ML20079Q460
| ML20079Q460 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/30/1984 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20079Q461 | List: |
| References | |
| NUDOCS 8402010175 | |
| Download: ML20079Q460 (12) | |
Text
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UNITED STATES y., (
g NUCLEAR REGULATORY COMMISSION
- ,' \\p
( t WASHINGTON, D. C. 20555 l
GPU NUCLEAR CORPORATION AND JERSEY CENTRAL PCWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION A'!ENDMENT TO THE AMENDED PROVISIONAL OPERATING LICENSE Amendment No. 71 License No. DPR-16 ll 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by GPU Nuclear Corporation and Jersey Central. Power and Light Company (the licensees) dated September 2, 1983, as supplemented. December 2, 1983. complies with the standards and requirements 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; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) 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 conducted in compliar.ce with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
8402010175 840130 PDR ADDCK 05000219 P
pg
, 2.
Accordingly, the license is amended by changes to the Technical Specifications as_ indicated in the attachment to this license amendment and Paragraph 2.C(2) of Provisional Operating License g
No. DPR-16 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 71, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f hl hp-4+
Dennis M. Crutchfield, C) fief Operating Reactors Brandh #5 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: Januery 30, 1984 s
[
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J
ATTACHMENT TO LICENSE AMENDMENT NO. 71 PROVISIONAL OPERATING LICENSE N0. OPR-16 DOCKET NO. 50-219 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are ider.tified by the captioned amendment number and contain vertical lines indicating the area of change.
PAGES 2.3-1 2.3-3 2.3-5 2.3-6 3.1-8 3.1-14 3.3-2a 4.1-4
}
4.1-6 s
2.3-1 l
I 2.3 LIMITIEG SAFETY SYSTD1 SETTINGS Applfc3hility:
Applies to trip settings on automatic protective devices related to variabics on which safety limits have been placed.
Obj ective:
To provide autoestic corrective ac, tion to prevent the safety limits from being exceeded.
Sp cification:
Limiting safety system settings shall be as follows:
FUNCTION LIMITING SAFETY SYSTDI SETTINGS 1)
Neutron Flux, Scram a)
APRM For recir'culation flow, M3,61 x 106 lb/hr:
j[ ([1.'34 x 10-6) W + 34.0) percent of-rated neutron flux when total peaking factors in all fuel types are less than or equal to those in Specif1-cation 2.1.A.1, or I
s The lowest value of:
PF
$ ([1,3'4 x 10-6] W '_34.0)
I 3
o PF percent of rated neutren flux frc=
among those calculations for cach fuel type with total peaking.facters.
PF > PF, where Pro = peaking factor in o
Specification 2.1.A.l.
6 For recirculation flow, W)61 x 10 lb/hr:
1 115.7 percent of rated neutron flux when total peaking f acters in all fuel types-are less than or equal'to those in: Specification 2.1.A.1, or PF The lowest value of 1.115.7 E PF } percent o
of rated neutron flux from among those
' calculations for each fuel type with totsi peaking f actors PF > PT, where PFo=peakingfactorinSpeci!1 cation 2.1.A l.
b)
IRM i 38.4 percent of rated neutron flux.
Amendment No.sid', 71
- - m _
m_
P.325 psig (initiated in IRM
- 7) Lov P ccoure Main Steam Lin2, range 10)
^
ESIT Closura f.10" Valve Closure fro = full
- 8) Main Stan= Line Isolation Valve ope closure, Scra=
- 9) Reactor Lov Water Level, Scra=.
E11',5" above the top of the ac:1ve fuel as indicated under nor=al operating c:nditic:
- 10) desetorLow-LovWaterLevel, t 7',2" above the top of the active fuel as. indicated under Main *S'tean Line Isolatien Valve no:=al operating condi.icns.
- Closure,
- 11) Reac:e Low-Lov Water Level, E 7'2" above the top of the active fuel.
Core Spray Initiatics E 7'2" above. the top of the
- 12) Reactor Low-Lov Water Level, active fuel with time delay Is,olation Condense Initiatics 13 seconds.
10 percent turbine atop valve (s
- 13) Turbine Trip Scra=
clesure fren full cpen.
