ML20091J446
| ML20091J446 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 05/29/1984 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8406050514 | |
| Download: ML20091J446 (62) | |
Text
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DUKE POWER GOMPANY P.O. HOx 33180 011ARLOTTE, N.C. 28242 som re me (7 37 EK31
" " ~ " " ' " * " "
May 29, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414
Dear Mr. Denton:
In regard to your April 10, 1984 letter requesting additional infonnation, please find attached three (3) copies of Duke's responses to the questions pertaining to the Offsite Dose Calculation Manual for the Catawba Nuclear Station. Please be advised that Section C4.3, " Fuel Cycle Calculations,"
is incomplete. This information will be provided by June 4, 1984.
Very truly yours, f.b. fAA lff Hal B. Tucker ,
RWO/php Attachments cc: (w/ attachment)
Mr. James P. O'Reilly Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 (w/oattachment)
NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild, Esq.
Attorney-at-Law P. O. Box 12097 Charleston, South Carolina 29412 Palmetto Alliance 8406050514 040529 PDR ADOCK 05000413 h 00{
21351 Devine Street A PDR ,3 Columbia, South Carolina 29205 L-
Mr. Harold R. Denton, Director May 29,~1984 Page 2 i> 1 cc: l (w/o attachment)
Mr. Jesse L. Riley Carolina Environmantal Study Group i 854 Henley Place !
Charlotte, North Carolina 28207 )
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Subj ect: Figure C1.0.1 - Liquid Radwaste Treatment System
- Commenting Agency: Franklin Research Center Commsnt: Effluent environmental release points not clearly identi-fied; end discharge point not designated.
Response: Effluent environmental release points are designated on Figure 5.1.3 of Catawba's Technical Specifications. Duke Power Company feels that this would be a duplication of information already presented in a controlled document.
With this explanation, no action is planned.
The end discharge points of liquid waste releases will be clarified on Figure C1.0.1.
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Subject:
Figure C.l.0.2 - Gaseous Radwaste Treatment System Commenting Agency: Franklin Research Center Comment: Monitors are not designated for the following effluent streams; Hydrogen Monitors at the recombiners; Vent System; Containment Purge System; Auxiliary Building Ventilation; Fuel Storage Area Ventilation System.
Response: EMF designation will be added to each of the above loca-tions except " Hydrogen Monitors at the Recombiners." This is not an effluent stream and no EMF monitor is required by Catawba's Technical Specifications.
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Subject:
Page C-4, Liquid Release Rate Calculation Commenting Agency: Franklin Research Center Comment: Method not provided for determining the liquid radioactivi-ty concentrations for both batch and continuous releases.
(See Section 2 of the attached NRC Staff Positions.)
Response: The requirement to list the methods and procedures for obtaining a representative sample and performing analyses in accordance with the ODCM has been deleted from Catawba's Technical Specifications. No action is planned.
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Subject:
Page C-8, Liquid Radiation Monitors Commenting Agency: Franklin Research Center Comment: In liquid setpoint calculations, Licensee has not addressed the possibility of simultaneous releases. Also setpoints for service water and component cooling water are not provided.
Response: Section 1.0, page 1-1, " Release Rate Calculations" address-es the possibility of simultaneous releases by stating that such releases could be administratively controlled so that the limitations addressed on page lii would not be exceed-ed. Duke Power Company uses this same type of control (via station procedures) at both its Oconee and McGuire plants without any problems.
The service water monitor and component cooling water monitor are not longer considered to be effluent monitors and have been deleted from Catawba's Technical Specifica-tions (Tables 3.3-12 and 4.3-8).
With this explanation, no action is planned.
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Subject:
Page C-11, Gas Monitors Commenting Agency: Franklin Research Center Comment: Licensee should justify using only Xe-133 in the setpoint calculation.
Response: Historical data as well as Table 11.3.3.1 in the Catawba FSAR show that >80% of the curies found in gaseous effluents are from the radionuclide Xe-133. This information will be added to Section C3.2.
Comment: Setpoir,ts should be calculated for all monitors proposed by the Licensee in Table 3.3-13 of the RETS submittal.
Responne; Typical setpoint calculations for each gaseous effluent stream monitor listed in Table 3.3-13 of the RETS submittal are shown on page C-11 of the ODCM. With this explanation, no action is planned.
Comment: Also, possible simultaneous releases from effluent lines are not addressed.
Response: Section 1.0, page 1-1, " Release Rate Calculations" address-es the possibility of simultaneous releases by stating that such releases could be administrative 1y controlled so that the limitations addressed on page til are not exceeded.
Duke Power Company uses the same type of control (via station procedures) at both its Oconec and McGuire plants without any problems. No action is planned.
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Subject:
Page C-13 Liquid Effluent Commenting Agency: Franklin Research Center Comment: 1) Licensee should provide the basis for the near field dilution factor Dy (37.7).
Response: Basis for Dy will be added to pages C-12 and C-13.
Comment: 2) Units are not assigned for dose D and data provided in WB Table C.4.0-3.
Response: The units " mrem /hr per pCi/ml," inadvertently lef t off, will be added.
Comment: 3) According to NUREG-0133 (Section 4.3.1) only adult fish consumption will be considered. Can Licensee justify child fish consumption is more conservative?
Response: Section C4.3.1, Case 2, will be changed to reflect adult fish consumption.
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- Subj ect: Page C-12, Section C4.2.2.2 Page C-15, Section C4.3.2.2 Commenting Agency: Franklin Research Center Comment:
As method to obtain doses from Tritium is different from that for Iodine-131, Licensee should also include the method for the Tritium dose calculation in these sections.
Response
Historical data has shown that I-131 contributes >95% of the dose to the thyroid of an infant. The factor "1.05" allows the remaining 5%, of which Tritium comprises some 3%, to be considered. Additionally, should a user need to calculate the Tritium dose, the method is provided in Section 3.1.2.2. With this explanation, no action is planned.
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Page 15, Gaseous Effluents (C4.3.2.1)
Commenting Agency: Franklin Research Center Comment: The assumption that Xe-133 contributes 45% to the dose is inconsistent with Section C3.2 where Licensee assumes 100%
contribution (is) from Xe-133 for dose rate calculation (s).
Response: Section C3.2 deals with typical setpoint calculation for gaseous radiation monitors. As previously stated, >80% of the curies in a gaseous effluent release is from Xe-133, and therefore, radiation monitors are calibrated to the most abundant radionuclide. Section C4.3.2.1 deals with simplified dose estimates. IIistorically, approximately 45%
of the Gamma dose and Beta air dose is from the radio-nuclide Xe-133. No further action is planned.
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Subject:
Table C5.0-2, Sampling Locations Coaumenting Agency: Franklin Research Center Conunent: Licensee has not identified pathsays that are carresponding to the sampling locations listed in the Table, i.e., fish, broad leaf vegetables, etc.
Response: The column headings, inadvertently left off, will be added.
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Subject:
Liquid Dose Projection Commenting Agency: Franklin Research Center Comment: Licensee has not specifically addressed (that) site-specific monthly dose projection (s). To (will) be used for Liquid Radwaste Treatment Actions.
Response: Technical Specification 4.11.1.3.1 states that " Doses ....
shall be projected .... in accordance with ... the ODCM."
Section C4.1 of the ODCM states " Dose projections shall be performed using simplified dose estimates." No action is planned.
Comment: Licensee claimed in Section 3.2 that site-specific informa-tion will be provided, but it was not found in Appendix C.
Response: All necessary site-specific information required to perform l dose calculations has been provided in Appendix C. Exam-ple: Section C4.3.1 contains the method and site-specific information to calculate doses from liquid effluents. No action is planned.
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Subject:
Gaseous Dose Projection Comunenting Agency: Franklin Research Center Comment: Licensee has not specifically addressed (that) site-specific monthly dose projection (s). To (will) be used for gaseous radwaste treatment actions.
Response: Technical Specification 4.11.2.4.1 states that " Doses ...
shall be projected ... in accordance with ... the ODCM."
Section C4.1 of the ODCM states " Dose projections shall be performed using simplified dose estimates." No action is planned.
Comment: Licensee claimed in Section 3.2 that site-specific informa-tion will be provided, but it was not found in Appendix C.
Response: All necessary site-specific information required to perform dose calculations has been provided in Appendix C. Exam-ple: Section C4.3.2 contains the method and site-specific information to calculate doses from gaseous effluents. No action is planned.
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Subject:
Page 3-7, Total Dose Commenting Agency: Franklin Research Center Comment: 1) Licensee should specifically address how total dose will be calculated with respect to 40 CRF 190 Require-ments, i.e., a method to sum up various doses calculated in (the) ODCM.
Response: Section C4.3 will be revised to provide this information.
- 2) A method to calculate direct dose should be provided.
Section 3.1.3 is not adequate to address the direct dose.
Response: Section 3.1.3 states that the point kernel method was used to calculate offsite dose rates. Since the calculated dose rates to an individual are negligible, 40.01 mrem /yr, Duke Power Company feels that it is unnecessary to provide these calculations in this document. Duke Power Company also feels that it is unnecessary to routinely re-calculate these doses unless warranted by a change in plant operating conditions. Additionally, this section has been previously approved by the Franklin Research
! Center and the Nuclear Regulatory Commission for both our Oconee and McGuire plants. With this response, no action is planned.
