ML20090A411
| ML20090A411 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/21/1992 |
| From: | Barrett R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20090A412 | List: |
| References | |
| NUDOCS 9203020323 | |
| Download: ML20090A411 (11) | |
Text
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10WA-!LLIN015 CAS AND ELECTRIC COMPANY D R ET NO 50 265 OUAD CITl[S NUCLEAR POWER STATION, t! NIT _2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 130 License No. DPR.30 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Commonwealth Edison Company (thelicensee)datedJune 28, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will o>erate in conformity with the application, the provisions of tie Act, and the rules and regulations of the Comission; C.
Thereisreasonableassurance(i)thattheactivitiesauthorized 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 Comission'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 Comission'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:
9203020323 920221 ADOCK0500ggS DR
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B.
T_echnical Specifications The Technical Specifications contained in Appendices A and D, as revised through Amendnient No. 130, are hereby incorporated in the lictnse. The licensee shall operate the f acility in accordance with the Technical Specifications.
3.
This license amendnent is etfettive innediately, to be implemented during the eleventh refueling outage.
FOR THE NUCLEAR REGULATORY C0littl5510N
- C I;ilhardJ.Barrett, Director Project Directorate 111-E Division of Reactor projects - lil/1V/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuence: February 21, 1992
.._ _.m ATTACHMENT TO LICENSE AMENDMENT NO. 130 FACILITY OPERATING LICENSE NO. DPR-30 DOCKET NO. 50-265 Revise the 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 marginal lines indicating the area of change.
REMOVE INSERT 3.2/4.2-6a 3.2/4.2-6a 3.2/4.2-7 3.2/4.2-7 3.2/4.2-8 3.2/4.2-8 3.2/4.2-11 3.2/4.2-11 3.2/4.2-11a 3.7/4.7-21 3.7/4.7-21 3.7/4.7-21a 3.7/4.7-22 3.7/4.7-22
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QUAD CITIES OPR 30 The RCIC and the HPCI high flow and temperature instrumentation are provided to detect a break in their respective piping. A trip of this instrumentation results in closure of the RCIC or HPCI steam supply isolation va:ves. The trip logic for this function is similar to that for the main steamline isolation valves, thus all sensors are required to be operable or in a tripped condition to meet single failure criteria. The trip settings of 170'F and 300%
of design flow and valve closure time are such that core uncovery is prevented and fission produri release is within limits. In addition, the steau supply valves for nach systsn, are closed on low steamline pressure to provide primary containment isolation when the reactor pressure, as sensed n the system steamlines, is below the required pressure for turbine operation.
Operation of the HPCI turbine will continue as long as reactor pressure is above 150 psig. When the reactor pressure falls below 150 psig, the s 30ed of the turbine-puma unit will decrease and would gradually be slowed c ue to stop friction and w ndage losses at low reactor pressures. The low reactor pressure isolation setpoint was developed in accordance with NEDC 31336, General Electric Instrument Setpoint Methodology," dated October,1986.
The trip setpoint of greater than or equal to 100 psig was calculated soch that the isolation will occur on decreasing reactor pressure to provide primary containment isolation when the reactor pressure, as sensed in the system steamlines, is below the required ores:ure for turbine operation. The external vacuum breaker line for 11e HPCI turbine will isolate on low steamline pressure concurrent with high drywell pressure signals. The instrumentation and controls ensure the pro 3er HPCI and primary containment response to a HPCI steamline areak (isolation of the steamline supply velves only), a large break inside the containment (closure of the steam su 3 ply and vacuum relief isolation valves) and a small or intermediate size brea< inside of containment (steam supply and vacuum breaker isolation valves remain open for HPCI operation).
