ML20094M248

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Amend 90 to License DPR-29,incorporating New MAPLHGR Curves for New Linear Fuel Change Calibr & Functional Test Frequencies for Instrumentation & Adding New Tech Specs Re Modified Scram Discharge Sys
ML20094M248
Person / Time
Site: Quad Cities Constellation icon.png
Issue date: 08/02/1984
From: Vassallo D
Office of Nuclear Reactor Regulation
To:
Commonwealth Edison Co, Iowa Illinois Gas & Electric Company
Shared Package
ML20094M251 List:
References
DPR-29-A-090 NUDOCS 8408150593
Download: ML20094M248 (23)
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{{#Wiki_filter:r .aneg I%, UNITt38TATES [' e NUCLEA] MEGULATORY COMMISSION i WASH 6NG f ON, O. C. 20066 o COMMONWEALTH EDISON COMPANY a IOWA-ILLINOIS GAS AND ELECTRIC COMPANY DOCKET NO. 50 254 00A0 CITIES NUCLEAR POWER STATION UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE ' Amendment No. 90 License No. DPR-29 1. The Nuclear Regulatory Comission (the Comission) has found thatt A. The applications for amendment by Commonwealth Edison Company (the Itcensee) dated February 21. February 28 and May 8,1984, + comply 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 It B. The facility will c arate in conformity with the application,

  • l the provisions of tie Act, and the rules and regulations of the l

Comissions I l C. There is reasonable assurance (1) that the activities authortred i by this amendment can be condugted without endangering the health l andsafetyofthepublic,and(11)thatsuchactivitieswillbe conducted in compliance with the Comission's regulationst D. The issuance of this amendment will not be inimical to the convnon defense and security or to the health and safety of the pub 1(ci and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all app 14 cable requirements have been satisfied. l 2. Accordingly, the Ifeense is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License flo. OPR Ig is here:iy amended to read as followst i e prar.g

2 (2)_TechnicalSpecifications The Technical Specifications contained in Appendices A and B as revised through Amendment No. 90, 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 C0ftMISS10N ^]f Domenic B. Vassallo. Chief Operating Reactors Branch *2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of 1ssuance: August 2, 1904

c. e ATTACHMENT TO LICENSE AMENDMENT NO. 90 FACILITY OPERATING LICENSE NO. OPR-29 DOCKET NO. 50-254 . Revise the Appendix "A" Technical Specifications by. replacement, addition or deletion of the following pages, as indicated: Remove Insert 3.1/4.1-2 3.1/4.1-2 3.1/4.1-2a 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 3.2/4.2-10a 3.2/4.2-14 3.2/4.2-14 3.2/4.2-16 3.2/4.2-16 3.2/4.2-17 3.2/4.2 - 3.3/4.3-3 3.3/4.3-3 Figure 3.5-1(Sheet 1of4) Figure (Sheet'3 of 4) Figure (Sheet 4 of 4) -e- -w

QUAD-CITIES DPR-29 3.1 LIMITING CONDITIONS FOR OPERATION BASES The reactor protection system automatically initiates a reactor scram to: a. preserve the integrity of the fuel cladding. b. preserve the integrity of the primary system, and c. minimize the energy which must be absorbed and prevent criticality following a loss-of-coolant accident. This specification provides the 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 instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for briefintervals to conduct required functional tests and calibrations. The reactor protection system is of the dual channel type (reference SAR. Section 7.7.1.2). The system is made up of two independent trip systems. each having two subchannels of tripping devices. Each subchannel has an input from at least one instrument channel which monitors a critical parameter. The outputs of the subchannels are combined in a one-out-of-two logic; i.e., an'd input signal on either one or both cf the subchannels will cause a trip system trip. The outputs of the trip systems are arranged so that a trip on both systems is required to produce a reactor scram. This system nieets the requirements of the IEEE 279 Standard for Nuclear Power Plant Protection Systems issued September 13.1966.The system has a reliability greater than that ora two-out-of-three system and somewhat less than that of a one-out-of-two system (reference APED 5179). With the exception of the average power range monitor (APRM) and intermediate range monitor (IRM) channels, each subchannel has one instrument channel. When the minimum condition for operation on tide number of operable instrument channels per untripped protection trip system is met.or ifit cannot be met and the affected protection trip system is placed in a tripped condition. the effectiveness of the protection system is preserved. i.e., the system can tolerate a single failure and still perform its intended function of scramming the reactor. Three APRM instrument channels are provided for each protection trip system. APRM's = 1 and =3 operam contacts in one subchannel,and APRM's #2 and =3 operate contacts in the other subchannel. APRM's =4. = 5. and =6 are arranged similarly in the other protection trip system. Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel.This allows the bypassing of one APRM per protection trip system for maintenance, testing, or calibration. Additional 1RM channels have also been provided to allow for bypassing of one such channel.The bases for the scram settings for the IRM APRM. 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 the drywell is provided to detect a loss-of-coolant accident and initiate the emergency core cooling equipment. The pressure. sensing instrumentation is a backup to the water-levelinstrumentation which is discussed in Specification 2.1. A scram is provided at the same setting as the emergency core cooling system (ECCS) initiation to minimize the energy which must be accommodated during a loss-of-coolant accident and to prevent the reactor from going critical following the accident. 3.1/ 4.1 -2 Amendment No. 90

