HL-4515, Semiannual Radioactive Effluent Release Rept for Jul-Dec 1993
| ML20064G603 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Hatch  |
| Issue date: | 12/31/1993 |
| From: | Beckman J GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| HL-4515, NUDOCS 9403160245 | |
| Download: ML20064G603 (348) | |
Text
{{#Wiki_filter:-_--__ _ _ _ _ - _ _ _ - - _ _ _ - _ _ _ - _ - _ - - - _ . Georgia Power Company
- 40 Ineemess Center Parkway Post Offce Box 1295 Birmingham, Alabama 35201 -
' Telephone 205 877-7279 J. T. Beckham, Jr. Georgia '
Power Vice President - Nuclear Hatch Project the wnhem eiwtrc system March 8, 1994 Docket Nos. 50-321 HL-4513 50-366 l i U.S. Nuclear Regulatory Commission ! ATTN: Document Control Desk 1 Washington, D.C. 20555 i Edwin 1. Hatch Nuclear Plant Semiannual Radioactive Efiluent Release Rep _o_rt Gentlemen: In accordance with the provisions of Plant Hatch Technical Specifications sections 6.9.1.8 and 6/9.1.9, Georgia Power Company is providing six copies of the Plant Hatch Units 1 and 2 Semianneal Radioactive Efiluent Release Report. This report covers the period July 1,1993 through December 31,1993. Should you have any questions, please advise. Sincerely,
, ') / N_-
J. T. Beckham, Jr. l SRM/cr
Enclosure:
Plant E. I. Hatch Units 1 & 2 Semiannual Radioactive Efiluent Release Report cc: (See next page.) 160032 9403160245 931231 h9' I PDR ADOCK 05000321 R FDR . i
Georgia Power A U.S. Nuclear Regulatory Commission Page Two
-March 8, 1994 cc: Georgia Power Commnv (w/o copics)
Mr. H. L. Sumner, Jr., General Manager - Nuclear Plant NORMS U.S Nuclear Regulatorv Commission. Washington. D.C. Mr. K. Jabbour, Licensing Project Manager - Hatch U.S Nuclear Reerdatory Commission. Region H Mr. S. D. Ebneter, Regional Administrator Mr. L. D. Wert, Senior Resident Inspector American NuclearInsurers Mr. M. Marugg State ofGeoreia Mr. J. L. Setser, Department of Natural Resources f hiahlCf 004515
GEORGIA POWER COMPANY PLANT E. I. HATCH UNITS NO.1 & 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT July 1,1993 - December 31,1993 _ . _ _ _ _ _ _ _ _ _ _ . - )
PLANT E. I. HATCH SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT SECTION TITLE PAGE 1 LIOUID EFFLUENTS 1 1.1 REGULATORY LIMITS 1 1.2 MAXIMUM PERMISSIBLE 5 CONCENTRATIONS 13 MEASUREMENTS & APPROXIMATIONS 6 OF TOTAL RADIOACTIVITY 1.4 LIQU'ID EFFLUENT RELEASE DATA 8 1.5 RADIOLOGICAL IMPACT ON MAN 10 DUE TO LIQUID RELEASES 2 GASEOUS EFFLUENTS 20 2i REGULATORY LIMITS 20 2.2 MEASUREMENTS & APPROXIMATIONS 27 OF TOTAL RADIOACTIVITY 2.3 GASEOUS EFFLUENT RELEASE DATA 32 2.4 RADIOLOGICAL IMPACT DUE TO 33 GASEOUS RELEASES 3 SOL.ID WASTE 49 3.1 REGULATORY REQUIREMENTS 49 3.2 SOLID WASTE DATA 49 m . . . .. .- .]
PLANT E.1. HATCH SEMIANNUAL RADIOACTIVE 4 EFFLUENT RELEASE REPORT SECTION TITLE PAGE 4 CHANGES TO THE PLANT HATCH ODCM 52 AND PCP 5 METEOROLOGY 53 5.1 1993 METEOROLOGICAL DATA 53
- n. - _ _ _ - -
1 PLANT E. I. HATCH SEMIANNUAL RADIOACTIVE C EFFLUENT RELEASE REPORT TABLE LIST OF TABLES PAGE l-1 TECHNICAL SPECIFICATION TABLE 3.14.1-1 3 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 1-2a LIQUID EFFLUENTS - SUMMATION OF 11 ALL RELEASES - UNIT 1 1-2b LIQUID EFFLUENTS - SUMMATION OF 12 ALL RELEASES - UNIT 2 1-2c LIQUID EFFLUENTS - SUMMATIONS OF 13 ALL RELEASES - SITE l-3a LIQUID EFFLUENTS - UNIT I 14 1-3b LIQUID EFFLUENTS - UNIT 2 15 1-3c LIQUID EFFLUENTS - SITE 16
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1-4a INDIVIDUAL DOSES i>UE 'lO LIQUID 17 RELEASES - UNIT l 1 -4b INDIVIDUAL DOSES DUE TO LIQUID 18 RELEASES - UNIT 2 1-5 LOWER LIMITS OF DETECTION - LIQUID 19 SAMPLE A.NALYSES 2-1 TECHNICAL SPECIFICATION TABLE 23 3.14.2-1 RADIO ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION c
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PLANT E. I. HATCH SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT TABLE LIST OF TABLBS PAGE 2-2a GASEOUS EFFLUENTS - SUMMATION OF 34 ALL RELEASES - UNIT 1 2-2b GASEOUS EFFLUENTS - SUMMATION fiF 35 ALL RELEASES - UNIT 2 2-2c GASEOUS EFFLUENTS - SUMMATION OF 36 ALL RELEASES - SITE 2-3a GASEOUS EFFLUENTS - ELEVATED 37 RELEASES - UNIT 1 2-3b GASEOUS EFFLUENTS - ELEVATED 38 PILEASES - UNIT 2 2-3c GASEOUS EFFLUENTS - ELEVATED 39 RELEASES - SITE 2-4a GASEOUS EFFLUENTS - GROUND- 40 LEVEL RELEASES - UNIT 1 2-4b GASEOUS EFFLUENTS - GROUND - 41 LEVEL RELEASES - UNIT 2 2-4c GASEOUS EFFLUENTS - GROUND - 42 LEVEL RELEASES - SITE 2-5 GASEOUS EFFLUENTS - DOSE 43 RATES - SITE 2-6a AIR DOSES DUE TO NOBLE GASES - 44 UNIT 1 2-6b AIR DOSES DUE TO NOBLE GASES - 45 UNIT 2
PLANT E. I. HATCH SEMI ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT TABLE LIST OF TABLES PAGE 2-7a INDIVIDUAL DOSES DUE TO RADIOIODINE, 46 TRITIUM, AND PARTICULATES IN GASEOUS RELEASES - UNIT 1 2-7b INDIVIDUAL DOSES DUE TO RADIOIODINE, 47 TRITIUM, AND PARTICULATES IN GASEOUS RELEASES - UNIT 2 2-8 LOWER LIMITS OF DETECTION - GASEOUS 48 SAMPLE ANALYSES 3- l a,b SOLID WASTE AND IRRADIATED FUEL 50,51 SHIPMENTS
RADIOACTIVE EFFLUENT RELEASE REPORT l LIOUID EFFLUENTS
- 11. REGULATORY LIMITS
- 1. The Technical Specifications presented in this section are for Unit 1.
Requirements for Unit 2 are the same as Unit 1; however, the Technical Specification numbers are not the same. TECHNICAL SPECIFICATIONS 314.1 The radioactive liquid efiluent monitoring instrumentation channels shown in table 3.14.1-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.15.1 are not exceeded. The alann/ trip setpoints of 3ese channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM). (Technical Specification Table 3.14.1-1 is included in this section as Table 1-1). 3.15.1.1 The concentration of radioactive material released at any time from the site to UNRESTRICTED AREAS (figure 3.15-1) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B Table II(column 2) for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2.0E-04 uCi/ml total activity. 315.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid efiluents released, from each reactor unit, from the site (figure 3.15-1) shall be limited to:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ.
- b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ 3151.3 The liquid radwaste treatment system, as described m the ODCM, shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid efiluent per Unit from the site (figure 315-1) when projected over the calendar quarter would exceed 0.18 mrem to the total body or 0.62 mrem to any organ.
(1)
l l 3151.4(a) The contents within any outside temporary tank chall be limited to less than or equal to 10 curies, excluding tritium and dissolved or t.ntrained noble l gases. (a) An outside temporary tank is not surrounded by liners, dikes, or walls that are capable of holding the tank contents and not having tank overflows and drains connected to the liquid radwaste treatment system. 6.9.l.9 states in part: "The Radioactive Effluent Release Report shall include (on a quarterly basis) unplanned releases from the site to unrestricted areas of l radioactive materials in gaseous and liquid efiluents that were in excess ofI Ci, excluding dissolved and entrained gases and tritium for liquid efUuents, or those I in excess of 150 Ci of noble gases or 0.02 Ci of radiciodines for gaseous releases." I l l l l 1 l l l l l l I I i (2)
l TABLE l-1 l l TECHNICAL SPECIFICATION ! TABLE 3.14.1-1 (SHEET 1 of 2) ! RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMEtTPATION Minimum Channels Instrument OPERABLE Applicability ACTION
- 1. Gross Radioactivity Monitors Providing Automatic Termina-tion of Release L1guld Radwaste Effluent Line 1 (a) 100
- 2. Gross Radioactivity Monitors not Providing Automatic Termination of I Release i
! Service Water j System Effluent Line 1 (b) 101 l 3. Flowrate Measure-( ment Devices ** l L1 quid Radwaste Effluent Line 1 (a) 102 Discharge Canal 1 (b) (a) 102
- 4. Service Water 1 At all times 103 System to Closed Cooling Water System Differential Pressure l
** Pump curves may be utilized to estimate flow; In such cases, ACTION statement 102 is not required.
(a) Whenever the radwaste discharge valves are not locked closed. (b) Whenever the service water system pressure is below the closed cooling water system pressure or differential pressure indication is not available. (3)
4 TABLE l-1 (OONTfNUED) TECHNICAL SPECIFZCATION TABLE 3.14.1-1 (SHEET 2 of 2 ) RADIOACTIVE LIOUID EFFLUEtfr MONITORING INSTRUMEtITATION l i TABLE NOTATIONS 4 i i ACTION 100 - With the number of channels OPERABLE less than l required by the Minimum Channels OPERABLE requirement, effluent releases may be continued, provided that prior to 5 initiating a release: j a. At least two independent samples are analyzed in y accordance with Specification 4.15.1.1.1. t
- b. At least two technically qua11 fled individuals '
j independently verify the release rate calcu ations ! and discharge valving. a L Otherwise, suspend release of radioactive effluents via this pathway. If the channel remains inoperable for over 3 i 30 days, an explanation of the circumstances shall be Included in the next semi-annual effluent release report. ACTION 101 - With the numbers of channels OPERABLE less than ! required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided 4 that once per shift grab samples are collected and analyzed for gross radioactivity (beta or garca) at a . Lower Limit of i 3 Detection of at least 10-7 uC1/nd. If the channel remains inoperable for over 30 days, an explanation of the - circumstances shall be included in the next semi-annual effluent release report. t ! ACTION 102 - With the nurber of channels OPERABLE less than i required by the Minimum Channels OPERABLE requirement, l effluent releases via this pathway may continue, provided
- the flowrate is estimated at least once per 4 hours during j actual releases. If the channel remains inoperable for over 30 days, an explanation of the circumstances shall be included in the next semi-annual effluent release report.
l ACTION 103 - With the number of channels OPERABLE less than i required by the Minimum Channels OPERABLE requirement, assure that the service water system effluent system monitor is OPERABLE. 4 ? J 1 (4) i i j
4 l.2 NIAXINIUN1 PERN11SSIBLE CONCENTRATIONS i The MPC values used in determining allowable liquid radwaste release rates and
' concentrations for principal gamma emitters,1-131, tritium, Sr-89, Sr-90 and Fe-55 are taken from 10CFR Part 20, Appendix B, Table II, Column 2.
For dissolved or entrained noble gases in liquid radwaste, the MPC is taken from Technical Specification 3.15.1.1 (Unit 1) and 3.11.1.1 (Unit 2) as 2.0E-04 uCi/ml. 4 For gross alpha in liquid radwaste, the MPC is taken from 10 CFR Part 20, Appendix B, Note 2.b as 3.0E-08 uCi/ml. ] Further, for all the above radionuclides or categories of radioactivity, the overall ! MPC fraction is determined in accordance with 10 CFR Part 20, Appendix B, Note 1. The method whereby the MPC fraction is used to determine release rates and liquid radwaste emuent radiation monitor setpoints is described in Section 1.3 of this report. 4 i d i il 4 4 1 i i a d
; (5)
1.3 MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY Prior to release of any tank containing liquid radwaste, and following the required recirculation, samples are collected and analyzed in accordance with Technical Specification Tables 4.15.1-1 (Unit 1) and 4.11.1-1(Unit 2). A sample from each tank planned for release is analyzed for principal gamma emitters, I-131, and dissolved and entrained noble gases by gamma spectrometry. Monthly and quarterly composites are prepared for analysis by extracting aliquots from each sample taken from tanks which are released. Liquid radwaste sample analyses are performed as follows: Measurement Frequency Method
- 1. Gamma Isotopic Each Batch Gamma spectroscopy with computerized data reduction
- 2. Dissolved or Each Batch Gamma spectroscopy Entrained with computerized Noble Gases data reduction
- 3. Tritium Monthly Distillation and Composite liquid scintillation counting
- 4. Gross Alpha Monthly Gas flow proportional Composite counting
- 5. Sr-89 and Sr-90 Quarterly Chemical separation and Composite gas flow proportional counting
- o. Fe-55 Quanerly Chemical separation and Composite low energy photon detector.
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Gamma isotopic measurements are performed in-house in the radiochemistry lab using germanium spectrometry. Three germanium detectors are available: a 20% efficient and two 15% l efficient intrinsic germanium detectors, with 2.0 FWHM resolution and housed in 4 inch-thick lead shields. A one-liter liquid radwaste sample is poured into a Marinelli beaker in preparation for a 3000 second count. A peak search of the resulting gamma ray spectrum is performed by the computer system. Energy and net count data for all significant peaks are determined, and quantitative reduction or LLD calculations are performed for the nuclides specified in Table Notation e of Technical Specification Tables 4.15.1-1 (Unit 1) and 4.11.1-1 (Unit 2): Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144. The quantitative calculations include corrections for counting time, decay time, sample volume, sample geometry, detector emeiency, baseline counts, and branching ratio. LLD calculations, including the above corrections, are made based on the counts in two standard deviations of the baseline count at the location on the spectrum where a peak for that radionuclide would be located if present. The radionuclide concentrations determined by gamma spectroscopic analysis of a sample taken from a tank planned for release and the most current sample analysis results available for tritium, gross alpha, Sr-89, Sr-90, and Fe-55 are used along with the corresponding MPC values to determine an MPC fraction for the tank planned for release. This MPC fraction is then used, with appropriate safety factors, abng with the expected dilution stream l i flow to calculate a maximum permissible release rate and a liquid emuent monitor setpoint The monitor setpoint is calculated to l assure that the limits of Technical Specifications 3.15.1.1 (Unit 1) or 3.11 1.1 (Unit 2) are not exceeded. l (7) l
A monitor reading in excess of the calculated setpoint therefore results in an automatic termination of the liquid radwaste discharge. Liquid effluent discharge is also automatically terminated if the dilution stream flow rate falls below the dilution flow rate used in the setpoint calculations and established as a setpoint on the dilution stream flow monitor. Radionuclide concentrations, safety factors, dilution stream flow rate, and liquid effluent radiation monitor calibration factor are entered into the computer and a prerelease printout is generated. If the release is not permissible appropriate warnings will be l included on the prerelease printout. If the release is permissible it is approved by the Chemistry Foreman on duty. The pertinent l information is transferred manually from the prerelease printout to a one-page release permit which is fonvarded to Radwaste Operations. When the release is completed the release permit is l returned from Radwaste Operations wi;h actual release data l included. These data are input to the computer and a post release l printout is generated. The post release printout contains actual release rates, actual release concentrations and quantities, actual dilution flow, and calculated doses to an individual. 1.4
- LIQUID EFFLUENT RELEASE DATA Regulatory Guide 1.21 Tables 2A and 2B are found in this report as Table 1-2a for Unit 1, Table 1-2b for Unit 2 and Table 1-2c for the site; and Table 1-3a for Unit 1,1-3b for Umt 2, and Table 1-3c for the site.
The values for the four categories of Tables 1-2a and 1-2b, and 1-2c are calculated and the Tables completed as follows:
- 1. Fission and activation products - The total release values (not including tritium, gases. and alpha) are comprised of the sum of the measured individual radionuclide activities. This sum is for each batch released to the river for the respective
*There were no monitors out of service greater than 30 days and no unplanned releases.
1 (8)
l quarter. Percent of applicable limit is determined from a mixed nuclide MPC fraction calculation. The average concentration for each nuclide over all released batches is divided by the corresponding individual MPC value., The sum over all nuclides 'of the C/MPC ratios times 100 is the percent of applicable limit for emuent releases during the quarter. l
- 2. Tritium - The measured tritium concentrations in the monthly composite j j
samples are used to calculate the total release and average diluted , concentration during each period. Average diluted concentration divided by l ' l the MPC limit,3.0E-03 uCi/ml, is converted to percent to give the percent of applicable limit.
- 3. Dissolved and entrained gases - Concentrations of dissolved and entrained gases in liquid effluents are measured by germanium spectroscopy on a one liter sample from each liquid radwaste batch. The average concentration of dissolved or entrained noble gases for all released batches is divided by the MPC value stated in Technical Specifications 3.15.1.1 and 3.11.1.1 (2.0E-04 uCi/ml) to determine the MPC fraction. The result x100 is the percent of l
applicable limit for noble gases in liquid effluent releases during the quarter. ! Radioisotopes ofiodine in any form t.re also determined during the isotopic analysis for each batch; therefore, a separate analysis for possible gaseous forms is not performed because it would not provide additional information. ! 4. Gross alpha radioactivity - The measured gross alpha concentrations in the j monthly composite samples are used to calculate the total release of alpha l radioactivity. I i l (9) l _i
! Other data peninent to batch releases of radioactive liquid effluent from both units are as follows: Number of batch releases: 669 l Total time period for batch releases: 86,226 minutes l Maximum time period for a ! batch release: 242.0 minutes Average time period for batch releases: 128.9 minutes Minimum time period for a batch release: 79 minutes Average stream flow during periods ! of release ofliquid effluent into a flowing stream: 3,400 CFS 1.5 RADIOLOGICAL IMPACT ON MAN DUE TO LIQUID RELEASES Doses to an individual, due to radior.ctivity in liquid effluent, were calculated in accordance with Technical Specifications 3/4.15.1.2 (Unit 1) and 3/4.11.1.2 (Unit 2) using the methodology presented in the Plant Edwin I. Hatch Offsite Dose Calculation Manual. As required by the above Technical Specifications, doses were calculated separately for Unit I and Unit 2. Results are presented in Table 1-4a for Unit I and Table 1-4b for Unit 2. 1 l i t l (10)
l TABLE 1-2a E. I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES l ____________________________________________________________l Unit Quarter 3 Quarter 4 Est. Tota Error (%) A. Fission and activation products
- 1. Total relc-ase Ci 2.67E-02 1.29E-02 4.70E+01 1
- 2. Average diluted congentration during period uCi/ml 2.34E-08 1.52E-08
- 3. % of applicable limit % 1.13E-01 9.50E-02 B. Tritium
- 1. Total release Ci 8.75E+00 5.61E+00 3.70E+01
- 2. Average diluted concentration during period uCi/ml 7.68E-06 6.58E-06
- 3. % of applicable i limit % 2.56E-01 2.19E-01 l C. Dissolved and entrained gases
- 1. Total release Ci 4.48E-04 8.11E-04 1.00E+02
- 2. Average diluted concentration during period uCi/ml 3.93E-10 9.52E-10
- 3. % of applicable limit % 1.96E-04 4.76E-04 )
D. Gross Alpha radioactivity
- 1. Total release Ci 0.00E+00 0.00E+00 1.20E+02 E. Volume of waste (prior to dilution) liters 4.96E+06 3.64E+06 1.00E+01 I
F. Volume of dilution water used liters 1.14E+09 8.52E+08 1.60E+02 l I (11) L
i TABLE 1-2b E. I. HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES 1 1 Unit Quarter 3 Quarter 4 Est. Total l Error (%) A. Fission and activation products
- 1. Total release Ci 4.22E-02 5.25E-02 4.70E+01 1
- 2. Average diluted concentration during period uCi/ml 3.29E-08 5.76E-08
- 3. % of applicable limit % 1.47E+00 1.66E+00 B. Tritium l Ci
- 1. Total release 4.33E+00 4.54E+00 3.70E+01
- 2. Average diluted concentration during period uCi/ml 3.38E-06 4.99E-06
- 3. % of applicable limit % 1.13E-01 1.66E-01 C. Dissolved and entrained gases
- 1. Total release Ci 6.94E-02 1.38E-01 1.00E+02
- 2. Average diluted concentration l
during period uCi/ml 5.42E-08 1.52E-07
- 3. % of applicable limit % 2.71E-02 7.58E-02 l
D. Gross Alpha radioactivity i
- 1. Total release Ci 0.00E+00 0.00E+00 1.20E+02 E. Volume of waste (prior to dilution) liters 5.51E+06 3.88E+06 1.00E+01 F. Volume of dilution water used liters 1.28E+09 9.10E+08 1.60E+02 l
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I TABLE 1-2c E. I. HATCH NUCLEAR PLANT - SITE i SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1L>3 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES l l l 1 Unit Quarter 3 Quarter 4 Est. Total Error (%) i A. Fission and I 1 activation products Ci 6.88E-02 6.54E-02 4.70E+01
- 1. Total re1 9 ase
- 2. Average diluted congentrat$on ,
' during period uCi/ml 2.84E-08 3.71E-08 1
- 3. %,of applicable limit % 8.29E-01 9.02E-01 B. Tritium
- 1. Total release C1 1.31E+01 1.01E+01 3.70E+01 I, 2. Average diluted t concentrat4on l during period uCi/ml 5.41E-06 5.76E-06 4
- 3. % of applicable limit % 1.80E-01 1.92E-01 C. Dissolved and i entrained gases i 1. Total release Ci 6.99E-02 1.39E-01 1.00E+02
- 2. Average diluted ,
concentration l 4 during period
- 3. % of applicable uCi/ml 2.89E-08 7.88E-08 l limit % 1.44E-02 3.94E-02 D. Gross Alpha
- radioactivity
- 1. Total release Ci 0.00E+00 0.00E+00 1.20E+02 E. Volume of waste (prior to j dilution) liters 1.05E+07 7.52E+06 1.00E+01 F. Volume of dilution i water used liters 2.42E+09 1.76E+09 1.60E+02 1
'{ (13) _..m-
.- - - , - , __r.-
TABLE 1-3ao E. I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 LIQUID EFFLUENTS Continuous Mode ** Batch Mode l Nuclides i Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 H-3 Ci 8.75E+00 5.61E+00 Fission and activation products: Na-24 C{ 7.03E-06 3.63E-05 Mn-54 C1 2.39E-04 5.63E-05 l Fe-55 C4 0.00E+00 2.73E-03 i Co-58 C1 7.01E-06 1.27E-05 l Co-60 C4 6.38E-03 2.10E-03 i Zn-65 C1 5.32E-03 2.51E-03 i As-76 C4 1.48E-05 7.50E-05 Sr-89 C1 5.91E-04 5.36E-04 Sr-90 C4 0.00E+00 6.52E-05 Sr-92 C1 0.00E+00 1.26E-06 Y-91m Cf 9.52E-07 2.17E-04 Nb-95 C1 2.10E-06 0.00E+00 Nb-97 C:. 0.00E+00 4.18E-06 Tc-99m C:. 5.90E-07 3.67E-06 Sb-125 C4 1.78E-04 2.15E-05 I-131 C1 3.75E-06 1.20E-05 I-133 C4 2.21E-05 1.66E-05 l I-134 C1 0.00E+00 2.73E-06 Cs-134 C$ 1.47E-03 4.41E-04 Cs-137 C1 1.22E-02 4.03E-03 Cs-138 Ci 7.91E-05 0.00E+00 Ba-139 C1 8.09E-06 2.83E-06 Ba-140 Ci C1 0.00E+00 1.05E-04 2.81E-06 3.40E-05 La-140 Total Ci 2.67E-02 1.29E-02 Dissolved and entrained gases: Kr-85 Ci 0.00E+00 7.51E-04 Xe-133 Ci 9.57E-05 1.71E-05 Xe-135 Ci 3.52E-04 4.31E-05 Total C1 4.48E-04 8.11E-04 Gr-Alpha Ci 0.00E+00 0.00E+00
- Zeroes in this table indicate that no radioactivity was resent above detectable levels. See Tablc 1-5 for typical ower limits of detection for liquid sample analyses
