 3.1 (T ab l e 3.1.1 ) , 3.7, 4.1, and 4.7.
- 2. . Extent of changes:
New specifications for undervoltage protection systems and station batteries. ,
- 3. Changes requested:
l l Add AReplace Page With Attached Page l 2.3-3 2.3-3
! 2.3-3a -
2.3-7 2.3-7 2.3-8 ,
2.3-8 3.1 -11 b -
! 3.1-12a 3.1-12a
! 3.1-12b -
l 3.7-1 3.7-1 l .
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: '
f The changes requested with regard to unde-woltage are proposed in order to incoporate Tcchnical Specifications pertaining to undervoltage protecticn systems which were . installed 'during the 1980 ref ueling outage. These protection systems are as described in our. letters of September 25, 1979, and November 1,
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1979, which were-in response to your letters of June 2,'1977 and August 11, 1979.
A. description of a modification 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.
d The safety.related loads which were previously powered from 125V DC d istr ibution center "A", are now powered from 125V- DC distribution center "C".
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3 t
, 4 - _ .
- F The Safety.related loads that were previously powered from panel E are now pcwered from Panel 00-F. An add itional motor control center, 00-2, has been
. cdded. The changes to Section 3.7 are editcrict changes to reflect these new
- power supplies .for saf ety related loads.
The change to Section 4.7 updates the battery surveillance requirements to reflect today's standards. This specification also requires that the battery bo tested only when the reactor is shutdown. At ali times when the plant i s op'erating, both batteries will be at full capacity; thereby, increasing the ssfety margin due to increased availability of the station battery.
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e n
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., n
s FUNCTION LIFi1 TING SAFETY SYSTEM SETTINGS
- 7) Low Pressure Main Steam Line, h8?9 psig
.MSIV Closure
- 8) Main Steam Line Isolation Valve A 10% 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 i under normal operating conditions.
- 10) -Reactor Low-Low Water Level, EE 7',2" above the top of the hbin Steam Line Isolation Valve active fuel as indicated Closure. under normal operating conditions.
- 11) Reactor Low-Low Water Level, it 7'2" above the top of the Core Spray Initiation active fuel
- 12) Reactor Low-Low Water Level, it 7'2" above the top of the Isolation Condenser Initiation active fuel with time delay 8 3 seconds.
13). Turbine Trip Scram ,
10 percent turbine stop valve (s) closure from full open.
- 14) Generator Load Rejection Scram Initiation upon loss of oil pressure from turbine acceleration relay.
- 15) Loss of Power
- n. 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 + 1% (36. 7) volts Undervoltage (Degraded Voltage) 10 + 1%-(.1) second time delay.
2.3-3
2.3-35 BASES: Safety limits have been estab lished in Specifications 2.1 and 2.2 to protect the integrity of the f uel cladding and reactor coolant system barriers.
Automatic protective devices have been prov ided 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 operator error or equipment mal f unction. This Specification establisnes the trip settings for these automatic protection devices.
'The Average Power Range Monitor, AP'RM , trip setting has been established to casure never reaching the fuel cladding integrity saf ety limit. The APRM system rcsponds 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 s:nsed is read out as percent of rated thermal power. For slow maneuvers, those tthere core thermal power, surf ace heat flux, and the power transf erred to the uster follow the neutron flux, the APRM will read reactor thermal power. For f ast transients, the neutron flux will lead the power tri:ns f erred from the cladding to the water due to the ef fect of the fuel time constant. Therefore ,
then the neutron flux increases to the scram sotting, the percent increase in hsat flux and power transferred to the water will be less than the percent increase in neutron flux.
The APRM teip setting wiII be variad automatleally wIth recirculation flow wIth the trip setting at rated flow 61.0 x 10' lb/hr or greater being 115.7% of rsted neutron flux. Based on a complete l
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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 b3 low that for which the fuel cladding integrity saf ety 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
. lsvel set point results in more than 4 feet of water remaining above the core efter isolation. (11).
During periods when the reactor is shut down, decay heat is present and adequate ucter 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 norrial reactor feedwater system and control rod drive hydraul ic system provide pr_otection f or the water level safety limit both when the reactor is operating at power or in the shutdown condition.
