ML20070E037
| ML20070E037 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/21/1991 |
| From: | Barrett R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070E042 | List: |
| References | |
| NUDOCS 9103040167 | |
| Download: ML20070E037 (10) | |
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UNITED STATES i
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4 NUCLEAR REGULATORY COMMISSION li WASHINO TON, D. C. 20556 k.....,of COMMONWEALTH EDISON COMPANY AND IOWA-ILLINOIS GAS AND ELECTRIC COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR IOWER STATION, UNIT 1 AMENDMENT TO FACIt'TY OPERATING LICENSE Amendment No. 129 License No. DPR-29 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
.The ar-11 cation f'r arnendment by romonwealth Edison Company (th n'nsee)di.'r De a ber M, 1990, complies with the standards end yOements of he Atomic Energy Act of 1954, as -amended (the Ac,t} 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 puisions of the Act, and the rules and regulations of the Comis; ion; C.
There is reasonable-assurance (1) that the activities authorized by this emendment 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 comon defense and security or to the health and safety-of-the publie.; ar.d E.
The issuance of this amendment is in accordance with 10 CFR Part 51 i
of the Comission's-regulations and all applicable requirements.-
have been satisfied.
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Accordingly, the license is amended by changes to the Technical-
-Specifications as indicated in.the attachment to this license amendment, and paragraph 3.8. of Facility Operating License No. DPR-29 is i
hereby amended to read as follows:
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Technical Specifications The Technical Specifications contained in Ap)endices A and B as revised through Amendment No.1?9, are here >y 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 Richard arrett, Director Project Directorate 111/2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance:
February 21, 1991
4 ATTACHMENT TO LICENSE _ AMENDMENT NO. 129 FACILITY _0PERATING LICENSE NO. DPR-29 DOCKET NO. 50-2E4 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attache d pages. The revised paces are identified by the captioned muendment nuncer and contain marginal lines indicatirig the area of change.
REMOVE INSERT 1.1/2.1-4 1.1/2.1-4 1.1/2.1-15 1.1/2.1-15 3.1/4.1 -8 3.1/4.1-8 3.1/4.1-10 3.1/4.1-10 3.1/4.1-13 3.1/4.1-13 3.1/4.1-14 3.1/4.1-14 3.1/4.1-17 3.1/4.1-17 a
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QUAD CITIES DPR-29 C.
Power Transient C.
Renctor low water level scram setting shall be 144 inches above the top of 1.
The neutron flux shall not the active fuel
- at normal operating exceed the scram setting estab-conditions.
lished in Specification 2.1.A for longer than 1.5 seconds as indicated by the process i
computer.
2.
When the process computer is out of service, this safety limit shall be assumed to be exceeded i
if the neutron flux exceeds the scram setting established by Specification 2.1.A and a con-
's ol rod scram does not occur.
D.
Reactor Water Level (Shutdown D.
Reactor low water level ECCS Condition) initiation shall be > 84 inches above the top of the active fuel
- Whenever the reactor is in the at normal operating conditions, shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be less than that
-corresponding to 12 inches above the top of the active fuel
- when it is seated in the core.
- Top of active fuel is defined to be 360 inches above vessel zero (See Bases 3.2).
E.
Turbine stop valve sciam shall be <
10% valve closure from full open. -
F.
The scram for turbine control valve fast closure due to actuation'of the fast acting solenoid valve shall be > 46f-psig EHC tluid pressure.
G.
Main steamline isolation valve closure scram shall be < 10% valve closure from full open.~
- Top of active fuel is defined to ba 360 inches atove vessel zero (See Bases 3.2).
1.1/2.1-4 Amendment No. 129
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Q'JAD CITIES OPR-29 F.
Turbine Control Valve Fast Closure Scram The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass, i.e., it prevents MCPR from becoming less than the MCPR fuel cladding integrity safety limit for this transient.
For the load rejection without bypass transient from 100%
power, the peak heat flux (and therefore LHGR) increases on the order of 15% which provides wide margin to the value corresponding to 1%
plastic'strair,of the cladding.
The trip setpoint of 1 460 psig EHC fluid pressure was developed to ensure that the pressure switch is actuated prior to the closure of the turbine control valves (at approximately 400 psig EHC fluid pressure) yet issure that the system is not actuated unnecessarily due to EHC system pressure transients which may cause EHC system pressure to momentarily decrease.
G.
Reactor Coolant Low Pressure Initiates Main Steam Isolation Valve Closure The low pressure isolation at 825 psig was provi? d to give protection against fast reactor depressurization and the resulting rapid cooldown of the vessel.
Advantage was taken of tne scram feature which oca rs in-the Run mode when the main steamline isolation valves are closed to provide for reactor shutdown so that operation at pressures lower than those specified in the thermal hydraulic safety limit does not occur, although operation at a pressure lower than 825 psig would not e
necessarily constitute a) unsafe condition.
H.
Main Steamline Isolatico Valve Closure Scram The low pressure iso'.ation of the main steamlines at 825 psig was prcvided to give protection against rapid reactor depressurization and the resulting rapid cooldown-of the vessel.
