ML20073J408
| ML20073J408 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/14/1983 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20073J401 | List: |
| References | |
| 6373N, NUDOCS 8304190285 | |
| Download: ML20073J408 (23) | |
Text
E ATTACHMENT 1 quad Cities Unit 1 DPR-29 Proposed Technical Specifications Revised
- Pages Previous Amendment No's.
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ATTACHMENT 1 Revised
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6373N l
l I
ATTACHMENT 2 Quad Cities Unit 2 DPR-30 Proposed Technical Specifications Revised
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< 3.8/4.8-9 30 I 3.8/4.8-10 36 3.8/4.8-11 X 3.8/4.8-12 40 l 3.8/4.8-13 40 3.8/4.8-14 40 3.8/4.8-15 40 3.8/4.8-16 40 3.8/4.8-17 40 3.8/4.8-17a Deleted 3.8/4.8-18* ----
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. ATTACHMENT 2 Revised
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6373N
QUAD-CITIES OPR - 29 II. DOSE EQUIVALENT l-131 - That concentration of l-131 (microcurle/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those IIsted in Table ill of Tl0-14844, ." Calculation of Distance Factors for Power and Test Reactor Si tes".
JJ.- PROCESS-CONTROL -PROGRAM (PCP) - Contains .them sampling, analysi s , and formulation determination by which ' solidi fication of' radioactive wastes from liquid systems is assured.
KK. OFFSITE DOSE CALCULATION MANUAL (00CM) - Contains the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, and in the calculation of gaseous and liquid effluent monitor alarm / trip setpoints.
LL. CHANNEL FUNCTIONAL TEST (RADIATION MONITOR) - Shall be the injection of-a simulated signal into the channel as close to the sensor as practicable to verify operability including alarm and/or trip functions.
MM. SOURCE CHECK -'The qualitative assessment of ins trumen t response when the sensor is exposed to a radioactive source.
NN. MEMBER (S) 0F THE PUBLIC - Shall include all persons who are not occuca-tionally associated with the plant. This category does not include employees of the utili ty, its contractors, or vendors. Also excluded from this category are persons who enter the site to service ecui: ment or_to make deliveries. This category does include cersons aho use por-tions of the site for recreational, occupational, or other purposes not associated with the plant.
1.0-5
3.2/4.2 PROTECTIVE INSTRUMENTATION 4
LIMITING,CO2ITIONS"?OR"0PERATION: SUMEILLANCE ? REQUIREMENTS '
Ap'plicability: Applicability:
Applies to the plant instru=entation Applies,to_the surveillance.:equire=ents which performs a protective function. of the instru=entation that performs a protective function.
Objective:
Objective To specify the type.and. frequency.of Totassureathe. operability'of' pro - surveillance to be: applied toaprotective tactive= instrumentation instrumentation.
SPECIFICATIONS A. Primary Containment Isolation A. Pri=ary Contain=en: Isolation Func: ions Functions Instru=entation and logic syste=s shall When pri=ary contain=ent integ- be func:1onally tested and calibrated rity is required, the limiting as indicated in Table 4.2-1.
conditions of operation for the instrumentation : hat initiates pri=ary contain=ent isolation are given in Table.3.2-1.
B. Core and Contain=ent Cooling Sys- 3. Core and Contain=ent Cooling Systens -
te=s - Initiation and Control Ini:iation and Control The IL=1cing :cadi: ions for opera- Instru=entation and
- gi: sys:e=s shal:
tion for the instru=entation : hat be func:ionally :ested and :slibra:ed initiates or controls the core and as indicated in Table 1.2-1.
contain=ent cooling syste=s are given in Table 3.2-2. This instru-
=entation sust be operable when :he syste=(s) 1: initiates or controls L are required :o be operable as L specified in Specification 3.5.
l C. Control Rod 31ock Ac:uation C. Con:rol Rod 31ock Ac:uation
- 1. The li=1:ing condi: ions of opera- Instru=entation and logic sys:e=s shall l.
- ion for :he instru=entation that be func:1onally :ested and calibra:ed i
l i
initiates control rod block are as indicated in Table 0.2-1.
t given in Table 3.2-3.
I
- 2. The =ini=um nu=ber of operable l instru=ent, channels specified in Table 3.2-3 for the rod block
! =enitor =ay be reduced by ane in one of :he : rip sys:e=s for =ain-tenance and/or :esting, provided t+ :ha: this condi: ion does no: last longer chan 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> in any 30-day period. If :his :endition exists for = ore than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30-day period, :he syste= shall be :rippec.
! 3.2/a.2-1
QUAD-CITIES DPR-29 9
D. . Refueling Floor Radiation-Monitors' D. Refueling Floor' Radiation-Moni: ors
- 1. Except as specified in Specifi- The.tvo' refueling floor radiation acni-cation 3.2.0.2, the two refueling. ors shall be fune:1onallyL tested and floor radiation monitors shall be calibrated as indicated in Table .2-1.
operable whenever irradiated fuel
~
Reactor building ventilation isolati:n or' components are present in the and stundby gas :rea:=en: sys:em int:i-fuel storage pool and;during re- ation shall be performed at leas: each fueling.or fuel movement opera- operating cycle.
tions.
- 2. One of.the two refueling. floor
. radiation monitors may be inopera-ble for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the inopera-I N"
ble' sonitor is not restored :o service in this :ime, :he reac:ce building ventilation system shall be isolated.and :ha standby gas creatment' operated until repairs are complete.
-3. The trip setting for :he refueling floor radiation moni: ors shall be se: at a value of 100 2R/hr.
4 'Jpon loss of both refueling floor radiation moni: ors while in use, the reac:or building ventilation system shall be isolated and :he standby 3as :reatment operated.
E. Postacciden: Ins:ru=en:ation E. Postaccident :ns:rumen:a:1:n Th'e limizing condi: ions for opera: ion ?os:acciden: ins rumen:a: ion shall be for the instrumenta: ion which is read func:icnally :es:ed and calibra:ed 2s out in the control room, required for inci:2:ed in Taole . 2-2.
.postaccident-conitoring are given in Table 3.2-4
, 3.2/'.2-2
QUAD-CITIES OPR-29 F
F. Control Roon Ventilation System F. Control Room Ventilation Systen isolation isolation The control room ventilation' system Surveillance for instrumentation is isolated from outside air on a which initiates isolation of signal of high drywell pressure, control room ventilation shall be
. low water level, high main stream- as speci fied in Table 4.2-1.
line. flow, or-hightradiation in-either;of.theireactor' building ventilation ~ exhaust ducts.
Limiting conditions.for operation shall be as indicated in Table 3 2-1 and Speci fication 3 2.H. l .
G. Radioactive Liquid Effluent G. Radioactive Licuid Ef#1uent Instrumentation Instrumentation The effluent monitoring instru- Each radioactive liquid effluent mentation shown in Table 3.2-5 monitoring instrument shown in shall be operable with alarm set- Table 4.2-3 shall be demonstrated points set to ensure tha t the limits operable by performance of the {
of Speci fication 3.8.3 are not given source check, instrument check,I exceeded. The alarm setpoints calibration, and functional test shall be determined in accordance operations at the frequencies with the 00CM. shown in Table L.2-3
- 1. With a radioactive liquid effluent monitoring instrument alarm / trio setpoint less conservative than required, without delay suspend the release of radioactive liquid effluents moni tored by the af fected instrument, or declare the instru-ment inoperable, or change the set-point so i t is acceptably conserva-tive.
- 2. With one or more radioactive licuid effluent' monitoring instruments inoperable, take the action shown in Table 3 2-5 Exert best efforts to return the instrument to operable status within 30 davs and, i f un-successful, explain in the next Semi-Annual Radioactive Effluent Release Report why the inoperability was not corrected'in a timely manner. This is in lieu of an LER.
3.2/4.2-3
= -
QUAD-CITIES OPR 29 e
~
3 In the event a limiting condition for operation and associated action requirements cannot be satisfied because of circumstances in excess of_those addressed in the specifi-cations, provide a 30-day written
- report to the NRC-pursuant to Specification 6.6.B.2., and no changes!are required in the opera.
tional _ condi t ion of the -plant , and
- this does not prevent the plant from entry'into an operational mode.
H. Radioactive Gaseous Effluent H. Radioactive Gaseous Effluent instrumentation Instrumentation The effluent monitoring instru- Each radioactive gaseous radiation mentation shown in Table 3.2-6 shall monitoring instrument in Table be operable with alarm / trip'setpoints _4.2-4 shall be demonstrated coerable set to ensure'that the limits of by performance of' the given source Speci fication 3.8.A. are not exceeded. check, instrument check, calibration, The alarm / trip setpoints shall be and functional test operations at determined in accordance with the- the f requency shown in Table 4.2-4 00CM.
- 1. With'a radioactive gaseous effluent monitoring-instrument alarm / trip setpoint less conser-
.vative than required, without delay suspend the release of radioactive gaseous effluents monitored by the affected instru-ment, or declare the instrument inoperable,or change the setpoint so it is acceptably conservative.
- 2. With one or more radioactive gaseous effluent monitoring instruments inoperable, take the action shown in Table 3.2-6. Exert best efforts to return the instrument to operable status within 30 days and, i f unsuccess ful, exp'ain in the next Semi-Annual Radioactive Effluent Release Report why the inoceracility was not corrected in a timely manner.
This is in lieu of an LER.
, 3 2/4.2-4
QUAD-CITIES OPR-29 3 In the event a limiting condition for operation and associated action requirements cannot be satisfled because of circumstances in ; excess of those addressed in the specifications, provide a 30-day written > report to the NRC pursuant to the Specification 6.6.B.2., and no changes are required in the operational con-dition of the plant, and this does not prevent the plant from entry into an operational mode.
?
1 3 2/4.2-5
QUAD CITIES DPR-29 3.2 LIMITING CDOITION FOR OPERATION BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consecpences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, control rod block and standby gas treatment systems. The objectives of the specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance and (2) to precribe the trip settings required to assure adequate performance. When necessary, one channel may be made inoperable for brief intervals to conWet required functional tests and calibrations. Some of the settings on the instrumentation that initiates or control cora r.nd contaJnment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. It should be noted that the setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by the protective instrumentation which serves the condition for which isolation is required (this instrumentation is shown in Table 3.2.1). Such instrumentation must be available whenever primary containment integrity is required. The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded during an accioent.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
here.
Thus the discussion given in the basis for Specification 3.1 is applicable The low reactor level instrumentation is set to trip at > 8 inches on the level instrument (top of active fuel is defined to be 360 inches above vessel zero) and after l allowing for the full power pressure drop across the steam dryer the low level trip is at 504 inches above vessel zero, or 144 inches above the top of active fuel. Retrofit 8x8 fuel has an active fuel length 1.24 inches longer than earlier fuel designs.
However, present trip setpoints were used in the LOCA analyses *. This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the l recirculation pu@s (reference SAR Section 7.7.2). For a trip setting of 504 inches above vessel zero and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum break: the setting is therefore adequate.
The low low reactor level instrunentation is set to trip when reactor water level is 444 inches above vessel zero (with top of active fuel defined as 360 inches above vessel zero, -59 inches is 84 inches above the top of active fuel). This trip initiates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subtstems starts the emergency diesel generator, and trips the recirculation pug s. This trip setting level was chosen to be high enough to prevent spurious operation but low enough to initiate ECCS operation and primary system isolation so that no mciting of the fuel cladding will occur and so that postaccident cooling can be accoglished ano the guidelines of 10 CFR 100 will not be exceeded. For the coglete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary isolation are initiated and in tire to meet the above ctiterib. The instrumentation also covers the full spectrum of breaks and meets the above criteria.
The high-drywell pressure instrumentation is a backup to the water level instrumentation and, in addition to initiating ECCS, it causes isolation of Group 2 isolation valves.
For the breaks discussed above, this instrumentation will initiate ECCS operation at about the same time as the low low water level instrumentation; thus the results given above are applicable here also Group 2 isolation valves include the drywell vent, purge and sum isolation valves. High-drywell pressure activates only these valves because high drywell pressure could occur as the result of non-safety-related causes such as not purging the drywell air & ring startup. Total system isolation is not desirable for these conditions, and ordy the valves in Grom 2 are required to close. The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip of Group 1 primary system isolation valves.
- Loss of coolant accident analysis for Dresden thits 2 & 3 and QJad Cities thits 1 & 2, NEDO-24146A, April,1979 6312N 3.2/4.2-5a AMENDMENT NO.
QUAD CITIES DPR-29 The APRM rod block function is flow biased and prevents a significant redJction in MCPR, especially during operation at reduced flow. The Am M provides gross core protection, i.e., limits the gross withdrawal of control rods in the normal withdrawal sequence.
In the refuel and startup/ hot stan@y 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 ROM rod block function provides local protection of the core, i.e., the prevention of transition boiling in a local region of the core for a single rod 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 assure that, with the specific trip settings, rod withdrawal is blocked before the PCR reaches the fuel cladding integrity safety limit.
Below 305 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 level. Analysis is such that the trip setting is less than a factor of 10 above the indicated of the worst-case accident results in rod block action before MCR approaches the MCPR fuel cladding integrity safety limit.
A downscale indication on an APRM is an indication the instrunent 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 $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 operator action. The alarm setpoint has been selected to provide adequate time to allow 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 if'CI system must function since to reactor operate pressure does not decrease rapidly enough to allow either core spray or LPCI in time.
The automatic pressure relief function is provided as a backup to the if'CI in the event the HPCI coes not operate. The arrangement of the tripping contacts is such as to provide this function when 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 l
l 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 trip logic is one out of two. Trip settings of 100 mR/hr for the monitors on the refueling floor are based upon initiatirg normal ventilation isolation and stan&y gas treatment system operation l
3.2/4.2-7 Amenchent No.
6312N l
l T ~4+-
' QUAD-CITIES DPR-29 so that none of the activity released during the refueling accident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby 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 monitoring of the containment following a loss-of-coolant accident. Information from this instrumentation naill 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 comes from 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 controi 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 steamline high flow, or high radiation in the reactor building ventilation duct, will cause isolation of the intake air to the control room. The above trio signals result in immedi-ate isolation of the control room ventilation system and thus minimize any radia tion dose.
The radioactive liquid and gaseous effluent instrumentation is provided to monitor the release of racioactive materials in liquid and gaseous ef'!uents during releases. The alarm setpoints for the instruments are rovided to ensure that the alarms will occur prior to exceeding the limi ts of 10 CFR 20.
s l
3 2/L 2-3 i --- - - - -
quad-CITIES DPR-29 Table 3 2-5 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum No.
of' Operable Total-No.
Channels of Channels Parameter' Act'icn ; (1 ) ,
1- I Service Water A Effluent Gross Activity Monitor l' 1 Liquid Radwaste C Effluent Flow Rate Monitor 1 1 Liquid Radwaste B Effluent Gross I Activity Monitor 1 I Soray Canal Discharge C f, Blowdown Flow Rate Monitor-Notes Action A: With less than the minimum number of operable channels, releases via this pathway may continue, provided tnat at-least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samoles are collected.and analyzed for beta or gamma-activity at an LLD of less than or equal to 10~/ uCi/ml .
l Action 3: With less than the minimum number of operable channels, ef fluent releases via this pathway may continue, provided that prior to initiating a release, at least 2 independent samples are analyzec,
! and at least 2 members of the facility staff indeoendently verify the release calculation and discharge valving. O the rw i se , 's uscend release of radioactive ef fluents via this oathway. i Action C: With less than the minimum number of operable Onannels, releases via l this pathway may continue, provided the flew rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. ?umo curves may be utilized to estimate flow.
f i
S.
3 2/4.2-156
QUAD-CITIES DPR- 29 Tablee3.2-6' RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
-Minimum N6. -
cf Operable Total No,
' Channels-(l) of Channels Parameter Ac t ion (2)
.ll 2 SJAE Radiation O' Mon i to rs' I 2 Main Chimney Noble A Gas Activity Monitor 1 1 Main Chimney lodine C Sampler l ~1 Main Chimney C Particulate Sampler 1 1 Reactor Bldg. Vent 3 Sampler Flow Race Monitor
~
l 1 Reactor 31dg. Vent C lodine Sampler
_. I 1 Reactor Bldg. Vent C Particulate Sampler 1 1 Main Chimney Samoter 3 Flow Rate Monitor i 1 Main Chimney Flow -3 i
Rate Monitor 1 2 Reactor Sidg. Vent E Noble Gas Monitor Notes
('1) For SJAE monitors, applicable during SJAE operation. For cener instrumentation, applicable at all times.
(2). Action A: With the number of ooerable channels less :han :ne minimum recuire-ment, e f fl uen t releases via this cathway may con t i nue , p rov i de gra samples are taken at least once per 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> shift and :nese samoles are analyced within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Action 3: With the number of operaole channels less : nan :he ninimum re uirec, effluent releases via this catnway may continue rovidec :na: :me flow rate is estimated at least once :er L Scurs.
3.2/4.2-15c
F
-QUAD-CITIES OPR-29*
J Action C: With less than .the mic' mum channels operable, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment, as required in Table 4.8-1.
Action 0: With'less than-the minimum channels' operable, gases.from the main condenser off gas system.may be released to the environment for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided at least.one chimney monitor is operable; i otherwi se, be in hot cs tand-by. in .12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />;
-Action Ei Wi th :less ' than the minimum channels ., operable , immediately suspend release of radioactive ef fluents via :this' pathway' .
3.2/4.2-15d
- y. . .m g p rwyyyte A .x f p ,%, s. , f_ .
'OPR-29
,N . 4 ~
'\
d [ ;rf > "
w Tcble 4. 2-1_(Cont'd)
,s.x . , ~ ,4 m w . .>
g'iA6 .ac " - '
instrument g ' C - ' I ns t rurnen th /
instrgnt
- fynAel W=% fIl x Funct2)
Test i(onal Calibration (2) Check t ..g e ,,
- - g w . .i -
'HPCI lsolatio.n * ;
r g
, g, e. - *
, I. A tesmline.high flow -(1) Once/3 months None
- 2. "Ste$ mil,pMarea high temperature Refuelinglutage Refueling Outage None
':3.-j.cwreaftorpressure .(1) Once/3 monthsw- None-.
- s. >
- i. y - ,
. s i'
, Reactor Building Vent isolation and SBGTS-Initiation
- 1. Refuel Floor Rad. Monitors (1) dnce/3 months once/ day
+d'
?
Control Room. Ventilation System.lsolation J W V'
% u Q jg %
- 1. Reactor low water level (1) @ ice /3 months Once/ day 4- '2. .Orywell high pressure (1) , Once/3 months None
- 3. MainT s teamline ' high flow (1) Once/3 months Once/ day.
s .. 7 . l
./ % ., l y.
Notes
+
_1,- '#
s.
v
- 1. I n i t i a l-1 once o[r month until exposurg hou'Is (M as defined on Figure 4.1-1) are 2.0' Q 0,; thereafter, according to figure,4.1-1 with an interval not less than 1 mch.tgor. m$retthan 3 months.N The concilation of instrument fai. lure rate data
, may lyclude ' data obtaired fromo othe'? boiling water reactors for which tne- sa.9e des i g'n? ins t rument operates . io af, 'pvi ronment similar to that of Quad-Cities Un i t s ' I arid J'. <v
. */*/
. e -
. /
- 2. Functiona!. tes tsi calibrations , and instrument enecks are not required wnen these in s t rumen t'i a ~re .,no t requi4 ^~ d toc.b'e operiole, or are tripoed.
~- . .
^
-3 This instrumentation is excepted from the func;ional test definition. The functional test shall consj,st of fnjecting c, simula ted elect rical signal into
'e ~'
-the measur56 channe1. ~
. %~ ,
- 4. ThisinsirhShtchannelisexceptedEroEthefunctionaltest definitions and shall
,- be~ calibrathd using si+iulated electrical siga!S 7 once every 3 months.
- . w. - . .y 'w
. 5. Func t i onas l ce .e .'s t s s ha l l be pe r fo rmed be fo r[ .each :startup wi tn a recui red f requency not to er;eed'once per week. Calibrations shat'l be performed during eacn startuo or during", controlled shutdowns with a recuired frequency not to excesc once per ,
.week. (;,
- 6. The poFiti[d+ing mechanism shall be caMbrated every refueling outage.
7 .~ Logic sys tem"fuiictional tests are performed as s:ecified in t e acclicao!e sec-tion for these systems.
3.2/h.2-17
- - - . . .-~ . . - . - -_ - - .-.
-QUAO-CITIES-
, , , -OPR-29 g , Tablee4.243' RADI0 ACTIVE Liqul0 EFFLUENT MONITORING lNSTRUMENTATION SURVEILLANCE" REQUIREMENTS-Instrument .
, Functional Source Instrument. . Checi: (1) Calibration -(l) (3) . .- Tes t ' (1 ) (2) 1 Check (1)~
l Liquid (RadwesterEffluent
0l Rv Q *,( 7 ) - -(6)1
- Grcss-Activity Monitort
- Ssrvice-Water. Effluent- :0' Rc Q=(7). .R
. Gross-Activity Monitor Liquid ~Radwas te 'Ef fluen t (4) R NA NA
- Flow - Ra te Mon i tor -
Slowdewn' Flow Rate Monitor (4) R MA , NA
- Spray _ Canal Olscharge
' No tas :
(1)~ 0 = once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
.M = once per 31 days Q.= once per 32 days R = once ser 18 months S = once per 6- months (2) The ' ins trument Functional Test shall also demons tra te tha t con trol room alarm annunciation occurs, .if fany of the following conditions exist, where aoolicable,
- a. Instrument indicates levels above the alarm setooint,
- b. Circuit ~ failure.
- c. instrument indicates a downscale failure,
- d. Instrument controls not set in OPERATE mode.
-(3).' Calibration,shall include performance of a functional test.
.(4).. instrument ~ Check to veri fy flow during periods of release.
1($) Calibration shall include performance of a source check.
I- (6) LSource check shall consist of observing instrument response during a disenarge.
((7)- Functional _tes t may be performed bv using trio eneck and test circuitry associatec with the monitor chassis.
3.2/4.2-19
h LQUAO-CITIES
'OPR-29' Table 4.2 RA010 ACTIVE GASEOUS EFFLUENT MONITORING' lNSTRUMENTATION SURVEILLANCE. REQUIREMENTS-Instrument Calibra . Functional Source-Instrument Mode (2) Check (1)' tion (l)(4)_ ' Tes t(!)(3) check (l)
EniniChimneyJNoble< Gas > n D. ER Q '. M-
~
Activity-Monitor z
Main'ChimneyLSampler B: O' R (y(6)' NA-Flow - Ra te ' Mon'i to r .