The turbine stop valve ( s) scram anticipates the pressure, neutron flux, heat flux increase caused by the rapid closure of the turbine stop valve (s) and falIure of the turbine bypass system. With a seram setting of 107, 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 approached (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 aimost identical to the turbine trip and the resultant peak pressure and MCHFR are essentially the same.
The undervoltage protection system is a 2 out of 3 coincident logic relay system dcsigned to shif t emergency buses C and D to on site power shoul d normal power be ' lost or degraded to an unacceptable level. The trip points and time delay s3ttings have been selected 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 features connected to the plant emergency power distribution system.
2.3-8 Rnterences s1) FOSAR. Vol ume I, Section V:1-4.2.4 (2) FDSAR, . Volt:me I, Section I-5.6 (3) Licensing Application Amendment 28, I ten l i t . A-12 (4 ) Licensing Application Amendment 32, Question 13 (3) Letters, Peter A. Morris, Director, 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) License Application Amendment 69, Section lil-D-5 (10) License Application Amendment 65, Section 8.I V.
(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 REQt;IREMENTS (CONTD)
Min.No.o f Min. No. of Operable Reactor Modes Operable or Instrument
. .in which Function Operating Channels Per Must be Operable (Tripped) Trip Operable Function Trip Setting Shutdown Refuel Startup Run. Systems Trip Systems Action Required
- N. loss of Power
- a. 4.16KV Emergency ** X (aa) X(aa) X (aa) X (sa) 2 1 Bus Undervoltage (Loss of Voltage)
- b. 4.16 KV Emergency ** X (aa) X(aa) X (aa) X (aa) 2 3 See Note Z Bus undervoltage (Degraded Voltage)
3.1-12a TABLE 3.1.1 (CON'D)
- 1. The interlock is not required during the start-up test program Land demonstration of plant electrical-output but shall be provided following-these ections.
- j. Not. required below 40% or turbine rated steam flow.
- k. - All four (4) drywell pressure instrument channels may be made inoperable during the integrated primary containment Icakage rate te.ct (See Specification 4.5), provided that primary containment integrity is not required and that no work is performed on tiu) reactor or its connected systems which could result
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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 rolsys 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.
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3.1-12b
- u. These functions not required to be operable.when primary containment integrity is not.rnquired.to be maintained.
- w. These functions must be operable only when' irradiated fuci is in the fuel pool or reactor vessel and secondary containment integrity is required per specification 3.5.B.
- y. The number of operable channels may be reduced to 2 por 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 operabic when the associated safety bus is not required to be energized or fully operable as per applicable sections of these technical specifications.
3.7-1 3.7 AUXILIARY ELECTRICAL POWER Acolicability: Applies to the operating status of the auxilary electrical power supply.
Objective: To assure the operability of the auxiliary electrical power supply.
pecificatior. A. The reactor shall not be made critical unless all of .the following requirements are satisfied:
- 1. The following buses or panels energized.
- a. 4153 volt buses 1C and 10 in the turbine building switchgear room.
- b. 460 volt buses IA2, 182, 1A21, 1821 vital MCC 1A2 and 1B2 in the reactor building switchgear room: 1A3 and 183 at the intake structure; 1 A21 A,1821 A,1 A21B, and 18218 and vital MCC 1 AB2 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 switchgear room.
d.120 voit protection panel 1 and. 2 in the cable room. ,_
e.125 volt 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 panel s A and B in the cable room.
- 2. One 230 KV line is fully operational and switch gear and both startup transformers are energizec 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 ED or EC is in the open position. l r
3.7-la B. The reactor shall be placed in the cold shut-oown' positior fit the1 avail ab il ity of power
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falls below.that required by Specification A above, except that;the reactor may remain
. in operation / for a period not to exceed 7
_ days in any 30 day period ^1f a startup transformer -is out of serv ice.
l, lI i
s tr - -- - e .w -
y - w m y
3.7-2 i I
!.one of the engineered safety feature equipmen: f J in I the rensining 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 i of feeding their designated 4160 volt buses. l
- 2. If one diesel generator becomes inoperable during I 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?4 rated power until repairs are completed. 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 diesel generacor may be out of service or the reactor 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 of power available 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 s 'I'. 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 charger failure. These chareers 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 source of AC power, since both 230 KV lines are on the same set of towers, either one-or both 230 KV lines 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.
