ML20077E930
| ML20077E930 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/05/1991 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | GPU Nuclear Corp, Jersey Central Power & Light Co |
| Shared Package | |
| ML20077E934 | List: |
| References | |
| DPR-16-A-152 NUDOCS 9106110306 | |
| Download: ML20077E930 (6) | |
Text
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!,_f,[f/i UNITED STATES i
NUCLEAR REGULATORY COMMISSION
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a GPU tJUC L,E A,R,,C,0,k,0p,R,A,TJ 0y bli_U JERSEY CEt4TPAL 00WER & LIGHT CC+! Patly 99.Cf,E,T,yp2 {0-219 OYSTER CREEh iUCLEAR GEt;EPATit;G STAT!Ott f!'.E!.p[1,E!J,T, Tp, ((pyJ,5)p!!,AL, p[Jf f,Tj !1,G, L J C E!g Amendn.ent flo.152 License llo. OPR-1C 1.
The t;uclear Regulatory Coneission (the Concission) has found that:
A.
The application for aniendnent by GPU t uclear Corporation, et al.,
(the licensee), dated tioven:ber 19, 1990, con: plies with the standards and requiren>ents of the Atomic Energy Act of 195t., as anended (the Act), and the Conmission's rules and regulations set forth in 10 EfP Chapter 1; C.
The f acility will operate in conforraity with the application, the provisions of the Act, and the rules and regulations of the Conmission; C.
There is reasonable assurance (i) th6t the activities authorized by this an.endn.ent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in con:pliance with the Consnission's regulations; D.
The issuance of this amendn:ent will not be ininiical to the coneon defense and security or to the health and safety of the public; and E.
The issuance of this an'endn.ent is in accordance with 10 CFR part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9106110306 910600 PDR ADocK 050002 W_
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2 2.
Accordingly, the liter te is atet.ded by changes to the Technical Specificatieris as indicated in the attachnent to this license artendnient, and paragraph ?.C.(2) of 0:ovisional Operating License flo. OPR-10 is hercby atiended to teed as follows:
(2) T e,c h n,1,c a,1,,Sp ec i f i c a t i on s The Technical Specifications contained in Appendices A and 0,, as itvised through Ainendnent tio,152, are hereby incorporated in the license. OpU t;uclear Corporation shall t>perate the f acility in accordance with the Technical Specifications.
3, This license an:endnent is of fective as of the date of issuance, to be inplorented within 30 days of issuance, FOR TFE tiUCLEAR REGULATORY COMMISS10tl
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JotrIf.Stolz,Directo P > ject Directorate 1-4
.ivision of Reactor Projects - 1/11 Office of fiuclear Reactor Regulation Att achnient:
Changes to the Technical Specifications Date of Issuarte: June 5, 1991
AT T AC Ft'E f 4.T..T O..L.I.C.E.f 4 5.E. Al'.E.t@!'.E.f4.T..I40 i s,
IPOV1510f4AL CIEFAllhG L1CEh5E fl0. DPR-16 00C r E l 110,., j,0 - ?,1,9 F(place the followirig pages of the Apper.dir A Technical Specifications with th( enclosed pages as indicated.
The reviset pages are identifitd by araendr ent r.uritier and contain vertical lities indicating the arcas of change.
I(eove I n s e r_t Pese 3.1-4 Page 3.1-4 Page 4.1-5 Page 4.1 5 Pese 4.1-8 Page 4.1-8
l parti:ular proteet t,n anstrucent ta not required: Or the plant is placeo an the protectt:n er safe conditt:n tnat tne instrunent inattates.
This is a :cmpitsneo in a normal manner without sac)ecting the plant t: acnermal Operattens condittens.
The action and out-of-service requirenents apply to all instrumentation within a particular function, e.g., a f the requirements on any one of the ten scram functi:ns cannot to wet : hen c:ntr:1 reds anall to inserted.
The trip level settings not specified in Specification 2.3 have teen included in this specification.
The cases for these settings are discussed telcw.
The nigh drywell pressure trip setting to s 3.5 peig.
This trip wtll scram the reactor, initiate centair. ment spray in conjunction with icw 1:w reacter water l
level, initiate core spray, initiate prtmary containment i ',o l a t io n, initiate depressurt:stien in c:n] unction with icw-low-low-reacter water level, autemati initiate the standby gas treateent system and isolate the reactor building.
The scram function shuts the core down during the less-of-coolant accidents.
A steam leak of about 15 gpm and a liquid leak of about 35 gpm from the primary system will
- ause drywell pressure to reach the scram points and, therefore, the scram provides protecticn for breaks greater than the above.
High drywell pressure provides a second means of initiating the core spray to mitigate the consequences of less-of-ecolant ace dent.
Its trip setting of 13.5 psig initiates the core spray in time to provide adequate Ore cooling.
The breax site ccverage of high drywell pressure was discussed above.
L:w-low water level and high drywell pressure in addition to initiating core spray also causes isolaticn valve closure.
These settings are adequate to cause isola',1cn to minimize the of f site dose within required lirits.
It is permissible to make the drywell pressure instrument channels inoperable during performance of the integrated primary containment leakage rate test provided the reacter is in the cold shutdown condition.