6

u-W The' control; rod ~ drive: scram system is-designed 'so that all off the . water which. is -discharged 'from the Reactor by a scram can be accommodated In'the discharge. piping. A part of this system'is an individual instrument volume for each of the south and north CRD accumulators.. These two1 volumes and their piping can hold in excess of C0 gallons of -water and is the low: point In'the piping. No credit was taken_.for these volumes in the design of the discharge piping relative to the amount of water which'must be acco=modated during a scram. During normal operations, the discharge volumes are empty; however, should either. volume fill with ' water. the water discharged to the piping from the Reactor may*not be accommodated 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 the Reactor when the volume. remaining in either instrument volume is approximately 40; gallons. For_ diversity of level sensing methods that will ensure and provide a scram, both' differential pressure switches and thermal ~ switches have been I,ncorporated,into the design and logic of the system. The setpoint for the scram signal has been chosen on' the basis of providing sufficient volume remaining to accommodate a scram even with 5 spm leakage per drive into the SDV. As indicated above, there is sufficient v.olume in the piping to accommodate'the scram without impairment aof the scram times or the amount of insertion of the control rods. This. function shuts the Reactor 'down while suf ficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function properly. 3.1/L.1-2a Amendment No. 90

QUAD-CITIES DPR-29 I l Loss of condenser vacuum occurs when the condenser can no longer handle heat input.. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient,' neutron flux rise, and an increase in surface heat flux..To prevent the cladding safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure. The tur:ine stop valve closure scram function alone is adequate to prevent the cladding safety limit from being exceeded-in the event of a turbine trip transient with bypass closure. The condenser low-vacuum scram is a backup ~to the stop valve closure scram and causes a' scram before the stop valves are closed, thus tne resulting transient is less severe. Scram occurs at 21 inches Hg vacuum, stco valve closure. occurs at 20 inches Hg vacuum, and bypass closure at 7 inches Hg vacuum. High radiation levels in the main steamline tunnel above that due to the normal nitogen and oxygen radioactivity are an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds seven times normal backgrcund. The purpose of this scram is to reduce the -source of such radiation to the extent necessary to prevent excessive tureine contamination. Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector off-gas monitors, which cause an isolation of the main conde'nser off-gas line provided the limit specified in Specification 3.8 is exceeded. r The main steamline isolation valve closure scram is set to scram when the isolation valves are 10% closed from full open. This sc~ ram anticipates the pressure and flux transient which would occur when the valves close. By sc amming at this setting, the resultant transient is insignificant. A reacto'r mode switch is provided which actuates or bypasses the various ~ scram functions appropriate to the particular plant operating status (reference SAP Section 7.7.1.2). Whenever the reactor mode switch is in the Refuel or Startup/ Hot Standby position, the turbine condenser low-vacuum scram and main steamline isolation valve closure scram are bypassed. This bypass has been provided for flexibility during startup ~- and to allow repairs to be made to the turbine condenser. While this bypass 15 in effect, protection is provided against pressure or flux increases by the high-pressure scram and APRM 15% scram, respectively, which are effective in this mode. If the reactor were brought to a hot standby condition for repairs to the turbine condenser,'the main steamline isolation valves would be closed. No hypothesized single'failute 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, thus 8P6viding 'for a manual means of rapidly inserting control rods during all modes of reactor operation. The IRM system provides protection against excessive power levels and short reactor periods in the startup.and intermediate power ranges (reference SAR Sections 7.4.4.2 and 7.4.4.3). A source range monitor (SRM) system is also provided to supply add information.during startup but has no scram ($1onal neutron level functions (reference SAR Setticn 7.4.3.2). Thus the IRM is required in the Refuel and Startup/ Hot Standby modes in addition, protection is provided in this range by the AP9M.13% scram as discussed in the bases for Specification 2.1. In the pewtr range, the APRM system provides required protection (reference SAR Section.7.4.5.2.). Thus, the IRM system is not required in the Run mode, the 3P9M's cover only the intermediate and power range; the IRM's provi$e adecuate coverage in the startup and intermediate range. The hi;h-reac' tor pressure, high-drywell pressure, reactor low water level, and scram discharge volume high level scrams are required f o r. the Startu:/ Hot Standby and Run modes of. plant operation. They are therefore l required to be operational'for these modes of reactor operation.. I The turbine condenser low-vacuum scram is recuired only during power operation and must be bypassed to start up the unit. 3.1/4.1-3 Amendment-No. 66, 90 ~.. i

