ML20126E670
| ML20126E670 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 05/30/1985 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Commonwealth Edison Co, Iowa Illinois Gas & Electric Company |
| Shared Package | |
| ML20126E672 | List: |
| References | |
| DPR-30-A-086 NUDOCS 8506170173 | |
| Download: ML20126E670 (28) | |
Text
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UNITED STATES
,9 _
_ p, NUCLEAR REGULATORY COMMISSION
.n 7j WASHINGTON, D. C. 20555
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COMMONWEALTH EDISON COMPANY AND IOWA-ILLIN0IS GAS KR5 ELECTRIC COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 86 License No. DPR-30
- 1. ~The Nuclear Regulatory Commission-(the Commission) has found that:
A.
The applications for amendment by Commonwealth Edison Company (the licensee) dated January 3 and February 4, 1985, comply 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;.
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 activities authorized by this amendment can be conducted without endangering the health
- s
-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.
. coninon defense and security or to the health and safety-of the public; and i
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.
2.
Accordingly, the license is amended by changes to the Technical" Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-30 is hereby amended to read as follows:
5 a
s (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 86, are hereby incorporated in the license. The licensee 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 Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 30, 1985
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- - i i
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ATTACHMENT TO LICENSE AMENDMENT NO. 86 FACILITY OPERATING LICENSE N0. DPR-30 l
DOCKET NO. 50-265 i
Revise the Technical Specifications by deleting the following pages and inserting the enclosed pages.
Remove Insert 3.1/4.1-2 3.1/4.1-2
^
3.1/4.1-2a 3.1/4.1-3 3.1/4.1-3 3.1/4.1-6 3.1/4.1-6 3.1/4.1-7 3.1/4.1-7 3.1/4.1-8 3.1/4.1-8 3.1/4.1-9 3.1/4.1-9
'3.1/4.1-10 3.1/4.1-10 3.1/4.1-12 3.1/4.1-12
-3.1/4.1-13 3.1/4.1-13 3.1/4.1-14 3.1/4.1-14 3.2/4.2-10 3.2/4.2-10 3.2/4.2-10a-3.2/4.2-14 3.2/4.2-14 3.2/4.2-14a 3.2/4.2-16 3.2/4.2-16 3.2/4.2-17 3.2/4.2-17 3.2/4.2-17a 3.3/4.3-3 3.3/4.3-3 Fig. 3.5-1 (5 pages)
Fig. 3.5-1 (6 pages) e O
1
QUAD-CITIES OPR-30 3.1 LIMITING W NDITIONS FOR OPERATION BASES h reactor protection system automatically initiates a reactor scram to:
- a. ' preserve h Integrity of the fuel cladding, j
b.
preserve the integrity of h primary system, and c.
minimize h energy which must be adsorbed and prevent criticality following a loss-of-coolant accident.
This specification provides % limiting conditions for operation necessary to preserve the ability of the system to tolerate single failures and still perform its intended function even during periods when instresment channels may be out of service because of maintenance. h n necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
The reactor protection system is of N dual channel type (reference SAR, Section 7.7.1.2).
h system is made up of two independent trip systems, each having two subchannels of tripping devices. Each subchannel has en input from at least one instrument channel which monitors a critical parameter.
The outputs of W subchannels are combined in a one-out-of-two-logic; i.e., an input sign.. on either one or both of the subchannels will cause a trip system trip. The outputs of h trip systems are arranged so h t a trip on both systems is required to produce a reactor scram.
This system meets W requirements of the IEEE 279 Standard for Nuclear Power Plant Protection Systems issued Septed er 13, 1966. The system has a reliability greater h n that of a two-out-of-three system and somewhat less h n that of a one-out-of-two system (reference APED 5179).
With h exception of h average power range monitor (APIDO and intermediate range monitor (IRM) channels, each subchannel has one instrument channel. When h minimum condition for operation on the i
nuder of operable instrument channels per untripped protection trip system is met, or if it cannot be met and h af fected protection trip system is placed in a tripped condition, h effectiveness of h protection system is preserved, i.e. h system can tolerate a single failure and still perform its intended function of scrammming h reactor. Three APIDI instrument channels are provided for each protection trip system.
APftt's #1 and #3 operate contacts in one subchannst, and APIDt's 82 and 93 operate contacts in h o h r subchannel. APIDt's 94 and 35, and 36 are arranged similarly in h o h r protection trip system. Each protection trip system has one more APitt than is necessary to meet h minimum number required per channel. This allows W bypassing of one APIDI per protection trip system for maintenance, testing, or calibration. Additional 1l54 channels have also been provided to allow for bypassing of one such channel.
The bases for h scram settings for h 1154, APIDI, high reactor pressure, reactor low water level, turbine control valve fast closure, and turbine stop valve closure are discussed in Specifications 2.1 and 2.2.