R:cstor-Sidg. Vent' Sampler ~ B D R gg(6) NA-tFlow Rate Monitor Main Chimney Flow' Rate .B 0 R Q NA-Monitor 1Rasetor. Bldg. Vent 3 0 R Q LQ-Activity Monitor
~SJAE ' Activi ty Moni tor' A 3 .R Q R
[^ Main Chimney lodine and -
B 0(5) 34 34 33
' ' Particulate Sampler Raactor Bldg. Ven t lodine B 0(5) 33 ng ng cod Particulate Sampler Notes--
- ( ): 0 =.once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-M.= once per 31 days. .
Q = once per 92 days R = once per-18 menths
-(2) A = during SJAE coeration 3: at all times (3) The In= trumen t Functiona l Tes t shal l a lso demons't ra te tha t con t rol room alarm annunciation occurs , i f any of the folloaing condi tions exis t, wnere acclicao!e:
- a. Instrument indicates levels above the alarm se tpoi n t
- b. Ci rcui t failure
- c. Instrument indicates a downscale failure
- d. Instrument controls not set in OPERATE mode (4) Calibration shall include performance of a functional test.
~(5) ins trumen t check ' to veri fy ope raci l i ty of samp ler; :na: t e sampier is i r.- p l ace
. and fu'nctioning procerly.
(6): Functional - tes t shall be cerformed on local swi:cnes crevi:;rg ;>. ?? u alarm.
3.2/L.2-20
. .a. , , _ __ . _ _ _ -_ .. _ _ . - . _ _ _ _ __
QUAD-CITIES' DPR-29 3.8/4.8 RADICACTIVE EFFLUENTS Limiting Conditions for . Operation Surveillance P-equirements fApplicability: Applicability:
Applies to the radioactive effluents Applies to the periodic measurements from the. plant. radioactive effluents.
Speci fi ca t ions A. Gaseous Effluents A. Gaseous Effluents
- 1. The dose rate in unrestricted areas at 1. The dose rates due--to radioactive or beyond the site boundary (Figure materials released in gaseous l'
4.3-1)'due to radioactive' materials effluents from the site shall be released in gaseous effluents from -determined to be within the pre-the site shall be limited to the scribed limits by obtaining recre- _
following: sentative samples in accordance with the samoling'and analysis
- a. For Noble Gases: program speci fied in Table 4.3 :
Thedoseratesarecalculatedusingl (1) Less than 300 mrem / methods prescribed in the Off-Sice year to the whole Dose Calculation Manual. (00CM) .
body.
i (2) -F Less than 3000 mrem /
year to the skin. I
- b. For iodine-131, for iodine -133, and for all radionuclides ~in car- :
ticulate form with half-lives greater than 3 days less than 1500 mrem / year. i
- c. If the dose rates exceed the ,
above-limits, without delay decrease the release rates to bring the dose-rates within '
the limits, and provide prompt 2, The air dose due to releases of notification to the Commission radioactive noble gases in gaseous ;
(6.6.8.1.) effluents snall be determined to
, be within the :rescribed limits by
- 2. The air dose in unrestricted areas at obte:aing representative samoles
'or beyond the sirte boundary due to in accordance witn the samoling Noble Gases released in gaseous effluents and analysis program soecifisc from~the unit shall be limited to the in sections A anc 3 of Table L.3-!.
l following: The allocation of ef fluents between units naving shared effluent con-
- a. For gamma radiation: trol systems and the air cases are determinec using metnocs crescribec (1) Less than or equal to in the 00CM st least :nce everv 31
- 5 mead during any cal- cays.
endar quarter.
3.3/4.3-1 e - -- r- - r m= -<e-w-+-e se e - ,v , - - - - -~r-
y QUAD-CITIES-DPR-29 (2) Less then or equal to 10 mrad during any calendar year.
t
- b. .For Beta radiation:
(1) Less thah or' equal to 10 mrad;during+any cal -
endar quarter (2). Less'than orcequal to~
20 mrad during any cal-endar year.
- c. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, pre-pare and submit to the Com-mission within 30 days, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the cor-rective actions to be taken to ensure that future releases are in' compliance with 3.3.A.2.a. S b.
This is in lieu of a Licensee Even t Report.
- d. With the calculated air dose from radioactive noble gases in gaseous ef fluents exceeding the limits of Specification
'3.8.A.2.a. or 3.8.A.2.b., prepare and submit a Special Report to the Commission within 30 days and limit the subsequent re-leases such that the doses or dose commitment to a memoer of the public from all uranium fuel cycle sources is limited to less than or equal to 25 meem to the total body or any organ (except thyroid, which is, limited to less than or equal to 75 meem) over 12 consecutive months.
This Special Report shall include an analysis which demonstrates that radiation exoosures to all emeers of the public from all uranium fuel cycle sources (including all efflu-(; ent pathways and direct radiation) are less than the 40 CFR Part 190 Standard. Otherwise, obtain a 3.3/h.3-2
QUAD-CITIES DPR-29 variance :from, the' Commission to permit releases which exceed the 40 CFR Part 190 Standard. The radiation exposure analysis contained in the Special Re-port shall'use the methods prescribed in the ODCM. This report is in lieu of a Licensee Event Report.
-3.. The dose to a member of thecpublic.in 3. The dose to,a3 member.of.the pub-unrestricted areas at or beyond'thecsite lic-due to releases of iodine-131, boundary from-iodine-131, iodine-133, iodine-133, tritium, and all radio-tritium, and all radionuclides in nuclides in particulate form particulate form wi th hal f-lives greater with half-lives greater than 8 than 8 days in gaseous effluents days shall be dete'rmined to be-released from the' unit-shall be limited within the prescribed limits by to.the following: obtaining reoresentative samoles in accordance with the sampling
- a. Less than or equal to 7.5 mrem and analysis program soeci fief n to any organ during any calendar Table 4.3-I.
quarter.
For radionuclides not determined
- b. Less than or equal to 15 mrem to in each batch or weekly comoosite.
.- any organ during any calendar the dose contribution to the cur-f year. rent calendar cuarter cumulative summation may be estimated by
- c. With the calculated dose from assuming an average monthly con-the release of: iodine-131, iodine- centration basec on tne crevious
~l33, tritium, and all radionuclides monthly or cuarterly composite in particulate form wi th hal f-lives analyses. However, for reoorting graater-than 3 days in gaseous purposes, the calculated dose effluents exceeding any of the contributions shall be based on above limits, prepare and submit the actual comoosite analyses
- > . to the Commission within 30 days, when possible.
a Special Report which identifies the cause(s) for exceeding the The ellocation of effluents cetween; limit and defines the corrective units having shared effluent con-actions taken and the proposed trol systems and the doses are actions to be taken to ensure determined using the methocs pre-that future releases are in scribed in the 00CM at least once compliance with 3.3.A 3. a. 5 b. every 31 days.
This is in lieu of a Licensee Event Report.
3.3/4.3-3
~.
QUAD-CITIES DPR-29
- d. With -the calculated dose from the release of iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents exceeding-the limi ts of Speci fication 3.3. A 3.a.
Eor 3.8.A.3 6., prepare and submit.
a Special Report to'the Commission within 30 days-and limit subsequent.
releases such that the dose or' dose commitment to a member of the public
. f rom all uranium ~ fuel cycle sources
.is limited to less than or equal to 25 mrem to the total body or organ (except the thyroid, which is limited to less than or equal to 75 wa em) over 12- consecutive months.
This Special Report shall include an analysis which demonstrates that radiation exposures to all members of the public from all uranium fuel cycle sources (including all ef-fluent pathways and direct radi-([ ation) are less than the 40 CFR Part 190 Standard. Ocherwise, obtain a variance from the Commission to permit releases which exceed the 40 CFR Part 190 Standard. The.
radiation exposure analysis con-
.tained in the Special Report shall use the methods prescribed in the 00CM. This report is'in lieu of a Licensee Event Recort.
3.3/h.3-L
' quad-CITIES DPR-29
- 4. Off-Gas System h. Off-Gas System
- a. At all times during . pro- Doses due to treated' gases cessing for discharge to released.to unrestricted areas the envi rons , process and at or beyond the si te- coundary control' equipment provided shall be projected at least once to reduce,the? amount.:or opere31;daysein accordance with concentratlon of radio- the 00CM.
active materials shall be.-
operated?
- b. The above specification shall not apply for the Off-Gas Charcoal Absorber Beds below 30 percent of rated thermal power.
5 Explosive Gas Mixture 5. Explosive Gas Mixture
- a. The concentration of hydrogen Once per 3. hours verification in the of f gas hold up system, will be made'that the unit is downstream of the recombiner operating within the allowable shall be limited by having a band of the base-line plot of recombiner operable within recombiner outlet temperature the allowable band of the vs.-reactor power.
base-line plot of recomoiner outlet temperature vs. reactor power, whenever the reactor is operating at a pressure greater than 900 psig.
- b. The recombiner may be inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
- 6. With~either the recombiners inoperable, or all charcoal beds bypassed for more than 7 days in a calendar quarter while-operating above 30.cercent of rated thermal power, prepare and submit to the Commission within 30 days a soecial L report which includes the following in fo rma t ion :
3.3/h.3-5 i
QUAD-CITIES DPR-29
- a. Identi ficatibn of'the Jdefe'ctive -
equipment.
- b. Cause of the defective equipment.
I'
- c. Action (s)' taken to restore the equipment to an operating status.
- d. : Length of' time the above-requirements were not' satisfied.
- e. Volume and' curie content of the waste discharged which was not processed by the inoperable equipment but '
which reluired processing.
- f. - Action (s) taken to prevent a recurrence of equipment failures.
This is in lieu of a Licensee Event Repo rt .
S. Liquid Effluents B. Liquid Effluents.
- 1. The concentration of radioactive 1. The concentration of radio-material released from the site active material in unre-
.to unrestricted areas at o'r beyond stricted areas shall be the site boundary (figure h.3-1) determined to be within the shall be limited to the concen- prescribed limits by ootain-trations speci fied in 10 CFR Part ing' representative samoles 20, Appendix B, Table II, Column 2. in accordance witn the samoling and analysis program soecified in Table 4.3-3 The samole analysis results will be used with tne calculational methods in the 00CM to tetermine that the ccncentrations are aithin With the concentration of the limits of Soeci fication radioactive material' released 3.3.B.1.
from the site to unrestricted areas exceeding the above limits, without delay decrease the release rate of radioactive materials and/or increase the dilution flow rate to restore the concentration to within the above limits.
3.3/h.3-6
- e. --
..w- e - - . -.- -
,, , , , y-
k :
-QUAD-CITIES s, DPR-29
'2. : Th'e ' dose or dose . commitment 2. a'. The dose contributions from Labove background to a member measured quantities of of 'the public' from radioactive . radioactive' material shall materials in. liquid effluents be determined by ca!culation 7 "~
_ . released'to unrestricted. areas at least once per 31 days
'at or beyond thel site boundary .
and a cumulative summation from each unit shall be limited of these total body and'any to s the- fol lowing : organ > doses +shall be maintained for'each: calendar quarter;
- a.m .During : any-. calendar quarter:
. (l) Less than or equal. to 3 mrem to the whole body.
(2k Less than or equal to 10 mrem to any organ.
- b. During any calendar year: .b. Doses computed at the nearest community water system will (I) Less:than or equal to 6 consider only the-drinking mrem to the whole body. water pathway and shall be projected using the~ methods (2) Less than or _ equal to 20 prescribed in the 00CM at
. mrem to any organ. least once per 92 days, f, ~
- c. With the calculated dose from the-release of radioactive materials in liquid ef fluents exceeding any of the aoove.
..l imi t s , prepare and submi t to the Commission within 30 days a Special Report which identi-fies . che cause(s) for exceed- .
ing the limit (s)- and. defines tha corrective actions'taken and the proposed actions to be taken to ensure that future -
releases are in compliance with 3.8.3.2.a. s b. This is in lieu of a Licensee Event Report.
- d. With the calculated dose from the release of radioactive materials in liquid effluents exceeding the limits of Speci ficat ion 3.3.8.2.a. or 3.3.3.2.b., prepare and submit
~
a.Special Report to the Commis-sion within 30 days and limit the subsequent releases such that the dose or dose commi t-ment to a real individual from all uranium fuel cycle sources is limited to less than or 3.3/4.3-7
QdAD-CITIES
~
equal co 25 mrem to the-total
- body or any organ (except thyroid,.
which is limited'to less than or equal ~to 75 mrem) over consecutive months. This Special Report shall' include an analysis which demonstrates that - radi-ation exposures to all.real individuals from-all uranium fue l ' cycle i sou rces , (Including all effluent pathways and direct- -
-radiation) are less than the 40 CFR Part 190 Standard. Other-wise obtain a variance from the Commission to permit releases which exceed the 40 CFR Part 190 Standard. The radiation exposure analysis contained in the Special Report shall use methods pre-
- scribed in the 00CM. This report
- is in . lieu of a Licensee Event Report.
- e. With the projected annual
- whole body or any internal organ dose comouted at the nearest downstream community-water system is equal to or exceeds 2 mrem from all radioactive materials re-leased.in liquid effluents 1
from the Station, prepare and submit a Special Report within 30 days to the opera-torlof the community water system. The report is prepared to assist the opera-tor in meeting the require-l
'ments of 50 CFR 141: EPA Primary Orinking Water Stan-dards. A copy of this report will be sent to the NRC.
This is in lieu of a Licensee Event Report.
3.3/4.3-7a e . ,,4- - - -.,--.,w.- y- - , , _ , ,c.,,y--- - - + + e-*- - = - ' = , - - - * - = r t'
QUAO-CITIES OPR-29 L31 At:al1Jtimes during processing Liquid atas te Treatment
~
3.
-prior ~ to. discharge to the envi rons ,
s process and. control equipment pro- a. Ooses'due to liquid releases vided to ' reduce the amount or con- to-unrestricted areas at or centration of radioactive materials beyond the s i te boundary shall' shall be operated .when the projected be projected at leas t once pe r.
' dose :due to liquid effluent releases . 31 days in 'accordance wi th the - l to / un res t ri c ted . a re'as (see-Fi gure 00CM; i 4.8-1).,,whensaveraged over 31 days,-
exceeds' 02.13, mrem' to the, total body.
or 0.42. mrem to any organ..-
4.. I f liquid. w'as te f has- to be or is being discharged wi thout treatment
. as - requi red above , prepare and sub-
' mi t-.to the Commi ssion wi thin . 30 days , a report which' includes the following information:
- a. Identification of the defective equipment.
- b. . Cause of the defective equip-ment.
- c. Ac'. ion (s) taken to res tore the equionent to an operating s ta tus.
d.. Length of time the aoove recui re-ments were not sati s fied.
- e. -Volume and curie content of the
. waste discharged which was not processed by. the appropriate equipment but which requi red processing.
- f. Action (s) taken to preven t a recurrence of equipmen t failures.
l This is-in lieu of a Licensee Event Report.
1 3.3/4.3-3
QUAD-CITIES DPR- 29 '
E.
C . Mechanical: Vacuum Pump C. Mechanical Vacuum Pump -l 1~. ~.The' mechanical vacuum At least.onca during each
'shall be capable of operating cycle, automatic-being isolated and . securing and isolation or secured on'a signal of the mechanical vacuum pump main steam high radi- shall-be verified.
ation or shall be. iso-laced and secured' ,
.whenever the main steam isolation; valves are
~open.
1 1
4 e
w 3.3/a.3-9
-QUAD-CITIES.
.0PR-29
- 0. EnvironmentalLMonit'oring Program D. Environmental Monitoring Program
- 1. The: environmental moni toring' l. The radiological environmental program given inLTable 4.8-4'
~
~
conitoring samples .nall.be
'shall be-conducted except as . collected pursuant to Table.4.3-3 specified.below. from the locations specified in the 00CM, and shall ce analyzed
- 2. LWith,the radiologicalienviron- ; pursuant to the requirements of
. mental ~ monitoring program not. Table:4~.8-4; 4.8-5 and.4'.3-6; being conducted as spesified.in
-Table 4.8-4,. prepare.and submit 2. The results of analyses performed to-the Commission, in ene Annual on radiological environmental moni-Radiological Operating Report, toring samples shall be-summarized ~
- a' description of the reasons for in the Annual Radiological.
not. conducting the program as Environmental Operating Recort.
required and the plans for preventing a recurrence.
Deviations are permitted from sthe required sampling schedule.
If specimens are unobtainable due to hazardous conditions, seasonal unavailability,
. contractor omission which'is J- corrected _as soon as
. discovered, mal function of sampling equipment, or if a person who participates in the program goes ou. of business.
If. the equipment malfunctions, corrective actions shall'be-completed as soon as practical. If a person supplying samples goes out of "
business, a replacement will be found as soon as possible. All deviations from the sampling schedule shall be described in
-the annual report.
3 Wi th the- level of radioactivi ty 3 The land use census snall be in an environmental sampiing med- conducted at least once per
. ium at one or more of the locations twelve months between the dates speci fied in the 00CM exceeding of June I and October 1 by a the 1.imits of Table 4.3-5 door-to-door survey. aerial when averaged over any calendar survey, road survey, or by :en-
' quarter, prepare end submit to sulting local agriculture the Commission within 30 days authorities.
from the end of the af fected calendar quarter, a Special .
-Report which includes an evaluation of any release conditions, environmental 3.3/L.3-10
QUAD CITIES DPR-29
. factors or other aspects which caused the .' limits of Table 4.8-5 to be exceeded.
This report is not required if l- the measured' level of radioactivity was not the result of plant effluents; however, in such an event the-condition shall be reported and described in the Annual Radiological Environmental Operating Report.
- 4. With milk samples unavailable 4. The results of the land use census from one or more of the sample shall be included in the Annual locations required by Table 4.3-4, Radiological Environmental identify locations for obtaining Operating Report.
replacement samples and add them to.the radiological environmental monitoring program within 30 days.
The locations from which samples were unavailable may then be deleted from the monitoring program. In lieu of a Licensee Event Report, identify the cause of the unavailability of samples and identi fy the new location (s) for obtaining replacement samples in the Annual Radiological En-vironmental Operating recort and also include in the report a re-vised figure (s) and table for the 00CM reflecting the new location (s).
5 A census of nearest residences and 5 The results of the anal.yses cer-of animals producing milk for , formed as part of the recuired human consumption snail be con- crosscheck crogram sha I ce in-ducted annually (during the grazing cluded in the Annual Radiological season for animals) to determine Environmental Operating Recort.
their location and number with The analyses shall be done in respect to the site. The nearest accordance sich the 00CM.
residence in each of the 16 meteorological sectors shall also be determined within a distance of five miles. The census shall be conducted under the following conditions:
- a. Within a 2-mile radius f rom
. the plant site, enumeration of animals and nearest residences by a coor-to-coor or equivalent counting technique.
3.3/L 3-il
QUAD-CITIES OPR-29 b'..LWishin'a'Sami l e -rad i es ',
enumeration of animals by using: referenced information from county agricultural-
.' agents or other reliable sources, c
- 6. With a land use census identi-fying location (s) of animais t which yleid(s) an 00CM calculated.-
-dose or dose-commitment greater.
than the values currently being-calculated in Specification-4.8.A.3, the' new location (s) shall be added to the' radio-logical envi ronmental moni toring program with 30 days, if possible.
The sampling location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted f rom this monitoring program af ter .0ctober 31 of the year in which this land use census was conducted.
7 Radiological analyses shall be-performed on samples representative of those in Table 4.3-3, suoplied as a part of the inter-laboratory Comparison Program which has been approved by the NRC.
- 8. With analyses not being performed as recuired, report the corrective actions taken to prevent a-recurrence to the Commission in the Annual Radiological Environmental Operating Report.
l l
l l
t i
3.8/h.3-12
QUAD-CITIES DPR-29 v
E. ~ Solid Radioactive Waste E. Solid Radioactive Waste
- 1. The solid radwaste system.shall 1. The PCP.shall specify the method be used as applicable in accor - and frquency to verify solidifi-
-dance with the PCP to process cation of radioactive waste.
wet radioactive wastes to meet Actions to be taken if solidi fi-shipping and burial ground cation is not verified shall also requirements. be specified.in the PCP.
- 2. 'With the provisionsm of the Process' Control Program not-satisfied, suspend shipments of defectively processed or. defectively packaged. ,
- solid radioactive waste from the site.
C' t
3.8/h.3-13
~
QUAO-CITIES
-OPR-29 F. =Miscalienious'Redio3ctiva Materials ~ F. Miscell'aneous Radioactive _ Materials
-Sources:- Sources
' Source Leakage: Test Each sealed' source shall be tested for leakage and/or contamination by the
. Specification licensee'or by other persons speci-s_ fically authorized by the Commission or Each sealed source containing radio- an Agreement state. The test method active materiel in excess of 100 shall have a detection sensitivi ty of microcuries of 'beca m and/or gamma: emit ' at least 0.005.microcuries per test ting material or 5-microcuries of' alpha ' sample.
emitting material shall be free;of.
10.005 microcuries of -removable con- Each category of' sealed sources shall camination. be tested at the frequency described below:
Each sealed source with removable cont'mination a in excess of the above 1. Sources. in use (excluding startuo limit shall be immediately withdrawn previously subjected to core flux', -
from use-and either decontaminated At least once per 6 months for and repaired or disposed of in ac- all sealed sources containing radio-cordance with Commission Regulations, active' material:
A complete inventory of radioactive-
- a. With a hal f-li fe greater than 30 materials in the licensee's posses- days (excluding Hydrogen 3), and 4
sion shall be maintained current at 4 all times. b. In any form other than gas.
- 2. Stored sources not in use - Each sealed source snall :e tested orior to the use or trans fer to another licensee unless tested within the
. previous 6 contns. Seale: sources trans ferred wi thout a :ertificate indicating the last test date shall be tested prior to ceing placed into use.
A Soecial Reoort shall ce precared and submitted to the Commission cursuant to Speci fication 6.6.C.3 i f source leakage tests reveal thecresenceof10.005 microcuries of removaole :0ntamination.
G. In the event a limiting condition for operation and/or associated action requirements identi fied in sections 3.3.A. through 3.3.E., and k.3.A.
throu'gh 4.8.E. cannot be satisfied because of circumstances in excess of those addressed in the specifi-cations, no changes are recuired in the operational condition of the plant, and this does not prevent the clant from entry into an operational mode.
3.3/h.3-la
QUAD-CITIES OPR-29 ~
BASES.