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Subject:
Interlaboratory Comparison Program Commenting Agency: Franklin Research Center Comment: Licensee has not provided a description on (the) Inter-laboratory Comparison Program as committed in the RETS submittal. !-
Response: Section C5.0, page C-17, in the ODCM states that Duke Power Company participates in the Environmental Protection Agency's Environmental Radioactivity Laboratory Inter-comparison Studies (crosscheck) Program. Verbatim para- '-
graphs have been previously approved by the Franklin Research Center and the Nuclear Regulatory Commission for both our Oconee and McGuire plants. As Duke Power Company feels that this section is still adequate, no action is planned.
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Subject:
Figure Cl.0-1, Liquid Radwaste Treatment System Commenting Agency: Nuclear Regulatory Commission Comment: 1) Indicate route of off-normal steam generator blowdown and how it is monitored.
Response: Figure C1.0-1 will be revised to provide this information.
Comment: 2) Indicate how flows from liquid radwaste, conventional waste water treatment, nuclear service water, and low pressure service water are monitored for flow, moni-tored or continuously sampled for radiation, and merged prior to discharge.
Response: Figure C1.0-1 will be revised to provide this information.
Comment: 3) Identify all environmental release points.
Response: Liquid effluent environmental release points are designated on Figure 5.1-3 of Catawba's Technical Specifications.
Duke Power Company feels that this would be a duplication of information already presented in a more tightly con-trolled document. With this explanation, no action is planned.
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Figure C1.0 Gaseous Treatment System Commenting Agency: Nuclear Regulatory Commission Comment: 1) Indicate route of off-normal steam generator blowdown vent exhaust.
Response: Figure C1.0-2 will be revised to provide this information.
Comment: 2) Indicate where hydrogen and oxygen are monitored in the waste gas holdup system.
Response: Oxygen and hydrogen monitors are located upstream and downstream of the hydrogen recombiners. Since these monitors are not on an effluent stream and do not monitor radiation, Duke Power Company feels that they should not be presented in a document which deals with off-site doses.
Comment: 3) Indicate (the) location of flow and radiation monitors and continuous samplers and type of continuous samplers for effluents from the waste gas holdup, condenser evacuation, auxiliary building ventilation, fuel storage area ventilation, containment purge, contain-ment ventilation, and vent systems.
Response: EMF designations will be added to each of above locations.
Comment: Identify all environmental release points.
Response: Gaseous ef tluent environmental release points are designat-ed on Figure 5.1-4 of Catawba's Technical Specifications.
Duke Power Company feels that this would be a duplication of information already presented in a more tightly con-trolled document. With this explanation, no action is planned.
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Subject:
Page C-4, Liquid Release Rate Calculations Commenting Agency: Nuclear Regulatory Commission Comment: Explain why simultaneous batch releases need not be consid-ered, or provide method for considering simultaneous batch releases.
Responce: Section 1.0, page 1-1, addresses the possibility of simul-taneous releases by stating that such releases could be administrative 1y controlled so that the limitations addres-sed on page til would not be exceeded. Duke Power Company has successfully used this type of control (via station procedures) at both its Oconee and ifcGuire plants. With this explanation, no action is planned.
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Paae C-5, Liquid Release Rate Calculations Commenting Agency: Nuclear Regulatory Conaission Comment: Provide method of considering releance of radioactive materials from the conventional waste water treatment (system) and how this is to be coimidered in the methodo1-ony for determining the undiluted effluent flow (f) in C2.1.1, Nesponset Section C2.1.2 states that three of the water sources that normally flow into the conventional waste water treatment I system will be diverted to the Radwante Treatment System '
should these water sources become radir> active. At this time their flow (t) will be the undiluted flow referenced !
in Section C2.1.1. With this explanation, no action is planned.
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Page C-7, Caseous Release Rate Information Commenting Agency: Nuclear Regulatory Commission Comment: Provide method for considering releases of radioactive ;
materials from the steam generator blowdown vent.
Response An off-normal modo of operation is provided to accommodate '
situations when it is not possible to either vent the steam :
! generator blowdown tank to the "D" heater or pump the '
! liquid to the condensate system. In this mode of opera-tion, the steam generator blowdown tank releases steam to l '
l the atmosphere while its liquid level is maintained by directing the pumped liquid to the Turbine Building sump.
Radioactivity Icvels in the Steam Generator Blowdown System are monitored for activity by EMF-34. If activity should l be detected, each blowdown flow control valve, the atmo-l spheric vent, and the valve to the Turbine Building Sump will close. Blowdown can only be continued by venting the !
! steam to "D" heater and pumping the liquid to the conden-sate system. As a result, no gaseous radioactive m.iterials can be released via this pathway. With this explanation.
{ no action is planned.
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Subject:
Page C-8, Liquid Radiation Monitor Setpoints Commenting Agency: Nucicar Regulatory Commission Comment: Address how simultaneous reicases from liquid radwaste and conventional waste water treatment will affect the setpoint of the radiation monitor for the Liquid Radwaste Effluent Line.
Response: Section C2.1.2 states that if any of the conventional waste water treatment water sources become radioactive, they can be diverted to the Radwaste Treatment System and be dis-charged through the Liquid Radwaste Effluent Line.
Section 1.0, page 1-1, " Release Rate Calculations" addresses the possiblity of simultaneous releases, by stating that such releases could be administrative 1y controlled so that the limitations addressed on page 111 are not exceeded. Duke Power Company uses this same type of control (via station procedures) at both its Oconee and McGuire plants without any problems.
With this explanation, no action is planned.
Subject:
Page C-11, Gaseous Radiation Monitor Setpoints Commenting Agency: Nuclear Regulatory Commission Comment: 1) Provide basis for using f = 151,000 cfm Response: This is only a typical flowrate. Exact flowrates that will be used to calculate radiation monitor setpoints will be listed in station procedures. The wording associated with this 151,000 cfm flowrate, as well as this 28,000 cfm flowrate listed on page C-11, will be changed to clarify this item.
Comment: 2) Provide method fcr considering simultaneous releases.
Response: Section 1.0, page 1-1, addresses the possibility of simul-taneous releases by stating that such releases could be administratively controlled so that the limitations ad-dressed on page lii would be not be exceeded. Duke Power Company has successfully used this type of control (via Station Procedures) at both its Oconee and McGuire plants.
With this explanation, no action is planned. '
Comment: 3) Provide method for determining setpoints for the waste gas holdup system, containment purge, and containment ventilation monitors.
Response: The monitor setpoint calculations are fully described in Section C3.2, page C-10. No action is planned.
Comment: 4) Provide method for determining setpoints for the radiation monitor on the steam generator blowdown to terminate venting to the atmosphere through the Turbine Building vent.
Response: No gaseous radiation monitor on the steam generator blow-down vent is required by Catawba's Technical Specifica-tions, since this is not a gaseous effluent release point.
Therefore, no setpoint calculation is given. Picase see our response to your comment on page C-7, Gaseous Release Rate Calculations. No action is planned.
Subject:
Page C-4.0, Dose Calculation Commenting Agency: Nuclear Regulatory Commission l Comment: 1) Provide a numbered and captioned figure showing the site boundary and the unrestricted area boundary for gaseous and liquid effluents.
Response: This information is provided in Figures 5.1-3 and 5.1-4 of Catawba's Technical Specifications. This placement of information in the Technical Specifications has been previously approved by the Franklin Research Center and the Nuclear Regulatory Commission for both our Oconce and ticGuire plants. As Duke Power Company feels that this would a duplication of information already presented in a more tightly controlled document, no action is planned, i Comment: 2) The ODCt! should include a table that contains the following information for each sector: sector, the distance to the controlling receptor locaM on, the_
pathway of exposure, the age group, the X/Q and D/Q j values. I 1
Response: a. Sector information was inadvertently left off Figure (
C5.0-1. It will be added. J
- b. The distance to the controlling receptor location is l listed in the Section of the ODCt! that bases its l calculations on this information, i.e., Section l C4.3.2.2, page C-15. Duplication of information is l unnecessary. No action is planned. l 1
- c. Exposure pathway information is also listed in this manner, i.e., Section C4.3.2.2, page C-15. Again this would be a duplication of information and is unneces-sary. No action is planned.
- d. Age group information has been provided in the assump-tion for each calculation where this information is used, i.e., Section C4.3.2.2, page C-15. Again this would be a duplication of information and is unneces- ,
sary. No action is planned. I
- c. X/Q and D/Q vt. lues are listed on Tables C-4.0-1 and C-4.0-2. No action is planned.
Comment: Provide the date of the land-use census that was used in identifying the controlling receptor locations. 1 1
Response: This information will be added to Section C5.0, page C-16. i l
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Subject:
Page C-5.0, Radiological Environmental Monitoring Commenting Agency: Nuclear Regulatory Commissi,n Comment: 1) Tables C5.0-1, 2 and 3 List the number of samples, but do not provide the locations for all samples. The ;
tables should contain the following columns: (1) !
exposure pathway and/or sample; (2) criteria for i selection of sample number and location; (3) sampling l and collecting frequency; (4) sample location number l (it should refer to a figure in the ODCH); (5) location (distance and direction); (6) type and frequency of analysis.