a t QUAD-CITIES D PR-29 to an out-ef-linits input...This type of failure for analog devices is a rare occurrecce and is detectable by an operator who observes that one signal does not track the other three. For purposes of analysis, it is assumed thr.t ibis rare failure vill be detected within 2. hours. The bistable trip circuit which is a part of the Group 2 devices can 4 sustain unsafe failures which are revealed only on test Therefore, it is necessary to test them periodically. A study uns conducted of the instrumentation channels included in the Greup 2 devices to calculate their ' unsafe' failure rates. The analog devices (sensors and ampli iers) are predicted to have an unsafe failure rcte of less than 20 x 10j failures / hour. predicted to have an unsafe failure rate of less than 2 r 10The bistable trigeircuits are I failures / hour. Considering the. 2-hour scnitoring interval for the analog devices as assuned above and a weekly test interval for the bistable trip circuits, the design reliability goal of 0.99999 is attainad with ample margin The bistable devices are monitored during plant operation. to record their failure h1 story and establish a test interval using the curve-of Figure 4.1-1 There are numerous identical histable device.i used throughout the plant instrumentation system Therefore, significant data on the failure rates for the bistable devices should be accumulated rapidly. The~ frequency of calibration of the APRM flow biasing network has been established. at each refueling outage. The-flow biasing network. is functionally of. the flow input to the flow-biasing netverk can be made during functional test by direct meter reading- (IEEE 239 Standard for Nuclear-l Power Plant Protect 1on Systems, Section 4.9, September 13, 1966). There are several instruments which. must be-calibrated, and. it. vill take several days to perfoer. the. calibration; of the entire network. 4 Wile the. calibration is being perf ormed, a. zero flow signal vill be sent to half' cf the APRM's, resulting in a half scram. and. rod block condition Thus,11 the calibration-vere perfonned during operation,. fluz shaping would nor be possible 3ased. on experience. at other generating stations, drift. of instrument sich. as those in the-flow biasing network is. not significant; therefore, tar avoid. spuriour scrams, s. calibratiotr frequency of each. refueling outage-is es tablished Reactor low water le' vel instrtssents 1-263-57A,.1-26]-575, L-263-58A, and. 1-263-583 have been modified. to be an analog trip system. '"he analog trip systes consists of an analog se,nsor (transmitter) and. a. master / slave. trip unit setup which ultimately drives a trip relay. The frequency of calibration-and SinctionaL testing for instrissent loops of the analog trip syster,. including reacter low water level, har been established. in Iicensing " Topical Reporr NEDO-21617-A (December 1978). logic, NEDO-21617-A states that each; trip unit bec subjected to. a.With. the one-cut-of-I calibratiorlfunctional test of one month. An. adequate. calibration / su rveillance test interval for the trans=itter is once per operating cycle. 3.1/4.1. Amendment No, 90 e . _., _. _ _ -.. - ~ -, -, -. _ _..,. _, _, -. - - _.,.,

QUAD-CITIES OPR-29 Group 3 devices are active only during a given portion of tne 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 perfoImed just prior to shutdown or startup, i.e., the tests that are perfcImed just prior to use of the instrment. Calibration frequency of the instrument channel is divided into two groups. 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 devices and detectors that drift or lose sensitivity. Experience with passive type instrments in Cbmmonwealth Edison generating stations and substations indicate that the specified calibrations are adequate. For those devices which employ amplifiers, etc. drift specifications call for drift to be less than O.4%/ month 1.e., in the period of a month a drift of 0.4% woulo occur, thus providing for adequate margin. The sensitiv1.ty of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. Changes is a power distribution and electronic drift also Iequire compensation. Tnis compensation is accomplisned by callorating the APRM system every 7 days using heat balance data by calibrating individual LPRM'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 comparisan of Tables 4.1-1 and 4.1-2 indicates that some instrument channels have not been included in the latter table. These are mode switch in shutoown, manual scram, high water level in scram discharge volume, main steamline isolation valve closure, turbine control valve fast closure, and turbine stop valve closure. All of the devices cr sensors associated with these scram functions are simple on-off switches,' hence calibration is not applicable, i.e., the switch is either on or off. 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 volme tank. Based on the above, no calibration is required for these instrument channels. B. The WLPD shall be checked once per day.to deteImine if the APRM scram Iequires adjustment. This may normally be done by checking the LPRM readings, TIP traces, or process computer calculations. Only a small nmber of control rods are moved daily, thus the p'eaking factors are not expected to change significantly and a daily check of the WLPD is adequate. ~ ~ Re ferences 1. I. M. Jacobs, " Reliability of EngineeIed Safety Features as a Function of Testing Frequency", f4) clear Safety, Vol. 9, No. 4, pp. 310-312, July-August 1968, 2. Licensing Topical Report NEDO-21617-A (D3cemoer 1978). 3.1/4.1-7 Amendment 61, 90

