ML20086T364
| ML20086T364 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/28/1984 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20086T354 | List: |
| References | |
| 8131N, NUDOCS 8403060279 | |
| Download: ML20086T364 (14) | |
Text
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ATTACHMENT 1 Proposed Change to Appendix A Technical Specifications to Operating License DPR-29 Revised pages: 3.1/4.1-8 3.1/4.1-9 3.1/4.1-10 3.1/4.1-12 3.1/4.1-13 3.1/4.1-14 3.1/4.1-2 3.1/4.1-2a 3.1/4.1-3 3.1/4.1-7 3.2/4.2-14 3.3/4.3+3 8131N 8403060279 840228 PDR ADOCK 05000254 P PDR
'
- QUAD-Cllll3
' DPR-29
~
f A8tt 3.1-)
inEACTOR PT01ECT10!! $TSTEH ($ CRAM)II STROMENTATION REQUIREMEllis REFUIL Wimum Romber of Op:retle er
- 1 ripped Instrument
.! Chann:Is per Actioam 1Ip runcties trip level setting Irly Systea*
A ..
3 Mode switch h shutdown '
A
'l 1 Manual scram i
M 5120/125 of fut scale A 3 Hth Ba Inoperative 3
Apggm Specifcation 2.1.A.2 A 2 High On (15% scram) '
A 2 InoperatW A I 4 40 gallons per bank 2(per bank) Hih water levet h scram discha se votame* - ^
s1060 psig A 2 Mth reacti presswa
$2 psi A 2 High drp.e5 presswe*
~
A Reactor Icw water level 28 hches*
2 221 hches Hg vacuum A 2 Turbhe condenser hw vacuum m
s7 X no mat fut power A Man steamine high 2 background radiationem A
Mah steamroe 'notatbn s10% vake cbsure 4
vake cbswe m ,
Amendment No. 66 3.1/41-8
QUAD-ClTIES DPR-29
.e ,
TABLE 3.12 REACIOR PR0l[Ciloti ST$ FEM ($CR1!.1) llis1RUI.1E!:llT10H REQUIREl1Effis UCDL sualmem 1: umber -
W epr Ws er Irlppi lastrument Chsaa:ts per Actiss2 U3 Trip Faaetles trip level Setting trip Systee A
1, Mode switch h shutdm A
1 Manua! scrern .
- RM A NghRur s120/125 of fut scale 3
A 3 Inoperative Apggle Specification 2.1.A.2 A Hgh flus (15% saam) 2 - A Inoperative 2
$1060 psig A 2 N';h reactor piessure S 52 pst A 2 H(h dtpel pressure A
Rea: tor Icw water level 28 inches
- 2 2 ( Per bank 1 High water level h scram 6 40 gallons per bank A discharge ve!ume'o 221 inches Hg va:vum A 2 Turthe cedenser bn vacuumW s? X normal fufi power A Main steamt.ne high 2
ur ba:1g'ound radiation !
A Mah steam!be isolation 510% valve cbswe 4
valve cbswe m ,
Amendment No. 66 3,1fts.,
QL'AD-CITIF.S DPR-29 I TABLE 3.14 REACTOR pBOTECilDN SISTEM (SCRAM)l451RUtaENT AT10N REQUIREMEH uini x ier of Operable er
- 1 sipped instrument Acream Channels per ,
trip level Settlag Trip Spte.W Trip Foncties A
t .
Mode swRch h shutdown .
A 1
Mamat soam
. APRMS Specircation 2.1.A.!
Aa8 2 Kth Rur (ton biased) Aw8 2
beperative 33/32$ or fug'scak Aw8 .
Downscalerm 2 A
<1060 psk Ifgh-reacts pesswe
't A s2 psk Ksh drywet pesswa .