Response: a) Sample locations are provided in these tables, i.e.,
Table C5.0-2 states that sample location 213 is located 7.5 m ESE at the Fort Hill water supply. No action is planned, b) Exposure pathway and/or sample information is also provided on Table C5.0-2. With the exception of adding the column heading addressed elsewhere in these re-sponses, no action is planned.
c) Section C5.0, page C-16, states that the Radiation Environmental Monitoring Program shall be conducted in accordance with Technient Specification 3/4.12. The cr.teria for selection of sample number and location is based on this specification. With this explanation, no action is planned.
d) Sample locations are listed by number on Tables C5.0-1 and C5.0-2. These numbers correspond to Figure C5.0-1. I No action is planned.
e) Tables C5.0-1 and C5.0-2 do provide sampic location and distance information, i.e., Table C5.0-2 states that sample location 213 is 7.5m ESE at the Fort Hill supply. No action is planned.
f) Type and frequency of analysis has been provided in Table C5.0-3. No action is planned.
Comment: 2) A foldout figure is needed for Figure C5.0-1 (in order) to make it readable.
Response: Ten copies of a two-color 85 x 17 foldout were included with the ODCH submittal. Additional copies will be submit-ted with these responses.
Subject:
Page C5.0, Radiological Environmental Monitoring (Continued)
Cosament: 3) Presimably the methodology described in the ODCH will be impicmented via computer codes. The computer codes sho 1d be verified. After the codes are verified, pr ovide a reference (individual or company name, title
- o. document, and date) in the ODCH to document the validation of the codes.
Responses As stated in the introduction to the ODCH, the programs "LADTAP" and "GASPAR", written and distributed by the Nuclear Regulatory Commission, will normally be used to determine done assessments. The method outlined in the ODCH is merely a hand calculational method to be used only when "LADTAP" and/or "0ASPAR" is unavailable. Ilowever, we do use our own computer programs to calculate release rates and prepare reports required by Regulatory Guioe 1.2.1.
The codes are benchmarked (verified by hand calculations) by Duke Power Company personnel prior to irrplementation.
Benchmarks are also audited by our QA department and documented. These records are available for NRC audit.
Therefore, as a result of our tight controls, Duke Power Company does not feel that this information need be duplicated in the ODCH.
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Subjects pageC6/C2.2,R76andIIfdValues l
Commenting AAency: Nuclear Mcgulatory Consission l Comment: Identifythelocation(distanceanddirection)forthe576 '
and li/Q values presented for awscous release rate calculations. !
Response The distance and direction information will t>e added to j those sections in question. '
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Subject:
C4.0, Dose Calculation /tfeteorology Commenting Agency: Nuclear Hegulatory commission !
Comment Provide a description of the atmospheric dispersion and._ ,
deposition models and methodology used to calculated X/Q and D/Q values for inclusion in the ODCit. Also identify the source of meteorological data and period of record used for these calculations.
L Response The X7ij and li[t} values provided in Tables C4.0-1 and C4.0 2 I were generated using the coinputer program X0QlXX} in NURE0/ l CR 2919 and all assumptions outlined in NRC Hegulatory !
Guide 1.111 USNRC, 1977. The period of record was Decem- ,
her 17, 1975 - December 16, 1977. With this explanation, no action is planned, f l
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Subject:
TablesC4.01(X[Q)andC4.02(57Q)
Commenting Agency: Nuclear Regulatory Commission Comment: The XTQ and D76 values presented in these tables differ significant ly from those calculated by the staf f. For example, at 1.0 mi. In the north northeast sector. TC4.0-1 indicates a k/jLvalue of 2.6E-07, while the staff has calculated a D/Q value of 7.7E-06_at the same location.
Similarly, TC4.0-2 indicates a D/Q value of l.,7E-09 at 0.S mi. in the south sector, while the statf's D/Q value for this location is 7.2E 08. These large differences are probably attributab!c to characterization of the aerody-namics of the releases from the plant vents. The staff has assumed that these relcanes occur at ground level with ~
mixing in the turbulent wake of plant structures. TheX/4 and D/Q values presented in TC4.0 1 and TC4.0-2 are typical of those calculated for a mixed-mode releanc. The staff did not accept the rationale presented in the Section 2.3.5 of the FSAR for deviating from the H.0, 1.111 position that states "For effluents released from points less than the weight of adjacent solid structures, a ground-level release should be assumed." The top of the vent stacks et Catawba are about 4m below the top of the reactor building.
Another source of dif ference is adjustments to the straight-line airflow model to consider spatial and temporal varia-tions in airflow. Either provide revised tables of f/d and I174 values reflecting assumptions used by the staff, or 4
provide additional information (boyond that aircady re-viewed by the staff) to substitute the R/d and 07d values proposed in the ODCtt.
Nesponse Tables C4.0 1 and C4.0 2 have been revised using the computer program X0QIXX) (NUNE0/CR 2919) and all assumption outlined in NHC Hegulatory Guide 1.111 (USNHC, 1977).
ADDITIONAL NRC COMMENTS TO GDCM REQUESTED BY TELECON
Subject:
Table C4.0-3 Commenting Agency: Ed Branagan, USNRC Comment: Remove comment on Table C4.0-3 which states " Methodology for Table provided by: M. E. Wrangler, RAB:NRR:NRC on 3/17/83," and take responsibility for values.
Response: Duke Power Company will only stand behind the values generated using the subroutine and site-specific informa-tion but not the subroutine itself as it was provided by the NRC. A copy.of the letter and subroutine is provided.
No action is planned.
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APPENDIX C CATAWBA NUCLEAR STATION SITE SPECIFIC INFORMATION
e - A-42 - + -- - O APPENDIX C - TABLE OF CONTENTS Page C1.0 ' CATAWBA NUCLEAR STATION RADWASTE SYSTEMS . . . . . . . . C-1 C2.0 RELEASE RATE CALCULATION . . . . . . . . . . . . . . . . C-4 C3.0 RADIATION MONITOR SETPOINTS . . . . . . . . . . . . . . C-8 C4.0. DOSE CALCULATIONS . . . . . . . . . . . . . . . . . . . C-12 C5.0' RADIOLOGICAL ENVIRONMENTAL MONITORING . . . . . . . . . C-16 f'
9
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- C1.0 CATAWBA NUCLEAR STATION RADWASTE SYSTEMS
^
C1.1 LIQUID RADWASTE PROCESSING
-The liquid radwaste system at Catawba Nuclear Station (CNS) is used to collect
!and treat fluid chemical and radiochemical.by products of unit operation. The system produces efflucents which can be reused in the plant or discharged in small, dilute quantities to the environment. The means of treatment vary with waste type and desired product in the various systems:
A) Filtration - All waste sources are filtered during processing. In some
. cases, such as the Floor Drain Tank (FDT) Subsystem of the Liquid Waste (WL)-System, filtration may be the only treatment required.
B) Adsorption - Adsorption of halides and organic chemicals by activated charcoal (Carbon Filter) is used primarily in treating waste in the
. Laundry and Hot Shower Tank (LHST) Subsystem of the WL System. FDT waste may also.be treated by this method.
C) Ion Exchange - Ion exchange is used to remove radioactive cations from solution, as in the case of either LHST or FDT waste in the WL System
~ af ter removal of organics by carbon filtration (adsorption). Ion exchange-is also used in. removing both cations (cobalt, manganese) and anions (chloride, fluoride) from evaporator distillates in order to purify the distillates for reuse as makeup water. Distillate from the Waste' Evaporator in the WL System and the Boron Recycle Evaporator in the Boron Recycle System (NB)_can be treated by this method, as well as FDT, LHST waste, and letdown.
D). . Gas Stripping'- Removal of gaseous radioactive fission products is accomplished ~in both the WL Evaporator and the NB Evaporator.
E) Distillation - Production of pure water from the waste by boiling it
.away from the' contaminated solution which originally contained it is accomplished by both evaporators. Proper control of the process'will y yield water which can be reused for makeup. Polishing of this product can be achieved by ion exchange as pointed out above.
F) - Concentration.- In both the WL and NB Evaporators, dissolved chemicals are concentrated in the lower shell as water is boiled away. In the case of the WL Evaporator, the volume of water containing waste chemicals and radioactive-cations is reduced so that the waste may be more easily and cheaply solidified and shipped for burial. In the NB Evaporator, the dilute boron is concentrated to 4% so that it may be reused for makeup to the. reactor coolant. system.
Figure C1.0-1 is a schematic representation of the liquid radwaste system at Catawba.