**There are no continuous mode radioactive liquid release pathways at Plant Hatch.
l l 1 (14)
1 TABLE 1-3b* E. I. HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 j LIQUID EFFLUENTS l 1 Continuous Mode ** Batch Mode j Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 H-3 Ci 4.33E+00 4.54E+00 l Fission and activation products: 1.94E-03 Na-24 C1 3.40E-03 Cr-51 7.91E-06 1.90E-03 C$ 2.16E-04 3.07E-04 Mn-54 C4 1.10E-05 2.66E-05 Mn-56 C4 l Fe-55 C1 0.00E+00 4.03E-04 ; Co-58 4.55E-05 4.74E-04 i C4 2.12E-03 1.49E-03 Co-60 C1 Zn-65 3.03E-03 8.25E-03 C4 1.43E-03 2.44E-03 As-76 C1 Sr-89 6.23E-04 6.75E-04 C4 3.23E-05 9.66E-05 Sr-90 C1 Sr-91 C$ 2.03E-03 2.10E-03 Sr-92 C1 1.98E-04 2.63E-04 Y-91m 3.93E-04 4.52E-04 C4 8.17E-05 3.74E-03 Y-92 C1 Nb-97 C4 8.60E-05 2.75E-04 Mo-99 C1 1.49E-03 2.56E-03 Tc-99m C:, 2.74E-03 5.08E-03 , I-131 C;. 2.32E-03 2.14E-03 I-132 C4 2.16E-04 1.10E-04 I-133 C1 8.78E-03 5.79E-03 I-135 C4 4.69E-03 3.21E-03 Cs-134 C1 7.74E-04 1.17E-04 Cs-136 Ci 1.33E-04 1.58E-04 Cs-137 C1 4.26E-03 9.78E-04 Ba-139 C4 0.00E+00 4.37E-06 Ba-140 C1 7.93E-04 1.20E-03 La-140 Ci 6.76E-04 2.66E-03 l Ce-141 C1 2.07E-04 2.29E-04 Np-239 C4 1.36E-03 3.35E-03 Zn-69m C1 5.98E-07 3.20E-05 Sb-124 C$ 9.74E-06 0.00E+00 Total C1 4.22E-02 5.25E-02 l Dissolved and entrained gases: Kr-85 C1 2.20E-04 0.00E+00 Xe-131m C$ 1.47E-03 2.87E-03 Xe-133m C1 1.06E-04 9.05E-04 Xe-133 C4 1.12E-02 2.91E-02 Xe-135m C1 1.09E-02 4.92E-03 Xe-135 C4 4.55E-02 1.00E-01 Ar-41 C1 2.63E-06 0.00E+00 Total Ci 6.94E-02 1.38E-01 Gr-Alpha Ci 0.00E+00 0.00E+00
- Zeroes in this table indicate that no radioactivity was l resent above detectable levels, See Table 1-5 for typical gowerlimitsofdetectionforliquidsampleanalyses
**There are no continuous mode radioactive liquid release pathways at Plant Hatch.
(15)
i TABLE 1-3c* i E. I. HATCH NUCLEAR PLANT - SITE' l SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 LIQUID EFFLUENTS I l ______________________ c entieugus Mgde** _ _____ Batch Mgde _ _______ Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 H-3 Ci 1.31E+01 1.01E+01 Fission and activation products: Na-24 C4 3.41E-03 1.98E Cr-51 C4 7.91E-06 1.90E-03 Mn-54 C1 4.55E-04 3.63E-04 Mn-56 C: 1.10E-05 2.66E-05 Fe-55 C:. 0.00E+00 3.14E-03 Co-58 C4 5.25E-05 4.87E-04 Co-50 C1 8.49E-03 3.59E-03 Zn-65 C4 8.36E-03 1.08E-02 As-76 C4 1.44E-03 2.52E-03 Sr-89 C4 1.21E-03 1.21E-03 Sr-90 C1 3.23E-05 1.62E-04 Sr-91 C$ 2.03E-03 2.10E-03 Sr-92 C1 1.98E-04 2.64E-04 Y-91m C:. 3.94E-04 6.69E-04 Y-92 C:. 8.17E-05 3.74E-03 Nb-95 C4 2.10E-06 0.00E+00 Nb-97 C1 8.60E-05 2.79E-04 Mo-99 C4 1.49E-03 2.56E-03 Tc-99m C: 2.74E-03 5.08E-03 Sb-125 C: 1.78E-04 2.15E-05 I-131 C:. 2.32E-03 2.15E-03 I-132 C:. 2.16E-04 1.10E-04 I-133 C:. 8.80E-03 5.81E-03 I-134 C4 0.00E+00 2.73E-06 I-135 C1 4.69E-03 3.21E-03 Cs-134 C% 2.25E-03 5.58E-04 Cs-136 C4 1.33E-04 1.58E-04 Cs-137 C4 1.65E-02 5.00E-03 Cs-138 C4 7.91E-05 0.00E+00 Ba-139 C1 8.09E-06 7.20E-06 Ba-140 Ci 7.93E-04 1.20E-03 La-140 Ci 7.82E-04 2.69E-03 Ce-141 C1 2.07E-04 2.29E-04 Np-239 Ci 1.36E-03 3.35E-03 i Zn-69m C1 5.98E-07 3.20E-05 l Sb-124 Ci 9.74E-06 0.00E+00 Total C1 6.88E-02 6.54E-02 Dissolved and entrained gases: Kr-85 Ci 2.20E-04 7.51E-04 Xe-131m C4 1.47E-03 2.87E-03 l Xe-133m C1 1.06E-04 9.05E-04 Xe-133 C4 1.13E-02 2.91E-02 Xe-135m C1 1.09E-02 4.92E-03 Xe-135 Ci 4.58E-02 1.00E-01 Ar-41 Ci 2.63E-06 0.00E+00 l Ci 6.99E-02 1.39E-01 Total Gr-Alpha Ci 0.00E+00 0.00E+00 l
- Zeroes in this table indicate that no radioactivity was resent ahove detectab ee Table 1-5 for typical l
' gower1mitsofdetect{elevels,onfor11gugdsampleanalyses
**There are no continuous mode radioactive liquid release pathways at Plant Hatch.
i (16)
' TABLE 1-4a E. I. HATCH NUCLEAR PLANT - Unit 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 INDIVIDUAL DOSES DUE TO LIQUID RELEASES l Cumulative Doses per Quarter _______________ ______________________________________________ter organ Tech Unit Quarter 3 % of Quar 4 % of Spec Tech Tech Limit Speg Speg Limit L1mit _______________________________________________6E-02 Bone 5.0 mrem 4.26E-02 8.52E-01 1.4 2.91E-01 Liver 5.0 mrem 6.28E-02 1.26E+00 2.13E-02 4.26E-01 TBody 1.5 mrem 4.24E-02 2.83E+00 1.44E-02 9.57E-01 Thyroid 5.0 mren 3.47E-05 1.69E-03 1.13E-04 2.25E-03 Kianey 5.0 mrem 2.17E-02 4.34E-01 7.45E-03 1.49E-01 Lung 5.0 mrem 6.90E-03 1.38E-01 2.33E-03 4.67E-02 GILLI 5.0 mrem 5.15E 03 1.03E 1.93E _____________3.86E 02 ______________________________________-01 __________-03 Cumulative Doses This Year Organ Tech Unit Quarters % of l 1,2,3,& 4 Tech Speg . Limit Spec l Limit l Bone 10.0 mrem 1.20E-01 1.20E+00 l Liver 10.0 mrem 1.98E-01 1.98E+00 TBody 3.0 mrem 1.27E-01 4.24E+00 Thyroid 30.0 mrem 5.08E 5.08E-02 Kianey 10.0 mrem 8.04E-02 8.04E-01 Lung 10.0 mrem 1.77E-02 1.77E-01 GILLI 10.0 mrem 5.69E 02 5.69E _______________________ ______________________________________-01 i l l (17)
d TABLE 1-4b E. I. HATCH NUCLEAR PLANT - Unit 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 INDIVIDUAL DOSES DUE TO LIQUID RELEASES 1 1 l l I Cumulative Doses per Quarter -------------------------- 1 Organ Tech Unit Quarter 3 of Tech Quarter 4 % of Tech Spec Speg Speg Limit Limit 2 Limit --------------------- i
---------------------------------------01 Bone 5.0 mrem 2.10E-02 4.19E-6.42E-01 6.01E-03 1.20E-01 1.06E-02 2.11E-01 Liver 5.0 mrem 3.21E-02 TBody 1.5 mrem 2.18E-02 1.46E+00 6.46E-03 4.31E-01 Thyroid 5.0 mrem 2.21E-02 4.43E-01 1.98E-02 3.97E-01 Kidney 5.0 mrem 1.15E-02 2.30E-01 1.12E-03 2.24E-02 Lung 5.0 mrem 3.49E-03 6.98E-02 7.95E-04 1.59E-02 GILLI 5.0 mrem 3.14E-03 6.27E-02 4.33E-03 8.66E-02 1
Cumulative Doses This Year Organ Tech Unit Quarters % of i Speg 1,2,3,& 4 Tech Limit Spec l Limit Bone 10.0 mrem 7.54E-02 7.54E-01
- Liver 10.0 mrem 1.19E-01 1.19E+00 l TBody 3.0 mrem 8.05E-02 2.68E+00 Thyroid 10.0 mrem 1.06E-01 1.06E+00 Kidney 10.0 mrem 4.01E-02 4.01E-01 Lung 10.0 mrem 1.24E-02 1.24E-01 GIILI 10.0 mrem 1.58E-02 1.58E-01
, 1 + d n (18)
i l TABLE l-5 LOWER LIMITS OF DETECTION - LIQUID SAMPLE ANALYSES The values in this table represent apciori lower limits of detection (LLD) which are typically achieved in laboratory analyses ofliquid radwaste samples. RADIONUCLIDE LLD UNITS Mn-54 5.38E-08 uCi/ml Fe-59 7.78E-08 Co-58 4.67E-08 Co-60 4.78E-08 Zn-65 1.31E-07 Mo-99 5.10E-07* Cs-134 7.18E-08 ! Cs-137 6.05E-08 Ce-141 1.41E-07 Ce-144 6.30E-07
- l l-131 6.51E-08 l Xe-135 8.45E-08 Fe-55 2.00E-06 l
l Sr-89 5.00E-08
- Sr-90 5.00E-08 H-3 1.00E-05
*In accordance with Technical Specification Tables 4.15.1-1 (Unit 1) and 4.11.1-1 (Unit 2),
Table Notation b, the permissible Lower Limit of Detection may be increased inversely proportional to the magnitude of the gamma yield. However, the LLD determined in this manner must not exceed 10 percent of the Maximum Permissible Concentration (MPC) value specified in 10 CFR20, Appendix B, Table II (Column 2). l t (19)
l l l 2 GASEOUS EFFLUFmNTS ! 2.1 REGULATORY LIMITS The Technical Specifications presented in this section are for Unit 1. Requhements for Unit 2 are the same as for Unit 1; however, the Technical Specification numbers are not the same. TECHNICAL SPECIFICATIONS l 3.14.2 The radioactive gaseous effluent monitoring instrumentation channels shown in table 3.14.2-1 shall be OPERABLE with their alarm / trip setpoints set to l ensure that the limits of Specification 3.15.2.1 (a) are not exceeded. The ! alarm / trip setpoints of these channels shall be determined in accordance with the i ODCM. (Technical Specification Table 314.2-1 is included in this section as l Table 2-1). j 3.15.21 The dose rate at any time in the UNRESTRICTED AREAS (figure l 3.15-1) due to radioactive materials released in gaseous emuents from the site shall be limited to the following values:
- a. The dose rate limit for noble gases shall be less than or equal to 500
- mrem / year to the total body and less than or equal to 3000 mrem / year to the skin.
l l l b. The dose rate limit for I-131, I-133, tritium, and for all radioactive ! materials in particulate form and radionuclides other than noble gases l with half-lives greater than 8 days shall be less than or equal to 1500 mrem / year to any organ. ; I 3.15.2.2 The air dose in UNRESTRICTED AREAS (figure 3.15-1) due to noble ) gases released in gaseous emuents from each reactor unit shall be limited to the l
- following
l a. During any calendar quarter, to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation. l
- b. During any calendar year, to less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
l l l (20) i
l l 3.15.2.3 The dose to any organ of a MEhfBER OF THE PUBLIC from l-131, l I-133, tritium , and all radionuclides in particulate form with half-lives greater i than 8 days in gaseous effluents released to UNRESTRICTED AREAS (figure 3.15-1) from each reactor unit shall be limited to the following: 1 1
- a. During any calendar quarter to less than or equal to 7.5 mrem to any organ.
- b. During any calendar year to less than or equal to 15 mrem to any organ. j 3.15.2.4 The GASEOUS RADWAs ' _ 'REATMENT SYSTEM as described in the ODCM shall be in operation. (Th;: icchnical Specification applies whenever the main condenser air ejector system is in operation.)
4.15.2.4 GASEOUS RADWASTE TREATMENT SYSTEM operability shall be demonstrated by administrative controls which assure that the offgas treatment system is not bypassed. 3.15.2.5 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem. l (With the calculated doses from the release of radioactive materials in liquid or gaseous efiluents exceeding twice the limits of Specifications 3.15.1.2(a), 3.15.1.2(b), 3.15.2.2.(a), 3.15.2.2(b), 3.15.2.3(a), or 3.15.2.3(b), calculations shall be made including direct radiation ccatributions from the reactor units and from outside 3torage tanks to determine whether the above l limits of Specification 3.15.2.5 have been exceeded.) 3.15.2.6 The concentration of hydrogen downstream of the recombiners in the main condens r offgas treatment system shall be limited to less than or equal to 4 percent by ' r,ume. 3.15.2.7 The gross gamma radioactivity rate of the noble gases Xe-133, Xe-135, ! Xe-138, Kr-85m, Kr-87, and Kr-88 measured at the main condenser evacuation system pretreatment monitor station shall be limited to less than or equal to l 240,000 uCi/second. I (21) l i
6 9. l.9 states in part:
"The Radioactive Emuent Release Report shallinclude (on a quarterly basis) unplanned releases from the site to unrestricted areas of radioactive materials in ga seous and liquid emuents that were in excess of I Ci, excluding dissolved and entrained gases and tritium for liquid efHuents, or those in excess of 150 Ci of noble gases or 0.02 Ci of radiciodines for gaseous releases."
r i l t l (22) l I
r f TABtt 3 . 1 86 . 2 - 1 ( ;llt t i 1 OF to ) HADIOACityt GA31005 (FFLUltil M04_ITORING INSTRUMENTATION M i sil mum Clearine i s Parametor ACTIOM OPE RAug Appiiga.bIiity Mitrument
- 1. Ma in Condense r. 0f t'gaa treatmont system Explosive cas Monitoring System
** % Hydrogen 106 Hydrogen Mont tor (1) q
- 2. Reactor Building vent Stack >
Monitoring System Of
- Radioactivity Rate 105 M Noble Gas Activity Monitor (1) Measurement + M a.
Verify Presence of 107 w
- I lodine Sampler Ca rtridge (1) Cartridge w C b.
U
- Verify Presence of 107
- c. Particulate Sampler F il ter (1) Filter _
t.n 7 (D Effluent Systum Itownate
- System flowrote 104 (D
- d. (1)
Measurement Dovice Measurement tt Sampler f Iowre to Moasurement
- Sampler Flowrote 104 e.
Device (1) Mea suremen t O m u
- 3. Recomberee Bul le ing vent t ia t iosi w Monitoring b/t'em
- Radioactivity Rate 105 t a. Moble Ca s Activi ty Moni tor (I) Measurement +
- VGrify Presence of 107 *
- b. todine Sampler Cartridge (1) Cart ridge l
- Verify Presence of 107 Particulate Sampler I iltor (t) Filter .
- c.
- i
- Sempler Flowrate 104
- d. Sampler Flowrate Measurceent (1) Measurement '
Device i I
I A8i t 3.18.2-1 (SitLET 2 0F 4) 4 INSTRUMENTATION
- 14A0lOACi lVE GASEOUS [F f t Ut fil__HJJPi1TORING .
Ministem
- Cttanno t s Pa ramete r ACTION OPERAtiLE Applicability Instrument
- 4. Main Stack Monitoring System Radioactivity Rate 105
- a. Nocle Ces Activity Monitor (1) Measurement +
i
- Verify Presence of 107
- b. Bodine Sampler Cartrldge (1) Cartridge , ,
- Vertry Presence or 107 d Particulate sempler F ilter (1) Filter c.
p trj Effluent System F lowrate
- System Flowrote 1G4 ,
^
d. Measuring Devices (1) Measurement w y
- Sampler Flowra te - 104 H 8
m
- e. Sampler F lowra te Meastering (1) Measurement Device -
g 5 Cosidenser Orrgas Pret reatment
- W Monitor C
*** Radioactivity Rate 108 O Noble Gas Activity Monitor (ij Measurement ct M
O r m da w i 4 e
~
TABLE 2-1 (Sheet 3 of 4) TABLE 3.14.2-1 (SHEET 3 0F 4) 4 5 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION
- Table Notations
+ Monitor must be capable of responding to a Lower Limit of i
- Detection of 1 x 10-4 yCi/m1.
*0uring releases via this pathway.
**During main condenser offgas treatment system operation.
***During operation of the main condenser air ejector.
ACTION 104 - With the number of channels.0PERABLE less than required by the Minimus Channels OPERABLE requirement, I effluent releases via this pathway may continue, provided the flowrate is estimated at least once per 4 hours. i i If the number of channels OPERABLE remains less than ~ j required by the Minimum Channels OPERABLE requirement for ' over 30 days, an explanation of the circumstances shall be included in the next semi-annual ef fluent release' report. ACTION 105 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, i effluent releases via this pathway may continue, provided i grab samples are taken daily and analyzed daily for gross activity within 24 hours. With the number of main stack monitoring system channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, without delay
- ' suspend drywell purge.
If the number of channels OPERABLE remains less than required by the Minimum Channels OPERABLE requirement for l' over 30 days, an explanation of the circumstances shall be included in the next semi-annual effluent release report. ACTION 106 - With the number of channels OPERABLE less than required by the Minimum Chanoels OPERABLE requirement, l operation of the main condenser offgas treatment system may 4 continue provided: (a) Gas samples are collected once per 4 hours and analyzed within the ensuing 4 hours, or [ (b) Using a temporary hydrogen analyzer installed in the i offgas system line downs,trean of the recombiner, - hydrogen concentration readings are taken and logged 7 every 4 hours. ( 4 ' (25) , 1 e-
TABLE 2-1 (Sheet 4 of 4) TABLE 3.14.2-1 (SHEET 4 0F 4) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION l Table Notations (Continued) i If the number of channels OPERABLE remains less than required by the Minimus Channels OPERABLE requirement for - over 30 days, an explanation of the circumstances shall be l included in the next semi-annual effluent release report. ACTION 107 - With the number of channels OPERABLE less than I required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided samples are continuously collected with auxiliary sampling equipment for periods on the order of 7 days and analyzed within 48 hours af ter the end of the sampling period. If the number of channels OPERABLE remains less than required by the Minimum Channels OPERABLE requirement for over 30 days, an explanation of the circumstances shall be included in the next semi-annual effluent release report. i ACTION 108 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, release to the environment may continue for up to 72 hours provided: ! a. The offgas system is not bypassed, and
- b. The offgas post-treatment monitor (011-K615) or the main l
t ' stack monitor (011-K600) is OPERABLE. l Otherwise, be in at least HOT STANDBY within 12 hours. If the number of channels OPERABLE remains less than required by the Minimum Channels OPERABLE requirement for over 30 days, an explanation of the circumstances shall be included in the next semi-annual ef fluent release report. i 1 (26) ,
22 MEASUREhfENT AND APPROXINIATIONS OF TOTAL RADIOACTIVITY Waste gas release at Plant Hatch is confmed to four paths: main stack (also called the offgas vent), Unit I reactor building vent; Unit 2 reactor building vent, and the recombiner building vent. Each of these four paths is continuously monitored for gaseous radioactivity. Each is equipped with an integrating-type sample collection device for collecting particulates and iodines. Sample collection is in accordance with Technical Specification Tables 4.15.2-1 (Unit 1) and 4.11.2-1 (Unit 2). Unless required more frequently under certain circumstances specified in Table Notations to the above mentioned tables, samples are collected as follows:
- 1. Noble gas samples are collected by grab sampling monthly.
- 2. Tritium samples are collected by grab sampling monthly.
- 3. Radioiodine samples are collected by pulling the sample stream through a charcoal cartridge over a 7-day period.
- 4. Particulates are collected by pulling the sample stream through a particulate filter over a 7-day period.
- 5. The 7-day particulate filters above are analyzed for gross alpha activity.