Pressure sensing of h drywell is provided to detect a loss-of-coolant accident and initiate h emergency core cooling equipment. The pressure-sensing Instrisnentation is a backup to the water-level instrumentation which is discussed in Specification 2.1.
A scram is provided at h same setting as W emergency core cooling system (ECCS) initiation to minimize h energy which must be acconnodated during a loss-of-coolant accident and to prevent the reactor from going critical following h accident.
Amendment No. 86 3.1/4.1-2 Ol54H
QUAD-CITIES DPR-30 The control rod drive scram system is designed so that all of h water which is discharged from h Reactor by a scram can be aera==adated in the discharge piping. A part of this system is an Individual instr oent volume for each of h south and north CRD accumulators. These two volismes and hir piping can hold in excess of 90 gallons of water and is N low point in N piping. No credit was taken for h se volines in h design of W discharge piping relative to the amount of water which must be ace % ted during a scram. During normal operations, h discharge volumes are empty; however, should elWr voline fill with water, h water discharged to h piping from W Reactor may not be accomodated which could result in slow scram times or partial or no control rod insertion. To preclude this occurrence, level switches have been Installed in both volumes which will alarm and scram W Reactor when h volismo remaining in eihr Instrument volume is approximately 40 gallons. For diversity of level sensing methods h t will ensure and provide a scram, both differential pressure switches and W rmal switches'have been incorporated into h design and logic of h system. The setpoint for h scram signal has been chosen on h basis of providing sufficient volume remaining to accomodate a scram even with 5 gpm leakage per drive into SDV. As indicated above, h re is sufficient volume in the piping to acca==adate h scram without impairment of the scram times or h amount of Insertion of the control rods. This function shuts h Reactor down while sufficient volume remains to area =nodate' h discharged water and precludes W situation in which a scram would be required but not be able to perform its function properly.
Amendment No. 86 3.1/4.1-2a Ol54H
QUAD-CITIES DPR-30 Loss of condensate vacuum occurs when W condenser can no longer handle heat input. Loss of condenser vacuum initiates a closure of the turbine stop vah,es and turbine bypass valves, which eliminates W heat input to h cundenser. Closure of W turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat flux. To prevent h cladding safety limit from being exceeded if this occurs, a reactor scram occurs on farbine stop valve closure. The turbine stop valve closure scram function alone is adequate to prevent h cladding safety limit from being exceeded in h I
event of a turbine trip transient with bypass clost-e.
h condenser low-vacuum scram is a backup to h stop valve closure scram and causes a scram before the stop valves are closed, h s h resulting transient is less severe. Scram occurs at 21 inches Hg vacuum, stop velve closure occurs at 20 inches Hg vacuum, and bypass closure at 7 inches Hg vacuum.
High radiation levels in h main steamline tunnel above h t due to the normal nitrogen and oxygen radioactivity are an Indication of leaking fuel. A scram is initiated whenever such radiation level exceeds seven times normal background. The purpose of this scram is to reduce h source of such radiation to N extent necessary to prevent excessive turbine contamination. Discharge of excessive amounts of radioactivity to N site environs is prevented by h air ejector off-gas monitors, which e.ause an isolation of h main condenser off-gas line provided h limit specified in Specification 3.8 is ioxceeded.
h main steamline isolation valve closure scram is set to sc am when h Isolation valves are 105 closed from full open. This scram anticipates h pressure and flux transient which would occur when N valves close. By scramming at this setting, h resultant transient is insignificant.
A reactor modo switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status (reference SAR Section 7.7.1.2).
Whenever h reactor modo switch Is In h Refuel or Startup/ Hot Stan ey position, the turbine condenser low-vacuum scram and main s+eamline isolation valve closure scram are bypassed. This bypass has been provided for flexibility during startup and to allow repairs to be made to h turbine condenser. While this bypass is in effect, protection is provided against pressure or flux increases by N high-pressure scram and APftt 155 scram, respectively, which are effective in this mode.
~
If h reactor were brought to a hot stoney condition for repairs to the turbine condenser, h main staamline isolation valves would be closed. No hypo h sized single failure or single operator action in this mode of operation can result in an unreviewed radiological release.
The manual scram function is active in all modes, hs providing for a manual means of rapidly inserting cortrol rods during all modes of reactor operation.
The llW4 system provides protection against excessive power levels and short reactor periods in h startup and intermodlate power ranges (reference SAR Section 7.4.4.2 and 7.4.4.3).
A source range monitor ($104) system is also provided to supply additional neutror. level information during startup but has no scram functions (reference SAR Section 7.4.3.2).
Thus h IIDI is required in the Refuel and Startup/ Hot Standby mo&s. In addition, protection is provided in this range by h APfti 155 scram as discussed in h bases a
for Specification 2.1.
In W power range the APftt system provides required protection (reference SAR Section 7.4.5.2).
