ML20213G731
| ML20213G731 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/31/1986 |
| From: | Donehew J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20213G733 | List: |
| References | |
| NUDOCS 8611180299 | |
| Download: ML20213G731 (12) | |
Text
.
, p@n%,[g UNITED STATES 4
~ 'g NUCLEAR REGULATORY COMMISSION g
y K
IVASHINGTON, D. C. 20555 b
J GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET N0. 50-219 0YSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No.112 License No. DPR-I6 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 11, 1986, complies with the standards and require-ments 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; B.
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 (i) that the activitics 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 compliance 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 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.
I i
8611180299 86103:1 DR ADOCK 05000219 PDR
. -. ~
. 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.(?) of Provisional Operating License 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.112, are hereby incorporated in the license. GPl! 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 h
9 Jack N. Donohew, J kojectManager BWR Pro.iect Directorate #1 ivision of BWR Licensing
Attachment:
Changes to the Technical l
Specifications Date of Issuance: October 31, 1986 i
p
ATTACHMENT TO LICENSE AMENDMENT N0.112 PROVISIONAL OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicatina the area of change.
REMOVE INSERT 3.1 4 3.1-4 3.1-5 3.1-5 3.1-6 3.1-6 3.1-7 3.1-7 3.1.8 3.1.8 3.1-11 3.1-11 3.1-12 3.1-12 3.1-13 3.1-13 3.1-17 3.1-17 i
.m particular protection instrument.is not required; or the plant is placed in the protection or safe condition that the instrument initiates.
This is accomplished in a normal manner without subjecting the plant to abnormal operating conditions.
The action and out-of-service requirements apply to all instrumentation within a particular function, e.g., if the requirements on any one of the ten scram l
functions cannot be met then control rods shall be inserted.
1 The trip level settings not specified in Specification 2.3 have been included in this specification.
The bases for these settings are discussed below.
The high drywell pressure trip setting is 53.5 psig.
This trip will scram the l
reactor, initiate reactor isolation, initiate containment spray in conjunction with low low reactor water level, initiate core spray, initiate primary contain-ment isolation, initiate automatic depressurizatien in conjunction with low-low-low reactor water level, initiate the standby gas treatment system and isolate i
the reactor building.
The scram function shuts the core down during the loss-of-coolant accidents. A steam leak of about 15 gpm and a liquid leak of about 35 gpm from the primary system will cause drywell pressure to reach the scram point; and, therefore the scram provides protection for breaks greater than the above.
High drywell pressure provides a second means of initiating the core spray to mitigate the consequences of a loss-of coolant accident.
Its trip setting of
<3.5 psig initiates the core spray in time to provide adequate core cooling.
The break-size coverage of high drywell pressure was discussed above.
Low-low water level and high drywell pressure in addition to initiating core spray also causes isolation valve closure. These settings are adequate to cause isolation to minimize the offsite dose within required limits.
i It is permissible to make the drywell pressure instrument channels inoperable during performance of the integrated primary containment lerkage rate test provided the reactor is in the cold shutdown condition.
The reason for this 4
is that the Engineered Safety Features, which are effective in case of a LOCA under these conditions, will still be effective because they will be activated i
by low-low reactor water level.
4 i
The scram discharge volume has two separate instrument volumes utilized to j
detect water accumulation.
The high water level is based on the design that the water in the SDIV's, as detected by either set of level instruments, shall not be allowed to exceed 29.0 gallons; thereby, permitting 137 control rods to To provide further margin, an accumulation of not more than 14.0 gallons scram.
of water, as detected by either instrument volume, will result in a rod block and an alarm.
The accumulation of not more than 7.0 gallons of water, as detected in either instrument volume will result in an alarm.
j Detailed analyses of transients have shown that sufficient protection is provided by other scrams below 45% power to permit bypassing of the turbine i
trip and generator load rejection scrams.
However, for operational convenience, 40% of rated power has been chosen as the setpoint below which these trips are i
bypassed.
This setpoint is coincident with bypass valve capacity.
A low condenser vacuum scram trip of 23" Hg has been provided to protect the eain condenser in the event that vacuum is lost.
A loss of condenser vacuum wsuld cause the turbine stop valves to close, resulting in a turbine trip OYSTER CREEK 3.1-4 Amendment No.: 20, 73, 79, 112
The low condenser vacuum trip anticipates this transient and scrams the reactor.
The condenser is capable of receiving bypass steam until 7" Hg vacuum thereby mitigating the transient and providing a margin.
