ML20091M650
| ML20091M650 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/23/1992 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20091M643 | List: |
| References | |
| NUDOCS 9201290169 | |
| Download: ML20091M650 (61) | |
Text
{{#Wiki_filter:__ FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 REQUEST NO. 173, REVISION 1 RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICAiJONS A. LICENSE DOCUMENT INVOLVED: Technical Specifications ' PORTIONS: INDEX 1.0 Definitions Solidification Page i Ventilation Exhaust Treatment System Page la Liquid Radwaste Treatment System Page la INDEX 3/4.3.3 Radioactive Liquid Effluent Monitoring Instrumentation Page IV INDEX 3/4.3.3 Radioactive Gaseous Effluent Monitoring Instrumentation Page IV INDEX 3/4.7.13 Liquid Radwaste Treatment System Page Vil INDEX 3/4.7.13 Waste Gas System Page Vil INDEX 3/4.7.13 Waste Solidification System Page Vil INDEX 3/4.11 Page Villa INDEX 3/4.12 Page Villa INDEX BASES 3/4.7.13.2 Page XII INDEX BASES 3/4,7.13.3 Page XII INDEX BASES 3/4.7.13.4 Page X11 INDEX BASES 3/4.11 Page XIlla INDEX BASES 3/4.12 Page XIlla INDEX 6.16 Page XVI 1.28 Page 1-6 1.29 Page 1-6 1.30 Page 1-6 1.32 Page 1-7 1.36 Page 1-7 3.3.3.8 Page 3/4 3-42 4.3.3.8 Page 3/4 3-42 TABLE 3.3-12 Pages 3/4 3-43 & 3/4 3-44 TABLE 4.3-8 Pages 3/4 3-45 & 3/4 3-46 3.3.3.9- Page 3/4 3-47 4.3.3.9 Page 3/4 4-47 TABLE 3.3-13 Page 3/4 3-48 thru 3/4 3-50 TABLE 4'.3-9 Pages 3/4 3-51 & 3/4 3-52 3.3.3.10 Page 3/4 5-53 3.7.13.2 Page 3/4 7-49 4.7.13.2 Page 3/4 7-50 3.7.13.3 Page 3/4 7-51 4.7.13.3 Page 3/4 7-52 l 3.7.13.4 Page 3/4 7-53 4.7.13.4 Page 3/4 7-53 3/4 11- Pages 3/4 11-1 thru 3/4 11-15 3/s.'2 Pages 3/4 12-1 thru 3/4 12-12 BASES 3/4.3.3.8- Page B 3/4 3-6 BASES 3/4.3.3.9 Page B 3/4 3-6 9201290169 920123 PDR ADOCK 05000302 P PDR
BASES 3/4.7.13.2 Page B 3/4 7-7 BASES 3/4.7.13.3 Page B 3/4 7-8 BASES 3/4.7.13.4 Page B 3/4 7-8 BASES 3/4.11, Pgs B 3/4 11-1 thru 8 3/4 11-4 BASES 3/4.12 Page B 3/4 12-1 6.8 _ Pages 6-12 thru 6-13b 6.9.1.S(c) Pages 6-14 and 6-14a 6.9.1.5(d) Pages 6 14a thru 6-14c 6.9.2 i thru p Pages 6-17 and 6-18 6.10 Page 6-19 6.14 Page 6 21 6.15 Page 6 21 6.16 Page 6-21 DESCRIPTION OF REQUEST: This submittal requests the deletion of the Radiological Effluent Technical Specification (RETS) requirements in the Technical Specificaticns delineated in the " PORTIONS" section above. This request is consistent with the guidance provided in Generic letter 89-01, " Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relmation of Procedural Details of RETS to the Offsite Dose Calculation Manua. or to the Process Control Program",. dated January 31, 1989. REASON FOX REQUEST: The removal of the RETS requirements is requested in accordance with the guidelines provided in Generic letter 89-01. This request will provide for the
. implementation of programmatic _ controls in Crystal River Unit 3's Technical Specifications (lS) conforming' to the applicable regulatory requirements for radioactive effluents and for radiological environmental monitoring. Inclusion of these controls in TS will allow for the relocation of the current radioactive effluent and environmental monitoring specifications to the Offsite Dose Calculation Manual (0DCM) and the solid ra'iioactive waste specifications to the Process Control Program (PCP).
'Specifically, this request (1) incorporates programmatic controls in the-
- Administrative Controls section of the TS that satisfy the requirements of 10 CFR 20.106, 40'CFR_Part 190, 10 CFR 50.36a. and Appendix 1 to 10 CFR Pait 50, (2) relocates the current specifications involving radioactive effluent monitoring
.. instrumentation, the control of liquid and gaseous effluents, equipment requirements for . liquid and gaseous effluents, radiological environmental monitoring, land radiological reporting details from the TS to the ODCM, (3)
-relocates the definition of solidification and the current specifications on
. solid radioactive wastes to the PCP, (4) siinplifies the associated rerwting requirements,-(5) simplifies the administrative controls for changes to tv ODCM and PCP, (6) adds record retention requirements for changes to the ODCM and PCP, and. (7). updates the definitions- of the' 0DCM and PCP consistent with these changes.
EVALUATION Of REQUEST: The relocation of the RETS requirements from TS to the ODCM and PCP is consistent with the Nuclear Regulatory Commission's Policy Statement on Technical Specification improvement. This request will not reduce the level of control over gaseous and liquid radioactive effluents or solid waste management since programmatic control of the RETS will be maintained in the Administrative Section of the Technical Specifications. I The model specifications listed in Enclosure 3 to Generic Letter 89-01 are requested to be incorporated in '9ystal River Unit 3's Technical Specifications to satisfy the requirements of 10 CFR 20.106, 40 CFR 190,10 CFR 50.36A, and Appendix 1 to 10 CFR 50. The definitions of the ODCM and PCP are to be updated i in accordance with Gene ic Letter 89-01 guidance, and the programmatic and t reporting requirements listed in the Generic letter are proposed for l- incorporation without change in substance as replacement for existing l specifications. The 10 CFR 50.50 process will be utilized as the control mechanism for the relocated specifications, and includes requirements for review and acceptance by the Plant Review Committee (PRC) and approval by the Director, Nuclear Plant Operations (DPNO) prior to implementation. This will allow florida Power
. Corporation to make changes to the specifications which will maintain conformance with federal, State, and other applicable regulations and will not adversely impact the accuracy and reliability of effluent, dose, or setpoint calculations.
The implementing procedures for the- relocated specifications shall also be controlled in accordance with 10 CFR 50.59 and require PRC and DNPO review and approval prior to use, l 1 l-
SHOLLY EVALUATION: Florida Power Corporation proposes that this amendment does not involve a significant hazards consideration. The removal or, as appropriate, update of the Radiological Effluent Technical . Specifications (RETs) will provide for the implementation of programmatic controls in Crystal River Unit 3's Technical Specifications conforming to the applicable regulatory requirements for radioactive effluents and for radiological environmental monitoring which will allow for relocation of these specifications to the Offsite Dose Calculation , Manual and the Process Control Program. This action is consistent with the 1 guidance provided in Generic Letter 89-01.
- 1. Operation of the facility in accordance with the )roposed amendment would not involve a significant increase in the probaatlity of occurrence or consequences of an accident previously evaluated. This change is !
administrative in nature since the existing RETS requirements will be , relocated to the ODCM and PCP and will be controlled by the requirements I stipulated in the Administrative Section of the Technical Specifications. Therefore, the probability of occurrence is not increased and the consequences of previously evaluated accidents is not affected.
- 2. Operation of the facility with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. As stated above, the requirements of RETS will be incorporated into the ODCM and PCP with specific administrative controls remaining in the Technical Specifications and that this change is administrative in nature and is consistent with the guidance provided in Generic Letter 89-01.
- 3. Operation of the facility in accordance with the proposed amendment would not involve a s(gnificant reduction in a margin of safety. These changes do not reduce the margin of safety as the existing requirements will be maintained as part of the ODCM and PCP and will provide for adequate control over radioactive effluent releases, solid waste management, and radiological environmental monitoring activities.
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i LIST Of ATTACliMEN15 ATTACilMENT 1 - Proposed replacement Technical Specification pages ' ATTAtllMENT 2 - Summary of changes for 1SCRN 173, Revision 1 ATTActiMENT 3 - Draft Offsite Dose Calculation Manual revision 1 ATTACllMENT 4 - Draft Process Control Program revision (affected pages only) , 1 n'
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-DEFINITIONS SECTION PEE 1.0' DEFINITIONS DEFINED TERMS 1-1
' THERMAL POWER 1-1 RATED THERMAL POWER l-1 l OPERATIONAL MODE l-1 -
ACTION 1-1 l OPERABLE - OPERABILITY l-1 REPORTABLE EVENT l-2 CONTAINMENT INTEGRITY l-2 CHANNEL CAllBRATION 1-2 CHANNEL CHECK 1-2 CHANNEL FUNCTIONAL TEST l-3
- CORE ALTERATION 1-3
-SHUTDOWN MARGIN 1-3 -
IDENTIFIED LEAKAGE l-3 UNIDENTIFIED LEAKAGE- 1-4 PRESSURE BOUNDARY LEAKAGE l-4 CONTROLLED LEAKAGE 1-4 QUADRANT POWER TILT l-4 DOSE EQUIVALENT l-131 1-4 E --AVERAGE DISINTEGRATION ENERGY l CRYSTAL RIVER - UNIT'3 1 Amendment No.
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1 lEl DEFINITIONS SECTION PE{ l.0 DEJINITIONS (Continued) STAGGERED TEST BASIS 1-5 FREQUENCY NOTATION 1-5 AXIAL POWER IMBALANCE l-5 REACTOR PROTECTION SYSTEM RESPONSE TIME l-5 ENGINEERED SAFETY FEATURE RESPONSE TIME l-6 PHYSICS TESTS 1-6 SOURCE CHECK 1-6 PROCESS CONTROL PROGRAM (PCP) 1-6 0FFSITE DOSE CALCULATION MANUAL l-6 WASTE GAS SYSTEM l-6 PURGE-PURGING l-7 VENTING l-7 INDEPENDENT VERIFICATION 1 MEMBER (S) 0F THE PUBLIC 1-8 SITE B0UNDARY' l-8 UNRESTRICTED AREA 1-8 CORE OPERATING LIMITS REPORT l-8 OPERATIONAL MODES-(TABLE 1.1)- 1-9 FREQUENCY NOTATION (TABLE 1.2) 1-10 CRYSTAL RIVER - UNIT 3 la Amendment No.
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- tlM" 'It#1 CONDITIONS FOR OPERATION AND SURVfill ANCE Rf0VIREMENTS SEC' ON EAGE 3/4.2 POWER DISTRIBUTION LIMITS
;3/4,2.1 AX1AL POWER lMBALANCE 3/4 2-1 3/4.2.2 NUCLEAR HEAT FLUX HOT CHANNEL FACTOR - fg 3/4 2-4 3/4.2.3 NUCLEAP ENTHALPY RISE HOT CHANNEL FACTOR - FN H 3/4 2-6 3/4.2.4 _QUADRANT POWER TILT 3/4 2-8 3/4.2.5 DNB PARAMETERS 3/4 2-12 3/4.3 INSTRUMENIATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM ,
-!NSTRUMENTATION. 3/4 3-9 *
-3/4.3.3 MONITORING INSTRUMENTATION
-Radiation Monitoring Instrumentation 3/4 3-22
-Incore Detectors 3/4 3-26 Seismic Instrumentation 3/4 3-28 Meteorological 7 strumentation 3/4 3-31 Remote Shutdown Instrumentation 3/4 3-34
- Post-accident instrumentation 3/4 .3-37 ,
Fire Detection Instrumentation 3/4 3-40 Waste Gas Decay Tank - Explosive Gas Monitoring Instrumentation 3/4 3-53
--Toxic Gas Systems Chlorine Detection 3/4 3-55 Sulfur Dioxide Detection 3/4 3-56 CRYSTAL RIVER - UNIT 3 IV Amendment No.
ll@ll llMITING CONDITIONS FOR OPERATION AND SURVElllANCE REQUIREMENTS-SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION 3/4 7-13 3/4.7.3 CLOSED CYCLE COOLING WATER SYSTEM Nuclear _ Services Closed Cycle Cooling System 3/4 7-14 Decay Heat Closed Cycle Cooling Water System 3/4 7-15 3/4.7.4 SEA WATER SYSTEM Nuclear Services Sea Water System 3/4 7-16 Decay Heat' Sea Water System 3/4 7 17
-3/4.7.5 ULTIMATE HEAT SINK 3/4 7-18 3/4.7.6 FLOOD PROTECTION 3/4 7-19 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM 3/4 7 20 3/4.7.8 AUXILIARY BUILDING VENTILATION EXHAUST SYSTEM 3/4 7-23 3/4.7.9 HYDRAUllc SNUBBERS 3/4 7-25 3/4.7.10 SEALED SOURCE CONTAMINATION 3/4 7-35 '3/4.7.11 FIRE SUPPRESSION SYSTEMS Water System 3/4 7-38 Deluge and Sprinkler Systems 3/4 7-41 Halon System 3/4 7-44 Fire Hose Stations' 3/4 7-45 3/4/7.12 PENETRATION FIRE BARRIERS 3/4 7 47 3/4.7.13 RADI0 ACTIVE WASTE SYSTEMS.
Whste Gas Decay Tanks 3/4 7-48 Waste Gas Decay Tank - Explosive Gas Mixture 3/4 7-54 ) CRYSTAL RIVER - UNIT 3 Vil Amendment No.
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3/4.7.1- TURBINE CYCLE B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION B 3/4 7-3 3/4.7.3 CLOSED CYCLE COOLING WATER SYSTEM B 3/4 7-3 3/4.7.4- SEA WATER SYSTEM B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK B 3/4 7-4 3/4,7.6 FLOOD PROTECTION B 3/4 7-4 V 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM B 3/4 7-4 3/4.7.8 AUXILIARY. BUILDING-VENTILATION EXHAUST SYSTEM B 3/4 7-5 3/4.7.9- HYDPAULIC SNUBBERS B 3/4 7- 5 3/4.7.10 SEALED SOURCE CONTAMINATION B 3/4 7-6 3/4.7,11 FIRE SUPPRESSION SYSTEMS B 3/4 76 L L 3/4.7.12 PENETRATION FIRE BARRIERS B 3/4 7-6 3/4.7.13.1 WASTE GAS DECAY TANKS B 3/4 .7-7 3/4.7.13.2 DELETED - c L 13/4.~7.13.3 DELETED - 3/4.7.13.4 DELETED - 3/4.7.13.5 EXPLOSIVE' GAS HlXTURE B 3/4 7-8 L L - CRYSTAL RIVER ~ UNIT.3
. XII Amendment No.
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l neu i ADMINISTRATIVE CONTROLS , 1((1108 PEE Heeting frequency 69 Quorum 69 Review 69 , Audits 6 10 ! Authority 6 11 Records 6 11 6.6 REPORTABLE EVENT ACTION 6 11 6.7 SAFETY LIMIT VIOLATION 3 12 - 6.8 PROCEDVRES 6-12 629 REPORTING RE0VIREMENTS 6.9.1 ROUTINE REPORTS 6 13 : Startup Reports 6-13 4 Annual Reports 6 14 Monthly Operating Report 6-15 6.9.2 SPECIAL REPORTS 6-17 6.10 RECORD RETENTION 6 18 bli RADIATION PROTECTION PROGRAE 6 19 6.d2 HIGH R@l Ai;0N ARE6 6-19 6.13 ENVIRONMENTAL QilallFICATION 6 20
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i l CRYSTAL RIVER - UNIT 3. XVI Amendc.ent No.- j l
l l DErlN1110NS . I ENGINEERED. SAFE 1Y fEA1VELf1SPONSE Tith j l 1.25 -The ENGINEERID SAFE 1Y f EATURE RESPONSE 11ME shall be that t8me interval . from when the monitored parameter exceeds its ESF actuation setioint at the i channel sensor until the ESF equipment is capable of perf orming its safety i function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). limes shall include diesel ; gensrator starting and sequence loading delays where applicable, j MY_ElCE llS31 < l.26 PHYSICS TESTS shall be those tests performed to measure the fundamental f' nuclear characteristics of the reactor core and related in't , mentation and 1) described in Chapter 11 of the FSAR, 2) authorized under the provisions of 10 CFR i 50.59, or 3) otherwise approved by the Commission, j iW3CL..Cl1G i 1.27 A SOURCE CHECK shall be the qualitative assessment of channel response when ! the channel sensor is exposed to a radioactive source. M0 CESS CONTROL PROGRAM (PCP) 1.28 The PROCESS CONTROL PROGRAM shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and . packaging of solid radioactive wastes based on demonstrated processing of actual
- or simulated wet solid wastes will be accomplished in such a way as to assure e compliance with 10 CfR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive :
waste. ! 1.29 DELETED OffSITE DOSE CALCVLATION PAWALLW(tQ l.30 The Of fSITE DOSE CALCULATION MANb L shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gescous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental ; Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive : Effluent Controls and Radiological Environmental elonitoring Programs required by , Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi annual Radioactive t Effluent Release Reports required by Specifications 6,9,1.Sc and 6.9.1.Sd. WASTE GAS SYS1EM
-1.31 A WASTE GAS SYSTEM is any equipment (e.g., tanks, vessels, piping) capable
-of collecting primary coolant system of fgases from- the primary system and providi_ng for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment, ,
i L l l l CRYSTAL RIVER - UNIT 3 1-6 Amendment No.
i DErlNITIONS 1.32 DELETED J E GE - PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration tw other operating condition, in such a manner that replacement air or gas is required to ! purify the confinement. YiffilllG l 1.34 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING ; process. . 1BDEPEHOENT VERiftf 2 0B 1.35 INDEPENDENT VI 'rICA110N is a separate act of confirming or substantiating that an activity or n.ndition has been completed or implemented, in accordance with specified requirements, by an individual not associated with the original determination that the activity or condition was completed or implemented in accordance with specified requirements. 1.36 DELETED i CRYSTAL RIVER - UNIT 3 1-7 Amendment No.
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lllS1 M ulIAlls WA11LGaS_QICAY 1 ANK EXPLO11yLR&S_30][l]MlliG_llislM[1{J31[0!j llMITING CONDITION FOR OPERATION __ , 3.3.3.10 The Waste Gas Decay Tanks shall have one hydrogen and one oxygen ; monitoring channel OPERABLE. APPLICABill1Y: During WASTE CAS SYSTEM operation. l ACTION: a. With the number of OPIRABLE channels less than required above, operation of this system, may continue, provided grab samples are collected and analyzed: (1) at least once per 4 hours during degassing operations (2) at least once per 24 hours during other operations
- b. If the affected channel (s) cannot be returned to OPERABLE status within 30 days, submit a special _ report to the Commission l pursuant to Specification 6.9.2 within 30 days describing the reasons for inoperability and a schedule for corrective action.
- c. The provisions of 3.0.3 and 3.0.4 are not applicabic. !
SURVEILLANCE RE0VIREM[lil5 4.1.3.10 The Waste Gas Decay Tank explotive gas monitoring instrumentation shall be demonstrated operable by performing the CHANNEL CilECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CAllBRATION at the frequencies shown in Table 4.310. S l l l r : l 1 CRYSTAL RIVER - UNIT 3 3/4 3 53 Amendment No.
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3/4.3 INSTRUMENTATION BASES 3/4.3.3.8 DELETED 3/4.3.3.9 DELETED 3/4.3.3 10 WASTE CAS DECAY TANK - EXPLOSIVE GAS MONilDB1HS INSTRUMENTATIQN The OPERABILITY of the Waste Gas Decay Tank explosive gas monitoring instrumentation or the sampling and analysis program required by this specification provides for the monitoring (and controlling) of potentially explosive gas mixtures in the Waste Gas Decay Tanks. 3/4.3.3.11 T0XIC GAS SYSTEMS The OPERABillTY of the toxic gas systems ensures that sufficient capability is available to promptly detect ana initiate protective action in the event of an accidental toxic gas release. This capability is required to protect control room personnel and is consistent with guidance provided in Regulatory Guide 1.78, " Assumptions for Evalnating the Habitability of a Nuclear Power Plant During a Postulated Chemical Release', June 1974 and Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release" Revision 1, January 1977. The chlorine detection system is designed so that a chlorine concentration of 15 ppm by voltme is not exceeded in the control room within 2 minutes after detection. The sulfur dioxide detection system is designed so that a sulfur dioxide concentration of 40 ppin by volume is _not exceeded in the control room within 2 minutes after detection. CRYSTAL RIVER - UNIT 3 B 3/4 3-6 Amendment No.
PLANT SYSTEMS BASES _ -3/4.7.13.1--WASTE GAS DECAY TANKS Restr_icting the quantity of radioactivity c.ontained in each waste gas decay tank provides assurance that in the event of a simultaneous uncontrolled release of all of the tanks' contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with FSAR accident analyses. 3/4.7.13.2 DELETED CRYSTAL RIVER - UNIT 3 8 3/4 7-7 Amendment No.
PLANT' SYSTEMS BASES
'3/4.7.13.3 DELETIQ 3/4.7.13.4 DELETED 3/4.7.13.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the Waste Gas Decay Tanks is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50, CRYSTAL RIVER - UNIT 3 8 3/4 7-8 Amendment No.
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- ADH;NISTRA Q CONTR0lS ,,,,, ,,
6.7 SAFETY tlMIT VI'XATION ! 6.7.1 The following actions shall be taken in the event a Safety Limit is i violated: ;
- a. The facilny shall be placed in at least H01 STANDBY within one hour. [
- b. The Safety Limit violation shall be reported to the Commission, the Vice President, Nuclear Operations. and to the NGRC within 24 hours,
- c. A Safety Limit Violation Report shall be prepared. The report '
shall be reviewed by the PRC. This re,iort shall describe (1) applicable circumstances preceding the violation, (2) effects of i the violation upon facility components, systems or structures and (3) corrective action taken to prevent recurrence,
- d. The Safety Limit Violation Report shall be submitted to the Commission, the NGRC and the Vice President, Nuclear Operations within 14 days of the violation. A separate Licensee Event Report '
- aeed not be submitted !f the Safety Limit Violation Report meets the requirements of 10 CFR 50.73 (b) in addition to the requirements above.
6.8' PROCEDURES ANILPROGRAMS 6.8.1 3 COPE Written procedures shall be established, implemented and maintained covering the activities referenced below:
- a. The - applicable proccJures recommended in Appendix "A" of Regulatory Guide 1.33, November, 1972.
- b. Refueling operations.
- c. Surveillance and test activities of safety related equipment,
- d. Security Plan implementation.
! e, Emergency Plan ' implementation. .
- f. . Fire Protection Plan implementation.--
i 9 Systems Integrity Program inplementation. h, lodine .enitoring Program implementation,
- i. PROCESS. CONTROL PROGRAM implementation.
CRYST L RIVER - UNIT 3 6-12 Amendment No.
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i ADMIN]STRATIV[ CONTR01S _ 6.8 EQLLOVELB1DE0fAM KmLhad '
- j. OffSilE DOSE CALCULAil0N MANUAL implementation,
- k. Quality Assurance Program for effluent and enstrohmental I monitoring. r 6.8.2 REVl[W PROCESS 6.8.2.1 Each procedure and administrative policy of 6.8.1 atove, and changes i thereto shall be reviewed and approved prior to implementation as v
follows:
- a. The Emergency Plan, Security Plan, fire Protection Plan and implementing procedures, Administrative Instructions and thos:
test procedures associated with plant modifications that affect nuclear safety shall be reviewed and approved b.y the PRC and the Director Nuclear Plant Operations prior to impicaentation,
- b. For all other procedures, the review cycle shall consist of: an intradepartmental review by a Qualified Reviewer, and inter-disciplinary review by Qualified Reviewer (s) in interfacing departments, as specified in administrative procedures, and approval by the responsible Superintendent or Manager, as s accified by administrative procedures. The PRC shall then review tle 10 CfR 50.59 evaluation within 14 days of approval.
6.8.2.2 The training and qualification of Qualified Reviewers shall be governed by administrative procedures, with final certification oy the Director, Nuclear Plant Operations. Recertification will t,e requ' red on a periodic basis and upon transfer between departments. As a minimum, ali Qualified Reviewers shall meet the requirements of ANSI N18.1 1971, Sections 4.2, 4.3, 4.4, or 4.6, or the equivalent. 6.8.2.3 Each procedure and administrative policy of 6.8.1 shall be reviewed on a periodic basis as set forth in administrative procedures. l CRYSTAL RIVER .VNil 3 6-12a Amendment No. i
f M l.fil M ! $ 1.}. H @ T 8 0l S LL_..fEWELRE.?200M1.ltonutmedj 6.8.3 lemporary changes to procedures of 6.8.1 above may be made provided:
- a. The intent of the original procedure is not altered,
- b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License,
- c. The change is documented and subsequently reviewed and approved within 14 days of implementation, in accordance with the requirements of Specification 6.8.2 6.8.4 The following programs shall be establishco, implemented, and maintained:
- a. Radioactive Effluent Controls Pr0Sr.ED A program shall be provideo conforming with 10 CFR $0.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS Of THE PUBLIC from radioactive effluents as low as reasonably achievabie. The program (1) shall be contained in the l ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- 1) Limitation: on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint cetermination in accordance with the methodology in the ODCM,
- 2) limitations on the concentrations of radioactive material released in liquid effluents to Ut1 RESTRICTED AREAS conforming to 10 CFR Part 20 Appendix B. Table !!, Column 2,
- 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CfR 20.106 and with the methodology and parameters in the ODCM,
- 4) Limitations on the annual and quarterly doses or dose I
commitment to a MEliBER OF THE PUBL!C from radioactive materials in liquid effluents released from the unit to Uf4RESTRICIED AREAS conforming to Appendix 1 to 10 CfR Part 50,
- 5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days, CRYSTAL RIVER - Utili 3 6-13 Amendment fio.
i l EI,),ilSIRAMMOM ROI S , ,
. 6.8,4a Pad.igAcilyc_U fly e n t C o n t ralLf_ tag r am IC on t i n uff0
- 6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce j releases of radioactivity when the projected dt.ses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose comnitment conforming to Appendix 1 to 10 CIR
; Part 50,
- 7) Limitations on the dose rate resulting fr9m radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CfR Part 20, Appendix B Table 11, Column 1, B) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from the unit to areas beyond the SITE BOUNDARY conforming to Appendix ! to 10 CFR Part 50,
- 9) Limitations on the annual and quarterly doses to a MEMBER Of !
THE PUBLIC from lodine-131, lodine-133, tritium, and all radio nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from the unit to
- areas beyond the Si1E BOUNDARY conforming to Appendix 1 to 10 CfR Part 50, and
- 10) Limitations on the annual dose or dose commitn.cnt to any MEMBER Of THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CfR Part-190,
- b. Radioloalcal Environmental Monitoring ProarAnj i
l A program shall be provided to monitor the radiation and radio-
- nuclides in the environs of the plant, The program shall provide (1) representative measurements of radioactivity in the highest -
potential exposure pathways, and (2) verification of the accuracy I of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the t ODCM, (2) conform to the guidance of Appendix 1 to 10 CfR Part 50, ! and (3) include the following: 1
- 1) Monitoring, sampling, analysis, and reporting of radiation and i
radionuclides in the environment in - accordance with the methodo'.ogy and parameters in the ODCM, CRYSTAL RIVER - UNIT 3 6-13a I w ndment No. _ - _ _ . _ _ _ _ _ _ _ _ . ___.__..__._____a
hMIMElRMLVF CONIRolS . 6.8.4b Eidl0hGkJLiny i ro nme nlaL110allgrin9fr00rEL K0Alinked)
- 2) A Land Use Census to ensure that changes in the use of areas at and beyond the SilE BOUNDARY are identified and that 4 modific&tions to the monitoring program are made if required
, by the results of this census, and
- 3) Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance progrem for environmental monitoring.
L.9 REPORilNG REQUIREMENIS E@J1RE REPORTS 6.9.1 In addition to tha appitcable reporting requirements of Title 10 Code of federal Regulations, the following reports shall be submitted to 'i the Director of the Regional Office of Inspection and Enforcement unless otherwise noted. STARTUP REPQHIS 6.9.1.1 A summary report of plant startup and power escalation testing will be submitted following (1) receipt ot' an operating license, (2) amendment to the license involving a planned increase in power level, -(3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that , may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. , 6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test ' program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional spe:ific details requested in license conditions based on other - commitments shall be included in this report. 6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of .the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events, (i.e., initial criticality, completion of startup test program, and the resumption or ; commencement of commercial power operation), supplementary reports shall'be submitted at least every three months until all three events , have been completed. CRYSTAL RIVER - UNIT 3 6 13b Amendment No.- l _ _ ~ . . _ . _ . _ . _ _ . . _ , . _ . _ . _ . , _ _ - , _ _ _ - - - _ , _ - . ,
ADMitilSTRATI,V,JF C0t41R0t S ANjML AND SIMI AtitiyAl REPORTS , 6.9.1.4 Annual repot ts covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March ! of each year. i The initial report shall be submitted prior to March 1 of the year folluwing ' initial criticality. 6.9.1.5 Reports required on an annual basis shall include:
~
- a. A tabulation of the number of station, utility, and other personnel (including contractors) receiving exposures greater than -
100 mrem /yr. and their assqciated man rem exposure according to , work and job functions 3, c.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20 percent of the individual total dose need not be accounted for. In the aggregate, at least 80 percent of the total whole body dose re"ived from external sources should be assigned to specific
-ma .,r work functions,
- b. A list of the reactor vessel material surveillance capsules installed in the reactor at the-end of the report period and a.
summary of any withdri.wals or insertions of capsules during the report period. In supplying this information, the ownership of each capsule shall be indicated and the irradiation location in the vessel of each capsule which was inserted during the report period shall be identified,
- c. Annual Radiolpaical Environmental Oneratina Rep.ati The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the 00CM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix 1 to 10 CFR Part 50.
IThis tabulation supplements the requirements of 20.407-of 10 CFR Part 20,
- CRYSTAL RIVER - UNIT 3 6-14 Amendment No.
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ADMIf()STRAllVE C0tilgt _ __ AjiL4UAL At4D 5EM1 ANtiVAl REPORTS (C.pnLinuedl
- d. $1rdannual RadipAr,1htJHhentlehntlenati The Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year. The report shall include a summary of the quantitles of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCH and PCP and (2) in conformance with 10 CrR 50.36a and Section IV.B.1 of Appendix ! to 10 CFR Part 50.
CRYSTAL RIVER - UNIT 3 6 14a Amendment No. { i l
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ADMINISTRA1)VE CONTROLS
- e. A list of a'il challenges to the Pressurizer Power Operated Relief Valve (POPV) and pressurizer safety valves for the report period.
d 0 [
- CRYSTAL RIVER - UNIT 3 6-14c Amendment No..
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6.9.2 Special reports shall be submitted to the Director of the Office of , inspection and Enf orce.nent, Regior. 11, within the time period specified for each report. These reports shall be submitted covering , the activities identified below. A separate Licensee Event Report, 1 when required by 10 CfR 50.73 (a), need not be submitted if the Special Report meets the requirements of 10 CFR 50.73 (b) in addition to the requirements of the applicable referenced Specification.
- a. ECCS Actuation, Specification 3.5.2 and 3.5.3.
- b. Inoperable Seismic Monitoring Instrumentation. Specification 3.3.3.3.
- c. Inoperable Meteorological Monitoring instrumentation, Specification 3.3.3.4.
- d. Seismic event analysis, Specification 4.3.3.3.2.
- e. Inoperable fire Detection Monitoring Instrumentation, Specification 3.3.3.7.
- f. Specific Activity, Specification 3.4.8. .
9 Results of S t e t.m Generator Tube Inspection. Specification 4.4.5.5 b.
- h. Inoperable fire Suppression System, Specification 3.7.11.1.,
3.7.11.2. 3.7.11.3 and 3.7.11.4. 4 1 DELETED ;
- j. DELETED ,
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- 1. DELETED ,
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l l CRYSTAL RIVER - UNil 3 6-17 Amendment No. 'T y-m-w w
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- p. DELETED
- q. Inoperab)c explosive gas monitoring instrumentation, Specification 3.3,3.10.
f.dQ R[C.QD RETEN110N , 6.10.1 The following records shall be retained for at least five years:
- a. Records and logs of facility operation covering time intervals at ,
each power level. l
- b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety,
- c. All REPORTABLE EVENTS submitted to the Commission,
- d. -Records of surveillance activities, inspections and calibrations rmuired by these Technical Specifications.
- c. Records of reactor tests and experiments,
- f. Records of changes made to Operating Procedures.
- g. Records of radicactive shipments.
- h. Records of sealed source and fission detector leak tests and results.
- 1. Records of annual physical inventory of all sealed source material of record.
6.10.2 The following records shall be retained for the duration of the facility Operating License:
- a. Records and drawing changes reflecting facility design
- modifications made to systems and equipment described in the Final Safety Analysis Report,
- b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup'historles.
- c. Records of facility radiation and contamination surveys. l
- d. Records of radiation ' exposure for all individuals entering radiation control areas.
i 1 CRYSTAL RIVER.- UNIT 3 6-18 Amendment No.