Thus, h IIDI system is not required in W Run mode, h APftt's cover only the inte rmediate and power range, h IlW4's provide adequate coverage in h startup an(Intermediate range.
The high-reactor pressure, high-drywell pressure, reactor low water level, and scram discharge volume high level scrans are required for h Startup/ Hot Stan&y and Run modes of plant operation. h y are W refore required to be operational for h se modes of reactor operation.
h turbine condenser low vacuisa scram is required only during power operation and must be bypassed to i
star? up h unit.
Amendment No. M 86 3,if4,i_3 Ol54H
}
QUAD-CITIES DPR-30 to an out-of-limits input. This type of failure for analog devices is a rare occurrence and is detectable by an operator who observes het on signal does not track h ohr three. For purposes of analysis, it is assumed h t this rare failure will be detected within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The bistable trip circuit which is a part of W Group 2 devices can sustain unsafe failures which are revealed only on test. Therefore, it is necessary to test them periodically.
A study was conducted of h Instrumentation channels included in the Group 2 devices to calculate their-
' unsafe' failure rates. The analog devices (sensors and amplifiers) are predicated to have an unsafe failure rate of less h n 20 X 104 failures / hour. The bistable trip circuits are predicated to have 4
an unsafe failure rate of less than 2 X 104 failures / hour. Considering h 2-hour monitoring interval for h analog devices as assumed above and a weekly test interval for the bistable trip circuits, h design reliability goal of 0.99999 is attained wity ample margin.
h bistable devices are monitored during plani operation to record hir failure history and establish a test interval using the curve of Figure 4.1-1.
There are n eerous identical bistable devices used
%roughout h plant instroentation system, h refore, significant data on h failure rates for N bistable devices should be - =alated rapidly.
The frequency of calibration of h APIDI flow blasing network has been established at each refueling outage. The flow blasing network is functionally tested at least once per month and, in addition, cross calibration checks of flow input to W flow-blasing network can be made during h functional test by direct meter reading (IEEE 279 Standard for Bluclear Power Plant Protection Systems, Section 4.9, Septen6er 13, 1966). There are several instruments which must be calibrated, and it will take several days to -
perform h callbration of h entire network. While h calibration is being performed, a zero flow signal will be sent to half of h APIW4's, resulting in a half scram and rod block condition. Thus, if h calibration were performed during operation, flux shaping would not be possible. Based on experience at o h r generating stations, drift of instrument such as those in h flow biasing network is not significant; h refore, to avoid spurlocs scrams, a calibration frequency of each refueling outage is established.
Reactor low water level Instruments 2-263-57A, 2-263-578, 2-263-58A, and 2-263-588 have been modified to be an analog trip system. The analog trip system consists of an analog sensor (transmitter) and a l
master / sieve trip unit setup which uttimeteIy drivos a trip reIay. The frequency of calibration and functional testing for instrument loops of h analog trip system, including reactor low water level, has been established I'n Licensing Topical Report IIEDO-21617-A (hr 1978). With h one-out-of-two-taken-twice logic, IIEDO-21617-A states h t each trip unit be subjected to a calibration / functional test frequency of one mon %. An adequate calibration / surveillance test interval for h transmitter is once per operating cycle.
- Group 3 devices are active only during a given portion of h operational cycle. For example, h Iltt is
]
active during startup and inactive during full-power operation. Thus, h only test h at is meaningful is h one performed just prior to shutdown or startup, i.e., h tests h t are performed Just prior to use of h Instrument.
Calibration frequency of h instreont channel is divided into two groups. These are as follows:
1.
passive type indicating devices that can be compared wi n like units on a continuous
- basis, and 2.
vacuum tube or semiconductor devices and detectors h t drift or lose sensitivity, i
i Amendment No. 86 3.1/4.1-6 Ol53H
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QUAD-CITIES DPR-30 Experience with passive type instroents in Commonwealth Edison generating stations and substations Indicates h t the specified calibrations are adequate. For N se devices which employ amplifiers, etc.
I drift specifications call for drift to be less h n 0.4%/ month I.e.,
in h period of a month a drift of 0.4% would occur, thus providing for adequate margin.
The sensitivity of LPfM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. Changes in power distribution and electronic drift also require cogensation. This compensation is accoglished by calibrating h APIM system every 7 days using heat balance data and by calibrating individual LPfM's at least every 1000 equivalent full-power hours using TIP traverse data.
Calibration on this frequency assures plant operation at or below thermal limits.
A comparison of Tables 4.1-1 and 4.1-2 indicates that some instrument channels have not been included in h latter table. These are mode switch in shutdown, manual scram, high water level in scram discharge vol me, main steamline isolation valve closure, turbine control valve fast closure, and turbine stop valve closure. All of h devices or sensors associated with these scram functions are simple on-off switches, hence calibration is not applicable, i.e., h switch is either on or off. Firther, h se switches are mounted solidly to W device and have a very low probability of moving; e.g., h Wrmal switches in the scram discharge volme tank. Based on h above, no calibration is required for h se instrument channels.