- 14) Generater Load' Rejection Scra=
Isi:iate upon less of oil pressure fren turbine acceleration relay.
3 ASIS:
Safety limits have been established,in Specifica:icus 2.1 and 2.2
- o protec: the integrity of the fuel cladding and reactor coolan:
systa= barriers.
Autc=atic protective devices have been p cvided in the plant design to take cc :ee:ive actica :o preven the saft:7 li=1:s fre= being exceeded in==:=al opera:ics or opera:ional
- ansients caused by reasonable expected single operator error c:
equip =en: =alfunction.
This Specification establishes the ::1p se:.ings fc: these au:==atic proeaction devices.
)
The Average Power Range Moniter, A?RM
, ::1p set-ing has been es:ablished to assure never reaching the / fuel cladding integrity safe:y 1 N r.
The APRM system responds to cha=ges in neutron flux.
However, near rated ther=al poue: the A??X is calibrated, using a plan: heat balance, so tha
- he neu::en flux that is sensed is read ou: as percent of rated ther.,.a1 power.
?or slov =aneuvers, those where core ther=al power, surface hes: flux, and the pcuer ::ansfer:ed to the water f ollow the nen::en flux, the A??3 vill :ead reactor ther=al power.
For fas: ::ansients, the neu::en flux vill lead the power
- ansf erred f c= the cladd:.ng to the va:e: due to the effect of the fuel ti=e constan:. Therefore, when :he neu:: n flux increases to the scra= se::ing, :he perce=: increase in heat flur. and pcVer transferred to the water vill be less than the pe:cen: i= crease in neu::en flu::.
"he A??M ::1p se::ing -ill be varied au:enatically vi:h recircula:ics 6 lb/h: c l
flev vi h :he ::1p se::ing a rated flev 61.0 x 10 grea:e: being 115.7" c f ra:ed ne --~ -
lased en a c =ple:e 2.3-3 Amendment No. 75,,2d, 71 i
h---
2.3-3 For operation in the Startup node while the reactor is at low pressure, the IRM rangs 3 High Flux scram setting of 12% of the rated power provides adequate thermal margin between the maximum power and the safety limit of 18.3% of rated power to accommodate anticipated maneuvers associated with power plant startup. There are a fewpossible sources of rapid reactivity input to the system in the low power / low flow ccr.dition. Effects of increasing pressure at zero or low void content are minor, because cold water from sources available during the startup is not much colder than that already in the system, temperature coefficients are small, and control rod sequcaces are constrained by operating procedures backed up by the rod worth minimizer. Worth of individual rods is very low in a constrained rod pattern. In a sequenced rod withdrawal approach to the scram level, the rate of power rise is no rore than five permnt of the rated per minute, and the IRM system would be more than adequate to assure a scram before the power could exceed the safety limit.
To continue operation beyond 12% of rated power, the IRMs must be transferred into range 10. The Reactor Protec' ion System is designed such that reactor pressure must be above 825 psig to successfully transfer the IRM3 into range 10, thus assuring protection for the fuel cladding safety limit. The IRM scram remains active until the mode switch is placed in the RUN position at which time the trip becomes a coincident IRM upscale, APRM downscale scram.
The adequacy of the IRM scram was determined by comparing the scram level on the IRM range 10 to the scram level on the APRMs at 30% of rated flow. The IRM scram is at 38.4% of rated power while the APRM scram is at 52.7% of rated power. The mininum flow for Oyster Creek is at 30% of rated power and this would be the lowest APRM scram point. The increased recirculation flow to 65% of flow will provide additional margin to CFR limits. The APRM scram at 65% of rate flow is 87.1% of rated power, while the IRM range 10 scram remains at 38.4% of rated power.
Therefore, transients requiring a scram based on flux excursion will be terminated sooner with a IRM range 10 scram then with an APRM scram. The transients requring a scram by nuclear instrumentation are.the loss of i
feedwater heating and the improper startup of an idle recirc loop. The loss of feedwater heating transient is not affected by the range 10 IRM since the feedwater heaters will not be put into service until after the LPRM downscales have cleared, thus insuring the operability of the APRM system. This will be administratively controlled. The improper startup of an idle recirc loop becomes less severe at lower power level and the IRM scram would be adequate to terminate the flux excursion.