3.8/4.8.A.1 GASEOUS EFFLUENTS i 00SE-This'specifi~ cation is provided to ensure"that the dose at'the. unrestricted area boundary from gaseous effluents from the units on the site will be within the annual dose : limits of 10. CFR Part 20 for unrestricted areas. The annual dose . limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table 11. -These' limits provide reasonable assurance that radioactive material. discharged in gaseous. effluents-will not resultain the-cxposure of.an individual in an unrestricted area to annual average.concentra-yttons exceeding the limits spec.ified in Appendix B, Table 11 of 10 .CFR Part 20; (10 CFR Part 20.106(b)). ,The .specified release rate limits restrict, at all _
times,~the corresponding. gamma and beta dose rates above background to an individual at or beyond the unrestricted area boundary to less .than or equal to
,500 mrem / year _to the total body or to not less than_ or equal to 3000 mrem / year to the skin. ~These release rate limits also restrict, at all times,-the cor-responding thyroid dose rate.above background to an - infant via the cow-milk-infant pathway to not less ' than or equal to 1500 mrem / year for the nearest cow to the plant. For purposes of calculating doses resulting from airborne releases the main chimney is considered to be an elevated release point, and the reactor vent stack is considered'to be a mixed mode release point.
3.8/4.8.A.2 00SE, NOBLE' GASES This specification is provided to imolement the requirements of Sections 11.3, Ill.A and IV.A of Appendix 1, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section 11.3 of Aopendix 1.
The statements provide the required coerating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix ! to assure that the releases of radioactive material in gaseous effluents ail! be kept "as low as-is reasonably achievable." The Surveillance Requirements imolement the requi remen ts- in Section ill . A of Apoendix l that conformance with the guides of- Appendix 1 is to be shown by calculational prodecures based on models and data such that the actual exposure of an individual through the acpropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the 00CM for calculating the doses due to the actual release rates of radiocctive noble gases in gaseous effleents will be consistent with he methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Ef fluents for the Purpose of Evaluating
-Compliance with 10 CFR.Part 50, Apoendix 1," Revision 1, October 1977 and Regula tory Guide 1.111, " Methods' for Est ima ting Atmoscheric Transcort and Dispersion of Gaseous Ef fluents in Routine Releases f rom Lignt-Water Cooled Reactors", Revision 1, July 1977 The 00CM equations provide for determining the air doses at the unrestricted boundary based upon the historical average atmosoheric conditions. NUREG-0133 provides methods for dose calculations consistent with Regu.latory Guides 1.109 and 1.111.
3.3/4.3-15
eva~nu
- - 3{8/h.8.A.3 00SE. RADICIOCINEs. EA310 ACTIVE i".ATERIAL in ? ARTICULATE =0RM R . .310.u.w,c.L i u :,5 u...-,nza ina.. t.m w L. :..--ura::
m..]
Thi s#speci fica: ion- i se provi ced' to"imalemen: :he' re:ui remen:s of Sections l ! .C,- ! ! ! . A and. I'/. A of Apoendix i , 10 CFR ?ar: 30.
The Limiting Ccndi:icas ..
rcr Operation are the gu.i des set tor:h .in sec::en-II.C or. mecend.ix i. ine s:a:emen:s provide :he repu.tred operating r.iax...l.:y i:i i and a: the same :.ime implement the gu.i des set torth in aec:icn 13/ . s. or mopendix ! o assure tha:
- he releases of radioac:!ve materials in gasecus affluents will be ke : "as icw as is reascnably achievaole." The 00CM calculattenal ~ me:heds s;ecified in
-:he surveillance requi remen:s .implemen ts :he requi remen ts in Sec:icn !!!.A of Appendix ! : hat conformance wi th :he guidas of A: endix 1 be snown,5y calcula-tional precedures based,cn'ecdels and data such :ha: the. cc:ual ex:csure .cf' an individual :hrough acpropriate pa:hways is unlikely to :e- subs:antially, uncer-estima:ed. The 00CM calcula:icnal me:heds ap: roved by MRC for calcula:ing :he doses due to :he ac:ual release rates of :ne subject ma:erials are re:uired to be consistent with One methodelegy provided in Regula:ory Guide 1.10c, "Calcula-ce :ne
- icn or Annual Coses :o. Man rrem Routine Releases of Reac:ce .r..luen:s r.
!", Revision 1, Pur:ose of Evalua:ing Comoliance with 10 CFR ?ar: 50, Ap encix Oc:Ober IC77 and Regulatory Guide 1.111, "Me:heds for Es timating A;mcsoneric .
.ranspor:
i and D.i spersion of Gaseous ,rricen:s .n i .3.ce:.ine Re.ieases 'r : 9. .rge:-
Va:er-Cooled Reac: Ors," Revision 1, July 1477 These ecua:Icns also ;revide for determining :he ac:ual doses based u:en :he hi s:orica l average a tmos:reric ine release ra:e spec...trica:icns ror raciciev,ines, radicac:.:ve c:nd.:.icns. i ma:erial in par:iculate form and radienuclides c:her than noble gases are dependent On :he exis ting radienuclice pathways to man , in :he unres:ric:ed .
i ne Oa thways w. .nica were examined .in :ne :evele: men: cr :nese s:ect.ica-area.
{.
x :icns were: 1) individual inhalatica of air:cene radienuclides, E) de:csi: ion . .
or re2wienuci s. .,es onto green lea ry vege atica wi:n su:secuen: cen s ume :icn :y man and 3) descs.i:t en on:0 grassy a reas wnere mi is. animais g.are at:n ::nsump;icn cr :.s.e mien by man.
3.a
- o.
w /n4. 0 . m, . ,4 unb Vw
,y$ ,.,,x.._.
3i I R a s. _o..e si .,I t r
The OPERASILITY of :he gasecus was:e :rea: men: which reduces amoun:s concentra:icas of radicac:ive ma:erials ensures :na: :he sys:em diil ':e available for use whenever gaseous ef fluen:s re;uire trea: men: prior :c elease :: :e l cnvironment. The r:cuiremen: :ha: :.' e a:Or:pri a te :ce:icns of :hi s sy s tem :e ccerable when 5:ecified provides reaschabie assurance tha: :He releases .. .. .
of
.is re25cnac,.y .
radicac:.i ve ma teri a ,i s i n ga secus a r-l uen:3 wi.1 ce <ect .,as icw as acnievaole". This s ecification im lemer.:s :re re:ui:0emen 3 Of 10 CFR ?ar:
10 CFR :ar: 30, and 30.36a, Generai Casign Criterion 60 cf A::encix A design objective Sec:!cn li .0 of A :endix I :: iG .R Par: 30.
3.3/4.3.A.3. EX?LO51VE OAS ltiXTURE This 5:ecif!::: ice is :rovided :: ens re :M:: :egas .. con:er: a:i rn O f :::en -
sys:am . 3 e . ::::
- ally ex:icsive gas mix:u.2s cen:sinec in :~e O r c:n formance wi:n :he requiremen:s of Oenerai Casi;n Cri:erien iC Of a::en:ix A
- 10 C.R : Par: 30.
- 3. . .2 To-f
. , _ . , . ~..
g
. Liquid EFFLUENTS-3.8/4.8.8.l.CONCENTRATl0N' This specification is provided to ensure that the concentration o'f radio-active materials , released in liquid waste ef fluents from the si te to unrestricted
. areas avill be,less ' than 'the concentration levels specified in 10 CFR Part 20,'
Appendix:B; Table ll, column 2. .The concentratior. limitifor noble gases, MPC in' air'(submersion), was converted:to an equivalent concentration in water Eusing;the. International Commission on Radiological Protection (ICRP). Publication 2.
3.8/4.8.B.2. DOSE This~ specification is provided to implement the , equirements of- Sections lI A, Ill.A and lV. A of Appendix 1,10 CFR Part 50. . The Limi ting .Condi tion for Operation implements the guides set forth in Section I I . A. of Appendix i . The statements provide the required. operating flexibility and at the same time implement the guides-set forth in Section IV.A of Appendix i to assure that the releases of radioactive material in liquid effluents will be kept "as low as Is. reasonably achievable". The dose calculacions in the ODCM implement the
~
requi rements: In Section lli .A of Appendix ! that conformance with the guides of Appendix -i tue shown by calculational procedures based on models and data such that' the actual exposure of 'an individual ' through appropriate pathways is
({}> unlikely to be substantially underestimated. The equations speci fied in the
- 00CM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents will be consis tent with the methodology provided in - Regulatory Guide 1.109, " Calculation of Annual Doses to Man # rom Routine Releases: of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.113,
" Estimating Aquatic Dispe'rsion of Effluents from Accidental and Routine Reactor Releases for the Purpose of implementing Appendix I", April 1977 NUREG-Oll3
-provides methods for dose calculations consistent with Reg Guide 1.109 and
.l.113
'3.3/4.8.3 3 Liquio WASTE TREATMENT The' operability of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents requi re treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be keot "as low as is reasonably achievable". This specification implements the requirements of
.10 CFR Part '50.36a, General Design Criterion 60 of Aopendix A to 10 CFR Part
~ $0 and . design objective Section 11.0 of Aopendix l to 10 CFR Part 30.
m 3.3/h.3-17
QUAD-C ITI ES OPR-29 t')
G 3.8/4.8.0.1 MONITORING PROGRAM The. radiological monitoring program required by'this specification provides measurements of radiation and of radioactive materials .in those ex-s posure pathways' and for those radionuclides, which lead to the: highest potenti -
= al radiation exposures of individuals resulting from the' station operation.
This monitoring program thereby supplements the radiological ef fluent monitoring program by verifying that the measureable concentrations of radioactive. materials
- and levels of radiation are not higher than= expected on-the basis of'the ef-fluent measurements and modeling of the environmental exposure pathways. Pro-gram changes may.be initiated based on operational experience.
The detection capabilities required by Table 4.8-6 are state-of-the-art for routine environmental measurements in industrial l abora to ries . The specified lower limits of detection for 1-131 in water, milk and other food products corre-spond to approximately one quarter of the Appendix ! to 10 CFR Part 50 design objective dose-equivalent of 15 mrem / year for atmospheric releases and 10 mrem /
year for . liquid releases to the most sensitive organ and individual. They are based on the assumptions given in Regulatory Guide 1.109, " Calculation of Annual-Doses to Man f rom Routine Releases of Reactor Effluents for the Purcose of Evalu-
.ating-Compliance with 10 CFR Part 30, Appendix 1", October 1977, except the change for an infant consuming 330 liter / year of drinking water instead of 510' liters /
year.
(f 3.8/4.3.0.6 LAND USE CENSUS This specification is orovided to ensure that changes in the use of un-restricted areas are identified and that modifications to the monitoring program are made if required by the resul ts of this census. This census satisfies the requirements cf Section IV.S.3 of Aopendix l to 10 CFR ? art 30.
3.8/4.8.0.7 CROSSCHECK ?ROGRAM The requirement for participation in the interlaboratory comparison crosscheck program is provided to ensure that independent checks on the precision and accuracy of the. measurements of radioactive material in envi ronmental sample matrices are performed as part of the quality assurance program for envi-onmental monitoring in order to demonstrate that the'results are reasonably valid.
3.8/4.3-13
- - _ - . - - . . . . ~ . - - . . .
. QUAD-CITIES DPR 'N .
_.. '3'.8/4.8.C' MECHANICAL VACUUM PUMP'
-The purpose of_isolatingLehe mechanical' vacuum line is.co limit. release of activity from the' main condenser. During'an' accident, fission products would.be' transported from-the. reactor through the mainLsteamline to.the main condenser.
The~ fission product radioactivity would be sensed.by the main 'steamline radio-activity _ monitors which initiate isolation.
- 3. 8 / 4. S ', F . MISCELLANEOUS 'MDI0 ACTIVE MATERIJLS SOURCES The' objective o'f this ; specification,is to assure that . leakage f rom byproduct,
. source-and special' nuclear material sources does'not exceed allowable limits.'
The limitations onLremovable contamination for sources requiring leak-testing, including alpha emitters, is based :on- 10 CFR 70.39(c) limits for plutonium.
3.8/4.8.E. SOLID RADIOACTIVE tJASTE The. operability of the-solid radioactive waste system ensures that the system-will be available'for use whenever solid tadwastes require processing'and packaging
' prior to-being shipped off-site. 'This specification Onplements the requirements of 10 CFR 50.36a. and -General Design Criteria 60 of Appendix A -to 10 CFR Part - 50.
~i 3.8/4.3-19
QUAD-CITIES OPR-29 TABLE ~4.S-1 RADICACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM MINIMUM- LOWER LIMIT OF GASEOUS' . SAMPLING, ANALYS;S TYPE OF' OETECTION (LLD)
RELEASE ~ TYP' 'R EO.UENCY: FREQUENCY ACTIVITY ANALYSIS (uci/ml)
Principal A. Main Chimney M M D
Gamma' Emitters
- 1 x -10'0 Reactor Bldg. Grab. Sample- Tritium I i x lo o
. Vent Stack M i
-3. All Release Con tinuousd we i.131 !x 10-12 Types as Charcoal' Listed in A Samole l-133 1 10-iG Above Princioal l Continuousd WC Gamma 'Emi::ers* I Ix 10'II Particulate (1-131, others) :
Samole !
~}
Con tinuous d Q SR-39 l1 x 10'II
. Comoosite SR-90 ;
Particulate I
Samole i x 10'Ii I:
Continuousd q, 33 z; n, l ; ,).1-Comoosite i Darticulate l
Samole :
C. Main Chimney Con t in uousd yea;, ,x 73-6 Gas Monitor .toole Gases D. Reactor Bldg Continuous d Noble Noble Gases 1 x 10-N ll Vent Stack Gas Monitor i
i 3.3/4.3-20
QUAD-CITIES
.ry
- W DPR-29 TABLE 4.8-1 (Continued)=
TABLE NOTATION.
a.' The lower limi t of detection (LLD) is defined in table notation a.
of Table 4.8-6.
,b. Sampling and analyses shall also befperformed following shutdown,-
startup, or a thermal power change exceeding 20 percent'of rated thermal power-in I hour i f'an abnormal change in radionuclide mix-ture or concentration. Is anticipated.
c .~ Samples ~~shall be changed at least once per-7 days and the analyses completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after removal from the sampler. Sampling shall 'also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following each shutdown, startup, or thermal power level change exceeding 20% of rated thermal power cin one hour i f. an abnormal change ~ in radionuclide mixture or
- concentration is anticipated. When samples collected for' 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> -
are' analyzed, the corresponding LLD's may be increased by a factor o f 10.
- d. ; The ' ratio of sample _ flow rate to the sampled stream flow rate shall be known.
' 3h e. . The principal gamma emi tters for which the -LLO speci fication applies
" . exclusively are the following radionuclides: :Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and .Xe-138 for gaseous emissions , and Mn-54, Fe-59.
Co-60, 2n-65, co-58, Mo-99, cs-135, cs-137, ce-141, and ce-lha for particulate emissions. Other peaks which are measurable and identi-fiable by gamma ray spectrometry, toge ther wi th the above nuclides ,
shall be also identified and reported when an actual analysis is per-formed on a _ sample. Nuclides whien are below the LLD for the analyses shall not be reported as being present at the LLO level for that nuclide.
J 3.8/4.8-21
- QUAD-CITIES
'OPR-29 TABLE-4.8-2 MAXIMUM PERMISSIBLE CONCENTRATION OF OISSOLVED OR ENTRAINED NOBLE' GASES RELEASED FROM THE SITE TO UNRESTRICTED AREAS IN LIQUID WASTE NUCLIDE . MPC(uC1/ml)*-
Kr-58m- 2x10~4
-Kr-85 5x10-4 Kr-87 4x10-5 Kr-88 9x10-5 Ar-41 7x10~3 Xe-131m 7x10-4 Xe-I33m- 15x10~h Xe-133 : 6x10-Xe-135m 2x10-4 Xe-135 2x10~4
- Computed f rom Equation 20 of ICRP Publication 2 (1959), adjusted for infinite cloud submersion in water, and R = 0.01 rem / week, density = 1.0 g/cc abd Pw/Pt = 1.0. .
J
,f 3.3/4.3-21a
,, -- - -- g +-a , - - a-
QUAO-CITIES DPR-29.-
(}
. . TABLE 4.8-3 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of +
' Liqu i d .. Release. Sampling Analysis. Type of Detection -(LLD):
Type Frequency. Frequency; Act iv-ity Analys is (uci/ml)
A _Satch Waste Re- Prl' or to Prior to- Principal lease Tanks Each Batch Each Batch Gamma Emitterse 5x10 ~7 i 1-131 lx10-6 P rior - to ~ M Gross Al'pha lx10 -7 Each Satch Composite b H-3 lx10 ~3 Prior to Q Fe-55 lxio-6
'h
. Each Batch Composite b
_ ,g i
Prior to M Dissolved 5 ~
One Satch/.* Entrained Gases'. lx10 '_
(Gamma Emitters)
- 3. Plant Contin- I-13I lx10-6 C Principal uous Releasesd 3c(Grab M
-7 Sample) Gamma Emitters
- ~3 l' Q Sr-39, Sr-90 5x10 (Grab ~,
Sample) Fe-55 lx10 '
.)
3.3/4.3-22
QUAD-CITIES DPR-29 3
b/ TABLE 4.8-3 (Continued)
TABLE NOTATION
- a. The LLD is defined in Notation a. of Table 4.8-6.
- 6. A composite sample is one in which the quantity of licuid samoles is proportional to the quanti ty of liquid was te discharged and in which the method of sampling employed results in a specimen which is repre-sentative.of the liquids released.
- c. I f the alarm setpoint of the service water ef fluent monitor as deter-mined in the ODCM is exceeded, the frequency of analysis shall be in-creased to daily until the condition no longer exists.
- d. A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed to assure representative sampling. A continuous re-lease is the discharge of liquid wastes of a nondiscrete volume; e.g. ,
from a volume or system that has an input flow during the release.
- e. The principal gamma emitters for which the LLD speci fication applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-60, Zn-65, Co-58, Mo-99, Cs-134, Cs-137, Ce-lkl, and Ce-144. Otner peaks which are measurable and identifiable by gamma ray spectrometry to-() gether with the above nuclides, shall be when the actual analysis is performed on also identified and recorted a samole. Nuclides which are below the LLD for the analyses shall not be recorted as being cresent at the LLD level for that nucli de.
- f. The dissolved and entrained gases (gamma emitters) for ohicn tne LLD specification applies exclusively are the following racionuclides.
Kr-87, Kr-38, xe-133, xe-133m, xa-135, anc xe-138. Other dissolved and entrained gases (gamma emitters) which are measuracie and identi-fiable by gamma-ray s oec t rome t ry , together wi th the above nuclides ,
shall also be identi fied and reported wnen an actual analysis is cerformed on a samole. Nuclides which are below the LLD for the analyses shall not te recorted as being cresent at the LLD level fo r that nuclide.
l 3.3/L.3-23
.__m.-. _ . =. __m. . . m
(.- - t .. )
QUAD-C ITIES DPR-29 TABLE 4.8-4 RADIOLOGICAL. ENVIRONMENTAL HONITORING PROGRAM Exposure Patliway Minimum Numlier of Samples . Sampling and Type and frequency-and/or Sample and Sample Locations
- Collection Frequency of Analysis -
i 1. AIRDORNE
- a. Particulates 16 locations Continuous operation of- Gross beta and ganuna sampler for a week isotopic as specified ~
in ODCH.
I P b. Radiolodine 16 locations Continuous operation of. 1-131 as specified in I
sampler for two weeks- ODCH.
7 2. OlRECT RADIATION Forty Locations Quarterly (Minimum of two ILDs 4
per packet)
- l' Sample locations are given on the figure and table..in the ODCH.
I I
1 e
v ~ .
I j-QtlAD-CITIES OPR-29 B BLE 4.8-4 (Continued) g RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM Exposure Pathway Minimum Number of Samples Sampling and Type and Frequency and/or Sample and Sample Locations
- Collection Frequency of Analysis 5 4
3 WATERBORNE
- a. Public Water 2 Localions Monthly composite of Ganma . isotopic analysis weekly collected samples o f. each compos i t e samp le
- y b. Sediment I downstream location in Annually Canma isotopic analysis of each sample
~
w receiving body of water v-w c. Plant Cooling intake, Discharge Veekly composite Gross Beta analysis l
g Water of each sample ISample locations are shown on the figure in the ODCM i
i !
i 6
I l e t
I I !
I
s i
s y
l y as c nn.
O q n
e s u i s
se i l ao ct i r i
e y sp po r l lyma op F a t d A n as n
oe.
sl n ah i b a f c i o 1 a ad e 3 e n u e p 1 n
y f an T 1 o G o s
e em ci y y nt c
n l
no r
. k o e e t e u eesh q wrat e aeo M r e l y A d F cs t l R n nl t a l G a n oaa a u
O o m h R g i t i ;t n P n t s nen n aar a uno i c G l p
e e -
N l l nt i I
m l esr m
) R a o t h ae e d O S C A wpp S e T u I n N i O S t H E n o L OT I
C A I ( T s C9 N e g
- 2 4 E l D - - M p n AR 8 N m
- i UP O a s v QD f i
R S n i H
I o e C E V f i c D L N o r e 0 B E a r -
A r c e T L e o nr h A b L i e t C m s t I u e n na n G N l p
o iow i O i L m m t t f d O u a a c
ao c b e
I m S oy D i o i r
A n d L l d
_ R i n o c
_ M a 2 I b s e
_ d e
r a
s
. y n a o w e i
_ h l N t t p 0 k h a
_ a m 1 l s c P a T i i o
S S H F l e E
. r r J e u o N . . l a p s /
o d p n I b m
a x a . S E 4 "-
_. (
~ F R.T, ilj . : l t .
i i!
QUAD-CITIES
, , . OPR-29
~
' TABLE 4.8-5 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS-IN ENVIRONMENTAL SAMPLES 1
Reporting Levels Analysis Water, Airborne Particulate. Fish Milk Food Products or- Gases (pCi/m3) (pCi/Kg, wet) . (pCI/1) (pC i /Kg ,we t) 4 H-3 2 x 10 (a) _
Mn-54 .1 x 103 3 x 10 4 Fe-59 4 x 102 i x ig 4 Co-53 1 x 103 3 x 10 4 f
Co-60 3 x-102 1 x lo" Zn-65 3 x 102 2 x 10 4 g-Nb-95 4 x 102 1-131 2 09 3 1x 10 2 Cs-134 30 10 Ix 10 3 60 I x 10 3 Cs-137- 50 20 1 x 10 3 70 2 x 103 3a-La-140 2 x 10 2 3 x 10 2 (a) for drinking water samples. This is 40 CFR Part 141 value.
i
./
1 t 3.3/4.3-27
-)
QUAD-CITIES OPR-29'
_c hy '
-- TABLE 4.8-6 PRACTICAL LOWER LIMITS OF DETECTION (LLD)
FOR STANDARD ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM Sample Media-- Analysis LLDa ,b Units (4.66r)
Ai rborn ed'Pa r t i cu l a t e"- Gross Beta + 0.01 pCi/m 3 Gammailsotopic 0.01 pCi/m3 Ai rborne 1-131 lodine-131 0.10 pCi/m3 Milk / Water 1-131' 5 pCi/l Cs-134 10 pCi/l Cs-137 10 o' -pci/l Tritium 200 pCi/l Gross Beta'+ 5 pCi/l Gamma Isotopic 20 pCi/1/nuclide Sediment Gross Seta + 2 pCi/g dry Gamma I so top ic- 0.2 pCi/g dry Fish Tissue -I-131 - Thyroid 0.1 pCi/g wet
() Cs-134, 137 Gross Beta +
0.1 1.0 oCi/g wet pCi/g wet Y lsotopic 0.2 pCi/g wet 0 0.5 pCi/l on milk camples collected during the pasture season.