B.
The MFLPD shall be checked once per day to determine if h APfm scram requires adjustment. This may norwelly be done by checking h LPfm readings, TIP traces, or process computer calculations. Only a small number of control rods are moved daily, thus h peaking factors are not expected to change significantly and a daily check of h MFLPD is adequate.
Refersaces I.
- 1. M. Jacobs, " Reliability of Engineered Safety Features as a Function of Testing Frequency", Nuclear Safety, Vol. 9, No. 4, pp. 310-312, July - August, 1968.
2.
Licensing Topical Report NEDO-21617-A (December 1978).
O e
I Amendment _NO.J, 86 3.1/4.i_7 Ol53H
QUAD-CITIES OPR-30 TAE.E 3.I-1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS REFUEL MODE Minimum Nunber cf Operable or Tripped Instrument Channels pe Trio System l)
Trio Function Trio Level Settina Action (2)
I b de Switch in shutdown A
I Manual scram A
Ilm 3
High flux
$ 120/125 of full scale A
3 Inoperative Appg(3) 2 High flux (155 scram)
Specification 2.1.A.2 A
2 Inoperative A
2 (per bank)
High water level in scram
$ 40 gallons per bank A
discharge volume (4) 2 High-reactor pressure
$ 1060 psig A
2 High-drywell pressure (5)
$ 2 psig A
2 Reactor low water level 1 8 inches (8)
A 2
Turbine condenser low 121 inches Hg vacuun A
vacuun(7) 2 Main steamline high
$ 7 X normal full power A
radiation (123 background 4
Main steamline Isolation
< 105 valve closure A
valve closure (7)
Amendment No FI, 86 3.1/4.1-8 Ol54H
J QUAD-CITIES OPR-30 TABLE 3.1-2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS STARTUP/ HOT STANDBY MODE Minlaun Number cf Operable or Tripped Instrument ChanneIs per Trio System (l) -
Trio Function Trio Level Settina Action (2) 1-Mode Switch in shutdown A
1 Manual scram A
Ilm 3
High flux
$ 120/125 of full scale A
3 Inoperative A
Appg(3) 2
. High flux (155 scram)
Specification 2.1.A.2 A
2 Inoperative A
2 High-reactor pressure
$ 1060 psig A
2 High-drywell pressure (5)
$ 2 psig A
2 Reactor low water level 1 8 inches (8) g l
- 2 (per bank)
High water level in scram
$ 40 gallons per bank A
discharge volume (4) 2 Turbine condenser low 1 21 inches Hg vacuum A
vacuun(7) 2 Main steamIIne high
$ 7 X normal full power A
radiation (12) background 4
Main steamline isolation
< 105 valve closure A
valve closure (7)
~
Amendment No.JAT, 86 3, l f4, g _9 Ol54H
QUAD-CITIES-DPR-30 TABLE 3.1-3 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS RUN MODE I
Minim e Number cf Operable or
. Tripped Instro ent ChanneIs per Trip System (l)
Trio Function Trio Level Settina Action (2)
I Mode Switch in shutdown A
I Manuel scram A
m (3) 2-High flux (flow biased)
Specification 2.1.A.l A or B 2
Inoperative A or B 2
Downscale(II) t 3/125 of full scale A or B '
2 High-reactor pressure
$ 1060 psig A
2 High-drywell pressure
$ 2 psig A
2 Reactor low water level 1 8 inches (8) g l
2 (per bank)
High-water level in scram 5 40 gallons per bank A
l
_ discharge volume f
(
2 Turbine condenser low 1 21 inches Hg vacuum A or C ~
r vacuum 2
Main steamline high
$ 7 X normal full power A or C radiation (12) background I,
4 Main steamline isolation
$ 105 valve closure A or C l.