The Rod Worth Minimizer is not required beyond 10% of rated power. The ability of the IRMs to terminate a rod withdrawal transient is limited due to the nunber and location of IRM detectors. An evaluation was performed that showed by maintaining a mininum recirculation flow of 39.65x106 lb/hr in range 10 a complete rod withdrawal initiated at 35%
of rated power or less would not result in violating the fuel cladding safety limit. Therefore, a rod block on the IRMs at less than 35% of rated power would be adequate protection against z. rod withdrawal transient.
Amendment No. 71
)
o 2.3-6 The reactor coolant system safety valves offer yet another protective feature for the reactor coolant system pressure safety limit since these valves are sized assuming no credit for other pressure relieving devices.
In compliance with,
Section I of the ASME Boiler and Pressure Vessel Code, the safety valves must be set to open at a pressure no higher than 103% of design pressure, and they must limit the reactor pressure to no more than 110% of design pressure.
The safety valves are cized according to the code for a condition of turbine stop valve closure while operating at 1930 MW(t),
followed by (1) a delay of all scrams, (2) failure of the turbine bypass valves to open, and (3) failure of the isolation condensers and relief valves to operate. Under these conditions, a total of 16 safety valves are required to turn the pressure transient. For analysis purposes, the void reactivity coefficient.as also pessimistically increased by 50%,
i.e.,
a void coefficient 1.5 times normal.
Uit'.
'e safety valves set as specified herein the maximum vessel piassure (at the bottom of the pressure vessel) would be about 1301 psig (9);
maximum pressure at the lowest point in the recirculation loop is approximately 1315 psig which is 60 psi below the safety limit. The ASME B&PV Code allows a 11% of working pressure (1250 psig) variation in the poy point of the valves.
This variation is recognized in Specification 4.3.
The low pressure isolation of the main steam lines at 825 psig was provided to give protection against fast reactor depressuri-zation and the resulting rapid cool-down of the vessel.
Advantage was taken of the scram feature which occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the startup position and the IRMs be in the range 9, or lover, where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux i
scram. Thus, the co=binati9n of the main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire rang 2 of appli-cability of the fuel cladding integrity safety limit.
In addition the isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure.
With the scrams set at 10% valve closure, there is no increase in neutron flux and the peak pressure is limited to 1110 psig (9).
Amendmentflo.J',71 l
3.1-8 TABl.E 3.1.1 l'ROTECr1VE INSTRilHF14TATION REQUIRDIENTS (CatITD)
Hin. No. of Hin. No. of Operable Reactor Hodes Operable or I ns t rtunent in Which Function Operating Chantiels Fer Must lie Operable (Tripped) Trip Operable Action Function Trip. Setting Shutdown Refuel Startuq_ Run Systems Trip Systesas Requireda Close main steam B.
Reacto. Isolation isolation valves
- 1. Ia.-Low Reactor X
X X
X 2
2 and close isola-tion condenser Water Level vent valves, or i
120% rated X (s)
X (s)
X X
2 2
place in cold
- 2. liigh Flow in 3
shutdown condt-Hain Steam-
" tion line A
- 3. liigh Flow in 3 120% rated X (s)
X (s)
X X
2 2
Main Steam-line B
- 4. liigh Tempera-3 Ambient at X (s)
.X (s)
X X
2 2
ture in Hain Power + 50*F Steamline Tunnel l
X X
2 2
- 5. 1.ow Pressure In Hain Steam-line
- 6. liigh Radiation 3 10X Normal X (s)
X(s)
X X
2 2
d in liain Steam Backstound Tunnel 1
C.
Isolation Condenser X(s)
X(s)
X X
2 2
Place plant in 1.
High Eeactor cold shutdown Preasure ennditton 2.
l.ow-Low Reactor 2, 7' 2" above X(s)
X(s)
X X
2 2
Water top of AmendmentNo.g
, 71
3.1-14
.~
_.1BLE 3.1.1 (Cont'd)
T Those functions not renuired to be operable when the ADS is not required to be operable.
v.
These functions must be operable only when irradiated fuel is-in the fuel pool or reactor w.
vessel ana secondary containment integrity is required per specification 3.5.8.