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ADMIN 1STRAllV[ 00fMROL$ m_
- e. Records nf gaseous and liquid radioactive material released to the environs.
- f. Records of transient or operational cycles for those facility components identified in Table 5.7. l.
- g. Records of training and qualification for current members of the plant staff.
- h. Records of inservice inspections performed pursuant to these lechnical Specifications.
- i. Records of Quality Assurance activities required by the QA Manual.
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
- k. Records of_ meetings of the PRC and NGRC.
- 1. Records for Environmental Qualification which are covered under the provisions of paragraph 6.13. ,
- m. Records of analytical results required by the Operational Radiological Environmental Monitoring Program,
- n. Records of reviews performed for changes made to the Off!!TE DOSE CALCULATION MANUAL and the PROCESS CMTROL PROGRAM.
L11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AR(a 6.12.1 In lieu of the " control device"or " alarm signal" required by paragraph 20.203(c) (2) of 10 CFR 20 a High Radiation Area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiation Work Permit and any individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:
- a. A radiation monitoring device which continuously indicates the radiation dose rate in the area, or
- b. An integrating alarming de,imeter which alarms when a preset integrated dose or-dose rate is received. Entry into such areas with this alarming dosimeter may be made after the dose - rate levels in the area have been established and personnel have been made knowledgeable of them, or CRYSTAL RIVER - UNIT 3 6-19 Amendment No.
_ - - _ _ _ _ . __ _ _ . __ _ .. _ _ _ _ _.- _ _ _ _ _ _ _ _ . ~ _ . . .___ __ ADMINISTRATIVf CONTR0!S __ _ _ _ _ _ _ _ _ _ _,,,, 6.14 PRACESS CONTROL PP0 GRAM (PCP1 Changes to the PCP:
- a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3n. This documentation shall contain:
- 1) Sufficient information to support the change t<.,gether with the appropriate analyses or evaluations justifying the change (s),
and
- 2) A determination that the change will malatain the overall conformance of the solidified wasts product to existing requirements of federal, State, or other applicable regulations.
- b. Shall become etfective after review and accept;.1ce by the PRC and the approval of the Director, Nuclear Plant Operations.
6.15 0FFSITE DQH_CALCMLAI.LQ!U2LWAt (0DCM) L Changes to the ODCH:
- a. Shall be documented and records of reviews performed shall be retained as required b.y Specification 6.10.3n. This documentation shall contain:
- 1) Suf ficient information to support the change together with the appropriate analyses or evaluations justifying the change (s),
and ,
- 2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190,10 CFR 50.36a, and Appendix 1 to 10 CfR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations,
- b. Shall become effective after review and acceptance by the PRC and the approval of the Director, Nuclear Plant Operations,
- c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCH was made. Each change shall be. identified by markings in the margin of the affected pages, clearly indicating the area of-the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.
6.16 DELE 1E.D
- CRYSTAL RIVER - UNIT 3 6-21 Amendment No.
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SUMMARY
Of CHANGES FOR THE REMOVAL Of RETS FROM THE TEEllNICAL SPECiflCATIONS All of the specifications listed below are to be relecated to the ODCM or the PCP, or revised as provided for in Generic letter 89 01. SPEClflCATION OLD PAGE DESCRIPTION 1.0 SOLIDiflCATION I Relocated in accordance with GL 89 M . (INDEX) 1.0 VENTILATION EXHAUST la Relocated since the only specification 1REA1 MINT SYSTEM referencing this definition was (INDEX) relocated per GL 89 01, however, GL 89 01 did not provide for this action in the der 1NITIONS section. 1.0 LIQUID RADWASTE la Relocated since the only specification TREA1 MENT SYSTEM referencing this definition was (INDEX) relocatedd per GL 89 01, however. GL 89-01 did not provide for this action in the DEflN1110NS section. 3/4.3.3 Radioactive Lis,uid IV Relocated in accordance with GL 89 01. Effluent Monitoring Instrumentation (INDEX) 3/4.3.3 Radioactive Gaseous IV Relocated in accordance with GL 89 01. Effluent Monitoring Instrumentation (INDEX) 3/4.7.13 Liquid Radwaste Vll Relocated in accordance with GL 89 01. Treatment System (INDEX) 3/4.7.13 Waste Gas System Vil Relocated in accordance with GL 89-01. (INDEX) 3/4.7.13 Waste Solidification Vil Relocated in accordance with GL 89 01. System (INDEX) 3/4.11 RADI0 ACTIVE EFFLUENTS Villa Relocated in accordance with GL 89-01. (INDEX) 3/4.12 RADIOLOGICAL ENVIRON- Villa Relocated in accordan.:e with GL 89 01. MENTAL MONITORING (INDEX) 3/4.7.13.2 LIQUID WASTE XII Relocated in accordance with GL 89 01.- TREATMENT
-(INDEX-8ASES)
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F SP(X1[1 CAT 10N __ QLD PAGE DEiGIPTION ' 3/4.7.13.3 WASTE GAS SYSTEM Xll Relocated in accordance with GL 89-01. (INDEX - BA'tS) 3/4.7.13.4 SOLID RADIDACTIVE X11 Relocated in accordance with GL 89-01. WASTE (INDEX BASES) 3/4.11 RADIDACTIVE EFFLUENTS Xilla Relocated in accordance with GL 89-01. (INDEX - BASES) 3/4.12 RADIOLOGICAL ENVIRON- X111a Relocated in accordance with GL 89-01. HENTAL MONITORING (INDEX - BASES) 6.16 MAJOR CHANGES TO RADIO. XVI Relocated in accordance with GL 89 01. ACTIVE WASTE (REATHENT SYSTEMS (INDEX) 1.28 PROCESS CONTROL PROGRAM l-6 Updated definition in accordance with GL 89 01; Listed as Section 1.22 in GL 89 01. 1.29 SOLIDiflCATION 1.6 Relocated in accordance with GL 89-01; Listed es Section 1.32 in GL 89 01. 1.30 OffSITE DOSE CALCULATION 1-6 Updated definition in accordance MANUAL with GL 89 01; Listed as Section 1.17 in SL 89 01. 1.32 VENTILATION EXHAUST l-7 Relocated since the only specification TREATHENT SYSTEM which referenced this definition was relocated in accordance with GL 89-01, 1.36 LIQUID RADWASTE l-7 Relocated since the only specification TREAlhENT SYS1LM which referenced this definition was relocated in accordance with GL 89-01, 3.3.3.8 RADIDACTIVE LIQUID 3-42 Relocated in accordance with GL 89 01; EFFLUENT MONITORING Listed as Section 3.3.3.10 in INSTRUMENTATION GL 89 01. TABLE 3.3 li 3-43 Relocated in accordance with Gl. 89-01; and Referenced only ir. Section 3.3.3.8. 3 44 TABLE 4.3 8- 3 45 Relocated in accordance with GL 89 01; and Referenced only in Section 3.3.3.8. 3 46 3.3.3.9 RADIDACTIVE GASEOUS 3-47 Relocated in accordance with GL 89 01; EffLVENT MONITORING Listed as Secticn 3.3.3,1) in INSTRUMENTATION GL 89 01.
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L SPEClflCal10N OLD PAGE _D[SIRIEJ10N l i TABLE 3.3 13 3 48 Relocated in accordance with GL 89 01; l thru Referenced only in Section 3.3.3.9. 3 50 TABLE 4.3-9 3 51 Relocated in accordance with GL 89-01; ; and Referenced only in Section 3.3.3.9. ! 3 52 3.3.3.10 WASTE GAS DECAY 3 53 Revised per GL 89 01, Listed as TANK EXPLOSIVE Section 3.3.3.11 in GL 89 01. GAS MONITORING 3.7.13.2 LIQUID RADWAS1E 7 49 Relocated in accordance with GL 89 0); TREATHENT SYSTEM and Listed as Section 3.11.1.3 in 7 50 GL 89 01. 3.7.13.3 WASTE CAS SYSTEM 7 51 Relocated in accordance with GL 89 01; and Listed as GASE0US RADWASTE TREA1 MENT 7 52 or VENTILATION EXHAUST TREATMENT SYSTEh, Section 3.11.2.4 in GL 89 01. 3.7.13.4 WASTE SOLIDiflCA110N 7-53 Relocated in accordance with GL 89-01; SYSTEM Listed as SOLID RADI0 ACTIVE WASTES, Section 3.11.3 in GL 89 01. 3.11.1.1 LIQUID EFFLUENTS: 11 1 Relocated in accordance with GL 89-01. CONCENTRATION TABLE 4.11-1 11-2 Relocated in accordance with GL 89 01; thru Referenced only in Section 3.11.1. 11-4 3.11.1.2 LIQUID EffLVENTS: 11-5 Relocated in accordance with GL 89 01. DOSE and 11-6 3.11.2.1 GASE0US EFFLUENTS: 11-7 Relocated in accordance with GL 89 01. DOSE RATE TABLE 4.11-2 11-8 Relocated in accordance with GL 89-01;
.thru Referenced only in Section 3.11.2.1.
11-10 t 3.11.2.2 DOSE - NOBLE GASES 11-11 Relocated in accordance with GL 89-01; Title differs from GL-89-01. 3.11.2.3 DGJE - 10 DINE-131, 11-12 Rr: located in accordance with GL 89-01; TRITIUM, AND RADIO- and Title differs from GL 89-01. ACTIVE PARTICULATES 11-13
SPECIFICrd01 010 PAGE DESCRIP110J4 f 3.11.3 TOTAL DOSE 11-14 Relocated in accordance with GL 89 01;
- and Listed as Section 3.11.4 in 11 15 GL 89 01.
3.12.1 MOH110 RING PROGRAM 12 1 Relocated in accordance with GL 89 01. and 12 2 ; TABLE 3.12-1 12 3 Relocated in accordance with GL 89 01; thru Referenced only in Secticn 3.12.1. 12 5 TABLE 3.12-2 12 6 Relocated in accordance with GL 00 01; Referenced only in Section 3.12.1. TABLE 4.12 1 12 7 Relocated in .v.cordance with GL 89 01; thru Referenced only in Section 3.lf.l. 12 9 3.12,2 LAND USE CENSUS 12 10 Relocated in accordance with GL 89 01, and 12 11 3.12.3 INTERLABORATCRY 12-12 Relocated in accordance with GL 89 01. ' COMPARISON PROGRAM B3/4.3.3.8 a B3/4.3.3,9 B3/4 3 6 Relocated in accordance with GL 89 01; Bases not specifically addressed in GL., B3/4.7.13,2 B3/4 7 7 Relocated in accordance with GL 89 01; Bases not specifically addressed in GL. B3/4.7,13.3 & B3/4.7.13.4 B3/4 7 8 Relocated in accordance with GL 89 01;- Bases not specifically addressed in GL. , B3/4,11.1 thru B3/4.11.3 B3/4 11-1 Relocated in accordance with GL 89 01; thru Bases not specifically addressed B3/4 11-4 in GL, B3/4.12,1 thru B3/4.12.3 B3/4 12 1 Relocated in accordance with GL 89 01; Dases not specifically addressed in GL. 6.8 PROCEDURES AND PROGRAMS 6 12 Changed title of.6.8 from PROCEDURES and to PROCEDURES AND PROGRAMS as 6 12a listed in GL 89 01. i 6,8.a a & b (Programs) 6-13 Added effluents control and environ-mental monitoring programs per GL 89-01; Listed as 6.8,4 g & h in the GL. , - . - . a _,,; - , . -- - . - - - - ..- .- -. -- --
=SPECiflC6110N QLD PAGE DESCRIP110N , !
6.9.1.5 c & d-(Reports) 6 14 Reworded Sections to match the GL thru 89 01 wording for Annual and l' 6 14c Semiannual reports Listed as 6.9.1.3 and 6.9.1.4 in GL 89 01. 6.9.2 SPECIAL REPOR15 6 17 Relocated parts 'i' thru 'p' in and accordance with GL 89 01 6 18 Specifications referencing these l sections relocated per GL 89 01. ! 6.2.10 (n) RECORDS 6 19 Added new part 'n' in accordance i with GL 89 01: Listed as Section : 6.10.3 (o) in the GL. ; 6.14 PROCESS CONTROL PROGRAM b 21 Revised the wording in accordance I with Gt 89 01; Listed as Section 6.13 ! in the GL. , 6.15 OffSITE DOSE 6 21 Revised the wording in accordanco CALCULATION MANUAL with GL 89 01: Listed as Section 6.14 in the GL. i 6.16 MAJOR CHANGES 10 6 21 Relocated per GL 89 011 f RAD 10ACTIVC WASTE Listed as Section 6.15 in the GL, , TREATMENT SYS1[MS ! i ADM110MLSNfJECAU,0M : ADDRESSLD ILGLJ2:01 3/4.11.1.4 LIQll10110LDUP TANKS This specification - was never i" incorporated into CR 3's Tech Specs because of hydrologic considerttions for the site (see Amndt 69). 3/4.11.2.5 EXPLOSIVE GAS HlXTURE Requiraments retained in accordance with GL 89 01 (CR 3 references are 3.3.3.10 and 3.7.13.5, pages 3/4 3 53 and 3/4 7-54,respectively). , 3/4.11.2.6 GAS STORAGE TANKS- Requirements retained in accordance with- ' GL 89 01 (CR 3 reference is 3.7.13.1, pg. 3/4 7-48). ' 5.1.3 SITE MAP- - Retained in accordance with u 89 01 (CR-3 reference is 5.1.3, pg. 5 1). i r 6 g..,,m+. N-,- , - - - - - -,r.-s-p+,.,.. , ,-w,, . p - - , --.,,.vn., w,. ,w ,-;.-n ,-m,v.~ ,
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CRYSTAL RIVER - UNIT 83 0FT SITE DOSE CALCULATION MANUAL l
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INTRODUCTION The _ Off-site Dose Calculation Manual (0DCH) is provided to support implementation of the Crystal River Unit 3 radiological effluent controls. The ODCM is divided into two parts. Part I contains the control specifications for liquid and gaseous radiological effluents which were relocated from the Technical Specifications in accordance with the provisions of Generic Letter 89 01 issued by the NRC in Jaluary, 1989. Part 11 of the ODCH contains the calculational methods to be used in determining the dose to members of the public resulting from routine radioactive effluents released from Crystal River Unit 3. More accurate estimation of doses is performed annually in preparation of the year end Semiannual Radioactive Effluent Release Report. Part 11 also contains the methodology used to determine effluent monitor alarm / trip setpoints which assure that releases of radioactive materials remain within specified concentrations. The ODCM will be controlled by the Site Nuclear Services Department and revisions should be made with the approval of the Manager, Site Nuclear I Services. The ODCM shall become effective after the review and approval of the Plant Review Committee and approval by the Director, Nuclear plant Operations in accordance with Technical Specification Section 6.15. Changes to the ODCM shall be documented and records of reviews performed shall be retained as required by Technical Specification Section 6.10.3n. This documentation shall contain sufficient information to support the change (including analyses or evaluations), and a determination that the change will maintain the level of radioactive effluent control required by the regulations listed in Technical Specification Section 6.15 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. Historical documentation and distribution of the ODCM shall be the responsibility of the Nuclear Operations Records Manager in accordance with N00-05, Document Control Program, in accordince with Technical Specification Section 6.15, changes shall be submitted to the NRC in the form of a complete and legible copy of the entire ODCM as part of, or concurrent with, the Semiannual Radioactive t.' fluent Palease Report for the period of the report in which any change to el 1 ODCM u s made. Each change shall be identified by markings in the ma ' gin of the affected pages, clearly indicating the area of the page that v.s changed, and shall indicate the date (e.g., month / year) the change was implemented. l m_ i
IABLE OF CONTENTS PART I - SPECIFICATIONS Section 1.0 Definitions 1.1 Channel Calibration 1.2 Channel '. heck 1.3 Channel functional lest 1.4 frequency 1.5 Independent Verification 1.6 Liquid Radwaste Treatment System 1.7 Member of the Public 1.8 Mode 1.9 Offsite Dose Calculation Manual 1.10 Opetable - Operability 1.11 Site Boundary 1.12 Source Check 1.13 Unrestricted Area 1.14 Ventilation Exhaust Treatment System 1.15 Waste Gas System 1.16 Purge - Purging 2.0 Specification 2.1 Radioactive Effluent Monitoring Instrumentation 2.2 Radioactive Gaseous Effluent Monitoring Instrumentation 2.3 Liquid Radwaste Treatment System 2.4 Wasta Gas System 2.5 Liquid Effluents Concentration 2.6 Liquid Effluents Dose
t t PARf 1 SPECIFICATIONS (CON'T) AttliDD ' 2.0 Specification (Con't) 2.7 Gaseous Effluents Dose Rate 2.8 Dose Noble Gases 2.9 Dose 1 131. Tritium, and Radioactive Particulates 2.10 Total Dose 2.11 Radiological Environmental Monitoring
- i 2.12 Land Use Census 2.13 Interlaboratory Comparison Prog'am 2.14 Special Reports 3.0 $pecification Bases 3.1 Radioactive Effluent Monitaring Instrumentation Basis 3.2 Radioactive Gaseous Effl',ent Monitoring Instrumentation Basis 3.3 Liquid Radwaste Treatment System Basiw 3.4 Waste Gas System Basis .
3.5 Liquid Effluents Concentration Basis 3.6 Liquid Effluents Dose Basis 3.7 Gaseous Effluents Dor.t Rate Basis 3.8 Dose Noble Cases Basis 3.9 Dose 1 131. Tritium, and Radioactive Particulates Basis 3.10 Total Dose Basis 3.11 Radiological Environmental Monitoring Basis 3.12 Land Use Census Basis 3.13 Interlaboratory Comparison Program Basis
IA EE OF CONTENTS PART !! - HETHODOLOGIES Section P_Ait 1.0 RADI0 ACTIVE EFFLUENTS MONITOR SETP0 INT SPECIFICAT!GN5 1 1.1 Effluent Monitor Setpoint Specifications 3 1.2 Nuclide Analyses 6 1.3 Pre Release Calculations 11 1.4 Setpoint Calculations 17 2.0 '
..ADl0ACT!;; EFFLUENTS DOSE REDUCTION SPECIFICATIONS 28 2.1 Waste Reduction Specifications 30 2.2 Dose Projection Methodology 32 2.3 Total Dose Specification 34 3.0 RADICACTIVE EFFLUENTS SAMPLING SPECIFICATIONS 38 3.1 1 Liquid Releases (Batch)
40 3.1 2 Liquid Releases (Continuous) 40 3.1 3 Gaseous Releases (Waste Gas Decay Tanks).
40 3.1 4 Gaseous Releases (RB 1 AB) 40 3.1 5 Reactor Bldg. with Personnel and Equipscnt Hatches Open 40a 4.0 RADI0 ACTIVE EFFLUENTS DOSE CALCULATION SPECIFICATIONS 41 4.1 Dose Specifications 43 4.2 Nuclide Analyses 46 4.3 Dose Calculations 51 4.4 Dose Factors 54 5.0 ENVIRONNENTAL N0NITORING 79 6.0- ADMINISTRAT!YE CONTROLS 86
.j.
EVSON 13
PART I LIST OF 1ABif} Idle 2-1 Radioactive Liquid Effluent Monitoring Instrumentation 22 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 23 -Radioactive Gaseous Effluent Monitoring Instrumentation 24 Radioactive Gassous Effluent Monitoring Instrumentation Surveillance Requirements 2-5 Radioactive Liquid Waste Sampling and Analysis Program 26 Radioactive Gaseous Waste Sampling and Analysis Program 2-7 Operational Radiological Environmental Monitoring Program 28 Reporting Levels for Radioactivity Concentrations in Environmental Samples 2-9 Maximum Values for the Lower Limits of Detection l l _ a
PART !! Jy f 0F TABLE 5 Table
. hit 1
RADI0 ACTIVE EFFLUENTS MONITOR SETP0INTS 2 RADWA51E REDUCTION SYSTEM-DOSE PROJECTIONS 29 GA5EDUS AND LIQUID EFFLUENT REPRESENTATIVE SAMPLING 39 IV CUMULATIVE DO5E CALCULATIONS 42 4.4 1 Dose Factors for (xposure to a Semi-Infinite Cloud of Noble Gases 54 4.4 2 inhalation Dose Factors Infant 56 4.4-3 Inhalation Dose Factors Child 57 4.4 4 inhalation Dose Factors Teen 58 4.4 5 Inhalation Dose Factors Adult 59 4.4 6 Ingestion Dose Factors, Grass Cow Hilk Infant 61 4 d, 7, Ingestion Dose Factors, Grass Cow Hilk Child 62 4.4 8 Ingestion DosJ Factors. Grass Cow Milk Teen 63 4.'4-9 Ingestion Dose factors, Grass Cow Milk-AdQlt 64 4.4 10 Ingestion Dose Factors. Grass Cow Meat Child 66 4,4 11 Ingestien Dose factors, Grass-Cow Heat Teen 67 4.4-12 Ingestion Dose Factors, Grass Cow 4 at Adult 68 4.4 13 Ingestion Dose i actors, Vegetation Child 70 4.4 14 Ingestion Dose Factors, Vogetation Teen 71 4.4 15 Ingestion Oose Factors Vegetation Adult 72 lEVLS!ON 13
LIST OF TABLES (Continued) M P_333 i 4.4-16 Dose Factors Ground Plane i i 74 4,4-17 Liquid Effluent Adult Ingestion Dose Factors 76 ] 311 Environmental Monitoring 5tation Location 80 i [
).1-2 Ring TLDs (Inner Ring) 81 S.1-3 Ring TLDs (3 Mlle Ring) 32
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PART I SPECIFICATIONS w
t 1.0 DEFINITIONS
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l.1 Channel Calibration A OW44EL CEIERATICH ahall be the adjustment, as necessary, of the marval output such that it resperds with neomemary range ard nomct:y to knen values of the parumstar 6 the channel amitars. the GWetEL CA125U23t shall anaampass the entire ciamai inchuiirg the senser and alamm an# cur trip functions, ard shall incluso the OpMEL FDCTs3EnL turf, ameaL CALIEHADtM any be pergamund by ag series of sepential, omelagplag er total channel steps sue that the entire mannal is cmMbruted. 1.2 Channel Check A CNNEL 0553 shall be the qualitative assessment of channel behavior drirq operstica by obserystian. This decemiration shall ircluse, Wants possMe, compariman of the chamal iniication arus/cr status with ot2mr irdications arWar status derivws tren irdependant ' instwmar,i chamals anamurirg the emer. paruuster + 1.3 Channel Functional Test
- 4. Ana;:g :nannels . :ne injec;i:n of a simu a:ac signa' :n:: :ne
- nannel as :1:sa :: :ne :r mary senser as : rte:':acle :: e er; 'y
;;U, aILI-"I inc!.cing alam anc/:r :ri: 'uncti:ns .
- 3. Bis.sola channels - the in.'ec:icn of a simu:a:ac s';ra H:
- ne enannel sanser :: ver'#y OP9)B:LITI in:Ncing Cam acci:r rd: 'unc;i:ns.
1,4 Frequency NOTATION FREQUENCY S At le' i'. once per 12 hours. D At least once per 24 hours. 5 At least once per 7 days. At least once per 31 days. At least once per 92 days. SA At least once per 6 months. R At least once per 18 months. S/U Prior to each reactor startup. l l P t t Completed prior to each release. , N.A. Not applicable, i I
i l 1.0 DEFINITIONS (CON'T) l.5 Independent Verification INDEPENDENT VERIFICATION is a separate act of confirming or substantiating that an activity or condition has been completed or implemented, in accordance with specified requirements, by an individual not associated with the original determination that the activity or condition was completed or implemanted in accordance with specified requirements. i 1.6 Liquid Radwaste Treatment System The I,IQUID RADWASTE TREATMENT SYSTEM shall be any available equipment (e.g. , filters, evaporators) capable of reducing the quantity dischasge.of radioactive material, in liquid effluents, prior to 1.7 Member of the Public
*EM8ER(S) 0F THE 6 dLIC shall include all individuals -no by virtue of thei occupational status have no formal association with the plant. This category shall include non emp byees of the Ilcenset who are permitted to use portions of the site for recreational, occupational, or other purposes not asseciated with plant functions, This category shall ng include non* employees such as vending machine servicemen or postmen who, as part of their normal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, 1.8 Mode REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION. K,ff THERMAL POWER
- TEMPERATURE
- 1. POWER OPERATION > 0.99 > 5% > 28m F
- 2. STARTUP > 0.99 < 5% 1 280* F
- 3. HOT STANDBY < 0.99 0 > 280*F
- 4. HOT SHUTDOWN < 0.99 0 280*F > T,yg > 200*F
- 5. COLD SHt/TDOWN < 0.99 0 < 200*F
- 6. REFUE!.ING** < 0.95 0 < 140'F Excluding decay heat.
Reactor vessel head unbolted or removed and fuel in the vessel. 1.9 Offsite Dose Calculation Manual (00CM) The 0FFSITE DOSE CALCULATION MANUAL contains the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM also contains the Radioactive Effluent Controls and Radiological Environmental Monitoring Program (T.S. 6.8.4), and descriptions of the information that should be included in the Annual Radiological Environmental 3perating ard Semi-annual Radioactive Effluent Release Reports (T.S. 6.9.1.Sc and 6.9.1.5d).
l 1 1.0 DEFINITIONS (CON'T) 1,10 Operable - Operability A system, subsystas, train component er device shall be OPERA 4Lt er have lapiteitOPtAASILITY when in tafs esfinitten f t isbe shall capa,ble of performing ita spect fied function the assvept . en that all necessary attendant Instrumentatten. centrols normal and emergency electrical power sovrees, cooling er seal water, Ivbrication er other auntliary equipment, that are required for the system, subsystes, train, component er device their related te perfpre support its function (s), are aise capable of performing functicals). 1.11 Site Boundary The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee. 1.12 Source Check A Sotntcy Carex shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactiva source. 1.13 Unrestricted Area An UNRESTRICTED AREA shall be any area at or beyond the site bounda access to which is not controlled by the licensee for purposes of protection of individuals from esposure to radiation and radioactive matarlais, or any area within the site boundary used for residential quarters er industrial, commercial, institutional, and/or recreational purposes. 1.14 Ventilation Exhaust Treatment System A VENTII.ATION EIRADST TREATMEFlT SYSTEN is any system designed and installed to reduce gaseous radiciodine or , radioactive material in particulats form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or EEPA filters for the purpose of removing iodines or particulates the environment from the gaseoes exhaust strema prior to release to effect on acele gas effluent.s).(such a- system is not considered to have Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be VENTILATICH EXIAUST TREATMENT SYSTEM components. 1.15 Waste Gas System piping)A WASTE CAS SYSTEM is any equipment (e.g., tanks, vassels/ capable of collecting primary ,:oclant system of fgases from the primary system and providing for delay or holdup for t!' purpose the environment. of reducing the total radioactivity prior to release to-i l 1.16 Purge - Purging i PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that I replacement air or gas is require _ to purify the confinement.
._ . - - = . . .
2.0 SPECIFICATIONS RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRCHENTATION l 21 The radioactive liquid effluent monitoring instrumentation channels shown in Table R-i shall be OPERABLE with their alara/ specification 2,5trip setpoints set to ensure that the limits of are not exceeded. The setpotnts shall be determined (ODCM). in accordance with the OFFSITE DOSM CALCULATION MANUAL APPLICABILITY: As shown on Table 1-/ ACTIo3:
- a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required above, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or change the setpoint so that-it is acceptably conservative, or declare the channel inoperable.
b. With.one or more radfoactive lioufd effluent monf tering instrumentation channels inoperable, take the ACTION shown in Table R-/. Exert best efforts to return ! thedays. 30 inoperable instrument (s) to OPERABLE status within If the af f ected instrument (s) cannot be returned to OPERABLE status within 30 days, provide information on the reasons for inoperability and lack of timely corrective action in the next Sealannual , Radioactive Effluent Release Report. ; I l i-SURVEILLANC2 RERER2NTS ; _ l 2././ Each radioactive 11guld effluent monitoring instrumentation channel shall be damonstrated OPERABLE by Performance of the CHANNEL CEECK, SOURCE CRECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in' Table 2-L s ,:
TABLE 2-/
. RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Mt31 MUM ,
INSTRUNENT CHM 4NELS APPLICABLE OPERABLE MODES ACTION 1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AU10MATIC TERNINATION OF RELEASE r <' a. Auxiliary Building Liquid Radwaste 1 Effluent Line (RM-L2) All MODES 21 (
- b. Secondary Drain Tank Liquid 1 '
Effluent Line (RM-L7) All MODES 22
- 2. FLOW RATE MEASUREMENT DEVICES
- a. Auxillary Building Liquid Radwaste f
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Effluent Line l All MODES 23
- b. Secondary Drain-Tank Liquid 1 Effluent Line All MODES 23 i
_-_7._._ Table 2-1 (cnntinund) 1 TABLE NOTATION _
. 1 ACTION 21 With less than the required number of OPERABLE channels effluent releases via this pathway may contfnue, provided that prior to initiating a release
- a. At least two independent samples are analyzed in accordance with specification P./. / , and
- b. Aa INDU ENDENT VERIFICATION of release rate calculations is perfarmed,-and ,
- c. An INDD ENDENT VERIFICATION of discharge valve lineup is performed.
Otherwise, suspend releases of radioactive materials via this pathway. ACT. ION 22 With less than the required number of OPERABLE chennels, effluent releases via this pathway may continue, provided that grah samples are collected and at least once per 8 analyzed hours, at anfor LLDgross of radioactivityi at least 10- microcuries/al, ' ACTION 23 With less than the required number of OPEJLABLE channels, offluent releases via this pathway may continue, provided that the flow rate is estimated at least once per 4 hours during actual releases,
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3 CHANNEL SOURCE CNAMISEL INSTRUMENT CNANNEL- FUBICTIOIIAL CHECK CllECR ' CALIBRATION TEST N, . 1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE
- n. Auni)lary nullding. Liquid D* r Redwaste Etfluent Line a(1) M (RM-L2)
.' b . Secondary Drain Tank De P Liquid.Etfluent Line- ag1) n- '
(RM-L7)
. 2. FLOef RATE MEASUREMENT DEVICES
.a. Auxiliary guilding'. Liquid D(2) M.A. A Raduaste Effluent Line L A.
- b. Secotdary Drain Tank Liquid D(2) N.A. R Etfluent.Line AA-p 4
- 4 - e
n TAB 22 NOTATION During periods of release. (1) C3ANNIL CALIBRATION shall be perfor:ned using av one or more standards traceable to the National Bureau of Standards, or
- b. Standards obtained from suppliers that participate in measurement assurance activities with the National Bureau of Standards, or
- c. Standards related to previous calibrations performed using (a) or (b) above.
(2) CHANNEL CHECX aball consist of verifying indication of flow during periods of release. A CHANNEL CHECK shall be perfor:ned at least once per day on any day that continuous, periodic or batch releases are made. t e P e __ , R 'p, . , -, ~ ~ ~
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i 4 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
;t . 2.
Table 2-3 , The radioactive gaseous effjuent monitoring nstrumentation charmels shown in set to ensure shall thatbe the OPERABLE limits ofwith tne cifluent trJease isolation alarmitrip setpoints Specification 27 art out exteeded. Tha
- setpoints Yl.NUAL (CDCM).
sa' all ba dete==ined is accordance vith tis oyTS1TI DOSI CALM,, !ON APPLICABILITY: At shown in Table J-3 A C"nON:
- a. ,
With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required above, without delay suspend the rejease of radioactive gaseous effluents mwtored by the affected enannel where appilcable, or enange the setpoint so that it it acceptably
- conservative, or ceclare the channel inoperable.
- b. ,
71th one or more radioactive gaseous effluent monitoring mstrumentation enannels inoperable, take the ACTIOf f .seown in Table 2- 3 Exert best effortsIfto retum the inoperable instru nent(s) to OPERABLE status eithin 30 days. the affected instrumenu cannot be returned to OoERABLE status withie JO days, provide mformation m reasons for inoperability and lack of timely corree.ive action in the next 'iffluent and Waste Disposal Semlannual Report. SURVEILLANCE REOUIREMENTS
- 2. 2. I.
Each radioactive gaseous effluent monitoring instnmentation enannel shall be demortstrated GPEO.ABLE by perf armance of the CMANNEL CHECK, SOURCE C CHANNEL MODES and frequenciesCALIBRATION shown inand TableCHANNEL 2- V. FUNCTIONAL TEST operations durm vo - _
t TABLE ~ 2- 3 2 RADIOACTIVE GASEOUS EFFLUENT A4ONITORING INSTRUMENTATION MINIMUM ! CHANNELS APPLICABLE - OPERABLE MCDES i
'l ACTION
- 1. Waste Gas Decay Tank Monitor (RM-All) '
a.- Noble Gas Activity Monitor * - 1 b.. Elliuent System Flow Rate Mon
- tor All MODES 24 I All MODES
' 26
- 2. Reactor Building Purge Exhaust Doct Monitor (RM-A i) f,,; a. ~ Noble Gas Actirity Monitor
- i. Operating Range = 1 **
ii. Mid Ranges. ' I ** 27 ; iii. liigh Range #- 29 I **
- b. lodine Sampler i **
29
- c. Particulate Sampler 2) i ** .