+ Referenced to Cs-137 6 5.0 pCi/l on milk camples l
,i i
l l
i -
.)
i l 3.8/4.3-23 i
\
) _ - - _ - _ ._
~ , . . . . - - .
. QUAD-CITIES- ,.
1 /[
i If TABLE 4.8-6 (Continued)
TABLE NOTATION '
A. The LLD is the smallest concentration of radioactive material z in a sample that will be detected with 95 percent probabili ty with only 5% probabili ty of false-ly concluding that a blank observation represen ts a "real" .sequal .
For a particular measurement _ system (which may-include radiochemical separation) 4.55.5b-LLD = ---------------------------------------'
A- E V.* 2.22 Y
- exp ,(-Aat)
- t
- Where:
s l LLD ls the "a priori" lower limit of detection for a blank sample or background 1
- analysis as defined above (as pCi per uni t mass or volume).
sb is the square root of the background countf or of a blank -samole' count; is toe. estimated standard error of a background count or a blank sample count as appropriate (in units of. counts) .
E is .the counting efficiency (as counts per disintegration).
A ~ is the numoer of gamma-rays emi tted per disintegration for gamma-ray radio-() nuclide analysi s- (A - 1.0 for gross alona and tri tium measuremen ts).
V is the sample size (in uni ts of mass or volume).
2.22 is the numcer of disintegrations per minute :er oicocurie.
Y is the fractional radio-chemical yield when applicable (otherwise Y = 1.0).
A is the radioactive decay ccnstant for the particular radionuclide (in uni ts of
- reciprocal minutes) .
! at is the elapsed time between the midpoint of sample collecticn and the s tart time of counting. (a t = 0.0 for environmental samples and for gross al:ha meas u reme n ts ) .
1 .
- is the duration of the count (in units of minutes),
i l The value of "Sb" used in the calculation of the LLD for a detection sys tem snail be based on an actual ooserved cackground count or a blank samole count (as acoro-priate) rather than on an unveri fied theoretically predicted value. Typical values of "E", "V", "Y", "t", and "at" shall be used in the calculation.
l
- For gamma-ray radionuclide. analyses the background coun ts are determined f rom tne total counts in tne channels wnich are within plus or minus one FWHM (Full Widen i at Hal f Maximum) of the gamma-ray phctopeak energy normally used for the cuanti-tative analysis for that radionuclide. Typical values of tne PJHM shall :e used l (f t .
in :ne calculation.
l I
i 3.3/L.3-29
, . , - - - - -..~. -
e.--------vm--em -
p y e=r er e- -- --
QUAD-CITIE5' OPR- 29
^)
./ TABLE 4.8-6 (Continued)
TABLE NOTATION The.LLO for all measurements is defined as an "A priori" (before the fact) limit representing the capabili ty -of a measurement system and not as .an "a posteriori"
- (af ter the- fact) limi t for a particular sample measurement.
8 .- Other radionuclides which are.measureable and identifiable by gamma-ray spectro-netry, together with the nuclides indicated in Table'4.8-6, shall also.be identi-fled :and ' reported when an actual analysis is. performed on a sample. Nuclides.which are below .the LLD for the analyses shall not be reported as being present at the LLD level for. tha t nuclide.
C. LLD for drinking water.
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[
l QUAD-CITIES OP - 29 1 50.59 to varify thse such occians did not cons:1:uce en untsviewed snf scy quescioni Woposed4 changes to the Quality Assurance Progra:stdescripcion
saall he reviewed and approved by che ! tanager of Quality Assurance.
- 2) Proposed :hanges to procedures, equipment or systems which involve an anraviewed safety question as defined in 10 CTR 50.59.
- 3) Proposed tesc.s or experiments which involve an unreviewed safety quescion as defined in 10 C7R 50.59.
- 4) P= sposed changes in Technical Specification NRC cperacing licenses. *
- 3) 3oncomplisace with NitC requirements, or of internal procedures. or in-scruccions-hmoing nuclear safety significance.
- 6) ffanificane operacing- abnormalities or deviacions f rom nor=41 at.' erf ect ed performance-of planc equipment.that aff ect nuclear safe:y as ref erred :o it hr the Cnsice Review and Investigative Fune:1on.
- 7) Emportable occurrences requiring 24-hour nacifi acion to the NRC.
- 8) AII. recogni. zed indications of an unancicioated deficiency in some aspect of daelan or ope.r,cion of saf ety-related structures, systems or ce=ponents.
- 9) Review and report findings and reco==endacicos regarding all :hanges to t.he Generac.ing Stations bergency Plan prior :s implementation of suen c! range.
- 10) Iaview and rmoore findings and reco==endations regarding all teess ref erred by the Ter % L 1 Scaf f Suoervisor. Scacion Superin:ecdent. Civision "Tice-Presidenc - Nuclear Sca:icos and Manager of Quali:y Assurance.
. h. lam r function .
The Lidtr Tunc: ton shall be the resoonsibility of the anager of Oualtcy Assur-ance 1:zdependeac of :he Production Depar:=ent. Such res;casibility is tele gaced to cha Director of Quality Assurance for Operating and to :he scalf Assistant O to che Manager of Quality Assurance for saucen nce quality assurance activities.
F.icher snail appeuve the audic agenda and :hecklists. :he findings and :he
. report of each amiic. Audies shall be performed in accordance with the 00moany Quali.ry Assuranca Program and Procedures. Audits snail be serfor:ec to assure th.at safety-relar.ad functicas are covered wi:hin a period of 2 years or less a.s designa:ed below.
- 1) Audit of :he conformance of f acili:7 overacion :o orovisions :oscained wi:h-in the Teewical Specifications and appli:able Li:ense coniiciens a: leas:
cuca per year.
- 2) Audic of the adherence to procedures. : raising and qualif t:ation of :he stacion scaff ac least once per year.
~
- 3) Audic of the results of actions tasen :o correct deficiencies occurring in faci.licy equipment. struc:ures, syste=s or sethods of ooeratica tha affect nuclear safety ac least once per 6 months.
- 4) Audic of -ha performance of ac:ivities ' required by the Quality Assurance Frogra:s :o meet :he Cri:eria of Appendix "T' 10 07R 30.
F 5) Audic of ::ne Tacility hersency Plan and i=ptementing procedures.
- 6) Audit of the Facility $ecurt:y Plan and implementing procedures.
, 7) Audic onsite and offsite reviews.
- 3) Audir :he racility Tire Proce::icn Progras and t=pLesenting procedares ac
< imast once pe.r 24 months.
I
- 9) The radiological' enviren= ental :onitoring program and :ce results :certof ,
ae lease onca per l*ll. :onths. }
=
.- l
- 10) The CCCM ami i=plemencing procedures ac ;eas: :nce per I'. :enens.
.j
- 11) The FC? and ispie=enting procedures f or solidtit:a: ton af radioac:'ve was:e ?
i
~
ac Lease once per 2,4 sencas.
l
. 6.1 -2.
,-n.+
7 ,9 7 - ,.- ,,_.m n wy,. --,m .-- n-g- ----v --, -- w-- , - - - - ,
" - - - - - h
. _. us . .
QUAO-CITIES OPR-23
- 9e SwcIrvisor of : e GFfsite Review and favesticat'<e -;nction; 39d (6) suomit to :9e
.]ffsite Review and investigative ranc:fon #oe concu- e-ce i n a
- elv an,e . - ose te~s descrices in 5;ecification 5.3.G.'.a amicn ave :ee9 accrose: by :ae :asite Review an:
Investigative ranc icn.
The responsibilities of t9e Personnei :er<:rmi,; :-is #2,c:!:n are stated below:
' l I) Review of (l) crocedures re:vire: ov S oe:i ' < :s t i on s.1 and ;"aages : ereto an: 0:
any other 3rocosed scocecur es or c9arges inereto as deter *iaec by :9e 81act Su:er-intence"t to affe : 9uclear safety.
f
- 2) Review of all Dro00 sed tests 39d ex:eri*erts :'at af#e : uc! ear sa#e: a
) 3) 2evica of all procoses cranges to : e Te:-nical 5ce ifica: ions.
- 4) Review o' all procosed c9anges or "oci#itations to ia": s< stems or e:ai: e,t 09at e
t' , affect suclear safets.
C i
! ; 5) investi;ation of all noncomoliance oita NRC ecuire+ emes ame snait : e:are an: 'er-rs I ard a resort covering evaluatian anc recorren:ati:ns :: reve-: e:u- e9ce *: :'e t Division vice Presicen:-suelear stat: ors an :: :se su:eresor :/ : e :*f si te Reviea k and investigat;se funct i on.
I 5) Review of #aci'itt Ocerations :o Oe:e:: :oteat al safe: 9 ara as.
- n =er<or a,
- e 3< see:isi e,iews a-e eves:i; :io s are e:or s : e e:- as e:.es:e:
- v :e sucer<isor of :-e Of fsi:e 8ev;e- an: : ves:,;a: <e 2.ne: :r j.
- 3) 2eview :f *e 3tation 3e:ari:v ?'ar ac: saa ' ..:-:- e::-cea:e: : a-:es : :e 1 aieision /;:e ? esi:er:-N.e ear sta: On s .
1 .
i n eview o<:e -rer;e,c, s'a, ine s: t-:e ' cie e-ti ; :r::e: res a : sr s:- !
, "e:JFFeSce: "4"gea : **. )Itisi:^ ie 3*es;* eat *Sv !e ' 5; * *s. I
[. . '
l0) Review of "e:ortaale Oc:a'erces ar: 40 cos ta*e" t: :" eve': 'e .r~e*te-aevie- - a,v .,oia re: 2n. .:. , ease e ar :ac: .e a:e- r :: :. e-o : i. l
. i io t i,cimai-; : :re: ara: :n 3,: < -4 o ; o f - :or:> ::,e-: : eva..a: :- e:: , :a-
- ens aac *is:Osi !i0n :* ;"e : r"t:0 ee e ! ' *: 3"ee.* "e '*e":e ~ ^e '~s O' e6
, ice > esi:e-:-,ucies- ::acers. a,: : -e s.,v ice :- -e s. e =e>e. a: ,
I . e.-
n.
i ':: 2evie, :' : a es :: :e 2': a-: :: ". " , a: a.:- : a :es : e a: .4 5 e 'ea: e' l isste*s. l 5
( :. ut-ce::v e s:ao :, I.:e- : -: a: a.
] re
,e re:,,,:ai s:a ~ su:e ,;s:r : s es:cesir e ::
e,:ae=,s ;, a ee~ .er :, ar a eas .
es . .e,: ,at , a= ..a-a
) i 039c-ol :aases :/ :laat ai-teaar e. ::e'a :n . a: 3: '- s:'a* se :r ;e:.*es *e's: :
f l o
- a:i l i t'/ : e"att:ns a* 5 ai' aee :"e 3 ~5:.es: I'e 10 ~ " :elii' j e,s r. :: : : ,,ce a :- -r es. re;a:stices 3 n :*o'
- :c ,* ce: es e- - -i :: - : , .: : :-
l !s 9ecessa"v **e 3:st :n 5.:e"'7:e" e** i'ai' *:i . 5.O* *e:0-*e* a ' :'s ' 5e'," 3 Sourse of a:! :n :*at 5 0*e :ase*va; ve ** * *; S a #e ::e *a 2" - :e :-
- l socn Oisagree eati s'a- :e e: r e: '*~e: ' : :e ' 5 :' e es :e*~* 'i' '
5:ation, and !"e 32:e eisar of :*e ##s+:. 2 < E*- 'cei! ist e *.~I "
e 1
g 5 Ne:Qr:s i
i 1 o te:or:s. evie s. ; veso ;a t-:ns, i,e e:: re-ca: :n i sa- :. ::: ,-:e ,-- :: es i 03 *e 3ivisi:9 / ice 3'esite" -1wceear 5:a; :ns. *e .:e* 3r ~~e . # 5 ** 2./ * ,
, 3r: ' ves: :a: <e :. ::i:n. :e s:ae:n :..:4- :t :e : 3-: : e a-a;e- 3; :.o -
, i :ss.Pa9ce.
e 00 es al' "e :P.s na: : : ..' e"! a ;
- s'a' 09 se:* :o e 9 e >:a:
. 3*:cS .'es
=ri*:e" a /e " :2..'es 5'ai: 'e .~e a'e a' n ' "e : "
C': . . *, *-
- i",>s si:e 2evig, g 3 ~a*.es: ; ! ' te .*: :".
- ese '*;;!..'!s *I -- Ce 'e * >*
.cnte?' a* : "e:'C: :" s. -ils
- t0 . e5t*ia: * . f a
~
- ' e' **
f
- i-so- :e > es ae-e ..: ea- ,: .4 3. e .:e- , - -- 9 . > . :o e. j a-: 1 ves: ;a: .e .: :~.
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- bi I $
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- P.1- 29 6.2 ?!. ANT QPERA!!NG ?ICCIDL*RES A. ' 3scatled v:1::an procedures, including 'sp?li:able the:koff '.is:s :overing i: ass liscad' be;:w snall ' e prepared, appr ved. and 20hered ::s:
1, Sar:a1 scartup, opera: sn. and snu:fo.m -af :ne : nae :r. and sener syscams and components . involving r.uclear saf e:y. of :he f acili:7
- 2. Refueltag operacions.
3 Actions :o be :aken :o :cere:: specift: and f:reseen pecencia; ra; fun::i:ss of systems or :o conents. inclu ing ressonses :s alar:s. susse::e: ;;1:arr syst a: Leaks, and acnor:41 4a0:171:7 :nanges. .
~;. E ergency conditi:ns involving po:encia; se actual ~;elease of :acica::ivi:y -
"0enerecing Sca:1:n+Esergency > ?lan" anc <stacion e:ergency ans a:nor:a; p roc edur es .
3 instrumencacion operacion whi:n :ould have an effec: :n :ne safe:y :f.:na facility.
- 6. ? eventive anc :orrective taintenance sperac ns vai:n :au.: nave an ef f e::-
on :ne safe:y sf :he fac111:7
- 7. Surveillance and :escing require =en:s.
- 3. Tescs and.experi en:s.
- 1. ? cesare :s ensure saf e snuc::wn of :ne olan:
- 10. Scacion Securt:y ?lan anc L::La:en*a::sn stoce: es.
.l.
. Tire ?:stetti:n ? 3gra: impia:ents:::n.
- 12. ODCM i olecenta:1:n.
- 13. ?C? 2:ola:en:2: ten.
4
- 3. Radiati:n :ent sl Orocecures snail :e ta;::sinec. a:e avai;as;e .:2 a;*. 4:2:1:n
- ersonnel, an
- a:Merec :: The 2 : ecures anal; sn v :ernisa t:;e : :11:::?. ex-
- osure and snail
- e ::nststen: st:n :ne to:2 :1:ents sf . 77. '*. *h ;4 ::-
actan sto:es: ton ze:gra: snail se ::;an::ac :2 ceet :ne re;u :enen:s :f .* :71 .:
- . ;. ? scaceres is: 1:e:s ;:en:if.e: '.n ines;ft:1::: i.*-4 an: anr : anges ::
suen :recedu;es saal; :e reviewe: an: aoorsve :y :ne ::ers: n; *ng;nter an:
- ne Tecnnical Staf f laservisor in :ne areas 2f :cers:::n :: f.e. tan:.;ng.
anc sy Main:enance Asse. iuot. an: Te:nni:41 3:tif Su:erv:s:: . : e areas af plant estatenance an: ;;an: ;ns:ec:;:n. 77::scures !s: .:ess .:en:;f.e:.
.a 3;ecificatton 3.2.3 ace saw ::anges :s su:n ;;:: :ures sna.. :e rev;a e:
anc aspro /es my :ne Te:nn :a13:af f 3.nervisor an: :na ?.acia:::n :he ::a; iuservtsor. A: '. e a s: :ne ;ers:n a;7::vtag ea:n :! :ne asove ::::stures saal; noid a talia senior :: era:st* s 11:ense. n a:: :::n. :nese ::::e:.res anc thanges there:3. Sus: ave au:nori:2:1:n ny ne i:a::: lu: erin en:en:
before 'seing L: ole:en:ec.
- 1. Work and instru::1:n : vee 3 : esures ent:n troia*ent a :::ve: ta' a: er.an: e or ocift:sti:n 2:2cecures sna;; se a;;;:vec an: au::::::ac :v :ne M2;n-
- enance Afst. Ig3C. .nere One # itten autner'.tv *as :een ::: vita: :v :ne ita:::n 54serta:endent. .e " Main:enance/M::ift: :::n ? ::e:.res* .:t;i:e:
fJ: safe:v Telated Vo 4 saal; se so a:3 Ove: niv *f . ::::s?ures referente:
.a :he ' Main:enancerMocift:1:1on ? :ce:ure' ave :een a:se:ve: as resuire:
sy 6.2.A. ? Scesures vnt:n : ioc fai; .t:nin :ne ret.;;ene*: :! 4. . .s 2 6.2.3. =ay se appenves sy :ne Oe: art en: fea:s.
- 3. Te:corary :hanges :o r:ce:utes 5.2.A. and 4.2.3. :ove av :e a:e :::v;:e::
- . -The inten
- af :ne se:gine; :::teure is 20: a;:ere:.
- 2. The ::ange is apersvec :y :vo se:ers :f :ne sian: 1:a g ene .: s:aif. 1:
- east 2ne of vno: -:':s
. . a lent:: ?.ea:::: : ers::: s "..;ense :- :na .nt:
affe::ed,
- t. . The ::ange is :::. en:e:. :evieve: *
- ne :?s t:e Re n av 1. : -ves: ;;:.
. e T'.n ::1:n an a:*:Ove: * : 7. e 3:a:13n 2.:e:1*. e**e9 . t .". ;n .. ;avs !f isoienen:2:::n.
I. 3:1;*s
. f **e t e gen *v :::stures *es* 1b ec ;* I: e a t f i
- a * *. * * . ! .a. - 1*1;. :e
- en:2::e: La a::or:ance . :n :ne :3I? :!anua L
- 3. *.
QUAD-CITIES .
i
' DPR- 29 ~
- 2. - Aitabulationashall be submitted on 'an annual basis of the' nu=ber off sta: ion, utility, and other personnel (including contrac: ors) receiving exposures 4 greater'than 100 =res/yr and : heir associated man.rea exposure according'to work-and: job function ~(Note: this tabulation supplements the' requirements of LSection 20.107 of 10~CFR 20),-e.g., reactor opera: ions and surve11'ance, inservice inspection, routine =aintenance, special-maintenance (describa -
mair.tenance),. waste. processing,.and refueling. The d,ose assign =en:s :o various. duty functions =ay be estimates based on'pockat; dosimeter, ILD, or
. film 3 badger:easurements. Small exposures totaling _lessichan 20" of :he
, individual total dose need not be accounted for. In the. aggregate, at leas:-
SO%tof the total whole body dose received from. external sources shall be assigned to specific. major work functions.
- 3. Monthly 0perating Report Routine reports of. operating s:stistics and shutdown experience shall be submi::ed on a sonthly basis :o he Director, Of fice o f Manage =en: nforma-tion and ? ogram Control, U.S. Nuclear Regulatory C ==1ssi:n, Tashing::n, DC .20535, vi:h a copy to the appropriate Regional Of fice, :o arrive no '.a:er -
chan :he '5th of each son:h following :he 221endar conth c:vered by :he repor:. -In addi: ion, any :hanges :o he ODCM shall be subsi::ed wi:h':he.
Monthly opera:Ing Repor: vi:hin 90~ days of.:he effec:ive date ;f :he :hange.
t
~
A report of major change :o che radioac:ive was:e :rea::en: sys: ens shall t .be submit:ed with :he Monthly. 0perating Repor: for. :he period in which :he ,
evaluation was reviewed and accepted by the Onsi:e review fun : ion. *f such change is.re-evalua:ed and no: ins ailed, no:ifi:a:i:n f :an:alla:1:n ,
of the change should be provided :o :he NRC.
- 3. Reportable Occurrences Repor:able occurrences, including correc:ive actions and measures :: preven:
recurrence, shall be repor:ed :o :he NRC. :n general, :he i por:ance of an occurrence with respect to safe:y significance determines :he inmediacy of re-
~
porting required. n some :ases, however, che significance of an even: nay no: be obvious a: :he :ine of i:s occurrence. In such :ases, :he NRC shal' he
- informed promptly of an increased significance in :he licensee *s assessmen: Of
- the. event. In adci
- ion, supplemental repor:s may be required :o fully describe i final resolution of :he occurrence. In :ase of :orrec:ed or supp'.ecen:al repor:3, 1-a licensee event repor: shall be :c=ple:ed and reference shall be made : :he
- original repor
- da:e.
t
- 1. Prompt Notification wi:n Wri::en Followup The :ypes of even:s lis:ed below shall be repor:ed as expedi:i:usly as i possible, but wi:hin 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> by :elephone and confirmac by :alegra:' . na .1- .
l gram, or facst=ile ::ansmission :o :he direc::: Of :he appr:pria:e regi:na.
office or his designa:e no la:er :han :he firs: working day foll: wing :ha even., . s. 3 ...a...,.... .....:,..., vu,e . ,. c. o c .4-.....
u...,. . .
. a . < . .,,.<a. . . - .....
, .a...,.
, . ,3, o ..a6
.. .. g a .i .t .4 .. e .n.a t a . , as a ~4..4 .. ur.. a .- - o. '
. . .2 . .= d. -.y"
. . .3 .= ' '
... a. a. . s >. =. *"2...
f
, repor: form. *n icrnation provided :n the li:ensee even: repor: f:r shal.
l be supplemen:ed as needed by addi:icnal narra:1ve ra:eria; :: pr:vi:e
(
I l
4 j o.: -
- - - - - ---__,.--,.,-,.---~w , _ - . - , - - . . - , - - -
QUAD-CITIES DPR-29 Note: This item is intended to providefor reporting of potentially generic problemr.