This- analysis used information on peaking diesels when synchroni ation 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 synchronization is not required.
As indicated 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 for -
3 days. As indicated in Amendment 32 of the Licensing 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 curtailed. It is reasonable to expect that within 8 ,
1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> conditions can be evaluated and the following loads I curtailed:
- 1. One Reactor Building Closed Cooling Water Pump.
- 2. One Core Spray Pump.
- 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. i l
References:
(1) Letter, Ivan R. Finfrock, Jr. to the Director of Nuclear Reactor Regulations dated April 14, 1978.
t Instrument Channel Check Calibrate Test Remarks (Applies to Test G Calibration)
- 19. -Manual Scram Buttons- NA NA 1/3 Mo
- 20. liigh Temperature Main .N A Each Refuel- Each refuel- Using heat source box Steamline Tunnel ing outage ing outage
- * *- Using built-in calibration equipment
- 21. _ SRM
- 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) oCalibrate prior to startup and normal shutdown and thereafter check 1/s and test 1/wk until no longer required. .'-
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4.7-1
. 4.7 AUXILI ARY ELECTRICAL POWER Aopticability: Applies to surveillance ruquirements of the auxiliary electrical supply.
Ob jective: To verify the availability of the auxiliary electrical supply.
S peci f icatien: 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 simulating a loss of offsite power in conjunction with a safety injection actuation test signal) and functionally tested during each ref uel ing outage by:
- a. Verifying de-energization of the emergency busses and load shedding from the emergency busses.
- b. Verifying the diesel starts from snblent conditions on the auto-start signal, energizes the energency busses with permanentl.y connected loads, er.ergizes the auto-connected energency loads through the load sequence timers listed in Table 3.1.1 and operates fore 5 minutes while its generator is loaded with the emergency loads.
- c. Verifying that on diesel generator trip, the loads are shed from the energency busses and the diesel restarts on the auto-start signal, the emergency busses are energized with permanently connected loads, the auto-connected emergency loads are energized through the load sequences and the diesel operates for E 5 minutes while its generator is loaded with -the emergency loads.
- 3. Each diesel generator shall be given a thorough inspection at least annually.
- 4. The diesel generators # fuel supply shall be checked following the above tests.
- 5. The diesel generators8 starting batteries shall be tested and monitored the same as the station batteries, Specification 4.7.B.
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407-la 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 overal l battery voltage is greater -
than or equal to 120 volts (Diesel battery; 112 volts).
- d. Pilot cell speci fic gravity, corrected to 77
- F, shal l be recorded for surveillance review.
- 2. Quarterly 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 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 specific gravity, 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 test 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 voit and to reset at 125 volts il volt (Diesel; 120 volts 1 1 volt) .
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- - i s h The biwcekly tests of the diesel generators are primarily to check for failures and deterioration in the system since last use. The manufacturer has recommended the two week test interval, based on experience with many of their engines.
One factor in determining 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 at least 20'. of rated power rintil engine and generator temperatures have stabilized (about one hour).
The minimum 20'. load will prevent soot formation in the cylinders and injection no::les. Operation up to an equilibrium temperature ensures that there is no over-heat problem. The tests also provide an engine and generator operating history to be compared 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 comprehensive, including procedures that are most effectively conducted at that time. These include automatic actuation and functional capability tests, to verify that the generators car. start and assumenload in less than 20 seconds and testing of the diesel gererator load sequence timers which provide protection from a possible diesel generator overload during LOCA conditions. The annual, thorough inspection will detect any signs of wear long before failure.
The manufacturer's instructions for battery care and mainten-ance with regard to the floating charge, the equalizing charge, and the addition of water will be followed. In addition, vritten records will be maintained of the battery performance. Station batteries will deteriorate with time, but precipitous failure is unlikely. The station surveillance procedures follow the recommended maintenance and testing practices of IEEE STD. 450 which have demonstreted, thorough experience, the ability to provide positive indications of cell deterioration tendencies long before such tendencies cause cell irregularity or improper cell performance.