- 6. Quarterly composite samples are prepared from the particulate filters collected over the previous quarter and the quarterly composite sample is analyzed for Sr-89 and Sr-90.
Sample analyses results and release flow rates from the four release points form the basis for calculating released quantities of radionuclide-specific radioactivity, dose rates associated with gaseous releases, and cumulative doses for the current quarter and year. This task is normally performed with computer assistance. l l l l l [ (27) f s
The noble gas grab sample analysis results are used along with maximum expected release flow rates from each of the four vents to calculate monitor setpoints for the gaseous effluent monitors serving the four release points, to assure that the limits of Technical Specifications 3.15.2.1.a (Unit 1) or 3.11.2.1.a (Unit 2) are not exceeded. Calculation of monitor setpoints is described in the Plant Hatch ODCM. With each release period released radioactivity, dose rates, and cumulative doses are calculated. Cumulative dose results are tabulated along with percent of Technical Specification limits (3.15.2.2 and 3.15.2.3 (Unit 1); 3. I1.2.2 and 3.11.2.3 (Unit 2) for each release, for the current quarter and year. After each calendar quarter (13 weeks) a summary of waste gas releases from the four vents is compiled for preparation of the Semiannual j Efiluent Release Report required by Technical Specifications 6.9.1.8 and 6.9.1.9 and described in NRC Regulatory Guide 1.21. The e ethods for determining released quantities of radioactivity, dose raws and cumulative doses are as follows:
- 1. FISSION AND ACTIVATION GAS
! The radionuclide-specific released radioactivity is determined from sample analyses results collected as described above and average release flow rates over the period represented by the collected sample. l l Instantaneous dose rates due to noble gases and due to radiciodines, tritium, and particulates are calculated (with computer assistance). Calculated dose rates are compared to the dose rate limits specified in 3.15.2.1.a (Unit 1) and 3.11.2.1.a (Unit 2) for noble gases; and 3.15.2.1.b (Unit 1) and 3.11.2.1.b (Unit 2) for radiciodine, tritium, and particulates. Dose rate calculation methodology is presented in the Plant Hatch ODCM. l l l l (2d) l l
+s- Ls-4 u . - . . n 1
I i Beta and gamma air doses due to noble gases are calculated for the location in the unrestricted area with the potential for the highest exposure due to l gaseous releases. Air doses are calculated for each release period and cumulative totals are kept for each unit for the current calendar quader and year. Cumulative air doses are compared to the dose limits specified in Technical Specifications 3.15.2.2 (Unit 1) and 3. I1.2.2 (Unit 2). Current percent of technical specification limits are shown on the printout for each
- release period. Air dose calculation methodology is presented in the Plant I
Hatch ODCM.
- 2. RADIOIODINE, TRITIUM, AND PARTICULATE RELEASES Released quantities of radiciodines are determined from the weekly samples and release flow rates for the four release points. Radioiodine concentrations are determined by gamma spectroscopy.
Released quantities of particulates are determined from the weekly (filter) samples and release flow rates for the four release points. Gamma spectroscopy is used to quantify concentrations of principal gamma emitters. Aftec each calendar quartar the paniculate filters from each vent are combined, fused, and strontium separation is performed. Since sample flows l and vent flows are almost constant over each quarterly period the filters l from each vent can be dissolved together. Decay corrections are made back ! to the middle of the quarterly collection period. Where significant Sr-89 or Sr-90 is not detected, LLD's are calculated. Strontium concentrations are input to the composite file of the computer to be used in release, dose rate and individual dose calculations. l Tritium samples are obtained monthly from each vent by passing the sample ; i l stream through a cold trap. The grams of water vapor / cubic foot gas is l measured upstream of the cold trap in order to alleviate the difficulties in l determining water vapor collection efliciencies. The tritium samples are l analyzed by an independent laboratory and results are fumished in uCi/ml of water. The tritium concentration in water is converted to tritium concentration in air and this value is input into the composite file of the computer to be used in release, dose rate, and individual dose calculations. j 1 l (29)
Dose rates due to radiciodine, tritium, and particulates are calculated for a hypothetical child, exposed to the inhalation pathway, at the !ocation in the ! unrestricted area where the potential dose rate is expected to be the highest. Dose rates are calculated for each release point, for each release period, and ! the total dose rate from all four release points are compared to the dose rate i limits specified in Technical Specifications 3.15.2.1,b (Unit 1) or 3.11.2.1.b l (Unit 2). Individual doses due to radiciodine, tritium, and particulates are calculated l l for the critical receptor, which is described in the Plant Hatch ODCM. Individual doses are calculated for each release period and cumulative totals l are kept for each unit for the current calendar quarter and year. Cumulative individual doses are compared to the dose limits specified in Technical l l Specifications 3.15.2.3 (Unit 1) and 3.11.2.3 (Unit 2). Current percent of technical specification limits are shown on the printout for each release period. i ! 3. GROSS ALPHA RELEASE The gross alpha release is computed each month by counting the particulate filters each week for gross alpha activity in a gas flow proportional counter. The four or five weeks' numbers are then recorded on a data sht and the activity is summed at the end of the month. The summed activity is then ! divided by the total monthly volume to determine the concentration. This concentration is input to the composite file of the computer and is used for release calculations.
- 4. ERROR ESTIMATES l
Regulatory Guide 1.21 requires that estimated total error in analysis l techniques be reported. These estimates are required for the total fission and ; l
- activation gas release, total I-131 re! ease, total particulates with half-lives I greater than 8-day release, and total tntium release.
l l (30) i __ _
l "The total or maximum error associated with the emuent measurement will include the cumulative errors resulting from the total operation of sampling l ! and measurement. Because it may be very diflicult to assign error terms for l each parameter affecting the final measurement, detailed statistical evaluation of error are not suggested. The objective should be to obtain an overall estimate of the error associated with measurements of radioactive materials released in liquid and gaseous efIluents and solid waste." l Estimated errors are based on errors in counting equipment calibration, counting ; l statistics, vent flow rates, vent sample flow rates, non-steady release rates, chemical yield factors, and sample losses for such items as charcoal canridges. (1) Fission and Activation Total Release was calculated from sample analysis results and release point flow rates. l l Statistical Error 60 % Counting Equipment Calibration 10 % i Vent Flow Rates 10 % l Non-Steady Release Rates 20 % 100 % l (2) 1-131 Release was calculated from each weekly sample: Statistical Error 60 % Counting Equipment Calibration 10 % l Vent Flow Rates 10 % Vent Sample Flow Rates 10 % l l Non-Steady Release Rates 10 % ! Losses From Charcoal Cartridge 10 % 110 % l (3) Particulates with half-lives greater than 8 days release was calculated from sample analysis results and release point flow rates. , l Statistical Error at LLD concentration 60 % Counting Equipment Calibration 10 % Vent Flow Rates 10 % Vent Sample Flow Rates 10 % l Non-Steady Release Rates 10% l 100 % l (31)
l (4) Total Tritium Release was dominated by the reactor building vent tritium l release; hence, the larger statistical errors of the otT-gas vent and recombiner building vent tritium releases do not atTect the error in the total tritium release-L Water Vapor in Sample Stream Determination 20% Vent Flow Rates 10 % l Counting Calibration and Statistics 10 % Non-Steady Release 50% 90 % l l 2.3
- GASEOUS EFFLUENT RELEASE DATA l Regulatory Guide 1.21 Tables 1 A, IB, and IC are found in this report as Table 2-2a for Unit 1,2-2b for Unit 2, and Table 2-2c for site; and Table 3-3a for Unit I, Table 3-3b for Unit 2 and Table 3-3c for site; and Table 2-4a for Unit 1 Table 2-4b for Unit 2, and Table 2-4c for site. Data are presented on a quarterly basis as required by Regulatory Guide 1.21.
1 To complete Tables 2-2a-c, total release for each of the four categories (fission i and activation gases; iodines; particulates; and tritium) was divided by the number of seconds in the quarter to obtain a release rate in uCi/second for each category. However, the applicable Technical Specification limits are not in terms of release rate in uCi/second but in terms of dose rate in mrem / year, as presented in Technical Specifications 315.2.1 (Unit 1) and 3.11.2.1 (Unit 2). Noble gases are l limited as specified in 3.15.2.1.a and 3.11.2.1.a. The other three categories (tritium, radioiodines, and particulates) are limited as a group as specified in 3.15.2.1.b and 3.11.2.1.b. Further the limits specified in Technical Specifications 3.15.2.1 and 3.11.2.1 are site limits, not unit limits. Dose rates due to noble gas releases and due to radiciodine, tritium, and particulates are presented in l Table 2-5 along with percent of technical specification limits. 1 Gross alpha radioacthity is reported in Tables 2-2a,2-2b, and 2-2c as curies ' released in each quarter. Limits for cumulative beta and gamma air doses, due to noble gases, are specified in Technical Specifications 3.15.2.2 (Unit 1) and 311.2.2 (Unit 2). These limits are unit limits. Cumulative air doses are presented in Tables 2-6a and 2-6b, along with percent of technical specification limits. i l
*There were no unplanned releases.(See Section 5 for Monitor out of service informuion)
(32)
- _. - - =_ _
! Limits for cumulative individual doses, due to radiciodine, tritium, and I particulates, are specified in Technical Specifications 3.15 2.3 (Unit 1) and 3.11.2.3 (Unit 2). These limits are also unit limits. Cumulative individual doses are presented in Tables 2-7a and 2-7b, with percent of technical specification l limits. l 2.4 RADIOLOGICAL IMPACT DUE TO GASEOUS RELEASES l Dose rates due to noble gas releases were calculated for the site in accordatice with Technical Specifications 3/4.15.2.1.a (Unit 1) and 3/4. I 1.2.1.a (Unit 2). Results are presented in Table 2-5. Dose rates due to radiciodine, tritium, and l particulates in gaseous releases were calculated in accordance with Technical l l Specifications 3/4.15.2.1.b (Unit 1) and 3/4.11.2.1.b (Unit 2). These results are also in Table 2-5. Cumulative air doses due to noble gas releases were calculated for each unit in I accordance with Technical Specification 3/4.15.2.2 (Unit 1) and 3/4.11.2.2 l (Unit 2). These results are presented in Table 2-6a for Unit I and Table 2-6b for Unit 2.. Cumulative doses to an individual due to radiciodine, tritium , and particulates were calculated for each unit in accordance with Technical Specifications 3/4.15.2.3 (Unit 1) and 3/4.11.2.3 (Unit 2). These results are presented in Table 2-7a for Unit I and Table 2-7b for Unit 2. l Dose rates and doses were calculated using the methodology presented in the Plant Hatch Offsite Dose Calculation Manual. l l t (33) l
TABLE 2-2a E. I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES Unit Quarter 3 Quarter 4 Est. Total Error (%) A. Fis ion and Act$vation Gases l
- 1. Total Ci 1.46E+02 5.67E+02 1.00E+02 i Release
- 2. Average uCi/sec 1.85E+01 7.55E+01 Release Rate For Period
*3. % of Tech %
Spec Limit i i B. Iodines ! 1. Total Ci 4.27E-02 7.08E-02 1.10E+02 l Iodine-131
- 2. Average uCi/sec 5.44E-03 9.42E-03 Release Rate For Period
! *3. % of Tech % l Spec Limit C. Particulates
- 1. Particulates Ci 2.42E-02 1.32E-02 1.00E+02 with half-lives > 8 days
- 2. Average uCi/sec 3.08E-03 1.76E-03 Release Rate For Period
*3. % of Tech %
Spec Limit
- 4. Gross Alpha Ci 1.35E-06 1.08E-06 Radioactivity D. Tritium
- 1. Total Ci 5.59E+00 7.48E+00 9.00E+01
- 2. Average uCi/sec 7.11E-01 9.95E-01 Release Rate For Period
*3. % of Tech %
Spec Limit
- Technical Specification limits are in terms of dose rate See Tables 2-5, 2-6a, 2-6b, l
(mrem an 2-7b dose (mrem).
.-7a,/yrgand l
l (34)
TABLE 2-2b E. I. HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASF REPORT 1993 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES Unit Quarter 3 Quarter 4 Est. Total Error (%) A. Fission and Activation Gases
- 1. Total Ci 9.04E+02 1.46E+03 1.00E+02 Release
- 2. Average uCi/sec 1.15E+02 1.94E+02 Release Rate For Period
*3. % of Tech %
Spec Limit B. Iodines
- 1. Total Ci 2.57E-02 5.13E-02 1.10E+02 Iodine-131
- 2. Average uCi/sec 3.26E-03 6.82E-03 Release Rat 9 For Period
*3. % of T9ch %
Spec Limit C. Particulates
- 1. Particulates Ci 3.40E-02 9.27E-03 1.00E+02 with half-lives > 8 days i 2. Average uCi/sec 4.33E-03 1.23E-03 Release Rate For Period
*3. % of Tech %
Spec Limit
- 4. Gross Alpha Ci 5.63E-07 8.51E-07 l Radioactivity l D. Tritium
- 1. Total Ci 1.55E+01 1.25E+01 9.00E+01
- 2. Average uCi/sec 1.97E+00 1.66E+00 Release Rate For Period
*3. % of Tech %
Spec Limit
- Technical Specification limits are in terms of dose rate l
' fmrem/yrganddose
-7a, an 2-7b (mrem). See Tables 2-5, 2-6a, 2-6b, l os)
l l TABLE 2-2c E. I. HATCH NUCLEAR PLANT - SITE SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES Unit Quarter 3 Quarter 4 Est. Total Error (%) A. Fis ion and Act$vation Gases
- 1. Total Ci 1.05E+03 2.03E+03 1.00E+02 Release
- 2. Average uCi/sec 1.34E+02 2.69E+02 Release Rate For Period
*3. % of Tech %
Spec Limit B. Iodines l 1. Total Ci 6.84E-02 1.22E-01 1.10E+02 Iodine-131
- 2. Average uCi/sec 8.70E-03 1.62E-02 Release Rate For Period
*3. % of Tech %
j, Spec Limit l C. Particulates
- 1. Particulates Ci 5.83E-02 2.25E-02 1.00E+02 with half-lives > 8 i days
- 2. Average uCi/sec 7.41E-03 2.99E-03 '
Release Rate For Period
*3. % of Tech %
Spec Limit
- 4. Gross Alpha Ci 1.92E-06 1.93E-06 Radioactivity D. Tritium
- 1. Total Qi 2.11E+01 2.00E+01 9.00E+01
- 2. Average UC1/sec 2.68E+00 2.66E+00 Release Rate For Period
*3. % of Tech %
Spec Limit
- Technical Specification limits are in terms of dose rate fmrem/yrganddose
-7a, an 2-7b (mrem). See Tables 2-5, 2-6a, 2-6b, (36) l
TABLE 2-3a l E. I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 GASEOUS EFFLUENTS - ELEVATED RELEASES
- 1 l
Continuous Mode Batch Mode ** Nuclides ( Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 i
- 1. Fission Gases Kr-85m C4 1.21E+00 6.35E+00 Kr-87 C1 3.62E+00 4.87E+00 t Kr-88 C4 0.00E+00 3.56E+00 Xe-133 C1 1.33E+01 4.66E+01 Xe-135m Ci 1.80E+01 1.90E+01 Xe-135 C1 3.90E+00 3.98E+00 Xe-138 Ci 6.89E+01 6.48E+01 Ar-41 Ci 0.00E+00 7.67E 01 1
__-_---__-___-____---_______----_--_ ____-________-_________- _ l Total For Period Ci 1.09E+02 1.50E+02
- 2. Iodines I-131 C$ 1.31E-04 3.78E-02 ,
! I-133 C1 9.06E-04 4.71E-02 ! 1.07E 02 l I I 135 Ci 1.58E 03
- -_______-_______________________-___ ___---_______________--- _ 1 Total For Period Ci 2.62E 9.56E ______-___ -_-___________
- - _ _ _ _ _ _ - - _ _ - _ - - - _ _ - _ _ ____-______-02
_ _ _- 0 3
- 3. Particulates Mn-54 Ci 0.00E+00 5.75E-08 Co-58 C4 0.00E+00 6.53E-08 Co-60 C1 1.20E-06 2.44E-07 Zn-65 Ci 0.00E+00 9.86E-07 Sr-89 Ci 1.70E-06 2.42E-06 Sr-90 Ci 0.00E+00 3.66E-07 5.02E-08 5.65E-07 Cs-137 C1 .
Ba-140 C4 8.75E-06 1.40E-04 l La-140 C1 1.25E-05 2.58E-04 ! I 131 Ci 1.56E 9.69E 07 1
-__--_ _ _-_-______---_-____-07 __________ __________________-_____
Total For Period Ci 2.47E 05 4.03E __-______-04 _________________________ l
- Zeroes in this table indicate that no radioactivity was See Table 2-8 for ty present above detectable levels. lower limits of detection for gaseous sample analyses.pical I **There are no batch mode radioactive gaseous release l pathways at Plant Hatch.
(37)
TABLE 2-3b E. I. HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 i GASEOUS EFFLUENTS - ELEVATED RELEASES * ., Continuous Mode Batch Mode ** Nuclides Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 l Released l
- 1. Fission Gases Kr-85m C4 1.21E+00 6.35E+00 Kr-87 C1 3.62E+00 4.87E+00 Kr-88 C4 0.00E+00 3.56E+00 Xe-133 C1 1.33E+01 4.66E+01 r
i Xe-135m Ci 1.80E+01 1.90E+01 Xe-135 C1 3.90E+00 3.98E+00 t Xe-138 C4 6.89E+01 6.48E+01 Ar C1 0.00E+00 7.67E ___________________ l __-____-41 ______ -________________-____-01 -____ Total For l Period Ci 1.09E+02 1.50E+02 i l 2.lodines l! _______________________________________________-_____________ I-131 C4 1.31E-04 3.78E-02 I-133 C1 9.06E-04 4.71E-02 I 135 Ci 1.58E 03 1.07E -________________________ _________________________________-__-02 Total For Period Ci 2.62E 03 9.56E 02 l _____________________________________________________________
- 3. Particulates
! Mn-54 Ci 0.00E+00 5.75E-08 Co-58 Ci 0.00E+00 6.53E-08 l Co-60 Ci 1.20E-06 2.44E-07 Zn-65 C$ 0.00E+00 9.86E-07 Sr-89 C4 1.70E-06 2.42E-06 ! Sr-90 C4 0.00E+00 5.02E-08 l Cs-137 C1 3.66E-07 5.65E-07 l Ba-140 Ci 8.75E-06 1.40E-04 La-140 C1 1.25E-05 2.58E-04 I 131 Ci 1.56E 9.69E 07 _____________-____-07 i l l Total For i Period Ci 2.47E 05 4.03E 1 _-_________-_________-______________-04 _________________________
- Zeroes in this table indicate that no radioactivity was See Table 2-8 for ty present above detectable levels. lower limits of detection for gaseous sample analyses.pical
**There are no batch mode radioactive gaseous release pathways at Plant Hatch.
l l i (38) I i
1 l TABLE 2-3c
- E. I. HATCH NUCLEAR PLANT - SITE SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 GASEOUS EFFLUENTS - ELEVATED RELEASES
- Continuous Mode Batch Mode ** !
----------------_-------------------------------------------- 1 l
Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 l l l 1. Fission Gases 1 l Kr-85m C4 2.42E+00 1.27E+01 Kr-87 C1 7.25E+00 9.73E+00 Kr-88 C4 0.00E+00 7.13E+00 Xe-133 C1 2.65E+01 9.32E+01 Xe-135m C4 3.59E+01 3.79E+01 Xe-135 C1 7.81E+00 7.96E+00 Xe-138 C4 1.38E+02 1.30E+02 Ar-41 C1 0.00E+00 1.53E+00 Total For Period Ci 2.18E+02 3.00E+02
- 2. Iodines I-131 C$ 2.62E-04 7.57E-02 I-133 C1 1.81E-03 9.41E-02 l I 135 Ci 3.16E-03 2.15E-02 Total For Period Ci 5.24E-03 1.91E-01 l 3. Particulates ~~~~~~~~~~~~~~~~~~~~~~
! ~~~~~E~n264~~~~~ Ci ~~~b5bbE+bb~~~5~5552b7 Co-58 C 0.00E+00 1.31E-07 Co-60 C:. 2.39E-06 4.89E-07 Zn-65 C4 0.00E+00 1.97E-06 Sr-89 C1 3.40E-06 4.85E-06 Sr-90 C$ 0.00E+00 1.00E-07 Cs-137 C4 7.33E-07 1.13E-06 Ba-140 C1 1.75E-05 2.80E-04 , La-140 Ci 2.50E-05 5.16E-04 I-131 Ci 3.12E-07 1.94E-06 \ Total For Period Ci 4.94E-05 8.06E-04
- Zeroes in this table indicate that no radioactivity was i See Table 2-8 for ty (
l present ahove detectable levels. lower limits of detection for gaseous sample analyses.pical
**There are no batch mode radioactive gaseous release pathways at Plant Hatch.
(39)
i i
\
TABLE 2-4a E. I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 1 1 GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES
- Continuous Mode Batch Modc**
_____________________________________________________________ l j Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4
- 1. Fission Gases Xe-133 C4 1.82E+01 4.08E+02 Xe 135 C1 1.87E+01 9.16E+00 Total For Period Ci 3.69E+01 4.18E+02
- 2. Iodines I-131 C4 4.26E-02 3.29E-02 I-133 C4 3.98E-01 3.02E-01 I 135 C1 8.23E 01 6.44E
_________________________ _ ___________-01 ___ _____________________ Total For Period Ci 1.26E+00 9.79E ____________________________________-01 ____ _____________________
- 3. Particulates Zn-65 Ci 1.32E-04 1.19E-04 Sr-89 CL 4.82E-03 4.73E-07 Sr-90 CL 1.09E-04 0.00E+00 Nb-95 CL 7.88E-06 5.72E-06 Cs-137 CL 0.00E+00 1.73E-06 Ba-140 C4 4.41E-03 3.01E-03 La-140 C1 8.74E-03 6.36E-03 Ce-141 CL 3.05E-04 2.59E-04 I-131 CL 3.04E-03 3.07E-03 Other2 CL 2.65E 0.00E+00 .