valve closure (6) l 2
Turbine control valve fast
> 405 turbine / nerator A or C closure (9) load mismatch 0) 2 Turbine stop valve
$ 105 valve closure A or C closure (9)
L 2
Turbine EHC control fluid 1 900 psig A or C low pressure (9)
AmendmentNo.[,86 3,1f4,i_lo Ol54H
l QUAD-CITIES DPR-30 i-TABLE 4.1-1 SCRAM INSTRUMENTATION AND LOGIC SYSTEMS FUNCTIONAL TESTS P:NIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTATION, LOGIC SYSTEMS, AND CONTROL CIRCulTS Instrument Channel Grouo(3)
Functional Test (7)
Minimum Freauencvi4)
Mode switch in shutdown A
Place mode switch in Each refueling outage shutdown Manual scram A
Trip channel and alarm Every 3 months l
litt l
High flux C
Trip channel and alarm (5)
Before each startup and l
weekly during refueling (6) l Inoperative C
Trip channel and alarm Before each startup and weekly during refueling (6)
APfti High flux B
Trip output releys(5)
Once each week inoperative B
Trip output relays Once each week i-Downscale B
Trip output relays (5)
Once each week High flux 155 C
Trip output relays (5)
Before each startup and weekly during refueling (6)
High reactor pressure A
Trip channel and alarm (1)
~
High drywell pressure A
Trip channel and alarm (1)
Reactor low water love,1(2)
B (81 (1)
{
High water level in scramO)
A Trip channel and alarm Every 3 months discharge voluna (thermal and dp switches)
Turbine condenser low vacuum A
Trip channel and alarm (1)
Main steamline high radiation (2)
B Trip channel and alarm (5)
Once each week Main steamline Isolation valve A
Trip channel and alarm (1) closure Turbine control valve fast A
Trip chansi and alarm
( l')
(
closure
{
Turbine stop valve closure A
Trip channel and alarm (1)
Turbine EHC control fluid low pressure A
Trip channel and alarm (1) 3.1/4.1-12 Amendment No. 86 Ol53H
QUAD-CITIES DPR-30 TABLE 4.1-1 (Cont'd)
I tiotes:
5 1.
Initially once per month until exposure hours (M as defined on Figure 4.1-1) are 2.0 X 10 ;
W reafter, according to Figure 4.1-1 with an interval not less than 1 month nor more than 3 months.
The compilation of instrument failure rate data may includa data obtained from o h r bolling water reactors for which h same design instrument operates in an environment similar to that of Quad-Cities Units I and 2.
2.
An Instrument check shall be performed on low reactor water level once per day and on high steamline radiation once per shift.
3.
A description of h three groups is included in the bases of this specification.
4.
Functional tests are not required whea N systems are not required to be operable or are tripped. if tests are missed, Ny shall be perfwhed prior to returning h systems to an operable status.
5.
This instrumentation is exempted from h Instrisment functional test definition (1.0 Definition F).
This instrtsment functional test will consist of injecting a r.imulated electrical signal into the measurement channels.
6.
Frequency need not exceed weekly.
l 7.
A functional test of h logic of each channel is performed as indicated. This coupled with placing h mode switch in shutdown each refueling outage constitutes a logic system functional test of N scram system.
8.
A functional test of N master and slave trip units is required monthly. A calibration of the trip unit is to be performed concurrent with h functional testing.
9.
Only h electronics portion of W h rmal switches will be tested using an electronic calibrator during h three month interval test. A water column or equivalent will be used to test h dp switches.
O Amendment NO. 86 3,,f4,,,,3 Ol53H
QUAIN lTIES OPR-30 TABLE 4.1-2 SCRAM INSTRUMENT CAllBRATION MINIMUM CAllBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS i.
Instruent Channel' Group (I)
Calibration Standard (5)
Minimum Frequencv(2)
High flux litt C
Ccaparism to APRM after Every controlled heat balance shutdown (4) 1
.High flux APfti Output signal B
Heat balance Once every 7 days Flow bias B
Standard pressure and Refueling outage voltage source
-~
LPfti B(6)
Using TIP system Every 1000 e wivalent full power hours High reactor pressure A
Standard pressure seurce Every 3 months High drywell pressure A
Standard pressure source Every 3 months Reactor low water level B
Water level (7)
Ttrbine condenser low vacuum A
Standard vacuun source Every 3 months Main steamline high radiation B
Appropriate radiation Refueling outage source (3) l Tgrbine EHC control fluid A
Pressure source Every 3 months low pressure High water level in scram A
Water level Refueling Outage discharge volume (dp only)
Notes:
1.
A description of the three groups is included in the bases of this specification.
2.
Calibration tests are not required when the systems are not required to be operable or are tripped.
If tests are alssed, they shall be performed prior to returning the systems to an operable status.
3.
A current source provides an Instrument channel alignment every 3 months.
4.
Maximun calibration frequency need not exceed once per week.
5.
Response time is not part of the routine instrument check and calibration but will be checked every
~
refueling outage.
6.
Does not provide scram function.
7.
Trip units are calibrated monthly concurrently with functiona'l testing. Transmitters are calibrated l
once per operating cycle.
(
l i
Amendment NO. 86 3,,f4,,_,4 0153H
.~...
QUAD-CITIES DPR-30 Optimizing each channel Independently may not truly optimize h system considering the overall rules of system operation. However, true system optimization is a complex problem. The optimums are broad, not sharp, and optimizing h indiwIduaI channeIs Is general 1y adequate for h system.
The formula given above minimizes h unavailability of a single channel which must be bypassed during testing. The minimization of h unavailability is Illustrated by curve I of Figure 4.2-2, which assumes h t a channel has a failure rate of 0.1 X 10-6/ hour and 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is required to test it. The unavailability is a minimum at a test interval I, of 3.6 X 103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />.
If two similar channels are used in a one-out-of-two configuration, h test interval for minimun availability changes as a fu mtion of h rules for testing. The slaplest case is to test each one independent of h o h r.