The instrumentation which initiates ECCS action is arranged In a one-out of-two taken twice logic circuit. Unlike the reactor scram circuits, however, there is one trip system associated with each function rather than the two trip systems in the reactor protection system. The sin 0 e fallure l
criteria are met by vidue of the fact that redundant core cooling functions are provided, e.g., sprays and automatic blowdown and high pressure coolant i
l injection. The specification requires that if a trip system becomes inoperable, the system which it activates is declared inoperable. For example, if the trip system for core spray A becomes inoperable, core spray A is declared inoperable and the out-of service specifications of Specification 3.5 govern.
This specification preserves the effectiveness of the system with respect to the single f ailure criteria even during periods when maintenance or testing is being performed.
The control rod block functions are provided to prevent excessive control rod witdrawal so that MCPR does not go below the MCPR Fuel Cladding integrity Safety Limit. The trip logic for this function is one out of n; e.g., any trip on one of the six APRM's, eight IRM's, four SRM's will result in a rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure that the single-f alkne criteria are met. The 3.2/4.2 6a Amendment No. 130
QUAD CITIES DPR-30 minimum instrument channel requirements for the RBM may be reduced by one for a short eriod of time to allow for maintenance, testing, or calibration. This time period is on! ~3% of the operating t!me in a month and does not significantly increase the risk o preventing an inadvertent control rod withdrawal.
The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at reduced flow. The APP 4 p.ovides gross core protection,i.e., limits the gross withdrawal of uutrol rods in il e wa mal withdrawal t
- sequence, in the refuel and startup! hot standby modes, the APRM rod block function is set at 12% of rated power. This control rod block provides the same type of protection in the Refuel and Startup/ Hot Standby modes as the APRM flow.blased rod block does in the Run mode, i.e., prevents control rod withdrawal before a scram is reached.
The RBM rod block function provides local protection of the core, i.e., the prevention of transition bolling in a local region of the core for a single ro,1 withdrawal error from a limiting control rod pattern. The trip point is flow biased. The worst case single control rod withdrawal error is analyzed for each reload to assere that, with the s 3ecific trip settings, rod withdr,.valis blocked before the MCPR reaches the fuel c adding integrity safety limit.
Below 30% power, the worst-case withdrawal of a single control rod without rod block action will not violate the fuel cladding integrity safety limit. Thus the RBM rod block function is not required below this power level.
The IRM block function provides local as well as gross core protection. The scaling arrangement m such that the trip setting is less than a factor of 10 above the indicated level. Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR fuel cladding integrity safety limit.
A downscale indication on an APRM is an indication the instrument has failed or is not sensitive enough, in either case the instrument will not respond to changes in control rod motion, and the control rod motion is thus prevented. The downscale trips are set at 3/125 of full scale.
The SRM rod block with s 100 CPS and the detector not full inserted assures that the SRM's are not withdrawn from the core prior to commencing rod withdrawal for startup. The scram discharge volume high water level block provide annunciation for o aerator action. The alarm setpoint has been selected to provide adequate time to a low determination of the cause of level increase and corrective action prior to automatic scram initiation.
For effective emergency core cooling for small pipe breaks the HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. T.% automatic pressure relief function is provided as a backup to the HPCl in the event the HPCI does not operate. The arrangement 3.2/4.2-7 Amendment No. 130
~ -
- QUAD CITIES DPR 30 of the tripping contacts is such as to provide this function whnn necessary and minimize spurious operation. The trip settings given in the specification are -
adequate to assure the above criteria are met (reference SAR Section 6.2.6.3). The specification preserves the effectiveness of the system during periods of maintenance, testing or calibration and also minimizes the risk of inadvertent operation, i.e., only one instrument channel out of service.
Two radiation monitors are provided on the refueling floor which initiate isolation of the reactor building and operation of the standby gas treatment systems. The tria logic is one out of two. Trip settings of s100 mR/hr for the monitors on the refue'ing floor are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none of the activity released during the refueling accident leaves the reactor build 5q via the normal ventilatinn stack but that all the activity is processed by the sta,% gas treatment system.