- d. Elliuent System Flow Rate Monitor 1- **
23
- e. Sampler Flow Rate Monitor I **
26 26 t
- 3. Auxiliary Building and Fuel 11andling '
Area Exhaust Duct Monitor (RM-A2)
- a. Noble Cas Activity Monitor
- 1. Operating Range
- I All MODES 28 ii. Mid Range # 1 1, 7, 3 & 4 lii. High Rangei 29 '
! I, 2, 3 & 4 29
- b. lodirre Sannpler "I
- c. Particulate Sampler All MODES 2)
I All MODES 25
.d. Elfluent System Flow Rate Monitor i 4
- e. Sampler Flow Rate Monitor All MODES 26 I All MODES 26
'
- Provides control room alarm and automatic termination of release. l
*
- During periods of reactor building purge. 1
'# There is no isolation setpoint or release termination function for this monitor. Alarm setpoints are determined by the appwpria e system procedures. .
a TABLE J-7 (Continued) . TABLE NOTATION t - ACTION 24 with less than the required number of OPERABLE cha to initiating a releaseWaste Gas Decay Tank may be released to the n , provided that prior 1. (RM A2)is OPERABLE with its setpoints Specification J. 7 Monitor set to e are not exceeded. re that the limits o' - 2. in accordance with the OFF5ITE DOSE CALCULATION , or MAN
- a. azordance At least with twoTable independent 2-6 and samples of the tanWs yzed in contents are b.performed, An INDEPENDENT and VERIFICATION culations ofis release ra
- c. An performed. -INDEPENDENT- YERIFICATION of discharge valve lineup is
., i Otherwise, suspend releases of radioactNe a way.effluents via this p th
),
ACTION 23affected via the Withpathway the number may continueof OPERABLE channels! less th n required, effluent releases I collected with an-Mey sampling equipmen,t as required in Table 3 ;' ACTION 26 Tith the number of OPERABLE channels less w th via this pathway may continue, provided flow east per 4 hours. rate is estimated a once ACTION 27 With _ the nurnbar of OPERABLE channels less t suspend PURCING of radioactive effluents via this pathway.an required, I 1
- -_- _- ------ . X. -
TABLE A -3 (Continued) . TABLE SOTA TION ,
~
ACTION 23 With the number of OPERABLE channels less than required
- hours and analyzed wimin 24 hours, and either th the contents of the taste Gas Decay Tanks.Part 2 are me ACTION 29 With the number of OPERA 8LE channels less than required b monitoririg the appropriate parameter (s), within 72 h 1)
Elther of restore the event, or the inoperable Channel (s) to OPERABLE status w 2) Prepare and submit a Special Report to the Commission pursuant to Specification J.d within the twrt 14 days out11 rung the acton taxen, causetoofOPERACLE system the inoperability status. and the plans and scr.edule for restormg tn NOTE: Action Sta tement 2.2.. && not applicable l l l
TABLE 2 */ RADIOACTIVE GASEO'U5' EFFLUENT MONITORING'INSTitUMENTATION SURVEILLANC"_ REQUIREMENTS CllANNEL MODES IN WillCil Cl!ANNEL SOURCE CHANNEL FUNCTIGNAL SURVEILLANCE
- INSTRUMEN T CllECK CflECK CALIBR ATION - -TEST REQUIRED !
- 1. WASTE CAS DECAY TANK I MONITOR (RM-All)
- a. Noble Gas Activity P P 11(1) M All MODES Monitor ,
- b. Efiluent System Flow P. N.A.. R M All MODES
, Rate Monitor r
. ,[
- 2. REACTOlt BUILDING PURGE EXilAUST DUCT MONITOlt (RM-AI)
- a. Noble Gas Activity Monitor ' -
- 1. Operating Range - D P R(I) M F 't li. Mid Range. W M R(I) M # ,
iii. High Range 'W M II(I) M i
- b. Iodine Sampler W N.A. N.A. N.A. # '
- c. Particulate Sampler W N.A. N.A. N.A. #
- d. Eliluent System Flow- D N.A. R M #
Rate Monitor
- e. Sampler Flow Rate D N.A. R M #
Monitor
- 3. AUXILIARY BUILDING & FUEL.-
flANDLING AREA EXilAUST DUCT MONITOR (RM-A2) .l a.- Noble Gas f.ctivity Monitor g
- i. Operating Range D N.A. Ril) M All MODES li. Mid Range W M R(1) M 1, 2, 3, 4 ,
lii. High Range W M ' R(I) . M 1, 2, 3, 4 l
- b. todine Sampler . W N.A. N.A. N.A. All MODES ,
- c. Particulate Sampler W N.A. N.A. N.A. All MODES
- d. E 11uent System Flow D N.A. R M Ali MODES ;
Rate Monitor !
- e. Sampler Flow Itate D N.A. R M All MODES Monitor i
e TABLE 2-4 (Conunue(s d During periods of Reactor Beijsmg Purge. (1) CNANNEL CALIBRATION shau be performed usingt
- a. One or more standards traceable to the National Bureau of Standarcis, or
' h. Standards obtamed from suppliers that partidpate in measurement assurance i act:ivities with the National Bureau of Standards, or
- c. Standards related to previous callbrations using (a) or (b) above.
1 i A I f i i 5 i i
?
i A h . v.
LIQUID RADWASTE TREA'DU!NT SYSTEM 2.3 The LIQUID RADWASTE TREATMENT SYSTEM shall bs used, as required, to reduce radioactive materials in liquid wastes prior to their discharge, when projected monthly doses due to licuid effluents discharged to UNRESTRICTED AREAS *
.tould exceed the following values:
- a. 0.06 area whole bodyr
- b. 0.2 area to any organ.
APPLICABILITY: At all times. ACTION: -a. When radioactive liquid waf.te, in excess of the I above limits, is discharged without prior treatment, prepare and submit to the Commission j within 30 days, a Special Report pursuant to Specification J./V information: , which includes the following i
- 1. Identification of inoperable equipment and the reasons for inoperability.
- 2. Actions taken to restore the inoperable equipment to CPERABLE status.
- 3. Actions taken to prevent recurrence.
'*4*' .
l I LICUID RADWASTT TREATMINT SYSTIM (Continued). I SERVEILLANCE REOUIRZy.EN'"5 4- -
-2.J./
Doses due to liquid re least once per 31 days, 2ases shall be projected at CE CULATION MANUAL (ODCM). in accordance with the CFFSITE DOSE O e 8 L l l
, +s-. -_ _ . . . _ - - , . . . , . . . _. . --- - . . ,_
s WA&TE GAS-SYSTDj 2.Y The WASTE GAS SYST M oball be used as required, to reduce the radioactivity of materials in gaseo,us waste prior to dischargd, when projected monthly air doses due to releases of : gaseous effluents from the site to areas at or beyond the SITE BOUNDARY would exceed:
- 1) 0.2 mrad gammar
- 2) 0.4 mrad betas and The VENTILATION EERAUST TREAT? GENT SYSTEM shall be used, as required, to reduce the quantity of radioactive materials in ,
gaseous waste prior to discharge when projected monthly air ' dose:s due to release of gaseous e,ffluents from the site to areas at or beyond the SITE BOUNDARY would exceed: {
- 1) 0.3 area to any organ.
APPLICA5ILITY: At all times. ! ACTICM
- a. When the WASTE GAS i SYSTEM and/or VENTILATION ,
EXEADST TREAMENT SYSTEM are not used and gaseous waste in excess of the above limits is discharged without prior treatment, prepare and submit to the commission, within 30 days a Special Report, pursuant to Specification .? . R which includes j l
- 1) Identification of the inoperable equipment and I i
the reason (s) for inoperability. l 1
- 2) Actions taken.to restore the inoperable i equipment to OPERABLE status.
- 3) Actions taken to prevent recurrence.
l l I
._ n - . . . - - -
PLANT SYSTDt3 l StTRVIILLANCE REQUIR2MENTS
- 2.4./-
projected at least once per 31 days, is accordance with the-Dose ' QFFSITE DOSE CALCULATION MANUAL (ODCN) . t 9 . - - , ,$$w-)
, _ ,- - , - _ . ., , . , - y ,.
. - - - . . - . - - . -.. - . - -- -. . . - . ~ . -. .
I lo lIOUID rrrLUENTS l CONCINTFA"'IOJ
- l 4.I ' The concentration of radioactive material released to UNRESTRICTED AREAS shall be less than or equal to the concentrations specified in 10 CTR Part 20, Appendix B, l' Table II, Column 2 for radionuclides otner than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be less than or equal to 2x10-4 microcuries/ml. total activity. 1 APPLICABILI"T At all times.
AC" ION:
- a. With the concentration of radioactive materials released to UNRESTRICTED AREAS exceeding the above limits, without delay restore the concentration of radioactive materials being released to UNRESTRICTID AREAS to within the above limits. If the concentration of radioactive materials being released in excess of the above limits. is related to a plant operating characteristic, appropriate corrective measures (e.g. , power reduction, plant aburdown) shall be taken to restore the concentration of radioactive materials being released to UNRESTRICTED AREAS to within the above 2.imits.
SURVEII.I.ANCE REQUIREMEN"5 2.J./ Radioactive liquid wastes shall be sampled and analyzed in accordance with the sampljng and analysis program of Table 2-5. 2 . 5. 2 The results of the radioactivity analyses shall be used in accordaned with the methods in the OFFSITE DCSE CALCULAT70N MANUAL (CDCM) to assure the concentrathns of radioactive material released from the site are stintained within the lini::s of Specification 2. 5 . t l I
\
TAELt_ 2-5
~
RAD IC A0*!vT LI C C ID WAS*E S A.$0 L I NG AC AN Al? S I S P R OC P.A.M d Lower Limit Minisum
- of Liquid Release sampling Analysis Type of Activity Detection Type -
Frequency (n;3) Frequency Analysis (VCi/ml)" ! 4 A. natch Waste P P Releage tach tatch Zach tatch Principa 5x10'7 Tanks taitters} Gamma
- 1. Evaporator Condensate I-131 storege .
1x10-6 Tanks (2) P M
- 2. Laundry & ' Dissolved and 1x10-5 One Batch /M Entrained Cases Shower Sump Tants (2) (Ga=ma 2: sit te r s )
4
- 3. Seconda ry P M E-3 1x10-5 Drain Tank Each Batch Composite 3 Gross Alpha 1x10~7 7 Q 'St-69, St-90 5x10~I Each Batch Ccaposite b Fe-55 1x10 '
~
- 3. Continuogs W Principal ca==4 5x10 ~7 Releases Continuous' Composite C Emitters *
- 1. Secondary Orain Tank I-131 ;x;0 -6 1x10 -5 M M Gr ab Sample Dissolved and Entrained Gases (Gamma E=1tters)
M H-3 1x10 -5 Continuous
- Composite C Gross Alpha 1x10'I
' Q. St-89, Sr-90 5x10"I Continuous' Composite C Pe-55 1x10 ~0 i
*?.
TA3LE J-T (CRntinundl TABLE NOTATION
- a. The LLD* is the smallest concentration of radioactive material In a sample that will be detected with 954 probability with 54 probability of falsely concluding that a blank observation represents a 'real' signal.
For a particular measurement system (which may include radiochemical separation): 4.66s b LLD = (E) (V) (2.22x10') (Y) (e -^4 5) Where LLD is the per
-microcurie lower unitlimitmass of or detection volume),as defined above (as sb is the standard deviation of the background criunting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as ctmnts per disintegration), V is the sampie size (in units of mass or volume), 2.22:10 6 is the number of disintegrations per minute per alcrocurie,
- Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclide,-and at is the elapsed time between aidpoint of sample collection ; and time of counting (for plant effluents, not environmental ' samples). Typical values of E, V, Y, and at shall be used in the calculation.
- The LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (af ter the fact) limit for a particulg,r measurement.
1 4 e,U ____m .- __. _ _ _ _ , , . . , . . .
-. .- - - . . . . - - _ . - . .- . ~ _ - - .. -
, 1 TABLE J-I (Continued)
TABLE NOTATION, b. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. ,
- c. To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent streaa.- Prior to analyses, all samples,,taken for the composite shall be thorzugMy atxed is order for ths ,
composite release. sample to be representative of the effluent
- 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 repress,ntative sampling.
e. A continuous nondiscrete release is the discharge of liquid wastes of a volume; e.g., from a volume or syncan that has an input flow during the continucur release, f. The principal gamma emitters for- which the LLD specification applies esclusively are the following radionuclides: Mn-54, Pe-59, co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses shall be reported ' as 'less than' the nuclide's LLD, and shall not be reported as being present at the LLD level for that nuclide. The
'less than' . values shall not be used in the required dose calculations.
1
-lDe '. , , - -
. - .- . . - . _ - - . . ~ _- - .- - .-- - . - .- .. - . . LIQUID EFFLtfENTS - DOSE
- 2. 6 Tho dose,or dose commitment to a MEMBER OF TER PUBLIC from radioactive-materials in liquid effluents released to tmlutSTR.TCTED ARRAS shall be limited as follows:
- a. During any calendar quarter to less than or equal to 1.5 area to the total body and less than or equal to 5 area to any organ,
- b. During any calendar year to less than or equal to 3 area to the total body and to less than or equal to 10 area to any organ.
APPLICABILITIY: At all times. ACTIONt
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above lir.its, prepare and submit to the l commission, to Spccification within 30 days, a special Report pursuant 2.N , which includes:
- 1. Identification of the cause for exceeding the lini.t (s) :
- 2. Corrective action taken to reduce the release of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the remainder of the current i calendar year so that the dose or dose !
comaltaent to a MEMBER OF THE PUBLIC from this source is less than or equal to 3 area total body and less then or equal to 10 area to any organ during the calendar year.
_ . _ . - . - _ . . . . - . ..-.. - - - - - -. . ~. - - - - - l P.ADICACTIVE EFTLtmtTS StT.T11 LANCE REQUIREMENTS 2,4./ DOSE CALCtrLATIONS. Cumulative dode contributions frca liquid effluents shall be determined in accordance with the OFFSITE DOSE-CALCULATION MANUAL (CDCM) at least once per 31 days, d e e
.... ? -
GAsb005 EFFLUENTS - DOME RATE J.7 The-dose rate at or beyond the SITE SOUNDARY, due to radioactive materials celeased ,e gaseous effluents shall be limited as foi ..nts:
- a. Wohle gases: less than or equal to 500 aren/ year total the skin. body and less than or equal to 3000 aren/ year to
- b. Iodine-131, Tritium, and radioactive particulates with half-lives of greater than 8 -days: less than or equal to 1500 mram/ year to any organ.
APPLICABILITYt At all times ACTION:
- a. With dose rate (s) exceeding the above limits, without delay decrease the dose rate to within the above limita(s). .If the dose rate et or beyond the SITE BOUNDARY due to radioactive,3pterials in gaseous effluents in escess of the above limits is related to a plant operating characteristic, appropriate corrective measures (e.g., power reduction, plant 1 shutdown) within shall belimits.
the above taken to decrease the dose rate to l l l
)
l SURVEILLANCE REQUIREMENTS I A. 7. / The dose rate due to noble gases in gaseous
.etfluents shall be dettu. mined to be withia the above limits in accordance with the not aods and procedures of the OFFSITE DOSE 1 l CALCULATION MANUAL (CDCM).
2 . 1 The dose rate due to radioactive materials specified' I above, other than noble gases, in gaseous effluents shall be determined - to be 91 thin the above limits in accordance with the OFFSITE DOSE CALC','LATION MANUAL (ODCM) by obtaining i
;Table representative J-6, suples and performing analyses in accordance with 1
a.:- '
TABLE 2-4 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM l Minimum Gaseous Release Type Sampling Analysia Type of Lower Limit of Frequency , Frequency Activity Analysia - Detection jLLD) (pCl/ml) P P A. Waste Gas Decay Each Tank Each Tank Tank Grab Principal Gamma EmittersI lx10 -4 t Sample B. Reactor Building Each Purge c Each Purge Purge Exhaust Grab Prinolpal Gamma ygittersb,C lx10 ~4
' Duct Honitor Sample (RM-A1) H-3 -
1x10
- C. Auxillary Building U M
and Fuel Handling Grab M Principal Gasuna Emitterab,f lx10'4 Area Exhaust Duct Sample M-3 lx10
- Honitor (RM-A2)
D. All-Release Types Continuous
- W ~
as Listed in A, B, Charcoal C above I-131 lx10
-12 Sample Continuous
- W Principal Gausta r.miU. rs' lul0 -Al Particulate (I-131, others)
Sample , Continuous
- M Gross Alpha Composite 1410 -11 Particulate 8emple Continuous
- O Sr-89, ar-90 Composite lx10 -##
Particulate Sample Continuous Noble Gas Noble Gases lx10 *
~
Monitor Gross Beta & Gamma
1 TABLE O-4 (Continued) TABLE le0TATI0tt a. The LLD*inisathe material sample smallest that concentration will be detected of radioactive with 954 probability with 54 probability of faisely concluding that a blank' observation represents a "real' signal. For a particular measurement systes (which may include raJiochemical separation): W-4.66s b 4 LLD = i' (E) (V) (2.22x10') (Y)' ( e'^ 4 t ) . where: o LLD is the lower unitlimit of or detection as defined above (as 1 alcrocurle.per mass volume), i-sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
- I-E is the counting efficiency (as cgints per disintegration),
- V is-the sample size (in units of mass or volume), ~
f 2.22x10 6 is tho' number of disintegrations per minute }..e , microcurie, Y is the fractional radiochemical-yield (when applicable), A is the radioactive eadionuclide, and decay constant for the particular at , is the elapsed time between aidpoint of sample collection and time ,f counting (for plant effluents,- not i environmental samples). Typical values of E, V, Y, and 4 t shall be used in the
- calculation.
i
- The LLD is defined.as an a priori (before the fact) limit ~
1 representing a posteriori the capability of a measurement system and not as an (after the fact) limit for a particular measurement, f 4
~
a _ a . .
. _ , . _ _ r i _ - . ._ .. _ _ _ _ . . . _ _ . _ , _ . _ _ _ . _ . - . _ _ - _ . - . . . . . -
TABLE A6 (continued 1 Ig.t.: noT m on
- b. Analyses shall also be perforsed betwee$ 2 and 4 hours following-shutdown, startup or a change in power excee(ing 15% A? /ED TEEMAL POWEA within ons ac c. level
- c. Tritium grab samplet shall be tan' en between 12 and 24 hours af ter flooding the refueling canal and at. least once days thereaf ter while the refueling canal is flooded. par 7 samples shall be changed at least once per 7 days and analyses shall be from (or af ter removal completed sampler)within 48 hours af ter changing
. sampling and analyses shall be performed at least once per 24 hours for at least 7 days following each shutdown, startup or change in power level exceeding 15% of RATED TIERNAL POWER within one hour, unless the Iodine Monitoring Channels in Radiation Monitors AM-Al aad RM-A2 show that tha Radiciodine concentration in the Ausiliary Building and Fuel Bandling Area or the Reactor Building Purge Exhaust Ducts will lead to e release which 12 less than lot of the 10 CFR 20, Appendia 3, Table II, Column I limits, at or beyond tba JiTE SQCNDARY. -
- e. The ratio of the sample flow rate to the sampled stream flow '
rate shall be known for the tima period covered by each dose or does rate calculation made in accordance with the Specifications 3.11. 2.1, 3.11. 2. 2 and 3.1. 2. 3. f. The principal samma emitters for which the LLD specification applies esclusivaly are the following radionuclides: Kr-47, 4 Kr-88, Ie-133, %e-133m,1e 135, and Xe-138 for gaseous waissions and Ma-54, Fe-59, Co-58, Co-60, 2n-45, Mo-99, Cs-124, Cs-137, co-141 and Co-144 for particulate esissions. S This list does not mean that only these nuclides are to be detected and reported. Other psaks which are esasurable { i and identifiable, together with the above nue1 > des, shall also be toontified and reported. Wuclides whuh are below the LLD for the analyses shall be reported as 61ess than' the nucUde's LLD and shall not be reported as being present at the LLD level for thC4 nuclide. The *1ess than* values shall r.ot to used in the required dose calculations.
s DQ85-WQ9fp3 GA#B8
- 2. 8 The~ air dose at or beyond the sITI SOUNDAM
, due to radioactive noble gases released in gaseous effluents shall be limited tot
- a. During any calendar quarter: less than or equal to 5 mrad gamma and less than or equal to 10 arad beta radiation, and 4
- b. During any calendar years less than or equal to 10 mrad gamma and less than or equel to 20 mrad beta radiation.
APPLICA3ILITY: At all times. ACTION: a. With the calculated air dose from radioactive noble lasesingaseouseffluentsexceedinganyofthsabove 1mits, ! 30 days, prepare a special and submitpersuant Report, to the Commission, within to specification J.N , which it:cludes: . 4 . 1) j Identification limit (s), of the cause for exceeding the . i 2) Corrective radioactive action taken to noble gases in reduce gaseousthe release of effluents during the remainder of the current calendar quarter and during the remainder of the current calendar year so that the average dose during the
- calendar year is less than or equal to 10 mrad l gamma and 20 mrad beta radiation.
I ( 5URVEIT iMC2 REQUTREKENTS R. 5. I DOSE CALCULATIONS
' the current calendar quarter and Cu.lulative dose conteibutions current calendar year shall be for determined (ODCM) at leastin accordance once per 31with the OFFSITE DOSE CALCUI.ATION MANUAL days, i
,,s.
DOSE - 10 DINE-131, TRITIt34, AND RADICACTIVE PAATICUI.ATTS
-4
.J.9 The dose to a MFXtF.R OF TEE PUBLIC from Iodine-131 Tritium, and radioactive particulates with half-lives g at or beyond the SITE 500MDARY limited as follows: shall be ,
- a. During any calendar quarter:
7.5 ares to any organ, and less than or equal to
- b. During any calendar year area to any organ. less than or equal to 15 APPLICABILITY: At all times.
ACTION: a. With the calculated 131, Tritium dose from t.he release of Icdine-e and radioactive greater than a day half-lives,particulates with in gaseous effinents, ;- exceeding any of the above limits, prepare and submit ' to the commist pursuant ion, within 30 to specification days, a Special Report,
- 2. N , which includes: 1 1)
Identification limi ts (s); of the cause for exceeding the 2) corrective action to reduce those releases during the remainder of the current calandar quarter and I the remainder of the current calander year so that or equalthe average to 15 area. dose to any organ is less than i l i i _ __ ,. --- ~
Dost I tw3 131. TRITIUM, AND RADroActivT PARTI N Trs SURvTILLANCE REQUIREMrwTs 4.9./ 00$E CALCULATICHS: Cuinulative dose calculations for ; the current calendar quarter and current calendar year shall be determined in accordance with the oFFSITE DOSE CALCULATION MANUAL (oocM) at least once per 31 days. i o o I 1
' " ' ~ -t r-vr--,-- , - . . m., , ,__
l
. N DOSE l
J./d OF Tux Pr::mL2c, be calendar due to year dose or dose cossaitment to any KEMBER releases of radioactivity and radiation, from uras.ium fuel cycle sources shall be limited to less than or equal to 25 areas to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 arens). APPI,ICAs u Y: At all times. ACTIOWs - a. with the calculated doses from the release of radi.sactive materials in liquid or gaseous ef fluents exceeding
- 2. ca b twice the limits of specification 2.4.
,, 29a , 2 86 , 2. s . , or J. 7 4 , calculations should be made, which include direct radiation contributions from the reactor, to determine whether the above limits of Specification J./0 have been escoeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to specification ,
2./V , a special Report that defises the corrective action to be taken to reduce i,
- subsequent releases to prevent recurrence of exceeding the above limits arid includes the schedule for achierving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405c, shall include an analysis that estimates the radiation em p sre (dose) to a MEMBER OF ME PUBLIC from uranium fuel cycle sources, including all of fluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also desc:ibe levels of radiation and concentrations of radioactive material involved, and the cause of the e p ere levels or concentrations. If the estimated dose :s) exceeds the above limits, and if the release edition resulting in violation of 40 CFR Part 190 has not already been corrected, the special Report shall 1 w-9 Me a request for a variance in accordance with the prcnrisions of 40 CTR Part 190. Submittal of the repoet is considered a timely request, and a variance is gra=:ed until staf f action on the request is complete, i
l
1 1974 0083 (continued) SURVEILLAMet RrcOIREMENTS *
- 2. /0. / D0$f CALCULATIONS 11guld and gaseous effluents - Cumulative shall be dose contributions determined from in accordance with Specifications 2.6./ 2 8./
, , and 2.9./ , and in accordance with the QFFSITE DOSE CALCULATION MANUE (CDC , , . . . . .. ,, - , - - - - - ~ ~ - - " ' ' ~ ' ' ' " ' " " ~ ~ ' ~ ' " " ~
/ RADTDLOGICAL ENVIRONMENhAL MONITORING
- 2. H be conducted ' The as radiological specified environmental in Table .27 monitoring program shall APPLICA3!LITY: At all times.
ACTION: 4. With the radiological environmental monitoring program not being conducted as specified in Table J7 , prepara and submit to the Commission, in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a eocur rence.
- b. With the level of radioactivity, resulting free plant offluents, in an environmental sampling medium escoeding the reporting levels of Table Js v '
/
averaged over any calendar quarter, prepare and A.Ait to the Commission, within 30 days of obtaining analytical results from the affected sampling period, a Special Report pursuant to specification a.N , i which identifies the cause(s) for escoeding the limit (s) and defines corrective actions to be taken to reduce radioactive effluents so that the potential annual calendar dose to a limits year MEMBER OF TRE PUBLIC is less than the of specifications a,7
.18 and J.1 .
When more than one of' Y$e radionuclides in Table 4-d are detected in the sampling medium, this report shall be submitted if e concentration (1) concentration (2) '
+ + ... > 1.0 limit level (1) limit level (2) 4 When radionuclides other than those in Table #-e are detected and are the result of plant effluents, l; this report shall be submitted if the potential annual ,;
dose to a MEMBER OF THE PUBLIC is greater than or equal to the calendar year limits of specifications l 2.7 , .t.# and 1. 4 . This report is not ; required if the measured level of radioactivity was : not the result of plant effluentar however, in such l an event, the condition shall be reprted and described in the Annual Radiological Environmental e Operating Report.
- i
ACTION (Con t inu ed ).
- c. With alik hr fresh leafy vegetable samples unavailable trou one Table J 7 or more of the sample locations required by of samples and idsntify locations for obtainingidentify the replacement sampisa in the next Annual Radiological Environmental Operating Report. The locations from which samples were unavailable may then be deleted from those required by Table A 7 , provided the locations from which t.be replacement samples were obtained are added to the environmental monitoring program as replacement locations. ,
$URVE!LLANCE REQUIREMENTS l 1
-' * /' * /
be collected pursuant to TableThe 2-7 radiological environmental monitoring sampl I from the locationa given in the (CDCM) table and figure (s) in the OFFSITE DOSE CALCULATION MANUAL ; Tables and .2 1 shallandbe 2 -analyzed pursuant to the requirements of , 9 .
. . , , , ~~ __ .__ _ _ --- . - - - - . - .-- --
~
TABLE 2-7 .. 08"I'R ATIONAL rat 100 LOGICAL ENVIRONGENTAL MONITMMC PROGRAM Espense P;O.__, NW et ?_ /; Mer Sasuple and Locations - i _- f" ,1 eMies : ,, j TMw
.g ,=_ ,;,, ,
! I. AIRBORNt* Orie sample each: Caritirivinas sampler / Radiosodine and COT, CIf, C40, C41, R M can&sters ' Weekly collection : particulates CM, and Castrol a) 1-131 analysis weekly Location C47 Partloslate L .! .s 4 a) Grossp at E 24 hours / followang weekIy fitace i dnange. { b) Conipesite gamma ptral
- analysis (by locatier.y .
gearterly. (Gamma Spectral l Analysis shall also be j perforsened en individael 2
- gis if groes beta activity of any semp4e is greeter than I.0 p C1/ml i and =tddiis also greater than l' ten times the control sample activity.) .
- 2. DIRECT RADIATION 8) Site Soundary: Contimsous Genima espesure rate /gaarserfy C60,C6t,C62,C63, placement / Quarterly
] C64,C63,C66,C67, co lection j C65, C69, C4 t, C70,
; C27.C73,C72,C73
- 2) Five Miles
- j. Cl8,CO),C04 C74, C73,C76,COS,C77 C09.C78,C16G COI,
- C79 4
)) C<wstro8 t.ocatient
- C47 .j i
_ _ . - _ , - --. - --- - - - -- -' - ' l
l [, J l I i [ 4 TARLE l-7 (Centinuedt i j I e
- OPERATMW4AL RAD 00 LOGICAL ENVIRONMP'NTAL MONETONING PROGRAM i
) ~ i - , O E ! ' Enyeeure Pe2_, i i andler Sessyle Nweeer of * : . "_ - E fig TypeJPregsmacy i and Lecaelens CeEsct6en Fre- ==cy f of."- ',.: t i 3
- 3. WATER 80RNE . One sample cactu !
! Seawater C14H, CI 4G Grab sample /Monehly C: 72 spectrat Canarel Location C13 ! t enelysis/ monthly !
- i. [
Tritiwa analysis en each sample ! er en a quarterly composite ! et smanthly : 455
- I s
Ground water One samples ! r C40 (Control Grab sample / l Locat6an) C:7 --2 spectral and j sem& annual I Trithsn analysis /cach sample ! Drinking water One samp% each:
- g COF,CiG,L13 Grab sampief9aarterly (All Control C
- - _ specarat and Tritlum ;
Locathms) analysis /each sample [ ,i i Shoreline sediment One sample each: .. r l ClWI,C16M,Cl4G Sem8 annual sample l l Centrol Locatlan C r12 spectral analysis /cach semple { CM !' . ~ .* ;
- 4. 3NGESTSON 4 *~
' i Fish & Invertebrates One sample each: Querterly: I 1
C29, Conteel Location Gamma socctral analysis on l Oysters and carnivorous edeble portions /each sample
- CM fedi (
a . .- ! r ; b t I
- i. I L
, , . . , , , . _ . . _,-m. , . _ . . - , . - . ~ . - . . , , , _ . , , . - _ . _
s i j v i g TABLE J-7 (Continued OPER ATIONAL R A000 LOG 9 CAL ENVIRONMENTAL ESONfTORING PROGRAM . Exposure Patieway Monber of 5 - ," - *1 , -j
*" t TypeNregsency i andler $1- ,'_ and Lecaelens r n=cties Freepsency of Aamiysis ,
4 Food Products - One sample each: Monthly (when available): Genome spectral and 1-131 C484* C48b*, Sarnple comprised of analys.s/each sample Control Location C47 three (3) types of broad ; leat vegetation from eadi Iccation , [ One sample: Annual during harvest Gamma spectral analysis / I Cl9 Citrus each sample i One sample Annual during harvest: Camnia spectral analysis / ' C04 Watermelon each sample 4 i i i 1 i i i '
- Stations C48a and C48b are located at or beyond the 4400 f t. site boundary for gaseous etfluents in the two sectors whicti yleid the highest historical annu.I average D/Q values. !
i i
- ' e i
1 t t _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . . . . ~_. .
TART.E 2-# REPORTING l_EVER_S Post RAIMOACTIVITY CONCENTRATBONS IN ENVIItON 4 , Water Airborne Particulate l *
. Q h' (pCl/t) Fish tomt
- er s'- (pC1/meD) Feed Products ' g (yC1/Kg, wet) (pCill) (yCE/Kg, wet)
H-3 20h* aan-14 n,o00 20.000 Fe-39 400 10,000 i Co-n t i,000 i 30,o00 , Co40 300 ! 10,000 In4) 300 20,000 t Zr-Nb-93(b) 400' ; 1-13I C 2 0.9 ) 100 Cs-114 30 lo 1,000 60 t,000 - t Cs-127 30 20 2,000 70 2,000 ' l Be-l.a-I 40(b) 200 ! 300 i i l
+
I F 1 (a) For drinking a value waterpCift of 30,000 samples. may heTliis is 40 CFR Part t41 value. Il no drinking water pathway esists, esw.t.
. l
-(b)
An the ergullibeluen pJf Citt $$dtOpC.enlutiare el the p4 ent .anil d,sughter Isotope wiekh contains (fie reporting value of ; I 2 (.-) : For iltinkang wa eer s.sunples *=aly. l i ! i i 4
TAate 2-9 I asAxsasuna VAttF.5 POIt TE LOWER LIAAITS OF DETECTIOpi(LLt3 aJ waeer Airborne Particulate Phh Asilk Feed Freescas Sedsment Analysis (pCl/I) er Games (pC1/mh (pCi/Kg, wet) (yCl/I) (yCl/Kg, Q) (pCI/KS, dry) gross beta . 0.01
'3 H'
34 g. 15 130 l[. 30 260 397 , . 38 g I3 (30 60g I3 130 63 30 260 h C 1 93 3, IS 335, II 0.08 g 60 , I34 g i3 0.03' 130 13 60 t30 137 g IS OA* I30 18 30 180 I403,_g , 13
# I)#
,_-..-.,_.m.---_-.-a6 - -s.-6... . . . . -.. . .