Advantage was taken of the scram foeture in the Run mode which occurs when the main steamline isolation valves are closed to provide for reactor shutdown so that high power operation at low reactor pressures 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, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron flux scrams.
Thus, the combination of main steamline low pressure isolation and isolation valve closure scram in the Run mode assures the availability of neutron flux scram protection over the entire range of applicability of fuel cladding integrity safety limit.
In addition, the isolation valve closure scram in the Run mode anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve c Msure.
With the scram? set-at 10%
valve closure in the Run mode, tPare is no increase in neutron flux.
1.1/2.1-15 Amendment No. 129
QUAD-CITIES OPR 49 To satisfy the long-term objective of maintaining an adequate levtl of safety throughout the plant lifetime, a minimum goal of 0.9999 at the 951 confidence level is proposed.
With the,ne-out-of-two taken twice logic, this requires that each sensor have an ava' lability of 0.993 at the 95% confidence level.
This level of availability tay be maintained by adjusting the test interval ts a function of the observed failure history (Reference 1).
To facilitate the implementation of this techaique, figure 4.1-1 is provided to indicate an appropriate trend in test interval.
The procedure is as follows:
1.
Like sensors are pooled into one group for the purpose of date acquisition.
2.
Tne factor M is the exposure hours and is equal to the number of sensors in a group, n, times the elapsed time T(H=nT).
3.
The accumulated number of unsafe failures is plotted as an ordinate against M as an abscissa on Figure 4.1-1.
4.
After a trend is established, the appropriate monthly test interval to satisfy the goal will be the test interval to the lef t of the plotted points.
5.
A test interval of 1 month will be used initially until a trend is established.
The turbine control vahe fast acting solenoid valve pressure switches directly measure the trip oil pressure that causes the turbine control valves to close in a rapid manner.
The reactor scram setpoint was developed in accordance with NEDC 31356 " General Electric Instrument Setpoint Methodology" dated October, 1986.
As part of the calculation, a calibration period is inputted to achieve a nominal trip point and an allowable setpoint (Technical Specification value).
The nominal set-point is precedurally controlled.
Based on the calculation input, the calibration period is defined to be every Refueling Outage.
Group 2 devices utilize an analog sensor followed by an amplifier and a bistable trip circuit.
The sensor and amplifier are active components, and a failure is almost always accompanied by an alarm and an indication of the source of trouble.
In the event of failure, repair or substitution can start immediately.
An as-is failure is one that " sticks" midscale and is not capable of going either up or down in response to an out-of-limits input.
This type of failure for analog devices is a rare occirrence and is detectable by an operator who observes that one signal does not track the other t1ree.
For purposes of analysis, it is assumed that 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 the Group 2 devices can sustain unsafe failures which are revealed only on test.
Thereft.re, it is necessary to test them periodically.
A study was conducted of the instrumentation channels included in the Group 2 devices to calculate their ' unsafe' failure rates.
Theanalogdevices(sensorgand amplifiers) are predicted to have an unsafe failure rate of less than 20 X 10 failures / hour.
Thebistagletripcircuitsarepredictedtohaveanunsafefailure rate of less than 2 X 10 failures / hours.
Considering the 2-hour monitoring interval for the analog devices as assumed above and a weekly test interval for the bistable trip circuitr, the design reliability goal of 0.99999 is attained with ample margin.
3.1/4.1-8 Amendment No.129
QUAD-CITIES DPR-29 Group 3 devices are active only during a given portion of the operation cycle.
For example, the IRM is active during startup and inactive during full power operation.
Thus, the only test that is meaningful is the one performed just prior to shutdown or startup, i.e., the tests that are performed just prior to use of the instrument.
Calibration frequency of the instrument channel is divided into two (,;
.gs.
These are as follows:
1.
Passive type indicating devices that can be compared with like units on a continuous basis, and 2.
Vacuum tube or semiconductor uevices and detectors that drif t or lose sensitivity.
Experience with pas'.,1ve type instruments in Commonwealth Edison generating stations and substations indicate that the specified calibrations are adequate.
For thost devices which employ amplifiers, etc. drift specifications call for drift to be less than 0.4%/ month 1.e., in the period of a month a drift of 0.4% would occur, thus providing for adequate margin.
The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approx kately constant rate.
Changes in a power distribdion and electronic drift also require compensation.
This compensation is accomplished by calibrating the APRM system every 7 days using heat balance data by ce:ibrating individual LPRM's at least every 1000 equivalent full power hours ut ag TIP traverse data.
Calibration on this frequency assures plant operation at ;r below th rmal limits.
A comparison of Tables 4.1-1 and 4.1-2 indicates that some instrument char els have not been included in the latter table.
These are mode switch in shutdown, manual scram, high water level in scram discharge volume, main steamline isolation valve clo:;ure, and turbine stop valve closure.
Allofthedevicesorsensorsassociatedj with these scram functions are simple on-off switches, hence calibration is not applicable, i.e., the switch is either on or of f.
Further, these switches are mounted solidly to the device and have a very low probability of moving; e.g., the thermal switches in the scram discharge volume tank.