Ir %Is case, h re is assumed to be a finite probability h t both may be bypassed at one time. This case is shown by curve 2.
Note ht h unavailability is lower, as expected for a redundant system, and h minimum occurs at h same test interval. Thus, if W two channels are tested independently, h equation above yields h test interval for minimen unavailability.
A more usual case is that h testing is not done Independently. If both channels are bypassed and tested at h same time, h result is shown in curve 3.
Note that h minimun occurs at about 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, much longer h n for Cases I and 2.
Also, h minimum is not nearly as low as Case 2, which indicates h t this method of testing does not take full advantage of h redundant channel. Bypassing both.
3 channels for simultaneous testing should be evolded.
The most likely case would be to stipulate that one channel be bypassed, tested, and restored, and h n lamediately following h second channel be bypassed, tested, and restored. This is shown by curve 4.
Note h t W re is not true miniewn. The curve does have a definite knee, and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by N equation for a single channel.
The best test procedure of all h se examined is to perfectly stagger W tests. This is, if h test interval is 4 months, test one of h o h r channels every 2 months. This is shown in curve 5.
The difference between Cases 4 and 5 is negligible. There may be o h r arguments, however, h t more strongly support h perfectly staggered tests, including reductions in human error.
The conclusions to be drawn are h se:
a.
A one-out-of-n system may be treated h same as a single channel in tenas of choosing a test Interval.
b.
More N n one channel should not be bypassed for testing at any one time.
Reactor water level instruments 2-263-73A & B, HPCI high steam flow Instruments 2-2389A-D, and HPCI steam line low pressure instruments 2-2352 & 2353 have been modified to be analog trip systems. The analog trip system consists of an analog sensor (transmitter) and a master / slave trip unit setup which ultimately drives a trip relay. The frequency of calibration and functional testing for instrument loops of the analog trip system has been established in Licensing Topical Report NEDO-21617-A (December 1978). With h one-out-of-two-taken-twice logic, NEDD-21617-A states that each trip unit be subjected to a calibration / functional test frequency of one month. An adequate calibration / surveillance test interval for h transmitter is once per operating cycle.
The radiation monitors in h ventilation duct and on h refueling floor which Initiate building isolation and stoney gas treatment operation are arranged in two one-out-of-two logic systems. The bases given above for h rod blocks apply here also and were used to arrive at h functional testing frequency.
Amendment No. 86 3.2/4.2-10 Ol53H 1
~
QUAD-CITIES DPR-30 Beses on experience at Dresden Unit I with Instruments of similar design, a testing interval of once every 3 months has been found to be adequate.
1 The automatic pressure relief instrumentation can be considered to be a one-out-of-two logic system, and h discussion above applies to it also.
The instrumentation which is required for N postaccident condition will be tested and calibrated at regularly scheduled intervals. The basis for h calibration and testing of this instrumentatin. Is h same as was discussed above for h reactor protection system and W emergency core cooling systems.
References 1.
- 8. Epstein and A. Shiff, " Improving Availability and Readiness of Field Equipment Through Periodic Inspection", UCRL-50451, Laurence Radiation Laboratory, p. 10, Equation (24), July 16, 1968.
3.2/4.2-10.
Amendment No. 86 Ol53H
QUAD-CITIES DPR-30 INSTRUMENTATION THAT INITIATES ROD BLOCK Minimum Number of Operable or Tripped Instrument Channels'oer Trio System (II Instrument Trio Level Setting
'2 APRM' upscale (flow bias)(7) 1 (0.58WD + 50)
FRP (2)
MFLPD 2
APRM upscale (Refuel and i 12/125 full scale Startup/ Hot Standby mode) 2 APRM'downscale(7) 1 3/125 full scale 1
Rod block monttor upscale 1 0.65WD + 42 (2)
(flow bias)(7) 1 Rod block monitor downscale(7) 1 3/125 full scale
'3 IRM downscale(3) (8) 1 3/125 full scale 3
IRM upscale (8) i 108/125 full scale 2(5)
SRM detector not in Startup 12 feet below core center line position (4) 3 IRM detector not in Startup 1 2 feet below core center line position (8) 5 2(5)(6)
SRM upscale i 10 counts /sec 2(5)
SRM downscale(9) 1 10 counts /sec 2
1 (per. bank)
High water level in scram i 25 gallons (per bank)
V discharge volume (SDV) 1 SDV high water level scram NA trip bypassed
' NOTE:
6 1,.For the Startup/ Hot Standby and Run positions of the reactor mode selector switch, there shall be two operable or tripped trip systems for each function except the SRM rod blocks. IRM upscale and IRM downscale need not be operable in the Run position. APRM downscale,- APRM upscale (flow biased), and RBM downscale need not be operable in the Startup/ Hot Standby mode. The RBM upscale need not be operable at less than~307, rated thermal power. One channel may be bypassed above 307 rated thermal power provided that a. limiting control rod pattern does not exist. For systems with more than one channel per trip 1,
if the first column cannot be met for one of the two trip systems, this condition may exist for up to 7 days provided that during that time the operable system is functionally tested intnediately and daily thereafter; if this condition lasts longer than 7 days the system shall be tripped. If the first column
'cannot be met for both trip systems. the systems shall be tripped.