- 2. Tlurty-Day Written Reports ,
The reportable occurrences discussed below have lesser immediate importance than those described un.".cr B.I. above. Such events shall be the subject of written reports to the director of the appropriate regional office within 30 days of occurrence of the event. The written report shall include, as a mmimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
- a. Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.
- b. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.
Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurations as described in Items B.2.a. and B.2.b. need not be reported except where test results themselves reveal a degraded mode as described above.
- c. Observed inadequacies in the implementation of administrative or procedural controls which .
threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems. --
- d. Abnormal degradation of systems other than those specified in Item B.l.c. above designed to contain radioactive material resulting from the fission process.
l Note: Sealed sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limitsfor identifed leakage setfort in technical specifcations need not be reported under this item.
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QUAD-CITIES
, DPR-29
.. C. Unique Reporting Requirements'
- l. Radioactive Effluent Release Report (Semi-Annual)
A semi-annual report shall be submitted to the Commission within 60 days after January I arid July 1 of each year specifying the quantity of each ~
of the radionucildes released to unrestricted areas in liquid and gaseous effluents during the previous 6 months. The format'snd content of the re-port shall be in accordance with Regulatory Guide 1.21 (Revision 1) dated June, 1974. Any changes to the PCP shall be included in this report.
- 2. Environmental Program Data (Annual Report)
An annual report containing the data taken in the standard radiological monitoring program (Table 4.8-4) shall be submitted prior to May 1 of each year. The content of the report shall include:
- a. Results of all environmental measurements summarized in the format of Regulatory Guide 4.8 Table I (December 1975). (Individual sample results will be retained at the Station). In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. Summaries, Interpretations, and analysis of trends of the
-results are to be provided.
- b. An assessment of the monitoring results and radiation dose via the
('
principal pathways of exposure resulting from plant emmissions of radioactivity including the maximum noble gas gamma and beta air doses in the unrestricted area. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (ODCM).
- c. Results of the census to determine the locations of nearest residences and of nearby animals producing milk for human consumption, and the pasture season feeding practices at dalries in the monitoring program (Table 4.8-4).
- d. The reason for the emission if the nearest dairy to the station is not in the monitoring program (Table 4.8-4).
- e. An annual summary of meteorological conditions concurrent with the releases of gaseous effluents in the form of joint frequency distri-butions of wind speed, wind direction, and atmospheric stability,
- f. The results of the Interlaboratory Comparison Program described in section 3.8.D.7
- g. The results of the 40 CFR 190 uranium fuel cycle dose analysis for each calendar year.
- h. A summary of the monitoring program, including maps showing sampling locations and tables giving distance and direction of sampling locations
( from the Station.
- 6. 6-5
l t QUAD-CITIES DPR-29 (g.> -
3 If a confirmed measured radionuclide concentration in'an environmental sampling medium averaged over any calendar quarter sampling period ex-ceeds the reporting level given in Table 4.8-4'and if the radioactivity is attributable to piant operation, a written report shall be submitted to the Director of the NRC Regional Office, with a copy to the Director, Office of Nuclear Reactor Regulation, within 30 days from the end of the.
quarter.
- a. When more than one of the radionuclides in Table 4.8-4 are detected in the medium, the reporting level shall have been exceeded if R. i )E 1 where Ci is the average quarterly concentration of the I th radionuclide in the medium and RL is the reporting level of radionuclide 1.
- b. If radionuclides other than those in Table 4.8-4 are detected and are due to plant effluents, a reporting level is exceeded if the potential annual dose to an individual is equal to or greater than the design objective doses of 10 CFR 50, Appendix 1.
- c. This report shall include an evaluation of any release conditions, C.w- environmental factors, or other aspects necessary to explain the anomalous effect.
- 4. Special Reports Special Reports shall be submitted as indicated in Table 6.6-1.
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-d 6.6-6 l .. - _. .. . . __ - - _ _ - _ _ .
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_ QUAD-CITIES 4.
1 OPR-29 i
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- TABLE' 6.6-1: ,
j .SPECIAL REPORTS.
. Specification Aron: -Reference Submittal Date-a: = Secondary conta inment~leakt ratestes t-5(1) 4.7.C;
. Uponicompletion of.each tes't .
'b. :S umma ry e s ta t us -o f fue l ' pe rfo'rmance . I'.1 Bases After each' refueling. outage.
t 4
, c. Materials' radiation surveillance 4.6.8.2 After each' specimen' removal.
specimens and completion.of analyses.
- d. Evaluation of EGC operation 3 3.F Bases' Upon completion-of
, initial tes ting.
- e. Radioactive Source Leak Testing (2) 4.8.F Annual Report
-f. .Special . Ef fluen ts' Reports .3.3.A. 30 days following occurrence.
3.8.S.
3.8.0.
- 6.6.C.3 j Notes-
[ 1. Each integrated leak rate test of the secondary containment'shall be the subject of c summary technical report. This report snould include data on the aind sceed, win'd direction, outside and inside temoeratures during the test, concurrent reactor ouild-ing pressure, and. emergency ventilation flow rate. The recort shall also include analyses and interpretations of those data which demonstrate cocollance with the j specified leak rate limits.
1
{ 2. This recort is required oni / if the tests reveal the presence of 0.00$ microcuries or more of removable contamination.
!I.
I 6.6-l I
4 9
QUAD-CITIES DPR-29 6.8 Offsite Dose Calculation Manual (0DCM)
A. The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable LCO's contained.in these' Technical Specifications. Method-ologies and calculational procedures acceptable to the Commission arer contained in NUREG-0133 The ODCK.shall be submitted to the Commission at the time of proposed Radiological Effluent Technical Specifications and shall be subject to review and approval by the Commission prior to implementation.
B. Licensee initiated changes to the ODCM may be made provided the change:
- 1. Shall be submitted to the Commission by inclusion in the Monthly Operating Report pursuant to Specification 6.6.A.3 within 90 days of the date the change (s) was made effective and shall contain:
- a. Sufficiently detailed information to support the change.
Information submitted should consist of a package of those pages of the ODCM to be changed together with appropriate analyses or evaluations justifying the change (s);
(])
- b. A determination that the change will not reduce the accuracy of reliability of dose calculations or set-point determinations; and
- c. Documentation of the fact that the change has been reviewed and found acceptable by the onsite review functions.
- 2. Shall become effective upon review and acceptance by the onsite review function.
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,-. - - - - - --,,n , . --- -- .- .- - ---- n .-
QUAD-CITIES
' ' 8 ." DPR- 29 A':
6.9 ProcessControlProgram(PCP)
A. The PCP shall contain the sampling, analysis, and formulation determina-tion by which solidification of radioactive wastes from liquid systems is assured.
B. The-PCP shall be approved by the Commission prior to implementation.
C. Licensee initiated changes may be made to the PCP provided the change:
- 1. Shall be submitted to the Commission in the Radioactive Effluent Release Report for the period in which the change was made and and shall contain:
- a. Sufficiently detailed information to support the change;
- b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- c. Documentation that the change has been reviewed and found acceptable by the onsite review function.
- 2. Shall become effective upon review and acceptance by the onsite
() review function.
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QUAD-CITIES DPR-29 lj ~.
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\QLe 6.510 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, Solid)
A. Licensee initiated major changes to the radioactive waste systems may, be made provided:
- 1. The change -is reported in the Monthly Operating Report for the period in which the evaluation was reviewed by the onsite review function. The discussion of each change shall contain:
- a. A summary of the evaluation that led to the determination >
that the change could be made in accordance with 10 CFR 50.59;
.b. Sufficient detailed information to support the reason for the change;
- c. A detailed description of the equipment,-components, and process involved and the interfaces with other plant systems;
- d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and (or quantity of solid waste that differ f rom those previously predicted in the license application and amendments);
- e. A comparison of the predicted releases of radioactive materials in IIquid and gaseous effluents and in solid waste to the actual
()
i releases for the period in which the changes were made;
- f. An estimate of the exposure to plant operating personnel as a result of the change; and
- g. Documentation of the fact that the change was reviewed and found acceptable by the onsite review function.
- 2. The change shall become effective upon review and acceptance by by onsite review function.
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t 6.10-1
QUAD-CITIES DPR - 30 O
- 11. DOSE EQUIVALENT l-131 - That concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and Isotopic mixture of I-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those IIsted in Table lli of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites".
JJ. PROCESS CONTROL PROGRAM (PCP) - Contains- the sampling, analysis, and formulation determination by which solidification of radioactive; wastes from IIquid systems is assured.
KK. OFFSITE DOSE CALCULATION MANUAL (ODCM) - Contains the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and IIquid effluents, and in the calculation of gaseous and liquid effluent monitor alarm / trip setpoints.
LL. CHANNEL FUNCTIONAL TEST (RADI ATION MONITOR) - Shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify operability including alarm and/or trip functions.
MM .' SOURCE CHECK - The qualitative assessment of instrument response when the sensor is exposed to a radioactive source.
NN. MEMBER (S) 0F THE PUBLIC - Shall include all persons who are not occupa-() tionally associated with the plant. This category does not' include employees of the utility, i ts contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use por-tions of the site for recreational, occupational, or other purposes not associated with the plant.
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- DPR- 33 3.2/4.2 PROTECTIVE INSTRUMENTATION LIMITING' CONDITIONS FOR OPERATION SURVEIi. LANCE' REQUIBFMENTS-Ap'plicability: Applicability:
Applies to the plant-instrumentation Applies,to-the surveillance _ requirements which performs a protective function. of the instrumentation that performs a protective function.
Objective:
Objective:
To.specify the type.and frequency of To assure the operability of pro - surveillance to be applied to protective tactive instrumentation instrumentation.
SPECIFICATIONS A. Primary Containment Isolation A. Primary Containment Isolation Functions Functions Instrumentation and logic systems shall When primary containment integ- be functionally tested and calibrated rity is required, the limiting as indicated in Table 4.2-1.
conditions of operation for the instrumentation that initiates primary containment isolation are given in Table 3.2-1.
fh B. Core and Containment Cooling Sys- B. Core and Containment Cooling Systems -
tems - Initiation and Control Initiation and Control The limiting conditions for opera- Insttumentaticn and logic systems shall tion for the instrumentation that be functionally tested and calibrated initiates or* controls the core and as indicated in Table 4.2-1.
containment cooling systems are given in Table 3.2-2. This instru-mentation must be operable when the system (s) it initiates or controls are required to be operable as specified in Specificatiun 3.5.
C. Control Rod Block Actuation C. Control Rod Block Actuation
- 1. The limiting conditions of opera- Instrumentation and logic systems shall l tion for the instrumentation that be functionally tested and calibrated initiates control rod block are , as indicated in Table 4.2-1.
given in Table 3.2-3. -
- 2. The minimum number of operable instrument channels specified in Table 3.2-3 for the rod block monitor may be reduced by one in
- one of the trip systems for nain-
'u C tenance and/or testing, provided that this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30-day period. If this condition exists for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 30-day period, the system shall be tripped.
3.2/4.2-1 . _
QUAD-CITIE3 DPR-30 D. Refueling Floor Radiation Monitors. D. Refueling Floor Radiation Monit'rs o 1.- Except as.specified in Specifi- The two refueling floor radiation me i-cation 3.2.D.2, the :wo refueling tors shall be functionally tested 2.a i calibrated as indicated in Table a.:-1.
floor radiation monitors shall be operable whenever irradiated fuel Reactor building ventilation isolat. -
or components are present in the and standby gas treatment system initi-fuel storage pool and during re- ation shall be performed at least each fueling or fuel movement opera- operating cycle.
cions.
- 2. One of the two refueling floor radiation monitors may be inopera-ble for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the inopera-
",) ble monitor is not restored to service in this time, the reactor building ventilation system shall be isolated and the standby gas treatment operated until repairs are complete.
i
- 3. The trip setting for the refueling floor radiation monitors shall be set at a value of 100 mR/hr.
- 4. Upon loss of both refueling floor
. radiation monitors while in use, the reactor building ventilation system shall be isolated and the standby gas treatment operated.
E. Postaccident Instrumentation E. Postaccident Instrumentation Th'e limiting conditions for operation Postaccident instrumentation-shall bu for the instrumentation which is read functionally tested and calibrated as
,.out in the control room, required for indicated in Table 4.2-2.
. postaccident monitoring are given in
,- Table 3.2-4.
';)[ 3.2/4.2-2
QUAC-CITIES" DP -30 ~
{A3 F. Control Room Ventilation System ' F. Control Room Ventilation System
- isolation isolation The control room. ventilation system Surveillance for instrumentation is isolated from outside air on a which initiates I' solation'of signal of high drywell pressure,. ' control room ventilation shall be low water level, high main stream- asispeci fied in Table 4.2-1.
line flow, or high radiation in either.of the reactor building ventilation exhaust ducts.
Limiting conditions for operation shall be as indicated in Table 3 2-1 and Specification 3.2.H.1.
G. - Radioactive Liquid Effluent G. Radioactive Liquid Effluent instrumentation Instrumentation The effluent monitoring instru- Each radioactive liquid effluent mentation shown in Table 3 2-5 monitoring instrument shown in shall be operable with alarm set- Table 4.2-3 shall be demonstrated points set to ensure that the limits operable by performance of the of Specification 3.8.B are not given source check, instrument check, exceeded. The alarm setpoints calibration, and functional test shall be determined in accordance ' operations at the frequencies
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s- with the ODCM. shown in Table 4.2-3
- 1. With a radioactive IIquid effluent monitoring instrument alarm / trip setpoint less conservative than
- required, without delay suspend the
- release of radioactive liquid effluents monitored by the affected instrument, or declare the instru-ment inoperable, or change the set-point.so it i s acceptably conserva-
[ tive.
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- 2. With one or more radioactive liquid effluent monitoring instruments l
Inoperable, take the action shown in Table 3.2-5 Exert best efforts to j return the instrument to operable status within 30 days and, if un-successful, explain in the next Semi-l' Annual Radioactive Effluent Release Report why the Inoperability was not corrected in a timely manner. This is in lieu of an LER.
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! 3.2/4.2-3 t
m.,---,- = - - ,-s.m e ,-- - o- - - - +e-m r v. s- , -m --w--r QUAD-CITIES DPR-30 Wv 3 In the event a ilmiting condition for operation and associated. action requirements cannot be satisfieo.
because of circumstances in excess of those addressed in the specifi-cations, provide a 30-day written report to the NRC pursuant to .
Specification 6.6.B.1., and no changes are required in the opera-+
tional condition of the plant, and-this does not prevent the plant from entry into an operational mode. .
H. Radioactive Gaseous Effluent H. Radioactive Gaseous Effluent Instrumentation Instrumentation The effluent monitoring instru- Each radioactive gaseous radiation mentation shown in Table 3.2-6 shall monitoring instrument in Table be operable with alarm / trip setpoints 4.2-4 shall be demonstrated operable set to ensure that the limits of by performance of the given source Specification 3.8.A. are not exceeded. check, instrument check, calibration, The alarm / trip setpoints shall be and functicaal test operations at determined in accordance with the the frequency shown in Table 4.2-4.
(() ODCM.
- 1. With a radioactive gaseous effluent monitoring instrument alarm / trip setpoint less conser-vative than required, without I
delay suspend the release of radioactive gaseous effluents monitored by the affected instru-ment, or declare the instrument inoperable,or change the setpoint e so it is acceptably conservative.
- 2. With one or more radioactive gaseous effluent monitoring Instruments inoperable, take the action shown in Table 3 2-6. Exert best efforts to return the instrument to operable status within 30 days and, if unsuccessful, explain in the next Semi-Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.
This is in lieu of an LER.
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._/ 3 2/4.2-4
QUAD-CITIES DPR-30
.("N 3 In the evelt a Ilmiting condition for operation and associated action requirements cannot be satisfied because of circumstances in excess of those addressed in the specifications, provide a 30-day written report to the NRC pursuant to the Specification 6.6.8.2., and no changes are required in the operational con-dition of the plant, and this does not prevent the plant from entry into an operational mode.
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QUAD CITIES DPR-30
'3.2 LIMITING COBOITION FOR OPERATION BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the conse@ences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious conse@ences. This set of specifications provides the limiting conditions of operation for the primary system isolation fmetion, initiation of the emergency care cooling system, control rod block and staney gas treatment systems. The objectives of the specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any cogonent of such systems even during periods when portions of such systems are out of service for maintenance and (2) to precribe the trip settings required to assure adequate performance. .When necessar), one channel may be made inoperable for brief intervals to conduct required functional tests and cellbrations. Some of the settings on the instrumentation that initiates or control core and containment cooling have tolerances explicitly stated where the high and 104 values are both critical and may have a substantial effect on safety. It should be noted that the setpoints of other instrumentation, where only the high or low and of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Isolation valves are installed in those lines that penetrate the primary containmer.t and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by the protective instrumentation which serves the condition for which isolation is' required (this instrumentation is shown in Table 3.2.1). Such instrumentation must be available whenever primary containment integrity is required. The objective is to -
isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded
& ring an accident.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement. .Thus the discussion given in the basis for Specification 3.1 is applicable here.
The low reactor level instrumentation is set to trip at > 8 inches on the level instrument (top of active fuel is defined to be 360 inches above vessel zero) and after allowing for the full power pressure droo across the steam dryer the low level trip is at 504 inches above vessel zero, or 144 inches above the top of active fuel. Retrofit 8x8 fuel has an active fuel length 1.24 inches longer than earlier fuel designs.
However, present trip setpoints were used in the LOCA analyses (PEDO-24146A, April 1979). This trip initiates closure of Group 2 and 3 primary contalment isolation valves but does not trip the recirculation peps (reference SAR Section 7.7.2). For a j trip setting of 504 inches above vessel zero (144 inches above top of active fuel) and a 60-second valve closure time, the valves will be closed before perforation of the 1
cladding occurs even for the maximum break: the setting is therefore adequate.
The low low reactor level instrumentation is set to trip when reactor water level is 444 inches above vessel zero (with top of active fuel defined as 360 inches above vessel zero, -59 inches is 84 inches above the top of active fuel). This trip initiates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems starts the emergency diesel generator, and trips the recirculation pmps. This trip setting level was chosen to be high enough to prevent spurious operation but low enough to initiate ECCS operation and primary system isolation so that no .91 ting of the fuel cladding will occur and so that postaccident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded. For l
the co@lete circumferential break of a 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary isolation are initiated and in time to meet the above criteria. The instrumentation clso covers the full spectra of breaks and meets the above criteria.
The high-drywell pressure instrumentation is a backup to the water level instrumentation and, in addition to initiating ECCS, it causes isolation of Group 2 isolation valves.
For the breaks discussed above, this instrumentation will initiate ECCS operation at about the same time as the low low water level instrumentation; thus the results given above are applicable here also Group 2 isolation valves include the drywell vent, purge
' and sump isolation valves. High-drywell pressure activates only these valves because high drywell pressure could occur as the result of non-safety-related causes such as not purging the drywell air & ring startup. Total system isolation is not desirable for these conditions, and only the valves in Group 2 are required to close. The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip of Group 1 primary system isolation valves.
6312N 3.2/4.2-5a AMENDMENT NO. ,
QUAD CITIES DPH G The APRMWring especially rod block function operation is flow biased at reduced flow. and prevents a significant re&ction in MCPR, The APRM provides gross core protection i.e., limits the gross withdrawal of control rods in the normal withdrawal sequence. ,
In rated of the refuel and startup/N t stan & y modes, the AR M rod block function is set at 12%
power.
, This control rod block provides the same type of protection in the Refuel and Startup/ Hot Stan&y modes as the APRM flow-biosed 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 boiling in a local region of the core for a single rod 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 assure that, with the y specific trip cladding settings, integrity rod withdrawal safety limit. is blocked before the >CPR reaches the fuel Below 305 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.
arrangement indicated level. is such that the trip setting is less than a factor of 10 above theThe scaling MCPR approaches the MCm fuel cladding integrity safety limit. Analysis of the worst A downscale sensitive enough. indication on an APRM is an indication the instrument has failed or is rod motion, and the control rod motion is thus prevented.In either case the instrument will n The downscale trips are set at 3/125 of full scale.
The SRM rod block with i 100 CPS and the detector not full inserted assures that the SRM's are not withdrawn from the core prior to connencing rod withdrawal for startup.
The scram discharge volume high water level block provide annunciation for operator action.
The alarm setpoint has been selected to provide adequate time to allow determination scram initiation. of the cause of level increase and corrective action prior to automatic For effective emergency core cooling for small pipe breaks the WCI system must function LPCI to operate in time.since reactor pressure does not decrease rapidly enough to allow either cor to the WCI in the event the WCI does not operate.The automatic pressure relief function is provi contacts is such as to provide this function when necessary and minimize spuriousThe arrangemen 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 specificaticn preserves the effectiveness of the system during periods of maintenance, testing or calibration and
- also of minimizes the risk of inadvertent operation, i.e., only one instrument channel out service.
Two radiation monitors are provided on the refueling floor which initiate isolation of the reactor out building and operation of the stanty gas treatment systems. The trip logic is one of two.
Trip settings of 100 mR/hr for the monitors on the refueling floor are based upon initiating normal ventilation isolation and stan&y gas treatment system operation 3.2/4.2-7 ts 6 nt it.
6312N
0.UAD-C I T1 ES .
OPR-30 f.,
so that none of the activity released during the refueling accident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby 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 monitoring of the containnent:
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. Thl's 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 comes from cutside the service building. In the event of an accident, this source of cir 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 s
monitor installed in the intake air duct, signals which Indicate an accident,
~/ l.e., high drywell pressure, low water level, main steamline 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 immedi-cte Isolation of the control room ventilation system and thus minimize any radiation dose.
The radioactive liquid and gaseous effluent instrumentation is provided to
, monitor the release of radioactive materials 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.
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QUAD-CITIES DPR (m L.
. Table 3 2-5 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum No.
of Operable Total No.
Channels of Channels Parameter Action (1) ,
1 1 Service Water A Effluent Gross Activity Monitor 1 1 Liquid Radwaste C Effluent Flow Rate Monitor 1 1 Liquid Radwaste B Effluent Gross Activity Monitor 1 1 Spray Canal Discharge C Blowdown Flow
() Rate Monitor Notes B
Action A: With less than the minimum number of operable channels, releases via i this pathway may continue, provided that at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples are collected and analyzed for beta or gamma activity l
at an LLD of less than or equal to 10-7 uCi/ml.
Action B: With less than the minimum number of operable channels, effluent releases via this pathway may continue, provided that prior to initiating a release, at least 2 independent samples are analyzed, and at least 2 members of the facility staff independently verify the release calculation and discharge valving. Otherwise, suspend i
release of radioactive effluents via this pathway.