, d' r
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p Table C1.0-1 ABBREVIATIONS Systems:
CM - Condensate System KC - Component Cooling NB - Boron Recycle RL - Low Pressure Service Water RN - Nuclear Service Water System WC - Conventional Waste Water Treatment WL - Liquid Waste Recycle WP - Turbine Building Sump WS - Nuclear Solid Waste Disposal Tanks:
BA - Boric Acid Tank FDT - Floor Drain Tank LHST - Laundry and Hot Shower Tank MST - Mixing and Settling Tank NCDT - Reactor Coolant Drain Tank RHT - Recycle Holdup Tank RMT - Recycle Monitor Tank RMWST - Reactor Makeup _ Water Storage Tank SGDT --Steam Generator Drain Tank VUCDT - Ventilation Unit Condensate Drain Tank WDT - Waste Drain Tank WEFT - Waste Evaporator Feed Tank WMT - Waste Monitor Tank TABLE C1.0-1
C1.2 GASEOUS RADWASTE SYSTEMS The_ gaseous waste disposal system for Catawba is designed with the capability of processing the fission-product gases from contaminated reactor coolant fluids resulting from operation. The system shown schematically in Fig. C1.0-2 is_ designed to allow for the retention, through the plant lifetime, of all the gaseous fission products to be discharged from the reactor coolant system to the_ chemical and volume control system or the boron recycle system, to limit the need for intentional discharge of radioactive gases from the waste gas
' holdup tanks. Thus, the only unavoidable sources of low-level radioactive gaseous discharge to the environment will be from periodic purging operations of the containment, from the auxiliary building ventilation system, and through the secondary system air ejector. With respect to the.former, the potential contamination is expected to arise from uncollectable reactor coolant leakage.
With respect to the. air, ejector, the potential source of contamination will be from leakage of the reactor coolant to the secondary system through defects in steam generator tubes. The gaseous waste disposal system includes two waste gas compressors, two catalytic hydrogen recombiners, six gas decay storage tanks for use during normal power generation, and two gas decay storage tanks for use during shutdown and startup operations.
C 1. 2.' 1 Gas Collection System The gas collection system combines the waste' hydrogen and fission gases from the volume control' tanks and that from the boron recycle gas stripper evaporator produced during normal operation with the gas collected during the shutdown degasification (high percentage of nitrogen) and will cycle it through the catalytic recombiners to convert all the hydrogen to water. After the water vapor is removed, the resulting gas stream will be transferred from the recom--
biner.into the' gas decay tanks, where the accumulated activity may be contained
- j. in six approximately equal parts. From the decay tanks the gas will flow back to the compressor suction-to complete the loop circuit.
C1.2.2 Containment and Auxiliary Building. Ventilation Nonrecyclable reactor coolant leakage _ occurring either inside the containment or inside the auxiliary building will generate gaseous activity. Gases result-ing from leakage inside the containment will be contained until the containment i air is released through the VQ or VP system. The containment atmosphere will j be discharged through a charcoal adsorber and a particulate filter prior to
!. release to the atmosphere.
Gases resulting from leakage inside the auxiliary building are released, with-out further decay, to the atmosphere via the auxiliary building ventilation
' system. The ventilation exhaust from potentially contaminated areas in the auxiliary building is normally unfiltered. .However, on a radiation monitor l ..
alarm, the exhaust is passed through charcoal adsorbers to reduce releases to the atmosphere.
C1.2.3 Secondary Systems Normally, condensate flow and steam generator blowdown will go parallel through 4 of the 5 condensate polishing demineralizers to remove activity and harmful ions from the water. Noncondensable gases will be taken from the i
!- C-2 L
e r.
secondary system by the condenser steam air ejector and are passed through a i
radiation monitor to the unit vent.
Figure C1.0-2 is a schematic representation of the gaseous radwaste system at ,
Catawba.
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{ CONDENSER j g E M F -35 STEAM C 60 cfm GENERATOR %s g -
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% BLOWOOWN RECYCLE TANK' O
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w_ad CONTROL TANK 4 _.___ .--__--- J TO TO UNIT 2 REACTOR 4 4 g R E ACTO R y y y 4 4 4 BORON RECYCLE GAS GAS TANKS SHIM 8LEED STORAGE 600 FT 3 EVAPORATOR TANKS 1000:4 G AS STRIPPE R g g g g g g '
i WASTE G AS SYSTEM l EMF-50 l NORMALLY CLOSED UPPER VOLUME h 18,000 cfm il
-tm -
ADDITION SYSTEM (VQ) p LOWER VOLUME A C jg
- W ---
PURGE OUTLET (VP)
PURGEINLET S -
C A P 6 EM F-38 4 ~~-
10,000 E M F-39 j cfm E M F-40 CONTAINMENT VENTILATION SYSTEMS (2) i INTERN ATL REClRCU. TR AINS (2) 8,000 cfm E ACH l EMF-42 l 35,700 cfm hJ p O FUEL HANDLING ] PER UNIT "I l AREA 4 l EMF-41 l
^ #
OUTSID E p PER U l AIR 4
- OTHER AREAS AUXILI ARY PER UNIT BUILDING SUPPLY FANS LEGENO:
AUXf LI ARY BUILDING SYSTEM (SHAREL)
P PREFILTER A HIGH-EFFICIENCY PARTICULATE FILTER C CHARCOAL ABSOR8ER FUEL HANDLING AREA IS NORMALLY UNFILTERED. UPON A RADIATION ALARM BY EMF-42.THE E5(HAUST WILL BE DIVERTED TO THE FILTERED MODE.
POTENTIALLY CONTAIMINATED AREAS OF THE AUXILIARY BUILDING ARE NORMALLY UNFILTERED. UPON A RADIATION ALARM BY EMF-41, THE EXHAUST WILL BE DIVERTED TO THE FILTERED MODE.
FIGURE C1.0-2 CATAWB A NUCLEAR STATION G ASEOUS R ADWASTE SYSTEM
C2.0 RELEASE RATE CALCULATION Generic release rate calculations are presented in Section 1.0; these calculations will be used to calculate release rates for Catawba Nuclear Station.
C2.1 LIQUID RELEASE RATE CALCULATIONS There are two potential release points at Catawba. They are as follows:
- 1. -Liquid Waste Effluent Discharge Line
- 2. Conventional Waste Water Treatment System Effluent Line C2.1.1 Liquid Waste Effluent Discharge Line There are three low-pressure service water pumps with a minimum flow rate of 16,500.gpm each and four nuclear service water pumps with a minimum flow rate of 9,000 gpm each which provide the required dilution water needed for a release. The LPSW system flow rate monitor has a variable setpoint which term-
~ inates the release by closing the isolation valve, 1 WL124 should-the dilution flow fall below the setpoint. The following equation shall be used to calculate a discharge flow, in gpm.
r
< ; n ,
f<F RL
' I i
- i=1 MPC f
- a where:
4 f = the undiluted effluent flow, in gpm.
F = actual low pressure service water flowrate, in gpm, from the sum RL
- of the flowrate monitors located in the Control Room.
a = the recirculation factor at equilibrim (dimensionless),1.027.
- 8
- a=1+O--R'
= 1 + 4400 efs
~Q H
where:
Q = average dilution flow (120 cfs)
R QH = average fl W.Past Wylie Dam (4400 cfs)
C. = the concentration of radionuclide, i, in undiluted effluent as
- determined by laboratory analyses, in pCi/ml.
MPC. = the concentration of radionuclide, i, from 10CFR20, Appendix B,
-Table II, Column 2. If radionuclide, i', is a dissolved noble gas, the MPC. = 2.0E-04'pci/ml.
1 C-4
C2.1.2 Conventional Waste Water Treatment System Effluent Line The conventional waste water treatment system effluent is normally considered nonradioactive; that is, it is unlikely the effluent will contain measurable
-activity above background. It is assumed that no activity is present in the effluent until indicated by radiation monitoring measurements and by periodic analyses of the composite sample collected on that line. The water sources listed below that are normally discharged via the conventional waste water treatment system and/or the Turbine Building Sump will be diverted if they become radioactive.
- a. Containment Ventilation Unit Condensate Effluent Line Normally the containment ventilation unit condensate effluent line would discharge into the Turbine Building sump, but if radiation is detected above background, the discharge will be terminated and an alarm actuated.
The containment ventilation unit condensate tank will then be recirculated,. sampled and then discharged through the liquid waste effluent line and monitored or processed thru the WL system.
- b. Auxiliary Feedwater Sump Pumps and Floor Drain Sump Pump Line Normally the discharge line coming from these sumps will discharge into the Turbine Building sump, but if radiation is detected above background, the discharge flow will automatically be routed to the floor drain tank for processing and later be discharged through the liquid waste effluent line.
- c. Turbine Building Sump Discharge Line Normally the discharge from the Turbine Building sump will go into the conventional waste water treatment system, but if radiation is detected above background, the sump pumps A, B, and C will stop and an alarm actuated. The Turbine Building sump discharge line can either be routed to the floor drain tank for processing or routed directly to the liquid waste effluent discharge line.
C2.2 GASEOUS RELEASE RATE CALCULATIONS The unit vent is the release. point for waste gas decay tanks, containment air releases, the condenser air ejector, and auxiliary building ventilation. The condenser air ejector effluent is normally considered nonradioactive; that is, it is unlikely the effluent will contain measurable activ.ty above background.