Based on the above, no cali-bration ja required for these instrument channels.
B.
The MFLPD shall be checked once per day to determine if the APRM scram requires adjustment.
This may normally be done by checking the LPRM readings, TIP traces, or j process computer calculations.
Only a small number of control rods are moved daily, l
thus the peaking factors are not expected to change significantly and a daily check of the MFLPD is adequate.
4 References 1.
I. 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 NE00-21617-A (December 1978),
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3.
NEOC - 31336 "Caneral Electric Instrument Setpoint Methodology" dated October, 1986 3.1/4.1-10 Amendment No.129 1
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6 QUAD-CITIES DPR-29 TABLE 3.1-3 RECTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS R Minisium Number of Operable or Tripped Instrument Channels pe TrioSystem{y3 Trip Function Trip Level Settino ActionI23
-1 Mode switch in shutdown A
1 Manual scram A
APRM[3]
2 High Flux (flow Liased)
Seecification 2.1.A.1 A or B 2
Inoperative A or B 2
Downscale [111
> 3 '125 of full scale A or B 2
High-reactor prewns i Ir,60 psig i.
2 High drywell pressure.
1 2.5 psig A
2 Reactor low water level 1 8 inches [8]
A 2 (per bank)
High-water level in scram discharge volume 1 40 gallons per bank A
2 Turbine condenser low 1 21 inches Hg vacuum A or C vacuum 2
Main Steamline high
< 15 X normal full power A or C radiation [12) power background (without hydrogen addition) 4 Ma 4 steamline isolation
< 10% valve closure A or C valve closure [6]-
.2 Turbine control valve fast 1 460 psig [10)
A or C closuref valve trip system oil pressure low [9]-
2 Turbine stop valve-1 10% valve closure A or C closure [9]
2 Turbine EHC control fluid 1 900 psig A or C low pressure [9]
3.1/4.1-13 Amendment No. 129
-QUAD-CITIES DPR-29 i
TABLE 3.1-4 NOTES FOR TABLES 3.1-1, 3.1-2, AND 3.1-3 (1) There shall-be two operable trip systems or one operable and one tripped system for each function.
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[2] 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 shali be taken:
A.
Initiate insertion of operable rods and complete insertion of all operable rods within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
B.
Reduce power level to IRM range and place mode switch in the Startup/ Hot Standby position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Reduce turbine load and close main steamline isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
[3] -Aa APRM will be considered inoperable if there are fewer than 2 LPRM inputs per C
level or there are less than 50% of the normal complement of LPRM's to an APRM.
[4] Permissible to bypass, with control rod block for reactor protection system reset in refuel and shutdown positions of the reactor mode switch.
[5] Not required-to be operable when primary con _tainment integrity is not required.
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[6] The design permits closure of any one litie without a swam being initiated.
[7] Automatically bypassed when reactor p-essure is < 1060 psig.
[8] The +8-inch trip point is the water letel-as measured by the instrumentation outside the shroud.
The water level inside the shroud will decrease as power is increased to 100% in comparison to the level outside-the shroud to a maximum of 7 inches.
This is due to the pressurti drop across the steam dryer.
Therefore, at 100% power,
-an indication of +8 inch water level will actually be +1 inch inside the shroud.
1 inch on the water level instrumentation is > SO4" above vessel zero.
(See Bases 3.2).
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[9] Permissible to bypass when first stage turbine pressure is less than that which corresponds to 45% rated steam flow.-(< 400 psi)
[10] Trip is-indicative of turbira control valve fast closure (due to low EHC fluid pressure) as a result of fast acting valve actuation.
[11] The APRM downscale trip function is automatically bypassed when the IRM instrumentation is operable and not high.
[12] Channel shared by the reactor protection and containment-isolation system.
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3.1/4.1-14 Amendment No.129 l
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QUAD-CITIES DPR-29 TABLE 4.1-2 SCRAM INSTRUMENT CALIBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHAN b13 Calibration Standard [5]
Minimum FrequencyE23 Instrument Channel Group High flux IRM C
Comparison to APRM after Every controlled heat balance shutdown [4]
High flux APRM Output signal B
Heat balance Once e;ery 7 days Flow bias B
Standard pressure and Refueling outage voltage source LPRM B[6]
Using TIP system Every 1000 equivalent full power hours High reactor pressure A
Standard pressure source Every 3 months i
High drywell pressure A
Standard pressure source Every 3 months Reactor low water level B
Water level
[7]
Turbine condenser low vacuum A
Standard vacuum source Every 3 months Main steamline high radiation 8 Appropriate radiation Refueling outage source [3]
Turbine EHC control fluid A
Pressure source Every 3 months low pressure Turbine control valve A
Pressure source Refueling outage fast closure Highwater 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 missed, 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)~ Maximum calibration frequency need not exceed once per week.
[5] Response time is not part of the routine instrument check cad calibration but will be checked every refueling outage.
[6] Does'not provide scram function.
[7] Trip units are calibrated monthly concurrently with functional testing.
Transmitters are calibrated once per operating cycle.
3.1/4.1-17 Amendment No. 129
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