- -- - - - ~ - -
6 TABLE 2-9 (Cetinued) TABLP NOTAT10N '
~ .
a. be detelted with 93% probabdity with 3%The LLD* is the smallest con
- blank observation represents a "real" signal. probability of falseJy concJudAng that a For a particular meuurement system (which may include radlochemical separatio ,
LLDe (E) (V) (2.22) (Y) (e ' ' ' ' ) ! There LLD 1 mass or the lower limit of detection u defined above las picoeurie per unit volume), 1 sb is the standard deviation of the background counting rate or of the counting rate of a blank ample u appropriate (u counts per minute). : E is the counting efficjency (u counts per disintegration), V la the umple size (in units of maas or volume), i i 2.22 la the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yleid (when appilcable). A is the radioactive decay constant for the particular radionuclide, and at is the elapsed time between sample collection (or end of the samole collection efiluent perod) and tame of counting (for environmental umples, not plant umples). Typcal values of E, Y. Y and4t, should be used in the calculation.
' The LLD is defined as arig gjg,i (before the fact) limit representing the capabdity '
measurement measurement. Analyses system shall and not be as an a posteriori (after the fact) 1Amit for a particular acNeved under routirw eenditions.performee in such a manner that the stated LLD's wdl be Occuser. ally, background fjuctuations, unavoidatile smalj sample sizes, the may render these LLD's unschievable. presence of interferring nuclides, or other uncontrollable circumstance identified and described in the Annual Radiological Environmental Op 99
,-..n- k- n,- -
. n,5 % .? a
~
a w.,- - ,.--, --cm. --- ,-n,nn-,,,----,,n,m,-m--- ,, w,,,-,n,- . ,m r .m-
TABLg 2 T (continued) } TA8LE NOTAT10N
~
b. LLD be used, for dr1nking water. If no drbking water pathway exits, a value el 3000 pCill m c. The specified LLD W for an equalbrium mixture of parent and daughter nucudes which contain,13 pel/l of the parent nuclide. d. Other peaks which are measurable and identifiable, together with the radionucudes in Table 4.121, shall be identitled and reported. e-C6134, and analysia, not to C4137 LLD's analyses apply only to the quarterly composite gamma spectral of sint' cyticulate futers. f. LLD foranaj lootopic drinking water. If no ysis may be used. armking water pathway exists, the LLD of gamma
- g. LLD for 1131 appiles to a single weekly filter.
I s
}
r 4 l l l k 99 9 e m , ., 3.y,y-rw -y1--vtv-' ' - - ' - * * - " ' ' ' ' ' ' ' ' '
RADIOt,00fcAL ENv!RoNH2:NTAL MOHITORIMO LAND USS CIWSUS
. s
.t.c A-land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence and the nearest garden
- of greater than 500 square f eet producing fresh lesfy vegetables in each of the land based meteorological sectors within a distance of five alles.
A7pt,1 CAST LITY: At all times. ACTION:
- a. with a land use census identifying a lecation(s) that yields a calculated dose or dose commitment greater than the values currently be!.ng calculated by s pecification J . 9. / , mdentify the new location in t se nest Annual Radiological Environmental operating Report.
- b. With a land use eensus identifying a location (s) which yields a calculated dose or dor,e corisitzent (via the same esposure pathway) whica is at least 20% greater than at a location from which samples are currently 1 being obtained in accordance with specification 3.12.1.1, this location shall be added to the radiological environmental monitoring program within 30 days. Tne new sampling location shall replace the present sampling location, which has the lower calculated dose or dose commitment (via the same esposure pathway), af ter June 30 following this land use census. Identification of the new location and revisions of the appropriate figures from the CFFSITE DOSI CAJ4tn ATION MANUAL (CDCM) shall be submitted with the nest Semiannual Radioactive Ef fluent Release Report.
- Broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census.
RADM)LOGCAL ENYtRONMENTAL MONITORING LAND USE CENSU$ (Centinued) , SURVEILLANCE REQUIREMENTS _
~
- 3. /J. /
The land use census shall be conducted at least once per 12 months during the growing ieann by a door to-door survey, terial surveyr or by consulting local agriculture authorities, using that information which will provide acequate results. (
.;,.t'*
RADIOLOGICAL ENVIRONMENTAL MdMITORING ' INTERLA50RATORY_ CCMPARISON PROG 1 TAM ' ' J. U Analyses aball be performed on radioactive materials supplied as part of an Interlaborator has been approved by the Commission. y Comparison A summary of Program the results which obtained from this program shall be included in the Annual Radiological Environmental.
- Operating Report.
.- -t ... .
APPLICABILITY: At all times. ACTION:
- a. with analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.
SURVEILLANCE REQUIREMENTS
- 2. h . / !
- No surveillance requirements other than those required
- by the Interlaboratory Comparison Program, i
l
-. . - . . - - , _ - . . , . . - . ~ . - . . - . - . . . . - . _ - . . - - , . .-- - ,.
ADMINISTRAf!VE CONTROLS J,N SPECIAL REPCRTS Special reports shall be submitted to the Director of the Office of Inspection and Enforcement, Region !!, within the time period specified for each report. These reports shall be submitted covering the activities identified below. A separate Licensee Event Report, when rrquired by 10 CFR 50.73 (. , need not be submitted if the Special Report meets the requiremen)ts of 10 CFR 50.73 (b) in addition to the Myuircaents of the applicable refereaced Specification. A Dose due to radioactive materials in liquid effluents in excess l of specified limits, Specification 26. J. Dose due to noble- gas in gaseous effluents in excess of specified limits, Specification J. S . C. Total calculated dose due to release of radioactive etfluents exceeding twice the limits of specifications J . s a-J. s h 2.&n p, 8 6
, . , , J. 9 a , or .t . 9 6 (required by Specification 2./0 ).
D. Dose due to Zodine-131, Tritium, and radioactive particulates with greater than eight day half-lives, in gaseous etfluents in excess of specified limits Specification a.9. K. Failure to process liquid radwaste, in excess of limits, prior to release, Specification 2. # . F. Tailure to process gaseous radwaste, in excess gf limits, prior 3 to release, Specification 2. V. 6, Measured levels of radioactivity in environmental sampling medium in excess of the. reporting levels of Table 2-p , when avetaged over any quarterly sampling period. Specification
- 2. //.
h Inoperable Mid or Nigh Range Noble Gas Effluent Monitoring Instrumentation, Speci.'ication 22. i y r - u n-- m __. ,- ,, - r - - . y _--y --_4,,-w--w --...-e-gw-4 ,,..w_.,r.m.--w.w_-a ,-y - - - , . ~-w-----
3.0 SPECIFICATION BASES 3.1 RADIDACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION BAS!$ , the radioactive liwid offluent instrumentation is pewided to acnitor ard
- oantzul, as esplicable, the ruleenen of radioactive antarials in liwid effluents ering actual or potential rsleases of lipid affluents. the alaru/
nooonsance trip setpointe for these with the pre instnaments shall be calmtletat in me in the cmin test memcm mNtn (mx30efto10ensure limita that20. Q R Part the alars/ trip will cocar prior to ammading the The CFDARI12Tt and uma of this instnamentation is consistant 64 of Apper 11xwith A tothe repirwacita 10 CFR Part 50.of Genessl Design critaria 60, 63 arti 3.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION BASIS the railonctive geneous offluent instnamentatlan is pewided to nortitor and ccritrol, as applicable, the rsleases of rdulve materials in genecmas offluenta ering actual or potential releases of geneous etfluents. the alaruvtrip estpoints for these instrumenta are calmalated in nooorstance with the proce& ares in the cr? NITE cost mtstArtts smNtn (acco of limita to 10 ensure that 20. GR part the alaravtrip will occar grier to ensanading the the oppakI1nY and une of this instnamentaticri is canaistant with the requirwasnts of C'anerel Danign critaria 60, 63 and 64 of Aspendix A to 10 CFR Port 50. 3.3 LIQUID RADWASTE TREATMENT SYSTEM BASIS The requirement that these systems be used when specified - provides assurance that the releases of radioactive materials in 11guld af fluents will be kept "as low as is reasonably achievable" (ALULA). This specification implements the requirements of 10 CFR Part 50.36a, General Design criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix ! to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of
. Appendix I,10 CFR Part 50, for liquid ef fluents,
, 4
i i 3.0 SPECIFICATION BASES (CON'T) l 4 3.4 WASTE GAS SYSTEM BASIS i The requirement - that these systans be used when specified j provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept 'as low as is reasonable achievable" (ALARA) . This specification implements i the requirements of 10 CFR Part 50.36a, General Design Criterion - 60 of Appendia A to 10 CFR Part 50, and the design objectives given in Se G M n It D of Appendia I to 10 CFR Part 50. The. specified 14%w t4Hrning the use of appropriate portions of the systems were geel/%d as a suitable fraction of the dose design ; objectives set &yth in Sections 11.3 and II.C of Appeedfx.!,10 ' CFR Part 50 J W #seous effluents. -~ i j 3 LIQUli0 Uf(WHTS CONCENTRA110N BASIS This specificatka :ls provided to ensure that the concentration of radioanite storials released in liquid waste of fluents to ,
- -UNRESTRICt4D AABAS Part'will 20, be less than3 theTable concentration levels t
specified 1h 10 CFR Appendia II, Column 2. This limit,ation. provides additional assu,rance that the levels of radioactive materials la bodies of water in UNRESTRICTED ARIAS i will jectives result in esposures within (1) the section II.A design.ob - of Appendia . I,' 10 CFR 50, to a MEMBER OF TER PUBLIC and (2) the lialta of 10 CFR 20.106(e) to the population. The s concentration limit for dissolved or entrained noble gases is based upon the assumption that Io-135 is the controlling radioisotops .and :lts Mpc in air (submersion was converted to an equivalent esmeentration in water using the methods. described in
-Internationsi Commission on Radiologloal protection (ICRP)
Pub 11 cation. 2.) 3.6 -LIQU10 EFFLUENTS DOSE BASIS This specification is provided to implement the requirementsIof-Sections II. A, III.A and IV.A of' Appendia,1,10 CFR Part 50. The ; Limiting condition for operation implement.si the- guides- set forth in 5ection II.A of Appendia'I. The ACTION statement provides the " 1 requiredEoperating flexibility and at the same time--isglements
- the guider,-set - forth in section- IV. A of Appendia I to assure that-tho' releanes kept "asilow as ofisradioactive reasonablymaterial in li quid effluents will be achievable" (ALARA). The dose calculations in the OFFSITE DOSE CALCUI.ATION MANUAL {(ODCM) i implementi the requirements in Section III. A' of Appendis _I that conformance:with the guides of Appendia I be- shown by calculational procedures based on models and data, such that-the
_ . . _ _ . _ . _ _ - . . . _ _ A
3,0 SPEClflCATION BASES (CON'T) 3.6 LIQUID EFFLUENTS DOSE BASIS (CON'T) actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the OFFSITE DOSE CALCULATION MANUAL (ODCM) for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose Appendix 1,* of Evaluating Compliance with 10 CFR Part 50, Revision 1, October 1977 and Regulatory Guide 1.113,
" Estimating Aquatic Dispersion of Ef fluents f rom Accidental and Routine I,* April Reactor 1977. Releases for the Purpose of Implementing Appendix 3.7 GASEOUS EFFLUENTS DOSE RATE BASIS This specification is provided to ensure that the dose at any time at and beyond the SITE SOUNDAAY from gaseoun, ef fluents will be within the annual dose limits of 10 Cn Part 20. The annual dose 10 Cnlimits Part 20, areAppendia the dosesBassociated with the concentrations of Table II Column 1. These limits provide reasonable assuranc,e that radioactive material discharged in gaseous effluents will not result in the orposure of a MEMBF.R -
CF TU PUBLIC, either uithin or outside the SITE SOUNDARY to * " annual average concentrations escoeding the limits specified in ! Appendix 3, Table II of 10 CFR Part 20 (10 CM Part 2 0.106 (b) (1) ) . For a MEMBER 0F ME PUBLIC who say at time be within the SITE SOUNDARY, the occupancy of the MEMBER OF TH PUELIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDART. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above to a MIMBER OF 'mE PUBLIC at or beyond the $1TE SOUNDARY to less than or equal to 500 ares / yea'r to the total body or to less than or equal to 3000 aren/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mram/ year. l l a _ . -- . .
__ _ _ _ _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ ~_--____ 3.0 SPEClflCATION BASES (CON'T) 3,8 GASEOUS EFFLUENTS DOSE f40BLE GASES BASIS This specification is provided to implement the requirements of Sections II.3, III.A (6nd IV.A of Appendix I, 10 CFR Part 50. The Limiting condition for operation implements the guides set forth in Section 11,5 of Ape ndiz 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendiz ! to assure that the releases of radioactive material in gaseous af fluents will be kept 'as low as is reasonably achievable" ( ALARA) . he surveillance Requirements Laplement the requirements in section III. A of AppendLa I that conformance with the guides of Appendix I be shown by calculationa:. procedures based on models and data such that the actual erposure of a MEMBER OF TER PUBLIC through appropriate pathways is unlikely to be substantially underestimated. -The dose calculations established in the orFSITI Dose CALCULATION MANUAL (CDCM) for calculating the doses due to the actual release rates of radioactive noble gases in gaseous ! etfluents are consistent with the methodology provided in Regulatory Guide 1.109, ' Calculation of Annual Doses to Man from Routine Releases of Reactor Ef fluents for the Purpose of Realuating Compliance with 10 CFR part 50, Appendim I,' Revision 1, October 1977 and Regulatory Guide 1.111 " Methods for Ratisating Atmospheric Transport and Disper,sion of Gaseous Effluents Revision 1 July 1977. in Routine Releases from Light-Nater Cooled Reactors
- equations ,provided for determinit.g the air doses at and beyo the SITE DOUNDART atmospheric are based upon the historical average conditions.
3.9 GASEOUS EFFLNENTS. DOSE l-131, TRITIUM, AND RADI0 ACTIVE PARTICUL BASIS This specification is provided to implennt the requirements of SecMons II.C, III.A and IV. A of Appendia I,10 Crr, nrt 50. The Limiting conditions for operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in section IV.A of Appendiz ! to assure that the releases of radioactive materials in gaseous effluents will be kept 'as low as is reasonably achievable" ( ALARA). The OFFSITE DO5E CALCULATION MANUAL (ODCM) calculational methods specifled in the surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual esposure of a MEMBER OF TIE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The orFSITE DOSE CALCULATION MANUAL (ODCM) methods for calculating the dose due.to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, ' Calculation of Annual Doses x . . _ - . . - _ _ - _ _ - - -
- . - . - - - - . - . - ~ . _ _ - - - . - _ _ - - - - - - _
I i 3.0 SPECIFICATION BASES (CON'T) 3.9 GASEOUS EFFLUENTS DOSE l 131,1RITIUM, AND RADl0AC11VE PARTICULATE
- BASIS (CON'T) !
4 to Man from Routine Releases of Reactor tifluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1,*
- Revision 1, October 1977 and Regulatory cuide 1.111, ' Methods for "j tatisating Atmospheric Transport and Duspersion of caseous i Effluents in Routine Releases from Light-Water-Cooled Reactors,' '
Revision 1, July 1977. These equations also provide fer determining the actual doses based upon the hlstorical average ; etmospheric conditions. The release rate specifications for i Iodine-131, Tritium, and radioactive particulates with half-life less than eight days are dependent on the existing radionuclide ' pathways to man, in areas at and beyond the SITE 300NDARY. The i pathways which were esamined in the development of these calcu-lations were: 1) Individual inhalation of airborne radionuclides, 1~
- 2) deposition of radionuclides onto green leaf vegetation with subsequent consumption by man, 3) deposition onto grassy areas i where alik animals and meet producing animals grase with con-sumption of the allk and meat by man, and 4) deposition on the ground with subsequent esposure of man.
3.10 TOTAL DOSE BASIS . t This specification is provided to meet the dose limitations of 40 C3 R Part 190 that have now been incorporated' into -10 CFR Part 20 by 44 FR 19525. The-spe-ification requires the preparation and t submittal of a Special Report whenever the calculated doses from , plant radioactive effluents esoeed twice the design objective doses of Appendia 1. For sites sentaining up to 4 reactors, it 3 is highly unlikely -that the resultant dose to a MEMBER OF TIE - PUBLIC wLil aseeed the dose limits of-40 CFR Part-190 if the individaal reactors remain within the reporting requirement : level. The hyesial Deport will describe a oourse of action that should ,re,sult in the limitation of the annual dose to a-MEMBER OF
' TER ptfBLIC to within the 40 CFR Part 190 limits. For the ';
0 purposes of the Special Report, it may be assumed that the dose l
- coma)Dat to the MEMBER OF TER PUBLIC from other uranium fuel i L
- cycle sources is negligible, with the eaception that dose contributions from other nuclear fuel cycle facilities at the same-site or within a radius-of 8 km must be considered. If the ,
dose to any MEMBER OF TIE PUBLIC is estimated to exceed the requirements.of 40 CFR Part 190, the Special Report with a request for a variance (provided the' release conditions resulting in - violation of 40. CFR- Part .190 have not already been- corrocted) , - in-.accordance with the provisions of 40 CFR Pairt -190.11 and 10 CFR Part 20.405c, is considered to be a timely request and F fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of i 40 CPR Part 190 and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed l in specifications a J nu .r , 9
. An individual is not considered a MEMBER OF TEE PUBLIC during any period in which i
he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle. y,- .-. , ,
l i 3.0 SPECIFICAil0N BASES (CON'T) 3.11 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM basis i The radiological monitoring program required by this t i specification provides measurements of radiation and of ! radioactive materials it those esposure pathways and for those radionuclides which lead to the afghest potential radiation ' esposures of MEMBERS OF' TIE PtT5LIC resulting from the station i operation. This monitoring program thereby supplements the
- radiological effluent monitoring program by verlfying that the measurabis radiation are?oncentrations not higher than oforpected radioactive onmaterials the basicand of the levels of ,
ef fluent measurements and modeling - of the environmental espesure pathways. esperience. Program changes may be initiated based on operational The LLD's required by Table 29 are considered optimum for routine environmental measurements in industrial laboratories. ' The LLD's for-drinking-water meet the requirements of 40 CFR 141. 3.12- RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LAND USE CENSUS BASIS This specificatian is provided to ensure that changes in the use of areas at or beyond the_ s:TE 30tmDARY are -identified and that modifications to the monitoring program are made if required by the results of this census. Adequate information gained from ' door-to-door or aerial surveys- or through consultation with local agricultural authorities shall be used. This census satisfies ; the requirements of Section IV.R.3 of Appendis I to-10 CFR Part
- 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways _via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 3 1.109 for consumption by a child. To determine this minimum ;
l garden size, the following : assumptions were used: 1) that-20% of the garden was used for growing broad leaf vegetation (i.e., similar-to lettuce and cabbage), and 2) a vegetation yield of 2 - kg/ square meter. ; 3.13-PROGRAM RADIOLOGICAL BASIS ENVIRONMENTAL MONITORING INTERLABORATORY COMP The requirement for participation in an Interitboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the me6surements of radioactive ; material in environmental sample matrices are-performed as part : of the quality assurance program for environmental monitoring in ! order-to demonstrate that the results are-reasonably valid, i W- , _ . - , - - ---,-,,__~-__-,,i.,...,~
- _2_m m.a42 -i--.** a-m -. - - __.dJ - -=h. a.A---4&aA++ _aM 4.=-MM A- N&J4---A---4hhJ+-ava+a--*AM+ee- e.#--*dhCMv+--- A EM 5--M===43. 4 J --.-.i..#,,Mah4-8J---g-45kh- w m&.$4AJ_. .
?
b b s PART 11 METl10001.0GIES ,
=,-,, ....., -W _. . . ,e,ae , , , , . .m- --'
i i i h i l I
*f 1
'I t
i 1 t i i i i i 1 i i SECTION 1.0 ,! RADIOACTIVE EFFLUENT ! MONITOR SETPOINTS SPECIPICATIONS ! i e
=;
i 1 t
.i i
i i I 5 h- t
.t.. n' N WEBuESlOX"O:
,,t, * ?T**{.,fk,,.*,.----,-,,%.w~,,
.-**-3---ee*"*-e'6++* *"*""*O#'Y'4""'-*"******#'*#'***-'*-' '"# ~~
TABLE I - RADIDACTIVE EFFLUDIT NONITOR SETPOINTS D i IT1 i
.g u
RELEASE TYPE SETPOINT NUCLIDE ANat. SETPOINT SETPOINT SPECIFICATION CALCULATION pe0NITOR ADJUSTNENT ("f) BATCH CONT. TYPE ** FREO. O- RM-Al X 1.1-1 1.2-1 P i.3-1 3.4-i g (Noble Gas) RM-Al X 1.1-1 1,2-1 g 3,3-3 3,4. 2 (Noble Gas) A RM-A2 X* X 1.1-1 1.2-2 W/P* 1.3-1 1.4-3 (Noble Gas) RM-All X 1.1-1 1.2-3 P 1.3-1 1.4-4
. (Noble Gas)
~
RM-L2 X 1.1-2 1.2-4 P 1.3-2 1.4-5 (Gamma) RK-L7 X X 1.1-2 , 1.2-5 W I 3-2 1.4-6 & l.4-7 iGamsa) RM-Al & RM-A2 N/A N/A 1.1-3 NA NA 1.3-3 NA tlodine Channels)
*This monitor 1;' used in conjunction with (or instead ef) RM-AII to monitor the release of the waste gas decay tanks. Fluclide analysis and setpoint calculation mus; be performed for this monitor prior to waste gas decay tank release., At all other times, it is a continuous saurce monitor and the setpoint is determined weekly.
**For composited samples the results from the most recently completed analysis are used.
l
y GASEOUS EFFl.UENT MONITOR 3 . SETPOINT SPECIFICATION 1.1-1 (Monitors RM-A1, RWA 2 and RM-All) The dose rate at or beyr.,ad the SITE BOUNDARY, due to radioactive materials teleased in gaseous ef fluents, is limited as follows: Noble Gases - 500 mrem / year (total body) 3000 mrem / year (skin) 1-131, Tritium and radioactive 1500 mrern/ year (any organ via particulates with the inhalation pathway.) greater than 8 day half-lives ,1 The radioactive gaseous ef fluent monitors (RLAl, RM-A2 and RM-All) shall have their alarm / trip setpoints set to ensure that the above total body, noble gas dose rate limit is not exceded. e
References:
( N
))' Technical-Specif ica tion-3r3:3;9tB33:3:9-
,2T Technical-Specification-3cl-hhttB3;rl2T
- 3) Plant Procedures I.
1 REV!SION "0"
LIQUID E E UENT N00lITORS SETPOINT SPECIFICATI00t 1.1-2 (flonitors RN-L2 , RR-L7 ) ~ The concentration ot' radioa';tive materials in liquid effluents, released to UNRESTRICTED AREAS, is lL@;ed to the concentrations speci!ied by 10 CTR 20. Appendix B, Table II, Column 2 for radionuclides other than noble 1ases, and is limited to 2E-4 pC1/al total activity concentra.*. ion for all 24 a 'olved or entrained noble gases. The radioactive liquid effluent monitots (RN-L2 and RM-L7) shall have their alarn/ trip setpoints set to ensure that the above gamma emitting concentration limits are not exceeded.
.~.
O .
References:
, 1) -Technical-Specification-37373 3
- 2) Technica1-Specification-3rt1-1+
- 3) Plant Procedure
. O REV S 0 M 10
. - . . - . .......-. - ~_~ - _ - . - - . . - _ - . - - - - . - - - - . . - . I l i GASEOUS EFFLUENT MONITORS - - SETPOINT SPECIFICATION 1.1-3 (lodine Channels in RM-A1 and RM-A2) Sampling and analyses of the Reactor Building Purge Exhaust, and the Auxiliary Building and Fuel Handling Area Exhaust for radiolodine and other gamma emitters, shall be performed at least once ser 24 hours for at least 7 days following each shutdown, startup or change in power leve, exceeding 15% of RATED THERMAL POWER within one hour, when the Radiolodine concentration in the Auxillary Building and Fuel Handling Area or the Reactor Building Purge Exhaust Ducts will lead to a release which is greater than or equal to 10% of the 10 CFR 20. ^apendix B. Table 11, Column I limits, at or beyond the SITE BOUNDARY. The lodine monitoring channels in radiation monitors RM-Al and RM-A2 shall have their alarm setpoints set to alarm when the above radiolodine concentration limits are exceeded.
~
,s. '
x Ref erences: .
- 1) Technical-S:e ifica+ ion 9rit?rirTable 4.14-2rFootnote4d)-
- 2) Plart Procedures
-s-
. ;,.c.......
~<J.ui- ,,
- -- . .. . . . _ - _ - . . . --- - - -- .. ~ . NUCLIDE ANALYSIS 1.2-1 REACTOR BUILDING PURGE EIRAUST NUCLIDE SAMPLE SOURCE LLD(b)(ucl/cc) O A. Principal Ganaa Emitters (*) Mn-54 1x10-4/1x10-11 Fe-59 1x10*4/1x10-11 Co-58 Pre-release grab sample for Batch 1x10-4/1x10-11 Co-60 Type release. Weekly Particulate In-65 1x10-4/1x10-11 Filter Analysis for continuous (c) - 1x10-4/1x10-11 Mo-99 type release. 1x10-4/1x10-11 Cs-134 1x10-4/1x10-11 Cs-137 1x10-4/1x10-11 Ce-141 1x10-4/1x10-11 Ce-144 1x10-4/1x10-11 Kr-87 Pre-release grab sample for Batch 1x10-4 Kr-88 type release. Noble Gas monitor in10-4 Xe-133 during batch and continuous releases 1x10-4 Xe-133m - Grab sample within 2-6 hr. following - 1x10-4 Xe-135 startup, shutdown or > 15% TTP 1x10-4 Xe-138 change in 1 hr. 1x10-4 B. Iodine 131 Pre-release grab sample for Batch NA/1 x 10-12 ,
%. type release. Weekly charcoal filter f
N. and once per 24 hr for 7 days following startup shutdown or > 15% RTP change in 1 nr unless I-131 concentration.at site boundary < 10% 10 CFR 20 limit. C. Tritita Pre-release Grab Sample and within 1x10-6 12-24 hr following flooding of ' refueling canal and once per 7 days while canal is flooded. D. Gross Alpha Monthly Particulate Filter Composite 1x10-11 E. Sr-89 Quarterly Particulate Filter Composite 1x10-11 F. Sr-90 Quarterly Particulate Filter Composite 1x10-11 (a) Other identified Gamma Emitters not listed in this table shall be included in dose and setpoint calculations.
-(b) The first value re'-rs to the LLD for pre-release grab sample; -the second value refers to the LLD for weekly Particulate Filter Analysis.
(c)- Reactor Building Purge is considered continuous after a minimum of one Reactor Building volume has been released on a continuous basis (i.e., first volume is a batch type). ; O
-e- REV SLN 11 ;
NUCLIDE ANALYSIS 1.2-2 AUXILIARY BUILDING AND FUEL HANDLING AREA EXHAUST-NUCLIDE SAMPLE SOURCE LLD (b) (uci/ml) (/ A. Principal Gamma Emitters IA) Mn-54 Fe-59 1x10-4/1x10-11 Co-58 1x10-4/1x10-11 Weekly Particulate Filter Analysis. 1x10-4/1x10-11 Co-60 En-65 1x10-4/1x10-11 Mo-99 1x10-4/1x10*11 Cs-134 1x10-4/1x10-11 Cs-137 1x10-4/1x10-11 Ce-141 1x10-4/1x10-11 Ce-144 1x10-4/1x10-11
- 1x10-4/1x 1011
' ~
Kr-87 Monthly Grab Sample and 1x10-4 Kr-88 Continuous Noble Gas monitor. 1x10-4 Xe-133 - Grab sample within 2-6 hr following - 1x10-4 Xe-133m startup, shutdown or *lP15% RTP 1x10-4 Xe-135 change in 1 hr. 1x10-4 Xe-138-1x10-4 B. Iodine 131 Weekly Charcoal Filter analysis and once 1x10-12
,' per 24 hr for 7 days following startup i shutdown or > 15% RTP change in.
I 1 hr unless I-131 cor. centration at site boundary < 10\ 10 CFR 20 limit. C. . Tritium- Monthly Grab Sample and within 1x10-6 12-24 hr following flooding of refueling canal-and once per 7 days while canal is flooded.
'D. ' Gross Alpha Monthly Particulate Filter Composite 1x10-11 E. St Quarterly Particulate Filter Composite 1x10-11
- r. St ' Quarterly Particulate Filter Composite 1x10'11 (a) Other in-dose identified Gamma and'setpoint Emitters not listed in this table shall be included calculations.
(b) The first value refers to the LLD.for pre-release grab sample; the second value refers to the LLD for weekly Particulate Filter Analysis. d O REVISION 11
NOCLIDE ANALYSIS 1.2-3 WASTE GAS DECAY TANKS NOCLIDE SAMPLE SOURCE LLD(b)(uC1/al) . A. Principal Gamma Emitters (a) Mn-54 Fe-59 1x10-4/1x10-11 Co-58 1x10-4/1x10-11 Co-60 1x10-4/1x10-11 En-65 - 1x10-4/1x10-11 Pre-release Grab sample and Weekly - 1x10-4/1x10-11 Mo-99 Particulate Filter Sample from RM-A2. Cs-134 1x10-4/1x10-11 Cs-137 1x10-4/1x10-11 Ce-141 1x10-4/1x10-11 Ce-144 1x10-4/1x10-11 1x10-4/1x10-11 Kr-87 1x10-4 Kr-88 1x10-4 Xe-133 - Pre-release Grab sample. - 1x10-4 Xe-133a 1x10-4 Xe-135 1x10-4 Xe-138 1x10-4 C\ V B. Iodine 131 Weekly Charcoal Filter froa RM-A2. 1x10-12 (a) Other identified and setpoint' Gamma -Emitters not listed in. this table shall be included in dose calculations, _( b)- The first value refers to the LLD for pre-release- grab sample the - second value refers to the LLD for weekly Particulate Filter Analysis. O
... REVISIS 11
NUCLIDE ANALYSIS 1.2-4 EVAPORATOR CONDENSATE STORAGE TANKS, LAUNDRY AND SHOWER SUMP TANKS, SECONDARY DRAIN TANK
. NUCLIDE SAMPLE SCURCE LLD(uCl/ml) ,
A. Principal Gamma Emitters (a)
. Mn-54 3 3x10-7 Fe-59 3x10-7 Co-58 3x10-7 Co-60 3x10-7 Zn-65 _ Pre-release Grab Sample 3x10-7 Mo-99 5x10-7 i Cs-134 5x10-7 Cs-137 5x10-7 Ce-141 5x10-7 Ce-141 Jx10-7 B.- todine t31 - Pre-Release Grab Sample 1x10-6
- C. Dissolved and-Entrained Noble
' Gases i Monthly Grab Sample 1x10-5
. D. . ' Tritium Monthly Composite. '
lx10-5 E. Gross Alpha Monthly Composite lx10-7 ' F. : 5r-89 Quarterly Composite 5x10-8
- G. Sr-90_ Quarterly Composite' 5x10-8
'H. Fe-55 Quarterly Composite ' lx10-6 (a)- : Other ' identified Gamma Emitters not listed in this table shall be included in dose and setpoi calculations, .
O REV S ON 7
NUCLIDE ANALYS15't.2-5 SECONDARY DRAIN TANK AND/OR PLANT CONDENSATE A
- (j NUCL1DE SAMPLE SOURCE '
LLD(uCl/ml) A. - Principal Gamma Emitters (a)
~
Mn-54
~5x10-7 Fe-59 5x10-7 Co-58 5x10-7 Co-60 5x10-7 Zn-65 Weekly Compustte 5x10-7 Mo-99 3x10-7 Cs-134 3x10-7 Cs-13/
3x10-7 Ce-141 5x10-7 Ce-lu _5x10-7 B. lodine 131 Weekly Composite Ix10-6 C. Dissolved and ' p ' Entrained Noble Gases Monthly Grab Sample lx10-5 D. Tritium Monthly Composite Ax10-5 E. Gross Alpha Monthly Composite Ix10-7 F. Sr-89 Quarterly Composite 5x10-8 G. Sr-90 Quarterly Composite 5x10-8 H. Fe-55 Quarterly Composite Ax10-6 (a) Other icentified Gamma Emitters not listed in this table shall be included in dose and setpoint , calculations.- g 10 - REVS ON 7
PRE-RELEASE CALCULATION 1.3-1 GASEOUS RADWASTE RELEASE [ I. INTRODUCTION Prior to initiating a release of gaseous radwaste, it must be determined that the concentration of radionuclides to be released, and the flow rates at which they are released will not cause the dose rate
-limitations of Specification 1.1-1 to be exceeded.