The instrumentation which is provided to monitor the postaccident condition is listed in Table 3.2 4. The instrumentation listed and the limiting conditions for operation on these systems ensure adequate monlioring of the containment following a loss-of-coolant accident. Information from this instrumentation will provide the operator with a detailed knowledge of the conditions resulting from the accident; based on this information he can make logical decisions regarding postaccident recovery.
The specifications allow for postaccident instrumentation to be out of service for a period of 7 days. This period is based on the fact that several diverse instruments are available for guiding the operator should an accident occur, on the low probability of an instrument being out of service and an accident occurring in the 7-day period, and on engineering judgment.
The normal supply of air for the control room ventilation system Trains "A" and *B" is outside the service building. In the event of an accident, this source of air may be required to be shut down to prevent high doses of radiation in the control room.
Rather than provide this isolation function on a radiation monitor installed in the intake air duct, signals which indicate an accident, i.e., high drywell pressure, low water level, main streamline high flow, or high radiation in the reactor building ventilation duct, will cause Isolation of the intake air to the control room. The above trip signals result in immediate isolation of the control room ventilation system and thus minimize any radiation dose. Manualisolation capability is also provided.
Isolation from high toxic chemical concentration has been added as a result of the
" Control Room H_ abitability Study" submitted to the NRC in December 1981 in response to NUREG 0737 Item lll D.3.4. As explained in Section 3 of this study, ammonia, chlorine, and sulphur dioxide detection capability has been provided. The setpoints chosen for the control room ventilation isolation are based on early detection in the outside air supply at the odor threshold, so that the toxic chemical will not achieve toxicity limit concentrations in the Control Room.
The radioactivie liquid and gaseous effluent instrumentation is provided to monitor the re' ease of radioactive matorials in liquid and gaseous effluents during releases.
The alarm setpoints for the instruments are provided to ensure that the alarms will occur prior to exceeding the limits of 10 CFR 20, 3.2/4.2-8 Amendment No.130 1
QUAD-C! TIES DPR-30 TABLE 3.2-1 INSTRUMENTATION THAT.N!TIATES PRIMARY CONTAINMENT ISOLATION FUNCTIONS Minimum Number of Operable or Tripped ChannelsL}.1 Instrumen Instruments Trip Level Setting Ac tionL2]
4 Reactor low wattr[5l
>144 inches above top of A
active fuel' 4
Reactor low low water 184 inches above top of A
active fuel
- 4 High drywell pressure [5]
12.5 psig [3]
A 16 High flow main steamitne[5]
1 40% of rated steam flow B 1
16 High temperature main 1200* F B
steamline tunnel 1 5 x normal rated power B
1 4
High radiation m31n steamline tunnelL63 background (without hydrogen addition) 4 Low main steam pressure [4]
1825 psig B
2 High flow RCIC steamline
<300 % of rated steam C
Tiow(7) 4 RCIC turbine area high 1170' F C
temperature 2
High flow HPCI steamitne
<300% of rated steam D
Ylow(7) 4 HPCI area hlgh temperature 1170* F 0
1 00 pstg 0
1 4
HPCI Steamitne pressure 3.2/4.2-11 Amendment No. 130
___.__._m_._
v.
QUAD-CITIES OPR..
TABLE 3.2-1 (Cont.)
INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION FUNCTIONS Notes 1..
Whenever_ primary containment integrity is required, there shall be two operable or tripped systems for each function, except for low pressure main steamline which only need be available in the Run position.
- 2. - Action, if the first column cannot be met for one of the trip systems..
that trip system shall be tripped.
If the first column cannot be met for both trip systems, the appropriate actions listed below shall. be taken.
A.
Initiate an orderly shutdown and have the reactor in Cold Shutdown condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an' orderly load reduction and have reactor tn Hot Standby within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Close isolation valves in RCIC system.
D.
Close _ isolation valves in HPCI subsystem.
4 3.
Need not be operable when primary containment integrity is not required.
4.
The isolation trip signal is bypassed when the mode switch is in Refuel or Startup/ Hot Shutdown.
5.
The instrumentation also isolates the control room ventilation system, 6.