2 at hches e A, 2
Rea: tor hw water level 2(per bank) ' tfgh-wata level h scram A l discharge volume 4 40 gallms per bank
~ m21 inches Hg vacuum AwC 2 Twthe conder:ser low
- vacuum Mah steamin high 57 X normal fut 2 power ba:1greand AwC ra6ationem 4 Mah steamim isolation s10% vake cbswe AaC valve cbswe*
240% turbine / generator AwC 2
Twthe contic! vake fast bad mhmatchn#
ebswe*
' AwC Tubhe stop valve 510% valve cbswe 2
sbswe* -
2900pst AwC 2
Twbhe EHC controlluid .
bw pesswe* .
Amendment No. 66 3,3,4,,,,,
- QUAD-CITIES DPR-29 L
TA8LE 4.11 SCRAM INSTRUMENTATION AND LOGIC SYSTLMS FilNCTIONAL TESTS MilllMUM FUIICTIONAL TEST FREQUENCIES FOR SAFETY INSTRUMENTATION, LOGIC SYSTEMS, AND C INuesa Frugs:ac/0 trauiP Fumetumet Test'73 lastumsstChamasi Each refueing outage A Place mode switch in shutdown
. Mode sentch a shutdown T4p channel and alarm Every 3 months A
Manual scram IRW Before each startup and weekly C Trp channel and alarm
- Hgh ihm dweg refueleg*
Before each startup and weekfy C Trp channel and alarm Inoperative durmg refueleg*
APIIM Once each week B Tre output relays
- Hgh Sur Once each week B Tre output relays inoperatnte Trp output relays
- Once each week B
Downscale Befo's each startup and weekly C Tre output relays
- Hgh tur 15% demg refueleg*
(1)
A Tre chanrel and alarm Hgh reactor presswe Trp channel and alarm (1)
A Heti dryweX presswe Trp channel and alarm (1)
A Reactor low water leveP Trp channel and alarm Every 3 months Hgh water level m scram A docharge volume (thermal and op switches)
(1)
A Tre channel and alarm Twt >me condenser low vacuum Once each week Man steamine hgh B Tre channel and alarm
- radeten*
Trp channel and alarm (1)
A Man steamime isolaton valve classe (1)
A Tre channel and alarm TwtHne control valve fast closwe Trp channel and alarm (1)
Twbine stop valve classe A Trp channel and alarm (1)
A Twbme EHC control Red low Pesswe i
3.1/41-12
QUAD-CITIES DPR-29
.. TABLE 4.11(Cent'd) g
=
5
- 1. buteily once per swth untd suposure hours thi as ee%d on Figure 41-1) are 2.0 : 10 . thersahe". accordes to Figure 41 1. with an interval not ser more than 3 months The compiiate of mstrument failure rate dita may eclude cata obts.ed from other boilms satii reactors for whch the same design hstrument operates r an envrone. tot sutilar to that of Quad Cities (m.ts 1 and 2.
- 2. . An instrument chech shall te periorN on low reactor water intel once per day and cm h@ ste* % radaten once per shift.
- 3. - A descripton of the three groups is sciudad a the baws of thrs speceluten.
^
4 ' funcional tests are not retowed when the systems are not retured to te operable or are tripped if tests are esseo they shall be performed prer to returnmg the systems to an operable status.
i Ths estrumentaten a enemited fre the estrument functua! test defmition (1.0. Definiten () This estrument functonal test mil consist of miecteg a smiated electreat signal sto the measurement dannels 6 frageancy need not anceed weeMy
- 7. A functional test of the lagt of each chainet is periore.ed as edcated Ifus coupled with places the mode sw.tch m shutdown ea:;h refueirg outage constitutes a legc system functenal tast of the scram system.
- 8. Only the electronics portion of the thermal switches will be teste 1 u:1rg an electronic calibratcr during the three month test. A water column or equivalent will be used to test the dp switches.