II. IE.0RMATION REQUIRED Results of appropriate Nuclide Analysis from Section 1.2 III. CALCULAT10E Noble Gas Gamma Emissions Dose Rate (Total Body) - I (X/Q)KjQt mrem /yr. (1.1) Noble Gas Beta Emissioni Osse Rate (Skin) = I (X/Q)Qi(Lj + 1.lMj) mrem /yr. (1.2) Iodine 131. Tritium. Radioactive Particulates Dose Rate (1,T,P) - I (X/Q)PjQj mrem /yr. (1.3) - where: Kj - The total body dose factor due to-gamma emissions for egch identified noble gas radionuclide, in mrem /yr per #C1/md. (See Table 4.4-1), j 9 Li The skin dose factor due to beta emissions for each identified noble gas radionuclide, in mrem /yr per pCi/m3 (See Table 4.4-1). l Mt The air dose factor due to gama emissions for each identified noble gas radionucide, in mrad /yr per gC1/m,- (unit ~ conversion constant of 1.1 mrem / mrad converts air dose to skin dose). (See Table 4.4-1). j Pj- - The dose. parameter for radionuclides other than noble gases for the inhalation pathway, in mrem /yr per pC1/m3 (See Table 4.4-3). l
-Qi -
The release rate of radionuclides, i, in gaseous effluent from individual release sources, in #Ci/sec (per unit, unless otherwise specified). Qj -Effluent stream nuclide concentration x flow rate. 7 [J R EVISION 18 "-
Flow Ratos (Verlable - based on satpoint needs, nosinal or cax-laus values listed below).
- 1) Reactor Building Purge Exhaust Duct =
50,000 cfm = 2.4 x 107 cc/sec .
- 2) Auxiliary Building and Fuel Handling Area Exhaust Duct = 156,000 cfm = 7.4 x 107 cc/sec
- 3) Waste Gas Decay Tank Release Line = 50 cia max = 2.4 x 104 cc/sec X/0 = 2.5x10-6 sec/m3 for all vent releases.
This value is - highuc calculated annual average relative concentration for any area at or beyond the unrestricted area boundary. Ir. : der for a gaseous release to be within the limits of specification 1.1-1, the Projected Dose Rate Ratio (PDRR) must not exceed 1. The PDPR for each limit is calculated as follows: PDRRTB = PDRTB / 500 (1.4) PDRRSK = PDRSK / 3000 (1.5) PDRRORG = PDRORG/ 1500 (1,6) PDRTB = Projected Dose ' Rate to the TOTAL BODY due to noble gas emaissions. PDRSK = Projected Dose Rate to the SKIN due "to noble gas emaissions. PDRORG= Projected Dose Rate to any organ due to inhalation of lodine, . tritium and particulates with half-lives grec er than 8 days. 500 = The allowable total body dose rate due to noble -' gas ganaa emissions in area /yr.
-3000 = The allowable skin dose rate due to noble gas beta emissions in area /yr.
_1500 = The allowable organ dose rate in area /yr. If the concentration- of radionuclides to be released is less
-than the effluent monitor LLD set-PDRR equal to 1.
' Equations 1.1, _1.2, and 1.3 are solved for each release type and release point currently releasing or awaiting release. It relationships - 1.4, .1.5, ' and 1.6 are satisfied, the release can be under the assumed flow rates. If one or more of the relationships 1.4, 1.5 and ' 1. 6 are not satisfied, &ction must be
_made
**"* ' d"c '" " d1 ""c21d ' t ' ' "'* ' * <D_
release (or to reduce the radionuclide release rate already in
*"1*i **"' -
progress). REVISKN 11
. . . . :, . , - = ..= . . . . . . .-
^ "
. ;u ~ ' T.
^~
~;._
l The following actions are available to reduce the release rates at the three release points.
- 1) Waste cas Degav Tanks a) Release Valve may be throttled b) Tank contents may be diluted c) Release may be delayed for longer riecay time.
- 2) Reactor Buildino Purae Exhaust Duet a) Dilution flow may be opened to reduce purge rate while maintaining the same flow rate.
- 3) Auxiliary Buildino and Fuel Handlino Area Exhaust a) Reduce inlet a upply to areas in Auxiliary Building to reduce radioactivir *
. r-te to vent, b) Identify av , e- -
soui: of radioactive releases into the Auxiliary D -k Etfluent Monitor-LLD Determina " The Technical Specification LLs < relationship given below may be used to calculate a monitc; LLD.
. . , _ 4.66'YB~
O Slope
=
B Average monitor background count rate in cpa. Slope = Slope of monitor calibration curve in cpa/pci/al
.o .
REVISK N 11
PRE-RELEASE CALCULATION 1,3-2 LIQUID RA0 WASTE RELEASE G V !. . INTRODUCTION . Prior to initiating a release of liquid radwaste, it must be determined that the concentration of radionuclides to be released and the flow rates at which they will be released will not lead to a release concentration greater than the limits of specificatinn 1.1-2 at the point of discharge. II. INFORMATION REQUIREQ Results of appropriate Nuclide Analysis from Section 1.2 III. CALCULATIONS Discharge- Cyj Cg 0+E Concentration = I + Ca CT Cs Ce F
+ + + + +
MPC7j 2E-4 MPCo MPCT MPCs MPCr.. E where: Cyj - The concentration of isotope i, in the gamma spectrum exluding dissolved or entrained noble gases. Cg =
. Total dissolved or entrained noble gas concentration. x CT
=
Tritium Concentration from most recent analysis. L Co - Gross alpha concentration from most recent analysis. l Cs Sr-89, 90 concentration from most recent analysis, n Cre Fe-55 concentration from most recent analysis. E - Effluent Stream Flow Rate 0 - Dilution Stream Flow Rate (Nuclear-Services seawater flow only)- M P'. - 10CFR20 Appendix B, Table II, Column 2 Maximum-Permissible Concentration by isotope, i If the Calculated Discharge Concentration is less than or equal to 1, the discharge may be initiated. If the calculated discharge concentration is greater than 1, action must be taken to reduce the effluent concentration or effluent stream flow rate prior to initiating discharge. O REV SIN 13 _.- . - . _ -
.+$1d .E* - ~ _,...h# .._..e.-.A.u .*
PRE-RELEASE CALCULATION 1.3-3 GASEOUS EFTLUENT IODINE MONITORS I. INTRODUCTION In order to determine the setpoints for these monitors, the following assumptions are used. A. The release rate through the Auxiliary Building and Fuel Handling Area exhaust duct is 7.4 x 107 cc/sec. (156,000 cfa). B. The release rate through the Reactor Building Purge Exhaust Duct is 2.4 x 107 cc/sec (50,000 cfa). C. A limitless supply of uniformly concentrated I-131 is available to supply.the Exhaust Ducts. D. The iodine filter has been installed for 8 hours and operating at a constant flow rate of 472 cc/sec (1 cfa Therefore, total flow through the filter has been 1.3f x 107 cc.) . II. CALCULATIONS The limiting. concentration of Iodine in the vent which would result in a concentration of one-tenth the 10 CFR 20 limit at the site boundary is - calculated as follows: - Cy. 0. q (3,7) (X/Q) FK where: Cy = The Concentration of Radiciodine in the vent in uCi/cc. CI = The-10 CFR 20 Appendix B, Table II Column 1 concentra-tion limit for Iodine 131, 1 x 10-IO uC1/cc. F = The duct flow rate: 2.4 x 107 cc/see for the Reactor. Building Purge Exhaust Duct and - 7.4 x 107 cc/sec for the Auxiliary Building and Fuel Handling Area Exhaust-Duct. K = Unit conversion constant 1 x 10-6 m3 /cc X/0 = The highest calculated annual average concentration for any area at or beyond the unrestricted area boundary 2.5 x 10-0 sec/m3 Solving egn. 1.7 for the' Reactor Building Purge exhaust vent yields: (RB) = 1.67 x 10-7 uCi/cc e \ Solving egn. 1.7 for the Auxiliary Building & Fuel Handling Area Exhaust vent yields: C V(AB) = 5.41 x 10-8 uCi/cc
- 's -
REVISLN 10
In ' order to determine the total quantity of Iodine 131 collected on the f filter, the values of Cy above are multiplied by the volume assumed to
.s .
have passed through the filter
- Og = fkcy (1.8) where:
Qg = The total quantity of Iodine 131 collected on the filter in uC1. Cy = The concentration of Iodine 131 in the vent in uCi/cc.
=
f The assumed total volume of vent atmosphere that has pas-sed through the filter, 1.36 x 107 cc (1 CFM for 8 hours). k = The Iodine removal efficiency of the filters: 90\ Solving egn. 1.8 for the Reactor Building vent yields: Og = 1.08 uCi r
~._. .f Solving egn.- 1.8 for the Auxiliary Building and Tuel Handling Area vent -
yields: Qg = 6.62 x 10~1 uCi These values are converted to counts per minute for the Iodine monitoring
-channels through use of the appropriate calibration curves.
l l n v
- 1e - REVISl0N 10
l l Setpoint Calculation 1.4-1 Reactor Building Purge Exhaust Duct Monitor (RM-A1) (batch Type Releases) INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded. METHODOLOGY Reactor Building. atmosphere is circulated through radiation monitor RM A6 ' (containment atmosphere noble gas monitor) and the count rate is observed. The-observed count rate is correlated to a corresponding count rate for RM-Al , (Reactor' Building purge exhaust duct monitor), and factors are applied-to ' account for background radiation, and the pressure difference batween the detector chambers and exhaust vent. The obtained value establishes the l maximum allowable setpoint. The alarm / trip setpoint is adjusted to this or a 4 more conservative value prior to initiating the release, if the concentration of radionuclides to be released is less than the effluent monitor LLD " Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 yCi/ml. CALCULATION Net CPM x VF /29.9 - Vlh (pCi/cc/ CPM)A6
.. JtM-Al Setpoint (CPM) - I
+ Bkg O where:
PORR ( 29.9 -V6/ (gCi/cc/ CPM)Al Net CPM - The observed RM A6 count rate, in cpm, less background, or obtained from the calibration curve. VF - The vent fraction; that wrtion of the total plant gaseous release associated with this vent and discharge type. Value can be set to a number between 0 and 1. The summation of the vent fractions of=RM-Al and RM-A2 cannot exceed 1.- PORR - The noble gas ganna emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable' dose rate referenced in Section 1.3-1, relationship 1.4. V6- - The actual gauge vacuum reading at RM-A6 at the time of sampling. VI - The actual or average gauge vacuum reading at RM Al during narmal operation. (gCi/cc/ CPM)A6 - #Ci/cc per cpm for RM-A6. This is based on an actual sample-or derived from the calibration curve.
~"~
t 1EV SiON 13
,,,,,,..,_m. . _
(yci/cc/ CPM)Al - pCi/cc per cpm for RM-A1.- This is based on an actual sample or derived from the calibration curve. Bkg - RM Al background count rate in cpm.
.~. .
U O
~~
REVIS ON 13
Setpoint Calculation 1.4-1A steactor. Building Purge Exhaust Duct Monitor (RM-A1) (Special Release for Functional Testing of the Reactor Building Purge System) INTRODUCTION Following completion of the analyses required by Section 1.2-1 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded. METHODOLOGY Auxiliary Building and Fuel Handling Area atmosphere is continuously passed through radiation monitor RM-A2 and the count rate is observed. The observed count rate is correlated to a corresponding count rate for RM-Al, and factors are applied to account for background radiation and the pressure difference between the detector chambers and exhaust vent. The obtained value establishes the maximum allowable setpoint. The alarm / trip setpoint 's adjusted to this or a more conservative value prior to initiating the release. If the concentration of radionuclides to be released is less than the effluent monitor LLD_" Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 #Ci/ml. CALCULATIDE Net CPM x VF [ 29.9 - V1h (gC1/cc/ CPM)A2
-RM Al Setpoint (CPM) - + Bkg PDRR
( 29.9 - V2) (gCi/cc/ CPM)Al _ where: - I Net CPM = The observed RM A2 count rate, in cpm, less background, or obtained from the calibration curve. VF = The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type.- VF can be set to a value from 0 and 1. The sum of RM-Al and RM-A2 vent fractions can not exceed 1. PDRR - The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate: referenced in Section 1,3-1, relationship 1.4. V2 = The actual gauge vacuum reading at RM A2 at the time of sampling. VI - The actual or average gauge vacuum reading at RM-Al during normal operation.
- O
- '** ~
REVISl5N 14
. , . - . . - - - . - _ . - - . - - - . - _ _ - . . . . - ~ . . _ - . . . . . . . . - - . - - . . _ ,
#ci/cc per cpm-for RM A2. This is based on an actual (yC1/cc/ CPM)A2 =
sample or derived. from the calibration curve. O .(#C1/cc/ CPM)Al
- pCi/cc per epm for RM.A1.
sample or derived.from the calibration curve. This is based on an actual B kg .- - - RM Al background count rate in cpm. q O O
.+
h t O REVISION 13 - 188 -
Setpoint Calculation 1.4-18 Reactor Building Purge Exhaust Duct Monitor (RM-A1) (Special Release Following ILRT of Reactor Building) INTRODUCTION Following completion of the analyses required by Section 1,2-1 and determination of release rates and concentration limits in accordance with Section 1.3 1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded. METHODOLOGY Reactor Building atmosphere is circulated through a sampling-apparatus. The Noble gas sample is analyzed to determine the projected dose rate ratio (PDRR). Het CPM is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 gCi/ml. These values are combined with the monitor background and vent fraction, to arrive at the monitor setpoint. The obtained value establishes the maximum allowable setpoint. The alarm / trip setpoint is adjusted to this or a more conservative value prior to initiating the release. Shortly, after beginning the purge, naw RM Al alarm / trip setpoints are determined using the methodology of Setpoint Calculation 1.4-2, LALCULATION
~ ~ . Net CPM x VF RM Al Setpoint (CPM) = + Bkg where:
Net CPM - A value derived from RM-Al calibration curve. VF = The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. 'VF can be set to a value from 0 and 1. The. sum of RM-Al and RM A2 vent fractions'can not exceed 1. PDRR - 1 Bkg - RM Al bacxground count rate in cpm. O
- 'ac -
L 3EVISION 13
# w ,- .w=.,+ r ---,,--,ge, # - - %-Sw , -
. - .- .w-,~ew. .
-.w- -- ,
Setpoint Calculation 1.4 2 Reactor Building Purge Exhaust Duct Monitor (RM A1) 7 (Continuous Type Releases) V - INTRODUCTION fo1 % wing completion of the analyses required by Section 1,2-1 and determination-of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded. METHODOLOGY React r Building atmosphere is passing through radiation monitor RM Al during a continuous type release. Factors are applied to the observed count rate to account for background radiation and vent fraction. The obtained value establishes the maximum allowable setpoint. The alarm / trip setpoint is l adjusted to this or a more conservative value weekly during continuous releases. If the concentration of radionuclides to be released is less than the effluent monitor LLD " Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 2.5E-2 pCi/ml. CALCULATION Net CPM x VF RM-Al Setpoint (CPM) = + Dkg PDRR
~ ~ . _.
where: Net CPM - The observed RM.Al count rate, in cpm, less background, or obtained from the calibration curve. VF = The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. Value can be set to a number between 0 and 1. The summation of the vent fractions of RM-Al and RM-A2 cannot exceed 1. PDRR - The noble gas gamma emission Projected Dose Rate Ratio calculated-in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1.3-1, relationship 1.4. Bkg- - RM-Al background count rate in cpm. O
-"~
REVISK N 13
. _ . . . . . - . . - . . - - - - - - - - . . . . - - - . . - - . . - ~ _ .
l 1 Setpoint Calculation 1,4-3 Auxiliary Building & Fuel Handling Area Exhaust Monitor (RM-A2) (Continuous Type Releases) INYRODUCTION
-following completion of the analyses required by Section 1.2 2 and determination of release rates and concentration limits in accor"ance with Section 1.3-1,'the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded.
METHODOLOGY Auxiliary Building and fuel Handling Area atmosphere is continuously passing through radiation monitor RM A2. Factors are applied to the observed count rate to account for background radiation and vent fraction. The obtained value establishes the maximum allowable setpoint. The alarm / trip setpoint is j adjusted to this or a more conservative value weekly during continuous releases. If the concentration of radionuclides to be released is less than the-effluent monitor LLD " Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 8E 3 pCi/ml. CALCULAT1qN Net CPM x VF RM-A2 Setpoints (CPM) = + Bkg PDRR where: Net CPM _ - The observed RM-A2 count rate, in cpm, less background, or obtained from the calibration curve. VF = The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. Value can be. set to a number between 0-and 1.- The summation of the vent fractions of RM Al and RM-A2 cannot exceed:1. PDRR - The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. .This ratio is the actual-projected dose rate divided by the allowable dose rate referenced in Section 1.3-1, relationship 1.4. Bkg - RM-A2_ background count rate in cpm. O RERSmN 33 n - n- .__ _ _ = _ . - - - - - - - - - - - ~ - - - - ~ - ~ ~ ~
Setpoint Calculation 1.4 4 Waste Gas Decay Tank Monitor (RM-All) (Batch Type Releases) INTRODUCTION Following completion of the analyses required by Section 1.2 3 and determination of release rates and concentration limits in accordance with Section 1.3-1, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentration limits are exceeded. METHODOLOGY Prior to initiating a Waste Gas Decay Tank release, its contents are drawn through radiation monitor RM-All and returned to the waste gas header. Factors are applied to the observed count rate to account for background radiation and vent fraction. The obtained value establishes the maximum allowable setpoint. The alarm / trip setpoint is adjusted to this or a more l conservative value weekly during continuous releases. If the concentration of radionuclides to be released is less than the effluent monitor LLD " Net CPM" is obtained from the calibration curve by determining the CPM which corresponds to 20 pCi/ml. CALCULATION Net CPM x VF x 24.7 RM All Setpoint-(CPM) = + Bkg
'~. PORR x P b - -
where:
~
Net CPM = The observed RM All count rate, in cpm, less background, or obtained from the calibration curve. VF - The vent fraction; that portion of the total plant gaseous release associated with this vent and discharge type. Value is equal to 0.5. PORR - The noble gas gamma emission Projected Dose Rate Ratio calculated in accordance with Section 1.3. This ratio is the actual projected dose rate divided by the allowable dose rate referenced in Section 1,3-1, relationship 1.4. 24.7 - The maximum pressure (psia) which RM All detector chamber should be ( Ujected to. This corresponds to a flow of 15 CFM from the release line to the vent. P - Pressure (psia) in RM-All at time of obtaining net CPM. Bkg - RM-All background count rate in cpm. L REVISION 13 . ,, .
. . _ . . , . . _, -m,.., _
I 1 1 Setpoint calcult. tion 1.4-5 l Plant Diedarge Line Itatitor (IN-L2) (Itstcit Type Relamaan) l 1NMODUCHCH. Follcwing completicn of the analyses Inquired by Sectico 1.2-4 an$ detamim-tien of mlease rates and wmuwtion limits in accortLvre with Sccticn 1.3-2, the monitor setpoint requires adjustment to ensure that alarm and pathway isolation occur if nuclide concentratico limits are exw-1 ME310001DGY Evaporator Csusrs. ate Storage Tank or Laundry and Shower Surrp 'Ihnk ocntents are circulated thrcugh radiation monitor IN-L2 and returned to the auxiliary buildirq sung to cbtain the actual courit rate at IN-L2 for the ocnoentration contained in the tank for release. '1he chmarved cxunt rate is adjustad for re-lease ficw, background and statistical counting variations, particular to this release ficw path, h resultian value is used as the alarnVtrip setpoint and PM-L2 is adjusted to this or a mcre conservative value prior to initiatin3 the release. If the ocncentratico of radionuclides to be released is less than the effluent monitor LID set "IC1/MIri" equal to 1 and derive " Net CIN" frun the calibration c'.trve by determinity the CEM which couwspanis to 3E-7 pC1/ml. CMDNCN IN-L2 Setpoint (CrM) = Net CTM x AP x (E + D) + Bkg + 3.3 YBkg (I C i/MPC1 ) x E where: Net CIN =
'Ibe observed IN-L2 count rate, in cpm, less s
back-<Jround, or obtained frun the calibration curve. AF = hininistration Factor to acocunt for error in setpoint determinaticn. AF = 0.8. EC3/MFCi =
'Ibe ratio of the actual gansna emittinJ whu-tions (excitdirq dissolved and entrained gases) of the tank contents to be released to the Mayi== Per-missihle Circahation (MPC) as listed in 10 CFR 20, Table II, Column 2 for unrestricted areas.
E =
'1he release flow rate of waste to be discharged in gallons per minute. A mavi== ficW rate of 100 gpn will be used for the Evaporator condensate Storage Tanks and 40 gpn for the Laundry and Shower Sunp Tanks.
D =
'Ihe dilutico flow frun the Nuclear Services Sec Water system in gallons per minute.
19q = PM-L2 background count rate in cpu.
- 3. 3VBkg =
A statistical spread cm the background count rate which represents a 99.95% confidence-level on monitor counting. 'Ihis factor is included to prevent inadver-( tent high/ trip alarms due to randcan counts on the nrmitor.
'~
REV SLON 12
_.-._...-_-___.__._.-...____._._.m.._ W irt * * % 1,4-4 A 'nzrbine Buildirq n'osamurt Discharge Line Mtmita (IM-L7)
- d. (Ctztiranaus Type peleasee)- .
ImmoucrnEt 4 1he activity released through the Turbine Ball mammarit Discharge Line ! Mcnitor IM-L7 is analyzed in accordance with on 1.2-5. The setpoint is a fixed ocricentraticrt Mami cs) worst case ruclide released at the worst case rate as described in the Methodology Section below. The monitor setpoint is i adjusted to ensure isolation of the ruleses pathway if ruclide w U.ution limits are aum==i=1 , MDIEXIIIXEf i The alarnvtrip setpoint determination is wa-i on the worst came assunption that I-131-is the only ruclide being this e===Tt.icn equates all counts on IM-L7 to I-131 with an MFC of 3 x 10' uci/ml. I-131 has the noot conservative MFC of the tv <:lides available to this release path and " visible"
- to IM-L7. The nietpoint A ased on assuring _1 MFC or less of I-131 in the discharge canal and is daarmined by deriving the cpn frta the QL7 calibraticm curve which oo . is is to a WWikutics) of 3 x-10 uci/ml.ar1d applying the flow dilutist factor, background _ocunts, and statistical countirq variations. The resulting value is used as the alarnVtrip setpoint and IM-L7 is
- adjusted to this or a note conservative value to maintain control on release .
--conditions. - d i
GIIIRJd'EE( CIN x (E + D) IM-L7 Satpoint (CEM) = E _ + Bkg + 3.3 V Bkg __ wter CFM - = 1he' counts per mirute u.w +41ng to 3 x-10~7 uci/ml (1 MPC I-131) frun the current IM-L7. calibration curve.
.E = 1he ==vi== release ficw rate of water able to be discharged in _gallculs per minute.
D- = The dilution flow frun the Nuclear Servit.es-Sea Water. system-in gallons par mirute. Bkg- = The background count rata at BM-L7 in cpn. 3.3V Bkg = A statistical spread on the birdvguand count rate which - sor_.- it.s a 99.95% oanfidence level on monitor counting. This factor is included to provent iradvertent high/ trip
- alarms due to rartican counts an the monitor.
REVISION 12.
Setpoint Calculaticm 1.4-7
'Nrbine tuiltiirq Bamament Discharge Line Mcnitor (IN-L7)
(Itatdt Type Ralenses) Irmmucnm
\
Followirg coupletion of the analyses required by Section 1.2-4 ard detennimtion of release rates and ccreentration limits in accordance with section 1.3-2, the acnitor setpoint requires adjustment to ensuru that alann and pathway isolation occur if nuclide v hation limits are awa-ki. ME310DGIXTL Station Drain Tank (SDP-1) contents are cirullated through radiation nonitor IM-L7 and returned to the sung to cbtain the actual count rate at TN-L7 for the u.nn.= hation cantained in Km tank for release. 'Ihe observed cn2nt rate is adjusted for release flow, background and statistical countirq variations, particular to this release flos path. 'Ihe resultirq value is used as the alanq/ trip setpoint and IN-L7 ja adjusted to this or a r. ore conservative value prior to initiatirg the release. If tM canoenLcaticn of radionuclides to be released is loss than the efi'luent mar., Jr LID set "DC1/Mici" equal to 1 ard derive " Net CIM" from the calibration curve by determinity the CIN whidi conasF= is to 3E-7 4C1/ml.
.CAIEULATI W IN-L7 Setpoint -(CEM) = + Bkg + 3. 3VBkg (E C1 / Ei) x E
~ . .
where: Net CIN =
'Ihe observed IN-L7 count rate, in cpa, less backgreurd.
AF = kaninistraticn Factor to account for error in setpoint determination. AF = 0.8. Iri/MPCi =
'Ibe ratio of the actual ganna emittiry ccncen-trstions (emludirq dissolved and entrained gases) of the tank contents to be released to the Maxitnum Permissible crsmhation (MPC) as listed in 10 CFR 20, Table n , Cblunn 2 for unrestricted areas.
E =
'Ihe release flow rate of waste to be discharged in gallons per minute. A maxinun flow rate of 600 gpm will be used.
D =
'Ihe diluticn flow frun the Nuclear Services Sea Water systan in gallons per minute.
Bkg = IN-L7 background ocunt rate in cpu. 3.3YBkg = A statistical spread on the backgound count rate which represents a 99.95% confidence level on monitor counting 'Ihis factor is incltried to prevent imdvertent high/ trip alarms due to rarden cnznts on the mcnitor.
\_
. . . . -~. .. .. .- ;
1EVIS.C N 12 wi. ,,s i
I
' OELETE0
- O . 25 -
l REVISION 13 L
~
_ ._-_ _ _ . _ . ~_ . - - - . . _. _ . . ._ - __..__.m . _ . _ _ . _ _ _ _ CALCULATION OF INHALATION PATHWAY DOSE FACTOR (Pj) Pt - K' {BR) DFAt mrem / year per uC1/m3 where: i K' - A constant unit of conversion - 106 pCl/uti BR The Breathing F, ate of the child age group - 3700 m3/ year DFAj - The maximum organ inhalation dose factor for the child age group for the ith radionuclide, in mrem /pti. The total body is conridered as an organ in the selection of DFA. NOTE: For the inhalation pathway Pj - Rj, so values of Pj may be taken from Table 4.4-3. O, 4 w* A
+
References:
l
- 1) NUREG-0133, Section 5.2.1.1
- 2) Regulatory Guide 1.109, Table E 5, and Table E 9
-l l
,F4 ' V 3EVIS10N 13 . ze . g e c ,- ,- - g w- +--y -- v v
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.j l DELETED O REV s ON 13 - -- . _ . _ _
. . , _ . _ - . _ _ _ _ . - ~ - . . _ -. . _ _ - -.- - --- - - .- _ _ _ _ _ ___ _ -
l 4 s I r-i i A r i i o i SECTION 2.0
.. 1
.~.. , '
9 RADICACT!YE eft'LUENTS , L uo$E REDUCTION SPEC!PICAT10N5 i t
-i
?
i ~ REVIS?OM "o" ;
- 28 - :
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- . '(N V-
- i. i 1
TABLE.E RADWASTE REDUCTION SYSTFM - DOSE PROJECTION 4 DO$E PROJECTION PROJECTm FLOW
, SYSTEM . SPECIFICATION CALCULATION FREQUENCY- DIAGRAM i.
t d Caste 2.1-1 2.2-l ' - M* 2.3-1 i' Gas
- Treatme nt 4i Ventillation 2.1-1 2.2-1 M* 2.3-1 j . Exhaust '
- Treatment i
Liquid 2.1-2 2.2-1 M' 2.3-2
. Radwaste Treatment
, ^. )
[O [ .*) 2-
- When a Radwaste Reduction System is not available for use..
{'.) J- --- O i.;)
TASTE REDUCTION SPECFICATION NO. 2.1 1 The WASTE GAS SYSTEM shall be used, as required, to reduce the radioactivity of , materials in gueous waste prior to discharge, when projected monthly air doses due to releases of gaseous effluents from the site to areas at or beyond the SITE BOUNDARY would exceed:
- 1) 0.2 mrad gamma / month *
- 2) 0.4 mrad beta / month
- AND The VENTILATION EXHAUST TREATV!.NT SYSTEM shall be used, as required, to reduce the quantity of radioactive materials in gaseous waste prior to discharge, when projected monthly air doses due to release of gaseous effluents from the site to areas at or beyond the SITE BOUNDARY would exceed:- ,
- 1) 0.3 mrem to any organ / month
- Doses due to gaseous releases from the site shall be projected at least once pet 31 days. .
-.~. .
V. 1 l l [ The !!mits of the 10CFR30, Appendix 1, paragraph B1 criteria were reduced to 1/4 of the monthly portion of the annual limit as explained in correspondence among AIF, Utilities and the NRC dated December 24,1981. l l i --
References:
- 1) Grystal-Rivec Unit-3-Technical 4pecifIcation-3.7,13.3--
- 2) - Plant Procedures O 3) Correspondence C.A. Willis (NRC) to S. Pandy (Franklin Research Center) dated 11/20/51 and AIF letter to AIF subcommittee on RETS dated 12/24/81.
4
1 i l l
) TASTE REDUCTION SPECIFICAT10N No. 2.1-2 The LIQUID RADWASTE TREATMENT SYSTEM shall be used, as required, to reduce !
radioactive materials in liquid wastes prior to their dierge;, when projected monthly ' doses due to liquid effluents discharged to UNRESTRICTED AREAS would exceed the ' following valuest
- a. 0.06 mretti whole body / month *
- b. 0.2 mrem to eny organ / month
- t Doses due to liquid releases shall be projected at least once per 31 days.
- w. .
O & I The limits of the 10CFR30, Appendix 1, paragraph A criteria were reduced to 1/4 of the monthly portion of the annual limit as explained in correspondence among AIF, Utilities and the NRC dated 12/24/81. Referencest
- 1) Crystal RiverttnittTechnical-Specification-3.7,13.2-
- 2) Plant Procedures s 3) Correspondence C.A. Willis (NRC) to S. Pandy (Franklin Research Center) dated i-11/20/81 and AIF letter to AIF subcommittee on RETS dated 12/24/81.
e ,,( (< ^ te dI *\ %
_ . _ - . _ _ _ . ..~.-________._._.-_.._._.._...___-.___________.._m._._.__ DOSE PROJECTION PIETHODOLOCY 2.2-1 CASE 005 RADWASTC i
!. IlfTRODUCI1QL Crystal River Unit 3 operating practices require use of the waste gas system (Waste Gas Decay Tanks). The normal release paths for gaseous etfluents are via ventilation exhaust treatment systems (HEPA and Charcoal Tilters). The operability of the ventilation exhaust treatment systems is controlled by t he -- Cr Unit Technical Specifications. A.6m M s/ A # 7 'I ^ ystal- N '" - River As long as these practices and specifications are maintained, the A E > i radwaste reduction requirements of Crystal-River--Unit-3-Technical .E /a. J A Specificat-fun-3r'Ir13r3 are set, and there is no need to project doses prior to the release of gaseous radwaste.
!!. CALCUIATIONS Dose projection calculations will be necessary if either systen is not available for use, 4
31D e Dp= 3DQ where: O Dp = Projected Dose (monthly). De
=
Current quarter cumulative dose, including projection for I release under ev duM ion. NDQ = Number of days into quarter, where the quarterly periods are: January 1 thrcugh March 31, yril 1 through June 30, July 1 through September 31, October 1 through December 31.
References:
i
- 1) 1 S--h 6 r4 r2 *,-3 rh 871.
- 2) FSAR 5.5.1, 5.5.2
.1 LO L
- 32 L REWSL0i\ l0 .
- , , , - . . -,-.,.. ..,.e-r=+n--y,e---, .---v. ,.w. - -- ,---- m , , - - ,..,-e . - , - - - '
DOSE PROJECTION ME750DOLOGY 2.2-2 LIQUID RADWASTE I. DrftoDocTIGE Crystal River Unit 3 operating practices require liquid radwastes (except for Laundry and Shower Sump waste and Secondary Drain Tank waste) to be processed prior to releasing then to the environment.
.c a al,,,2 3 d MI J s ercirt As long as these(practices are maintained the radwaste reduction requirements of / crystal- Alver-Unit-3-Techaical-specification-3:7.13r2 are net, and there is no need to project doses prior to the release of liquid radwaste.
- 11. CAIGLkf1(ES .
Dose projection calculations will be necessary if there is a malfunction of 11guld radwaste treatment systes equipment and liquid radwaste must be released without prior treatment. 31De Dp= NDQ where Op = Projected Dose (monthly). a De Current quarter cumulative dose, including projection for release under evaluation. NDQ = Number of days into quarter, wher$ the quarterly periods are: January 1 through March 31, April 1 through June 30 July 1 through September 31, October i through December 31.
References:
7 opcm is,1 I, ,% f d .3+A3 0 REVIS.ON 10
. - . . _. - - .- - .- ~ ..- - - - - .-_
-_ .. ._- - ~.- - - - . - _ - - . - - -~
'IUTAL COSE 6MICIFICATIGi 2.3 (LIGHD AND CWunm HEurAsrs)
'Ihm calendar year dose or done cr==nitsmant to any menbar of the public, due to Islaamam of radicactivity and radiation frun uranium foal cycle ocurces shall be limitad to less than or equal to 25 anos to the Wole bcxty or any organ, (except the thyroid Wich shall be 11mitad to less than or equal to 75 mriens).
'Ihis specificaticri is satisfied by meetirq specificatlans 4.1-1, 4.1-2, !
and 4.1-3. If h exceed twice the limits of specificaticris 4.1-1, 4.1-2, ard 4.1-3 then an analysis shall be performed to confirm contirued ocepliance with 40CrR190(b) .