QUAD-(TIIF.S DPR-29 TABLE 3.11 REACTOP Pt01ECil0ll ST5f tl.1(50 RAM)ll:STROMENTATION REQUIREl'El(15 REFUEL 1.',00E Wimum t mber of Op: rat!s y Tripped instre:nt a Act!:a5 Chanz:!a perm Talp Tsact!as trip level setting Irly $pte:a A 1 Mode switch h shutdewn A ), 1 Manual seram RM -3 ' High N s120/125 of fut s: ale A 3 Inoperathe ~ gpggm 2

  • High hi (15% suam)

SpeciG:ation 2.1.A.2 A A 2

  • hoperative I

2(per ba.k) H',gh water level h scam 4 40 gallens per bank A dhcha.ge vobme'U - 2 High rea: tor pressure s1060 psig A m s2 psig A 2 High drywel pressure 2 Rea: tor bw water level 28 hches* A 2 Turthe condenser low 221hd.es Hg vacwm A m vacuum 2 ,Mah steamine high 57 X normai fut pmer A radiatbanD backgrcund 4 Mah steamine sc'atbn s10% vake cbsure A vake cbsurem t A. endnent No. 66, 90 3.1/41-C k

QUAD-CITIES DPR 29

  • F TABLE 3.12 RE AC10R PRDIECTIO!! $T!IEM (5:R11.1)ll;51RUI.1E!!Till0ll REQUIRE!,tEll15 ST ARIUP/ HOT STA!;CET 110DL

~ l'.irlm:m Kumber el 0;:ratie er Trl;-:1! lastrustet Chanre!s per Trip sptis trip Funciles Trip level Setting Acties5 m A 1, Mode switch h shutdow A Manual sca5 1 1RM 3 Egh is 5120/125 of ful scale A "A 3 Ino;erative' Apggn 2 IQh Ga (15% saa.-d Spe:itcation 2.1.A.2 A A 2 be;erative 2 H@. renter pressure $1060 psig A S $2 psig A 2 High-6Tatt pressure 2 Renter hw water level 2 8 hches'81 A ~ 2 ( per bank i High r;ater level h scam t81 6 40 ganons per bank A dh:harge volume 2 Tu:bhe c:rdenser bw 221hches Hg va:un A m va:uum 2 Mah steamhe high $7 X normat fuH power A rafiatior:423 hadttound 4 Mah steambe hotation s!C% valve closure A rake chsurem 1 /cendrnent No. 66, 90 3.1/41-9

Ol'AD CITIF.S DPR-29 TABLE 3.14 '~ REACTOR PROTECTION SYSTEM ($ CRAM)(N$iRUljtWT Afl0N RIOU1REttEllT MirJmsm Kr.har of OperaMe er Tripp:t Inst. aast v Acti:a2 Chana:!s per Trfs tertl 3 tting fD frip Fan: tion Trip System A Mode swRch h shutdown 1 ~ A 1 Manual saam Mmm ICgh ths (C:n biased) $;<cifatbn 2.1.A.I. Aw8 AW8 2 2 k,4rative Ocwnscatenn 23/125 of fut stak Aw8 2 A <1050 psig '2 High-teactor pesswe A 2 High47 wet presswa 52 psy A, 81 Rea:tw kw rater kvel 28 bches 2 2( per bsnk) ' High. water hvel h scram A \\ dis:harge vobrne 6 40 gallons per bank A or C Twthe condenser bw 2:1hches Hg vacwm ~ 2 va::a Mah steamtoe high s7 I notte.al ful AwC 2 power background radstionnr1 Mah steamice isolatbn 4 s10% valve cbswe AwC vake cleswe'll 240% tubine/ generator A or C Twthe centici vake fast 2 bad mhmatchns) cleswe* Twthe stop vake s10% vake eleswe A or C 2 m chswe A or C Tubhe DIO control Suid 2900 psig 2 5 bw pesswe 4 /cendrnent No. 66, 90 3.1/41-10

~ QUAD-CITIES DPR-29 TABLF. 4.1-1 SCRAM INSTRUMINTATION AND II)GIC SYSTIMS MNCTIONAL TISTS i MINIMUM.NNCTIONAL TIST FRIQUINCIIS ?UR SATITT INSTROENTATION, 7 IC STSTIMS, AND CONTG CIRCUITS' Mini =um Frecuenev(4) Functional Test Ins tn=:ent Channel Greue Mods switch in shutdown A Place node switch in Zach refueli=g outage shutdevn Mancal scram A Trip channel and alam Ivery 3 months IRM Trip channel and. ala=x(5) 3efore each startup -High flur C and week 1 d . refu eling[6 }uring Inoperative .C Trip channet and alatz 3efore each scartup and veek1 d T6 "Ti E refueling APU Trip output relays (5) Onca each h High flur. 3 Inoperative 3 Trip output reJr.ys Once each week Downscale 3 Trip output rels.ys Once each week. High flur 15 - C M p output. rels.ys-Before each startup i and weekigduring I refueling High rescror pressure A. Trip channel and. alarx (1) Eigh drywell pressure A Tripr channel and. alam (1) Rerctor low water level (2) 3 (8) -(l) .High water level in. scraz (9) ' A. Trip channel and, alam IverT 3 months diccharge volume. (thermal and dp switches) Turbine. condenser low vacuun A Trip channel and alam (1) Main ste

e. high 3

Trip channet and alar.n(5) Once each we'c raiiation - Main. sreamline isolation valve A Trip channel and. alam (1) closure. Turbine control valve fas: 1. Trip ch'andefand alaer (1) closure Turbine'stop valve closure A. Trip channel and ala=x (1) Turbine IEC control fluid. low 'A Mp channel and. ala=n. (1) l pecssure 6 3.1/4.1-12. Amendment No. 90 i l e i ~ ~..