s e
0
/
me 3.1/4.1-13
QUAD-CITIES DPR-29 TABLE 4.1-2 SCRAM lil5TRUMENT CAllBRATION MINIMUM CALIBRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS troup'D e
Cailtration Standergs) ghannum Frequency'D lustruinest Chemmet Comparison to APRM after Every controlled shutdown'8)
Hgh lux IRM C heat balance Ngh Scx APRM B Heat balance Once every 7 days Output senei B Standard pressure and voltage Refueira outage Flow bias source B4' Usmg Tir system Every 1000 equivalent full LPRM power hours A Sta-dard pressure source Every 3 conths Hgh reactor pressure A Standard pressure source Every 3 monns Hgh drywell pressure A Water level Every 3 months Reactor low water level A Standard vacuum sour:e Every 3 months Turbne condenser low vacuum B Approprete radaten source (3' Refuelog outage Mari steamine hgh radaten A Pressure source Every 3 months Turbne EHC control Sud low pressure HigNater level in scra:n discharge A Water level Refueling Outage volume (dp only)
Notes:
L A descripton of the three groups a ecluded m the bases of this specificaten 2 Cahtssten tests are not requeed when the systems are not requeed to be operable or are tripped if tests are mssed. they shall tr performed pror to returnerg the systems to an operable status 3 A curent source provil'es an estrument channel shgnment every 3 mor ths 4 Gianmum cateraten frequency need not exceed once per week
$ Response twne is not part of the routme mstrument check and cabbraton but udt be checked every refuelmt outage 6 Does not provide scram functen g
3.1/4.1-14
QUAD-CITIES DPR-29 3.1 LIMITING CONDITIONS FOR OPERATION BASES The reactor protection syetem 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 orservice because of maintenance. When necessary, one channel may be made inoperable for briefintervals to con 6 ct required functicnal 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.. and input signal an either one or both of the subchannels will cause a trip system trip. The outputs of the trip systems are arranged so that a trip ori both systems is required to produce a reactor scram.
This system meets the requirements of the IEEE 279 Standard for Nuclear Power Plant Protec' ion Systems issued September 13.1966.The system has a reliability greater than that of a two-out-of three system and somewhat less than that of a one-out-of-two system (referenca APED $179).
With the exception of the average power range monitor (APRM) and intermediate range monitor (IRM) channels each subchanael has one instrument channel. When the minimum condition for operation on the 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
- I and m3 operate contacts in one subchannel, and APRM's 22 and #3 operate contacts in the other subchannel. APRM's *4. # 5. and z6 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 IRM 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 contro! valve fast closure, and turbine stop valve closure are discussed in Specifkations 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-level instrumentation which is discussed in Specification 2.1. A scram is provided at the same setting as the emergency core coolirrg 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
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 turbine'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 the resulting transient is less severe. Scram occurs at 21 inches Hg vacuum, stop 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 background. The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent excessive turbine 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 condenser off-gas line provided the limit specified in Specification 3.8 is exceeded.
The main steamline isolation valve closure scram is set to scram when the isolation valves are 10% closed from full open. This scram anticipates the pressure and flux transient which.would occu! when the valves close.
-By scramming at this setting, the resultant transient is insignificant.
A reactor 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 flexihility during startup and to allow repairs to be made to the turbine condenser. While this bypass is in effect, protection is provided 'against pressure or flux incre4Ses 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 condenset, the main steamline isolation valves would be closed.
No hypothesized 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, thus providing for a manual means of rapidly inserting control roos during-all mooes of reactor operation.
The IRM system provides protect jan against excessive power levels and short reactor periods in the startup and intermediate power ranges (referer:ce SAR Sections 7.4.4.2 and 7.4.4.3). A source range monitor (SRM) system is also provided to supply additional neutron level information during startup bui has no scram functions (reference SAR Section 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 t APRM 15% scram as discussed in the bases *or Specification 2.1. In the power 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 APRM's cover only the intermediate and power range; the IRM's provide adequate coverage in the startup and intermediate range.
The high-reactor pressure, high-drywell pressure, reactor low water level, and scram discharge volume high level scrams are required for the
..Startup/ Hot Standby and Run modes of plant operation. They are therefore required to be operational for these modes of reactor operation..
'The turbine condenser low-vacuum scram is required only during power operation and must be bypassed to start up the unit.
3.1/4.1-3 Amendment No. 66
, QUAD-CITIES DPR-29 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 movir.g; e.g. , the thermal swltches in the scram discharge volume tank. Based on the above, no calibration is required for these instru' ment 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 process computer calculations.