Action C: With less than the minimum number of operable channels, releases via this pathway may continue, provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves may be utilized to estimete flow.
k_,
3.2/4.2-15b
QUAD-CITIES DPR- 30 Table 3 2-6 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum No.
of Operable Total No.
Channels (1) of Channels- Parameter- Action (2) 1 2. SJAE. Radiation. D Monitors 1 2 Main Chimney Noble A Gas Activity Monitor 1 1 Main Chimney lodine C Sampler 1 1 Main Chimney C Particulate Sampler 1 1 Reactor Bldg. Vent B Sampler Flow Rate Monitor Reactor Bldg. Vent C (f" ) 1 1 lodine Sampler 1 1 Reactor Bldg. Vent C Particulate Sampler 1 1 Main Chimney Sampler B Flow Rate Monitor 1 1 Main Chimney Flow B Rate Monitor l '
1 2 Reactor Bldg. Vent E Noble Gas Monitor Notes (1) For SJAE monitors, applicable during SJAE operation. For other instrumentation, applicable at all times.
(2) Action A: With the number of operable channels less than the minimum require-l ment, effluent releases via this pathway may continue, provided grab samples are taken at least once per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift and these samples are analyzed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
() Action 8: With the number of operable channels less than the minimum required, effluent releases via this pathway may continue provided that the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3 2/4.2-15c l
QUAD-CITIES DPR-30 Action C: With less than the minimum channels operable, effluent releases via l
this pathway may continue provided samples are continuously collected with auxiliary sampling equipment, as required in Table 4.8-1.
Action 0: With less than the minimum channels operable, gases from the main condenser off gas system may be released to the environment for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided at least one chimney monitor is operable; otherwise, be in hot stand-by in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Action E: With less than the minimum channels operable, immediately suspend
. release of radioactive effluents via this pathway.
E
- t. '. .
I I
l l
(~)
3 2/4.2-15d
QUAD-C: TIES DPR-30 Table ~4.2-1 (Cont'd) f~ ^ Instrument instrument al Instr Channel Functi(
Test Calibration (2) Checkgnt HPCI Isolation
- 1. Steamline high flow (1) Once/3 months None
- 2. Steamline area high temperature. Refueling Outage. Refueling Outagen None 3 Low reactor pressure. (1) Once/3 months None-Reactor Building Vent Isolation and SBGTS Initiation
- 1. Refuel Floor Rad. Monitors (1) Once/3 months Once/ day Control Room Ventilation System Isolation
- 1. Reactor low water level (1) Once/3 months once/ day
- 2. Drywell high pressure (1) Once/3 months None 3 Main steamline high flow (1) Once/3 months Once/ day Notes
- 1. Initial 1 once per month until exposure hours (M as defined on Figure 4.1-1) are 2.0 x 10 ; thereafter, according to Figure 4.1-1 with an interval not less than 1 month nor more than 3 months. The compilation of instrument failure rate data may include data obtained from other boiling water reactors for which the same design Instrument operates in an environment similar to that of Quad-Cities Units 1 and 2. I
- 2. Functional tests, calibrations, and instrument checks are not required when these instruments are not required to be operable, or are tripped.
l l
3 This instrumentation is excepted from the functional' test definition. The functional test shall consist of injecting a simulated electrical signal into the measured channel.
- 4. This instrument channel is excepted from the functional test definitions and shall be calibrated using simulated electrical signals once every 3 months.
5 Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibrations shall be performed during each startup l or during controlled shutdowns with a required frequency not to exceed once per
( week.
l
- 6. The positioning mechanism shall be calibrated every refueling outage.
7 Logic system functional tests are performed as specified in the applicable sec-C tion for these systems.
3.2/4.2-17
QUAD-CITIES DPR-30
- (Psa Table 4.2-3 RADIOACTIVE Liqul0 EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Functional Source Instrument Check (1) Calibration (1) (3) Test (1) (2) Check (1)
Liquid Radwaste Effluent D R Q'(7) (6)
Grass Activity Monitor S2rvice Water Effluent D R Q (7) R Grcss Activity Monitor Liquid Radwaste Effluent (4) R NA NA Flow Rate Monitor Blowdown Flow Rate Monitor (4) R NPL NA Sprey Canal Discharge Notes: -
("31) D = once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> M = once per 31 days Q = once per 92 days R = once per 18 months S = once per 6 months (2) The Instrument Functional Test shall also demonstrate that control room alarm annunciation occurs, if any of the following conditions exist, where applicable.
I
- a. Instrument indicates levels above the alarm setpoint.
- b. Circuit failure.
. c. Instrument indicates a downscale failure.
l d. Instrument controls not set in OPERATE mode.
(3) Calibration shall include performance of a functional test.
(4) Instrument Check to verify flow during periods of release.
(5) Calibration shall include performance of a source check.
(6) Source check shall consist of observing instrument response during a discharge.
l l (7) Functional test may be performed by using trip check and test circuitry associated l with the monitor chassis.
G 3 2/4.2-19
QUAD-CITIES DPR-30 grry Table 4.2-4 w.
RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Calibra- Functional Source Instrument Mode (2) Check (1) tion (l)(4) Tes t (1) (3) Check (1)
Main Chimney Noble Gas B D R Q M Activity Monitor Main Chimney Sampler B D R
()(6) NA l Flow Rate itonitor Re ctor Bldg. Vent Sampler B D R G)(6) NA Flow Rate Monitor Main Chimney Flow Rate B D R Q NA Monitor R2cctor Bldg. Vent B D R Q Q Activity Monitor l SJAE Activity Monitor A D R Q R min Chimney lodine and B D(5) NA NA NA articulate Sampler Re ctor Bldg. Vent lodine B D(5) NA NA NA and Particulate Sampler Notns (1) D = once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> M = once per 31 days '
l Q = once per 92 days R = once per 18 months (2) A = during SJAE operation i
B - at all times (3) The Instrument Functional Test shall also demonstrate that control room alarm annunciation occurs, if any of the following conditions exist, where applicable:
- a. Instrument indicates levels above the alarm setpoint
- b. Circuit failure
- c. Instrument indicates a downscale failure
- d. Instrument controls not set in OPERATE mode.
(4) Calibration shall include performance of a functional test.
(5) instrument check to veri fy operabili ty of sampler; that the sampler is in place l (' and functioning properly.
I (6) Functional test shall be performed on local switches providing low fl ow a larm.
3.2/4.2-20
^
, QUAD-CITIES OPR-30 g 3.8/4.8 RADICACTIVE EFFLUENTS.
p .
Limiting Conditions for Operation Surveillance Requirements Applicability: Applicability: !
Applies to the radioactive effluents. Applies 'to the periodic measurements l from the plant. radioactive effluents. ,
)
Speci fications - l A. Gaseous Effluents A. Gaseous Effluents
- 1. - The dose rate in unrestricted areas at 1. The' dose rates due to radioactive or beyond the site boundary (Figure materials released in gaseous 4.8-1) due to radioactive materials effluents from the site shall be released in gaseous effluents from determined to be within the pre-the site shall be limited to the scribed limits by obtaining repre-following: sentative samples in accordance .
with the sampling and analysis
- a. For Noble Gases: program specified in Table 4.8-1.
The dose rates are calculated using (1) Less than 500 mrem / methods prescribed in the off-Sits year to the whole Dose Calculation Manual (ODCM).
body.
(2) Less than'3000 mrem /
year to the skin.
- b. For iodine-131, for lodine -133, and..for all radionuclides in par .
ticulate form with half-lives !
greater than 8 days less than 1500 mrem / year.
- c. If the dose rates exceed the above limits, without delay
- decrease the release rates to i bring the dose rates within the 1
- mits, and provide prompt 2. The air dose due to releases of notification to the Commission radioactive noble gases in gaseous (6. 6. 8.1. ) effluents shall be determined to be within the prescribed limits by
- 2. The air dose in unrestricted areas at obtaining representative samples or beyond the site boundary due to in accordance with the sampling
- Noble Gases released in gaseous effluents and analysis program specified f' rom"the unit shall be limited to the in sections A and B of Table 4.8-1.
following: The allocation of effluents between units having shared effluent con-
- a. For gamma radiation: trol systems and the air doses are determined using methods prescribed (7 , -(1) Less than or equal to in the ODCM at least once every 31 5 mrad during any cal- days.
endar quarter.
3.8/4.8-1
QUAD-CITIES' DPR- 30 ,
( .)
(2) Less then or equal to 10 mrad during any calendar year.
- b. For Beta radiation:
(1) Less than or equal to 10 mrad during any cal -
endar quarter ,
(2) Less than or equal to 20 mrad during any cal-endar year.
- c. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, pre-pare :nd submit to the Com-4 mission within 30 days, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the cor-rective actions to be taken to
' ({; ensure that future releases are in compliance with 3.8.A.2.a. & b.
This is in lieu of a Licensee Event Report.
- d. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding .
the limits of Specification 3.8.A.2.a. or 3.8.A.2.b., prepare and submit a Special Report to
( the Commission within 30 days e and limit the subsequent re-leases such that the doses or dose con 1mitamnt to a member of the public from all uranium fuel cycle sources is limited to less than or equal to 25 mrem to the tots 1 body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months.
l This Special Report shall include an analysis which demonstrates that radiation exposures to all members of the public from all uranitn fuel l
cycle sources (including all efflu-
-' ent pathways and direct radiation) are less than the 40 CFR Part 190 Standard. Otherwise, obtain a 1
I 3.8/4.8-2
- - - - _ - . ~ . - --_ . - - - - , __ . - . - ..
QUAD-CITIES DPR-30 (x,
variance ~ from the Commission to permit releases which exceed the 40 CFR Part 190 Standard. The radiation exposure analysis contained in the Special Re-port shall use the methods prescribed in the ODCM. This report is in lieu of a Licensee Event Report.
3 The dose to a member of the public in 3 The dose to a member of the pub-unrestricted areas at or bey ~ond the site lic due to releases of iodine-131, boundary from iodine-131, iodine-133, iodine-133, tritium, and all radio-tritium, and all radionuclides in nuclides in particulate form particulate form wi th hal f-lives greater with half-lives greater than 8 than 8 days in gaseous effluents days shall be determined to be released from the unit shall be limited within the prescribed limits by to the following: -
obtaining representative samples in accordance with the sampling
- a. Less than or equal to 7 5 mrem and analysis program specified in to any organ during any calendar Table 4.8-1.
quarter.
For radionuclides not determined
- b. Less than or equal to 15 mrem to in each batch or weekly composite, any organ during any calendar the dose contribution to the cur-() year. rent calendar quarter cumulative summation may be estimated by
- c. With the calculated dose from assuming an average monthly con-the release of iodine-131, iodine- centration based on the previous 133, tritium, and all radionuclides monthly or quarterly composite in particulate form with half-lives analyses. However, for reporting greater than 8 days in gaseous purposes, the calculated dose effluents exceeding any of the contributions shall be based on l
above limits, prepare and submit the actual composite analyses l to the Commission within 30 days, when possible.
a Special Report which identifies the cause(s) for exceeding the The allocation of effluents betweens i
limit and defines the corrective units having shared effluent con-actions taken and the proposed trol systems and the doses are t
actions to be taken to ensure determined using the methods pre-that future releases are in scribed in the ODCM at least once
! compliance with 3.8.A.3. a. s b. every 31 days.
l This is in lieu of a Licensee Event Report.
l 3.8/4.8-3
1 QUAD-CITIES DPR- 30
~
- d. With the calculated dose from the release of Iodine-131, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents exceeding the limits of Specification 3.8.A.3.a.
or 3.8.A.3.b., prepare and submit a Special Report to the Commission within 30 days and limit subsequent releases such that the dose or dose commitment to a member of the public from all uranium fuel cycle sources is limited to less than or equal to 25 mrem to the total body or organ (except the thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months.
This Special Report shall include an analysis which demonstrates that radiation exposures to all members of the pubile from all uranium fuel cycle sources (including all ef-fluent pathways and direct radi-ation) are less than the 40 CFR Part 190 Standard. Otherwise, obtain a variance from the Commission to permit releases which exceed the 40 CFR Part 190 Standard. The radiation exposure analysis con-tained in the Special Report shall use the methods prescribed in the ODCM. This report is in lieu of a Licensee Event Report.
3.8/4.8-4
QUAD-CITIES DPR- 30 0
- 4. Off-Gas System 4. Off-Gas System
- a. At all times during pro- Doses due to treated gases cessing for discharge to released to unrestricted areas the environs,' process and . at or beyond the site boundary control equipment provided shall be-projected at least once to reduce the amount or per 31 days in accordance with concentration of radio- the ODCH..
active materials shall be operated.
- b. The above specification shall not apply for the Off-Gas Charcoal Absorber Beds below 30 percent of rated thermal power.
5 Explosive Gas Mixture 5 Explos'ive Gas Mixture
- a. The concentration of hydrogen Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> verification in the off gas hold up system, will be made that the unit is
! downstream of the recombiner operating within the allowable shall be limited by having a band of the base-line plot of
. recombiner operable within recombiner outlet temperature
{^/
~
the allowable band of the vs. reactor power.
base-line plot of recembiner outlet temperature vs. reactor power, whenever the reactor is operating at a pressure greater than 900 psig.
- b. The recombiner may be inoperable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
- 6. With either the recombiners inoperable, I
or all charcoal beds bypassed for more than 7 days in a calendar quarter while operating above 30 percent of rated thermal power, prepare and submit to l
the Commission within 30 days a special report which includes the following in format ion:
l 3.8/4.8-5
qyAD-CITIES DPR-30
(?i' sa a. Identification of the defective equipment.
- b. Cause of the defective equipment.
- c. Action (s) taken to restore the equipment to an operating status.
- d. Length of time the above requirements were not satisfied.
- e. Volume'and curie content of die waste discharged which was not processed by the inoperable equipment but which required processing.
- f. Action (s) taken to prevent a recurrence of equipment failures.
This is in lieu of a Licensee Event Report.
() 8. Liquid Effluents B. Liquid Effluents.
- 1. The concentration of radioactive 1. The concentration of radio-material released from the site active material in unre-to unrestricted areas at or beyond stricted areas shall be the site boundary (figure 4.8-1) determined to be within the shall be limited to the concen
- prescribed limits by obtain-trations specified in 10 CFR Part ing representative samples 20, Appendix B, Table 11, Column 2. in accordance with the sampling and analysis program specified in Table 4.8-3 The sample analysis results will be used with the calculational methods in the ODCM to determine that the concentrations are within With the concentration of the limits of Specification radioactive material released 3.8.B.1.
from the site to unrestricted areas exceeding the above limits, without delay decrease the release rate of radioactive materials and/or increase the dilution flow rate to restore the concentration to within the
,. above limits.
t
,k/ u l
3.8/4.8-6
QUAD-CITIES
' DPR- 30 O
- 2. The dose or dose commitment -
- 2. a. The dose contributions from
.above background to a member measured quantitics of of the public from radioactive radioactive material shall materials in IIquid effluents be determined by calculation released to unrestricted areas at least once per 31 days at or beyond the site boundary and a cumulative summation from each unit shall be limited of these' total body and any to the following: organ doses shall be maintained for each calendar quarter.
- a. During any calendar quarter:
(1) Less than or equal to 3 mrem to the whole body.
(2) Less than or equal to 10 mrem to any organ.
- b. During any calendar year: b. Doses computed at the nearest community water system will
. (1) Less than or equal to 6 consider only the drinking mrem to the whole body. water pathway and shall be projected using the methods (2) Less than or equal to 20 prescribed in the ODCM at mrem to any organ. least once per 92 days.
- c. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report which identi-fies the cause(s) for exceed-ing the limit (s) and defines the corrective actions taken and the proposed actions to be taken to ensure that future i releases are in compliance with l 3.8.B.2.a. & b. This is in lieu l of a Licensee Event Report.
- d. With the calculated dose from the release of radioactive
. materials in liquid effluents exceeding the limits of
- Specification 3.8.B.2.a. or l 3.8.B.2.b., prepare and submit
! a Special Report to the Commis-sion within 30 days and limit the subsequent releases such
( ). that the dose or dose commi t-ment to a real individual from all uranium fuel cycle sources l Is , limited to less than or 3.8/4.8-7 .
QUAD-CITIES DPR-30 C equal to 25 mrem to the total l body or any organ (except thyroid, ' -
g which is limited to less than or equal to 75 mrem) over 12 consecutive months. This Special Report shall include an analysis.
which demonstrates that radi-ation exposures to all real individuals from all uranium fuel cycle sources (including all effluent pathways and direct radiation) are less than the
. 40 CFR Part 190 Standard. Other-wise obtain a variance from the Commission to permit releases which exceed the 40 CFR Part 190 Standard. The radiation exposure analysis contained in the Special Report shall use methods pre-scribed in the ODCH. This report is in lieu of a Licensee Event Report.
- e. With the projected annual
. (,/ whole body or any internal organ dose computed at the nearest downstream community water system is equal to or exceeds 2 mrem from all radioactive materials re-leased in liquid effluents from the Station, prepare and submit a Special Report within 30 days to the opera-tor of the community water l
system. The report is prepared to assist the opera-tor in meeting the require-ments of 40 CFR 141: EPA Primary Drinking Water Stan-
' dards. A copy of this report will be sent to the NRC.
This is in lieu of a Licensee l Event Report.
()
3.8/4.8-7a
QUAD-CITIES DPR-30
' C. 3 At all times during processing 3. Liquid Vaste Treatment prior to discharge to the environs, process and control equipment pro ' - a. Doses due to liquid releases vided to reduce the amount or con- to unrestricted areas at or centration of radioactive materials - beyond the site boundary shall shall be operated when the projected be projected at least once per dose due to liquid effluent releases 31 days in accordance with the to unrestricted areas (see Figure
- 0DCM.
4.8-1) , when averaged over. ' 31 days ,
exceeds 0.13 mram to the total body ..
or 0.42 mrem to any organ.
-4. If liquid waste has to be or is being discharged wi thout treatment as requi red above, prepare and sub-mi t to the Commission wi thin 30 days, a report which includes the following information:
- a. Identification of the defective equipment.
l
- b. Cause of the defective equip-ment.
() c. Action (s) taken to restore the equipment to an operating status,
- d. Length of time the abov.e require-ments were not satisfied.
l
- e. Volume and curie content of the waste discharged which was not processed by the appropriate equipment but which requi red processing.
- f. Action (s) taken to prevent a
! recurrence of equipment failures.
This is in lieu of a Licensee Event Report.
O 3.8/4.8-8
[
-~,e ,-r---- - , ,r- ,.m, n r,e-, - . - - - , ,_-,e>,n w,re ,,,e-- ,-,,p - - , , . - - - - - - . - - , - - - - - , - - - , , - , - - , - - --
QUAD-CITIES-DPR- 30
&p ,
C. Mechanical Vacuum ? =p C. Mechanical Vacuum Pump l
- 1. The mechanical vacuum At least once during each shall be capable of operating cycle,' automatic being isolated and securing and isolation of secured on a signal of the mechanical vacuum pump main steam high radi- shall be verified.
ation or shall be iso-laced and secured whenever the main steam isolation valves are open.
<C -
'f r <
3.8/4.8-9
QUAD-CIT!I:
OPR--30 (S) O. . . Environmental Monitoring Program '
D. Environnental P.cnitoring Program
- l. The environmental monitoring 1. The radiological environmental program given in Table 4.8-4 monitoring samples shall be shall be conducted except as- collected pursuant to Table 4.8-3 specified below. from the locations specified in the ODCM, and shall be analyzed
- 2. With the radiological environ- pursuant to the requirements of <
mental monitoring program not Table 4.8-4,.4.8-5 and 4.8-6.
being conducted as specified in Table.4.8-4, prepare and submit 2. The results of analyses performed to the Commission, in the Annual on radiological environmental moni-Radiological Operating Report, toring samples shall be summarized a description of the reasons for in the Annual Radiological
- not conducting the program as Environmental Operating Report.
i required and the plans for preventing a recurrence.
Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, contractor omission which is
, corrected as soon as discovered, malfunction of l (])
sampling equipment, or if a person who participates in the program goes out of business.
If the equipment malfunctions, corrective actions shall be completed as soon as practical. - If a person supplying samples goes out of business, a replacement will be found as soon as possible. All deviations from the sampling i schedule shall be described in the annual report.
l 3 With the level of radioactivity 3 The land use census shall be l In an environmental sampling med- conducted at least once per lum at one or more of the locations twelve months between the dates specified in the ODCM exceeding of June 1 and October 1 by a
- the limits of Table 4.8-5 door-to-door survey , aerial when averaged over any calendar survey; road survey, or by con-quarter, prepare and submit to sulting local agriculture 1 the Commission within 30 days authorities.
from the end of the affected calendar quarter, a Special
, f3 Report which includes an iU evaluation of any release conditions, environmental 3.8/4.8-10 v.,., 3 . _ - . . , ,,.- . --.-,,,y, .,- - ,-_.,,,.y , . . . _ , , . ,.,,_.,._,,---.g.,.-_m_,..,.__....__,_.,_____,,m,-,, --
7
> >.< 3 : . d
1-30 d,c o-s or otner ascects wnicn
.=_sec :he IImi ts of Table
' 1-5 to be sxceeded.
Thi: report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event the.
condition shall be reported and described in the Annual Radiological Environmental Operating Report.
- 4. With milk samples unavailable 4. The results of the land use census f-om one or more of the sample shall be included in the Annual locations required by Table 4.8-4, Radiological Environmental identify locations for obtaining Operating Report.
replacement samples and add them to the radiological environmental monitoring program within 30 days.
The locations from which samples were unavailable may then be deleted from the monitoring program. In lieu of a Licensee Event Report, identify the cause O of the unavailability of samples and identi fy the new -location (s) for obtaining replacement samples in the Annual Radiological En-vironmental Operating report and also include in the report a re-vised figure (s) and table for the ODCM reflecting the new location (s).
l S. A census of nearest residences and 5. The results of the analyses per-of animals producing milk for formed as part of the required human consumption shall be con- crosscheck program shall be in-ducted annually (during the grazing cluded in the Annual Radiological
, season for animals) to determine Envi ronmental Operat ing Report.
l their location and number with The analyses shall be done in respect to the site. The nearest accordance with the ODCM.
residence in each of the 16 meteorological sectors shal t also be determined within a r distance of five miles. Thc l census shall be conducted under the following-conditions:
- a. Within a 2-mile radius from *
.- the plant site, enumeration
(,, of animals and nearest residences by a door-to-door l
or equivalent counting technique.
3.8/4.8-11
QUAD-CITIES OPR- 30 .
l
(?? b. *iichln'a 5-mile radius,
'- enumeration of animals by using referenced information from county agricultural agents or other reliable sources.