_________________________-03 ____________________________________ l Total For Period Ci 2.42E 02 1.28E ___________-02 ________________________
- Zeroes in this table indicate that no radioactivity was See Table 2-8 for ty present above detectable levels. lower limits of detection for gaseous sample analyses.pical
**There are no batch mode radioactive gaseous release pathways at Plant Hatch.
(40 )
TABLE 2-4b E. I. HATCH NUCLEAR PLANT - UNIT 2 l SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 i GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES
- Continuous Mode Batch Mode **
Nuclides Relgased_,,,,_gnig,gua{pe{_3__gga{ge{,j__gua{}g{, 3,_gga{gg{_3 1
- 1. Fission Gases Kr-85m Ci 1.19E+01 3.19E+00 Kr-87 C1 6.71E+01 0.00E+00 Xe-133 Ci 3.06E+02 1.20E+03 Xe-135m C1 1.09E+02 5.04E+01 Xe-135 Ci 1.04E+02 5.49E+01 Xe 138 C1 1.98E+02 0.00E+00 Total For Period Ci 7.96E+02 1.31E+03
- 2. Iodines ~~~ ~ ~~~ ~~~~~~~~~~~~~~~~~~~~~~
~~~~~
~~~~ i {35 Ci 5 55E205 57355205 I-133 C: 2.22E-01 1.18E-01 I 135 C1 4.27E 2.41E 01
______ _ ___________________-01 ___________________________________ Total For Period Ci 6.74E 01 3.72E ____________________________________-01 _________________________ , f
- 3. Particulates I Zn-65 Ci 1.48E-04 1.30E-04 Sr-89 Ci 2.81E-03 2.48E-06 Sr-90 C1 3.20E-05 0.00E+00 Cs-137 Ci C1 1.79E-05 3.98E-03 1.30E-06 2.20E-03 Ba-140 1 La-140 Ci C1 2.27E-02 2.69E-04 4.43E-03 1.71E-04 i
Ce-141 1 I-131 Ci C1 2.30E-03 1.76E 03 1.93E-03 0.00E+00 i Other2 ; ___________________________________________--__________-_____ \ Total For l Period Ci 3.40E 8.87E _________________________-02 ___________-03 _________________________ l
- Zeroes in this table indicate that no radioactivity was See Table 2-8 for ty l
present above detectable levels. lower limits of detection for gaseous sample analyses.pical l l l
**There are no batch mode radioactive gaseous release pathways at Plant Hatch.
i l l t l l (41) l
l l l 1 TABLE 2-4c E. I. HATCH NUCLEAR PLANT - SITE SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 ) GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES
- Continuous Mode Batch Mode **
Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 ! _____________________________________________________________ j i
- 1. Fission Gases )
Kr-85m C$ 1.19E+01 3.19E+00 Kr-87 C4 6.71E+01 0.00E+00 i t Xe-133 C1 3.25E+02 1.61E+03 i Xe-135m C1 1.09E+02 5.04E+01 ! Xe-135 Ci 1.22E+02 6.40E+01 Xe 138 C1 1.98E+02 0.00E+00 l Total For I ~ Period Ci 8.32E+02 1.73E+03 l _____________________________________________________________ 1
- 2. Iodines I-131 C: 6.81E-02 4.64E-02 r
I-133 C: 6.21E-01 4.20E-01 I I 135 C1 1.25E+00 8.85E 01 ! Total For Period Ci 1.94E+00 1.35E+00 l
- 3. Particulates j
~ " ""
l ~~~~~{~Zbb~n Ci b~79bZO4 ~5 bObZO~~ Sr-89 Ci 7.63E-03 2.95E-06 Sr-90 C1 1.41E-04 0.00E+00 Nb-95 Ci 7.88E-06 5.72E-06 l Cs-137 C1 1.79E-05 3.03E-06 Ba-140 Ci 8.40E-03 5.21E-03 La-140 C1 3.14E-02 1.08E-02 Ce-141 C:. 5.74E-04 4.30E-04 J I-131 C:. 5.34E-03 5.00E-03 i Other2 C1 4.41E 0.00E+00 _________________________-03 ____________________________________ Total For Period Ci 5.82E 02 2.17E 02
- Zeroes in this table indicate that no radioactivity was See Table 2-8 for ty present,ahove detectable levels. lower limits of detection for gaseous sample analyses.pical
**There are no batch mode radioactive gaseous release pathways at Plant Hatch.
I l l l (42)
l l Table 2-5 , I E. I. HATCH NUCLEAR PLANT - SITE SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 I GASEOUS EFFLUENTS - DOSE RATES Dose Rates Due to Noble Gases Organ Tech Unit Quarter 3 % of Quarter 4 % of Speg Tech Tech Limit Spec Spec Limit Limit TBgdy 500 mrem /yr 3.01E+00 6.01E-01 8.49E-01 1.70E-01 Skin 1.80E+00 6.01E 3000 mrem /yr 5.50E+00 1.83E --_____---_-__-_--_-_-02
---_-_---_____--------_.----__--_---__-01 -_
Dose Rates Due to Radiciodine, Tritium, and Particulates Organ Tech Unit Quarter 3 % of Quarter 4 % of Spec Tech Tech l Limit Speg Speg Limit Limit
- Bgne 1500 mrem yr 3.52E-02 2.35E-03 1.07E-02 7.16E-04 Liver 1500 mre yr 4.17E-02 2.78E-03 3.59E-02 2.39E-03 TBody 1500 mrem yr 3.30E-02 2.20E-03 2.89E-02 1.93E-03 Thyroid 1500 mre yr 3.75E+00 2.50E-01 2.67E+00 1.78E-01 Kidney 1500 mrem yr 5.28E-02 3.52E-03 4.38E-02 2.92E-03 Lung 1500 mrem yr 6.68E-02 4.45E-03 3.67E-02 2.45E-03 GILLI 1500 mrem yr 3.85E-02 2.56E-03 2.99E-02 1.99E i
__--__-------_---_---------------__-------------______-_-_-03 --- i l i l l l l l (43) i - - _l
i TABLE 2-6a E. I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 ! AIR DOSES DUE TO NOBLE GAS RELEASES Cumulative Doses per Quarter _ _ _ _ _ , . _ l Wpe Tech Unit Quarter 3 % of Quarter 4 of or Spec Tech Tech Rad'- Limit Spec Speg ation Limit Limit \ ____________________________-________________________________ I l Gamma 5.0 mrad 1.29E-02 2.58E-01 4.47E-02 8.93E-01 I Beta 10.0 mrad 1.78E 02 1.78E 01 1.203 01 1.20E+00 i 1 Cumulative Doses This Year T pe Tech Unit Quarters % of l o , Speg 1,2,3,& 4 Tech Radi- Limit Speg ation Limit Gamma 10.0 mrad 8.08E-02 8.08E-01 Beta 20.0 mrad 1.71E 8.56E 01 ___________________________-01 __________________________________ t i l L (44)
l ! TABLE 2-6b E. I. IIATC11 NUCLEAR PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 AIR DOSES DUE TO NOBLE GAS RELEASES l Cumulative Doses per Quarter ______ l _____________________________________________________% T pe Tech Unit Quarter 1 % of Quarter 4 of o Spec Tech Tech Radi- Limit Speg SpeQ ation Limit Limit Gamma 5.0 mrad 7.77E-01 1.55E+01 1.88E-01 3.76E+00 Beta 10.0 mrad 6.13E 6.13E+00 3.82E 01 3.82E+00 ________________________-__-01 ___________-______________________ Cumulative Doses This Year T pe Tech Unit Quarters % of o Spec 1,2,3,& 4 Tech Radi- Limit Speg ation Limit Gamma 10.0 mrad 1.04E+00 1.04E+01 > Beta 20.0 mrad 1.09E+00 5.44E+00 i i 1 ( { t i f i \ (4 5) r i L_ b
1 l l TABLE 2-7a i E. I. HATCH NUCLEAR PLANT - Unit 1 l SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 l INDIVIDUAL DOSES DUE TO RADIOIODINE TRITIUM, ' I l AND PARTICULATES IN GASEOUS REL8ASES i I Cumulative Doses per Quarter l Organ Tech Unit Quarter 3 % of Quarter 4 % of Speg Tech Tech Limit Spec Spec Limit Limit Bone 7.5 mrem 9.22E-02 1.23E+00 3.77E-03 5.02E-02 Liver 7.5 mrem 7.25E-03 9.66E-02 7.38E-03 9.83E-02 TBody 7.5 mrem 1.67E-02 2.22E-01 5.99E-03 7.99E-02 Thyroid 7.5 mrem 8.84E-01 1.18E+01 7.62E-01 1.02E+01 Kidney 7.5 mrem 9.31E-03 1.24E-01 9.11E-03 1.21E-01 Lung 7.5 mrem 6.12E-03 8.16E-02 5.25E-03 6.99E-02 GILLI 7.5 urem 7.54E 03 1.01E-01 5.43E-03 7.25E 02 Cumulative Doses This Year Organ Tech Unit Quarters % of l Speg 1,2 ' & 4 Tech l Limit Spec l Limit l i Bone 15.0 mrem 1.31E-01 8.74E-01 Liver 15.0 mrem 2.22E-02 1.48E-01 TBody 15.0 mrem 3.46E-02 2.30E-01 Thyroid 15.0 mrem 2.01E+00 1.34E+01 Kidney 15.0 mrem 2.67E-02 1.78E-01 Lung 15.0 mrem 1.82E-02 1.21E-01 GILLI 15.0 mrem 2.02E 02 1.35E __ __-_______________________________-01 ________________________ (46)
l l l TABLE 2-7b E. I. HATCH NUCLEAR PLANT - Unit 2 i SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1993 l INDIVIDUAL DOSES DUE TO RADIOIODINE TRITIUM, l AND PARTICULATES IN GASEOUS REL$ASES l CEmg agive ggsgs ggg_ggag333_________________________________ Organ Tech Unit Quarter 3 % of Quarter 4 % of i l Spe Tech Tech
$mSt $mSt Bone 7.5 mrem 4.28E-02 5.71E-01 1.84E-03 2.46E-02 Liver 7.5 mrem 1.08E-02 1.44E-01 7.97E-03 1.06E-01 TBody 7.5 mrem 1.34E-02 1.78E-01 7.33E-03 9.78E-02 Thyroid 7.5 mrem 5.26E-01 7.01E+00 3.61E-01 4.82E+00 Kidney 7.5 mrem 1.19E-02 1.59E-01 8.74E-03 1.17E-01 Lung 7.5 mrem 1.05E-02 1.40E-01 7.13E-03 9.51E-02 GILLI 7.5 mrem 1.13E-02 1.50E 7.13E-03 9.50E-02
-----__-------_----_--_--_--_--_-----_-01 _-_---------_---------_
Cumulative Doses This Year Organ Tech Unit Quarters % of Spec 1,2,3,& 4 Tech I Limit Speg Limit Bone 15.0 mrem 9.28E-02 6.18E-01 Liver 15.0 mrem 2.84E-02 1.89E-01 TBody 15.0 mrem 3.43E-02 2.29E-01 Thyroid 15.0 mrem 1.29E+00 8.58E+00 Kianey 15.0 mrem 3.09E-02 2.06E-01 Lung 15.0 mrem 2.71E-02 1.80E-01 mrem 2.84E 1.89E-01
---------------_-------__-_-02 GILLI 15.0 _------__------_--_--______-_____-
l l I (47)
TABLE 2-8 LOWER LIMITS OF DETECTION - GASEOUS SAMPLE ANALYSES l The values in this table represent apriori lower limits of I detection (LLD) which are typically achieved in laboratory l analyses of gaseous radwaste samples. l LLD UNITS ! RADIONUCLIDE l Kr-87 1.31E ^i7 uCi/ml Kr-88 2.10E-07 Xe-133 1.62E-07 Xe-133m 6.07E-08 Xe-l 35 5.77E-08 Xe-138 2.85E-06 1 l 1-131 4.37E-14 I-133 6.16E-13 Mn-54 2.78E- 14 Fe-59 4.62E-14 Co-58 2.46E-- 14 Co-60 2.88E-14 Zn-65 7.51 E- 14 Mo-99 6.02E-13 Cs 134 3.64 E- 14 Cs-137 2.88 E- 14 Cs-141 4.94 E- 14 Cs-144 2.02E-13 ,, Sr-89 1.00E-I l ; 1.00E-11 : Sr-90 H-3 1.00E-06 (48)
- 3. SOLID WASTE 3.1 REGULATORY REQUIREMENTS l The Technical Specifications presented in this section are for Unit 1.
l Requirements for Unit 2 are the same as for Unit 1; however, the Technical Specification numbers are not the same. TECHNICAL SPECIFICATIONS 3.15.3.1 The solid radwaste system shall be used in accordance with the PROCESS CONTROL PROGRAM to provide for the SOLIDIFICATION of wet solid wastes and for the SOLIDIFICATION and packaging of other radioactive wastes, as required, to ensure the meeting of the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site. 6.9.1.9 states in part: The Radioactive Effluent Release Report shall include the following information for each type of solid waste shipped offsite during the report period:
- a. Container volume
- b. Total curie quantity (specify whether determined by measurement or estimate)
- c. Principal radionuclides (specify whether determined by measurement or estimate)
- d. Type of waste, e g., spent resin, compacted dry waste, evaporator bottoms e Type of container, e g.. LS A, type A, type B, large quantity
- f. Solidification agent, e.g., cement.
32 SOLID WASTE DATA Regulatory guide 1.21 Table 3 is found in this report as Table 3-la. and 3-lb. (49)
EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1993 JULY 1.1993 - DECEMBER 31,1993 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS FOR UNITS I AND II ' TABLE 3-1A FORM TITLE: REG GUIDr.1.21 ErrLUENT AND HASTE DISPOSAL srNTunst11!. REPorr or ont f n u1ETE AltD f eenf Aven ruter REf PWnrTS I / I / 93 TO I2 / 31 /93 - FOR UNIT 142 PERIOD COVERED: FRON A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel) i
- UNIT 6 month Est. Total
- 1. Type of waste runneg
_oeriod
, ,3 3 6.85e+0
- a. Spent resins, filter sludges, evaporator c1 1.97 e0: 1.0 a+01 bott me. see.
_a 3 4.07a&Of
- b. Dry compressible waste, contaminated 2 .0 e01 c1 2 .2 B a&O(
equin. etc.
- c. Irradiated components, control rods, , ,_a 3 1.50 w01 c1 1.07a&O! 2 .0 F01 3 . E
- d. Control Rod Drive Filters _3 ci . E . E 3 . E
- e. Other (describe) _3 cl . I E acuin. etc.
- 2. Estimate of major nuclide composition (by type of waste)
Prat --i (MInfr_E IknurE 6.82E+01 1.34E+03 l
- a. 2n-65 Eg. .jj 7.11E+00 1.40E+02 Co-60 1.35EMI 2.66E+02 Cc-137 1.11E+00 2.18E+01 1.99E+02 4
1.01E@ l Or ne.r 5.59E+01 1.2BE+00
- b. 2r-65 Fe-55 1.57E+01 3.59E-01 Co-60 1.61E+01 3.75E-01 Cn-137 2.39E+00 5.46-02
{ Orber 9.61E+00 2.20E-01
- c. 2n-65 0.00E+00 0.00E+00 Fe-% 5.5hE+01 5.59E+04 Co-60 3.66E+01 3.90E+04 C2-137 f 36E-03 1.45E+00 Other 7.54E+00 a.04E+03 d.
e.
- 3. Solid Waste Disposition Humber of Shiments ligde of Transportation Destination 22 Tractor Trailer Barnwell B. IRRADIATED FUEL SHIPMENTS (Disposition)
Hm.htt 21 Shipments ligde of Transoortatigm Dmitination
!;one None None (50)
l l EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1993 JULY 1,1993 - DECEMBER 31,1993 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS FOR UNITS I AND 11 l TABLE 3-IB FORM TITLE: UNIT 1 AND 2 TECH SFEC 6.9.1.9 ErrLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT OF SOLID WASTE AND famanIATED FUEL SHIPMENTS PERIOD COVERED FRO ( 7/ 1 / 93 MI I2 /31 /93 FOR UNIT: 1&2 rYPE Ctut Pmxru stuA. Ntwam v0auE Tm xumn. OF QUANTTTY NUCtJDES/ CONTAJNEJt CONT, CONT. 3HIPWINT CATION WASTE DETTRMNE DET N T DESCRIP SHIPPED CL1NC fT CONT. AGENT DEWATERED ISt7 Za65.Cdo HIGH 2 132 4 TYPEB RESNS WIA$t1 TID Cst 37 INTEGitJTY CASK N/A FsSS CONTAJNER 10 142 WEAStU.D DEWATT. RID 383 In65 Ca60 HIGH 10 202 1 TYPE A RESNS WEASUltED Cal 37 (NTEGRJTY CASK N< A Fs55 CONTAINER I 132 4 l & 210 MEAstMD DAW 2 28 Za65 Ca60 STRONO IS 95 FTRONO (DRY) EST1 MATED Cal 31 TIGHT TIGHT N/A (ACTIVE) FsSS CONT BOXES (WASTE) MEAst1ED & 25 & 25 lidLADLATED 101000 Cedo. Fe SS STIEL 5 57 4 TYPE 8 NA HARDWARE ESTlWA1T.D Cel31 HIGH SHIPPING INTEGRffY 2 121 CASK CONTANER 8-l20 AP101 (51)
l 4 CHANGES TO THE PLANT HATCH ODCM & PCP Technical Specifications 6.9.1.8 and 6.9.1.9 require in part that changes to the Plant Hatch Offsite Dose Calculation Manual (ODCM) be reported to the Commission in the l next Semi-annual Efiluent Release Report.