Amendment No. 86 3.2/4.2-14 0154H
QUAD-CITIES DPR-30 is h percent of ' rive flow required to produce a rated core flow of 98 million Ib/hr. Trip level 2.
W d
D setting is in percent of rated power (2511 left).
3.
IM downscale may be bypassed when it is on its lowest range.
4.
This function is bypassed when the count rate is > 100 eps.
5.
One of W four S m inputs may be bypassed.
6.
This'SM function may be bypassed in h high IM ranges (ranges 8, 9, and 10) when the IM upscale rod block is operable.
- 7.
Not required to be operable when perfoming low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.
8.
This IM function occurs when h reactor modo switch is in W Refuel or Startup/ Hot Standby position.
9.
This trip is bypassed when h SM is fully inserted.
d
- O i
Amendment No. 86 3.2/4.2-14.
Ol54H 4
. ;c g;.
- ig <
~
~
QUAD-CITIES?
DPR-30 TABLE 4.2-1 MINIMUM TEST AND CALIBRATION FREQUENCY FOR CORE AND
~
j '
CONTAINMENT COOLING SYSTEMS INSTRUMENTATION R00 BLOCKS. AND'ISOLATIONSI7)~
Instrument Functional Instrument Channel I311(2)
.Calib' ration (2)
Instrument Check (2)
TECCS Instrumentation T.
Reactor low-1cw water' level' (1)
Once/3 months Once/ day 2.
Drywell-h?gh pressure:
(1)
Once/3 months None
- 3. : Reactor low pressure (1)
-.Once/3 months None
.4
. Containment. spray interlock
- a.-
l 2/3 core height -
- (1) (10)
('10 ) -
None
.b.
Containment pressure' (1)
-Once/3 months None
.5.
Low-pressure core cooling
.(1)
Once/3 months-None
, u. - ; +
pump. discharge:
['
- 6.
Undervoltage 4-KV essential Refueling outage Refueling outage None
- 7.
Degraded voltage Refueling outage (8). Refueling outage Once/ month -
.4-KV essential busses.
Rod Blocks 1.
APRM downscale
.(1)'(3)
Once/3 months None
- 2..APRM flow variable.
(1) (3)
Refueling outage None
- 3.
IRM upscale -
(5) (3)
(5) (3)
None
- 4..IRM downscale (5) (3)
(5) (3)
None
-5.
RBM upscale ~
(1) (3)
Refueling outage None
- 6.
RBM downscale-(1) (3)
-Once/3 months None
- 7..SPM upscale
.(5) (3)
(5)'(3)
None
- 8..SRM detector not in.startup (5) (3)
(6)
None position
-9.
IRM detector not'in.startup (5).
(6)
None
-position
- 10. SRM downscale
-(5) (3) -
(5) (3)
None-
-11..High water level in scram Once/3 months Not applicable None discharge volume (50V)-
- 12. SOV high level trip Refueling outage Not applicable None 7
- bypassed.
-Main Steamline Isolation s
- 11. : Steam tunnel high temperature Refueling outage Refueling outage None L2. ~ Steam 11ne high flow (1)
Once/3 months Once/ day
- 3.. Steam 11ne' low pressure (1).
Once/3 clonths None
-4.
Steamline high radiation (1) (4)
Refueling outage Once/ day
- 15. iReactor. low low water level (1) (10)
(10)
Unce/ day-
.RCIC' Isolation-
[1. Steam 1tne high flow Once/3 month > (9)
Once/3 months (9)
None
- 2. ' Turbine area high temperature
' Refueling outage Refueling outage None
.Once/3 months Once/3 months None 3.
Low reactor: pressure' Amendment:No. 86 3.2/4.2-16 L0153H
1
=
QUAD-CITIES OPR-30 TABLE 4.2-1 (Cont'd)
Instrument Functional IDstrument Channel Ig11(2)
Calibration (2)
Instrument check (2)
HPCI Isolation
.1. 'Steamline high flow (1) (9) (10)
(9) (10)
Ncne 2.
Steamline' area high temperature Refueling outage Refueling outage None 3.
Low reactor pressure (1) (10)
(10)
None-
' Reactor Building Ventilation ~$ystem Isolation and SBGTS Initiation
~ - '
1.
Refueling floor radiation monitors (1)
Once/3 months Once/ day
' Control Room Ventilation System Isolation 1.
Reactor low water level-(1)
Once/3 months Once/ day.
-2.