- 6. With a land use census identi-fying location (s) of animals which yicid(s) an ODCM calculated _
dose or dose commitment greater than the values currently being calculated in Specification
- 4. 8. A. 3, the new location (s) shall be added to the radio-logical environmental monitoring program with 30 days, if possible.
The sampling location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.
7 Radiological analyses shall be performed on samples representative of those in Table 4.8-3, supplied as a part of the intar-laboratory Comparison Program which has been approved by the NRC.
- 8. With analyses not being performed as required, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.
l O
3.8/4.8-12
QUAD-CITIES DPR-30 V .
(~J E. Solid Radioactive Waste E. Solid Radioactive Waste
- 1. The solid radwaste system shall 1. The PCP shall specify the method be used as applicable in accor- and frquency to verify solidifi-dance with the PCP to process cation of radioactive waste.
wet radioactive wastes to meet Actions to be taken if solidifi-shipping and burial ground cation is not verified shall also requirements, be-specified in the-PCP.
- 2. With the provisions of the Process Control Program not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive waste from the site.
A j 0 .
3.8/4.8-13
~
QUAD-CITIES DPR-30 F. M i s ce l l aneous 4 Rad i oact i ve . Mate r i a l s F. Miscellaneous; Radioactive Materials Sources' Sources
(,
Source Leakage Test Each sealed source shall be tested for leakage and/or contamination by the Specification licensee or by other persons speci-fically authorized by the Commission or Each sealed source containing radio- an Agreement state. The test method active material in excess of 100 shall have'a detection sensitivity of microcuries of beta and/or gamma emit - at.least 0.005 microcuries.per test ting material or-5 microcuries of alpha. sample.
i emitting material shall be free of 1 0.005 microcuries of removable con-- Each category of sealed sources shall tamination. . be tested at the frequency described below:
'Each sealed source with removable contamination in excess of the above 1. Sources in use (excluding startup
- limit shall be immediately withdrawn previously subjected to core flux) -
from use and either decontaminated At least once per 6 months for and repaired or disposed of in ac- all sealed sources containing radio-cordance with Commission Regulations. active material:
A complete inventory of radioactive a. With a half-life greater than 30 materials in the licensee's posses- days (excluding Hydrogen 3), and sion shall be maintained current at all times, b. In any form other than gas.
g 2. Stored sources not in use - Each
< - sealed source shall be tested prior to the use or transfer to another licensee unless tested within the previous 6 months. Sea' led sources transferred without a certificate Indicating the last test date shall be tested prior to being placed into use.
A Special Report shall be prepared and e submitted to the Commission pursuant to 3'
Specification 6.6.C.3 i f source leakage tests reveal the presence of 1 0.005 microcuries of removable contamination.
- G. In the event a limiting condition
- for operation and/or associated action requirements identified in sections 3.8.A. through 3.8.E., and 4.8.A.
through 4.8.E. cannot be satisfied l because of circumstances in excess of those addressed in the speci fi-
- cations, no changes are required in the operational condition of the plant, and this coes not prevent the
(,
plant from entry into an operational mode.
3.8/4.S-14
IUAD-CITIES OPR-30 BASES-
. ' 3 3.8/4.8.A.1 GASEOUS EFFLUENTS - DOSE I
This specification is provided to ensure that the dose at the unrestricted area boundary from gaseous effluents from the units on the site will be within the annual dose limits of 10 CFR Part 20 for. unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table 11. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an Individual in an unrestricted area to-annual average concentra-tions exceeding the limits specified in Appendix B, Table 11 of 10 CFR Part 20
-(10 CFR Part 20.106(b)). The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an Individual at or beyond the unrestricted area boundary to less than or equal to 500 mrem / year to the total body or to not less than or equal to 3000 mrem / year to the skin. These release rate limits also restrict, at all times, the cor-responding thyroid dose rate above background to an infant via the cow-milk-infant i
pathway to not less than or equal to 1500 mrem / year for the nearest cow to the plant. For purposes of calculating doses resulting from airborne releases the
' main chimney is considered to be an elevated release point, and the reactor vent stack is considered to be a mixed mode release point.
3.8/4.8.A.2 DOSE, NOBLE GASES This specification is provided to implement the requirements of Sections s 11.8, Ill.A and IV.A of Appendix 1, 10 CFR Part 50. The Limiting Condition i C./
for Operation implements the guides set forth"in Section 11.8 of Appendix 1.
The statements provide the required operating flexibility and at the same time implement the guides set forth in Section' IV. A of Appendix 1 to assure that the releases of radioactive material in gaseous effluents will be kept "as low es is reasonably achievable." The Surveillance Requirements implement the requirements in Section I l l . A of Appendix 1 that conformance with the guides of Appendix i is to be shown by calculational prodecures based on models and data such that the actual exposure of an Individual through the appropriate
- pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with tie methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Ef fluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors", Revision 1, July 1977 The ODCM equations provide for determining the air doses at the unrestricted boundary based upon the historical average j atmospheric conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.
3.8/4.8-15
QUAD-CITIE5 DPR-30 3.8/4.8.A.3 DOSE,RA010100lNES.RA010ACTIVkMATERIALIN?ARTICULATEFORM aN3 RADICMUCLIGES OTHER THAM MCBLE GASES
,7, 1.- This specification is proviced to imolement the requirements of Sections ll.C, Ill.A and IV. A of Appendix 1,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section li,C of Appendix 1. The statements provide the required operating flexibility and at the same ties implement the guides set forth in Section IV.A of Appendix ! to assure that
'the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The 00CM calculational methods specified in the surveillance requirements implements the requirements in Section Ili.A of Appendix i that conformance with the guides of Appendix 1 be shown by calcula-tional procedures based on models and data such that the actual exposure.of an individual through appropriate pathways is unlikely to be substantially under-estimated. The 00CM calculational methods approved by NRC for calculating the doses due to the actual release rates of the subject materials are required to be consistent wi th the methodology provided in Regulatory Guide 1.109, "Calcula-tien of Annual Doses to. Man from Routine Releases of ReactorI",Effluents Revisionfor 1,the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actuaf doses based upon the historical average atmospheric conditions. The release rate specifications for radiciodines, radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted
- f area. The pathways which were examined.in the development of these specifica-N- tions were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumptien by man and 3) deposition onto grassy areas where milk animais graze wi th consumption of the milk by man.
3.8/4.8.A.4 GASEOUS WASTE TREATMENT The OPERABILITY of the gaseous waste treatment which reduces amounts or concentrations of radioactive materials ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be operable when specified provides reaschable assurance that the releases of radioactive materials in gaseous. effluents will be kept "as low as is reasonably achievable". This speci fication implements the requirements of 10 CFR Part 50.36a, General casign Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section II.0 of Appendix I to 10 CFR Part 50.
3.8/4.8.A.S. EXPLOSlVE GAS MIXTURE This specification is provided to ensare that the concentration of poten-in tially explosive gas mixturas contained in the of f gas system is minimi:ed conformance with the requirements of General Design Cri terion 60 of Apcendix A to 10 CFR Part 50.
b ss 3.8/4.8-16 ~
49
QUAD-CITIES OPR-30 Liqul0 EFFLUENTS 3.8/4.8.8.1 CONCENTP.ATION This specification is provided to ensure that the concentration of radio-cctive materials released in liquid waste effluents from the site to unrestricted creas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B,' Table 11, column 2. The concentration limit for noble gases, MPC in air (submersion), was converted to an equivalent concentration in water using the International Commission on Radiological Protection (ICRP) Publication 2.
4 3.8/4.8.8.2. DOSE This specification is provided to implement the requirements of Sections ll A, Ill. A and IV.A of Appendix 1,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section l l .A. of Appendix ! . The statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix ! to assure that the releases of radioactive material in liquid effluents will be kept "as icw as is reasonably achievable". The dose calculations in the ODCM implement the requirements in Section lil.A of Appendix ! that conformance with the guides of Appendix ! be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is 1
C/ unlikely to be substantially underestimated. The equations speci fied in the ODCM for calculating the doses due to the actual release rates of radioactive materials in IIquid effluents will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases o.f Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.113,
" Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of implementing Appendix I", April 1977 NUREG-0113 provides methods for dose calculations consistent with Reg Guide 1.109 and 1.113 3.8/4.8.8 3 LIQUID WAS'E TREATMENT ,
- The operability of the IIquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section 11.D of Appendix i to 10 CFR Part 50.
I
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3.8/4.8-17
QUAD-CITIES OPR- 30 3.8/4.8.0.1 HONITORING-PROGRAM The radiologica! monitoring program required by this specification provides measurements of radiation and of radioactive materials In those ex-posure pathways and for those radionuclides, which lead to the highest potenti-al radiation exposures of individuals resulting from the station operation.
This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measureable concentrations of radioactive materials cnd levels of radiation are not higher than expected on the basis of the ef--
fluent measurements and modeling of the environmental exposure pathways. Pro--
gram changes saay be initiated based on operational experience.
The detection capabilities required by Table 4.8-6 are state-of-the-art for routine environmental measurements in industrial laboratories. The specifled lower limits of detection for 1-131 in water, milk and other food products corre-spond to approximately one-quarter of the Appendix I to 10 CFR Part 50 design cbjective dose-equivalent of 15 mrem / year for atmospheric. releases and 10 mrem /
year for liquid releases to the most sensitive organ and individual. They are based on the assumptions given in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evalu-cting Compliance with 10 CFR Part 50, Appendix I", October 1977, except the change for an infant consuming 330 liter / year of drinking water instead of 510 liters /
year.
3.8/4.8.D.6 LAND USE CENSUS This specification is provided to ensure that changes in the use of un-restricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.
i 3.8/4.8.D.7 CROSSCHECK PROGRAM The requirement for participation in the interlaboratory comparison crosscheck program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for envi.onmental l monitoring in order to demonstrate that the results are reasonably valid.
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QUAD-CITIES DPR-30
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3.8/4.8.C MFCHANICAL VACUUM PUMP
. The purpose of isolating the mechanical vacuum line is to limit release of cctivity from the main condenser. During an accident, fission products would be transported from the reactor through the main steamline to the main condenser.
The fission product radioactivity would be sensed by the main steamline radio-cctivity monitorc which initiate isolation.
3.8/4.8.F. MISCELLANEOUS RADI0 ACTIVE MATERIALS SOURCES The objective of this specificat, ion is to assure that leakage from- byproduct, source and special nuclear material sources does not exceed allowable limits.
The limitations on removable contamination for sources requiring leak testing, including alpha maitters, is based on 10 CFR 70.39(c) limits for plutonium.
3.8/4.8.E. SOLID RADIOACTIVE WASTE The operability of the solid radioactive waste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped off-site. This specification implements the requirements of 10 CFR 50.36a. and General Design Criteria 60 of Appendix A to 10 CFR Part 50.
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i 3.8/4.8-19
~
QUAD-CITIES OPR- 30 TABLE 4.3-1 RADICACT.IVE GASECUS WASTE SAMFLING AND ANALYSIS PROGRAM MINitiUM LOWER LIMIT or GASEOUS SAMPLING ANALYSIS TYPE OF' OETECTION (LLO)
RELEASE TYPE' 'REQt1ENCY ~ :
.REQUENCY ACTl'ilTY ANALYS l 5 (uci/mi)
Principal -
A. Main Chimney b M M Gamma Emitters
- 1 x 10 Reactor Bldg. Grab Sample Tritium i x 10-0
' lent Stack N
- 9. All Release Continuousd ye 1-131 Ix 10~I2 Tyces as Cha rcoa l ~
1-133 1 10~10 Listed in A Samole Above P rinci pa l I ConeInuousd WC Gamma Emieters" 1 x 10'II Particulate (I-131,others) i Samole _ _ , .
]> Continuous d q gg.gg !;x ig-il I-Comoosice [ SR-90 i I
Particulate Samole li 1x 10 ' I '. l-ij Continuousd M Gross Ai:na ,
- x 'O
Comoosite l
? articulate l l
Samole i >
C. Main Chimney Continuousd Noble 'x 'O'6 Gas Monitor Noole Gases D. Reactor Bldg Continuous d Noble Noble Gases 1 x 10-b Vent Stack ( Gas Monitor s
3.3/h.3-20
QUAD-CITIESL
, DPR-30 TABLE 4.8-1 (Continued)
TABLE NOTATION
- a. The lower limit of detection (LLD) is defined in table notation a.
of Table 4.8-6.
- b. Sampling and analyses shall also be performed following shutdown, startup, or a thermal power, change exceeding 20 percent of rated thermal power in I hour if an abnormal change in radionuclide mix-ture or concentration is anticipated.
- c. Samples shall be changed at least once per 7 days and the analyses completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after removal from the sampler. Sampling shall also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following each shutdown, startup, or thermal power level change exceeding 20% of rated thermal
. power in one hour if an abnormal change in radionuclide mixture or concentration is anticipated. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLD's may be increased by a factor of 10.
, d. The ratio of sample flow rate to the sampled stream flow rate shall be known.
i
, (e - e. The principal gamma emitters for which the LLD specification applies i\- . exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions, and Mn-54, Fe-59, j Co-60, Zn-65, Co-58, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. Other peaks which are measurable and identi-fiable by gamma ray spectrometry, together with the above nuclides, shall be also identified and reported when an actual analysis is per-formed on a sample. Nuclides which are below the LLD for the analyses shall not be reported as being present at the LLD level for that nuclide.
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~ ~ , _ _ . _ - - - -. _ _ __ __ _ - - _ _ _ - . _. _ _ _ _ _ _ . _ _ - -
- he : ~ i5
.-< 30
- na i .u :
('N MAXIMUM PERMISSIBLE CONCENTRATION OF DISSOLVED OR ENTRA!NED N0BLE GASES '
RELEASED FROM THE SITE TO UNRESTRICTED AREAS IN LIQUID WASTE ,'
I NUCLIDE MPC (uC l /ml )*f
~
, Kr-87 4x10-5 Kr-88 9x10-5 Ar-41 7x10-5 Xe-131m 7x10-4
() Xe-133m 5x10-4 6x10
-4 Xe-133 Xe-135m 2x10-4 2x10
-4 Xe-135 I
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- Computed f rom Equation 20 of ICRP Publication 2 (1959), adjusted for infinite cloud submersion in water, and R = 0.01 rem / week, density = 1.0 g/cc abd Pw/Pt = 1.0.
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U 3.8/4.8-21a
'l QUAD-CT;IS DFR 30
'O TABLE 4.8-3 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Liquid Release Sampling Analysis Type of Detection (LLD)
Typer Frequency Frequency Activity Analysis (uci/ml) i A. Batch Waste Re- Prior to Prior to Principal lease Tanks Each Batch Each Batch Gamma Emitters
Prior to q Fe-55 lxio-6 Each Batch Composite b Sr-89, Sr-90
'{ 5x10-6 Dissolved &
Prior to M 1x10 -5 I
One Batch /M Entrained Gases (Gamma Emitters)
B. Plant Contin- I-131 lx10-6 uous Releasesd gc(Grab M c ' Princi pa l Sample) Gamma Emitters
- 5x10 -7 Dissolved & Entrained GasesI(Gamma emmi ters 1x10-5 H-3 lx10-5 Gross Alpha 1x10-7 Q Sr-89, Sr-90 -8 Q" 5x10 (Grab lx10 -6 i
Sample) Fe-55 v
3.8/4.8-22 l . . _ . _ _ _ , _ _ . __
QUAD-CITIES DPR-30
(')
TABLE 4.8-3 (Continued)
TABLE NOTATION
- a. The LLD is defined in Notation a. of Table 4.8-6.
- b. A composite sample is one in which the quantity of liquid samples is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is repre- '
sentative of the liquids released.
- c. If the alarm setpoint of the service water effluent moni.or as aeter-mined in the ODCM is exceeded, the frequency of analysis shall be in-creased to daily until the condition no longer exists.
- d. A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed to assure representative sempling. A continuous re-lease is the discharge of liquid wastes of a nondiscrete volume; e.g.,
from a volume or system that has an input flow during the release.
- e. The principal gamma emitters for which the LLD speci fication applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-60, Zn-65, Co-58, Ho-99, Cs-134, Cs-137, Ce-141, and Ce-14h. Other peaks which are measurable and identifiable by gamma ray spectrometry to-
,~ gether with the above nuclides, shall be also identified and reported
(/ when the actual analysis is performed on a sample. Nuclides which are below the LL3 for the analyses shall not be reported as being present at the LLD level for that nuclide.
- f. The dissolved and entrained gases (gamma emitters) for which the LLD specification applies exclusively are the following radionuclides:
Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138. Other dissolved and entrained gases (gamma emitters) which are measurable and identi-fiable by gamma-ray spectrometry, together wi th the above nuclides, shall also be identified and reported when an actual analysis is performed on a sample. Nuclides which are below the LLD for the
- analyses shall not be reported as being present at the LLD level for i
that nuclide.
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l 3.8/4.8-23
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TABLE 4.8-4 RADl01.0GICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Minimum Number of Samples Sampling and Type and Frequency and/or Sample and Sample Locations
- Collection Frequency of Analysis i
l I. AIRBORNE 3
- a. Particulates 16 locations Continuous operation of Gross beta and gamma sampler for a week lsotopic as specified In ODCH.
Y" b. Radioiodine 16 locations Coricinuous operation of I-131 as specified in
, E! samp ler for two weeks ODCH.
Si 2. DIRECT RADIATION Forty Locations Quarterly (Minimum of two TLDs
- per packet) f ^ Sample locations are given on the figure and table in the ODEM.
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QUAD-CITIES DPR-30 TABLE 4.8-4 (Continued)
RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM Exposure Pathway Minimum Number of Samples Sampling and Type and Frequency and/or Sample and Sample Locations
- Collection Frequency of Analysis 3 WATERBORNE
- a. Public Water 2 Locations Monthly composite of Gamma isotopic analysis weekly collected samples of each composite sample y b. Sediment I downstream location in Ar.aua l l y Gamma isotopic analysis oo receiving body of water of each sample oa c. Plant Cooling Intake, Discharge Weekly composite Gross Beta analysis 0; Water of each sample
^ Sample locations are shown on the figure in the ODCH 1
1 i
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QUAD-CITIES DFR-30 i
TABLE 4.3-4 (Continued)
RADIOLOGICAL. ENVIRONMENTAL MONITORING PROGRAM l
Exposure Pathway Minimum Number of Samples Sampling and Type and Frequency
- and/or Sample and Sample LocationsA _ Collection Frequency of Analysis
- 4. INJESTION
- a. Milk 2 Locations At least once weekly 1-131 analysis when animals are on of each sample pasture; at least once Y' per month at other times.
R f" b. Fish I location in receiving Semi-annually Gamma isotopic analysis 2
body of water on edible portions 5
2
- Sample locations are described in the ODCM i
QUAD-CITIES DPR-30 (3
s .e TABLE 4.8-5 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Anclysis- Water.- Airborne Particulate. Fish Milk F6cd Products or Gases (pC1/m3) (pCl/Kg, wet) (pCl/1) (pCi/Kg, wet)
H-3 2 x 100(a)
Mn-54 1 x 103 3 x 10 4
,Fc-59 4 x 102 1 x 10 4
,Co-58 1 x 103 3 x 10 4 Co-60 3 x 10 2 1 x 10 b Zn-65 3 x 10 2 2 x 10 4 Zr-Nb-95 4 x 102 l 131 2 09 3 1 x 10 2 Cs-134 30 10 1 x 10 3 60 I x 10 3 Cs-137 50 20 1 x 10 3 70 2 x 103 Be-La-140 2 x 10 2 3 x 102 (a) for drinking water samples. This is 40 CFR Part 141 value.
(-._
4 3.8/4.8-27
QUAD-CITIES DPR-30
()? TABLE 4.8-6 PRACTICAL LOWER LIMITS OF DETECTION (LLD)
FOR STANDARD ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM Sample Media Analysis LL0a ,b Units (4.66r)^
Ai rborne " Particulate" Gross Beta + 0.01 pCI/m 3 Gamma Isotopic 0.01 pCi/m3.
Airborne I-131 lodine-131 0.10 pC1/m3 Milk / Water 1-131" 5 pci/1 Cs-134 10 pCi/1
, Cs-137 10 A pCi/1 Tri t i um 200 pCi/1 Gross Beta + 5 pCi/1 Gamma Isotopic 20 pCi/t/nucilde Sediment Gross Esta + 2 pCi/g dry I Gamma Isotopic C.2 pCi/9 dry
,, Fish Tissue 1-131 - Thyroid 0.1 ocl/g wet
(~ Cs-134, 137 0.1 pCi/g wet Gross. Beta + 1.0 pCi/g wet Y isotopic 0.2 pCi/g wet O. 0.5 pCi/l on milk camples collected during the pasture season.
+ Referenced to Cs-137 l'
l db 5.0 pCi/1 on milk camples t
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i QUAD-CITIES DPR-30 S. TABLE 4.8-6 (Continued) i{"- TABLE NOTATION -
A. The LLD is the smallest concentration of radioactive material 'in a sample that t
will be detected with 95 percent probability with only 5% probability of false-ly concluding that a blank observation represents a "real" sequal.
For a particular measurement system (which may include radiochemical separation) 4.55.5b LLD = ---------------------------------------
. A* E
- V
- 2.22 -Y
- exp (-iat)
- t Whe re:
LLD is the "a priori" lower limit of detection for a blank sample or background
' analysis as defined above (as pCi per uni t mass or volume).
sb is the square root of the background count or cf a blank sample count; is the estimated standard error of a background count or a blank sample ccunt as appropriate (in units of countt).
E is the counting efficiency (as counts per disintegration).
A is the number of gamma-rays emitted per disintegration for gamma-ray radio-e nuclide analysis (A - 1.0 for gross alpha and tritium measurements).
(s )
V is tne sample size (in units of mass or volume).
2.22 is the number of disintegrations por minute per picocurie.
Y is the fractienal radio-chemical yield when applicable (otherwise Y = 1.0) .
A is the radioactive decay constant for the particular radionuclide (in units of reciprocal minutes) .
At is the elapsed time between the midpoint of sample collection and the start time of counting. (at = 0.0 for environmental samples and for gross alpha measurements).
t is the duration of the count (in units of minutes).
The value of "Sb" used in the calculation of the LLD for a detection system shall i
be based on an actual observed background count or a blank sample count (as appro-priate) rather than on an unverified theoretically predicted value. Typical values of "E", "V", "Y", "t", and "at" shall be used in the calculation.
I
! For gamma-ray radionuclide analyses the background counts are determined from the
- total counts in the channels which are within plus or minus one FWHM (Full Width
! at Half Maximum) of the gamma-ray photopeak energy normally used for the quanti-m tative analysis for that radionuclide. Typical values of the FWHM shall be used j (j,, in the calculation.