It is assumed that no activity is present in the effluent until indicated by radiation monitoring measurements and/or by analyses of periodic samples collected on that line. -Radiation monitoring alarm / trip setpoints in con-
-junction with administrative controls assure that release limits are not exceeded; see section C.3.0 on radiation monitoring setpoints.
The following calculations, when solved for flowrate, are the release rates for nobic gases and for radioiodines, particulates and other radionuclides with half-lives greater than 8 days; the most conservative of release rates
- calculated in C2.2.1 and C2.2.2 shall control the release rate for a single release point.
C-5
C2.2.1 Noble Gases IK g [(X/Q)Qg ] < 500 mrem /yr, and 1
I (L g;+ 1.1 Mg) [(X/Q)Qg ] < 3000 mrem /yr 1
where the terms are defined below.
C2.2.2 Radiciodines, Particulates, and Other Radionuclides With T 1/2 > 8 Days s
I P; [W Qg ] < 1500 mrem /yr 1
where:
K. = The total body dose factor due to gamma emissions for each identified 1
noble gas radionuclide, in mrem /yr per pCi/m3 from Table 1.2-1.
L.
1
= The skin dose factor due to beta emissions for each identified noble 3
gas _radionuclide, in mrem /yr per pCi/m from Table 1.2-1.
M = The air dose factor due to gamma emissions for each identified noble I 3 gas radionuclide, in mrad /yr per pCi/m from Table 1.2-1 (unit conver-sion constant of 1.1 mrem / mrad converts air dose to skin dose).
P. = The dose parameter for radionuclides other than noble gases for the
- inhalation pathway, in mrem /yr per pCi/m3 and for the food and ground plane pathways in m2 .(mrem /yr) per pCi/sec from Table 1.2-2. The dose factors are based on the critical individual organ and most restrictive age group (child or infant).
= The release rate of radionuclides, i, in gaseous effluent from all D*. release points at the site, in pCi/sec.
(X/Q) = 3.10E-05 sec/m 3 . The highest calculated annual average relative concen-tration for any area at or beyond the unrestricted area boundary. The location is the NNE sector @ 0.5 miles.
W = The highest calculated annual average dispersion parameter for estimating the dose to an individual at the controlling location:
W = 3.1E-05 sec/m 3 , for the inhalation pathway. The location is the unrestricted area in the NNE sector @ 0.5 miles.
W = 1.1E-07 meter 2, for the food and ground plane pathways. The location is the unrestricted area boundary in the NE/NNE sector
@ 0.5 miles (nearest residence, and vegetable garden).
\
l C-6 l
. l 4
i.
., 6 1
I
-1=kCf+k2 1 i 1 = 4.72E+2C.f 1 i:
where:
C. ~=-the concentration of radionuclide, i, in undiluted gaseous effluent, in pCi/ml.
f' = the undiluted effluent flow, in efm 3
kg = conversion factor,.2.83E4 ml/ft
,k2 = conversion factor, 6El sec/ min a
C-7
4
~
C3.0 FADIATION MONITOR SETPOINTS Using the generic calculations presented in Section 2.0, radiation monitoring ,
setpoints are calculated for monitoring as required by the Technical Specifications. l All radiation monitors for Catawba are off-line except EMF-50 (Waste Gas 1 System) which is in-line. These monitors alarm on low flow; the minimum flow ,
alarm level for both the liquid monitors and the gas monitors is based on the manufacturer's recommendations. These monitors measure the activity in the
~
liquid or gas volume exposed to the detector and are independent of flow rate if a minimum flow rate is assured.
Radiation monitoring setpoints calculated in the.following sections are expressed in activity concentrations; in. reality the monitor readout is in counts per minute. The relationship between concentration and counts per minute is estab-lished by a station procedure using the following relationship:
- 2.22 x 10 6e V
. where:
c = the gross activity, in pCi/ml
- r = the count rate, in cpm 2.22 x 108 = the disintegration per minute per pCi e = the counting efficiency, cpm /dpm V =~the volume of fluid exposed to the detector, in ml.
For those occurrences when simultaneous releases of radioactive material must be made, monitor.setpoints will be adjusted downward in accordance with Station Procedures:to insure that instantaneous concentrations will not be exceeded.
.C3.1 LIQUID RADIATION MONITORS C3.1.1 Waste Liquid Effluent Line .
As described in Section.C2.1.11on release rate calculations for the waste liquid effluent, the release is controlled by limiting the flow rate of effluent from the station. Although the release rate is flow rate controlled, the radiation monitor setpoint shall be set to terminate the release if the effluent activity should exceed.that determined by laboratory analyses and used to calculate the release rate. A typical radiation eonitor setpoint may be calculated as follows:
c $ " of $ 2.48E-05 pCi/ml where:
c = the gross activity in undiluted effluent, in pCi/ml f = the flow from the tank may vary from 0-100 gpm but, for this calculation, is assumed to be 100 gpm.
C-8 4
. - , - ,. _ . , , - - .--,.,-, . . . , . . .. , . . , , - ..--~y
L i'
I'
- MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture
- o = 1.027'(See Section C2.1.1)
F = the dilution flow may vary as described in section C2.1.1, but is
~
conservatively estimated at 25,500 gpm, the minimum flow available.
C3.1.2' Containment Ventilation Unit Condensate Effluent Line - EMF-44 As described in Section C2.1.2 on release rate calculations for the conta nment i
- ventilation unit condensate effluent,.it is possible but unlikely that the effluent'will.contain measurable activity above background. It is assumed that no activity is present in the effluent until indicated by radiation monitoring.
Since the tank contents are discharged automatically, the radiation monitor
- setpoint will be set at 1.0E-06 pCi/ml (the monitor's lowest level of detection) plus background to assure that release limits are not exceeded.
C3.1.3 Auxiliary Feedwater Sump Pumps and Floor Drain Sump Pump - EMF 52 As described in Section C2.1.2 on release rate calculations for the auxiliary feedwater sump pumps and floor drain sump pump effluents, it is possible but
.unlikely.that the effluents will.contain measurable activity above background.
ItLis assumed that no activity is present in the effluent until indicated by radiation monitoring. Since the sumps are discharged automatically, the
' ~
radiation monitor setpoint will be set at 1.0E-06 pCi/m1'(the monitor's lowest level-of detection) plus background to assure that release limits are not exceeded.
C3.1.4 . Turbine Building Sump Discharge Line - EHF 31 As described in Section C2.1.2 on release rate calculations for the turbine
. building sumps, it is possible but unlikely that the effluents will contain
' measurable activity above. background. It is assumed that no activity is I -present in the effluent until indicated by radiation monitoring. Since the sump contents are discharged automatically, the radiation monitor setpoint will be set at 1.0E-06 pCi/ml (the monitor'n lowest level of detection) plus background to assure that release limits are not exceeded.
I i
+
1 C-9
+ . - - - - .. _ . , , . - _ . .. , , - . . - - - - - - , - - - - _ - - , - - . - . -
C3.2 GAS MONITORS The following equation shall be used to calculate noble gas radiation monitor
'setpoints based on Xe-133 (iiistorical data shows that Xc-133 is most predominant isotope):
K(X/Q)Qg _ < 500 (see Section C2.2.1)
$g=4.72E+02Cf(seeSectionC2.2.2) g Cg < 1.16/f where:
Cg = the gross activity in undiluted effluent, in pCi/ml f = the flow from the tank or building sources, in efm K ~= from Table 1.2-1 for Xe-133, 2.94E+2 mrem /yr per pCi/m 3
- X/Q = 3.1E-05, as defined in Section C.2.2.2 As stated in Section C2.2, the unit vent is the release point for the contain-ment purge ventilation system, the containment air release and addition system, the condenser air ejector, and auxiliary building ventilation.
For releases from the containment purge ventilation system, a typical radiation-monitor setpoint may be calculated as follows:
Cg < 1.16/f = 6.5E-06 where:
f = 151,000 cfm (auxiliary building ventilation) + 28,000 cfm (containment purge) = 179,000 cfm For, release from the containment air release and addition system, the waste gas decay. tanks, the condenser air ejectors, and the auxiliary building ventilation, a typical radiation monitor setpoint may be calculated as follows:
Cg < 1.16/f = 7.7E-06 where:
f = 151,000 cfm (auxiliary building ventilatior.)
C-10
C4.0 DOSE CALCULATIONS C4.1 FREQUENCY OF CALCULATIONS Lose contributions to the maximum exposed individual shall be calculated every 31 days, quarterly, semiannually, and annually (as reqaired by Technical Spec-ifications) using the methodology in the generic information sections. This methodology shall also be used for any special reports. Dose projections shall be performed using simplified estimates. Fuel cycle dose calculations shall be performed annually or as required by special reports. Dose contributions may be calculated using the methodology in the appropriate generic inforination sections.
C4.2 DOSE MODELS FOR MAXIMUM EXPOSED INDIVIDUAL C4.2.1 Liquid Effluents For dose contributions from liquid radioactive releases, one of the two following cases will apply:
- 1. If the radionuclides Co-58 and/or Co-60 have been detected and Cs-134 and/or Cs-137 have not been detected (i.e., plants without any fuel failure) dose calculations will be based upon an adult who consumed fish caught in the discharge canal and who drank water from the nearest " downstream" potable water intake. The dose from these two radionuclides has been calculated to be 13% of that individual's total body dose.