)
A revision to the Plant Hatch Offsite Dose Calculation Manual was made to I incorporate compliance to NRC Generic Letter 89-01 and to incorporate the requirements of the new 10CFR20. Approval of Revision 8 immediately followed approval of Revision l 7, and Revision 7 was, therefore, never distributed. Revision 8 represents a complete . replacement of the Plant Hatch Offsite Dose Calculation Manual. i ~ Appendix A of this report contains the complete replacement of the Plant Hatch j Offsite Dose Calculation Manual. This replacement is identified as Revision 8 of the Plant Hatch Offsite Dose Calculation Manual. i i L l i 1 (52)
l l l l 5 METEOROLOGY In accordance with Technical Specification 6.9.1.9, the annual summary of meteorological data collected at Plant Hatch over 1993 is presented in this section. 51 1993 Meteorological Data Attachment i Joint Frequency Tables of Wind Speed and Wind Direction 10m vs Delta Temperature 60-10m. Attachment 2 Joint Frequency Tables of Wind Speed and Wind Direction 60m vs Delta Temperature 60-10m. Attachment 3 Joint Frequency Tables of Wind Speed and Wind Direction 100m vs Delta Temperature 100-10m. Attachment 4 Wind Roses from 10m (Seasonal and Annual). Attachment 5 Wind Roses from 60m (Seasonal and Annual). t l Attachment 6 Wind Roses from 100m (Seasonal and Annual). Attachment 7 Wind Roses from 23m Backup (Seasonal and Annual). Attachment 8 Percent Data Recovery by Parameter and for Pertinent l Composite Parameters for 1993. Attachment 9 Plots of Monthly Averages and Averages of Daily Extremes of Ambient Temperature and Dew Point Temperature. Attachment 10 Daily, Monthly, and Annual Precipitation for the Period of January through December 1993. Attachment 11 Plant Hatch 1993 Meteorological Summary. l l l l (53)
PLANT HATCH JOINT FREQUENCY TABLES OF WIND SPEED 1 of 8 ATTACHMENT 1. AND WIND DIRECTION 10m VS DELTA TEMPERATURE 60-10m JANUARY 1,1993 THROUGH DECEMBER 31, 1993 l l HOURS AT EACH UIND SPEED AND DIRECTION PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: A DT/DZ ELEVATION: SPEED:SPD10M DIRECTION:DIR10M LAPSE:DT60-UIND SPEED (MPH) WItiD DIRECTIOtt 1-3 4-7 8-12 13-18 19-24 >24 TOTAL 1 34 3 0 0 0 38 N 4 2S S 0 0 0 34 NNE 8 93 9 0 0 0 110 NE 4 74 3 0 0 0 81 , Erie E 3 52 0 0 0 0 55 3 42 2 0 0 0 47 ESE SE 8 31 4 0 0 0 43 SSE 9 46 3 0 0 0 58 S 7 47 12 0 0 0 66 SSU 8 S3 2 0 0 0 63 SU 9 84 11 0 0 0 104 USU 8 80 9 0 0 0 97 U 7 69 20 0 0 0 96 UNU 6 65 16 0 0 0 87 tiu 4 72 12 0 0 0 88 litlU 0 36 3 0 0 0 39 TOTAL 89 903 114 0 0 0 1106 PERIODS OF CALM (HOURS): 0 UARIABLE DIRECTION 398 HOURS OF MISSING DATA: 65
ATTACHMENT 1 (continued) 2 of 8 HOURS AT EACH UIND SPEED AND DIRECTI0ti PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: B DT/DZ ELEVATION: SPEED:SPD10M DIRECTION:DIR10M LAPSE:DT60-UIND SPEED (MPH) WIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL li 3 13 2 0 0 0 18 NNE 1 13 0 0 0 0 14 NE 5 39 1 0 0 0 45 ENE 5 37 0 0 0 0 42 E 3 15 1 0 0 0 19 ESE 3 14 0 0 0 0 17 SE 2 16 0 0 0 0 18 SSE 4 15 0 1 0 0 20 S 4 17 3 0 0 0 24 SSU 11 9 1 0 0 0 21 SU 22 22 0 0 0 0 44 USU 9 23 0 1 0 0 33 U 8 22 7 0 0 0 37 UNU 8 21 6 0 0 0 35 NU 11 17 3 0 0 0 31 IINU 5 18 0 0 0 0 23 TOTAL 104 311 24 2 0 0 441 PERIODS OF CALM (HOURS): 0 UARIABLE DIRECTION 279 HOURS OF MISSING DATA: 65
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ATTACHMENT 1 (continued) 4 of 8 HOURS AT EACH UIND SPEED Arid DIRECTION PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: D DT/DZ ELEVATION: SPEED:SPD10M DIRECTION:DIR10M LAPSE:DT60-UIND SPEED (MPH) WIflD DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL 30 70 16 0 0 0 116 tl tlNE 27 53 2 0 0 0 82 f1E 63 227 23 0 0 0 313 ENE 82 141 0 0 0 0 223 E 46 50 2 0 0 0 98 ESE 38 37 2 0 0 0 77 SE 30 44 1 0 0 0 75 SSE 37 43 6 1 0 0 87 S 47 42 8 0 0 0 97 SSU 49 53 4 0 0 0 106 SU 54 62 3 1 0 0 120 . USU 68 47 5 2 0 0 122 U 48 74 6 6 0 0 134 UNU 32 77 17 1 0 0 127 NU 45 57 9 0 0 0 111 liNU 37 48 4 0 0 0 89 TOTAL 733 1125 108 11 0 0 1977 PERIODS OF CALM (HOURS): 0 UARIABLE DIRECTION 1246 HOURS OF MISSING DATA: 65
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ATTACHMENT 2. PLANT llATCil JOINT FRE0VENCY TABLES OF WIND SPEED 1 of 8 AND WIND DIRECTION 60m VS DELTA TEMPERATURE 60-10m JANUARY 1, 1993 TilROUGH OECEMBER 31, 1993 HOURS AT EACH UItiD SPEED AtiD DIRECTI0ti PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: A DT/DZ ELEVATI0ti: SPEED:SPD60M DIRECTI0ti:DIR60M LAPSE:DT60-UIriD SPEED (MPH) UItiD DIRECTIOrt 1-3 4-7 8-12 13-18 19-24 >24 TOTAL 12 19 1 0 0 33 ti 1 titiE 2 12 20 8 0 0 42 2 33 65 17 2 0 119 tie EtiE 1 25 50 4 0 0 80 E O 17 35 8 0 0 60 ESE 1 25 13 4 0 0 43 13 23 4 0 0 41 SE - 1 SSE 2 16 15 2 0 0 35 S 2 15 28 9 0 0 54 SSlJ 1 18 23 7 0 0 49 SU 0 18 44 22 8 0 92 USU 1 16 42 33 5 0 97 U 1 31 37 32 2 0 103 LJiiU 2 30 30 30 1 0 93 tiU 0 23 27 5 0 0 55 IlilU 1 13 19 0 0 0 33 TOTAL 18 317 490 186 18 0 1029 j PERIODS OF CALM (HOURS): 0 VARIABLE DIRECTI0ti 82
- HOURS OF MISSItiG DATA
- 1020
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ATTACHMENT 2 (continued) 4 of 8 HOURS AT EACH UIND SPEED AND DIRECTIOti PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: D DT/DZ ELEVATION: SPEED:SPD60M DIRECTION:DIR60M LAPSE:DT60-UIND SPEED (MPH) UIND DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL i , N 7 22 31 13 1 0 74 NNE 5 27 46 30 3 0 111 tie 9 66 177 71 2 0 325 EtiE 14 59 91 11 0 0 175 E 15 28 38 9 1 0 91 ESE 5 16 31 11 0 0 63 SE 12 21 34 3 0 0 70 SSE 7 20 24 15 0 0 66 S 7 33 14 15 0 0 69 SSU 8 35 23 4 1 0 71 SlJ 13 36 30 8 2 1 90 USU 14 45 46 14 5 2 126 U 10 32 48 32 8 9 139 Utiu 5 29 32 20 5 0 91 NU 7 36 36 7 1 0 87 NiiU 6 56 36 8 0 0 106 , TOTAL 144 561 737 271 29 12 1754 PERIODS OF CALM (HOURS): 0 , VARIABLE DIRECTIOri 237 : HOURS OF MISSING DATA: 1020
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8_290486560031 1 _ 6932222321 _ 4 _ 5 N E2 N _ 1 1 4 1 _ E P1 O _ H - _ _ 2 _ M _ I _ 3 _ H C A S3 D1 2 T_P C _ ( M 1 _ _ . T N3 E_ D 2 _ 56001 52476927843 _ 9 T A I9 E 1 _ 339661 9691 270854 _ 5 _ 2 U ZIR __ E - _ 1 1 1 1 _ 1 0 1 DD _ P 8 _ _ 0 _ H0/ _ S C1 T _ _ 2 - A0DM_D 7- __ 3849969277654970 _ 5 0 0 _ 233454344431 21 _ 4 1 6_N 1 1 . _ E1 _ 5 8 _ 0 I 4 _ _ T3 D _ U _ 6 _ A9 P _ _ A ES _ R T S : _ 3 _ 20933O0961 5691 42 _ 0 _ 8 b A R D _ - _ 1 1 U E _ 1 _ _ 0ND O= E _ _ HO H P _ _ hIG D: S _ _ 1 tTN RS _ _ LCI OS _ _ AES CA _ _ CRS _ _ II EL _ _ DM RC 0 EF FYf _ t N _ _ hLO OT0 _ O _ II _ DI _ _ DB DLT_ NT_ _ L OAS OIA _ IC _ _ A IR IBV_ WE _ E E E E U U U U_T PRU RAE _ R _ l El i SES SUS iUl _ O l i EAO PUH t ETL _ I _ f t i lf EEESSSSSUUUil l _ T t I PSE _ D _
l ATTACHMEtiT 2 (continued) 6 of 8 HOURS AT EACH UItiD SPEED AtiD DIRECTION PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: F DT/DZ ELEVATI0ti: SPEED:SPD60M DIRECTION:DIR60M LAPSE:DT60-UIND SPEED (MPH) WIND DIRECTI0t1 1-3 4-7 8-12 13-18 19-24 >24 TOTAL ti 7 7 11 9 0 0 34 IINE 1 8 8 G 0 0 23 tie 1 5 35 15 0 0 56 ENE 2 10 38 4 0 0 54 E 1 20 48 7 0 0 76 ESE 3 21 42 5 1 0 72 SE 9 45 23. 0 0 0 77 SSE 5 30 56 4 0 0 95 S 3 32 63 2 0 0 100 SSU 4 18 61 6 1 0 90 SU 3 19 96 9 0 0 127 USU 4 13 77 14 0 0 108 U 1 9 64 4 0 0 78 UtlU 4 7 21 3 0 0 35 NW 1 9 25 1 0 0 36 titlU 4 2 12 5 0 0 23 TOTAL 53 255 680 94 2 0 1884 PERIODS OF CALM (HOURS): 0 VARIABLE DIRECTION 33 HOURS OF MISSING DATA: 1020 4
ATTACHMENT 2 (continued) 7 of 8 HOURS AT EACH UIND SPEED AND DIRECTION PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: G DT/DZ ELEVATION: SPEED:SPD60M DIRECTION:DIR60M LAPSE:DT60-UIND SPEED (MPH) UItiD DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL (1 5 11 12 2 0 0 30 flNE 6 18 5 1 0 0 30 llE 7 21 35 4 0 0 67 ENE 2 24 33 1 0 0 60 E 4 19 27 1 0 0 51 ESE 5 19 23 0 0 0 47 SE 5 29 13 0 0 0 47 SSE 5 25 27 0 0 0 57 S 6 20 37 3 0 0 66 SSU 4 25 37 6 0 0 72 SU 7 26 48 13 0 0 94 USU S 22 45 8 0 0 80 W 3 13 66 5 0 0 87 UNU 4 13 31 2 0 0 50 NU 2 9 2? 1 0 0 39 NNU 3 9 16 3 0 0 31 TOTAL 73 303 482' 50 0 0 908 PERIODS OF CALM (HOURS): 0 I UARIABLE DIRECTION 36 HOURS OF MISSING DATA: 1020
8 of 8 ATTACllMENT 2 (continued) i l HOURS AT EACH LJItiD SPEED AND DIRECTION PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: ALL DT/DZ ELEVATIOti: SPEED:SPD60M DIRECTION:DIR60M LAPSE:DT60-UIND SPEED (MPH) UItiD DIRECTION 1-3 4-7 8-12 13-18 19-24 >24 TOTAL tl 23 82 116 39 1 0 261 titiE 28 97 124 G9 6 0 324 ilE 30 185 439 160 6 0 820 EtiE 27 196 301 25 0 0 549 E 24 145 221 37 1 1 429 ESE 15 143 235 28 2 0 423 SE 40 182 201 12 0 0 435 SSE 28 147 197 41 1 1 415 S 26 157 252 50 1 2 488 SSU 22 164 271 45 3 0 505 SlJ 32 180 363 83 11 1 670 LJSlJ 31 169 305 110 12 3 630 (J 33 158 345 120 19 9 684 LJtild 21 136 222 81 7 0 467 illJ 16 127 185 19 1 0 348 liillJ 18 111 144 18 1 0 292 TOTAL 414 2379 3921 937 72 17 7740 i PERIODS OF CALM (HOURS): 3 VARIABLE DIRECTION 633 HOURS OF MISSING DATA: 1020
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ATTACHMENT 3 (continued) 4 of 8 HOURS AT EACH UItiD SPEED AtID DIRECTIOil PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: D DT.'DZ ELEVATI0ti: SPEED:SP100M DIRECTI0ti:DI100M LAPSE:DT100-UIriD SPEED (MPH) UltiD DIRECTIOli 1-3 4-7 8-12 13-18 19-24 >24 TOTAL II 7 19 43 29 2 0 100 llHE 5 23 34 40 19 0 121 flE 7 50 80 71 3 0 211 EllE 5 56 67 23 0 0 151 E 9 26 32 32 3 0 102 ESE 7 23 33 20 1 0 84 SE 11 36 29 7 0 0 83 SSE 11 26 27 19 8 1 92 S 7 32 27 13 3 1 83 SSU 7 45 33 14 1 0 100 SU 14 60 44 27 5 4 154 USU 12 68 56 28 17 7 188 U 10 58 49 60 35 17 229 Utiu 8 65 54 34 10 0 171 flu 8 41 54 17 2 0 122 lillu 8 45 45 15 2 0 115 TOTAL 136 673 707 449 111 30 2106 PERIODS OF CALM (HOURS): 0 VARIABLE DIRECTI0ft 285 HOURS OF MISSItiG DATA: 1669
ATTACHMENT 3 (continued) 5 of 8 HOURS AT EACH UIND SPEED AtiD DIRECTI0tt PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: E DT/DZ ELEVATION: SPEED:SP100M DIRECTION:DI100M LAPSE:DT100-UIND SPEED (MPH) WIrlD DIRECTIOtt 1-3 4-7 8-12 13-18 19-24 >24 TOTAL tl 4 9 20 32 5 6 70 flNE 6 12 37 41 17 0 113 tie 6 7 59 86 9 0 167 EllE 2 18 61 52 2 0 135 E 4 18 59 69 0 1 151 ESE 6 31 87 67 5 B 196 SE 9 50 98 28 0 0 185 SSE 4 12 53 76 9 3 157 5 3 21 G5 93 8 1 191 SSU 2 26 71 90 11 1 201 i SU 4 21 92 100 20 1 238
- USU 6 16 45 67 22 1 157 U 5 23 34 93 17 3 175 UNU 4 6 32 74 12 0 128 NU 3 11 31 44 1 0 90 i
tlNU 3 13 23 27 2 0 68 TOTAL 71 294 867 1039 140 11 2422 , PERIODS OF CALM (HOURS): 0 j UARIABLE DIRECTION 58 4 HOURS OF MISSING DATA: 1669 i i
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i ATTACllMENT 3 (continued) 7 of 8 HOURS AT EACH UItiD SPEED AtiD DIRECTION PERIOD OF RECORD = 93010101-93123124 STABILITY CLASS: G DT/DZ ELEVATI0ti: SPEED:SP100M DIRECTI0ti:DI100M LAPSE:DT100- < UIIID SPEED (MPH) WIllD DIRECTI0ti 1-3 4-7 8-12 13-1;3 19-24 >24 TOTAL , NtiE 3 11 8 3 0 0 25 tie 2 5 22 13 0 0 42 EtiE 1 7 13 20 1 0 42 E 1 6 23 12 3 0 45 ESE 4 5 9 4 0 0 22 SE 4 17 12 7 1 0 41 SSE 2 1 15 16 5 0 39 S 2 6 13 14 1 0 36 SSU 3 7 11 13 5 2 44 SU 3 17 23 24 7 0 74 USU 4 14 21 17 1 0 57 U 3 9 17 22 5 0 56 UtiU 1 4 14 19 3 0 41 IlU 2 7 11 20 1 0 41 tillU 2 4 10 11 2 0 29 35 0 TOTAL 40 127 232 225 659 PERIOOS OF CALM (HOURS): 0 UARIABLE DIRECTIOil 12 HOURS OF MISSItiG DATA: 1669
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ATTACllMINT 7a. PLANT llATCil 23m 13ACKUP METEOROLOGICAL TOWER WIND ROSE JANUARY 1,1993 THROUGil DECIMBER 31, 1993
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ATTACllMENT 7b. PLANT HATCH 23m BACKUP METEOROLOGICAL TOWER WIND ROSE JANUARY l,1993 THROUGH MARCH 31, 1993
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ATTACHMENT 7e. PLANT HATCH 23m BACKUP METEOROLOGICAL TOWER WIND ROSE OCTOBER 1, 1993 THROUGH DECEMBER 31, 1993
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i ! ATTACHMENT 8. PLANT HATCH HETEOR0 LOGICAL TOWER 1993 DATA RECOVERY (PRIMARY AND BACKUP TOWERS) 1
- I f Parameter Recovery J (Percent) l Wind Speed 10m 99.3 Wind Speed 60m 92.6 Wind Speed 100m 93.8
- Wind Speed 23m (Backup) 94.8
- Wind Direction 10m 99.6
- Wind Direction 60m 89.2
! Wind Direction 100m 81.3
- Wind Direction 23m (Backup) 96.5 Delta Temperature 60-10m 99.3 Delta Temperature 100-10m .
94.1 Delta Temperature 45-10m (Backup) 95.9 Temperature 10m 99.5
- Dew Point Temperature 10m 94.1
- Temperature 23m 97.1 Rainfall 98.6 l Composite
- l Wind Speed and Direction 10m, Delta Temperature 60-10m 99.3
- Wind Speed and Direction 60m, Delta Temperature 60-10m 88.4 Wind Speed and Direction 100m, Delta Temperature 100-10m 81.0 Wind Speed and Direction 23m, Delta Temperature 45-10m 93.5 4
i 4 0740LO20794
i l l ATTACHMENT 9. PLANT HATCH HONTHLY AVERAGE AND AVERAGE l I 0F THE DAILY EXTREMES OF AMBIENT TEMPERATURE 10m JANUARY l, 1993 THROUGH DECEMBER 31, 1993 \ l I l 1 j l l 1 \ l :. 2 _- . _y 3. _._q_ ; q. , . j .._. 3 . . _ . : _ . -
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l l 1 ATTACHMENT 10. DAILY, MONTHLY AND ANNUAL PRECIPITATION TOTALS l January 1, 1993 through Decemocc 31, 1993 Precipitation Precipitation Month (Inches) Month (Inches) January March 4 0.35 2 0.03 5 0.01 3 1.44 6 0.57 4 0.11 7 1.56 12 0.15 8 2.24 13 1.03 9 0.15 17 0.17 10 0.01 18 0.01 11 0.25 23 0.77 12 1.73 24 0.21 15 0.11 25 0.17 16 0.13 26 0.90 19 0.01 27 0.82 20 0.43 31 DE 21 0.32 6.75 22 0.05 24 Q2 April 8.12 4 0.03 ! 5 1.06 Februarv 9 0.14 5 0.29 26 0.51 6 0.04 27 DE ! 7 1.12 1.81 8 0.14 10 0.05 tiay 11 0.04 4 0.72 16 0.47 5 0.19 18 0.05 13 0.81
- 20 0.14 14 0.01 l
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I ATTACHMENT 10 (continued) Precipitation Precipitation Month (Inches) Month (Inches) ALOR 1 0.04 September 10 0.98 4 0.10 11 0.94 5 0.04 12 0.14 6 0.02 22 0.48 7 1.11 23 0.09 9 0.17 24 2,20 10 0.04 25 0.14 14 0.01 26 0.23 20 0.38 27 0.14 21 1.37 28 1.22 22 0.04 29 0.04 23 0.49 30 Q11B 24 0.11 6.82 26 0.01 27 0.33 July u 28 Qm.02 1 0.15 4.29 8 0.80 , 9 0.01 October ' 16 0.10 6 0.36 17 0.05 7 0.55 , 29 Emil 8 0.41 l 1.68 9 0.45 i 10 0.09 Avaust 11 0.36 2 0.52 12 2.03 3 0.88 16 0.41 4 0.60 23 0.02 5 0.31 2d 0.02 6 0.01 26 0.08 9 0.17 26 0.95 14 0.21 30 1.27 18 0.16 31 Q102 l 19 0.32 ' 7.07 ! 20 0.06 l 23 0.29 24 0.07 28 1.02 29 0.38 30 0d9 5.69 82525020393/2
ATTACHMENT 10 (continued) l l Precipitation Month (Inches) ; l November 5 1.35 6 1.52 7 0.13 8 0.01 9 0.62 j 10 0.20 27 1.50 l 28 QE 5.34
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d ATTACHMENT 11. PLANT HATCH 1993 METEOROLOGICAL
SUMMARY
4 The data collection on the Plant Hatch meteorological tower was about average in 1993 which reversed a trend of steady data improvement over the last few years. With the exception of a two day period in April, all of the data were collected on the site's DRDT computer. The strip charts were used only for that two day period. The main problem with the instrumentation was with the 100m wind ' direction that was out-of-service for various periods in January, March, June and November. Similar problems affected the 60m wind direction in October. The overall data collection was down somewhat because of a problem with an interface board between the primary meteorological tower and the strip charts and DRDT computer. On at least ten occasions, a problem with this board caused the data collection from the primary tower to be bad for periods from a few hours to about four days. During these time periods, data from the backup tower were used to replace the 10m wind speed and direction, the delta temperature 60-10m and the ambient temperature at 10m. The wind speed and delta temperature we adjusted to account for the different instrument levels on the backup tower. The table below summarizes the data collection over the past five years. The 1993 data recovery for all parameters averaged better than 95%. 1989 1990 1991 1992 1993 5-Yr Wind Speed 10m 96.7 98.6 99.3 99.7 99.3 98.7 Wind Speed 60m 83.9 98.6 89.8 99.7 92.6 92.9 Wind Speed 100m 92.8 95.2 99.6 99.4 93.8 96.2 Wind Direction 10m 98.4 98.7 99.4 99.7 99.6 99.2 Wind Direction 60m 83.8 97.0 95.9 99.7 89.2 93.1 Wind Direction 100m 97.6 78.4 94.4 99.7 81.3 90.3 Delta Temperature 60-10m 93.2 98.2 99.4 99.7 99.3 98.0 Del ta Temperature 100-10m 92.2 98.3 99.3 98.2 94.1 96.4 Temperature 10m 98.2 96.4 99.3 99.7 99.5 98.6 Dew Point 10m 91.2 98.6 99.5 98.3 94.1 96.3 Precipitation 91.6 98.7 99.4 99.8 98.6 97.6 CILmgo11tg WS, HD 10m, DT 60-10 90.0 98.2 99.2 99.7 99.3 97.3 WS, WD 60m, DT 60-10 77.8 96.6 89.2 99.6 88.4 90.3 WS, WD 100m, DT 100-10 85.6 73.1 94.4 98.2 81.0 86.5 Attachments 1, 2 and 3 show joint frequency tables for the 10, 60 and 100m levels on the primary tower. The table below summarizes the last five years of joint frequency distributions using the wind speed and 0740LO20894
ATTACHMENT 11 (continued) direction at 10m versus the delta temperature 60-10m from the primary tower. The 1993 data shows a more normal distribution of stability classes when compared to the five year average than in r'ecent years. There were more stable hours than in recent years which can be attributed to the hot, dry summer with many clear nights. Plant Hatch Stability Classification Stability Percent Stability E-Year Group 1989 1990 1991 1992 1993 Average A 11.8 16.3 14.3 17.8 12.7 14.6 B 4.0 4.7 5.1 4.3 5.1 4.7 C 3.3 3.9 4.6 4.1 5.2 4.2 0 22.0 19.1 24.7 23.6 22.7 22.4 E 33.1 29.3 31.0 28.7 28.3 30.1 F 12.8 12.8 9.8 11.7 14.0 12.2 G 13.0 13.9 10.5 9.8 12.0 11.8 Total Hours 7886 8604 8692 8753 8695 l The wind rose data in Attachments 4 through 7 shows that the primary wind , direction is from the southwest with a close secondary peak from the ' northeast. There is good agreement between all levels on the primary tower with some minor differences with the backup tower. The backup tower has a peak wind direction of winds from the west and a secondary peak from the east-northeast. On a seasonal basis, the northeast winds occur primarily in the winter and the southwest winds are predominant in the summer months. ; Both the 10m ambient and dew point temperatures shown in Attachment 9 were about average in terms of long-term normals. In recent years, conditions, particularly in the summer, have been cooler than normal. However, in 1993 the summer months of July and August were quite hot and somewhat dry, particularly when compared with recent years. The annual rainfall total in Attachment 10 for 1993 was 56.35 inches which was the highest total recorded in recent years. It was a year with changeable totals including six months of more than 5 inches of rain and dry periods such as July with only about 1.5 inches. The 56 inch total is well above the normal of between 45 and 50 inches. 0740LO20894 k -- - - om -e-,. 5, w--.m
l Georgia Power Company Plant E. I. Hatch 1 Semi-annual Report Plant Radioactive Efiluent Releases July I through December 31,1993 ! Appendix A l i l l l l I
i f l l i l l l i 1 1 l l i OFFSITE DOSE CALCULATION MANUAL FOR l GEORGIA POWER COMPANY EDWIN I. HATCH NUCLEAR PLANT l 4 l Revision 8 January 1, 1994
Hatch ODCM DISTPIBUTION LIST Location
~
Coov No. Pecioient Plant Hatch - Simulator Bldg 1 Reddick, R. G. Baxley, Georgia Plant Hatch - Service Bldg 2 Sorrell, E. C. Document Control Baxley, Georgia Plant Hatch - Service Bldg 3 Lewis, J. C. Baxley, Georgia Plant Hatch - Service Bldg 4 Arnold, Brian Baxley, Georgia Plant Fatch - Service Bldg 5 Smith, Dorsey Baxley, Georgia Plant Hatch - Service Bldg 6 Bennett, Deryle Baxley, Georgia Plant Hatch - Simulator Bldg 7 Tipps, S. B. Baxley, Georgia , Plant Hatch - Service Bldg 8 Sorrell, E. C. Document Control Baxley, Georgia Southern Nuclear Operating Company 11 , Hopper, D. M. Inverness Center - Bldg 40, 6th Floor l Birmingham, Alabama Nichols, Georgia Power Company 14 M. C. Environmental Center l 5131 Maner Road l l Smyrna, Georoia 1 l Southern Company Services 16 Wehrenberg, J. A. Inverness Center, Bldg 42 Birroingham, Alabama S. Southern Company Services 18 l Hempstead, J. l Inverness Center, Bldg 42 Birmingham, Alabama l Georgia Power Company 21 Nichols, M. C. Environmental Center i b131 Maner Road l Smyrna, Georgia Southern Nuclear Operating Company 22 Robson, G. W. Inverness Center,. Bldg 40 Birmingham, Alabama i Rev. 8, 1/94 i--_---____.-_____ .A. , - , ,
I Match ODCM TABLE OF CONTENTS PAGE l i I DISTRIBUTION LIST . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 I I LIST OF TABLES ............................. v I l
............................. vil i l
LIST OF FIGURES I REFERENCES ............................... viii 1-1 CHAPTER 1: INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . ! i 2-1 CRAPTER 2: LIQUID EFFLUENTS . . . . . . . . . . . . . . . . . . . . . . 2-1 I l 2.1 LIMITS OF OPERATION 2-1 2.1.1 Licuid Ef fluent Monitorino Instrumentation Control 2-7 i ' 2.1.2 Licuid Ef fluent Concentration Control 2-10 i 2.1.3 Licuid Effluent Dose control Licuid Radwaste Treatment System Control 2-12 2.1.4 Systems 2-13 2.1.5 Maior Chances to Licuid Radioactive Waste Treatment l 2-14 2.2 LIQUID RADWASTE TREATMENT SYSTEM l 2 .7 2.3 LIQUID EFFLUENT MONITOR SETPOINTS 2-17 2.3.1 General Provisions Recardino Setpoints 2-19 2.3.2 Setooints for Radwaste Svetem Discharae Monitors 2.3.3 Setooints for Monitors on Normally Low-Radioactivity Streams 2-27 2-28 2.4 LIQUID EFFLUENT DOSE CALCULATIONS 2-28 2.4.1 Calculation of Dose 2-29 2.4.2 Calculation of A g 2-30 2.4.3 Calculation of CF 3 2-40 2.5 LIQUID EFFLUENT DOSE PROJECTIONS 2-40 2.5.1 Thirty-One Day Dose Proiections 2-40 2.5.2 pose Proiections for Specific Releases 2-41 l 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS
. . . . . . . . . . . . . . . . . . . . 3-1 CHAPTER 3: GASEOUS EFFLUENTS 3-1 l
3.1 LIMITS OF OPERATION 3-1 3.1.1 Gaseous Effluent Monitorina Instrumentation Control 3-7 l 3.1.2 Gaeeous Effluent Dose Rate Control l 3-11 3.1.3 Gasecus Effluent Air Dose Control 3-13 3.1.4 Control en Gaseous Effluent Dose to a Member of the Public ii Rev. 8, 1/94
Hatch OQQ3 TABLE OF CONTENTS (COntinuedi PAGE 3-15 3.1.5 Gaseous Radwaste Treatment System control Systems 3-16 3.1.6 Maior Chanaes to Gaseous Radioactive Waste Treatment f 3-17 3.2 GASEOUS RADWASTE TREATMENT SYSTEM l 3-19 3.3 CASEOUS EFFLUENT MONITOR SETPOINTS 3.-19 3.3.1 General Provisions Rocardina Noble Gas Monitor Setooints ' 3.3.2 Setooint for the Final Noble Gas Monitor on Each Release l 3-21 F.ith. MAX Source Streams 3-26 , 3.3.3 Setooints for Noble Gas Monitors on Ef fluent I 3-29 3.3.4 Determination of Allocation Factors. AG 3-31 3.3.5 Setooints for hoble Gas Monitors with soecial Reauirements 3-32 3.3.6 Setoolnts for Particulate and Iodine Monitors 3-33 3.4 GASEOUS EFFLUENT COMPLIANCE CALCULATIONS 3-33 l 3.4.1 Dose Rates at and Beyond the site Boundary 3-35 3.4.2 Noble Gas Air Dose at or Beyond Site Boundary ( J-39 I 3.4.3 Dose to a Member of the Public at or Beyond Site Boundary 3-42 3.4.4 Dose Calculations to Sucoort Other Requirements 3-47 3.5 GASEOUS EFFLUENT DOSE PROJECTIONS Thirtv-One Day Dose Proiections 3-47 3.5.1 Dose Proiections for Specific Releases 3-48 3.5.2 3-49 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS
. . . . . . . 4-1 CHAPTER 4: RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4-1 4.1 LIMITS OF OPERATION 4-1 4.1.1 Radiolocical Environmental Monitorinc 4-9 4.1.2 Land Use Census Interlaboratory Comparison Procram 4-11 4.1.3 4-12 j 4.2 RADIOLOGICAL KNVIRONMENTAL MONITORING LOCATIONS l
. . . . . . . . . . . . . 5-1 CHAPTER 5: TOTAL DOSE DETERMINATIONS . .