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QUAD-CITIES DPR-30 TABLE 4.8-6 (Continued)
{'
TABLE NOTATION The LLD for all measurements is defined as an "A priori" (before the fact) limit representing the capability of a measurement system and not as an "a posterior!"
(after the fact) limit for a particular sample measurement.
B. Other-radionuclides which are measureable and identifiable by gamma-ray spectro-metry, together with the nuclides indicated in Table 4.8-6, shall also be 'identi-fled and' reported when an actual analysis 'is performed on a sample. Nuclides which are below the LLD for the analyses shall not be reported as being present at the LLD level for that nuclide.
C. LLD for drinking water.
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- s. e/ 4.s- 30 ,
- .3-C~;13 m-30 50.!9 to edf7 that su:n acti:::s did not 0:nstituta an u .revir ad ssf etv questien. Prmsed ensetes o -ae naality usurat ce'tropaavaerutica
- snail be reviewee sna approvea 'by the JFanager a C uality a.s.uracea.-
~ '
Z) Freposed cha:ges to ;rocedures, aquip=ent or syscams which involve an unruviewed saf ety question as defined in 10 C7R 50.39.
- 3) Proposed tescs or experiments which involve an unreviewed safety question as defined is 10 CTR 50.59.
- 4) Ptaposed changes in Technical Specification NRC operating licenses. *
- 5) sancompliance with NRC requirements, or of internal procedures, or in-structions beving nuclear saf ety significance.
- 6) ' fignificanc operating abnormalities or deviations from normal and expected.
performance of plant equipment that affect nuclear safety as ref erred to it hy the Onsita Review and lavestigative Function.
- 7) Esportable occurrences requiring 24-hour notification to the ?fRC.
- 8) . AII. recognized indications of an unanticipated deficiency in some aspect of w gn or operation of safety-related structures. systems or components.
- 9) teriew and report findings and recommendations regarding all changes to ther Generacing Stations Emergency Plan prior to implementation of such desage.
- 10) Raeiew and report fiMings ard reco.asendations regarding all items referred hv the tac =f eal Raff Supervisor. Station Saperintendent. Oivisice Vice-Prwideric - % clear Seatiers, and "anager ef Quality As sutaace.
. b. Anadic Function The Audit Function shall be the responsibility of the Manager of Quality Assur-anee ina'opendrac of the froduction Department. Such responsibili y is delegated f
en t!'a Director of Quality Assurance for Operating and to the Staf f Assistant (y to the Manager of Qaality Assu.arce for saintenance gaality assurance actisities.
Pf Heme shall approws the es.dic seer.da and chacklists. the findings and the
. * . report: of each assiit. Audits shall be performed in accordance with the Company Quality Assursace Program and Procedures. Audits shall be per#cr=ed to assure ,
thste safety-rele=d functions are covered within a period of 2 years or less as designated below. ,
- 1) Aedit of the conformance of f acility operation to provisions contained with-1st the Tecluaical Specifications and applicable license conditions at least ance per year.
- 2) Audit of the adherence to procedures, training and qualification of the station staff at least once per year.
- 3) hadit of the results of actions taken to correct deficiencies occurring in facility egaipment, structures systems, or methods of operation that affect nuclear saf ety at least once per 6 months.
- 4) Asadit of the performance of activities required by the Quality Assurance Frogram to amet the Criteria of Appendix "P 10 C7R 50.
b 5) Audit of ttan Facility Emergency Plan and implementing procedures.
- 6) hadir of the Facility Security Plan and implementing procedures.
f 7) .hadit onsite and of fsite reviews.
- 8) Audit the Facility Fire Protection Program and implementing procedures at iaast once, per 24 months.
- 9) The radiological environmental monitoring program and the results thereof ac least onem per 12. months. .
- 10) The CDCM and implementing procedures at least once per 24 months.
g.
- 11) The PCP and implementing procedures for solidification of radioactive vaste ac least once per 2d. months.
i 61-1
1, CyAD-CITIES I DPR-30 ,
y , 4 the Supervisor of the Offsite Review and investigative Function; and (6) submit to the p" j ~
Offsite Review and investigative Function for concurrence in a timely manner, those items
(~ described in Specification 6.1.G.I.a which have been approved by the Onsite Review and
~
i investigative Function. -
t The responsibilities of the Personnel performing this function are stated below:
i I) Review of (1) procedures required by Specification 6.2 and changes thereto and (2)
- any other proposed procedures or changes thereto as determined by the Plant Super-
,f Intendent to affect nuclear safety.
, 3
- 2) Review of all proposed tests and experiments that affect nuclear safety.
v . 2 Q
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- 3) Review of all proposed changes to the Technical Specifications.
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g .< %) Review of all proposed changes or modifications to plant systems or equipment that 3# '
af fect nuclear safety.
L %6ui ji 5) Investigation of all noncompliance with NRC requirements and shall prepare and for-l
". 4 3
Umr h ward a report covering evaluation and recorrmendations to prevent recurrence to the Division Vice President-Nuclear Stations and to the Supervisor of the Offsite Review
, and investigative Function.
3j .
- 6) Review of facility operations to detect potential safety hazards.
1, f): 7) Performance of special reviews and investigations and reports thereon as requested gi by the Supervisor of the Of fsite Review and investigative Ft.nction.
p r
- 8) Review of the Station Security Plan and shall submit recornmended changes to the y Division Vice President-Nuclear Stations.
4 4 9) Review of the Emergency Plan and station implement'ng procedures and shall suomie g reconsnended changes to the Division Vice PresloentWuclear Stations.
7 l 10) Review of reportable occureices, and actions taken to prevent recurrence.
,, q
- 11) Rev'ew of any unplanned on-site release of radioactive material to the environs, T +
Including the preparation and forwarding of reports covering evaluation recommenda-I
- j l! l tions and' disposition of the torret,tive action to prevent recurrence to the Division Vice President-Nuclear Stations, and to the Supervisor of the Offsite Review and L
- j Investigative Function.
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b! 12) Review of changes to the PCP and 00CM, and major changes to the radwaste treatmerit 1 systems.
3 j b. Authority The Technical Staff Supervisor is responsible to the Station Supe-intendent and shall make f
recormnendations in a timely manner in all areas of review, investigations, and quality
- control phases of plant maintenance, operation, and administrative procedures relating
,3 , to facility operations and shall have the authority to request the action necessary to ensure compliance with rules, reguletions, and procedures when in his opinion such action j%, 3 is necessary. The Station Superintendent shall follow such recomrnendations or . select a course of action that is more conservative regarding safe operation of the facility. All '
,g, Au such disagreements shall be r? ported immediately to the Division Vice President-Nuclear y p Stations and the Supervisor of the Of fsite Review and Investigative Function. j
- c. Records
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b I) Reports, reviews, investiga'tions, and recommendations shall be documented with copies to the Division Vice President-Nuclear Stations, the Supervisor of the Of fsite Review
' [tP 3Y and investigative Function, the Station Superintendent, and the Manager of Quality Assurance.
- 2) Copies of all records and documentation shall be kept on file at the station,
- d. Procedures 4 ,
Written administrative procedures shall be prepared and maintained for conduct of the On-
' site Review and investigative function. These procedures shall include the following:
[j !) Content and method of submission and presentation to the Station Superintendent, f
Division Vice President-Nuclear Stations, and the Supervisor of the Of fsite Review and investigative Function.
6.1-5 i
- .'.C- C T IJ
- ?R- 30
" . 2 7'_.3! ??E' CN 2ROCIX3I2 r'.'
'~
Ar lacailad written-procecuras. inclucina 2aoiscaoia ?neckoef .:acs coverine itcas .
alstad below snail se precarac, acpre rec. Acc acnered to:
- 1. Nor=al startup, operacion, anc anuccown of the reactor. and other systems and ccmoonents involving nuclear saf ety of the facility.
- 2. Refueling operations.
- 3. Actions to be taken to correct specific and foreseen potential zalfunctions of systems or components, including responses to alarms, suspected pri=ary
- system leaks, and annormal reactivity changes. .
- 4. Emergency conaicions involving potential or actual release of radioactivity -
" Generating Station Emergency Plan
- and station emergency and abnormal proc edur es.
- 3. Instrumentation operacion which could have an affect on the safety of the facility.
- 6. Preventive and corrective saintenance operations which could have an effect on the safety of the facility.
- 7. Surveillanc e and testing esquirements-
- 3. Tests asd experiments.
- 9. ?rocedure to essure safe shutdown of the plaut.
- 10. Station 3ecurity Plan ano implementation procecures.
- 11. Fire Procaction Program Laplementation.
- 12. ODCM i.olementation.
( 13. ?:P tsplemarcation.
- 3. Radiatica conczol procedures shall be saintained, made available to all station pertoanel, and adnered to. The procatures shall show permissible radiacion ex-posure and sna11 be consistent with the requirements of 10 C'IR-20. This radi-acion pectaccion program shall be organized to meet the requirements of 10 C7R 20.
C. 1. Procedures for items identified in Specificacion 6.2-A and any changes to such procedures shall be reviewed and approved by the Operating Engineer and the Technical Scaff Supervisor in the areas of operacion or fuel handling, and by Maintenance Asst. Supt. and Technical Staff Supervisor in the areas of plant saintenance and plant inspection. Procedures for items identified l La Specification 3.2.3 and any changes to such procedures shall be reviewed and. approved by the Technical Staff Super'risor and the Radiation Chemical Supervisor. At least one person approving each of the above procedures shall hold a valid senior operator's license. In addition, these procedures and changes chereco zust have authorization by the Station Superintendent before being implemenced.
- 2. *4ork and instruction type procedures which implement approved saintenance or sodification procedur.as shall be approved and authorized by the Main-tenance Asst. Supt. where the written authority has been provided by the Station Superintendent. The " Maintenance /Modificacion Procedures" utilized for safety related work shall be so approved only if procedures referenced La the " Maintenance /Modificacion ??ocedure* have been approved as required by 6.2.A. Procedures vnich do not fall within the requirement of 6.2.A or 6.2.B. say be approved by the Department Heads.
D. Temporary changes to procedures 6.2.A. and 6.2.3. above may be made provided:
- l. The intent of the original procedure is not altered.
- 2. The change is approved by two members of the plant management staff. at
,- least one of whom holds a Senior Reactor operacor's 'icense
. on :he unic
,g,,j affected.
3- 3. The change is documented. reviewed by the Onsite Review and :nvestigative Function and approved by the Station Suoerintendent vienin la days of implementacion.
E. Drills of the emergency ;rocedures described in $pecificacion 6.2.A.a. snall be conducted in accorcance with the GSEP Manual.
6.2-1
. QUAD-CITIES DPR- 30 (5) 2 .^ A tabulation shall be submitted on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures gre'ater than'100 mram/yr and their associated man rem exposure according~to work and job function (Note: this tabulation supplements the-requirements of 1 Section 20.407 of 10 CFR 20), e.g... reactor operations and surveillance, 1 - _ inservice inspection, routine maintenance, special maintenance -(describe 3- maintenance), waste processing, and refueling. The d,ose assignments to various duty functions may be estimates based on pocket dosimeter,.TLD, or
) film badge measurements. Small exposures totaling less than 20% of the individual total dose need.not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources ~shall be.
4 assigned to specific major work functions. *
- 3. Monthly Operating Report.
. Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Management Informa-
- tion. and Program Control, U.S. Nuclear Regulatcry Commission, Washington,
- DC 20555, with a copy te the appropriate Regional office, to arrive ne later than the 15th of each month follow;ng the calendar month cevered by the l report. In addition,-cay changes te the CDCM shall be submitted with the Monthly Operating Report within 90 days of the effective date gf the change.
t A report of major change to the radioactive vaste trastment systems shall
-s be submitted with the Monthly Operating Repcre for the period in which the evaluation was reviewed and accepted by the onsite review function. If
- such change is re-evaluated and not inscalled , notification of cancellation '
of the change should be provided to the NRC.
- 3. Reportable Occurrences Reportable occurrences, including corrective actions 'and measures to prevent
+
recurrence, shall be reported to the NRC. In generaI, the importance of an j occurrence with respect to safety significance determines the immediacy of re-l porting required. In some cases, however, the significance of an event may l not be obvious at the time of its occurrence. In such cases, the NRC shall be
- informed promptly of an increased significance in the licensee's assessment of l the event. In addition, supplemental reports may be required to fully descr1be l
final resolution of the occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.
i 1. Prompt Notification with Written Followup The types of events listed below shall be reported as expeditiously as '
possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mail-gram, or facsimile transmission to the director of the appropriate regional off'ce i or his designate no later than the first working day following the event, with a written followup report within 2 weeks. The written followup
~ report shall include, as a minimum, a completed copy of a licensee event fi A. report fo rm. Information provided on the licensee event report form shall be supplemented as needed by additional narrative material to provide
- ..,- _.-_._ _ _._ _ ,6. 6-- 2_ ,
QUAD-CITIES DPR-30 MN4 Note: This item is intended to providefor reporting of potentially generte problems.
(.
- 2. Thirty-Day Written Reports The reportable occurrences discussed below have lesser immediate importance than those described under B.I. above. Such events shall be the subject of written reports to the director of the appropriate regional orlice within 30 days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented as needed. by additional narrative material to provide complete explanation of the circumstances surrounding the event.
- a. Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the technical specificarlons but which do not prevent the fulfillment of the functional requirements of afected systems.
b.
Conditions leading to operadon in a debraded mode permitted by a limiting condition for operation or plant shutdown required by a limi:ing condition for operation.
Note: Routine surveillance testing, instrument calibration or preventative maintenance which require system confrurations as described in items B.?.a. and B.2.b. need r.ot be -eperted except where te:rresults themselves reveal a degraded mode as de:cribed above. s
- c. Obs: ved inadequacies in the implementation of administrative or procedur.ti contrcls whicts threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineeted safety feature systems. .
d.
Abnormal degradation of systems.other than those specisied in item B.l c. abcve destgrad ta contain radioactive material resulting from the 5ssion process.
Note: Scaled sources or catbratic r sources are not included under this item. Leakage of valve packing or gaskets within the limitsfor identifed leakage setfort on technicalspectfcations need not be reported under this item.
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D' .
l 6.6-4 Amendment 40
T :;-::-'I; 34-30 C. Ja s.- m or*. i aa 3eoui renenn.- !
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';i ; set:.: I.'"!uent P,e!aase Report (Semi-Annual)
]
,a : - an u.a i recort shall be submitted to the Commission within 60 days a/cer January 1 and July I of each year specifying'the quantity of each of the radionuclides released to unrestricted areas in liquid and gaseous effluents during the previous 6 months. The format and content of the re-port shall be in accordance with Regulatory Guide 1.21 (Revision 1) dated June, 1974. Any changes to the PCP shall be included in this report.
2 '. Envi ronmental Program Data (Annual Report)
An annual report containing the data taken in the standard radiological menitoring program (Table 4.8-4) shall be submitted prior to May 1 of
~
each year. The content of the report shall include:
- a. Results of all environmental measurements summarized in the format of Regulatory Guide 4.8 Table 1 (December 1975). (Individual sample results will be retained at the Station). In the event that some results are not available for inclusion with tne report, the report shall be submitted noting and explaining the reasons for the missing results. Summaries, interpretations, and analysis of trends of the resalts are to be provided.
- 6. An assessment of the monitoring results and radiation dose via the
'(?
N' principal cathways of exposure resulting from piant emmissions of radioactivity including the maximum ncble gas gamma and beta air doses in the unrestricted area. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (ODCM).
- c. Results of the census to determine the locations of nearest residences and of nearby animals producing milk for human consumption, and the pasture season feeding practices at dairies in the monitoring program (Table 4.8-4).
- d. The reason for the emission if the nearest dairy to the station is not in the monitoring program (Table 4.8-4).
- e. An annual summary of meteorological conditions concurrent with the releases of gaseous effluents in the form of joint frequency distri -
butions of wind speed, wind direction, and atmospheric stability.
- f. The results of the interlaboratory Comparison Program described in section 3.8.D.7
- g. The results of the 40 CFR 190 uranium fuel cycle dose analysis for each calendar year.
- h. A summary of the monitoring program, including maps showing sampling e locations and tables giving distance and direction of sampling locations k./ from the Station.
6.6-5
QUAD-CITIES DPR '()
- 3. If a confirmed measured radionuclide concentration in an environmental sampling medium averaged over any calendar quarter sampling period ex-ceeds th'e reporting level given in Table 4.3-4 and if the radioactivity is attributable to plant operation, a written report shall be submitted to the Director of the NRC Regional Office, with a copy to the Director, Office of Nuclear Reactor Regulation, within 30 days from the end of the quarter,
- a. When more than one of the radionuclides in Table 4.8-4 are detected in the medium, the reporting level shall have been exceeded if R. i )k 1 where Ci is the average quarte.rly concentration of *he i th radionuclide in the medium and RL is the reporting level of radionuclide- i.
- b. If radionuclides other than those in Table 4.8-4 are detected and
, are due to piant affluents, a reporting level is exceeded if the potential ane.ual dose to an individual is equal to or greater than the design objective coses of 10 CFR 50,- Appendix 1.
x, c. This report shall includa an evaluacicn of .3ny release ccnditions,
> enviror. mental factors, or other aspects necessary to expla!n the ancmalcus effect.
- 4. Special Reports Special Reports shall be submitted as indicated in Table 6.6-1.
J 6
= e 6.6-6
QUAD-CITIES DPR- 30 0'~ TABLE 6.6-1 SPECIAL REPORTS Specification Aran Re ference Submittal Date a.. Secondary containment leak rate test (1) 4.7.C Upon completion of each test
- b. Summary statusrof fuel performance 1.1 Bases After each refueling outage.
- c. Materials radiation surveillance 4.6.8.2 After each specimen removal specimens and completion of analyses. ,
- d. Evaluation of EGC operation 3 3.F Bases Upon completion of initial testing.
- e. Radioactive Source Leak Testing (2) 4.8.F Annual Report
- f. Special Effluents Reports 3.8.A. 30 days following occurrence.
3.8.B.
3.8.D.
6.6.c.3
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Notss
. I. Each integrated leak rate test of the secondary containment shall be the subject of a summary technical report. This report should include data on the wind speed, wind direction, outside and inside temperatures during the test, concurrent reactor build-Ing pressure, and emergency ventilation flow rate. The report shall also include cnalyses and interpretations of those data which demonstrate compliance with the specified leak rate limits.
- 2. This report is required only if the tests reveal the presence of 0.005 microcuries or more of removable contamination.
L.
6.6-7
QUAD-CITIES DPR- 30 i
(l%
6.7 Offsite Dose Calculation Manual (ODCM)
A. The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and
, liquid effluents and in the calculation of gaseous and IIquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable LCO's contained in these_ Technical Specifications. Method-ologies and calculational procedures acceptable to the Commission are contained in NUREG-0133.-
The ODCM shall be submitted to the Commission at the time of proposed Radiological Effluent Technical Specifications and shall be subject.to review and approval by the Commission prior to implementation.
B. Licensee initiated changes to the ODCit msy be made provided the change:
- 1. Shall be submitted to the Commission by inclusion in the Morithly Operating Report pursuant to Soecification 6.6.A.3 within 90 days of the date the change (s) was made effective and shall contain:
- a. Sufficiently detailed information to support the change.
Information submitted should consist of a package of those paget of the ODCM to be changed tcgether with appropriate
[; analyses or evaluations justifying the change (s);
- b. A determination that the change w!11 not reduce the accuracy of reliability of oose calculations or set-point determinations; and
- c. Ddcumentation of the fact that the change has been reviewed an'd found acceptable by the onsite review functions.
l 2. Shall become effective upon review and acceptance by the onsite l review function.
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6.7-1 l
i i
QUAD-CITIES DPR- 30 t
/Ti L'
6.8 Offsite Dose Calculation Manual (ODCM)
A. The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and IIquid effluent monitoring instrumentation alarm / trip setpoints consistent with the applicable f.CO's contained in these Technical Specifications. Method-ologies and calculational procedures acceptable to the Commission are contained in NUREG-0133.,
The ODCM.shall be submitted to the Commission at the time of proposed Radiological Effluent Technical Specifications and shall be subject to review and approval by the Commission prior to implementation.
B. Licensee initiated changes to the ODCM may be made provided the change:
- 1. Shall be submitted to the Commission by inclusion in the Monthly Operating Report pursuant to Specification 6.6.A.3 within 90 days of the date the change (s) was made effective and shall contain:
- a. Sufficiently detailed information to support the change.
Inforretion submitted should consist of a package of those pages of the ODCM to be Shanged together with aporopriate .
analyses or evaluations justifying the change (s);
{ .
- b. A determination that the change w!11 not reduce the
, accuracy of reliability of dose calculations or set-point determinations; and
- c. Ddeumentation of the fact that the change has been reviewed an'd found acceptable by the onsite review functions.
- 2. Shall become effective upon review and acceptance by the onsite review function.
t L
i 6.3-1
( , .
~
QUAD-CITIES C'8 ,
OPR 30
.p
.v 6.9 ProcessControlProgram(PCP)
A. The PCP shall contain the sampling, analysis, and formulation determina-tion by which solidification of radioactive wastes from liquid systems is assured.
B. The PCP shall be. approved by the Commission prior to implementation..
C. Licensee initiated changes may be made. to the PCP provided the change:
- 1. Shall be submitted to the Commission in the Radioactive Effluent Release Report for the period in which the change was made and and shall contain:
- a. Sufficiently detailed information to support the change;
- b. A determination that the change did not reduca the overall
' conformance of the solidified waste product to existing criteria for solid wastes; and
- c. Documentation that the change has been reviewed and fcund acceptable by the ensite review function.
- 2. Shall become effective upon review and acceptance by the onsite
(.' review function.
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- l QUAD-CITIES OPR-30 i
.,-x.g
$_< 6.10 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, Solid)
A. Licensee initiated major changes to the radioactive waste' systems may be made provided:
- 1. The change is reported in the Monthly Operating Report for the period in which the evaluation was reviewed by the onsite review function. The discussion of each change shall contain:
- a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;
- b. Sufficient detailed information to support the reason for the change;
- c. A detailed description of the equipment, components, and process involved and the interfaces with other plant systems;
- d. An evaluation of the change which shows the predicted releases of radicactive e.aterials in liquid and gaseous effluents and y (or quantity of solid waste that differ f rom those previously predicted in the license application and amendments);
s
- c. A comparison of the predicted releases of radioactive materials in liquid and gasecus efflucnts and in solid waste to the actual l(,pn_'
j releases for the period in which the changes were made;
- f. An estimate of the exposure to plant operating personnel as a result of the change; and j
- g. Docunentation of the fact that the change was reviewed and found acceptable by the onsite review function.
i
- 2. The change shall become effective upon review and acceptance by by onsite review function.
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9, 6.D@-D ,