n =. I ~ ..i QUAD-CITIES DFR-29 TABLE 4.1-1 (Cont' d) Notes: 1. Initially once par nonth until exposure hours (M as defined on Figure 4.1-1) are 2.0 I 10); thereaf ter, according to Figure 4.1-1 vith an interval not less than 1 month nor more than 3 months. The ccupilation of instrument failure rate data nay include data obtained from other boiling vater reactors for which the same design instrument operates in an envirotxnent si=ilar to that of Quad-Cities Units 1 and 2, 2. An instru=ent check shall be perfonned on low reactor water level once per day and on high steamline radiation once per shif t. 3. A description of the three groups is included in the bases of this specification. 4 Functional tests are not required when the' systems are not required to be operable.or are tripped. If tests are nissed, they shall be performed prior to returning the systems to an operable status. 5. This instrumentation is exempted frcm the instrument functional test definition (1.0 refinition F) s This instrument functional test vill consist of injecting a simulated electrical signal into the measurement chann els. 6. Frequency need net exceed'veekly. ~ 7. A functional test of the logic of each ' channel is performed as indicated. This coupled with placing the mode switch in shutdown each refueling outage constitutes a logic system functional test of the scram systen. 8. A functional test of the master and slave trip units is required mo'nthly. A calibration of t he trip unit is to be perf ocned concurrent with the functional testing.

9. Only the electro 6ics portion of the thermal switches will be.

tested using an electronic calibrator during the three month test. A water column or equivalent will be used to test the dp switches. 3.1/4.1-13 Amendment No. 90

L@Fs) TA3LE 4,1-2 SCRAM INSIRUMENT CALIBRATION MINnfUM CAL:3 RATION FRMUENCIES FOR REACIOR FROTECTION INS F2EST CHANNELS Inst unent Channel Grous Calibration Standard Mini =um Frecuenev(2) Eigh flux Ip.M C Comparison to APRM af ter Ivery cogolled P heat balance shutdown High fluz AFFM Cutpu t signal-3 Heat balance Once every 7 days 71ov bias 3 Standard pressure and Refueling outage voltage source - 0) LyRM 3 Using IIP systen Ivery 1000 equivalent full power hours High reactor pressure A Standard, pressure source Ivery 3 acuths High dryvell pressare A. Standard pressure source

  • Every 3 months Raactor low water level 3

Water level (7) Tut 5ine condenser lov vacutzn A S tandard. vacut= source Ivery 3 months Main stes:nline hi'gh. radiation 3 Appropgte radiation Refueling outage source Turbine EEC control fluid A Pressure source Every 3 months low pressure Highwater level in scram A Water level Refuelirig outage discharge volume (dp only) Notes: 1. A description of the three groups is included in the bases of this stecification. 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 c annel aliganent everr 3" months.

  • 4 Maximum calibration frequeneT need nor exceed once per veek.
5.. Response tine is nor pare of the routine instrtment check and calibration but vill be checked every' refueling outage.

t 6. Does not provide scran function. l 7. rip units are calibrated monthly concurrently with functional testing. Trans=itters are calibrated once per operating cycle. 3.1/4.1-14 /.nendment No. 90 +

QUAD-CITIES DPR-29 Optimizing each channel independently may not truly optimize the systen considering the overall rules of system operation. However, true syst em optimi:.ation is a canplex problem. The optimums are broad, not sharp, and optimizing the individual, channels is generally adequate for the system. 7 The formula given above minimizes the unavailability of a single channel which must be bypassed during testing. The minimization of the unavailability is illustrated by :urve 1 of Figure 4.2-2, which asmsmes that a channel has a failure rate of 0.1 x 104/ hour and 0.5 hour is required to test it. The unavailability La a minimum at a test interval 1, of 3.6 X 103 hours. J If two similar :hannels are used in a one-out-of-two ccafiguration, the test interval fot miaimum availability changes as a function of the rules for i te s ting..- The simplest case is to test each one independent of the other. In this case, tl' art is assamed to be a finite probability that both may be bypassed at one time. This case is shown by curve 2. Note that the unavailability is lower, as expected for a redundant systen, and the minimum occurs at the sane test interval. Thus, if the two channels are tested independently, the equation above yields the test interval for minimum unavailability. A more usual case is that the testing is not done independently. If both l channels are bypassed and tested at the same time, the result is shown in curve 3. Note that the minimum occurs at about 40,000 hours, much longer _than r for Cases 1 and 2. Also, the minimum is not nearly as low as Case 2, whie.h indicates that this method of testing does not take full advantage of the redund nt channel. Bypassing both channels for simultaneous testing should be avoidei. The most likely case would be to stipulate that one channel be bypassed, tested, and restored, and then immediately following the second channel be s bypassed, tested, and restored. This is shown by curve 4 Note that there is not true minimum. The curve does have a definite knee, and very little reduction in syst am unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel, t The best test. procedure of all those examined is to perfectly stagger the ~ tests. This is, if the test interval is 4 months, test one of the other l channels every 2 months. This is shown in curve 5. The difference between Cases 4 and 5 is negligible. There may be other arguments, however,' that more i - strongly support the perfectly staggered tests, including reductions in tsiman l l error. The conclusions to be drawn are these: a. A one-out-of-n systen may be treated the same as a single channel in' i terms of choosing a test interval. ) b. More than one channel 'should not be bypassed for testing at any one time. 1 i 3.2/4.2-10 Amendment No. 90 4