- 2. If the radionuclides Cs-134 and/or Cs-137 have been detected, doce calculations indicate that the maximum exposed individual would be an adult who consumed fish caught in the discharge canal and who drank water from the nearest " downstream" potabic water intake. The dose from these two radionuclides has been calculated to be 90% of that individual's total body dose.
C4.2.2 Gaseous Effluents C4.2.2.1 Noble Gases For dose contributions from exposure to beta and gamma radiation from noble gares, it is assumed that the maximum c:: posed individual is an adult on the site boundary in each meteorological sectors.
C4.2.2.2 Radioiodines, Particulates, and Other Radionuclides T 1/2 > 8 days For dose contributions from radiciodines, particulates and other radionuclides; it is assumed that the maximum exposed individual is an infant who breathes the air and consumes milk from the nearest goat or cow in each meteorological sector.
C-11
r C4.3 sit!PLIFIED DOSE ESTI?! ATE C4.3.1 Liquid Effluents For dose estimates, two simplified calculations using the assumptions presented in Section C4.2.1 and source terms presented in the FSAR are presented. Once operational source term data is available, this information shall be used to revise these calculations, if necessary.
Case 1 - No Cs-134 or Cs-137 present in effluent.
DWB = 1.57E+03 (Ff)(T g) (CCo-60 + 0.35 CCo-58) f=1 where:
1.57E+03 = 1.14E+05 (U,,/D y + U,g BFg ) DF,17 (7.69) where:
1.14E+05 = 10 6 pci/pci x 103 ml/kg + 8760 hr/yr U, = 730 kg/yr, adult water consumption D, = 37.7, dilution factor from the near field area to the nearest potable water intake, where:
D a (See Section C2.1.1) g=NN R !!
U,g = 21 kg/yr, adult fish consumption BFg = 5.0E+01, bioaccumulation factor for Cobalt (Table 3.1-1)
DF,gT = 1.67E-06, adult, total t>ody, ingestion dose factor for Co-60 (Tabic 3.1-2) 7.69 = factor derived from assumption that 13% of dose is from Co-58 and Co-60 or 100% + 13% = 7.09 And where:
I" FA= F+f where:
f = liquid radwaste flow, in gpm o = recirculation factor at equilibrium,1.027 (see Section C2.1.1)
F = dilution flow, in gpm C-12 i
i i
l F
And where:
Tg = The length of time, in hours, over which CCo-58' Co-60, and Fg are averaged.
C = the average concentration of Co-58 in undiluted effluent, in Co-58 pCi/ml, during the time period considered.
C the average concentration of Co-60 in undiluted effluent, in pCi/ml, Co-60 = during the time period considered.
0.35 = The ratio of the adult total body ingestion dose factors for Co-58 and Co-60 or 1.67E-06 + 4.72E Table 3.1-2.
Case 2 - Cs-134 and/or Cs-137 present in effluent.
m Dg = 6.38E+05 I (Fg)(T g) (CCs-134 & 0.59 CCs-137)
E=1 where:
6.38E+05 = 1.14E+05 (U, /D,+ U,f BFg ) DF,g7 (1,10) where:
1.14E+05 = 10 8 pci/pci x 103 ml/kg + 8760 hr/yr U,,= 730 kg/yr, adult water consumption D" = 37.7, dilution factor from the near field area to the nearest potable ;
water intake, where:
D = o (see Section C2.1.1)
QR H U,g = 21 kg/yr, adult fish consumption BFg = 2.00E+03, bioaccumulation factor for Cesium (Table 3.1-1)
DF,g = 1.21E-04, adult, total body, ingestion dose factor for Cs-134 (Table 3.1-2) 1.10 = factor derived from the assumption that 90% of dose is from Cs-134 and Cs-137 or 100% +.90% = 1.10 And where:
to Fg=F+f C-13
4.
i
-where:
f =' liquid radwaste: flow, in gpm a = recirculation factor at equilibrium, 1.027 (see Section C2.1.1)
L i F = dilution flow, in gpm And where:
Fg
'CCs-134 = the average concentration of Cs-134 in undiluted effluent, in
- pCi/ml, durin'g the time period considered.
i-C t e average c ncen ra i n s-13 in undiluted e nluent, in pCi/ml, Cs-137 = during the time period considered.
0.59 = The' ratio of.the adult total body ingestion dose factors for Cs-134 and Cs-137 or 7.14E-05 +.1.21E-04 = 0.59 C4.3.2 Gaseous Effluents Meteorological data is provided in Tables C4.0-1 and C4.0-2.
C4.3.2.1 Noble Gases For dose estimates, simplified dose estimates using the assumptions in C4.2.2.1 and source terms in the FSAR are presented below. Once operational source term data is available, this information shall be used to revise these calculations, D if necessary. These calculations further assume that the annual average dispersion parameter is used and that Xenon-133 contributes 45% of the dose.
}
=3'.47E-10[$]Xe-133(*
y D p=1.03E-09[D}Xe-133(* }
where:
3 47E -10 = (3.17E-8)(353) (X/Q), derived from-equation presented in Section 3.1.2.1.
,1.03E-09. = (3.17E-08) (1050) (X/Q), derived from equation presented in Section 3.1.2.1.
X/Q = 3.1E-05 sec/m3 , as= defined in Section C2.2.2
'[D)Xe-133 = the total' Xenon-133 activity released in pCi 2.22 = factor derived from the conservative assumption (based on historical data)-that'45% of-the dose is contributed by Xe-133.
C-14 ,
C4.3.2.2 Radioiodines, Particulates, and Other Radionuclides with T 1/2 > 8 days For dose estimates, simplified dose estimates using the assumptions in C4.2.2.2 and source terms in the FSAT ,re presented below. Once operational source term data is available, this information shall be used to revise these calculations, if necessary. These calculations further assume that the annual average
' dispersion / deposition parameter is used and that 95% of the dose is from Iodine-131 concentrated in goat's milk. The simplified dose estimate to the thyroid of an infant is:
D=1.84E+04w(D)I-131(1.05) where:
w = 7.3E-10 = D/Q for food and ground plane pathway, in m2 from Table C4.0-2 for location of nearest real goat (NW sector at 2.5 miles).
(D)I-131 = the total Iodine-131 activity released in pCi.
1.84E+04 =(3.17E-08)(Rf[D/Q])withtheappropriatesubstitutgonsfor goat's milk in the grass-cow-milk pathway factor, R.
1
[D/Q] for Iodine-131. 'See Section 3.1.2.2.
1.05 = factor derived from the conservative assumption (based on historical-data) that 95% of the dose is contributed by I-131.
C4.3 FUEL CYCLE CALCULATIONS This section will be submitted when complete.
C-15
TABLE C4.0-1 (1 of 1)
' CATAWBA NUCLEAR STATION DISPERSION PARAMETER (X/Q) FOR LONG TERM RELEASES > 500 HR/YR OR > 125 HR/QTR.