5-1 l 5.1 LIMIT OF OPERATION 5-1 5.1.1 Aeolicability 5-1 5.1.2 Actions 5-2 5.1.3 Surveillance Reauirements i i 5-2 f 5.1.4 pasis 5-3 5.2 DEMONSTRATION OF COMPLIANCE CHAPTER 6: POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR
. . . . . . . . . . . . 6-1 ACTIVITIES INSIDE THE SITE BOUNDARY 6-1 6.1 REQUIREMENT FOR CALCULATION 6-1 I 6.2 CALCULATIONAL METHOD iii Rev. 8, 1/94
Hatch ODCM ! TABLE OF CONTENTS (Cont inued 1 1 PAGE 7-1 CHAPTER 7 REPORTS . . . . . . . ................... 7-1 7.1 ANNUAL RADIOLOGICAL ENVIRONMEN*fAL SURVEILLANCE REPORT 7-1 7.1.1 Recuirement for Report 7-1 7.1.2 Report Contents 7-3 7.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 7-3 7.2.1 Recru irement for Reoort 7-3 7.2.2 Reoort Contents 7-7 7.3 MONTHLY OPERATING REPORT 7-7 7.4 SPECIAL ' WS
. . . . . . . . . . . . . . . . . . . 8-1 CHAPTER 8: METEO:. ni LL MODELS 8-1 8.1 ATMOSPHERIC Dr. tSION 8-1 8.1.1 Ground-Level Releases 8-3 8.1.2 Elevated Releases 8-4 8.1.3 tiixed-Mode Releases 8-5
! 8.2 RELATIVE DEPOSITION 8-5 ! 8.2.1 Ground-Level Releases 8-5 8.2.2 Elevated Releases 8-6 8.2.3 Mixed-Mode Releases 8-7 l ! 8.3 ELEVATED PLUME DOSE FACTORS l l CHAPTER 9: METHODS AND PARAMETERS FOR CALCULATION OF GASEOUS EFFLUENT PATHWAY DOSE FACTORS, Rg . . . . . . . . . . . . . . . . . 9-1 a 9-1 9.1 INHALATION PATHWAY FACTOR 9-2 9.2 GROUND PLANE PATHWAY FACTOR 9-3 9.3 GARDEN VEGETATION PATHWAY FACTOR 9-6 9.4 GRASS-COW-MILK PATHWAY FACTOR 9-9 9.5 GRASS-GOAT-MILK PATHWAY FACTOR c 9-12 9.6 GRASS-COW-MEAT PATHWAY FACTOR
. . . . . . . . . . . 10-1 CHAPTER los DEFINITIONS OF EFFLUENT CONTROL TERMS 10-1 l 10.1 TERMS SPECIFIC TO THE ODCM 10-5 10.2 TERMS DEFINED IN THE TECHNICAL SPECIFICATIONS l
l iv Rev. 8, 1/94 l f -. __
0 Hatch ODCM ] LIST OF TABLES .f PAGE 2-3 Table 2-1. Radioactive Liquid Et fluent Monitoring Instrumentation Table 2-2. Radioactive Liquid Effluent Monitoring Instrumentation
- 2-5
' Surveillance Requirements 2-9 Table 2-3. Radioactive Liquid Waste Sampling and Analysis Program t 2-18 Table 2-4. Applicability of Liquid Monitor Setpoint Methodologies Table 2-5. Parameters for Calculation of Doses Due to Liquid 2-33 Effluent Releases 2-34 $ Table 2-6. Element Transfer Factors 2-35 Table 2-7. Adult Ingestion Dose Factors 2-38 Table 2-8. Site-Related Ingestion Dose Factors, Ag i 3-3 Table 3-1. Radioactive Gaseous Effluent Monitoring Instrumentation ' Table 3-2. Radioactive Gaseous Effluent Monitoring Instrumentation 3-5 Surveillance Requirements 3-9 Table 3-3. Radioactive Gaseous Waste Sampling and Analysis Program 3-20 Table 3-4. Applicability of Gaseous Monitor Setpoint Methodologies Table 3-5. Dose Factors for Exposure to a Semi-Infinite Cloud of 3-37 Noble Cases Table 3-6. Dose Factors for Exposure to Direct Radiation from Noble 3-38 Gases in an Elevated Finite Plume 3-41 Table 3-7. Attributes of the Controlling Receptor 3-44 Table 3-8. R,,p; for Ground Plane Pathway, All Age Groups 3-45 Table 3-9. R,jp j for Inhalation Pathway, Child Age Group for Garden Vegetation Pathway, Child Age Group 3-46 Table 3-10. R aipj 4-4 Table 4-1. Radiological Environmental Monitoring Program Table 4-2. Reporting Levels for Radioactivity Concentrations in 4-7 Environmental Samples 4-8 Table 4-3. Values for the Minimum Detectable Concentration 4-13 Table 4-4. Radiological Environmental Monitoring Locations Table 6-1. Attributes of Member of the Public Receptor Locations Inside the 6-3 SITE BOUNDARY 8-10 Table 8-1. Terrain Elevation Above Plant Site Grade for the Garden Vegetation Pathway 9-5 Table 9-1. Miscellaneous Parameters 9-8 Table 9-2. Miscellaneous Parameters for the Grass-Cow-Milk Pathway 9-11 Table 9-3. Miscellaneous Parameters for the Grass-Goat-Hilk Pathway v Rev. 8, 1/94
Hatch OfLQM I LIST OF TABLES (Continued) PAGE 9-14 Table 9-4. Miscellaneous Parameters for the Grass-Cow-Meat Pathway 9-15 Table 9-5. Individual Usage Factors 9-16 Table 9-6. Stable Element Transfer Data 9-17 Table 9-7. Inhalation Dose Factors for the' Infant Age Group 9-20 Table 9-8. Inhalation Dose Factors for the Child Age Group 9-23 Table 9-9. Inhalation Dose Factors for the Teenager Age Group 9-26 Table 9-10. Inhalation Dose Factors for the Adult Age Group 9-29 Table 9-11. Ingestion Dose Factors for the Infant Age Group 9-32 Table 9-12. Ingestion Dose Factors for the Child Age Group 9-35 Table 9-13. Ingestion Dose Factors for the Teenager Age Group 9-38 Table 9-14. Ingestion Dose Factors for the Adult Age Group 9-41 Ta- ,o 9-15. External Dose Factors for Standing on Contaminated Ground 1 1 vi Rev. 8, 1/94 1 1
- _ - - - _ _ _ _ _ _ __ ' ' ~ " ' ' e n --nen,en
Hatch ODCM LIST OF FIGURES PAGE 2-15 Figure 2-1. Unit 1 Liquid Radwaste Treatment System 2-16 Figure 2-2. Unit 2 Liquid Radwaste Treatment System Schematic Diagram of the Condenser offgas Treatment System 3-18 Figure 3-1. 4-15 Figure 4-1. Sampling Location Map, Site Periphery Figure 4-2. Sampling Location Map Beyond Site Periphery, North and 4-16 West of Site Figure 4-3. Sampling Location Map Beyond Site Periphery, South and 4-17 West of Site of Site 4-18 Figure 4-4. Sampling Location Map Beyond Site Periphery, East Figure 4-5. Location of Additional Control Station for TLDs and 4-19 Vegetation 8-11 Figure 8-1. Vertical Standard Deviation of Material in a Plume (og) 0~12 Figure 8-2. Terrain Recirculation Factor (K r) 8-13 Figure 8-3. Plume Depletion Effect for Ground Level Releases 8-14 Figure 8-4. Plume Depletion Effect for 30-Meter Releases 8-15 Figure 8-5. Plume Depletion Effect for 60-Meter Releases 8-16 Figure 8-6. Plume Depletion Effect for 100-Meter Releases Relative Deposition for Ground-Level Releases 8-17 Figure 8-7 8-18 Figure 8-8. Relative Deposition f or 30-Meter Releases 8-19 Figure 8-9. Relative Deposition for 60-Meter Releases 8-20 Figure 8-10. Relative Deposition for 100-Meter (or Greater) Releases 10-9 Figure 10-1. Site Map for Effluent Controls vii Rev. 8, 1/94
\
f d Hatch ODCM i PEFERENCES Britz, and R.L. Waterfield,
- Preparation ,
J.S. Boegli, R.R. Bellamy, W.L. 1 , of Radiological Effluent Technical Specifications f or Nuclear Power
' Plants," NUREG-103 3, October 19 78.
- 2. " Calculation of Annual Doses to Man from Routine Releases of Reactor Ef fluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," U.S. KRC Reculatory Guide 1.109, March 1976.
- 3. " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," U.S.
KRC Raoulatory Guide 1.109, Revision 1, October 1977. j 4. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous U.S. NRC 1 Effluents in Routine Releases from Light-Water-Cooled Reactors," Reculatory Guide 1.111, March 1976, i 5.
" Methods for Estimating Atmospheric Transport and Dispersion of GaseousKRC Effluents in Routine Releases from Light-Water-cooled Reactors," U.S.
- Reculatory Guide 1.111. Revision 1, July 1977.
4 4 6.
" Estimating Aquatic Dispersion of Ef fluents f rom AccidentalI," and Routine U.S. NRC Reactor Releases for the Purpose of Implementing Appendix Reculatory Guide 1.113, April 1977.
- 7. Edwin I. Hatch Nuclear Plant - Unit 1 Final Safety Analysis Report, l Georgia Power Company.
- 8. Edwin I. Hatch Nuclear Plant - Unit 2 Final Safety Analysis Report, Georgia Power Company.
- 9. HNP-2 Environmental Report - Operatino License Stace,Hatch Georgia Power Plant - Unit 1 Company, July 1975 (including Appendix A, "Edwin I. Number 1," March 1975).
Preoperational Environmental Surveillance Report
- 10. Hatch Nuclear Plant Land Use Survev - 1987, Georgia Power Company,
' February 1987.
- 11. Hatch Nuclear Plant Land Use Survey, Georgia Power Company, November 1987.
Hatch Nuclear Plant Land Use Survey - 1988, Georgia Power Company,
- 12. 1 November 16, 1988 and December 20, 1988.
i 1
- 13. Letter to Georcia Power Company from Pickard, Lowe. and Garrick, Inc.,
Washington, D.C., May II, 1987.
- 14. Letter to Georcia Power Company from Pickard, Lowe, and Garrick. Inc.,
Washington, D.C., June 3, 1987.
- 15. Letter to Georcia Power Company from Pickard, Lowe, and Garrick. Inc.,
Washington, D.C., June 11, 1987. I
- 16. Letter to Georcia Power Company from Pickard, Lowe, and Garrick, Inc.,
Washington, D.C., November 30, 1987. a
- 17. L.A. Currie, Lower Limit of Detection: Definition and Elaboration of a Proposed Position of Radiolooical Effluent and Environmental Measurements, U.S. NRC Report NU REG / CR- 4007, 1984.
" Radiological Assessment Branch Technical Position", U.S. Nuclear 18.
Regulatory Commission, Revision 1, November 1979. J viii Rev. 8, 1/94
Hatch ODCM REFERENCES (Continueql U.S. Department l
- 19. D.H. Slade (ed.), Meteoroloov and Atomic Enerov - 1968, i of Commerce, July 1968.
- 20. M. Abramowitz and I.A. Stegun (eds.), Handbook of Mathematical Functions with Formulas. Graohs, and Mathematical Tables, National Bureau of Standards, U.S. Department of Commerce, 1965. ]
- 21. H.C. Nichols and S.D. Holder, Plant Edwin I. Hatch Units 1 and 2 Thermal Ela-a Modal verification, Georgia Power Company Environmental Affairs Center, March 1981. l Bill Duval to Howard Rocers, Georgia Power Company,
- 22. Internal F.crorandum.
October 11, 1990. HooDer, Georgia Power Company,
- 23. Internal Memorandum. W.H. Ollinoer to D.M.
June 9, 1987.
- 24. Letter to Georcia Power Company from Ouantum Technoloov. Inc., Marietta, l Georgia, June 17, 1987.
25 W.W. Meinke and T.H. Essig, "Offsite Dose Calculation Manual Guidance: i Standard Radiological Effluent Controls for Boiling Water Reactors," 1, NUREG-1302, April 1991. f Generic Letter 89-01 Supplement No.
- 26. D.C. Kocher, " Radioactive Decay Data Tables," U.S. DOE Report DOE / TIC- l 11026, 1981.
U.S. NRC Report
- 27. J.E. Till and H.R. Meyer, eds., Badiolooleal Assessment, NUREG/CR-3332, 1983.
/
- 28. Letter to Georola Power Company from J.H. Davis. Health Physics Consultant, Lilburn, Georgia, September 17, 1990.
Davis. Health Physics
- 29. Letter to Georcia Power Company from J.H.
Consultant, Lilburn, Georgia, March 25, 1991. ix Rev. 8, 1/94
i Hatch 00CM l 1 i I CHAPTER 1 i a
~ INTPODUCTION i
j r
- l i
of the Technical l The Of f site Dose Calculation Manual is a supporting document it describes the methodology and parameters to be used i Specifications. As such, in the calculation of offsite doses due to radioactive liquid and gaseous effluents, and in the calculation of liquid and gaseous effluent monitoring instrumentation alarm setpoints. In addition, it contains the following j 4 The controls required by the Technical Specifications, governing the
- I radioactive effluent and radiological environmental monitoring programs.
} e Schematics of liquid and gaseous radwaste etfluent treatment systems, which include designation of release points to UNRESTRICTED AREAS. 4 i i e A list and maps indicating the specific sample locations f or the Radio-I d logical Environmental Monitoring Program. i be included e Specifications and descriptions of the information that must j in the Annual Radiological Environmental Surveillance Report and the j Annual Radioactive Effluent Release Report required by the Technical Specifications. 1 The ODCM will be maintained at the plant for use as a reference guide and l training document of accepted methodologies and calculations. Changes in the l calculational methods or parameters will be incorporated into the ODCM in order j to ensure that it represents current methodology in all applicable areas. Any l computer sof tware used to perform the calculations described will be maintained current with the ODCM. Equations and metbods used in the ODCM are based on those presented in NUREG-0133 { (Reference 1), in Regulatory Guide 1.109 (References 2 and 3), in Regulatory Guide 1.111 (References 4 and 5), and in Regulatory Guide 1.113 (Reference 6). l I i i I i f l-1 Rev. 8, 1/94 j i i
Hatch 00CM CHAPTER 2 LIOU!D EFFLUENTS I l 2.1 LIMITS OF OPERATION l The following Liquid Effluent Controls implement requirements established by l ' Technical Specifications Section 6.0. Terms printed in all capital letters are defined in Chapter 10. 2.1.1 Licuid Ef fluent Monitorino Instrumentation Control l the radioactive liquid In accordance with Technical Specification 6.18(1), ef fluent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits specified in The alarm / trip setpotnts cf tnese channels shall Section 2.1.2 are not exceeded. l l be determined in accordance with Section 2.3. ! 2.1.1.1 Applicability I l As shown in Table 2-1. 1 ( 2.1.1.2 Actions I With a radioactive liquid ef fluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, declare the channel inoperable, or change the setpoint to a conservative value. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 2-1. When the ACTION statement or other requirements of this control cannot be met, Entry into steps need not be taken to change the Operational Mode of the unit. as a minimum, an Operational Mode or other specified CONDITION may be made if, the requirements of the ACTION statement are satisfied. 2.1.1.3 Surveillance Requirements Each radioactive liquid effluent monitoring instrumentation channel chall be SOURCE CHECK, CHANNEL demonstrated OPERABLE by performance of the CHANNEL CHECK, 2-1 Rev. 8, 1/94 l 1
l Hat:. ODCM tre frequencies s.:wn in CA1!BRATION, and CHANNE: FUNCTIONAL TEST Operations at Table 2-2. 2.1.1.4 Basis The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid ef fluents The Alam/ Trip during actual or potential releases of liquid effluents. Setpoints for these instruments shall be calculated and adjusted in accordance with.the methodology and parameters in Section 2.3 to ensure that the alarm / trip The OPERABILITY and will occur prior to exceeding the limits of Section 2.1.2. use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. l l i r i l i I 2-2 Rev. 3, 1/94 i l i i I . - , , ' -
- - - ~ - . - , , , , . . , . _ _,
H4tch OOCH Tacle ;-1. Radtoactive Ltqutd Effluent Mentter.ng Instrumentateen a OPERABILITY Requirements Minimum Channels b ACTION OPERABLE Applicability Instrument 1. Gross Radioactivity Honitors Providing Automatic Termination of Release Liquid Radwaste 1 (1) 100 Effluent Line Providing Automatic Termination of
- 2. Gross Radioactivity Monitors not Release Service Water System (2) 101 1
Effluent Line C
- 3. Flowrate Measurement Oevices
- a. Liquid Radwaste (1) 102 1
Effluent Line (1),(2) 102 1
- b. Discharge Canal
- 4. Dif ferential Pressure Measurement Sty.ces Service Water System to closed Cowling At all times 103 1
Water System
- a. All requirements in this Table apply to each uni.. ,
l i
- b. Applicability of requirements is as fo11cws:
(1) Whenever the radwaste discharge valves are not locked closed. (2) Whenever the Service Water System pressure is below the Closed Cooling Water System pressure, or AP indication is not available. an such cases, ACTION statement
- c. Pump curves may be used to estimate flow; 102 is not required.
l l l 2-3 Rev. 8, 1/94 l
Hatch 00CM Table 2-1 (contd). Notatacn f or Table 2 ACT!CN Statements
)
ACTION 100 - With the number of channels OPERABLE less than required by the a Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a releases a. At least two independent samples are analyzed in accordance with Section 2.1.2.3, and ]
- b. At least two technically qualified individuals independently the release verify rate discharge line valving and verify the 1 calculations.
If Otherwise, suspend release of radioactive ef fluents via this patt.way. the channel remains inoperable for over 30 days, an explanation of the circumstances must be included in the next Annual Radioactive Effluent Release Report. 4 ACTION 101 - With the number of channels OPERABLE less than required by the ef fluent releases via this pathway Minimum Channels OPERABLE requirement, may continue, provided that once per shift grab samples are collected and analyzed for gross gadioactivity at If a MINIMUM DETECTABLE the channel CONCENTRATION remains inoperable for over no higher than 1 x 10' uC i / mi. . 30 days, an explanation of the circumstances must be included in the next Annual Radioactive Effluent Release Report. 1 ACTION 102 - With the number of channels OPERABLE less than required by the
- Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided that theIfflowrate is estimated the channel at least once remains inoperable f orper over 4 hours during actual releases.
30 days, an explanation of the circumstances must be included in the next Annual Radioactive Effluent Release Report. 1 ACTION 103 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, assure that the Service Water System ef fluent monitor is OPERABLE. J l f 1 2-4 Rev. 8, 1/94
Hntch OQCM
!acie -2. Radicactive Liquid Ef fluent M:nitoring Instr. mentation Surveillance Requirements a
Surveillance Requirements CHANNEL. SOURCE CHANNEL FUNCTIONAL CHANNEL CHECK CALIBRATION TEST Instrument CHECK
- 1. Gross Radioactivity Monitors Providing Automatic Termination of Release c
I.iquid Radwaste Db Pc R Q Effluent Line
- 2. Gross Radioactivity Monitors not Providing Automatic Termination of Release Service Water System f Db M R Q Effluent Line
- 3. Flowrate Measurement Devices
- a. Liquid Radwaste Effluent g
Db,d g3 9
- .ine
- b. Discharge Db,d R Q NA Canal
- 4. Differential Proesure Measurement Devices Service Water System to closed cooling NA D NA R Water System l
l i 2-5 Rev. 8, 1/94
H*tch ODCM Tacle 2-2 (contd). Notation f or Table 2 Surve.llance Requirements
- a. All requirements in this Table apply to each unit.