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~ r ,-,,n,eym.-__~,,w,,,,,,ww_,, .,,,-mn,wn~-,e ,m.--, w w-m, -e--,w-en--,.--- ,nua,,--o

QUAD CITIES DPR-29

  • eactor water level instruments' l-263-73A & 3, HPCI high stean flow instruments 1-23S9A-D, and HPCI steam line low pressure instruments 1-2352 & 2353 have been modified to be analog trip systems. The analog trip system consists of an anales sensor (transmitter) and a master / slave trip unit setup which ultimately drives a trip relay. The frequency of calib' ration and functional testing for instrument loops of the analog trip system has been established in Licensing Topical Report NEO-21617-A (December 1978). With the one-out-of-bro-taken-twice logic, NEDO-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 the transmitter is once per operating cycle. The radiation monitors in the ventilation duct and on the refueling floor which initiate building isolation and standby gas treatment operation are arranged in two one-out-of-two logic systems. The bases given above for the rod blocks appiy here also and were used to arrive at the functional testing frequ ency. Based on experience at Dresden Unit 1 with instruments of similar design, a testing intersal of cuce every 3 months has been found to. be adequate. The autcx atic pressure relief instrumentation can be considered to be a one-out-of-two logic system, and the discussion above applies to it also. The instrumentation which is required for the postaccident condition vill be tested and calibrated at regularly scheduled intervals. The basis for the calibration and testing of this instrumentation is the same as uns discussed above for :he reactor protection systen and the amargency core cooling systems. References 1.

3. Epstein and A. Shiff, " Improving Availability and Readiness of Field Equipment Through Periodic Inspection", UCRL-50451, Laurence Radiation Labora tory, p. IC, Equation (24), July 16,1968 3.2/4.2-1D %

Amendment ho. 90

QUAI)-CITIES. DPR-29 TABLE 3.24 l RSTRUl4ENTATION THAT LMIT11TES RCD BLOCK uweium swiner er operanse er inspec inswmast Cheensk per frtp $rsieg* bestroset h lam artW 2) 2 APRM gscale (b biasyn d$ O.5BW3 + 50 ~ N h F.M.PD APRM gsule (Re*uel and Start @/ Hot 512/ID M me 2 Standby mode) 2 APRM downsesk* A3'IU M "U so. 65Wp + 4 2 ( 2) 1 Rod bbci montar escale (b basyn 1 Rod tiock monitor downscale'n 23/125 fua scale 3 IRM downscale mm 23/125 fun sak 3 IRM esuem s108/125 M scak p SRM detector not m Startup posrtorf8) 22 feet below core center. Ime 3 ftM dete: tor not in Startup posterp86 22 feet bebw core center-Ine pm SRM sscak s105 counts /sec p SRM downscalem 2102 counts /sec 1(pertank) Hgh mater kvel m scram d:s:harge volume (SDV) f 25 gal 1%per bank) 1 SD7 high water level scram NA trip bypassed Notes 1. For the Startuo/H:t Stan:ty arrd Run positions of the reactor mode selector switen, there shall be two operaole or tripped trip systems for eacn function exce:t the SRM rod blocks. IRM upscale and IRM cownscale need not De operable in :ne Run positicn, APRM downscale, A RM uosale (flow biased), and RBM downscale need not be operable in the Startup/ Hat Stancby mode. The PSM umscale need not be ocersole at 12ss than 3C3 rated thermal power. One channel may te cypassed above 3CX rated thermal power proviced that a limiting control rod pattern coes not exist. For systems with more than one channel per trip system, if the first Coli.rin cannot be met for one of the two trip systems, this concition pay exist for up to 7 days provided that during that time the a: era:1e system is fmetionally tested imediately and cally thereaf ter; if tnis condition lasts longer than 7 days the system snall te tripped. If the first column cannot ce met for coth trip systems, the systems shall De trisced. 2. wp is the percent of crive' flow recuired to produce a rated core flpw of 98 million Ib/hr. Trip level setting is it} percent of rated power (2511 WWE). 3. IRM co.nscale may be bypassed unen it is on its lowest range.~ 4 Tnis function is cypassed enen the count rate is GT/E 100 CPS. 5. One of the four SRM inouts may De bypassed. 6.