Main steamline high radiation is an indication of excessive fuel failure.
Scram and reactor isolation are initated when high activity is detected in the main steam lines.
These actions prevent further release of fission products to the environment.
This is accomplished by setting the trip at 10 times normal rated power background.
Although these actions are initiated at this level, at lower activities the monitoring system also provides for continuous monitoring of radioactivity in the primary steam lines as discussed in Section VII-6 of the FDSAR.
Such capability provides the operator with a prompt indication of any release of fission products from the fuel to the reactor coolant above normal rated power background. The gross failure of any single fuel rod could release a sufficient amount of activity to approximately double the background activity at normal rated power.
This would be indicative of the onset of fuel failures and would alert the operator to the need for appropriate action, as defined by Section 6 of these specifications.
The settings to isolate the isolation condenser in the event of a break in the steam or condensate lines are based on the predicted maximum flows that these systems would experience during operation, thus permitting operation while affording protection in the event of a break.
The settings correspond to a flow rate of less than three times the normal flow rate of 3.2 x 105 lb/hr.
Upon initiation of the alternate shutdown panel, this function is bypassed to prevent spurious isolation due to fire induced circuit faults.
The setting of ten times the stack release limit for isolation of the air-ejector offgas line is to permit the operator to perform normal, immediate remedial action if the stack limit is exceeded.
The time necessary for this action would be extremely short when considering the annual averaging which is allowed under 10 CFR 20.106, and, therefore, would produce insignificant effects on doses to the public.
Four radiation monitors are provided which initiate isolation of the reactor 4
building and operation of the standby gas treatment system.
Two monitors are located in the ventilation ducts, one is located in the area of the refueling pool and one is located in the reactor vessel head storage area.
The trip logic is basically a 1 out of 4 system.
Any upscale trip will cause the desired action.
Trip settings of 17 mr/hr in the duct and 100 mr/hr on the refueling floor are based upon initiating standby gas treatment system so as not to exceed allowed dose rates of 10 CFR 20 at the nearest site boundary.
The SRM upscale of 5x105 CPS initiates a rod block so that the chamber can be relocated to a lower flux area to maintain SRM capability as power is increased to the IRM range.
Full scale reading is 1 x 106 CPS.
This rod block is bypassed in IRM Ranges 8 and higher since a level of 5 x 105 CPS is reached and the SRM chamber is at its fully withdrawn position.
The SRM downscale rod block of 100 CPS prevents the instrument chamber from being withdrawn too far from the core during the period that it is required to monitor the neutron flux.
This downscale rod block is also bypassed in IRM OYSTER CREEK 3.1-5 Amendment No.: 2,7. 112 01/5/71, 11/5/71 i
.7L-.
~* ~~ ~ ~ ~~
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Ranges 8 and higher.
It is not required at this power level since good indica-tion exists in the Intermediate Range and the SRM will be reading approximately 5 x 105 CPS when using IRM Ranges 8 and higher.
The IRM downscale rod block in conjunction with the chamber full-in position and range switch setting, provides a rod block to assure that the IRM is in its most sensitive condition before startup.
If the two latter conditions are satisfied, control rod withdrawal may commence even if the IRM is not reading at least 5%.
However, after a substantial neutron flux is obtained, the rod block setting prevents the chamber from being withdrawn to an insensitive area of the core.
The APRM downscale setting of > 2/150 full scale is provided in the run mode to prevent control rod withdrawal without adequate neutron monitoring.
High flow in the main steamline is set at 120% of rated flow. At this setting the isolation valves close and in the event of a steam line break limit the loss of inventory so that fuel clad perforation does not occur. The 120% flow would correspond to the thermal power so this would either indicate a line break or too high a power.
Temperature sensors are provided in the steam line tunnel to provide for closure of the main steamline isolation valves should a break or leak occur in this area of the plant. The trip is set at 50 F above ambient temperature at rated power. This setting will cause isolation to occur for main steamline breaks which result in a flow of a few pounds per minute or greater.
Isolation occurs soon enough to meet the criterion of no clad perforation.
The low-low-low water level trip point is set at 4'8" above the top of the active fuel and will prevent spurious operation of the automatic relief system.
The trip point established will initiate the automatic depressurization system in time to provide adequate core cooling.
Specification 3.1.B.1 defines the minimum number of APRM channel inputs required to permit accurate average core power monitoring.