Only a small number of control rods are moved daily, thus the peaking factors are not expected to change significantly and a daily check of the MFLPD is adequate.
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, i
i 3.1/4.1-7
[. Amendment No. 61
,-r- - , - - - , , , , --, .,,...-~.-----.-r, , , , - - - - --,,n-. . - ~ - - . - - - - - - - . - - + - - - , . - - - , - - .
I The control rod drive scram system is designed so that all of 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 two volumes and their piping can hold in excess of 90 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 accommodated during a scram. During normal operations, the discharge volumes are empty; however, should either volune fill with
-water, the water discharged to the piping from the Reactor may*not be accommodated which could result in slow scram timet 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 incorporated 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 gpm leakage per drive into the SDV. As indicated above, there is .
sufficient volume in the piping to accommodate the scram without impairment- of the scram times or the amount of insertion of the control rods. This function shuts the Reactor down while sufficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be requit ed but not be able to perform its function properly.
I i'
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i l
I 3 1/4.1-2a p:
- ~ _._ _- __ -.
QUAD-CmES .
- 1 TAKE 3.24 BtSTRUMENTATION THAt lMITIATES R00 K0CK smisen shedw of Operous er Trtesed lastrumsst Casessh per 1rtp Systen* hatrusset 31, gave! W (2) 2 APRM escale (Gow biasYD 4 0.5Swp'+ 50 N Y.T1.PD 2 APRM upscah (Refuel and Start @/ Hot s12/125 M scae Standby mode) 2 APRM doomscate'4 23/125 M scak 1 Red block montor escale (tow besyD so.'6SWp + 42 I2) 1 Rod block monitor downscale'" 23/125 M scale 3 IIM downscak m to 23/125 M scale 3 554 opscate* s108/125 M scale 28 SRM detector not m Startup posrten'" 22 feet bebe core center.
Ime 3 ftM detector not in Startup postion* 22 feet bebw core center.
ine 28 #1 SRM upscale s 103 counts /sec
( 28 SRM downscale'S kl02 counts /sec 1(perbank) Hgh water level m scram dschsge volurne (SUV) f25gallyperbank) 1 SDV high water level scram NA trip bypassed Notes
- 1. For the Startup/ Hot Standby amd 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, ARTM upsale (flow biased), and RBM downscale need not be operable in the Startup/ Hot Standby mode. The RBM upscale aeed not be operable at less than 30% rated thermal power. One charnel may be bypassed above 30E rated thermal power provided that a limiting control rod pattern does not exist. For systems with more than one channel per trip system, if the first coltsnn 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 immediately and daily thereaf ter; if this condition lasts longer than 7 days the system shall be tripped. If the first column cannet be met for both trip systems, the systems shall be tripped.
i
- 2. WP is the percent of drive flow required to produce a rated core flow of 98 mil; ion Ib/hr. Trip level setting is in percent of rated power (2511 MWt).
- 3. IRM downscale may be bypassed when it is on its lowest ranga.
- 4. This function is bypassed when the count rate is GT/E 100 CPS.
- 5. One of tne four SRM inputs may be bypassed.
- 6. This SRM function may be bypass?d in th? higher IRM ranges (ranges 8,9, and 10) when the IRM upscale rod block .s operable.
- 7. Not required to be operable while performing low power physics tests at atmospheric pressuce during or after refueling at power levels not to eceec' 94Wt .
8.
This IRM function occurs when the reactor mode switch is in the Refuel or Startup/H)t Standby position.
- 9. This trip is bypassed when the SRM is fully inserted.