- b. During releases via this pathway.
- c. In addition to the basic functions of a CHANNEL FUNCTIONAL TEST (Section 10.2), the CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
(1) Instrument indicates measured levels above the alarm / trip setpoint; (2) Instrument indicates an isolation on high alarm; or (3) Instrument controls are not set in operate mode.
- d. CHANNEL CHECK shall consist of verif ying indication of flow during periods of release. CRANNEL CHECK shall be made at least once daily on any day on which CONTINUOUS, periodic, or BATCH releases are made,
- e. The SOURCE CHECK shall consist of verifying that the instrument is reading onseale.
- f. In addition to the basic functions of a CHANNEL FUNCTIONAL TEST (Section 10.2), the CHANNEL FUNCTIONAL TEST shall aisc demonstrate that centrol room alarm annunciation occurs if any of the following conditions exists:
(1) Instrument indicates measured levels acove the alarm setpoint; (2) Instrument indicates a downscale failure; or (3) Instrument controls are not set in operate mode, j 2-6 Rev. 2, 1/94 . t l
Hatch ODCM l 2.I.2 L;;u;d Effluent C:ncentrat ton Cent rol and 6.18(3), the { i In accordance with Technical Specifications 6.18(2) l liquid effluents to concentration of radioactive material released in l l UNRESTRICTED AREAS (see Figure 10-1) shall be limited at all times to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for f radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 1x 10-4 pci/mL ; i total activity. l 2.1.2.1 Applicability This limit applies at all times. l i 2.1.2.2 Act ier.s With the concentration of radioactive material released in liquid effluents to 2.1.2, immediately UNRESTRICTED AREAS exceeding the limits stated in Section restore the concentration to within the stated limits. When the ACTION statement or other requirements of this control cannot be met, l steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified CONDITION may be made if , as a minimum, the requirements of the ACTION statement are satisfied. l I 2.1.2.3 Surveillance Requirements l The radioactivity content of each batch of radioactive liquid waste shall be determined by sampling and analysis in accordance with Table 2-3. The results j of radioactive analyses shall be used with the calculational methods in Section 2.3 to assure that the concentration at the point of release is maintained within the limits of Section 2.1.2. 2.1.2.4 Basis l l This control is provided to ensure that the concentration of radioactive materials released in liquid waste ef fluents to UNRESTRICTED AREAS will be less than ten times the concentration levels specified in 10 CFR 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of l radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the section II. A design objectives of Appendix I,10 CFR 50, to a KEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.1301 to the l population. The concentration limit for dissolved or entrained noble gases is I 2-7 Rev. 8, 1/94 i f i i
, . - . - . . .. - - _ . -- . . . ~ . - - - - . _ . _ - . - _ -_ - --.
l Hetch ODCM t I
' cased upon the assumption that Xe-135 ts the centro.';;ng radic;sotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological4Protection f was then (ICRP) Publication 2 (1959). The resulting concentration of 2 x 10 for Xe-135, stated l
multiplied by the ratio of the eftluent concentration limit j in Appendix B, Table 2, Column 1 of 10 CFR 20 (paragraphs 20.1001 to 20.2401), l I to the MPC for Xe-135, stated in Appendix B, Table II, Column 1 of 10 CFR 20 l (paragraphs 20.1 to 20.601), to obtain the limiting concentration of I x 10-4 uCL/mL. i l I l 1 i i I i 2-8 Rev. 8, 1/94 1
l Hat:r. 00CM I I l Tacle I-3. Padioactive Ltquid Waste Sampl.ng and Analysis Program a Sampling and Analysis Requirements .b MINIMUM DETECTABLE CONCENTRATION Minimum Liquid Sampling Analysis Type of Activity (MDC) FREQUENCY Analysis (uci/mL) Release Type FREQUENCY PRINCIPAL CAMMA 5 E-7 C P EMITTERS P f Each BATCH Each BATCH I-131 1 E-6 Dissolved and 1 E-5 P I One M Entrained Gases Batch Waste BATCH /M (Gamma Emitters) Release Tanks "~ M l P Each BATCH COMPOSITE Gross Alpha 1 E-7 i Sr-89, Sr-90 5 E-B p g Each BATCH COMPOSITE Fe-55 2 E-6
)
- a. All requirements in this table apply to each unit.
Terms printed in all capital letters are defined in Chapter 10. l
- b. l l
- c. For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radio- Under nuclides at or near the required MINIMUM DETECTABLE CONCENTRATION.
these circumstances, the required MINIMUM DETECTABLE CONCENTRATION may be increased inversely proportionally to the magnitude of the ganna yield (i.e., SE-7/I, where I
= photon abundance expressed as a decimal l fraction). In no case shall the MINIMUM DETECTABLE CONCENTRATION, as calculated in this manner for a specific radionuclide, be greater than 10 value spectfied percent of the corresponding Ef fluent Concentration Limit in 10 CFR 20 Appendix B, Table 2, Colamn 2.
l l 1 Rev. 8, 1/94 2-9 i
^ Hatch ODCM 2.1.3 Liould Effluent Dose Control In accordance with Technical Specifications 6.18(4) and 6.18(5), the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid of fluents released, f rom each unit, to UNRESTRICTED AREAS (see Figure 10-1) sha11 be limited:
- a. During any calendar quarter to less than or equal to 1.5 mram to the total and body and to less than or equal to 5 mrem to any organ,
- b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.
l 2.1.3.1 Applicability These limits apply at all times. 2.1.3.2 Actions With the calculated dose from the release of radioactive materials in liquid ef fluents exceeding any of the limits of Section 2.1.3, prepare and submit to the ' Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s); defines the corrective actions to be taken to reduce the releases; and defines the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the limits of Section 2.1.3. For Unit 2, this report is in lieu of any other report required by Technical Specification 6.9.1. This report shall also include (1) the results of radiological analyses of the drinking water source, and (2) the radiological impact on finished water supplies with regard to the requirements of 40 CFR 141, the Safe Drinking Water Act. 1 i When the ACTION statement or other requirements of this control cannot be met, j Entry into steps need not be ta' ten to change the Operational Mode of the unit. j an Operational Mode or other specified CONDITION may be made if, as a minimum, the requirements of the ACTION statement are satisfied. 2.1.3.3 Surveillance Requirements At least once per 31 days, cumulative dose contributions f rom liquid ef fluents for the current calendar quarter and the current calendar year shall be determined, for each unit, in accordance with Section 2.4. f 2-10 Fev. 8, 1/94 j
I i I Hatch ODCM l 1 I.1.3.4 Basis This control is provided to implement the requirements of Sections II.A, III.A and IV. A of Appendix I, 10 CFR Part 50. The limits stated in Section 2.1.3 implement the guides set forth in Section II. A of Appendix I. The ACTIONS stated in Section 2.1.3.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid ef fluents will be kept "as low as is reasonably achievable." Also, for f resh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the f acility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculations in Section 2.4 I, which state that implement the requirements in Section III.A of Appendix l l conformance.with the guides of Appendix I be shown by calculational procedures l based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC , The f through appropriate pathways is unlikely to be substantially underestimated. l equations specified in Section 2.4 for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with I the methodology provided in Regulatory Guide 1.109 (Reference 3) and Regulatory Guide 1.113 (Reference 6). This control applies to the release of liquid effluents from each unit at the site. The liquid ef fluents f rom shared LIQUID RADWASTE TREATMENT SYSTEMS are to be proportioned between the units. l l 1 2-11 Rev. 8, 1/94
Hatch Oggg i 2.1.4 Ltcutd Radwaste Treatront Sys e- Tcntre'. . In accordance with Technical Specification 6.18(6), the LIQUID RADWASTE TREATKENT SYSTEM shall be OPERABLE. The apprcpriate portions of the system shall be used to reduce radioactivity in liquid wastes pr'ior to their discharge when the f rom each unit, to UNRESTRICTED AREAS projected doses due to the liquid ef fluent, (see Figure 10-1) would exceed 0.06 mrom to the total body or O.2 mrom to any ! organ of a MEMBER OF THE PUBLIC in 31 days. 2.1.4.1 Applicability This limit applies at all times. i 2.1.4.2 Actions j With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the LIQUID RADWASTE TREATMENT SYSTEM not in operation, prepare and submit to the Nuclear Regulatory. Commission within 30 days pursuant to Technical Specification 6.9.2 a Special Report which includes the following information:
- a. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems and the reason for inoperability, l
l l t b. Action (s) taken to restore the inoperable equipment to OPERABLE status, and l l c. Summary description of action (s) taken to prevent a recurrence. ' When the ACTION statement or other requirements of this controi cannot be met, steps need not be taken to change the operattonal Mode of the unit. Er.try into I t an Operational Mode or other specified CONDITION may be made if, as a minimum, the requirements of the ACTION statement are satisfied. l 2.1.4.3 Surveillance Requirements least Doses due to liquid releases to UNRESTRICTED AREAS shall be projected at once per 31 days, in accordance with Section 2.5, during periods in which the discharge of untreated liquid ef fluent containing radioactive materials occurs or is expected to occur. 2-12 Rev. 8, 1/94
i ! Hatch OCCM The installed LIQU:D IUCWASTE TREATMENT SYSTEM shal. ce derenstrated OPERABI.E oy ( meeting the controls of Section 2.1.2 and 2.1.3. ' l. 2.1.4.4 Basis The OPERABILITY of the LIQUID RADWASTE TREATMENT SYSTEM ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the UNRESTRICTED AREAS. The requirement that the appropriate portions the releases of of this system be used when specified provides assurance that radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This control 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 I to 10 CFR Part 50. The specified f limits governing the use of appropriate portions of the LIQUID RADWASTE TREATMENT f 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, f or liquid ef fluents. This control applies to the release of radioactive materials in liquid ef fluents f rom each unit at the site. For units with shared radwaste systems, the liquid ef fluents f rom the shared system are to be proportioned among the units sharing that system. l l I 2.1.5 Maior Chances to Licuid Radioactive Waste Treatment Systems l l I Licensee initiated MAJOR CHANGES TO LIQUID RADIOACTIVE WASTE TREATMENT SYSTEMS:
- a. Shall be reported to the Nuclear Regulatory Commission in the Annual f
Radioactive Effluents Release Report for the period in which the change f was implemented, in accordance with Section 7.2.2.7.
- b. Shall becomte ef fective upon review and approval by the Plant Review Board.
2-13 Rev. 8, 1/94 i !- - - - - - - - - - - - _ - - _ _ _ _ _ _ _ ~ " " ' **"T?'7 Y e m y ye, w99.y , , _
i l Hatch ODCM 2.2 LIQUID RADWASTE TRIATMENT SYSTEM The Edwin I. Hatch Nuclear Plant is located on the south bank of the Altamaha River, which supplies make-up water to the circulating water system and receives blowdown from the cooling tower. There are two boiling water reactors on the j site. Each unit is served by a separate liquid radwaste treatment system. l Schematics of the liquid radwaste treatment systems are presented in Figure 2-1 and Figure 2-2. The dotted lines indicate alternate pathways through which liquid radwaste may be routed. The two units release liquid radwaste to separate discharge lines. Dilution flow l is furnished by the cooling tower blowdown and plant service water systems, if necessary. Releases from plant service water systems are to the main condenser circulating flume, or to the cooling tower blowdown discharge line when needed for additional dilution. Since each unit is served by a separate dilution l stream, liquid releases may be made independently from each of the two units. l t l Although no significant quantities of radioactivity are expected in the plant service water systems, these ef fluent pathways are monitored as a precautionary measure. I l l l l l l l l 2-14 Rev. 8, 1/94 l
l Haten 00CM l l i i narft op, ST ANOaRO atCYCL E SURGE e- .---- .. I ... 8 TANE s a 1 I 8 088 ST AJeOARO RECYCLE I c - - - - - - - - -- - - - - - - - - - xi f
- - - -- - - - II i e 3 a i l saastg , TO WASTt WAstg ,
nasti . saaelg i ConsOttsSATE COLLICTOA penum 8 e COLLECTom TAmK 4 FILTt m T W TAhR a i ETOMAGE T AAs u l I t g g 8 1 4 g l 4 I i l i I
+ -====-- I i
0F8 (TANQaRD RECYCLE M ^ j l r-----.--....,,,,. I i
- i I FLOOR PLOOM i R A06 ATION PLOom FLOOR f OmalN I ORAIN I ORA 8N I heOeslTOR Omath '
g [ FILTE R DEMIM SAMPLE COLLICTOR ,
, I AM E {
' TAha > 3 I l . I i 6..-..a TO l l
' D:scMARGE CANAL
! l i !
t CM t uiCA L I p ggg( ; 6
**UI ' ar44TE ,
- i TANS - ga, pg g ,
TANK i .
. i i i I I i i I ! .
' l i i i o t uie, I !
7tto Tamm !!
\
l Lav=Omv l Lauwomy D R *"a i O m a ni. _ l Ta%< L T E R Figure 2-1. " nit 1 Liquid Radwaste T r e a t.T e nt System 2-15 Rev. 8, 1/94
i l I i
'iatCn 00CM l
"O 088 STANDARO atCYCti 1 Sua05 . . _ . .,
, _ . . q TANK i i '
I l 8 i I e j *
; I I '
I y Of f STANDA AD RECYCLE ( ..........-.- - _. 4 ---_- -... I
; I '
- i i t i
I gagyg ,gg7g , 70 MAKTE pmg s coj,'/c' a *
- LI g
gi C a'N g I
== .
I c O'L"l^" TANK I
- I le 1 t
l'...--.... g I l l I
' OFFSTANOAmo atCYCLE I i ;
_--_---_-.A._______^_- - l l I w' i 1 4 i MAD 6ATION g g
#LOOR #LOOR , ptoon e WONITOR FLO4m i oEAIN '- OR AIN ' ' A Da AIN -= Oh3M COLLECTom Pt LTi m ; Ot wN , g33,pg g l TANK TANK 9
', I g
, t to i.._....' OttCHAmot li I
CAhAL I. CM a wcAL mg u,ug wAsTt ] .4,yg l TANK J , pg g , TANK ! ! i : 1 4 1 I i T Ce t 6d eCA L n A.st NigT A ALs2 g a TAAR f I. l l Figure 2-2. Unit 2 Liquid Radwaste Treatment System 2-16 Rev. 8, 1/94
I i 1 Hatch ODCM i i
.3 LIQUID EFFLUENT MONITOR SETPOINTS 2.3.1 General Provisions Recardina Setooints Liquid monitor setpoints calculated in accordance with the methodology presented in this section will be regarded as upper bounds for the actual high alarm setpoints. That is, a lower value for the high alarm setpoint may be established or retained on the monitor, if desired. Intermediate level setpoints should be established at an appropriate level to give suf ficient warning prior to reaching the high alarm setpoint. If no release is planned for a particular pathway, or if there is no detectable activity in the planned release, the monitor setpoint l
should be established as close to background as practical to prevent spurious alarms, and yet alarm should an inadvertent release occur. l Two basic setpoint methodologies are presented below. For radwaste system discharge monitors, setpoints are determined to assure that the lirrits of Secuon 2.1.2 are not exceeded. For monitors on streams that are not expected to contain significant radioactivity, the purpose of the monitor setpoints is to cause an alarm on low levels of radioactivity, and to terminate the release where this is possible. Section 2.1.1 establishes the requirements for liquid effluent monitoring instrumentation. Table 2-4 lists the monitors for which each of the setpoint methodologies is applicable, l l l l
}
I i l 2-17 Rev. 8, 1/94 i k
i d Natch QDE Tacle 2-4. Appl 1Cability Of Liquid MOni;Or Setpoint Meta dologies l ' Liquid Radweste Discharge Monitors Setpoint Method: Section 2.3.2 1 Release Type: BATCH i Unit 1 or Unit 2 Liould Radwaste System Effluent Monitor: 1D3' 9 07 / 2Dll-N007 4 4 Normally Low-Radioactivity Streams with Termination or Diversion upon Alarm J Plant Hatch has no liquid ef fluent streams in this category. I Normally Low-Radioactivity Streams wi- t' rm Only Setpoint Method: Section 2.s.3 Release Type: CONTINUOUS i 4 Unit 1 or Unit 2 Plant Service Water System Effluent 4 Monitor: IDil-N008 / 2Dil-N008 .I 4 4 } i i a I } 1 4 4 0 1 k
! 2-18 Rev. 8, 1/94 i
Hatch ODCM l i l 2.3.2 Setroints fcr Padwnste Svstem 2.senarte Mon tors . t t l 2.3.2.1 overview of Method LIQUID RADWASTE TREATMENT SYSTEM effluent line radioactivity monitors are intended to provide alarm and automatic termination of release prior to exceeding the limits specified in Section 2.1.2 at the point of release of the diluted I affluent into the UNRESTRICTED AREA. Therefore, their alars/ trip eetpoints are established to ensure compliance with the following equation (equation adapted l from Addendum to Reference 1}: c f 5 TT CECL (2.1) T+f where the Ff fluent Concentration Limit corresponding to the mix CECL = of radionuclides in the effluent being considered for discharge, in uCi/mL. e= the setpoint , in pCi/mL, of the radioactivity monitor measuring the concentration of rad.oactivity in the effluent line prior to i dilution and subsequent release. The setpoint represents a concen-tration which, if exceeded, could result in concentrations exceeding the limits of Section 2.1.2 in the UNRESTRICTED AREA. 1 f= the ef fluent flowrate at the location of the radioactivity monitor, i l r in gpm. l F= the dilution stream flowrate which can be assured prior to the l release point to the UNRESTRICTED AREA, in gpm. A predetermined l dilution flowrate must be assured f or use in the calculation of the radioactivity monitor setpoint. TF = the tolerance factor selected to allow flexibility in the establishment of a practical monitor setpoint which could l accommodate ef fluent releases at concentrations higher than the ECL values stated in 10 CFR 20, Appendix B, Table 2, Column 2; the tolerance factor must not exceed a value of 10. While equation (2.1) shows the relationships of the critical parameters that-determine the setpoint, it cannot be applied practically to a mixture of radio-2-19 Rev. 8, 1/94 _ . . _ _ ._ .~ _ _ _ _, _ _ ,
l l l Hntch 00CM-I i ' n.C.tces with different Effluent Concentratten Lim;ts (ECLs). For a mixture of radtonuclides, equation (2.1) is gattsfied in a practicable manner based on the calcula*,ed ECL fraction of the radionuclide mixture and the dilution stream f l flov.-ate that can be assured for the duration of the release (F ),dby calculating
'.ne maximum permissible effluent flowrate (f ,) and the radioactivity monitor l
- setpoint (c).
l The setpoint method presented below is applicable to the release of only one tank of liquid radwaste per reactor unit at a given time. Liquid releares must be controlled administrative 1y to ensure that this condition is met; otherwise, the t i setpoint method may not ensure that the limits of Section 2.1.2 are not exceeded. l l 2.3.2.2 Setpoint Calculation Steps l l l Step 1: Determine the radionuclide concentrations in the liquid waste being I considered for release in accordance with the sampling and analysis requirements of Section 2.1.2. All liquid radwastes are collected in tanks for sampling and analysis prior to release. To ensure that a representative sample can be taken from a tank, its 7 l contents will be recirculated for a minimum time per; d to allow adequate mixing of the contents. Minimum recirculation times are as follows (Reference 22): f l Minimum Pecircu;ation Time (minutes) l Tank (s) Unit,_1 Unit 2 waste sample tanks 40 40 Floor drain sample tanks 70 105 I chemical waste sample tanks 65 65 Demineralizer feed tank 115 NA Laundry drain tanks 50 NA l Tne total concentration of the liquid waste is determ:ned by the results of all required analyses on the collected sample, as follows: {Cf = Ca+bCs+Cf + C, + { C g (2.2) i 5 % where: C, = the gross concentration of alpha emitters in the liquid waste, not less than that measured in the most recent applicable composite sample. 2-20 Rev. 8, 1/94
Maten ODCM C, = tne concentration Of strontium radiolsetope s (Sr-89 or Sr-90) in the liquid waste, not less than that measured in the most recent applicable composite sample. the concentration of Fe-55 in the liquid waste, not less than that Cf= measured in the most recent applicable composite sample. C, = the concentration of H-3 in the liquid waste, not less than that measured in the most recent applicable composite sample. Cg = the concentration of gamma emitter g in the liquid waste as j neasured by gamma ray spectroscopy performed on the sample f or the release under consideration. The Cg term will be included in the analysis of each waste sample; terms for gross concentrations of alpha emitters, Sr-89, Sr-90, Fe-55, and tritium will be included in accordance with the sampling and analysis program required for the waste stream (see Section 2.1.2). For each analysis, only radionuclides l identified and detected above background for the given measurement should be included in the calculation. When using the alternate setpoint methodology of step 5.b, the historical maximum values of C , a C,g Cr, and C, shall be used. Step ?* Determine the required dilution f actor f or the mix of radionuclides detected in the waste. l Measured radionuclide concentrations are used to calculate ECL fractions. The ECL fractions are used along with a safety factor to calculate the required dilution f actor; this is the minimum ratio of dilution flowrate to waste flowrate that must be maintained throughout the release to ensure that the limits of Section 2.1.2 are not exceeded at the point of discharge into the UNRESTRICTED AREA. The required dilution factor, RDF, is calculated as the sum of the dilution factors required for gamma emitters (RDF y ) and for non-gamma-emitters (RDFnyI C-RDF = ' ( ( b { ECL'-. 8 (2.3) i i
=
RDF y
- RDFw, 2-21 Rev. 8, 1/94
Hatch ODCM I C" E ECL # (2.4)
. z ED ,7 (SF) (TE)
Ca C, C Cg
.g . f.