Specifications 3.1.B.2 and 3.1.C.1 further define the distribution of the operable chambers to provide monitoring of local power changes that might be caused by.a single rod withdrawal.
Any nearby, operable LPRM chamber can provide the required input for average core monitoring.
A Travelling Incore Probe or Probes can be used temporarily to provide APRM input (s) until LPRM replacement is possible.
Since APRM rod block protection is not required below 61% of rated power,(1) as discussed in t
Section 2.3, Limitina Safety System Settinas, operation may continue below 61%
as long as Specification 3.1.B.1 and the requirements of Table 3.1.1 are met.
In order to maintain reliability of core monitoring in that quadrant where an i
APRM is inoperable, it is permitted to remove the operable APRM from service for calibration and/or test provided that the same core protection is maintained by alternate means.
In the rare event that Travelling In-core Probes (TIPS) are used to meet the requirements 3.1.B or 3.1.C, the licensee may perform an analysis of substitute LPRM inputs to the APRM system using spare (non-APRM input) LPRM detectors and change the APRM system as permitted by 10 CFR 50.59.
Change: 6 OYSTER CREEK 3.1-6 Amendment No.: 9, 15, 112 01/4/80 g
3 "S
l Under assumed loss-of-coolant accident conditions and under certain loss of offsite power conditions with no assumed loss-of-coolant accident, it is inadvisable to allow the simultaneous starting of emergency core cooling and heavy load auxiliary systems in order to minimize the voltage drop across the cmergency buses and to protect against a potential diesel generator overload.
The diesel generator load sequence time delay relays provide this protective function and are set accordingly.
The repetitive accuracy rating of the timer mechanism as well as parametric analyses to evalua'.e the maximum acceptable tolerances for the diesel loading sequence timers wre considered in the establishment of the appropriate load sequencing.
Manual actuation can be accomplished by the operator and is considered appro-priate only when the automatic load sequencing has been completed.
This will prevent simultaneous starting of heavy load auxiliary systems and protect against the potential for diesel generator overload.
Also, the Closed Cooling Water and Service Water pump circuit breakers will trip whenever a loss-of-coolant accident condition exists.
This is justified by Amendment 42 of the Licensing Application which determined that these pumps were not required during this accident condition.
Reference:
(1) NE00-10189 "An Analysis of Functional Common Mode Failures in GE BWR Pro-tection and Control Instrumentation," L. G. Frederick, et al., July 1970.
Change: 6 OYSTER CREEK 3.1-7 Amendment No.: 9, 15, 44, 60, 112 U
i
g S
TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS n
9 n
Reactor Modes Min. No. of Min. No. of M
in which Function Operable or Instrument E
Must Be Operable Operating Channels Per
[ tripped]
Operable Action Function Trip Setting Shutdown Refuel Startup Run Trip Systems Trip Systems Required
- A.
Scram Insert control rods 1.
X X
X 2
1 2.
High Reactor X(s)
X X
2 2
Pressure 3.
High Drywell 5 3.5 psig.
X(u)
X(u)
X 2
2 w
Pressure a>
4.
Low Reactor X
X X
2 2
Water Level Fg
- 5. a. High Water 5 29 gal.
X(a)
X(z)
X(z) 2 2
i Level in Scram g
Discharge Volume
<+
North Side E
- b. Higher Water 5 29 gal.
X(a)
X(z)
X(z) 2 2
Level in Scram Discharge Volume South Side 6.
Low Condenser
> 23" hg.
X(b)
X(b)
X 2
2
.O Vacuum d
7.
High Radiation
< 10 x normal X(s)
X X
2 2
in Main Steam
Background
g Line Tunnel
.~
N
?
e Q
TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS (CONTD) 49 n
Reactor Modes Min. No. of Min. No. of g
in which Function Operable or Instrument p
Must Be Operable Operating Channels Per
[ tripped]
Operable Action Function Trip Setting Shutdown Refuel Startup Run Trip Systems Trip Systems Required
- D.
Core Spray Consider the respective 1.
Low-Low Reactor X(t)
X(t)
X(t)
X 2
2 core spray Water Level loop inoper-able, and 2.
High Drywell 5 3.5 psig X(t)
X(t)
X(t)
X 2(k) 2(k) comply with Pressure Spec. 3.4 3.
Low Reactor 1 285 psig X(t)
X(t)
X(t)
X 2
2 w
Pressure (valve
.y permissive)
E.
Containment spray Consider the 1.