.~..
References:
- 1) . Technical-Specification-3 r1173- + W A J I ; R A # #
- 2) Plant Prtrwhtras
- 3) 40 CFR 190 O !
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7 d THIS PAGE Ltry ggggg t l O iiEVS ON 10
O' . k SECTION 3.0 - f RADIOACTIVE EFFLUENTS
]
SAMPLING $PECIFICATIONS- ! i e
.i l
l O REV SiON "O" 38 - W $ W 'r@$99 &m'New-WM'N ye$ '99'e1'?t*W-4r-
*hP-=e**--V- 'y*W Pe'k*TM7+PP'g e'%
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TABLE 10 [} v GASEOUS AND LIQUID EFFLUENT REPRESENTATIVE SAMPLING RELEASE TYPE REPRESENTATIVE SOURCE OF SAMPLING EFFLUENT BATCH CONT. METHOD Evaporator X 3.11 Condensate Storage Tanks Laundry and X 3.11 Shower Sump Tanks Secondary X X 3.1 1, 3.1 2 Drain Tanks Plant Condensate X 3.12
~ . .
Waste Gas X 3.13 Decay Tanks Reactor Bldg. X X 3.14 Purge Exhaust Auxiliary Bldg. X 3.14
& Fuel Handling Area Purge Exhaust Reactor Bldg. X 3.13 with Both Personnel and Equipmer't Hatches Open u,
.,,. REV S 0 \ 5
Repreheatative Sampling feethod leo. 3.1-1 (Evaporator Condensate Storage Tanks, Laundry & Shower Sump Tanks, Secondary Drain Tank) O . To obtain representative samples from these tanks, the contents of the tank to be sampled will be recirculated through two contained volumes and a grab sample will be collected upon completion. No additions of liquid waste will ' be made to this tank until completion of the release. Representative Sampling Method No. 3.1-2 (Secondary Drain Tank and/or Plant Condensate) A representative sample may be obtained via grab sample of the Turbine Building Sump or the Secondary Drain Tank, Plant Condensate, or from the release compositor. Representative Sampling Method No. 3.1-3 (Waste Gas Decay Tank) T 'liepresentative gas, lodine, and particulate samples are drawn from the waste
~
gas decay tank staple lines. No additions of waste gas is allowed into a tank following sampling until the release has been completed. Representative Sampling Method No. 3.1-4 (Reactor Building & Auxiliary Building & Fuel Handling Area EIhaust) Representative gas, iodine, particulate and tritius _ samples are taken from these ' ducts at the location of the radiation monitors. The - sample for the Reactor Building purge Duct is taken from radiation monitor RM-A6 prior to a purge and is drawn from radiation monitor-~ RM-A1 during a purge. - The sample for the Auxiliary Building and Fuel Handling Area Exhaust Duct is drawn f gon RM-A2 during venting since this-is a continuous release pathway. If samples cannot be obtained from tho ducts of the Reactor or Auxiliary Building, samples can be - obtained from areas of these buildings that - are considered to be representative of the : radionuclide concentrations present throughout the respective -buildings. Sampling times and volumes should be established to assure the LLD Limits of Sections 1.2 and 4.2 for the radionuclides can be set. O 40 REV S CN 9 !
Re resentative Sampling Method No. 3.1-5 (Reactor But ding With Personnel And Equipment Hatch Opened) , The following conditions w .t be satisfied prior to having the RB personnel hatch and equipment hatchn --ad at the same time: o The Reactor Building purge exhaust fans are operational and the make up fans are shut down; o The initial purge must run for at least 2 days at 30,000 SCFM. o The Reactor Coolarit System is degassed and depressurized; o The Reactor Building recirculation system is continuously monitored by RM A6 (all channels); or air samples are taken daily on each elevation of the Reactor Butiding; o A particulate sampler is installed and operating on the Reactor Building refueling floor; o Air s3mples are taken as jobs in the Reactor Building necessitates (i.e., jobs that risk increasing the particulate, iodine, or gaseous concentrations of radionuclides in the Reactor Building); If the purge exhaust fans must be shut down, then either the personnel hatch or equipment hatch openings must be closed (e.g., temporary door installed in the personnel hatch). Representative Sampling Method No. 3.1 6 (Reactor Building During Integrated Leak Rate Test) Due to building overpressure, prepurge samples cannot be taken trom RM A6. Representative gas, todine, particulate and tritium samples may be obtained from the Intermediate Building-containment sampling apparatus or the Post Accident Sampling System.
Reference:
- Telecon fPC (Dan Green, Dan Wilder) to NRC (Charles Willis) 1ated 03/15/85 at 0930;
Subject:
Personnel and Equipment Hatch Openings. O o " '~ R:V S{N 13
9 SECTION 4.0 .
.m, .
RADIOACTIVE EFFLUENTS DOSE CALCULATION SPECIFICATIONS O
. . . . ., . n .o , r. ,
h $&h Y i l l
i O O I TABLE IV CuanULATIVE DOSE CALCULATION - DOSE NUCLIDE CALCULATION DOSE PATHWAY SPECIFICATK2N ANALYSIS METHOOOLOGY FACTORS Noble Cases 4.1-1 4.2-1,4.22 4.3-1 4.4-1 4.'2-3 Radiciodines, Radimetive Particulates 4.1-2 4.2-1, 4.2-2 4.3-2 4.4-2 to 4.4-16 Radiortuclides 4.2-3 other than Noble Gases .. Liquid Ef fluents 4.1-3 4.2-4, 4.2-5 4.3-3 4.4-17 s U.. M
~
4 r C) c-1: U i O: i
.. j s
Q Q DOSE SPECIFICATION 4.1-1 (NOBLE GASES) . The air dose at or beyond the SITE BOUNDARY due to radioactive noble gases released in gaseous elfluents shall be liralted as follows:
- 1) During any calendar quarter, d 5 mrad gamma, and 410 mrad beta radiation.
- 2) During any calendar year, 410 mrad gamma, and 6 to mrad beta radiation.
Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined at least once per 31 days.
,, ~ 't t
References:
- t. erscm A>.tI, Geh~ :y J; i $ 0 N gg a l
C r y s t al- R iver- U ni t T echn ical- S p eci f ica t ion -3.1 1,2. 2
DOSE SPECIFICATION 4.1-2 (RADIOIODINE & PARTICULATES) The dose to a MEMBER OF THE PUBLIC from lodine-131, Tritium and radioactive particulates with half lives of greater than 8 days in gaseous ef'luents released from the site to areas at or beyond the SITE BOUNDARY shall be limited as follows:
- 1) During any calendar quarter, s 7,3 mrem to any organ.
- 2) During any calendar year, s 13 mrem to any organ.
Cumulative dese calculations for the current calendar quarter and current calendar year shall be determined at least once per 31 days. O
==*
References i es:n A J1> % % , Cry 5tal-River-4) nit-3-Technical-Specif tearton-3.11.2J q . II " ? ?
t .1 4 i DOSE SPECIFICATION 4.13 i O* * (LIQUID EFFLUENTS) ,i i The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials : in liquid effluents released to UNRESTRICTED AREAS shall be limited as follows: [
- 1) During any calendar quarter, s 1.3 mrem total body. !
- 2) During any calendar quarter, s 3 mrem any organ.
- 3) During any calendar year, s 3 mrem total body.
- 4) During any calendar year, s 10 mrem any organ.
Cumulative dose contributions frot.. liquid effluents shall be determined at least once per ! 31 days. f O o J
- Referencesi-
# - , s. *" '
Grystal River-Unit + Technical 4peelfica tion-3d l.la R*V'S"O" 'N l'0"
DOSE SPECIFICATION 4.1-4 (RADIOACTIVE EFFLUENT RELEASE REPORT) , A Semiannual Radioactive Effluent Release Repcrt covering the operatien of the Unit during the previous six months of operation shall be submitted within 60 days after January i and July 1 of each year. The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the hypothetical worst case individual from releases (including doses from primary effluent pathways and direct radiation) for the previous calendar year. The assessment of radiation doses shall be performed in accordance with the off-site Dose calculation Manual (oDcM .
-.~..
O
-45a- wein 16.- V , dl u 4 e6
NUCLIDE ANALYSIS 4.2-1 REACTOR BUILDING PURGE EIMAUST NOCLIDE EAMPLE SOURCE LLD(b)(uci/al) A. Principal Gamma Esitters (a) Mn-54 1x10-4/1x10-11 Fe-59 1x10-4/1x10-11 Co-58 Batch release particulate filter 1x10-4/1x10-11 Co-60 for Batch Releases. Jeekly 1x10-4/1x10-11 Zn-65 - Particulate Filter Analysis for - 1x10*4/1x10-11 Mo-99 continuous (c) type release. 1x10*4/1x10-11 C4-134 1x10-4/1x10-11 Cs-137 1x10-4/1x10-11 Ce-141 1x10-4/1x10-11 Ce-144 1x10-4/1x10-11 Kr-87 Pre-release grab sample for Batch 1x10-4 Kr-88 type release. Weekly grab sample 1x10-4 Xe-133 - for continuouc type release. - 1x10-4 , Xe-133m 1x10-4 Xe-135 1x10-4 Xe-138 1x10-4 B.' C Iodine 131 Batch release charcoal filter for NA/1 x 10-12 \ Batch Releases. Weekly charcoal filter for continuous releases, C. Tritium Pre-release Grab Sample. 1x10-6 D. Gross Alpha Monthly Particulate Filter Composite 1x10-11 E. Sr-89 Quarterly Particulate Filter Composite 1x10-11 T. Sr-90' Quarterly Particulate Filter Composite 1x10-11 (a) Other identified Gamma Emitters not listed in this table shall be included in dose calculations. (b) The_first value refers to the LLD for pre-release grab sample; the second value refers to the LLD for weekly Particulate Filter Analysis. (c) Reactor Building Purge i's considered continuous after minisua of one Reactor Building voltaes have been released on a continuous basis (i.e., first one volume is a batch type). O
- 4. . REVIS 0 \ 11
. _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ - _ _ _ . _ . __.. _ -- ________,m_ . . . - _ _ . _ _ _ . . _ _ _ _ _ _
NUCLIDE ANALYSIS 4.2-2 AUI!LIARY BUILDING AND FUEL *.NDLING AREA EXHAUST NOCLIDE SAMPLE SOURCE LLD(b)( UCi/al) . A. Principal Ganaa Enitters (a) Mn-54 Fe-59 1x10-4/1x10-11 Co 1x10-4/1x10-li Weekly Particulate Filter Analysis. 1x10-4/1x10-11 Co-60 Zn-65 - 1x10-4/1x10-11 Mo-99 1x10 /1x10-11 Cs-134 1x10 /1x10-11 Cs-137 1x10 /1x10-11 Ce-141 1x10 /1x10-11 Ce-144, 1x10* /1x10-11 1x10-4/1x10-1' Kr-87 Monthly Grab Sample. 1x10*4 Kr-88 1x10-4 Xe-133 - 1x10-4 Xe-133a 1x10-4 Xe-135 1x10-4 Xe-138 ix10-4
'B'i' Iodine 131 Weekly Charcoal Filter Analysis.
O 1x10-12 C. Tritium Monthly Grab Sample. 1x10-6
- 0. Gross Alpha Monthly Particulate Filter Composite 1x10-11 -
E. St-89 Quarterly Particulate Filter Composite 1x10-11 ! F. 'Sr-90 Quarterly Particulate Filter Composite 1x10-11 (a) Other in dose identified calculations.Gamma Enitters not listed in this table shall be included (b) _The first value refers to the LLD for pre-release grab sample the second
.value refers to the LLD for weekly Particulate Filter Analysis.
O 4, . REVS {N ll
. = . . . - . . . . .- - - - .
.______7...___.-_._
h NUCLIDE ANALYSIS 4.2-3 WASTE Gas DECAT YANKS NOCLIDE SAMPLE SOURCE LLD(b){ UCi/al) A. Principal canna Enitters (*)
~~
Mn-54 Fe-59 1x10 /1x10-11 Co-58 1x10 /1x10-11 Co-60 1x10* /1x10-11 In-65 1x10 /1x10-11 Mo-99 1x10 /1x10-11 Weekly Particulate Filter sample (from RM-A2) 1x10 /1x10-11 Cs-134 Cs-137 1x10* /1x10-11 Co-141 1x10' /1x10-11
- 1x10* /1x10-11
~
Ce-144 1x10*4 Kr-87 1x10~4 Kr-88 1x10"4 Xe-133 - Pre-release Grab sample Xe-133a 1x10'4 - Xe-135 1x10*4 1x10"4 Xe-138 1x10*4 .!
- 8. Iodine 131 Weekly Charcoal Tilter (from RM-A2) 1x10*12 (al- other-identified dose and setpoint Gamma calculations. Enitters not listed in this table shall be included-in (b) The first value refers to the LLD for. pre-release 9L sample the second value refers to the LLO for weekly Particulate Filter Analysis.
O l l _ 48 REV S ON 11
. , . . . _ . . . - _ _ _ _ . . _ _ _ _ . . , - , _ _ . _ _ _ _ _ _ _ ~ _ - _ ,
NUCLIDE ANALYSI5 4.2-4 EVAPORATOR CONDENSATE STORAGE TANK 5, LAUNDRY AND.5HOWER SUMP TANKS, SECONDARY DRAIN TANK NUCLIDE SAMPLE SOURCE LLD(uCL/ml) ._ A. Principal Gamma Emitters (a) Mn-34 3x10 7 Fe-$9 3x10 7 i Co-58 3x10 7 Co-40
$x10-7 n-65 Pre release Grab Sample
, 5x10-7
~
Mo-99 3x10-7 Cel34 $x10-7 Cul37 3x10-7 Ce-141 3x10-7 Ce-14 t. 3x10-7 B._ lodine 131 Pre-Release Grab Sample lx10-6 C. Dissolved and " Intrained Noble O Gases Monthly Grab Sample 1x10-5 D. Tritium Monthly Composite lx10-5 1
- E2 Gross Alpha Monthly Composite Ix10 7
- F. Sr-59 Quarterly Composite 5x10 8 C. Sr-90. Quarterly Composite 5x10-8 H. Fe-55 Quarterly Composite Ix10-6 :
l l (c) Other identified Gamma Emitters not listed in this table shall be included In' dose calculat O 1
"::V, S 0\ 7 1
NUCLIDE ANALYSI5 4.2-3 t ' SECONDARY DRAIN TANK AND/OR ; PLANT CONDENSATE ' O MUCLIDE i SAMPLE SOURCE I,LD(uCl/ml) A. Principal Gamma Emitters (a) ! Mn34 3x10-7 r 1 e 59 3x10 7 ! Co-38 3x10 7 Co-60 3x10 7 t In-6$ Weekly Composite
$x10-7 Mo-99 3x10 7 Cs-134 3x10 7 Cs-137 3x10 7 Ce 141 3x10 7 Ce-144
~~ 3x10-7 B. todine 131 Weekly Composite c
Ix10-6 C. Dissolved and O Entralned Noble Gates Monthly Grab Sample , lx10-) D. Tritium Monthly Composite lx10-5 E. Gross Alpha Monthly Composite Ix10-7 F.. Sr-89 Quarterly Composite 5x10-8 G. Sr Quarterly Composite Sx10-8
- H. Fe-5) - Quarterly Composite Isl0-6 ,
(a) Other identUled Gamma Emitters not listed in this table shall be included in dose calcu O REV SIN 7
i
-l l
i DOSE CALCU!.ATION 4.3*1 (NOBLE CAS) The air dose at or beyond (ne SITE BOUNDARY due to noble gases released in gaseous effluents is calculated as follows: D1 = 3.17 x 10-8 E Mi (X/0)01 arad Dg = 3.17 x 10-8 E Ni(X/0)01 nrad where D1 = The air dose at or beyond the SITE BOUNDARY due to ganna . emissions from noble gases in gaseous effluents in arad/ time ' period. Dp = The air dose at or beyond the SITE BOUNDARY due to beta emissions from noble gases in gaseous effluents in arad/ time period. ; i 3.17 x 10-8 = The number of years in one second, yr/sec.
=
Mi The air dose f actor due to ganna emissions for each identi- . fled noble gas radionuclide, in arad/ year per uCi/m3
=
Ni The air dose factor due to beta emissions for each identi- ' fled noble gas radionuclide, in arad/ year per uCi/m3
=
K/Q. The highest calculated annual average relative concentration for areas at or beyond the UNRESTRICTED AREA Boundary, 2.5 x . 10-6 sec/m3 g = Total uCi of isotope i released during the calendar quarter or calendar year, as appropriate.
.c .
9 R O
' ~
REV'SION 10
- - - _ - . . - . - ~ _ -
i O DOSE CALCULATION 4.3 2 (RADl0100lNES&PARTICULATES) ,, s the dose to an-individual at or beyond the SITE BOUNDARY (,vc so lodine 131,
- Tritium and radioactive particulates with half lives of greatar than 8 days is calculated as follows:
0 3.17 x 10 8 I WRjQt ; where: 0 = The radiation dose to an individual at or beyond the UNRESTRICTED AREA BOUNDARY, in mrem.
=
Rt The dose factor for each identified radionuclide m2(mrem / year)peruCi/seeororem/yearperuCi/m$.1,in W = X/Q for inhalatior, boundary and 7.5 x patgway, 10- 2 5 x 10 6 sec/m3 at the site l sec/mi at the critical receptor. ' W = thway, 1.9 x 10*8m 2 at 0/Qforfoodandgroundplanesgm2atthecritical the site boundary and 5.7 x 10-receptor. ~;
=
Qt Total 901 of isotope i released during the calenaar quarter or calendar year, as appropriate. 3.17 x 10 8 . The number sf years in one second, yr/sec.
Reference:
NUREG 0133, Section 5.3.1 ' FSAR, Table 2 20 1 O REVISIC N 13
. 52
DOSE-CALCULATION 4.3 3 (LIQUID EFFLUENTS) The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS is calculated as : follows: DI ArItCk i ik Fk i k where: 0 The cumulative dose ccmmitment to the total body or any organ, r, from the liquid effluents for the total time period I tk in mrem. l tk - The length of the kth time period over which C ik 15 ' averaged for all liquid releases, in hours. : Cik - The average concentration of radionuclide, 1, in undiluted itquid effluent during time period tk from any liquid release, in gCi/ml. t Ari - The site related injestion dose commitment factor to the j total body or any organ for each identified principal gamma and beta emitter as shown in Table 4.417 of this manual, in mrem /hr per pC1/ml.
... lH !
Fk Waste release flowrate + (Waste flow rate + Dilution flow rate)
~
Oilution flowrate is equal to the total circulating water flow and/or Nuclear Services seawater flow av:ilable during the release. .
References:
1)-NUREG0133.Section4.3.
'O 2) Telecon/ Meeting Summary with C. Willis (USNRC) dated 01/16/85 regarding F-k REV SION 13 - >> -
1 DOSE CALCULATION 4.3-4 (RADIOACTIVE EFFLUENT RELEASE REPORT) The individual and population _ doses are calculated using Gl. SPAR I (for gaseous effluents) and LADTAP (for liquid effluents) computer codes obtained f rom the Nuclear Regulatory Commiasion, and are i revised to' include site specific data whenever possible. Both computer codes incorporate the calculational models and parameters documented in Re<3ula .ory Guide 1,109. Direct radiation doses are taken from the Plume Immersion Pathway calculated by GASPAR. Meteorological inly . lata consisting of aserage relative concen-trations- (X/Q's) ard average relative deposition values (D/Q's) are provided by- coi:pling GASPAR with the Nuclear Regulatory Commission computer code XOQDOQ (NUREG-0324, " Program for the ' Meteorological Evaluation of Routine Effluent P.eleases at Nuclear Power Stations"), The summations of gaseous and liquid ef fluents - and solid waste shipments listed in this Report are in accordance with the tables in Regulatory Guide 1.21*(Rev. 1, 6/74), The summation of solid radioactive waste is derived from Radio-active Shipment Records and reported in accordance with Technical
Specification Section 6.9.1,5(d).
4 6 O
.,,,_ IVLSL\ 7
. , _ . - . . - - - - . - .---. - . - . - . - - - - - . . ~ . . - - , ~ . . - -
g TABLE 4.4-1 o DOSE FACTORS FOR EXPOSURE TO' A SEMI-INFINITE CLOUD OF NOBLE GASES Ni Li Mi gl Nucilde D-Air * (DFl, ) - o-Skin * * (DFSI ) v- Air * (DFI') v. Body * * (DFBI ) Kr-83m 2.88E + 2 -- 1.93E + 1 7.56E - 2 Kr-85m 1.97E+3 1.46E + 3 1.23E + 3 1.17E + 3
. Kr-85 1.95E+3 1.34E + 3 1.72E + 1 1.61E + 1 a Kr-87 1.03E+4 -9.73E + 3 - 6.17E + 3 5.92E + 3 Kr-88 2.93E+3 2.37E + 3 1.52E + 4 1.47E + 4 Kr 1.06E + 6 - 1.01 E + 4 1.73E + 4 1.66E + 4 Kr-90 7.83E+3 7.29E + 3 1.63E + 4 1.56E + 4 Xe-131m 1.IIE+3 4.76E + 2 1.56E + 2 9.15E + 1 Xe-133m 1.48E+3 9.94E + 2 3.27E + 2 2.51E + 2
- Xe-133 1.05E+3 3.06E + 2 3.53E + 2 2.94E + 2 LXed3fm -
O Xe-135 7.39E +2 2.46 E + 3 ' 7.l l E + 2 l.86E + 3 3.36E + 3 1.92E + 3 3.12E + 3 1.81E + 3 Xe-137 1.27E+4 1.22E + 4 1.51E + 3 1.42E + 3 Xe-138 - 4.75 E+3 - - 4.13E + 3 9.21E + 3 d.83E + 3 - Ar-41 = 3.28E +3 2.69E + 3 9.30E + 3
- 8.84E + 3 -
- mrad-m3 --
uCi-yr
** mrem-m3 wCl-yr
References:
, 1) NUREG 0133
- 2) USNRC Regulatory Guide 1.109, Table B-1 n
mJr--,SsN.
. 7 a .- - -- -
uALCULATION OF INHAl.ATION FATHWAY DOSE FACTOR1(9 ) , Rj - K' (BR) DFAt mrem / year per uC1/m3 Where: K' - A constant unit of conversion - 106 pCi/uci BR - The Breathing Rate of the represented age group: ' 1400 m3/yr - infant 3700 m /yr - child 8000 m3/yr - teen 8000 m3/yr - adult DFAj - The maximum organ inhalation dose factor for the represented age group for the ith radionuclide, in mrem /pci.
~~~. -
V
References:
- 1) NUREG 0133, Section 5.3.1.1
- 2) Regulatory Guide 1.109, Table E-5, and Tables E-7 through E-10 i
O REVISION 13 - -
TAC 1.E 4.4-2 Inhalation Dose Fa.: tors - Infant
.4uclide ' Bone Liver T. Body Thyroid Kidney M 3Gy-H-3 ND 6.47 E2 6.47 E2 6.47E2 6.47E2 6.47E2 6.47E2 Cr-51 -ND ND 8.95El 1.32El 1.32El 1.28E4 3.57 E2 Mn-54 ND 2.53E4 4.98E3 4.98E3 4.98E3 9.95E5 7.06E3 Fe-55 1.97E4 1.17E4 3.33E3 ND ND 8.69E4 1.09E3 Fe-59 1.36E4 2.35E4 9.48E3 ND ND 1.02E6 2.48E4 Co-58 ND 1.22E3 l~82E3 ND ND 7.77E5 1.11E4 Co-60 ND 8.02E3 1.18E4 ND ND 4.51 E6 3.19E4. 3 N1-63 3.39E5 2.04E4 1.16E4 ND ND 2.09E5 2.42E3 In-65 1.93E4 l 6.26E4 3.l lE4 ND 3.25E4 6.47E5 5.14E4 Rb-86 ND 1.90E5 8.82E4 ND ND ND 3.04 E3 Sr 3.98E5 ND 1.14E4 ND ND 2.03E6 6.40E4 Sr-90 4.09E7 ND 2.59E6 ND ND 1.12E7 1.31E5 Y-91 5.88E5 ND 1.57 E4 - ND ND 2.45E6 7.07E4 Zr-95 - 1.15E5 2.79E4 2.03E4 ND 3.ll E 4 1.75E6 2.17E4 Nb-95 1.57E4 6.43E3 3.78E3 ND 4.72E3 4.79E5 1.27E4 u-103 2.02 E3 ND 6.79E2 ND 4.24E3 5.52E5 1.61E4 du-106 8.68E4 ND 1.0SE4 ND 1.07E3 1.16E7 1.64 E5 Ag-110m . 9.98E3 7.22E3 5.00E3 ND ~
1.09E4 3.67E6 3.30E4 To-125m 4.74 E3 1.99E3 6.58E2 1.62E3 ND 4.47E3 1.29E4 Te-127m 1.67E4 - 6.90E3 2.07E3 4.87E3 3.75E4 1.31E6 2.73 E4 -
= Tc-129m 1.41E4 6.09E3 - 2.23E3 5.47E3 3.18E4 1.68E6 6.90E4 !-131 - 3.79E4 - - 4.44 E4 1.96E4 1.48E7 5.18E4 ND 1.06E3 - Cs-134 3.96E5 7.03E5 ' 7.45E4 ND 1.90E5 7.97E4 l.33 E3 -
Cs-136 -4.83E4 1.35E5 - 3.29E4 ND 5.64E4 ' l.lSE4 1.43E3 Cs-137 5.49E5 6.12E5 4.55E4 - ND 1.72E5 7.13E4 1.33t 3 B:-l40- -5.60E4 5.60El 2.90E3 ND .1.34 El 1.60E6 - 3.84E4 Ce-141 2.77E4 1.67E4 1.99E3 ND 5.25E3 5.17E3 2.16E4
' Cc-144 3.19E6 1.21E6 1.76E5 ND 5.38E5 9.84E6 1.48E5 'Pr-143 1.40E4 5.24E3 - 6.99E2 ND 1.97E3 4.33E5 - 3.72E4 '
Nd-147 . 7.94E3 S.13E3 5.00E2 ND 3.15E3 3.22E5 3.12E4 O
. , , . RD/lS10N "O" l
- y r --, - r ,--
_. .. _ __ _ _ _ = _ < _ - . _ . . _ - . . _ _ __ _ _ ___. _ .. _._._ ___.__.,._.._ t
,i TABLE 4.4-3 inhalation Done Factors - Child Nuclide Bone Liver T. Body = Thyroid Kidney Lung Cl-LLI H-3 ND ^1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 Cr-31 ND ND 1.34 E2 8.33El 2.43El 1.70E4 1.08E3 Mr.-34 ND 4.29E4 9.31 E3 ND 1.00E4 1.38E6 2.29E4 +
- Fe-33 4.74E4 - 2.32E4 7.77E3 ND ND 1.llE3 2.87E3 Fe 39 2.07E4 3.34 E4 1.67E4 ND ND l.27E6 7.07E4 Co ND 1.77E3 3.16E3 ND ND 1llE6 3.44 E 4 Co-60 ND 1.3tE4 2.26E4 ND ND 7.07E6 9.62E4 NI-63 8.21 E3 4.63E4 2.80E4 ND ND 2.73E3 6.33E3 Zn-63 4.26E4 1.13E3 7.03E4 ND~ 7.14E4 9.93E3 1.63E4 Rb-86 ND 'l.98E3 1.14E3 ND ,ND ND 7.99E3 .
Sr-89 3.99E3 ND- 1.72E4 ND ND - 2.16E6 1.67E3 Sr-90 1.01E8 ND 6.44 E 6 ' ND ND 1.48E7 3.43E3 Y-91 9.14E3 ND 2.44 E4 ND ND 2.63E6 - 1.84E3 Er-93 1.90E3 4.18E4 3.70E4 ND- 3.96E4 2.23E6 6.ll E4 Nb-93 2.33E4 9.18E3 6.33E3 ND 3.62E3 6.14E3 3.70E4 .
>103~ 2.79E3 ND 1.07E3 ND 7.03E3 6.62E3 4.48E4 Ru-106 1.36E3 ~ ND 1.69E4 ND 1.84E3 1.43E7 4.29E3 Ag-!!0m 1.69E4 1.14E4; 9.14E3 ND 2,12E4 3.48E6 1.00E3 Te-123ni 6.73E3 2.33E3 9.14E2 1.92E3 - ND 4.77E3 3.38E4
- Te-127m - 2.49E4 8.33E3 3.02E3 ' 6.07E3 - 6.36E 4 1.48E6 7.14E4 Te-129m l.92E4 6.83E3 3.04E3 6.33E3 3.03E4 1.76E6 - 1.82E 3 1-131 4.81 E4 - 4.81E4 2.73E4 1.62E7. 7.88E4 -ND 2.84 E3 -
- Cs-134 6.3l E3 1.01E6- 2.23E3 ND 3.30E3 1.21E3 : 3.83E3 -
- ' Cs-136' 6.31 E4 1.71E3- 1.16E3 ND 9.35E4 1.43E4 4.18E3 -
- Cs-!J7 7 9.07E3 8.23E3 1.28E3 ND 2.82E 3 - 1.04E3 3.62 E3 -
'Ba-140 7.40E4 - 6.48El 4.33E3 ND 2.11 E1 1.74E6 1.02E3
- Ce-141 - 3.92E4 1.93E4 - :. 2.90E3 ND '
8.35E3 3.44E3 3.66E4
- Ce-144 - > 6.77E6 2.12E6 3.61ES ND 1.17E6 1.20E7 3.89E3 Pr-143- :1.83E4 3.35E3 9.14 E2 - ND 3.00E3 4.33E3 9.73E4 Nd-147 !!.08E4 8.73E3 6.81 E2 ND- ~ 4.8 t E 3 3.2SES 8.2LE4 i
a *
- . 1-
- 37
, . . - - ..- -- . - ~. -
- . . - - - - . . _ ~ . - - -. -
s TABLE 4.4-4' inhalation Dose Factors - Teen Nucilde Bone Liver T. Body Thyroid Kldwy L 3 CI-LLI
- H ND 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 1 Cr ND ND 1.35E2 7.49 El 3.07El 2.09E4 3.00E3 Mn-54 ND 1.70E0 8.40E3 ND 1.27E4 1.98E6- 6.68E4
, Fe-55 3.34E4 2.38E4 5.54E3 ND ND 1.24E5 6.39E3. ,
Fo-59 1.59E4 3.70E4 1.43E4 ND ND 1.53E6 1.78E5 Co-58 ND 2.07E3 2.78E3 ND ND 1.34E6 9.52E4 Co-60 ND 1.51E4 1.98E4 ND ND 8.72E6 2.59E5 , N1-63. 5.80E5 4.34E4 1.98E4 ND ND 3.07E5 1.42E4 .
--Zn-65 3.86E4 1.34E5 6.24E4 ND 8.64E4 1.24E6 4.66E4 Rb-86 -HD- 1.90E5 - 8.40E4 ND ND ND 1.77E4 Sr-89 4.34E5 - ND 1.25E4 ND ND 2.42E6 3.71E5 Sr 1.08E8 ND 6.68E6 ND ND 1.65E7 7.65E5
- Y 6.61E5 ND 1.77E4 ND ND 2.94E6 - 4.09E5
-Zr-95-- 1.48E5 4.58E4' 3.15E4 ND 6.74E4 2.69E6 - 1.49E5
%95 ~< l.86E4 1.03E4 5.66E3 ND 1.00E4 7.51E5 9.68E4 u-103- 2.10E3 - ND 8.96E3 ND 7.43E3 7.83E5 1.09E5
- Ru-106 9.84E4 - ND - 1.24E4 ND - 1.90E5 1.6IE7 9.60E5 Ag-110m : 1.38E4 -- .l.31E4 . 7.99 E3 ND 2.50E4 6.75E6 - 2.73E5 i Tc-125m 4.88E3 2.24E3 6.67E2 1.40E3 ND 5.36E5 7.50E4.
Te-127m 1.80E4 8.16E3 2.18E3 4.38E3 6.54E4 1.66E6 1.59E5 : Te-129m- 1.39E4 6.58E3 - 2.25E3 4.58E3 -5.1954 1.98E6 ' 4.05E5 I-131
- 3.54E4 _ _4.91 E4 2.64E4 1.46E7 8.40E4 ND 6.49E3 - -- Cs-134 5.02E5 - l'.13 E6 5.49E5 ND 3.75E5 1.46E5 9.76E3 Cs-136 L 5.15E4 ' l.94E5 - 1.37E5 ND 1.10E5 1.78E4 1.09E4 ' Cs-137 6.70E5 S.48E5 3.11E5 ND 3.04E5 1.21 E5 . 8.48E3 '
Be-140 5.47E4 6.70 El - 3.52E3 ND 2.28El - 2.03E6 ? 2.29E5 Ce-141 ' 2.84E4 - 1.90E4 - 2.17E3 ND 8.88E3 6.14E5 1.26E5 Ce-144 4.89E6 2.02E6 ' 2.62E5 ND 1.21E6 1.34 E7 -- 8.64E5 Pr-1431 - 1.34E4 - 5.31E3 6.62E2 ND 3.09E3 : 4.83E5 2.14 E5 - E Nd-147 7.86E3 8.56E3 15.13E2 :ND 5.02E3 3.72E5 1.82E5
. ,. do ; J. v t b ud.