The reason for this is that the Engineered Safety Features, which are effective in case of a LOCA under these conditi ns, will still be effective tecause they will be activated (when the Engineered safety Teatures system is required as identified in the tecnnical specificatten of the system) by low-Icw reacter water level.*
The scram discharge volume has two separate instrument volumes utilized to detect water a;;umulation.
The high water level is based on the design that the water in t he 501s, as detected by either set of level instruments, shall not be allowed to exceed 29.0 gallons; thereby, permitting 137 control rods to scram.
To provide further margin, an accumulation of not more than 14.0 gallons of water, as detected by either instrument volume, will result in a rod block and an alarm.
The accumulati n of not more thaa 7.0 gallens of water, as detected in either instrument volume will result in an alarm.
Oetailed analyses of transients have shown that sufficient protection is provided by other scrime below 45% power to permit bypassing of the turbine trip and generator load rejection scrams.
However, for operational convenience, 40% of rated power has been chosen ss the setpoint below which these trips are bypassed.
This setpoint is coincident with bypass valve capacity.
A low co.'jenser vacuum scram trip of 23 inches Hg has been provided to protect the main condenser in the event that vacuum is lost.
A loss of condenser vacuum would cause the turbine stop valves to close, resulting in a turbine trip Oyster Creek 3.1-4 Amendment tio n 20, 73, 79, 112 152
- Correction:
11/20/87
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TABLE 4.1.1 HINIMUM CllECK. CALIBRATION AND TEST FNEOUENCY FOR PROTECTIVE INSTRUMENTATION 1
M i
N o
Instrument Channel
~
3 N
g ec_h Calibrate Ien H+-marks I AILQies t o T.-st
-in.1 ca l i t> rat iool 1
M f1 M
1.
liigh Heactor Pressure l
1/d Note 3 1/Mo l
3 2.
liigh Drywel1 Pressure (Scram)
N/A 1/3 sno.
Note 1 By application of test pressure i
i 3.
Low Reactor Water Level 1/d Note 3 1/Mo j
4.
Low-Low Water Level 1/d Note 3 1/Mo t
5.
Iligh Water Level in Scram Discharge Volume a.
Digital N/A 1/3 mo.
1/3 mo.
By varying level in sensor columns b.
Analog N/A Note 3 1 mo.
6.
Low-Low-Low Water Level N/A 1/3 mo.
Note 1 By application of test paessume i
b
{
47.
Ifigh Flow in Main Steamline 1/d 1/3 mo.
Note 1 By applicat ion of test ps' **suu s -
8.
Low Pressure in Main Steamline N/A 1/3 mo.
Note 1 By application of test prerssure l
9.
Iligh Drywell Oressure 1/d Note 1 By application of test pressuce l
(Core Cooling) i;
]
10.
Main Steam Isolation Valve (Scram) N/A N/A 1/3 mo.
By exercising valve.
O
- r>
. g NOTE 1:
y3 Initially once/ month, thereafter according to Figure 4.1.1, with an interval not three months.
less than one a enth nut own e
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" NOTE 2 At least daily during reactor power operation, the reactor neutron flux peaking factor shall le estimat.31 an.1 i
flow-referenced APRM scram and rod block settings shall be adjusted, if necessary, as specified in Section 2.3
]
v' Specifications (1) (a) and (2) (a).
1 C NOTE 3:
Calibrate electronic bistable trips by injection of an external test current once per 3 smonths.
Ca l a br et e transmitters by application of test pressure once per 12 months.
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TABLE 4. ld r
M (cont *d) t 28
)
O Instrument Channel Check Calibrate Test Pemarks (Applies to Test and Calibrationi a
l N
- 27. Scram Discharge Volume (Rod Block)
{
a) Water level N/A Each re-2/3 Mo Calibrate by varying level in j
high fueling senscr column outage J
)
b) Scram Trip N/A N/A Each re-j bypass i
fueling outage J
j
- 28. Loss of Power a) 4.16 KV Daily 1/24 mos.
1/mo.
Emergency Bus Undervoltage u
(Loss of voltage) i e
I m
j b) 4.16 KV Daily 1/24 mos.
I/mo.
j Emergency Bus Undervoltage (Degraded Voltage)
I'
- 29. DrywelI High N/A Each re-Each re-
{
g Radiation fueling fueling g3 outage outage e
4 1,
O-
- Calibrate prior to startup and normal shutdown and thereaf ter check I/s and test 1/wk until no longer required.
tt
~
-ag Leaend: t4/A = Not Applicable; 1/s = Gnce per shift; 1/d = Once per day; 1/3d = Once per three days; y
1/wk = Once perweeks 1/3 mo = Once every 3 months; 1/18 mos. = Once every 18 months jo The following notes are only for Item 15 of Table 4.1.1:
I g A channel may be taken out of service for the purpose of a check, calibration, test or maintenance without dec t.s r ug the channel to be inoperable.
"o,e a.
The channel functional test shall also demonstrate that control room alarm annunciation occurs it any of t i.e following conditions exists:
y
{
g
- 1) Instrument indicates measured levels above the alarm setpoint.
j
- 2) Instrument indicates a downscale failure.
- 3) Instrument controls not set in operate mode.
H
- 4) Instrument electrical power loss.
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