High Drywell 5 3.5 psig X(u)
X(u)
X(u)
X 2(k) 2(k) containment Pressure spray loop inoperable 2.
Low-Low Reactor 1 7'2" above X(u)
X(u)
X(u)
X 2
2 and comply Water Level top of active with Spec.
fuel 3.4
[ F.
Primary Containment Isolation Isolate con-
=
tainment or
[
1.
High Drywell 5 3.5 psig X(u)
X(u)
X(u)
X 2(k) 2(k) place in g
Pressure cold shutdown condition p$
2.
Low-Low Reactor 1 7'2" above X(u)
X(u)
X(u)
X 2
2
=
Water Level top of active j g fuel M
N e
S TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS (CONTD)
M i
n E
Reactor Modes Min. No. of Min. No. of p
in which Function Operable or Instrument Must Be Operable Operating Channels Per
[ tripped]
Operable Action Function Trip Setting Shutdown Refuel Sta ri.up Run Trip Systems Trip Systems Required
- G.
Automatic Depressurization 1.
High Drywell
$ 3.5 psig X(v)
X(v)
X(v)
X 2(k) 2(k)
See note h Pressure 2.
Low-Low-tow
> 4'8" above X(v)
X(v)
X(v)
X 2
2 See note h Reactor Water top of active Level fuel 3.
AC Voltage NA X(v)
X 2
2 Prevent auto w
depressuriza-g O
tion on loss m
of AC power.
See note i H.
Isolation condenser Isolation (See Note hh) g g
1.
High Flow 5 20 psig P X(s)
X(s)
X X
2 2
Isolate M
Steam Line k
Affected Isolation
- g 2.
High Flow 1 27" P H O X(s)
X(s)
X X
2 2
condensor, g
Condensate 2
pg Line comply with q2 Spec. 3.8 WD O See note dd S '
I.
Offgas System Isolation 0 g 1.
High
< 10 x Stack X(s)
X(s)
~X X
1 2
reactor or Isolate Radiation Release ilmit t
In Offgas (See 3.6-A.1) trip the In-m w y*
Line (e) operable in-sg instrument E co channel M
M
i
.r S
TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS (CONTD)
E n
Reactor Modes Min. No. of Min. No. of g
in which Function Operable or Instrument i
p Must Be Operable Operating Channels Per
[ tripped]
Operable Action i
Function Trip Setting Shutdown Refuel Startup Run Trip Systems Trip Systems Required
- J.
Reactor Building Isolation and Standby Gas Treatment System Initiation Isolate Reactor Bldg.
1.
High Radiation
-< 100 Mr/Hr X(w)
X(w)
X 1
1 and Initiate Reactor Building Operation Floor Standby Gas Treatment Sysm tem or Man-2.
Reactor Bldg.
5 17 Mr/Hr X(w)
X(w)
X X
1 1
ual Surveil-Ventilation Exhaust lance for w
not more than 24 s,
hours "w
(total for 3.
High Drywell 5 3.5 psig X(u)
X(u)
X X
1(k) 2(k) all instru-Pressure ments under J) in any 4.
Low Low 1 7'2" above X(gg)
X X
X 1
2 30-day period Reactor Water top of active
[
Level fuel
=
p ( K.
Rod Block No control
, s 5
rod k[
1.
SRM Upscale 5 5 x 10 cps X
X(1) 1 2
withdrawals
- o II) permitted 2.
SRM Downscale 1 100 cps X
.X(1) 1 2
g-3.
IRM Downscale 1 5/125 fu11 scale (g)
X X
2 3
suk*
4.
APRM Upscale X(s)
X X
2 3(c) oo e 5.
APRM Downscale 1 2/150 fu11 scale X
2 3(c)
SE 4
TABLE 3.1.1 (CONTD)
E
- gg.These functions are not required to be operable when secondary containment is not required to be maintained cs Ni or when the conditions of Sections 3.5.b.1.a b, c, and d are met, and reactor water level is closely 92 monitored and logged hourly.
The Standby Gas Treatment System will be manually initiated if reactor water level drops to the low level trip set point.
hh. The high ilow trip function for "B" Isolation Condenser is bypassed upon initiation of the alternate shutdown panel.
This prevents a spurious trip of the isolation condenser in the event of fire induced l
circuit damage.
l I
l l
P ya o$
I i
- 5.
a kI 5
I~
Ef L
um 1
This note is applicable only during the Cycle 10M outage.
d" m
,