This signal also automatically closes the mechanical vacuum pump discharge
~
line isolation _ valves.
7.
Includes'a time delay of 31 t 19 seconds.
- Top of active fuel:is. defined as 360" above vessel zero for all water levels
'used in the LOCA analysis-(see Bases 3.2).
'I 3.2/4.2-Ila Amendment No.130
QUAD-CITIES DPR-30 TABLE 3.7-1 (Cont'd)
Valve Number of Maximum Number for Power-Operated Operating Normal '
Action on Isolation Valve Units Valves Time Operating Initiating Group IdentifIcatinD 1 and 2 inh 0ELQath0Ed 11td EM1110D ilgaal Radwaste 2
Drywell floor drain discharge A0-2001-3 1
120 0
GC 2
Drywell floor drain discharge A0-2001-4 1
120 0
GC 2
Drywell equipment drain discharge A0-2001-15 1
120 0
GC 2
Drywell equipment drain discharge A0-2001-16 1
120 0
GC Nate: Valve can be reopened after isolation for sampling Oxygen Analyzer 2
Oxygen analyzer valve A0-8801-A, 4
110 0
GC B,C.D 2
0xygen analyzer valve A0-8802-A, 4
110 0
GC B.C.D 2
Oxygen analyzer valve A0-8803A 1
110 0
GC 2
Oxygen analyzer valve A0-8803B 1
110 0
GC i
l 3.7/4.7-21 Amendment No. 130
TABLE 3.7-1-(Cont'd)
Valve Number of Maximum Number for Power-Operated Operating Normal Action on Isolation
-Valve Units Valves Time Operating Initiating Gtaup identifica.tlon Land _2 Inknard_0ntticard 1 sed Eosition Signal Traversing Incore Probe 2
On isolation
-signal, the TIP Tip Ball detector is Valve 4
withdrawn if in 700-733 use: five ball valves and one nitrogen purge are closed.
Tip Purge Valve Assembly 700-743 Reactor Water Cleanup 3
Pump suction isolation valve M0-1201-2 1
130 0
GC 3
Pump suction isolation valve MO-1201-5 1
130 0
GC HPCI 4
Steam isolation valve K0-2301-4 1
150 0
GC 4
Steam isolation valve M0-2301-a 1
150 0
GC 4
Vacuum breaker isolation M0-2399-40 1
550 0
GC 4
Vacuum breaker isolation M0-2399-41 1
150 0
GC RCIC 5'
Turbine steam supply MO-1301-16 1
525 0
GC 5
Turbine steam supply MO-1301-17 1
125 0
GC-3.7/4.7-21a Amendment No, 130
QUAD-CITIES DPR-30 TABLE 3.7-1 (Cont'd)
Ley:
0:
spen C:
closed SC:
stays closed GC: goes closed Note:
Isolation groupings are as follows:
Group 1: The valves in Group 1 are closed upon any one of the following conditions:
1.
Reactor low-low water level 2.
Main steamline high radiation 3.
Main steamline high iiow 4.
Main steamline tunnel high temperature 5.
Main steamline low pressure Group 2: The actions in Group 2 are initiated by any one of the following conditions:
1, Reactor low water level 2.
High drywell pressure Group 3: Reactor low water level alone initiates the following:
1.
C'teanup demineralizer system isolation Group 4: The steam supply isolation valves in the high pressure coolant injection system (HPCI) are closed upon any one of the following signals:
1.
HPCI steamline high flow 2.
High temperature in the vicinity of the HPCI steamline 3.
Low reactor pressure The turbine exhaust vacuum breaker isolation valves close when both of the following signals are present (simultaneously):
1.
High drywell pressure 2.
Low reactor pressure Group 5: Isolation valves in the reactor core isolation cooling system (RCIC) are closed upon any one of the following signals:
1.
RCIC steamline high flow 2.
High temperature in the vicinity of the RCIC steamline 3.
Low reactor pressure 3.7/4.7-22 Amendment No.130 1
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