  • his $2M function may te typassed in the hi;ner IT!M ranges (ranges B,9, and 10)

=nen :ne l'M upscale red elec< is coeraole. 7. NO: rewired to te coera:le enile performing low PO=er pnysics tests at ater:scretic pressure a; ring or after refueling at poser levels not to exceed Sda t. 8. T9151"M ' unction cecurs nen the reactor mode seiten is in the Refuel or Start.c/tt Stan00y pcsition. 9. This trio is bypassed unen the SRM is fully inserted. Am e.dment No.70,/j,90 3.2/4.2-14 f

QUAD-CITI'S DPR-29 TA3I.E 4.2-1 MIN MCM IS! Ah0 CAL:3 RATION FKKUENCT FCR. CORI. AND. II) CONT.AINMIN~ C00 LINO STSTDtS INST".M.NEATION, ROD 3LOCIS, AND ISOIATICNS Ins trumant - Ins t: ment Funct gal i Ins tr~ent U channel Test Calibration ( ) Check RCCS Instr =.a.ntation 4 1. Recctor low-lov ater level (1) once/3 =onths once/ day. 2. Dryvell high pressure (1) Once/3 months yone 3 Restor low pressure (1) Once/3 months None. 4 Contai=sent spray interlock. a. 1/3 core taight (1) (10) (10) None } b Contai ant. ; essure (1) Ouce/3 mon:hs None 5.. Iow-pressure coes cooling (1) Once/3 months No ne. i pung discharge 6. Underroltaga 4-k7 essential-Refueling outage Refueling outage None 7.. Degraded voltage R,efueling (8) Refueling outage, Once/ month 4-kv essential buses outage Ltd. 31ceks l. AP?.M downscale .(1) (3) ~ Once/3' months-None m2. APOM. flow variaale. (1) (3) Refueling outage-None. 3 II.M. up scale (5) (3) (3) (3) Kone i'4 IZX. d ownscale (5) (3) (5) - (3) None-5 13M 'up scale. (1) (3) Refueling outage None 6 R3M downscale.. (1) (3) Once/3 months None 7 SMf pecale (5) (3) (5) (3) rene , 3. SRM. ietector :or in. stz::=p (5) (3) (6) None. i pocitiett a 9. IRM. derector see in. starmy pcoitiotr (5) (6) None 10 513.Lownscale. (5) (3) (5) (3)~ None-11 High. us.ter level in. scram once/3 months. Not app 14 r shle. None; diccharge volu=e (SD7) 12 SDT high level trip Refueling outage Not applicable. None-t bypassed. l Main.St= -1# e Isolatiotr l Steact. tunnel high tenperature Refueling outage._ Refneling outage Yone-2 St -1'ne: hig fl.v (1) Once/3 months onee/ day-3 S ta=1'

e. Icw pressure (1)

Once/3 months None-4 Sta=1'ne: hig= radiatiott (1) (4). Refneling outage Once/ day-5 Reactor Icw low vata = level (1) (10) (10) Cace/da=-l RCIC' Isolatiotr 1 Ste=-14,e high !!av-Once/3' conths (9) Once/3 months (9) Fone l 2 ' m:bine area. his:. :esperac.tre Refueling outage Refueling outage None 3 Low reactor press::e Once/3 months once/3 months None - 3.2/4.2-16 AmDndment No. g7, p, gg, 90

QUAD. CITIES OPR-29 TA8.E 4.2-1 (cont'c) Instrunent 1onal Instrument Fune;W CallDration(2) C ec4(2)_ Instru ent Te st Charrel rf I Isolation 1. Steamline nign flow (1) (9) (10) (9) (10) tone l 2. Steamline ama nigh temperature ~ fueling outage % fueling outage tone l 3. Low reactor pressu:e (10) (10) rene Feactor Guilcing wnt11ation System Isolation and Stancey Tmatment System Initiation 1. Ventilation exnaust cuct (1) Once/3 months Once/ cay raciation montiors 2. Refueling floor radiation (1) @ce/3 months Chce/ cay monitcrs Steam Jet Air Ejector Off-Cas Isolation 1. Off-gas raciation monitors (1) (4) Re fueling outage Once/ cay Centrol Room wntilation System Isolation 1. Nacter low water level (1) Once/3 montns Once/ cay 2. Crywell hign pressure (1) Chee/3 months tone 3. min steamline high flow (1) mee/3 months Once/ cay 4 Ventilation exhaust cuct (1) Once/3 months once/ cay rotes: 1. Initially once per montn until exposuIt nours (M as cefined on Figure 4.1-1) am 2.0 X 10 ; thereafter, accorcing to Figum 4.1-1 with an interval not less than 1 month nor 5 more tnan 3 months. The compilation of instrument failure rate cata may incluce cata _ octainea from otner coiling water reactors for wncn the same oesign instrunent operates in an envircrvnent similar to tnat of @ad Cities triits 1 ana 2. 2. Functional tests, calibrations, and instrtsnent chetks a:e not requirec wnen tnese instnrents are not requirec to be operaule or tripped. 3. This instru entation is excepted from the functional test cefinition. The fmetion test - snall consist of injecting a simulated electric signal into the measurement cnannel. 4 This instrunent enannel is exceptea from the functional test cefinitions and shall te calibrated using simulatea electrical signals once every 3 montns. 5. Functional tests shall be perfo:med before each startw with a requirec frequency not to exceed once per week. Cilibrations shall be perfo:mec curing eacn startup or curing controlled snutco ns witn a required frequency not to exceec once per weeg. 6. The positionirg mecnaryism snail be cal 1Dratec eiety refueling outage. 7. Logic system functional tests are performed as specified in the applicaole section for tnese sytems. 8. Functional tests shall incluce verification of operation'of the cegrama voltage. ~ 9. Verification of the time celay setting of 3 3y 10 seconds snall be perfo:mec curing eacn refueling outage.