TABLE 4.4-5 Inhalation Dose Factors - Adult
- Nuclide Bone Liver T. Body Thyrold Kidney M GI-l.LI H 'l ND 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 Cr-51 ND ND 1.00E2 5.95El 2.28E! 1.44E4 3.32E3 Mn-54 ND 3.96E4 6.30E3 ND 9.84E3 1.40E6 7.74E4
- Fe-55 2.46E4 1.70E4 3.94E3 ND ND 7.21 E4 6.03E3 Fo-59 1.18E4 2.78E4 1.06E4 ND ND 1.02E6 1.88E5 Co-58 ND 1.58E3 2.07 E3 ND ND 9.28E5 1.06E5 Co-60 ND 1.15E4 1.48E4 ND ND 5.97E6 2.85E5 N1-63 4.32E5 3.14E4 1.45E4 - ND ND 1.78E5 1.34 E4 2n-65 3.24E4 1.03E5 4.66E4 ND 6.90E4 8.64E5 5.34E4 Rb 86 ~ND 1.35E5 5.90E4 ND ND ND 1.66E4 Sr-89 3.04E5 ND 8.72E3 ND ND 1.4E6 3.5E5 Sr-90 9.92E7 ND 6.10E6 ND ND 9.60E6 7.22E5 ,, _
' 91 -
- --. 4.62E5 ND 1.24E4 ND ND "
1.70E6 3.85E5 C r-95' - 1.07E5 3.44E4 2.33E4 ND 5.36E4 1.77E6 1.50E5 Nb-95 ' l.41 E4 7.76E3 4.21 E3 ND - 7.74E3 5.05E5 1.04 E5 + Ru-103 1.53E3 ND 6.58E2 ND 5.83E3 -5.05E5 1.10E5 Ru-106 6.91E4 ND 8.7?E3 ND' l.34E5 9.36E6 9.12E5 .
- Ag-110m 1.08E4 1.00E4 - 5.94E3 ND 1.97E4 4.63E6 3.02E5 To-125m 3.42E3 'l.58E3 4.67E2 1.05E3 - 1.24E4 3.14E5 7.06E4 To-127m 1.26E4 5.77E3 1.57E3 3.29E3 4.58E4 9.60E5 1.50E6 -
To-129m - 9.76E3 ' 4.67E3 . l.58 E3 3.44E3 3.66E4 1.16E6 3.83E5 1-131 2.52E4 - 3.58E4 2.05E4 1.19E7 6.13E4 ND 6.28E3 Cs-134 3.73E5 - 8.48E5 7.28E5 ND 2.87E5 9.76E4 1.04E4 Cs-136 3.90E4 1.46E5 1.10E5 ND -8.56E4 ' l.20E4 1.17E4 Cs-137 : 4.78E5 6.21 E5 4.28E5 ND 2.22E5 7.52E4 8.40E3-B6140? ~3.90E4 4.90E1 2.57 E3 ND 1.67El 1.27E6 2.18E5 - Co-141- 1.99E4 1.35E4 1.53E3 ND 6.26E3 3.62E5 1.20E5 Cc-144 '3.43E6 1.43E6 1.84E5 ND- 8.48E5 ' 7.78E6 3.16E5 Pr-143 : 9.36E3 3.75E3 4.64E2 ND 2.16E3 2.81E5 2.00E5.
-147 5.27E3. 6.10E3 3.65E2 ND 3.56E3 2.21 E5 1.73E5 5, -
REVISl01\ "O"
Calculation of ingestion Dose Factor Grass Cow Milk Pathway c - R 0/Q - K' QF(Vap) Fm(r)(DFLj)a fp fs
+
(1 - f pfs) e'A tih - i- e*A tl7 Aj + Aw Yp Ys where: Unit - m2 . mrem /yr per pCi/sec Reference Table R.G. 1.109 K' - A constant of unit conversion, 106 pC1/ C1. QF The cow's consumption rate, 50 kg/ day (wet weight) E-3 U ap
=
The receptor's milk consumption rate for age (a), E5 in liters, yr Infant & Child - 330 Teen - 400 Adult - 310 Yp - The agricultural productivity by unit area of pasture E 15 feed grass 0.7 kg/m2 Ys
- The agricultural productivity of unit area of E 15 stored feed 2.0 kg/mz Fm
_ The stable element transfer coefficients, in days /kg. E1 r = Fraction of depositea activity retained on cow's E 15 ,
%. feed grass 1.0 radioiodine 0.2 particulates -l tr - Transport time from pasture to receptor, in sec. E-15 1.73x10D sec (2 days) th- Transpe t time from crop field to receptor, in sec. E-15 7.78x100 sec. (90 days)
(DFLj)a The maximum organ ingestion dose factor for the ith E-ll to radionuclide for the receptor in age group (a), E 14 in mrem /pCi. Aj - The-decay constant for the ith radionuclide,_in sec'-l Aw: - The decay constant for removal of activity on eaf and E-15 plant surfaces _by weathering 5.73 x 10-' sec .
-(corresponding to a 14. day half-life),
fp - Fraction of the year that the cow is on pasture ---- (dimensionless) = 1*. fs - Fraction of the cow feed that is pasture grass ---- while the cow is on pasture (dimensionless)-= 1*.
- Milk cattle arc _ cons-idered to be fed from two potential sources, pasture grass
.and: stored feeds.
O 1 . e0 15 R E \n._.o:a . 2 m..>mmm. . . .
,- . .~ -_._ .-. _ . =_ . . _ . _ - . . . _ _ _ - _ . _ , . _ . _ . _ . _ _ _ .
Note: LThe above-equation does-ng1 apply to the concentration of tritium in meat. A separate equation is provided in NUREG 0133, section 5.3.1.4
-to determine Tritium value.
Reference:
The equation for R 6.3.1.3 i (D/Q) was taken from NOREG 0133 Section - 4 1 4 T O + REVISION 13 - 60a--
. = - . .
TABLE 4.4-6 Ingestion Dose Factors Grass-Cow-Mllk Pathway Qnfant) OJ ., Nuclide Bone Liver T. Body Thyroid Ki h y L GI-LLI 3 H-3 ND 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 - Cr-51 ND ND 1.61E5 1.03E5 2.30E4 2.03E5 4.71E6 Mn-54 ND 3.89E7 8.83E6 ND 8.63E6 ND 1.43E7 Fe-55 -1.35E8 8.72E7 2.33E7 ND ND 4.26E7 1.llE7 Fe-59 2.26E8 3.94E8 1.53E8 ND ND 1.17E8 1.88E8 Co-58 ND 2.43E7 6.06E7 ND ND ND 6.05E7 Co-60 ND 8.81 E7 2.08E8 ND ND ND 2.10E8 N1-63 3.49E10 2.16E9 1.21E9 ND ND ND 1.07E8 Zn-65 5.55E9 1.90E10 8.78E9 ND 9.24E9 ND 1.61E10 Rb 86 ND 2.23E10 1.10E10 ND ND ND 5.70E8 Sr-89 ND 1.45E6 9.98E5 ND ND ND 4.93E5 Sr-90 1.22 Ell ND 3.10E10 ND ND ND 1.52E9 Y-91 7.33E4 ND 1.95E3 ND ND ND 5.26E6 Zr-95 6.34 E3 1.67 E3 1.18E3 ND 1.80E3 ND 8.30E5 . pNb-95 .~. 5.93E5 2.44E5 1.41 E5 ND 1.75E5 ND 2.06E8 dau-103. 8.68E3 ND 2.90E3 ND 1.81E4 ND 1.06E5 Ru-106 1.90E5 ND 2.38E4 ND 2.25E5 ND 1.44E6. Ag-110m 3.86E8 -- 2.82E8 1.87E8 ND- 4.03E8 ND 1.46E10 Te-125m 1.51E8 5.04E7 2.04 E7 5.07E7 ND ND 7.18E7, Te-127m . 4.21E8 1.40E8 5.10E7 1.22E8 1.04E9 ND 1.70E8 Tc-129m 5.60E8 1.92E8 8.62E7 2.15E8 1.40E9 ND 3.34E8 I-131- 2.72E9 3.21 E9 1.41E9 1.03E12 3.75E9 ND 1.15ES ' Cs-134 ' 3.65E10 6.80E10 6.87E9 ND 1.75E10 7.18E9 1.85E8 Cs-136 2.03E9 5.96E9 2.22E9 ND 2.37E9 . 4.85E8 9.05E7. Cs-137 5.15E10 6.02E10 4.27E9 ND - 1.62 E10 - 6.55E9 1.88E8 Be-140 2.41E8 - 2.41 E5 1.24E7 ND. 5.73E4 1.48E5 5.92E7
- Ce-141 - 4.34 E4 2.64E4 3.llE3 ND 8.16E3 ND 1.37E7 Cc-144 2.33E6 - 9.52E5 1.30E5 ND 3.85E5 ND 1.33E8 . Pr-143 1.49E3 5.56E2 7.37El ND 2.07E2 ND 7.85E5 -
Nd-147 ' 8.86E2 - 9.10E2 5.57El- ND 3.51E2 ~ ND 5.77E5 O REVSON "O"
TABLE 4.4-7 IrinMtion Dose Factors Grass-dow-Mik Pathway (Child) Nuclide _Borm Liver .T. Body Thyroid Kidney L GI-LLI 3 H-3 ND 1.57E3 1.57E3 1.57E3 1.57E3 1.57E3 1.57 E3 Cr-51 ND ND 1.02E5 5.66E4 1.55E4 1.03E5 5.41E6
. Mn-34 ND 2.09E7 5.58E6 ND 5.87E6 ND 1.76E7 ,
Fe-55 1.12E8 5.93E7 1.84E7 ND ND 3.35E7 1.10E7 Fe-59 1.21E8 1.96E8 9.75E7 ND ND 3.67E7 2.04E8 Co-58 ND 1.21E7 3.72E7 ND ND ND 7.08E7 Co-60 ND 4.32E7 1.27E8 ND ND ND 2.39 E8 Ni-63 2.96E10 1.59E9 1.01E9 ND ND ND 1.07E8 Zn-65 4.13E9 1.10E10 6.85 E9 ND 6.94E9 ND 1.93E9 Rb-86 ND 8.77 E9 5.39E9 ND ND ND 5.64E8 Sr-89 6.69E9 ND 1.91E8 ND ND ND 2.59ES Sr-90 1.12 Ell ND 2.83E10 ND ND ND 1.50E9 Y-91 3.91E4 ND 1.fe4E3 ND ND ND 5.2tE6 Zr-95= 3.85E3 8.46E2 7.53E2 ND 1.21E3 ND 8.83E5 Jb-95 . 3.18E5 1.24E5 8.84E4 ND 1.16E5 ND 2.29ES ~ ~ \ .u-103 4.29E3 ND 1.65E3 ND 1.08E4 ND 1.1 lES ' Ru-106 9.24 E4 ND 1.15E4 ND 1.25E5 ND 1.44E6 Pg-ll om 2.09E8 1.4lE8 1.13E8- ND 2.63E8 ND 1.68E10 Te-125m 7.38E7 2.00E7 9.84E6 2.07E7 ND ND 7.12E7 To-127m 2.08E8 5.60E7 2.47E7 4.97E7 ~5.93E8 ND 1.68E8 < Te-129m 3.17E8 8.85E7 4.92E7 1.02E8 9.31 E8 - ND 3.87E8
' !-131' l.30E9 1.31E9 7.46E8 4.34 Ell 2.15E9 ND 1.17E8 ' Cs-134 - 2.26E10 3.71E10 - 7.84E9 ND 1.15E10 4.13E9 2.00E8 Cs-136 1.04E9 2.85E9 1.84E9 NO - 1.52E9 2.26ES 1.00E8 Cs-137 - 3.22E10 -- 3.09E10 4.53E9 ND 1.0lE10 3.62E9 L.93E8 'Bt-140 1.17E8 1.03E5 6.84E6 ND ' 3.34E4 ~ 6.12E4 5.94E7-Cc-141 2.19E4 1.09E4 1.62E3 ND- 4.78E3 ND 1.36E7 Co-144 1.62E6 5.09E5 8.66E4 ND 2.82E5 ND 1.33E8 ?r-143- 7.19E2 2.16E2 3.57El ND 1.17E2 ND 7.76E5 Nd-147.- 4.47E2 3.62E2 2.80El ND 1.99E2 ND 5.73E5 -
O REVlSlrdiN u vw
TABLE 4.4-8 Ingestion Dose Factors Grass-Cow-Mllk Pathway (Teen) G Nuclide Bone Liver T. Body Thyroid Kidney L3 Cl-LLI H-3 ND 9.94E2 9.94E2 9.94E2 9.94 E2 9.94 E2 9.94E2 Or-51 ND ND 5.00E4 2.78 E4 1.09E4 7.13E4 8.40E6 Mn-54 ND 1.40E7 2.78E6 ND 4.18E6 ND 2.87 E7 Fc-55 4.45E7 3.16E7 7.36E6 ND ND 2.00E7 1.37E7 Fc-59 5.21E7 1.22E8 4.70E7 ND ND 3.87 E7 2.88E8 Co-58 ND 7.95E6 1.83E7 ND ND ND 1.10E8 Co-60 ND 1.64E6 3.70E6 ND ND ND 3.14E7 N1-63 1.82E10 8.35E8 4.01 E8 ND ND ND 1.33E8 Zn-65 2. ll E9 7.32E9 3.41 E9 ND 4.68E9 ND 3.10E9 Rb-86 ND 4.73E9 2.22E9 ND ND ND 6.99E8 Sr-89 2.70E9 ND 7.73E7 ND ND ND 3.22E8 Sr-90 6.61E10 ND 1.63E10 ND ND ND 1.86E9 Y-91 1.58E4 ND 4.24E2 ND ND ND 6.48E6 Zr-95 1.66E3 5.22E2 3.59E2 ND 7.68E2 ND 1.21E6 r$'b-95 ~~.1.41E5 7.80E4 4.29E4 ND 7.56E4 ND 3.34 E8 - Clu-103 1.81E3 ND 7.74E2 ND 6.39E3 ND 1.51E5 Ru-106 3.75E4 ND 4.73E3 ND _. 7.24E4 ND 1.80E6 Ag-110m 9.64 E7 9.12E7 5.55E7 ND 1.74E8 ND 2.56E10 Tc-125m 3.00E7 1.08E7 4.02E6 8.39E6 ND ND 8.86E7 To-127m 8.44E7 2.99 E7 1.00E7 2.01 E7 3.42E8 ND 2.10E8 Tc-129m 1.llES 4.ll E7 1.75E7 3.57E7 4.63E8 ND 4.16E8 I-131 5.38E8 7.53 E8 4.05ES 2.20 Ell 1.30E9 ND 1.49E8 Cs-134 9.81 E9 2.31E10 1.07E10 ND 7.34E9 2.SOE9 2.87E8 Cs-136 4.59E8 1.80E9 1.21 E9 ND 9.82E8 1.55ES 1.45E8 Cs-137 1.34E10 1.78E10 6.20E9 ND 6.06E9 2.35E9 2.53E8 Bt-140 4.87 E7 5.96E4 3.14 E6 ND 2.02E4 4.01 E4 7.51E7 Cc-141 8.89E3 5.93E3 6.81E2 ND 2.79E3 ND 1.70E7 C;-144 6.58E5 2.72E5 3.54E4 ND 1.63E5 ND 1.65E8 Pr-143 2.89E2 1.15E2 1.44El ND 6.73E1 ND 9.53E5 Nd-147 1.82E2 1.93E2 1.19El ND 1.16E2 ND 7.15E5 (q l 1 m n , . p i .^, "iD.,1
.a
,; , ~:
t, TABLE 4.4-9 Ingestion Dose Factors Grass-Cow-Milk Pathway (Adult) Nuclide Bone Liver T. Body Thyroid Kidney Lang 3 GI-LLI ' H-3 ND 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 Cr-51 ND ND 2.86E4 1.71E4 6.27E3 3.80E4 7.20E6 Mn-54 ND 8.40E6 1.60E6 ND 2.50E6 ND 2.57E7 - Fo-55 2.51E7 1.73E7 4.04E6 ND ND 9.67E6 9.95E6 Ps-59 2.99 E7 7.02E7 2.69E7 ND ND 1.96E7 2.34 E8 Co-58 ND 4.72E6 1.06E7 ND ND ND 9.51E7
~ Co-60 ND 1.64E7 3.62E7 ND ND ND 3.08E8 Ni-63 6.73E9 4.66E8 2.27E8 ND ND ND 9.73E7 Zn-65 1.37E9 4.37E9 1.97E9 ND 2.92E9 ND 2.75E9 Rb-86 ND 2.59E9 1.21E9 ND ND ND 5.ll E8
.Sr-89 1.47E9 ND 4.2 t E7 ND ND ND 2.35E8 Sr 4.69E10 ND 1.15E10 ND ND ND 1.35E9'
'8.60E3 ND 2.29E2 ND ND ND 4.73E6 Zr-95 1.06E3 3.04E2 2.06E2 ND 4.77E2 ND 9.63E5 Nb-95 ' 3.65E5 2.44E5 9.59E3 ND ~ 2.43E5 ND 1.93E9 Ru-103 1.02E3 ND 4.39E2 ND 3.89E3 ND 1.19ES Ru-106 2.04E4 ND 2.58E3 ND 3.94E4 ND IJ2E6 ~
- Ag-110m 5.83E7 5.39E7 3.20E7 ND 1.06E8 ND 2.20E10 -
. Te-125m 1.63E7 5.90E6 2.18E6 4.90E6 6.J3E7 ND 6.50E7 Te-127m 4.58E7 1.64E7 5.58E6 1.17E7 1.86E8 ND 1.54E8 Te-129m 6.05E7 2.26E7 9.58E6 2.08E7 2.53E8 ND 3.05E8 I-131 2.97 E8 4.24E8 2.43E8 1.39 Ell 7.27E8 ND 1.12E8 Cs-134 - 5.65E9 1.34E10 1.10E10 ND 4.33E9 1.44E9 2.35E8 Cs-136 2.69E8 1.06E9 7.65E8 ND 5.92E8 8.11 E7 1.21 E8 -
Cs-137 - 7.38E9 1.01E10 6.61E9 ND 3.43E9 1.14E9 1.95 E8 - Bt-140 2.70E7 - 3.39E4 1.77E6 ND 1.15E4 1.94E4 5.55E7 Cc-141 4.85E3 3.28E3 3.72E2 - ND 1.52E3 ND 1.25E7 Cc-144 3.58E5 .l.30E5 1.92E4 ND 8.87E4 ND 1.21ES Pr-143 1.94E2 7.79El 9.62E0 ND 4.47El ND 8.50E5
-147 9.49El 1.10E2 6.56E 0 ND 6.41 El ND 5.26E3
.. REVIS10i% "O"
_ _ . _ - . . _ _ _ _- . _ _ . _-__ - . _ _ . _ . _ _ _ . _ _ _ _ _ _ ~. _ Calculation of Ingestion Dose Factor Grass Cow-Meat Pathway M
-R QF-(Vap) fp Is (1 - f pfs) e-A tjh
[ D/Q] = K' Fr(r)(DFLj)a '
+ 0'A tiy i 19 + Aw Yp Ys where: Unit = m2 . mrem /yr per #Ci/sec Reference Table R.G. 1.109 K' =
A constant of unit conversion, 106 pCi/ Ci. Qp = The cow's consumption rate, 50 kg/ day (wet weight) E-3 Ugp = The receptor's meat consumption rate for age (a), E-5 in kg/yr Infant - 0 Child - 41 Teen - 65 Adult - 110 Yp The agricultural productivity by unit area of pasture E-15 feed grass 0.7 kg/m2 Ys = - The agricultural productivity of unit area of E-15 stored feed 2.0 k;:/m2 Fr ' = The stable element transfer coefficients, in days /kg. E-1 - O - Fraction of' deposited activity retained on cow's E-15 feed grass 1.0 radiciodine 0.2 particulattes tr = Transport time from pasture to receptor, in sec. E-15 1.73x106 see (20 days) th --- Transport time from crop field te receptor, in sec. E-15 7.78x100 (9Ldays)_sec. (DFt.j)a - The maximum organ ingestion dose factor for the ith E-ll to radionuclide for the receptor in age group (a), E-14 in area /pCi Aj = The decay constant for the ith radionuclide', in-sec ---- Aw.= The decay constant for removal of activity on E-15 leaf and plant surfaces by weathering, 5.73 -x 10-7 sec. -1 (corresponding to a 14 day half-life). fp = Fraction of the year that-the cow is on ptsture ---- (dimensionless) = 1*. fs - Fraction of the cow feed that is pasture grass ---- while the cow is on pasture (dimensionless) - 1.
- REV SION 13 - es -
L.
- Milk'and grass cattle -
are considered to be fed from two potential sources, pastura storod-feeds.- Following the development in Regulatory Guide 1.109,
- the values of fp and fs will be considered unity, in lieu of site specific information provided in the annual land census report by the licensee. .
^
Note: The above equation does no.t apply to the conceitration of tritium in ~ meat. A separate equation is provided in NUREG 0133, section 5.3.1.4 to-determine Tritium value.
.~,
~
m
Reference:
The equation deriving.Rj-(D/Q) was taken-from NUREG 0133, Section 5.3.1.4. t f;--in NUREG-0133 is~ equivalent to ts in R.G. 1.109 Table E-15. OlEVISMN 13 1
- 6ba -
' [ r i-W TABLE 4.4-10
- Ingestion Dose Factors Grass-Cow-Meat Pathway (Child) ,
Nuclide Bone Liver T. Body Kidney Thyroid M GI-LLI H-3 ND- 2.34E2 2.34E2 2.34E2 2.34E2 2.34E2 2.34 E2 Cr-51 ND - ND 8.82E3 4.89E3 1.34E3 8.93E3 4.68E5
- Mn-54 ND 7.99E6 2.13E6 ND 2.24E6 ND 6.70E6 Fe-55 4.57E8 2.42E8 7.50E7 ND ND 1.37E8 4.49E7 Fe-59 3.81E8 6.16E8 3.07E8 ND NO !.79E8 6.42E8 -
Co-58 ND 1.65E7 5.04E7 ND ND ND 9.60E7 Co-60 ND 6.93E7 2.04E8 ND ND ND 3.84E8 N1-63 2.91E10 1.56E9 9.91E8 ND ND ND 1.05E8 Zn-65 3.76E8 1.00E9 6.22E8 ND 6.30E8 ND 1.76E8 Rb-86 ND- 5.77E8 3.53E8 ND ND ND 3.71E7 Sr-89 4.92E8 ND -1.40E7 ND ND ND 1.90E7 Sr-90 1.0t 310 ND 2.64E9 ND ND ND 1.40E8 Y-91 1.81E6 ND 4.83E4 ND ND ND 2.4 t E8 Zr 2.69E6 5.91E5 5.26E5 ND 8.46E5 ND 6.16E8 Nb-95 ._3.09E6 1.20E6 8.61E5 ND 1.13E6 ND' 2.23E9 (u-103 1.55E8 ND 5.97E7 ND 3.91 E8 ND 4.02E9 - Ru-106 4.44E9 .ND 5.54E8 ND . 5.99E9 ND 6.90E10 , Ag-110m . 8.4lE6 5.68E6 4.54E6 ND 1.06E7 ND 6.76E8 Te-125m 5.69E8 1.54E8 7.59E7 1.60E8 -ND ND 5.49E8 Te-127m 1.77E9 - 4.78E8 2.11E8 4.24E8 5.06E9 ND 1.44E9 - To-129m 4.78E9 5.05E8 2.81E8 5.83E3 5.31E9 ND 2.21 E9 --
- I-131 -- 1.66E7. _ l.67E7 .- 9.49E6 5.52E9 2.74E7 ND 1.49E6-Cs-134 9.22E8' l.51 E9 ' 3.19E8 ND 4.69ES - 1.68E8- 8.16E6
.-Cs-136 1.73E7 4.74E7 ' 3.07E7 ND 2.53E7 3.77E6 1.67E6
' Cs-137 1.33E9 1.28E9 - 1.88 E8 ' ND 4.16E8 1.50E8 - 7.99E6 -
Be-140 - 4.39E7- 3.85E4 2.56E6 ND 1,25E4 - 2.29E4 2.22E7 --
-Co-141 2.22E4 1.11E4 1.64E3 ND - 4.86E3 ND 'l.38E7 Co-144 2.32E6 7.26E5 1.24E5 ND 4.02E5 ND 1.89E8-jPr-143- ( 3.35E4 - 1.01E4 1.66E3 ND- 5.45E3 ND _ 3.61 E7 Nd-147 1.18E4 9.60E3 7.43E2 ND ' 5.27E3 ND 1.52E7-
- ., Q *' C ' ,-l.1 ~ i. l
-I' '
TACLE 4.4-11 Ingestion Dose Factors Grass-Cow-Meat Pathway (Teen) n UNuclid_e Bone Liver T. Body Thyroid KJdney g GI-LLi* H-3 ND 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 Cr-51 ND ND 5.65E3 3.14 E3 1.24E3 8.07E3 9.49E5 Mn-54 ND 6.98 E6 1.39E6 ND 2.08E6 ND 1.43E7 Fe-55 2.38E8 1.69E8 3.93E7 ND ND l.07E8 7.30E7 Fc-59 2.15E8 5.01E8 1.94E8 ND ND 1.58E8 1.19E9 Co-58 ND 1.41E7 3.25E7 ND ND ND 1.94E8 Co-60 ND 5.83E7 1.31E8 ND ND ND 7.60E8 N1-63 1.52E10 1.07 E9 5.15E8 ND ND ND 1.71E8
- - Zn-65 2.50E8 8.69E8 4.06E8 ND 5.56E8 ND 3.68E8 Rb-86 ND 4.06E8 1.91E8 ND ND ND 6.01E7 Sr-89 2.60E8 ND 7.44E6 ND ND ND 3.09E7 Sr-90 8.05E9 ND 1.99E9 ND ND ND 2.26E8 Y-91 9.56E5 ND 2.56E4 ~ ND ND ND 3.92E8 Zr-95 1.51 E6 4.78E5 3.28E5 ND 7.02E5 ND 1.10E9 Nb-95 1.79E6 9.93E5 5.47E5 ND 9.63E5 ND 4.25E9 "
J-103 8.58E7 ND 3.67E7 ND 3.03E8 ND 7.17E9
]'Ru-106 2.36E9 ND 2.97E8 ND 4.55E9 ND 1.13E11 Ag-ilom 5.07E6 4.80E6 2.92E6 ND ~ 9.15E6 ND 1.35E9 Te-125m 3.03E8 1.09E8 4.03E7 8.47E7 ND ND 8.94 E8 ,
Tc-127m 9.42E8 3.34 E8 1.12E8 2.24E8 3.82E9 ND 2.3J E9 - Te-129m - 9.61E8 3.57E8 1.52E8 3.10E8 4.02E9 ND 3.61 E9 - I-131 8.97E6 1.26E7 6.75E6 3.66E9 2.16E7 ND 2.48E6 Cs-134 5.2358 1.23E9 5.71E8 ND 3.91E8 1.49E8 1.53E7 , Cs-136 9.96E6 3.92E7 2.63E7 - ND 2.13E7 3.36E6 3.15E6 CS-137 7.24E8 9.63E8 3.36E8 ND 3.28E8 1.27E8 - 1.37E7 Bn-140 2.39E7 2.93E4 1.54E6 ND 9.94E3 1.97E4 3.69E7 ; Cs-141 1.lSE4 7.88E3 9.05E2 ND 3.71E3 ND 2.25E7 Cc-144 1.23E6 5.08E5 6.60E4 ND 3.04E5 ND 3.09E8
- Pr-143 1.76E4 7.03E3 8.76E2 ND 4.09E3 ND 5.79E7
- Nd-147 6.32E3 6.87 E3 4 .12E2 ND 4.04E3 ND 2.48E7 O
,_ REV!SION "O"
TACLE 4.4-12 Ingestion Dose Factors Grass-Cow-Meat Pathway (Adult) duclide Bone Liver T. Body Thyroid Kidney Lung GI-LLl-H-3 ND 3.25E2 3.25E2 3.25El 3.25E2 3.25E2 3.25E2 Cr-51 ND ND 7.06E3 4.22E3 1.56E3 9.37E3 1.78E6 Mn-54 ND 9.16E6 1.75E6 ND 2.72E6 ND 2.80E7 Fe-55 2.93E8 2.02E8 4.72E7 ND ND 1.13E8 1.16E8 Fc-59 2.69E8 6.32E8 2.42E8 ND ND 1.76E8 2.ll E9 Co-58 ND 1.83E7 4.10E7 ND ND ND 3.70E8 Co-60 ND 7.52E7 1.66ES ND ND ND 1.41E9 N1-63 1.89E10 ' l.31 E9 6.33E8 ND ND ND 2.73E8 Zn-65 3.56E8 1.13E9 5.12E8 ND 7.58E8 ND 7.13 E8 Rb-86 ND 4.86E8 2.27ES ND ND ND 9.59E7 Sr-89 3.08E8 ND 8.83E6 ND ND ND 4.93E7 Sr-90 1.24E10 ND 3.05E9 ND ND ND 3.59E8 Y-91 1.13E6 ND 3.03E4 ND ND ND 6.24E8 Zr-95 1.89E6 6.06E5 4.10E5 ND 9.51 E5 ND 1.92E9 Nb-95 2.29E6 1.28E6 6.86E5 ND 1.26E6 ND 7.74E9~ g] u-103 1,05E8 ND 4.54E7 ND 4.02E8 ND 1.23E10 W lu-106 2.80E9 ND 3.54E8 ND 5.40E9 ND 1.81 Ell Ag-110m 6.70E6 6.19 E6 3.69E6 ND ~ 1.22E7 ND 2.53E9 ~ Tc-125m 3.59ES 1.30E8 4.81 E7 1.08E8 1.46E9 ND 1.43E9 Tc-127m 1.12E9 3.99ES 1.36E8 2.85ES 4.53E9 ND 3.74E9-Tc-129m 1.15E9 4.28E8 1.82E8 3.94E8 4.79E9 ND 5.78E9 I-131 1.08E7 1.54E7 8.85E6 5.06E9 2.65E7 ND 4.07E6 Cs-134 6.57 E8 1.56E9 1.29E9 ND 5.06E8 1.68E8 2.74E7 Cs-136 1.28E7 5.04E7 3.63E7 ND 2.80E7 3.84 E6 5.73E6 Cs-137 8.72E8 1.19E9 7.81E8 ND 4.05E8 1.35E8 2.31 E7 Bn-140 2.90E7 3.64 E4 1.90E6 ND 1.24E4 2.08E4 5.96E7 Cc-141 1.41E4 9.51E3 1.08E3 ND 4.41 E3 ND 3.63E7 Cc-144 1.46E6 6.09 E5 7.82E4 ND 3.61 E5 ND 4.93E8 Pr-143 2.09E4 8.39E3 1.04E3 ND 4.85E3 ND 9.17E7 Nd-147 7.17E3 8.29E3 4.96E2 ND 4.85E3 ND 3.99E7 l l m R :y SLON "O" ' Calculaticn of Ingestion Dose Factor Vegetation Pathway Rj [D/Q] = K' (DFLj)a Va fL e*A i th + U fgeAtjh Yv (Ai + Aw)
~ '
where: Units - m2
- mrem /yr per uti/sec. Reference Tame. R.G. 1.iO9 K' -
A constant of unit conversion, 106 pCi/pC1. Ua The consumptien rate of fresh leafy vegetation by the E-5 receptor in age group (a), in kg/yr, infant - 0 Child - 26 Teen - 42 Adult - 64 S Ua The consumption rate of stored vegetation by the E-5 receptor in age group (a), in kg/yr Infant - 0 Child - 520 Teen - 630 Adult - 520 (DFLj)a - The maximum organ ingesting dose factor for the ith E-11 to E-14 radionuclide for the receptor in age group (a), ~
~. in mrem /pC1. '
~U ft - .The fraction of the annual intake of fresh leafy E-15 vegetation grown locally. (default 1.0) l fg - The fraction of the annual intake of stored vegetation - E-15 grown locally. (default 0.76) l tt - The average time between harvest of leafy vegetation _ E-15 l and its consumption, 8.6 x 10', seconds, (1 day) th - The average time between harvest'of stored vegetation E-15 and its consumption, 5.18 x 106 seconds, (60 days) Yv - The vegetation a real density, 2.0 kg/m2 E-15 r - Fr;; tion of deposited activity retained on the E-15 vegetation 1.0 radioiodine 0.2 particulates Ai - The decay constant for the ith radionuclide, in sec 'I --- Aw
- The decay constant for removal of activijy on lgaf and E-15 plant surfaces by weathering, 5.73 x 10' sec '
_(corresponding to a 14 day half-life). O 9EVISION 13
- Note: The above equation does ad apply to the concentrations of tritium in vegetation. A separate equation is provided in NUREG 0133; section 5.3.1.5 to determine tritium values.
O .
Reference:
The' equation deriving R (D/Q) was taken from NUREG 0133, i Section 5.3.1.5. i
- s. .
#8Ehe e O
**-' s e
?
l 1 O REVISION 13 . eSa .