Drywell high pressure (1)
Once/3 months None
- 3. - Main steamline high flow (1)
Once/3 months Once/ day Notes:
5 1.
Initially.once per month untti exposure hours (M as defined on Figure 4.1-1) are 2.0 X 10 ;
thereafter, according to Figure 4.1-1 with an interval not less than 1 month nor more than 3 months.
The compilation of instrument failure rate data may include data obtained from other butling water reactors for which the same design instrument operates in an environment similar to that of Quad-Cities Units 1 and 2.
. 2.
Functional tests.. calibrations, and instrunent checks are not required when these instruments are not required to be operable, or are tripped.
i
- 3..This instrumentation is excepted from the functional test definition. The functional test shall consist of injecting a simulated electrical signal into the measurement channel.
~
4.
This instrument channel is excepted from the functional test definitions and shall be calibrated using simulated electrical signals once every 3 months.
5.
Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibrations shall be performed during each startup or during controlled shutdowns with a required frequency not to exceed once per week.
Amendment No. 86 3.2/4.2-17 I
0153H I
QUAD-CITIES DPR-30 Notes: (Cont) 6.
The positioning mechanism shall be calibrated every refueling outage.
7.
Logic system functional tests are performed as specified in W applicable section for h se systems.
8.
Functional tests shall include verification of operation of h degraded voltage. 5 minute timer and 7 second inherent timer.
9.
Verification of h time delay setting of 3 $ T $ 10 seconds shall be performed during each refueling outage.
- 10. Trip units are functionally tested monthly. A calibration of h trip unit is to be performed concurrent with h functional testing. Transmitters are calibrated once per operating cycle.
)
Amendment No. 86 3.2/4.2-17e Ol53H
QUAD-CITIES OPR-30 3.
h control rod drive housing support 3.
The correctness of h control rod system shall be in place during reactor withdrawal sequence input to N Rlm power operation and when h reactor computer shall be verified after loading coolant system is pressurized above h sequence.
atmospheric pressure with fuel in h reactor vessel unless all control rods Prior to h start of control rod I
are fully inserted and Specification withdrawal towards criticality, h 3.3.A.I is met.
capability of h rod worth minimizer.to properly fulfill its function shall be l
a.
Control rod withdrawal sequences verlflod by h following checks:
shall be established so ht maximun I
reactivity h t could be added by a.
The RbM computer on line diagnostic dropout of any increment of any one test shall be successfully performed.
control blade would be such h t h rod drop accident design limit of 280 b.
Proper annunciation of h selection cal /gn. Is not exceeded.
error of one out-of-sequence control rod shalI be verifiod.
b.
Whenever h reactor is in h Startup/ Hot Stand >y or Run mode below c.
The rod block function of h RtM 205 rated h rmal power, h rod shall be verified by withdrawing h worth minimizer shall be operable. A first rod as an out-of-sequence second operator or quellflod control rod no more than to h bloc'k technical person may be used as a point.
substitute for an inoperable rod worth minimizer which falls after 4.
Prior to control rod withdrawal for withdrawal of at least 12 control startup or during refueling, verify that rods to h fully withdrawn a least two source range channels have an position. h rod worth minimizer observed count rate of at least three may also be bypassed for low power counts per second.
physics testing to demonstrate h shutdown margin requirements of 5.
h n a limiting control rod pattern Specification 3.3.A If a nuclear exists, an instrument functional test of, engineer is present and verifles h h IEIM shall be performed prior to step-by-step rod movements of N withdrawal of N designated rod (s) and test procedure.
delly thereafter.
4.
Control rods shall not be withdrawn for 6.
The scram discharge volume vent and drain startup or refueling unless at least two valves shall be verified open at least source range channels have an observed once per 31 days. Theses valves may be count rate equal to or greater than three closed Intermittently for testing under counts per seconds and h se SFN s are administrative control and at least once fully inserted.
per 92 days, each valve shall be cycled through at least one complete cycle of 5.
During operation with Ilmiting control full travel. At least once each rod patterns, as determined by the Refueling Outage, h scram discharge nuclear engineer, olhrt volume vent and drain valves will be demonstrated tot a.
both IEIM channels shall be operable.
s.
Close within 30 seconds a,fter receipt l b.
control rod withdrawal shall be of a signal for control rods to blocked; or scram, and b.
Open when h scram signal is reset.
3*3'4*3-3 Amendment No.ff,)f, 86 Ol55H
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Generation Rate (MAPLHGR) l vs. Planar Average Exposure Amendment No. Jf, 69, 86
~
4UAD UrtT e
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l E~- '! :H I 'a i g u O 10,000 20,000 30,000 ho,000 Planar Average Exposure (MWD /ST) Figure 3 5-1 Maximum Average Planar Linear Heat Generation (Sheet 6 of 6) Rate (MAPLHGR) vs. Planar Average Exposure Amendment No. 86}}