10. Trio units are functionally testea monthly. A calibration of t% trip unit is to oe
erfc
ec concurrent with the functional testing. Transmitters are calltrated once per c;eratirg cycle.

3.2/4.2-17 ccencnent to./y, 90

QUAD-CITIES DPR-29

3. ' The control rod drive housing support
3. The correctness of the control rod system shall be in plac: during reactor withdrawal secuence input to the power operadon and when the reactor RWM computer shall be verified after coolant system is pr:ssurized above loading the scouence, atmospheric pressure with fuel in the Prior to the start of control rod with-reactor vessel, unless all control rods rawal wards en..ticality, the capabil-are fully inserted'and Specification sty of the rod worth mimmtzer to 3.3 A l b met.

properly fulfill its function shall be a. Control rod withdrawal sequene:s verified by the following checks: shall be established so that max- - a-The RWM computer onh.ne d.iag- -imum reactivity that could be n stic test shall be successfully added by dropout of any incre-performed. ment of any one control blad:

b. Prop,er annunciaden of the s:lec-oule ec suen tnat the roe crop a :icent cesign limit or 280 caugm is not excee:ec.

tion error of one out of. sequence e nti i r d shall be verified.

b. When:ver the reactor is in the Starti:p/ Hot Standby or Run c.

The rod block function of the mode below 20% rated thermal -RWM shall be verified by with'- power, the rod worth minimia:r drawing the first rod as an out-shall be operable. A second opera-of secuene: contre! rod no mor: tor cr qualified technical person than to the block point. may be used as a substitute for an inoperable rod worth minimizer 4. Prior to control rod withdrawal for which fails after withdrawal of at startup or during r:fueline. verify that ~ least two source ran2e cha'anels least 12 control rods to the fully at withdrawn position. The rod have an observed count rate of at least ~ worth minimizer may also be three counts per second. bypassed for low power physics testing to demonstrate the shut-

5. Wh:n a limiting control rod pattern down margin requirements of C

8 2n 19struent functional test of ' Specification 3.3.A if a nuclear engineer is present and verifies the he RBM shall be p:rformed prior to step-by-step rod movements of the [n, "al of t ' ciesignated rod (s) I h u tg test procedure. 4. Control rods shall not be withdrawn 5. The scram disenarge volume vent ano crain for startup or refueling unless at least valves shall De verified 0;en at least once two source range Channeis have an per 31 days. These valves may be closed observed Count rate equal to or grealCT intermittentlv for testi.'g under than three counts Per second and these acministrative control anc at least once per 92 days, eacn valve shall te cycled througn at least ene c molete cycle or rull travel. SRM Y are fully,nserted. i At least Once eacn Rerceling Outage, the .ill ee eemon;stratee to: scram dischar e volu*

5. Durine operation with limiting con-trol rod patterns, as determined by the nuc!:ar engineer. either:

or a si; al for control rocs to scram, and a. both RBM channels, shall be operabitf. L' Open.wn the s::am signal is reset. b. control rod withdrawal shall be blocked: or 3.3/4.3-3 Amendment No. $7, g/,90

J DPR-29 - Tig. 3 J-1 Maxirum Avung2 Plan:r Lirnar (Shnt 1 of k) E tt gee rc. tion R.ts (MAPLRCR) vs'. Planar Avera6e Exposure

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7 Quad Citics DPL-29 Maximus Avareg) Plannr Lin=r Hrt G:ncr; tion R:t'; (MAPLHCR) vs. Planar Average Lgosure (Sheet 3 of h) Fig. 3 5-1 .._y..._=n=-n. n = n = = : r::, _-,. =- :n e.-- :=== _r :r u._.. ._.:n.-- = .=.!u = =-:

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