TABLE 4.4-13 ~ Ingestion Dose Factors Yegetation Pathway (Child) ' Nuclide Bone Liver T. Body Thyroid Kidney L GI LLI 3 H-3 .ND 4.01E3 4.01 E3 4.01 E3 4.01E3 4.01 E3 4.01E3 Cr-51 ND ND- 1.18E5 6.54 E4 1.79E4 1.19E5 6.25E6 Mn-54 ND 4.61E8 1.76E8 ND 1.85E8 ND 5.55E8 Fc-55 8.00E8 4.24E8 1.31E8 ND ND 2.40E8 7.86E7 Fs-59 4.07E8 6.58E8 3.28E8 ND ND 1.91E8 6.85E8 Co ND 6.47 E7 1.98E8 ND ND ND 3.77 E8 Co-60 ND- 3.78E8 1.12E9 ND ND ND 2.10E9 Ni 3.95E10 2.llE9 1.34E9 ND ND ND 1.42E8 Zn-65 8.13E8 2.17E9 1.35E9 ND 1.36E9 ND 3.80E8 Rb-86 ND 4.52EE 2.78E8 ND ND ND 2.91 E7 Sr-89 3.74 E10 - ND 1.07E9 ND ND ND 1.45E9 Sr-90 1.24E12 - ND 3.15 Ell ND ND ND 1.67E10 Y-91 1.87E7 - ND 5.01E5 ND ND ND 2.49E9~ u
' 3.92E6 8.63E5 7.68E5 ND 1.23E6 ND 9.00E8 Nb-95 4.10E5 1.60E5 1.14E5 ND 1.50E5 ND 2.95E8 Ru-103 1.54E7 'ND 5.92E6 ND ' 3.88E7 ND 3.98E8 '
Ru-106 7.45E8 -ND 9.30E7 ND 1.01E9 ND 1.16E10-Ag-110m 3.23E7 -1.18E7 1.74E7 ND 4.06E7 ND 2.59E9'
- To-125m 3.51E8 9.50E7 4.67E7 9.84E7 ND ND 3.38E8 Tc-127m - 1.32E9 3.56E8 1.57E8 3.16E3 1.94E9 ND 1.07E9
- Tc-129m 8.58E8 2.40E8 1.33E8 2.77E8 2.52E9 ND 1.05E9 -
I-131 -1.43E8 1.44E8 8.18E7 4.76E10 2.36E8 ND . l.28 E7 Cs-134 : 1.60E10 2.63E10 3.55E9 ND 8.15E9 2.92E9 1.42E8 Cs-136 ' . 4.44 E8 1.22E9 7.9CE8 ND 6.50E8 9.69E7 4.29E7
.Cs-137 2.39 E10 -- 2.29E10 - 3.38E9 ND 7.46E9 2.68E9 1.43E8 Bt-140 2.77E8 2.43E5 - 1,62E7 ND 7.91E4 1.45E5 1.40E8 -
-- Ce- 141 - 6.56E5 3.27E5 - 4.86E4 ND 1.43E5 ND 4.08E8 -
'Ce-144= 1.27E8 3.98E7 6.78E6 ND 2.21 E7 ND 1.04E10 -
Pr-143 1.46E5 4.39E4 7.26E3 .JD 2.38E4 ND 1.58E8 ; Nd-147 7.23E4 5.86E4 4.54E3 ND 5.47El ND 9.28E7 - O 1 1 RWIS!OR"0"
TABLE 4.4-14 Ingestion Dose Factors Vegetation Pathway (Teen) Nuclide Bone Liver T. Body Thyroid Kidney L3 GI.LLI H-3 ND 4.10E3 4.10E3 4.10E3 4.10E3 4.10E3 4.10E3 P-32 1.60E9 9.91 E7 6.20E7 ND ND ND 1.34E8 Cr-51 ND ND 6.19E4 3.44E4 1.36E4 8.84E4 1.04E7 Mn-54 ND 4.52E8 8.97E7 ND 1.35E8 ND 9.27E8 Fe-55 3.25E8 2.31E8 5.38E7 ND ND 1.46E8 9.98E7 Fe-59 1.83E8 4.23E8 1.65E8 ND ND 1.35E8 1.01E9 Co-58 ND 4.38E7 1.01E8 ND ND ND 6.04E8 Co-60 ND- 2.49E8 5.60E8 ND ND ND 3.24 E9 Ni 1.61E10 1.13E9 5.44E8 ND ND ND 1.81E8 Zn-65 4.24E8 1.47E9 6.87ES ND . 9.43E8 ND 6.24E8 Rb-86 ND 2.73ES 1.28E8 ND ND ND- 4.04E7
- Sr 89 1.57E10 ND 4.50E8 ND ND ND 1.87 E9 '.
/-90 7.51 Ell ND 1.85 Ell ND ND ND 2.llE10 Y-91 7.87 E6 'ND 2.llE3 ND ND ND 3.23E9 Zr-95 1.75E6 5.52E5 3.80E5 ND 8.12E5 ND 1.27E9 Nb-95 1.92E5- 1.06E5 5.85E4 ND 1.03E5 ND 4.54E8
- Ru-103 - 6.85E6 ND 2.93E6 .ND 2.41E7 ND 5.72E8'
. Ru-106 3.09E8 ND 3.90E7 ND- 5.97E8 ND - 1.48E10
. Ag-110m 1.52E7 ' t.44E7 8.76E6 ND 2.75E7. ND 4.04E9 .
Te-125m . 1.48E8 5.34E7 1.98E7 4.14 E7 - .ND ND 4.37E8 Tc-127m 5.52E8 1.96E8 6.56E7 1.31E8 2.24E9 ND 1.37E9 Tc-129m - 3.69E8 - - 1.37E8 '5.84E7 1.19E8' l.54E9 ND 1.39E9
.1-131- 7.70E7 - 1.08E8 5.79E7 3.15E10 1.86E8 ND 2.13E7 Cs-134 7.10E9 1.67E10 7.75E9 ND 5.31E9 2.03E9 2.08E8 Cs-136 4.65E7 1.83E8 1.23E8 ND 9.96E7 1.57E7 1.47E7 Cs-137 - 1.01E10 1.35E10 4.69E9 ND 4.59E9 1.78E9 1.92E8 -
i B2-140 - 1.39E8 1.71E5 8.97E6 ND 5.78E4 1.15E5 2.15E8 Cz-14.1 2.83E5 1.89E5- 2.17E4 ND 8.90E4 ND 5.4IE8 Cc-144 5.27E7 2.18E7 2.82E6 ND. 1.30E7 ND 1.33E10
-143 6.99E4 2.79E4 3.48E3 ND 1.62E4 ND 2.30E8 Nd-147 3.66E4 3.98E4 - 2.39E3 ND 2.34E4 ND 1.44E8 Q)}{\ONggAll
--. . - - ~ . . _
TABLE 4.4-15 l Ingestion Dose Factors Vegetation Pathway (Adult) Nuclide Bone Liver T. Body Thyrold Kidney g CI-LLI H-3 ND 5.llE3 5.llE3 5.llE3 5.11E3 5.llE3 5.!!E3 i Cr-51 ND ND 4.66E4 2.79E4 1.03E4 6.18E4 1.17E7 Mn-54 ND 3.llE8 5.94E7 ND 9.27E7 ND 9.54E8 Fo-55 2.09E8 - 1.45E8 3.37E7 ND ND ' 8.06E7 8.29E7 Fe-59 , 1.29E8 3.02E8 1.16E8 ND ND 8.45E7 1.01E9 i Co-58 ND - 3.09E7 6.92E7 ND ND ND 6.26E8 Co-60 ND 1.67E8 3.69ES - ND ND ND 3.14 E9
' N1-63 1.04E10 - 7.21 E8 3.49ES - ND ND ND 1.50E8
.In 3.18E8 1.01E9 4.57E8 ND 6.76E8 ND 6.37E8 Rb ND 2.19E8 1.02E8 ND ND ND 4.32E7 Sr-89.~ 1.03E10 ND 2.96E8 ND. ND ND 1.65E9 Sr-90 6.05 Ell ND 1.48 Ell ND ND ND 1.75E10'.
I -91 5.13E6 ND 1.37E5 Nr) ND ND 2.82E9 Zr-95 1.19E6 3.83E5 2.59ES ND 6.00E5 ND 1.21E9 Nb-95 1.42E5 7.90E4 4.24E4 ND ~ 7.81E4 ND 4.79E8 - Ru-103 J 4.79E6 - -ND 2.06E6 ' ND 1.83E7 ND 5.59E8' Ru-106 1.93E8 ND 2.44E7 ND 3.72E8 ND .1.25 Eld - Ag-110m 1.06E7 9.78E6 5.81E6 ND 1.92E7 ND 3.99E9 To-125m 9.66E7 3.50E7 1.29E7 2.90E7 3.93E8 ND - 3.86E8 Te-127m 3.49E8 1.25E8 4.26E7 8.93E7 1.42E9 ND ' 1.17E9 Te-129m - 2.56E8 9.55E7 4.05E7 8.79E7 1.07E9 ND 1.29E9 - I-131 8.09E7 - 1.16E8 6.63E7 3.79E10 1.98E8 ND 3.03E7 , Cs-134 4.66E9 1.11E10 9.07E9 ND - 3.59E9 1.19E9 1.94E8
.Cs-136 -- 4.47E7 1.77E8 1.27E8 ND 9.82E7 1.35E7- 2.01E7L Cs-137 6.36E9 8.70E9 - 5.70E9 ND 2.95E9 9.81E8 1.68E8 Ba-140- 1.29E8 - 1 62E5
. 8.47E6 ND 5.32E4 9.29E4 2.66ES
- Ce-141 1.97E5 1.33E 5 - 1.51E4 ND- 6.20E4 - . ND 5.10E8 Cc-144 3.29E7 1.37E7 1.77E6 ND 8.15E6 NO -1.llE10
, Pr-143 - 6.25E4 2.51E4 3.10E3 ND 1.45E4 ND 2.74E8 L bd-147 3.36E4 3.89E4 2.33E3 ND 2.27E4 ND 1.87E8 REVislON "O"
i m Calculation of Dose Factors \ in the Ground Plane Pathway (Rf [D/Q] ) , j Rf(D/Q) $ K' K"(SF) DFG1 ((1-e' I ) / Al) units a m2 mrem /yr per uC1/sec i where Reference Table, R.G. 1.101 K' = A constant unit of conversion,106 pCl/ wCl. K" = A constant unit of conversion,8760 hr/yr
-SF = The shleiding factor, (dimension less, 0.7) E-15 A1 = The decay constant for the Ith radionuclide, sec-I t- = The exposure period,4.73 x 10 8sec (15 years)
DFG 1=-- The ground plane dose conversion factor for the Ith radionuclide (mrem /hr per pC1/m2) E-6 V
Reference:
The equation deriving Rf(D/Q) was taken from NUREG 0133, Section 5.3.1.2. n U EVISION "O"
Table 4.4-16 -(]d y Dose Factors Ground Plane Pathway - T. Body Skin Cr-51 4.65E6 5.5E6 Mn-54 1.39E9 1.63E9 Fe-55 0 0 Fe-59 2.73E8 3.21E8 Co-58 3.79E8 4.44 E8 Co-60 2.15E10 2.53E10 N1-63 0 0 Zn-65 7.47E8 8.57 E8 Rb-86 8.98E6 1.02E7 Sr-89 2.17E4 2.52E4 Y-91 1.07E6 1.21E6 .
~
' Z r-9 5 O 2.45E8 2.84E8
- Nb-95 1.41E7 1.66E7 Ru-106 4.22E8 3.07E8 ~,-
Ag-llom 3.44E9 4.02E9 Te-125m 1.53E6 2.13E6 +
- Te-127m . 9.17E4 1.08E3 Te-129m - 1.98E7 2.31E7 l-131 1.72E7 2.08E7 Cs-134 6.85E9 ' 8.0E9 Cs-136 1.51E8 1.72E8 Cs-137 1.03E10 1.20E10 Ba-140 2.06E7 2.35E7 Ce-141 - 1.37E7 - 1.54E7 Ce-144 6.95E7 8.0$E7
!4-143 0 0 M-147 8.40E6 -1.01E7 i 1
[ v geyson"c" l
/ CALCULATION OF LIQUID EFFLUENT ADULT INGESTION-(j DOSE FACTORS Ai r - 1.14E5 (21BFj + 5 BIj) DFj l
Ajy - Composite dose parameter for the total body or critical or l of an adult for nuclide, i, for all appropriate pathways, gan mrem /hr per gi/ml 1.14E5 = units conversion factor 106 pci/ sci x 10 3ml/kg + 8760 hr/yr l BFj - Bioaccumulation factor for nuclide, i, in fish, pCi/kg per pCi/L, from Table A-1 of Regulatory Guide 1.109 (Rev.1) or Table A 8 of Regulatory Guide 1.109 (original draft).- BIj - Bioaccumulation factor for nuclide, i, in invertebrates, pCi/kg per pC1/L, from Table A-1 of Regulatory Guide 1.109 Table A 8 of Regulatory Guide 1.109 (original draft)(.Rev. 1) or DFj - Dose conversion factor for nuclide, i, for adults in
- pre-selected : organ, r, in mrem /pci, from Table E-Il or j Regulatory Guide 1.109 (Rev. 1) or Table A-3 of Regulatory Guide 1.109 (original draft).
.m,
Reference:
The equation for Saltwater sites from NUREG 0133, Section 4.3.1,
~
where Uw/0w .0 since no drinking water pathway exists. p q-
-"~
REVISKN 13 l
(] Table 4.4-17 C/ Llquid Effluent - Adult irgestion Dose Factors Nuclide Bone Liver Thyroid T. Body Kidney g GI-LLI H-3 ND 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 NO-24 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57 E-1 4.57 E-1 4.57E-1 Cr-51 ND ND 5.58E0 3.34E0 1.23E0 7.40E0 1.40E3 Mn-54 ND 7.06E3 1.35E3 ND 2.10E3 ND 2.16E4 Mn-56 ND 1.78E2 3.15El ND 2.26E2 ND 5.67E3 Fc-55 5.11E4 3.53E4 8.23E3 ND ND 1.97E4 2.03E4 Fo-59 8.06E4 1.%E5 7.27E4 ND ND 5.30E4 6.32E5 Co-58 ND 6.03E2 1.35E3 ND ND ND 1.22E4 Co-60 ND 1.73E3 3.82E3 ND ND ND 3.25E4 N1-63 4.96E4 3.44E 3 1.67E3 ND ND ND 7.18E2 N1-65 2.02E2 3.31El 1.20E1 ND ND ND 6.65E2 Cu-64 ND 2.14E2 1.0!E2 ND 5.40E2 ND 1.83E4
65 ~ J.61E5 5.13E5 2.32E5 ND 3.43 E5 ND 3.23E5 i ;-69 3.43E2 6.56E2 4.56El ND 4.26E2 ND 9.85El Br-83 ND ND 7.25 E-2 ND ..ND ND 1.04 E-1 Br-84 ND ND 9.39E-2 ND ND ND 7.37E-7 Br-85 ND ND 3.86E-3 ND ND ND LE-18 Rb-86 ND 6.24E2 2.91 E2 ND ND ND 1.23E2 Rb-88 ND 1.79E0 9.49E-1 ND ND ND 2.47 E-I l Rb-89 ND 1.19E0 8.34 E-1 ND ND ND 6.89E-14 Sr-89 4.99E3 ND 1.43E2 ND ND ND 8.00E2 Sr-90 1.23E5 ND 3.01 E4 ND ND ND 3.55E3 Sr-91 9.!8El ND 3.71E0 ND NO ND 4.37E2 Sr-92 3.48El ND 1.51E0 ND ND ND 6.90E2 Y-90 6.06E0 ND 1.63E-1 ND ND ND 6.42E4 Y-91rn 5.73E-2 ND 2.22E-3 ND ND ND 1.68E-1 Y-91 8.88El ND 2.37E0 ND ND ND 4.89E4 Y-92 5.32E-1 ND 1.56E-2 ND ND ND 9.32E3 Y-93 1.69E0 ND 4.66E-2 ND ND ND 5.35E4 -
Zr-95 1.59El 5.llE0 3.46E0 ND 8.02E0 ND 1.62E4 17 8.81 E- 1 1.78E-1 8.13E-2 ND 2.68 E-1 ND 5.51E4 m -* * * * ' '" E N'
' ', , .) *. 5
, v
Tablo 4.4-17 Liquid Effluent - Adult Ing: tion Do00 Fccters gligg gggg Liggg T. Body IhEIgid Eidney Lung E-LLI b-95 4.47E2 2.49E2 1.34E2 ND 2.46E2 ND 1.51E6 No ND 9.05E-4 1.72E-4 ND 2.055-3 ND 2.10E-3 Tc-99a' 1.30E-2 3.66E-2 4.66E-1 ND 5.56E-1 1.79E-2 2.17E1 Tc-101 1.33E-2 -1.92E-2 1.88E-1 ND 3.46E-1 9.81E-3 5.77E-14 Ru-103 1.07E2 ND 4.60E1 ND 4.07E2 ND 1.25E4 Au-105 8.89E0 ND 3.51E0 ND 1.15E2 ND 5.44E3
-Ru-106 1.59E3 ND 2.01E2 ND 3.06E3 ND =1.03E5 Ag-110s. 1.57E3 .1.45E3 1.33E1 ND 2.85E3 ND 5.91ES:
Sb-124- 2.77E2 5.23E0 1.09E2 6.70E1 ND 2.15E2 7.83E3 ' Sb-125 2.20E2 2.37EO 4.41E1 1.95E1 ND 2.30E4 1.94E4 Sb-126 1.13E2 2.31E0 4.09E1 6.95E1 ND 6.95E1 9.27E3 To-125m 2.17E2 7.86E1 2.91E1 6.52E1 8.82E2 ND 8.66E2-To-127a 5.48E2 1.96E2 6.68E1 1.40E2 2.23E3 ND 1.84E3 70-127 8.90E0 3.20E0 1.93E0 6.600') 3.63E1 ND 7.03E2 Te-129s 9.31E2 3.47E2 1.47E2 - 3.20E2 3.69E3 ND 4. 6 9 E3 -- 7o-129 2.54E0 9.55E-1 6.19E-1 1.95E0 1.07El ND 1.92E0 's C .-131a 1.40E2 6.85E1 5.71E1 1.08E2 6.94E2 ND 6.80E3 ro-131 1.59E0 6.66E-1 5.03E-1 1.31E0 6.99E0 ND 2.26E-1 Te-132 2.04E2 1.32E2 1.24E2 1.46E2 ~ .27E3 1 ND 6.24E3 I-130 3.968"- .1.17E2 4.61El 9.91E3 1.82E2 ND 1.01E2: I-131 2.1' 3.12E2 1.79E2 1.02E5 5.35E2 ND 8.23E1
- I-132 1.06 2.85E1 9.96E0 9.96E2 4.54E1 ND 5.35E0 .
'!-133 7.54E1 1.30E2 3.95E1 1.90E4 2.26E2 ND -1.16E2 I-134= 5.56M .1.51Et 5'.40E0 2.62E2 2.40E1 ND 1.32E-2 I-135 2.31E1 6.08E1- 2.24E1 4.01E3 9.75El ND 6.87E1 Cs-134 6.84E3 1.63E4 1.33E4 ND 5.27E3 1.75E3 2.85E2 Cs-136 7.16E2- 2.83E3 2.04E3 ND 1.57E3- 2.16E2 3.21E2
-Cs-137 8.78E3 1.20E4- 7.85E3 ND- 4.07E3- 1.35E3 2.32E2-Cs-138 - 6.07EO- 1.20E1 5.94E0 ND 8.81EO. 8.70E-1 -5.12E-5 Ba-139- 7.85E0- 5.59E-3 2.30Z-1 ND 5.23E-3 3.17E-3 1.39E1-Ba-140- 1.64E3 2.06E0 1.0SE2 MD 7.02E-1 1.18E0 3.38E3 Ba-141 3.81E0 3.69E-3 1.29E-1 ND 2.68E-3 1.63E-3 1.80E-9
,Ba-142 1.72E0 1.77E-3 1.08E-1 ND 1.50E-3 1.00E-3 2.43E-18
- N -140 1.5750 7.94E 2.10E-1 ND ND ND 5.83E4
)
wa-142 8.06E-2 3.67E-2 9.13E-3 ND ND ND 2.68E2 IVS10N 8
,m (), Table 4.4-17 Liquid Effluent - Adult ingestion Dose Factors Nucilde Bone Liver T. Body Tleyroid Ljjng Gl-LLI Kidne_t Cc-141 3.43E0 2.32E0 2.63E 1 ND 1.08E0 ND 8.86E3 CG-143 6.04E 1 4.46E2 4.94E-2 ND 1.97 E 1 ND 1.67E6 Cc-140 1.79E2 7.47El 9.39E0 ND 4.43El ND 6.0lL4 Pr-143 3.79E0 2.3200 2.87E 7 ND 1.34 EC ND 2.34E4 Pr-144 1.90E 2 7.87E-3 9.64 E-4 ND 4.44E-3 ND 2.73E 9 Nd-147 3.96E0 4.38E0 2.74E-1 ND 2.68E0 ND 2.20E4 W-187 9.16E0 7.6GEO 2.68E0 fl0 ND ND 2.31 E' Np-239 3.3352 3.47 E-3 1.91E-3 ND 1.08 E-2 Nd 7.llE2
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9 0 . SECTION 5.0 ENVIRONMENTAL MONITORING O REV!SION "O"
- - . . . . .- _ -_ ._._ - - . . - _ . - _ . - _ ~
t Table 5.1-1 Environmental Radiological Monitoring - Stations Locations DIRECTION DISTANCE STATION LOCA110N FROM PLANT BOM PLAN 1 (t[1 C04 State Park Old Dam on River ENE 6,3 near road intersection C07 Crystal River Public Water Plant ESE 7.5 C09 Fort Island Gulf Beach S 3.2 C10 Indian Waters Public Water Supply ESE 5.9 C13 Houth of Intake Canal WSW 3.4 l Cl4H Head of.Ofscharge Canal NW 0.1 Cl4M : Midpoint of Discharge Canni W. 1.2
-Cl4G Discharge Canal at Gulf of Mexico W 2.8 !
C18 - Yankeetown City Well N 5.2 C19 NW Corner State Roads 488 & 495 ENE 8.5 ,
~
C29 Discharge Area N 2.0 C30 Intake Area WSW 3.6 C40 Near N.E. Site Boundary E 3.5 near excavated pond & pump station C41 Onsite Meteorological tower SW 0.4 ! C46 North Pump Station N 0.4 C47 Office of Radiation Control, Orlando ESE 67 C48A Onsite North of CR 4 & 5 N 0.8 ; C48B Onsite NNE of CR 4 & 5 NNE 0.8 O
. : 80 -
REVIS ON 14 _j
i TABLE 3.1-2 l RING TLDs (INNER RING) LOCAT10N DIRECTION, DLSTANCE (Pt.) i C27 W 3400 ; C60 N 4400 C61 NNE 4400 , C62 NE $300 C63 ENE 4400 C64 E 4400 C65 ESE 1740 ! C66 - SE 1600 C67 SSE 1430 C68 'S 1500
. . ~ . .
(A C69 SSW i780 C41 SW 2l00
~
C70 WSW 4400 ' C71 WNW 3500 C72 NW 2400 C73 NNW 20(. g REfS!CN J i 81-l' x
TABLE 5.1-3 RING TLDs (5 MILE RING) facarios DitzcuQE- DIS"Mct imi 1 C18 N 5.2 C03 NNE 5.3 C04 NE '3 C74 ENE $.5 C75 E 4.2 - C76 ESE 5.4
-C06 SE 3.5 C77 SSE 3.2 C09 5 3,2 C78 WSW 4.1 C140 W 2.8 C01. NW 4.9 C79 NNW 5.0
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OI . SecriON 6.0
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ADMWISTRATJyE CONTROLS . O
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6.1 gricin and purggse of the OffsitLQose calculation Manual The Offsite Dose Calculation Manual was developed to support the implementation of the Radiological Effluent Technical Specifications O reautred by 10 CFR 50, Appendix 1, and 10 r,fR 50.36. The purpose of the manual is to provide the NRC with sufficient information relative to effluent monitor setpoint calculations, effluent related dose calculations, and environmental monitoring to demonstrate compliance with the-fadiological,EffluentTechnical-Specifications. en. M k < 6.2 Channes , gg,, f.gf 9. / , It is recognized that changes to the Offsite Oose Calculation Manual may be required during the operational life of Crystal River Unit 3. All changes shall be reviewed and approved by the PRCs arior to implementation. The NRC shall be informed of all changes to t1e 00CM by providing a description of the change (s) in the first Semiannual Radioactive Effluent Release Report following the date the change became effective. A m /. </ n u ia... pA,~ ! m d..~ <. 4, L e 0(#' .eLu *. &n ~.J:/ . .J n 6* J k A .GJD e f;L sfu.s 15~ N "<m de'*a * * **N ha lw* l e f 6'" 6.3 Review / f v/. ' In addition to the change review in 6.2 above, the NGRC shall n ziew the ODCM and its implementing procedures at least once per 24 months. l 6.4 Unolanned..Reles.ses In order- to better ensure that Technical Specification 6.9.1.5.d and
\ 6.5.1.6.k are met, the following definition of " unplanned release" was developed. This definition should be used as " guidance only."
An ' UNPLANNED RELEASE' is:
- 1) A release of radioactive waste to the plant environs which has not '
been evaluated and released in accordance with approved procedures and Technical Specifications. Radioactive waste in this context means radioactive material that is awaiting evaluation before being released in a controlled fashion. This includes plant conden: ate, and the contents of all of the waste tanks (i.e., ECST's, LSS1 3, WGOT's,SDT.1). Examples: Releasing the wrong waste tank, or sampling the wrong tank and releasing the correct tank;
- 2) A release of radioactive material through - a designated effluent pathway, which is due to equipment failure or human error, and causes actuation of an effluent monitor warning alarm; i.
O E REVISION 15 i -
. = . = = . - = - .- - - _ . _-
- 3) Any sustained release of gaseous radioactive material from the RCA, but not through a normal effluent pathway, due to equipment f ailure or human error, which exceeds 1 HPC, restricted area.
7 C Examples: Releases from the RB equipment hatch RCA in excess of I HPC, restricted area. Releases l' rom the 10 RCA in excess of I HPC restricted area. Not Examples of Unplanned Releases: Short term increases in effluent monitor count rate due to sampling and analysis aethities. Controlled releases due to planned maintenance activities. Minor transient releases due to normal plant operations, such as waste processing or power changes which may cause an effluent monitor's warning alarm to actuate. Events within the Auxiliary Building which may actuate a process monitor alarm but are not of sufficient magnitude to actuate an effluent monitor warning alarm.
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ATTACHMENT 4 4-m+ we. .weq eus s.-s--w-p-g- -mes-,w - c-p.m ,sq-y., gg.ey-qs 993--g+qugw g -1;p rp w -w g qe,n S-w g e ma.*y- - pp---gr9 pwg.-- ge
PROCESS CONTROL PROGRAM g ,5 ' < ~ 3 bMki k L
4 3.0 REGULATORY REQUIREMENTS 3.1 Technical Specifications 3.1.1 Section 6.9.1.5d The Semiannual Radioactive Effluent Release Report shall include a summary of the quantities of radioactive solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the Process Control Program. 3.1.2 Section 6.14 Changes to the PCP:
- a. Shall be documented and records of reviews performed shall be retained as required by Technical Specification 6.10.3n. This documentation shall contain:
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s), and
- 2) A determination that the change will maintain the overall confromande of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.
- b. Shall become effective after review and acceptance by the PRC and the approval of the Director, Nuclear Plant Operations.
3.2 Code of Federal Regulations 10 CFR 61.56, " Waste Characteristics" (a) (2) Liquid waste must be solidified or packaged in sufficient absorbent material to absorb twice the volume of the l i n.u i d . (a) (3) Solid waste containing liquid shall contain as little free standing and noncorrosive liquid as reasonable, but in no case shall the liquid exceed 1% of the volume. (b) (2) Notwithstandingtheprovisionsof(a)(2)and(3), liquid wastes or wastes containing liquid must be converted into a form that contains as little free standing end l noncorrosive liquid as is reasonably achievable, but !n no case shall the liquid exceed 1% of the volume of waste when the waste is in a container designed to assure
- stability, or 0.5% of the volume of the waste for waste processed-to a stable form.
3
l l 3.3 Commitments 3.1.1 Waste Solidt fication Syst em S pec i f' :",6- - The solid rad aste system shall l,e used at all Lives in accordan:) with a Process Control Program to process wet r.vtit active wastes tte meut shil. ping and burial ground requirements. - Action: With the provisions of the Process Control Program not satisfied, suspend shipments of defectively processt1 or defectively packaged solid r adicar,tive wastos from the site. Surveillanca: Thc Process Control Progra'n shall be used to verify the solidification of at least. one representative test specimen form at least every tenth batch nf each type of wet radioactive wast I (e.g., filter aludges, spent resins, evaporator bottoms, and boric acid solutluns). L l l - 3a-l l l x- ~ - ~
- v. , - ,..,,..,...m.-.,a., . , , . . , , . _ , _ _ _ . , , , , , , _ . , , _ _ _ _ , - . , . , , ,
i 4.0 ADMINISTRATIVE CONTRo!,5 4.1 Responsibility / Revisions Changes to the process Control pregian (PCP) are the responsability of the f Manager, Site Nuclear Services. Technical Specification 6.14 stipulates the ' required approvals necessary to modify the Process Control Progran prior to implementing any changes to the process (See Section 3.2). , 4.2 Reporting 4.2.1 Changes to the PCP - i i- Major changes to the rariloactive waste treatment systems (liquid, gaseous - and solid) initiated by FPC shall rep;rted to the MC in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was r%lewed by the Plant Review Committee (PRC) or be included as part of t!.e annual FSAR update. A major change to a radioactive waste treatment system is any change which would alter the aoility of the plant or system ) to meet the requirementr. of 10 CRf 50, Appendix 1. The change to tFo system may be implemented upon review t.nd acceptanut by the PRC. 4 - 4.2.2 Nonconformances Reporting of noncot'9rmances with the requirements of the PCP are documented i on Problem Reports and controlled in accordance with the appropriate procedures, i 4.3 Documentation All documentation associated with t' e verification of the process Control
- Progias is controlled in .accordance with the appropriate implementing proce-dures.
4.4 Definitions Batch - (1) For sampling or processing, a batch is ;the largest homogeneous I volume of waste that has bee:, recirculated and controlled as per , the PCP.
.( ?) For solidification testing, a batch is taken to be a disposal container (i.e., 55 gallon drums, etc.) utilized in the solidi-fication of the waste.
- Solidification.- This process shall be the conversion of wet wastes into-a form that meets shipping and burial ground requirements.
f
-s- REV S ON 1 m __ - _. __ _ _ _ _ - . _ __
l 5.3.2 Test frequency 5.1.2.1 Process Test frequency l A process test solidification shall be made prior to f ull scale solidittcation l to determine ratios and additives as per section 5.3.1.1. ' 5.3.2.2 Solidt i 4 tion Test Frequency l The PCp shall be used to verify the solidification of at least one (1; terre-s*<tative test specimen f rom at least every tenth batch of each type of wet rt .'ioactive waste . t 5.3.3 Acceptance Criteria ; 4 #7 /43, ' Soliolfication Test Datch Verification progras,' stipiilo en the :" activittes and documentation necessary to verify acceptancs of susadified unste.
' The solidified watte acceptance criteria is verified by e
- s. Visually inspecting for defects in the structure.. '
- b. tiniformity in color and density.
- c. No free standing liquid (<0.5% of total waste volume)
- d. pree stand!Sg monolith,
- e. Af te r 24 hours from solidt ficatioa, the final cured puduct shall resist ,
penetration when probed by hand with a spatula or firm object ()S0 pan). ~1 i t If any portion of the specisen f ails to pass the - A;ceptance Cr' 6.14 the applicable actions of Section 5.3.4 must be met.
)
5.3.4 Corrective Action
- 4. If the initial setc specimen from a batch of waste fril' t o verify solidi- )
fication, representative test specimens from each consecutive batch of the same' type of wet weste shall be collected and tested unt'l at least-3 consecutive initial test _ specimens dominstre:t solidi f ica tion . The 3 process and/or additives shall be - modified at required, as i roviNd in , Section 4.1, to assure solidification of tsub' sequent L5tcheu of vaaie.
. t
- b. If any test specimen f ails to vertf y solidificction, the solidific at .on of I the batch under test shall be suspended until Luch time as add;tiona.1 test specisens can be obtained, alternate solidificat?,on pararetet 1 can be determined in accordance with the process Control progrn, and _a arNequent
- test verifies -solidification. -Solidification of the batch may -_ then be '
resumed -using- the alternative sciidificolon paraattets xistermined by the i process Control-_Progras. i
- c. With installed equipment inchable of meet /ng tha requirements of-i Section 3.3 or declared inoperable, re., tore the equipment of operable status or provide- for contract capabiliy to process wistos 35 necessary to _ satisfy all appitcable- transportation and dhpose requirements. .
, 1 -
r i REVISICN 02 :
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