(sec/m3 )
Distance to the control location, (miles)
Sector 0.5 1.0 1.5 2.0 12 . 5 3.0 3.5 4.0 4.5' 5.0 N 2.6E-5 6.5E-6 2.7E-6 1.5E-6 9.7E-7 6.9E 5.2E-7 4.1E-7 3.3E-7 2.8E-7 NNE 3.1E-5 8.1E-6 3.3E-6 1.8E-6 1.2E-6 8.2E-7 6.2E-7 4.9E-7 4.0E-7 3.3E-7 NE 3.0E-5 7.8E-6 3.2E-6 1.8E-6 1.1E-6 8.0E-7 6.0E-7 4.7E-7 3.9E-7 3.2E-7 ENE 1.5E-5 3.9E-6 1.6E-6 8.9E-7 5.7E-7 4.1E-7 3.1E-7 2.4E-7 2.0E-7 1.6E-7 E 1.4E-5 3.7E-6 1.5E-6 8.4E-7 5.4E-7 3.8E-7 2.9E-7 2.3E-7 1.9E-7 1.6E-7 ESE 9.0E-6 2.3E-6 9.5E-7 5.3E-7 3.4E-7 2.4E-7 1.8E-7 1.4E-7 1.2E-7 9.7E-8 SE 9.2E-6 2.4E-6 9.8E-7 5.4E-7 3.5E-7 2.4E-7 1.8E-7 1.4E-7 1.2E-7 9.8E-8 SSE 1.1E-5 2.9E-6 1.2E-6 6.4E-7 4.1E-7 2.9E-7 2.2E-7 1.7E-7 1.4E-7 1.1E-7 S 2.5E-5 6.4E-6 2.6F-6 1.5E-6 9.3E-7 6.6E-7 5.0E-7 3.9E-7 3.2E-7 2.7E-7 SSW 1.7E-5 4.4E-6 1.8E-6 1.0E-6 6.4E-7 4.5E-7 3.4E-7 2.7E-7 2.2E-7 1.8E-7 SW 1.3E-5 3.4E-6 1.4E-6 7.4E-7 4.7E-7 3.3E-7 2.4E-7 1.9E-7 1.5E-7 1.3E-7 WSW 7.0E-6 1.8E-6 7.2E-7 3.9E-7 2.5E-7 1.7E-7 1.3E-7 . SE-7 8.2E-8 6.8E-8 W 8.9E-6 2.3E-6 9.3E-7 5.0E-7 3.2E-7 2.2E-7 1.7E-7 1.3E-7 1.1E-7 8.7E-8 WNW 6.6E-6 1.7E-6 6.8E-7 3.7E-7 2.4E-7 1.7E-7 1.3E-7 9.8E-8 8.0E-8 6.6E-8 NW 1.0E-5 2.6E-6 1.1E-6 5.9E 3.8E-7 2.7E-7 2.0E-7 1.6E-7 1.3E-7 1.1E-7 NNW 1.3E-5 3.3E-6 1.4E-6 7.5E-7 4.8E-7 3.4E-7 2.6E-7 2.0E-7 1.6E-7 1.4E-7 L_
TABLE C4.0-2 -
(1 of 1)
CATdWBA NUCLEAR STATION DIPERSION PARAMETER (D/Q) FOR'LONG TERM RELEASES > 500 HR/YR OR') 125 HR/QTR (meter 2)
Distance to the control location, (miles) 1.5 2.5 3.0 3.5 4.0 5.0 Sectuc 0.5 1.0 2.0 ___4 . 5 N 6.4E-8 1.6E-8 5.6E-9 2.8E-9 1.6E-9 1.1E-9 7.5E-10 5.6E-10 4.3E-10 3.4E-10 NNE- 1.1E-7 2.7E 9.6E-9 4.7E-9 2.8E-9 1.8E-9 1.3E-9 -9.5E 7.4E-10 5.8E-10 NE 1.1E-7 2.6E-8 9.3E-9 4.6E-9 2.7E-9 1.8E-9 1.3E-9 9.3E-10 7.2E-10 3.7E-10 ENE 4.1E-8 1.0E-8 3.6E-9 1.8E-9 1.1E-9. 6.9E-10 4.9E-10 3.6E-10 2.8E-10 2.2E-10 E '3.6E-8 8.8E-9 3.2E-9 1.6E-9 9.3E-10 6.1E-10 4.3E-10 3.2E-10 2.4E-10 1.9E-10 ESE 2.5E-8 6.0E-9 2.2E-9 1.1E-9 6.3E-10 4.2E-10 2.9E-10 2.2E-10 1.7E-10 1.3E-10 SE 3.0E-8 7.3E-9 2.6E-9 1.3E-9 7.7E-10 5.0E-10 3.5E-10 2.6E-10 2.0E-10 1.6E-10 SSE 3.8E-8 9.3E-9 3.3E-9 1.7E-9 9.7E-10 6.4E-10 4.5E-10 3.3E-10 2.6E-10 2.0E-10 7.2E-8 1.8E-8 6.3E-9 3.1E-9 1.8E-9 1.2E-9 8.5E-10 6.3E-10 4.8E-10 3.8E-10 S
SSW 6.6E-8 1.6E-8 5.8E-9 2.9E-9 1.7E-9 1.1E-9 7.8E-10 5.8E-10 4.4E-10 3.5E-10 SW 5.7E-8 1.4E-8 5.0E-9 2.5E-9 1.5E-9 9.6E-10 6.7E-10 5.0E-10 3.9E-10 3.1E-10 WSW 2.4E-8 5.7E-9 2.1E-9 1.0E-9 6.0E-10 4.0E-10 2.8E-10 2.1E-10 1.6E-10 1.3E-10 W 2.8E-8 6.7E-9 2.4E-9 1.2E-9 7.0E-10 4.6E-10 3.2E-10 2.4E-10 1.9E-10 1.5E-10 WNW 1.9E-8 4.6E-9 1.7E-9 8.2E-10 4.8E-10 3.2E-10 2.2E-10 1.6E-10 1.3E-10 1.0E-10 2.9E-8 7.0E-9 2.5E-9 1.3E-9 7.3E-10 4.8E 3.4E-10 2.5E-10 1.9E-10 1.5E-10 NW 4.1E-8 9.9E-9 3.6E-9 1.8E-9 1.0E-9 6.8E-10 4.8E-10 3.6E-10 2.7E-10 2.2E-10 NNW
TABLE C4.0-3 *
(1 of 3)
CATAWBA NUCLEAR STATION ADULT A,gT DOSE M ETERS mrem /hr per pCi/ml NUCLIDE BONE LIVER T.B0DY THYROID KIDNEY LUNG GI-LII H 3 0.0 4.58E-01 4.58E-01 4.58E-01 4.58E-01 4.58E-01 4.58E-01 NA 24 4.11E+02 4.11E+02 4.11E+02 4.11E+02 4.11E+02 4.11E+02 4.11E+02 CR 51 0.0 0.0 1.28E+00 7.65E-01 2.82E-01 1.70E+00 3.22E+02 MN 54 0.0 4.39E+03 8.37E+02 0.0 1.31E+03 0.0 1.34E+04 MN 56 0.0 1.10E+02 1.96E+01 0.0 1.40E+02 0.0 3.52E+03 FE 55 6.64Et02 4.59E+02 1.07E+02 0.0 1.0 2.56E+02 2.63E502 FE 59 1.05E+03 2.46E+03 9.45E+02 0.0 0.0 6.89E+02 8.21E403 C0 58 0.0 9.08E+01 2 04E+02 0.0 0.0 0.0 1.84E+03 C0 60 0.0 2.61E+02 5.75E+02 0.0 0.0 0.0 4.90E+03 NI 63 3.14E+04 2.18E+03 1.05E+03 0.0 0.0 0.0 4.54E+t2 NI 65 1.28E+02 1.66E+01 7.56E+00 0.0 0.0 0.0 4.20E+02 CU 64 0.0 1.02E+01 4.77E+00 0.0 2.56E+01 0.0 8.66E+02 ZN 65 2.32E+04 7.38E+04 3.33E+04 0.0 4.93E+04 0.0 4.65E+04 ZN 69 4.93E+01 9.44Et01 6.56E+00 0.0 6.13E+01 0.0 1.42E+01 BR 83 0.0 0.0 4.05E+01 0.0 0.0 0.0 5.83E+01 BR 84 0.0 0.0 5.25E+01 0.0 0.0 0.0 4.12E-04 BR 85 0.0 0.0 2.16E+00 0.0 0.0 0.0 0.0 RB 86 0.0 1.01E+05 4.71E+04 0.0 0.0 0.0 1.99E+04 RB 88 0.0 2.90E+02 1.54E+02 0.0 0.0 0.0 4.00E-09 dB 89 0.0 1.92E+02 1.35E+02 0.0 0.0 0.0 1.12E-11 SR 89 2.28E+04 0.0 6.54E+02 0.0 0.0 0.0 3.66E+03 SR 90 2.84E+05 0.0 7.62E+04 0.0 0.0 0.0 1.62E+04 SR 91 4.20E+02 0.0 1.70E+01 0.0 0.0 0.0 2.00E+03 SR 92 1.59E+02 0.0 6.88E+00 0.0 0.0 0.0 3.15E+03 Y 90 5.97E-01 0.0 1.60E-02 0.0 0.0 0.0 6.33E+03 Y 91M 5.64E-03 0.0 2.18E-04 0.0 0.0 0.0 1.66E-02 Y 91 8.75E+00 0.0 2.34E-01 0.0 ,0.0 0.0 4.82E+03 Y 92 5.24E-02 0.0 1.53E-03 0.0 0.0 0.0 9.18E+02
- Methodology for table provided by: M. E. Wrangler, RAB:NRR:NRC on 3/17/83 TABLE C4.0-3 (1 of 3) t-
i 6
TABLE C4.0-3 i
(2 of 3) l CATAWBA NUCLEAR STATION '
ADULT A DOSE PARAMETERS ait mrem /hr per pCi/mi NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LII Y 93 1.66E-01 0.0 4.59E-03 0.0 0.0 0.0 5.27E+03 ZR 95 3.07E-01 9.85E-02 6.67E-02 0.0 1.55E-01 0.0 3.12E+02 2R 97 1.70E-02 3.43E-03 1.57E-03 0.0 5.18E-03 0.0 1.06E+03 NB 95 4.47E+02 2.49E+02 1.34E+02 0.0 2.46E+02 0.0 1.51E+06 MO 99 0.0 1.13E+02 2.14E+01 0.0 2.55E+02 0.0 2.61E+02 TC 99M 9.41E-03 2.66E-02 3.39E-01 0.0 4.04E-01 1.30E-02 1.57E+01 TC 101 9.68E-03 1.40E-02 1.37E-01 0.0 2.51E-01 7.13E-03 4.19E-14 RU 103 4.84E+00 0.0 2.08E+00 0.0 1.85E+01 0.0 5.65E+02 RU 105 4.03E-01 0.0 1.59E-01 0.0 5.20E+00 0.0 2.46E+02 RU 106 7.19E+01 0.0 9.10E+00 0.0 1.39E+02 0.0 4.65E+03 A0 110M 1.23E+00 1.14E+00 6.78E-01 0.0 2.24E+00 0.0 4.66E+02 TE 125M 2.57E+03 9.32E+02 3.45E+02 7.74E+02 1.05E+04 0.0 1.03E+04 TE 127M 6.50E+03 2.32E+03 7.92E+02 1.66E+03 2.64E+04 0.0 2.18E+04 TE 127 1.06E+02 3.79E+01 2.28E+01 7.82E+01 4.30E+02 0.0 8.33E+03 I