Amendment No. 70, 84 3.2/4.2-14 L-
7-QUAD-CITIES
- 3. The control rod drive housing support '
- 3. The correctness of the control rod system shall be in place during reactor withdrawal sequence input to the power operation and when the reactor RWM computer shall be verified after
- cociant system is pressurized above loading the sequence.
atmospheric pressure with fuel in the Prior to the start of control rod with-reactor vessel, unless all control rods drawal towards criticality, the capabil-are fully inserted and Spect6 cation ity of the rod worth minimizer to 3.3.A.I is mn properly ful611 its function shall be
- a. Control rod withdrawal sequences verified by the following checks:
shall be established so that max- ,, .fhe RWM computer online diac-imum reactivity that could be a'dded by dropout of any incre-M c n s M k s e sfd -
performed.
ment of any one control blade '
would be such that the rod drop accident - b. Proper annunciadon of the selec.
design limit of 280 cal /gm is not exceeded. tion error of one out-of-sequence c nti i r d shall be verified.
- b. Whenever th'e reactor is in the Start'up/ 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 minimizer drawing the first rod as an out-shall be operable. A second opera- of-sequence control rod no more tor or quali5ed technical person - than to the block point.
may be used as a substitute for an
- 4. Prior to control rod withdrawal for inoperable rod worth minimizer .
which fails after withdravil of at startup or during refueling. verifv that least 12 control rods to the fully at least two source range cha'anels withdrawn position. The rod have an observed count rate of at least werd minimizer may also be three counts per second.
bypassed for low power physics testing to demonstrate the shut-
- 5. When a limitine control rod Pattern down margin requirements of * " "
f" "Specincation 3.3.A if a nuclear th B *l l0** f s : Per ormed pnor to engineer is present and verines the .
' .al of t. e designated rod (s) step.by step rod movements of the a I tw at ~
test procedure.
- 4. Control rods shall not be withdrawn 5. The scram disenarge volume vent and drain for startup or refueljng unless at least valves shall be verified open dt least once
..two source range Channeis have an per 31 days. These valves may be clused observed count rate equal to or greater intermittently for testing under asinistrative ,;ontrol and at least once per than three counts per second and these .a2 ears, eacn valve snali e cycled enruagn SRM T are fully inserted. at least one comlete cycle or rull travel.
At least once each Refueling Outage, the
- 5. During operation wi'.h limiting con-scram discharge volume vent and drain valves will be demonstrated to:
trol Tod patterns. as determined by the a.
nuclear engineer. either: 30 ses & W
' or a signal for control rocs to scram, and
- a. both RBM channels shall **be .
operabM. b. Open when the serain signal is reset.
- b. control rod withdrawal shall be ,
blocked; or 3.3/4.3-3 Amendment No. 57, 84
c ATTACHMENT 2 1
Evaluation of Significant Hazards Consideration Description'of Amendment Request-
, Subsequent to a failure of.76 of 185 control rods to fully insert at Browns Ferry Unit 3 in response to a manual scram signal, the
' Commission had embarked on an indepth review of the BWR control rod drive system which identified a number of. design issues requiring both short and long term-corrective measures. On October 1, 1980 letters were sent to all BWR licensees requesting commitments to reevaluate the present scram system and modifying it as necessary to meet both the design and performance criteria as developed by.the BWR Owners Subgroup.
Accordingly, a confirmatory order was written June 24, 1983 for Quad Cities Unit 1.regarding a schedule for implementation of the long term corrective actions. That Confirmatory Order also provided model technical specification changes. Based on cur final design and upon a review of the model technical. specifications, Commonwealth Edison is proposing a number of changes to Appendix A of the Technical Specification for Quad Cities Unit 1.in accordance with the forementioned Confirmatory Order.
Basis 1for Proposed No-Significant Hazards Consideration Determination The Commission has provided guiance concerning the application of' standards for determining whether a significant hazards consideration exists by providing specific examples. The examples of- actions involving no significant hazards consideration include: (ii) changes that constitute an additional limitation or restriction or control not presently within the technical' specifications e.g., a more stringent surveillance requirement..
The changes proposed in this application for amendment is encompassed by this example because of the additional limitations and
. restrictions _that'will be added by this Technical Specification amendment.
Therefore, since the application for amendment involves a proposed change thatfis similar to an example for which no significant hazards consideration exists, Commonwealth Edison has made a proposed determination that the-application involves no significant hazards consideration.
8131N i