The number _of operable channels may be reduced to 2 per Specificatior 3.9.E and F.
y.
The bypass function to permit scram reset'in the shutdown or refuel mode with control z.
rod block must be operable in this mode.
Pump circuit breakers will be tripped in 10 seconds t 15% during a LOCA by relays SK7A aa.
(
and SK8A.
bb.
Pump circuit breakers will trip instantaneously during a LOCA.
Only applicable during'STARTUP Mode while operating in IRM Range 10.
l cc.
Amendment No.,4(,)RI,)PI, 71
i 3.3-2a G.
Prirary Coolant System Pressure Isolation Valve _s_
Acolicability:
Operational Conditions - Startup and Run Modes; applies to the cperational. status of the prirary coolant syatem pressure isolation valves.
g ective:
To increase the reliability of prirary coolant system pressure isolation valves thereby reducing the potential of an intersystem loss of coolant accident.
Specification:
1.
Duri'ng reactor power operating conditions, the integrity of all pressure isolation valves listed in Table 3.3.1 shall be demonstrated. Valve leakage shall not exceed the amounts indicated.
2.
If Specification 1 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
H.
Recaired Minimum Recirculation Flow Rate for Ooeration in IRM Range 10 1.
During STARTUP mode operation, a minimum recirculation flow rate is required before operating in IRM range 10 to insure that technical specification transient MCPR limits for operation are not exceeded. Tnis minimum flow rate is no longer required once the reactor is in the RUN mode.
2.
39.65 x 106 lo/hr is the minimum recirculation flow rate necessary for operation in IRM range 10 at this time. This flow rate leaves sufficient margin between the minimum flow required by tne R E analysis performed and the minimum flow used while operatir.g in IRM range 10.
llRC Order dated April 20, 1981 Amendment flo. 71
r
=
n 4.1-4 The logic of the instrument safety systems in Table 4.1.1 is such that testing the instrument channels also trips the trip system, verif ying that it is operable. However, certain systems require coincident instrument channel trips to completely test their trip systems. Therefore, Table 4.1.2 specifies the minimum trip system test f requency for these tripped. systems. This assures that all trip systems f-or protective instrumentation are adequately tested, f rom sensors through the trip system.
Evury element of electrical circuitry for the reactor protection system is to be verified cperable prior to plant startup by functional testing. Parallel elements of circuits which do not permit functional verification of freedom from shorts by routine channel trips are to be verified functional during refueling shutdown.
IRM calibration is to be performed during reactor startup. The calibration of the IRMs during startup will be significant since the IKMs will be relied on for neutron monitoring and reactor protection up to 38.4% of rated power during a reactor startup.
Refere nc es: (1)
" Reliability of Engineered Saf ety Features as a Function of Testing Frequency," I. M. Jacobs, Nuclear Saf ety, Volume 9, No. 4, July-Aubest, 1968.
(2)
" Reactor Protection System, A Reliability Analysis,"
I. M. Ja'cobi, APED-5179, Eng. A-16, June,1966.
6 Amendment No. 71
4.1-6 TABLE 4,1.1 (cont'd)
Instriment Channel
- Cheqk, Calibrate Test Remarks (applies to Test & Calibration) 14.
Iligh Radiation in Reactor Building-j Operating Floor 1/s 1/3 mo,
1/wk Using gamma source for calibration Ventijation Exhaust 1/s 1/3 mo, 1/wk Using gamma source for calibration 15.
liigh Radiation on Air 1/a l'/3 no 1/wk Using builtrin calibration equipment Ejector Of f-Cas each NA 16.
IRH Level HA startup Using built-in calibration equipment' 1RM Scram 17.
IRN Blocks NA Prior to Prior to Upscale and downscale startup and startup &
shutdown shutdown 18.
Condenser Low Vacuum NA Each refuel-Each ing outage refueling.
outage
- Calibrate prior to startup and normal shutdown and thereafter check 1/s and test 1/wk unti1 no longer requierd.
Legend 3./d = Once per day; 1/3d = Once per 3 days; 1/wk = Once per week; NA = Not applicable; 1/s = Once per shift; 1/3 mo = Once every 3 months.
Asendment No.M 'l
w-mm--w--------'
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