TE 129M 1.10E+04 4.12E+03 1.75E+03 3.79E+03 4.61E+04 0.0 5.56E+04 TE 129 3.01E+01 1.13E+01 7.3*E+00 2.31E+01 1.27E+02 0.0 2.27E+01 TE 131M 1.66E+03 8.12E+02 6.77E+02 1.29E+03 8.23E+03 0.0 8.06E+04 TE 131 1.89E+01 7.90E+00 5.97E+00 1.55E+01 8.28E+01 0.0 2.68E+00 TE 132 2.42E+03 1.56E+03 1.47E+03 1.73E+03 1.51E+04 0.0 7.40E+04 I 130 2.88E+01 8.50E+01 3.35E+01 7.20E+03 1.33E+02 0.0 7.32E+01 I 131 1.59E+02 2.27E+02 1.30E+02 7.43E+04 3.89E+02 0.0 5.98E+01
-1 132 7.74E+00 2.07E+01 7.24E+00 7.24E+02 3.30E+01 0.0 3.89E+00 I 133 5.41E+01 9.41E+01 2.87E+01 1.38E+04 1.64E+02 0.0 8.46E+01 I 134 4.04E+00 1.10E+01 3.93E+00 1.90E+02 1.75E+01 0.0 9.57E-03 I 135 1.69E+01 4.42E+01 1.63E+01 2.92E+03 7.09E+01 0.0 4.99E+01 CS 134 2.98E+05 7.09E+05 '5.80E+05 0.0 2.29E+05 7.62E+04 1.24E+04 CS 136 3.12E+04 1.23E+05 8.86E+04 0.0 6.85E+04 9.39E+03 1.40E+04 CS 137 3.82E+05 5.22E+05 3.42E+05 0.0 1.77E+05 5.89E+04 1.01E+04 CS 138 2.64E+02 5.22E+02 2.59E+02 0.0 3.84E+02 3.79E+01 2.23E-03 BA 139 1.14E+00 8.14E-04 3.35E-01 0.0 7.61E-04 4.62E-04 2.03E+00 TABLE C4.0-3 (2 of 3)
TABLE C4.0-3 (3 of 3)
CATAWBA NUCLEAR STATION ADULT A DOSE PARAMETERS ait mrem /hr per pCi/mi NUCLIDE BONE LIVER T. BODY TlIYROID KIDNEY LUNG GI-LII BA 140 2.39E+02 3.00E-01 1.57E+01 0.0 1.0?E-01 1.72E-01 4 93E+02 BA 141 5.55E-01 4.19E-04 1.87E-02 0.0 3.90E-04 2.38E-04 2.62E-10 BA 142 2.51E-01 2.58E-04 1.58E-02 0.0 2.18E-04 1.46E-04 3.54E-19 LA 140 1.55E-01 7.82E-02 2.07E-02 0.0 0.0 0.0 5.74E+03 LA 142 7.94E-03 3.61E-03 9.00E-04 0.0 0.0 0.0 2.64E+01 CE 141 4.31E-02 2.91E-02 3.30E-03 0.0 1.35E-02 0.0 1.11E+02 CE 143 7.59E-03 5.61E+00 6.21E-04 0.0 2.47E-03 0.0 2.10E+02 CE 144 2.25E+00 9.39E-01 1.21E-01 0.0 5.57E-01 0.0 7.59E+02 PR 143 5.71E-01 2.29E-01 2.83E-02 0.0 1.32E-01 0.0 2.50E+03 PR 144 1.87E-03 7.76E-04 9.49E-05 0.0 4.38E-04 0.0 2.69E-10 ND 147 3.90E-01 4.51E-01 2.70E-02 0.0 2.64E-01 0.0 2.17E+03 W 187 2.96E+02 2.48E+02 8.65E+01 0.0 0.0 0.0 8.11E+04 NP 239 3.11E-02 3.06E-03 1.69E-03 0.0 9.54E-03 0.0 6.28E+02 TABLE C4.0-3 (3 of 3)
\
C5.0 Radiological Environmental Monitoring The Radiological Environmental Monitoring ."rogram shall be conducted in accordance with Technical Specification, Section 3/4.12.
The monitoring program locations and analyses are given in Tables C5.0-1 through C5.0-3 and Figure C5.0-1.
Site specific characteristics make groundwater sampling, special low-Icvel I-131 analyses on drinking water, and food product sampling unnecessary. Ground-water recharge is from precipitation and the groundwater gradient is toward the effluent discharge area; therefore, contamination of groundwater from liquid effluents is highly improbable. Special low-level I-131 analyses in drinking water will not be performed routinely since the expected I-131 dose from this pathway is less than 1 mrem / year. Food products will not be sampled since lakewater irrigation is not practiced in the vicinity.
The laboratory performing the radiological environmental analyses shall parti-cipate in an interlaboratory comparison program which has been approved by the NRC. This program is the Environmental Protection Agency's (EPA's)
Environmental Radioactivity Laboratory Intercomparsion Studies (crosscheck)
Program, our participation code is CP.
The dates of the land-use census that was used to identify the controlling receptor locations was 10/26/82 - 10/28/82. These dates will nat be changed ualess a subsequent census changes a controlling receptor's location.
C-16
- _ - _ - - - - . _ _ - . _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _______________________________1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
L
+
- ; TABLE C5.0-1 ,
(l'of 1) .
. n
.. . CATAWBA' RAD 10 LOGICAL MONITORING. PROGRAM SAMPLING LOCATIONS l .
, .! A j : (TLD LOCATIONS).
V ,7 g.;
p j ~ SAMPLING LOCATION DESCRIPTION SAMPLING LOCATION DESCRIPTION' ,
j ~
! 'i.
200 , SITE BOUNDARY (0.7M NNE) 232 4-5 MILE RADIUS (4.1M NE) i 201 SITE BOUNDARY -(0.5M NE) ,' 233, _
4-5-MILE RADIUS (4.0M ENE) 2 202 SITE BOUNDARY (0.6M ENE) 234
~
4.5 MILE RADIUS, . (4'.5MI E) i 203 . SITE BOUNDARY (0.5M SE)s - 235- 4.5 MILE RADIUS (4.0M'ESE).
204 SITE BOUNDARY .(0.5M SSW) 4 236. 4-5 MILE RADIUS (4.2M SE)_
, 205 SITE BOUNDARY (0.6M.SW); 237 4-5 MILE RADIUS (4.8M SSE)
' 206 SITE BOUNDARY (0.7M WNW) 238 4-5 MILE RADIUS (4.2M S)-
207 SITE BOUNDARY- (0.8M NNW) 239 4-5 MILE RADIUS (4.6M SSW)
'212 SPECIAL INTEREST (2.7M ESE) 240: 4-5 MILE RADIUS (4.1M SW) 217 CONTROL (10.0M SSE) 241 4-5 MILE RADIUS :(4.7M WSW).
222 SITE BOUNDARY: (0.7M N) 242 4-5 MILE RADIUS- (4.6M W)
J 223 SITE BOUNDARY (0.5M E) 243 4-5 MILE RADIUS (4.6M WNW)
$ 224 SITE BOUNDARY (0.7M ESE) 244 4-5 MILE RADIUS (4.1H NW) 225 SITE BOUNDARY (0.5M SSE) 245 4-5 MILE RADIUS (4.2M NNW)l J 226 SITE BOUNDARY. (0.5M S) 246 SPECIAL INTEREST (8.1M ENE) i 227 SITE BOUNDARY -(0.5M WSW) 247 CONTROL (7.5M ESE) ,
228 SITE BOUNDARY. (0.6M W) 248 SPECIAL INTEREST (8.2M SSE) 229 SITE BOUNDARY (0.9M NW) 249 SPECIAL INTEREST (8.1M S) j 230 4-5 MILE RADIUS (4.4M N) 250 SPECIAL INTEREST (10.3M WSW)
- 231 4-5 MILE RADIUS (4.2M NNE) 251 CONTROL (9.8M WNW) -
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a TABLE C5.0-3 (1 of'1)
CATAWBA RADIOLOGICAL-MONITORING PROGRAM ANALYSES
. ANALYSES SAMPLE MEDIUM- ' ANALYSIS SCHEDULE ' GAMMA ISOTOPIC TRITIUM LOW LEVEL I-131 GROSS BETA TLD.
- 1. . Air Radioiodine and Particulates- Weekly X
'2 . Direct Radiation Quarterly. X
- 3. . Surface Water Monthly -X-Quarterly Composite X
- 4. Drinking Water Monthly X X Quarterly Composite. X
- 5. Shoreline Sediment Semiannually X
- 6. Milk Semimonthly X X
- 7. Fish Semiannually X
- 8. Broadleaf Vegetation Monthly X
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