ECL a s ECL, ECL f ECL g, g2,s} RDF = 9 (SF) (TF) l l where: C, = the measured concentration of radionuclide i as defined in step 1, in pCi/mL. The C,, C, 3 cf, and Ct terms will be included in the calculation as appropriate. I ECL, = the Effluent Concentration Limit for radionuclide i from 10 CFR Part 20, Appendix B, Table 2, Column 2 (except for noble gases as discussed below). In the absence of information regarding the f solubility classification of a given radionuclide in the waste stream, the solubility class with the lowest ECL shall be assumed. l For dissolved or entrained noble gases, the concentration shall be l d pCi/mL. For gross alpha, the ECL shall be 2x10~9 limited to lxlO uCi/mL; if specific alpha-emitting radionuclides are measured, the ECL for the specific radionuclide(s) should be used. SF = the safety factor selected to compensate for statistical fluctuations and errors of measurement. The value for the safety factor must be between 0 and 1. A value of 0.5 is reasonable for liquid releases; a more precise value may be developed if desired. TF = the tolerance factor (as defined in Secti n 2.3.2.1). Sten 3: Determine the release-specific assured dilution stream flowrate. Determine the dilution stream flowrate that can be assured during the release period, designated F;d this value is the setpoint for the dilution stream flowrate measurement device. If simultaneous radioactive releases are plan: f rom the same reactor unit, the unit's dilution stream must be allocated among all the simultaneous releases, whether or not they are monitored during release. Normally, only the batch tank 2-22 Rev. 8, 1/94
__ ~ - . - . _ . = - . . - . - . . . . .. . .- Hatch ODCM effluents need be considered, unless there is detectaDie radioactivity in one cf the normally low-radioactivity s t re ams (see Table 2-4). Allocation of the dilution stream to multiple release paths is accomplished as follows: Tgp
= Tg (Ar p) (2.6) where F 4p =
the dilution flowrate allocated to release pathway p, in gym. b AF p = the dilution allocation f actor f or release pathway p. AFp may be assigned any value between 0 and 1 f or each active release pathway, under the condition that the sum of the Ar p for all active release pathways for each unit does not exceed 1. [ Note: Because the two ur;its have separate dilution streams, the two units do not affect each other with respect to dilution allocation.] F3= the assured minimum dilut ian flow for the unit, in gpm. For Plant Hatch, Fd is normally established at 10,000 gpm. If more precise allocation factor values are desired, they may be determined cased on the relative radiological impact of each active release pathway; this may be approximated by multiplying the RDF of each effluent stream by its respective planned release flowrate, and comparing these values. If only one I release pathway for a given reactor unit contains detectable radioactivity, its f AFp may be assigned the value of 1, making Fdp equal to Fd ' l For the case where RDF s 1, the planned release meets the limits of Section 2.1.2 without dilution, and may be released with any desired effluent flowrate and dilution flowrate. l Step 4: Determine the maximum allowable waste discharge flowrate. l For the case where RDF > 1, the maximum permissible ef fluent discharge flowrate for this release pathway, fmp (in gpm), is calculated as follows: l Idp (2 7) I mp (RDF) For the case RDF s 1, equation (2.7) is not valid. However, as discussed above, when RDF 51, the release may be made at full discharge pump capacity; the radio-l 2-23 Rev. 8, 1/94 l
I Hatch CDCM activity mentter setpotnt must still be calculated in accordance with Step 5 below. i NOTE la Discharge flowrates are actually limited by the discharge pump capacity. When the calculated maximum permissible release flowrate exceeds the pump capacity, the release may be made at full capacity. Discharge flowrates less than the pump capacity must be achieved by throttling if this is available; if throttling is not available, the release may not be made as planned. NOTE 2: If, at the time of the planned release, there is detectable radio-activity due to plant operations in the dilution stream, the diluting capacity of the dilution stream is diminished. (In addition, sampling and analysis of the other radioactive ef fluents af fecting the dilution stream must be suf ficient to ensure that the liquid effluent dose limits specified in the controls of Section 2.1.3 are not exceeded.) Under these conditions, equation (2.7) must be modified to account for the radioactivity present in the dilution stream prior to the introduction of the planned release: Tgp i C' t mp
- 1-[ - { (2.8)
RDT , r Ed a \ ECL,, t I where: l l Cir = the measured concentration of radionuclide i in release pathway r that is contributing to radioactivity in the dilution stream. f = the effluent discharge flewrate of release pathway r. l r If the entire dilution stream contains detectable activity due to ; plant operations, whether or not its source is identified, fr = rd' and C ar is the concentration in the total dilution system. This i note does not apply: a) if the RDF of the planned release is s 1; or b) if the release contributing radioactivity to the dilution-stream has been accounted for by the assignment of an allocation factor. l 2-24 Rev. 8, 1/94 i i
Hatch ODCM Steo 5: Determine the maximum radioactivity monitor setpo . nt concentration. Based on the values determined in previous steps, the radioactivity monitor setpoint for the planned release is calculated to ensure that the limits of Section 2.1.2 will not be exceeded. Because the radioactivity monitor responds primarily to gamma radiation, the monitor setpoint cp for release pathway p (in pCi/mL) is based on the concentration of ganna emitters in the waste stream, as follows: C p
- Ap EC g (2.9) where Ap = an adjustment factor which will allow the setpoint to be established in a practical manner to prevent spurious alarms while allowing a-margin between measured concentrations and the limits of Section 2.1.2.
Steo 5.a. If the concentration of gamma emitters in the af fluent to be released is sufficient that the high alarm setpoint , t can be established at a level that will prevent spurious f f ' alarms, A should be calculated as follows: l p 1 I A = n ADF P gy 1 (2.10) RDF f ap 1 I t where: ADF = the assured dilution factor. f,p = the anticipated actual discharge flowrate for the planned release (in gpm), a value less than f mp* The release must then be controlled so that the actual effluent discharge flowrate does not exceed f,p at any time. 1 l Steo 5.b. Alternatively, A may p be calculated as follows: l 2-25 Rev. 8, 1/94 i l _. ._. .
Hatch ODCM ADF ~ RDF,,7 A = (2 11) P any Steo 5.c. Evaluate the computed value of A p as follows: If Apt 1, calculate the monitor setpoint, c. p However, if cp is within about 10 percent of cg , it may be impractical to use this value of e. p This situation indicates that measured concentrations are approaching values which would cause the limits of Section 2.1.2 to be exceeded. Therefore, steps should be taken to reduce potential con-centrations at the point of discharge; these steps may include decreasing the planned effluent discharge flowrate, increasing the dilution stream flowrate, postponing simultaneous releases, and/or decreasing the ef fluent concentrations by further processing the liquid planned for release. Alternatively, allocation factors for the active liquid release pathways may be reassigned. When one or more of these actions has been taken, repeat Steps 1-5 to calculate a new radioactivity monitor setpoint. If Ap < 1, the release may not be made as planned. Consider the alternatives discussed in the paragraph above, and ) calculate a new setpoint based on the results of the f actions taken. 2.3.2.3 Use of the Calculated setpoint The setpoint calculated above is in the units uCi/mi. The monitor actually measures a count rate that includes background, so that the calculated setpoint I must be converted accordingly: cp = p p +B p cE (2.1a) where: c = the monitor setpoint as a count rate. Ep = the monitor calibration factor, in count rate /(pci/mL). Monitor calibration data for conversion between count rate and ; concentration may include operational data obtained from 2-26 Rev. 8, 1/94
_. -. __ ..__. _ _ _ . . . - _ -._ =_ i .. - . _ l l Hatch ODCM l l determining tne mon;t0r response t: strea concentrata ns measurec by liquid sample analysis. l l the monitor backgru nd count rate. In all cases, monitor Sp = background must be controlled so that the monitor is capable of responding to concentrations in the range of the setpoint value. 1 The count rate units of e p, E ,p and B in p equation (2.7a) must be the same (cpn or cps). i 1 I I l l 2.3.3 Setooints for Monitors on Normally Low-Radioactivity Streams i l i Radioactivity in these streams (listed in Table 2-4 above) is expected to be at very low levels, generally below detection limits. Accordingly, the purpose of these monitors is to alarm upon the occurrt.;ce of significant radioactivity in I these streams, and to terminate or divert the release where this is possible. ) l 2.3.3.1 Normal Conditions i l
*When radioactivity in one of these streams is at its normal low level, its radio-activity 1.nitor setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should an inadvertent release occur.
2.3.3.2 Conditions Requiring an Elevated Setpoint ! Under the following conditions, radionuclide concentrations must be determined and an elevated radioactivity monitor setpoint determined for these pathways:
- For streams that can be diverted or isolated, a new monitor setpoint must be established when it is desired to discharge the stream directly to the dilution water even though the radioactivity in the stream exceeds the l
! level which would normally be diverted or isolated. I e For streams that cannot be diverted or isolated, a new monitor setpoint must be established whenever the radioactivity in the stream becomes detectable above the background levels of the applicable laboratory analyses; or the associated radioact'ivity monitor detects activity in the stream at levels above the established alarm setpoint. When an elevated monitor setpoint is required for any of these ef fluent streams, 2.3.2. it should be determined in the same manner as described in Section 2-27 Rev. 8, 1/94
-. -_ _ _ _ . . . _ . _ - . , -._ _ . . _ _ , _ , , - . _ , _ . - . ~ _ . . . _ . _ , .
Hatch ODCM However, special consideration must be given to Step 3. Ar. allocation factor must be assigned to the normally low-radioactivity release pathway under I consideration, and allocation factors for other release pathways discharging simultaneously must be adjusted downward (if necessary) to ensure that the sum of the allocation factors does not exceed 1. Sampling and analysis of the l I normally low-radioactivity streams must be auf ficient to ensure that the liquid ef fluent dose limits specified in the controls of section 2.1.3 are not exceeded. ' I ! \ I 1 I 1 l I 1 l 1 l i i l l i i l l 2-28 Rev. 8, 1/94
l l Hatch ODCM ; l 2.4 LIQUID EFFLUENT DOSE CALCULA!!ONS The following sub-sections present the methods required for liquid ef fluent dose calculatione, in deepening levels of detail. Applicable site-specific pathways and parameter values for the calculation of D7 , A g, and CFjy are summarized in Table 2-5. 2.4.1 calculation of Dose The dose limits for a MEMBER OF THE PUBLIC specified in Section 2.1.3 are on a per-unit basis. Therefore, the doses calculated in accordance with this section j must be determined and recorded on a per-unit basis, including apportionment of releases shared between the two units. For the purpose of implementing Section 2.1.3, the dose to the maximum exposed individual due to radionuclides identified in liquid ef fluents released f rom each as follows (equation from Ref-unit to UNRESTRICTED AREAS will be calculated erence 1, page 15): m (2.12) D7 = { Air [ ( Att Cil TII i . i= 1 l where: Dr = the cumulative dose commitment to the total body or to any organ T, l in mrem, due to radioactivity in liquid ef fluents released during j the total of the m time periods Atg. Ag = the site-related adult ingestien dose commitment factor, for the total body or for any organ 7, due to identified radionuelade i, in (mrem mL)/(h*pC1). Methods for the calculation of A g are presented below in Section 2.4.2. The values of A;7 to be used in dose calculations for releases from the plant site are listed in Table 2-8. Atj = the length of time period 1, over which Cd and Fg are averaged f or liquid releases, in h. I I C;; = the average concentration of radionuclide i in. undiluted liquid effluent during time period 1, in ;ci/mL. Only radionuclides l l 2-29 Rev. 8, 1/94
- - - - _ - . . . ~ - -. .- - -_ - . - . - -- .- -.
3 I, Hatch ODCM i identifted and detected above background in their respective I samples should be included in the calculation. d i 4 Fg = the near-field average dilution factor in the receiving water of ,\ d the UNRESTRICTED AREA: 5 fg j ff
= (2.13)
TxZg i where: l fg= the average undiluted liquid waste flowrate ~ actually j observed during the period of radioactivity release, in
- gpn.
l
'F t = the average dilution stream flowrate actually observed i during the period of radioactivity release, in gpm.
I } Z= the applicable dilution f actor for the receiving water body, in the near field of the discharge structure, during the period of radioactivity release, from ! Table 2-5. i In equation (2.13), the product (F x Z) is limited to NOTE: 1000 cfs (= 448,000 gpm) or less. (Reference 1, Section 4.3.) 3 2.4.2 calculation of Aj7 J The site-related adult ingestion dose commitment factor, A ir, is calculated as follows (equation adapted from Reference 1, page 16, by addition of the irrigated i garden vegetation pathway):
-A l 'w (2.14)
Ajf = 1.14 x 10 5 , . q , - Ai 'f + u,, crg, org i ) where: i i 1.14 x 105 = a units conversion factor, determined by: 106 pci/pci = 10 3 mL/L + 8760 h/y. 1 l 1 2-30 Rev. 8, 1/94 4
}
4, , -
-_ - . , . - . ..- . - - _ - = - . - - - . - - . - - . - . . . .-.
.. ~ . - - - . .-._ --.. - . , - - -- --
Hatch ODCM l l Uw = the adult drinking water consumption rate applicable to the plant l site (L/y). D, = the dilution f actor f rom the near field of the discharge structure for the plant site to the potable water intake location. the decay constant f or radionuclide 1 (h'I) . Values of 1;used in f 1; = effluent calculations should be based on decay data from a recognized and current source, such as Reference 26. tw = the transit time from release to receptor for potable water consumption (h). 1 l Ur = the adult rate of fish consumption applicable to the plant site l (kg/y). BFj= the bicaccumulation factor for radionuclide i applicable to l freshwater fish in the receiving water body for the plant site, in i (pci/kg)/(pci/L) = (L/kg). For specific values applicable to the I plant site, see Table 2-6. ! l tg = the transit time f rom release to receptor for fish consumption (h) . f \ i Uy = the adult consumption rate for irrigated garden vegetation i applicable to the plant site (kg/y). ! CFgy
= the concentration factor for radionuclide i in irrigated garden i
vegetation, as applicable to the vicinity of the plant site, in (pci/kg)/(pci/L). Methods for calculation of CF ng are presented below in Section 2.4.3. DFj7 = the dose conversion factor for radionuclide i for adults, in organ T (mrem /pci). For specific values, see Table 2-7. 2.4.3 Calculation of CFjy The concentration factor for radionuclide i in irrigated garden vegetation, CFn, in (L/kg), is calculated as follows: 2-31 Rev. 8, 1/94 e m
3 i Hntch ODCM o For radionuclides other than tri:Aum (equat.cn adapted from Reference 3. l equations A-8 and A-9):
~ '
I ) #1 8iv (1 ~
- I - Al 'h (2.15) r (1 - e + e CF;y = M.I YvAEi Pkg i
l o For tritium (equation adapted from Reference 3, equations A-9 and A-10): f CT;y . = M*L y (2.16) j where l M= the additional river dilution facter from the near field of the l discharge structure for the plant site to the point of irrigation water usage. l 2 I= the average irrigation rate during the growing season (L)/(m h). , i r= the fraction of irrigation-deposited activity retained on the edible portions of leafy garden vegetation. Yy = the areal density (agricultural productivity) of leafy garden vegetation (kg/m 2) f; = the fraction of the year that garden vegetation is irrigated. B 3y = the crop to soil concentration factor applicable to radionuelide' i (pci/kg garden vegetation)/(pci/kg soil). , l P= the ef fective surf ace density of soll (kg/m 2 ). lj = the decay. constant for radionuclide 1 (h-I). Values of 1; used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 26. 1, = the rate constant for removal of activity from plant leaves by weathering (h-I). 1 2-32 Rev. 8, 1/94 l
. . _ . . _ _ . - _ _ - . , _ . . -._i
Hatch 00CM r l A = the effective rem val rate f:r activity cepesited on crop . eaves Ei (h-I) calculated as lEn " la'A w' i t, = the period of leafy garden vegetation exposure during the growing season (h). tb= the period of long-term buildup of activity in soil (h). the time between harvest of garden vegetation and human consumption th= ! (h). Ly = the water content of leafy garden vegetation edible parts (L/kg). 2-33 Rev. 8, 1/94
l Hatch ODCM Table 2-5. Parameters f or Calculation of Doses Due to Liquid Ef fluent Releases Dose Calculation Receptor Locations: I Vicinity of plant discharge Einh f Drinkino Water: None++ l Irricated Carden Veoetation: None+ + l Mumerical Parameters Parameter Value Referene.g l [ Ref. 9, Sec. 5.1; Ref. 6, Sec. B; I 2 10 f Ref. 21 I U, O L/y (Pathway not applicable) l Dw 1.0 12 h Ref. 3, Sec. A.2 tw 21 kg/y Ref. 3, Table E-5 f Uf l 24 h Ref. 3, Sec. A.2 tg Uy 0 kg/y (Pathway not applicable) M 1.0 I No value r 0.25 Ref. 3, Table E-15 Yy 2.0 kg/m2 Ref. 3, Table E-15 fg 1.0 P 240 kg/m2 Ref. 3, Table E-15 1, 0.0021 h*I (i.e., half- Ref. 3, Table E-15 life of 14 d) l t, 1440 h (= 60 d) Ref. 3, Table E-15 tb 1.31 x 105 h (= 15 y) Ref. 3, Table E-15 t 24 h Ref. 3, Table E-15 h Ly 0.92 L/kg Based on Ref. 27, Table 5.16 (for lettuce, cabbage, etc.) 1 l
* - Becaune there is no known drinking water pathway or irrigated garden i veget ation pathway downstream of the plant site, the parameters for these pat'4 ways are def ault values, and the usage factors are set to 0.
+ ",nere is no established default value for this parameter. The most conservative physically realistic value is 1.0.
* * - There is no established def ault value for this parameter. A value will be 2-34 Rev. 8, 1/94
Hatch ODCM Table 2-6. Element Transfer Factors Freshwater Fish Element . BFj H 9.0 E-01 C 4.6 E+03 Na 6.6 E+01 P 2.5 E+04 Cr 1.5 E+02 Mn 8.9 E+01 Fe 6.0 E+00 Co 1.7 E+02 NL 1.0 E+02 Cu 4.4 E+01 Zn 2.9 E+02 Br 4.2 E+02 Rb 2.0 E+03 Sr 3.8 E+00 Y 2.5 E+01 Zr 1.9 E+02 Nb 4.1 E+01 Mo 1.8 E+02 Tc 1.5 E+01 Ru 4.6 E+00 Rh 1.0 E+01 Ag 3.5 E+02 Sb 1.0 E+00 Te 4.0 E+02 I 4.3 E+01 Cs 5.8 E+02 Ba 5.0 E+00 La 2.5 E+01 Ce 8.4 E+01 Pr 2.5 E+01 Nd 4.6 E+01 W 1.2 E+03 Np 1.0 E+01 l *- Bioaccumulation Factors for freshwater fish, in (pci/kg)/(pCi/L). l They are obtained from Reference 9 (Appendix A, Table 2.3-1), l except as follows: Reference 2 (Table A-8) for Sb. l 2-35 Rev. 8, 1/94
4 Hatch ODCM f 7able 2-7. Adult Ingestion Dose Factors Bone Liver T. Body Thyroid Kidney Lung GI-LLI 5 Nuclide i 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 H-3 No Data C-14 2.84E-06 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 No Data 2.17E-05 P-32 1.93E-04 1.20E-05 7.46E-06 No Data No Data l Cr-51 No Data No Data 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 1 1 4 Mn-54 No Data 4.57E-06 8.72E-07 No Data 1.36E-06 No Data 1.40E-05 ! i Mn-56 No Data 1.15E-07 2.04E-08 No Data 1.46E-07 No Data 3.67E-06 Fe-55 2.75E-06 1.90E-06 4.43E-07 No Data No Data 1.06E-06 1.09E-06 ! Fe-59 4.34E-06 1.02E-05 3.91E-06 No Data No Data 2.85E-06 3.40E-05 , 1 Co-58 No Data 7.45E-07 1.67E-06 No Data No Data No Data 1.51E-05 Co-60 No Data 2.14E-06 4.72E-06 No Data No Data No Data 4.02E-05 Ni-63 1.30E-04 9.01E-06 4.36E-06 No Data No Data No Data 1.88E-06 Ni-65 5.28E-07 6.86E-08 3.13E-08 No Data No Data No Data 1.74E-06 Cu-64 No Data 8.33E-08 3.91E-08 No Data 2.10E-07 No Data 7.10E-06 Zn-65 4.84E-06 1.54E-05 6.96E-06 No Data 1.03E-05 No Data 9.70E-06 Zn-69 1.03E-08 1.97E-08 1.37E-09 No Data 1.28E-08 No Data 2.96E-09 Br-83 No Data No Data 4.02E-08 No Data No Data No Data 5.79E-08 Br-84 No Data No Data 5.21E-08 No Data No Data No Data 4.09E-13 Br-55 No Data No Data 2.14E-09 No Data No Data No Data No Data Rb-86 No Data 2.11E-05 9.83E-06 No Data No Data No Data 4.16E-06 Rb-88 No Data 6.05E-08 3.21E-08 No Data No Data No Data 8.36E-19 Rb-89 No Data 4.01E-08 2.82E-08 No Data No Data No Data 2.33E-21 Sr-89 3.08E-04 No Data 8.84E-06 No Data No Data No Data 4.94E-05 Sr-90 7.58E-03 No Data 1.86E-03 No Data No Data No Data 2.19E-04 Sr-91 5.67E-06 No Data 2.29E-07 No Data No Data No Data 2.70E-05 _ All values are in (mrem /pci ingested). They are obtained from Reference 3 (Table E-11), except as follows: Reference 2 (Table A-3) for Rh-105, Sb-124, and Sb-125, 2-36 Rev. 8, 1/94
Hatch ODCM Table 2-7 (contd). Adult Ingestion Dose Factors Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 2.15E-06 No Data 9.30E-08 No Data No Data No Data 4.26E-05 Y-90 9.62E-09 No Data 2.58E-10 No Data No Data No Data 1.02E-04 l Y-91m 9.09E-11 No Data 3.52E-12 No Data No Data No Data 2.67E-10 Y-91 1.41E-07 No Data 3.77E-09 No Data No Data No Data 7.76E-05 Y-92 8.45E-10 No Data 2.47E-11 No Data No Data No Data 1.48E-05 Y-93 2.68E-09 No Data 7.40E-11 No Data No Data No Data 8.50E-05 Zr-95 3.04E-08 9.75E-09 6.60E-09 No Data 1.53E-08 No Data 3.09E-05 l Zr-97 1.68E-09 3.39E-10 1.55E-10 No Data 5.12E-10 No Data 1.05E-04 Nb-95 6.22E-09 3.46E-09 1.86E-09 No Data 3.42E-09 No Data 2.10E-05 Ho-99 No Data 4.31E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06 Tc-99m 2.47E-10 6.98E-10 8.89E-09 No Data 1.06E-08 3.42E-10 4.13E-07 Tc-101 2.54E-10 3.66E-10 3.59E-09 No Data 6.59E-09 1.87E-10 1.10E-21 Ru-103 1.85E-07 No Data 7.97E-08 No Data 7.06E-07 No Data 2.16E-05 Ru-105 1.54E-08 No Data 6.08E-09 No Data 1.99E-07 No Data 9.42E-06 Ru-106 2.75E-06 No Data 3.48E-07 No Data 5.31E-06 No Data 1.78E-04 Rh-105 1.22E-07 8.86E-08 5.83E-08 No Data 3.76E-07 No Data 1.41E-05 Ag-110m 1.60E-07 1.48E-07 8.79E-08 No Data 2.91E-07 No Data 6.04E-05 Sb-124 2.81E-06 5.30E-08 1.11E-06 6.79E-09 No Data 2.18E-06 7.95E-05 Sb-125 2.23E-06 2.40E-08 4.48E-07 1.98E-09 No Data 2.33E-04 1.97E-05 Te-125m 2.68E-06 9.71E-07 3.59E-07 8.06E-07 1.09E-05 No Data 1.07E-05 Te-127m 6.77E-06 2.42E-06 8.25E-07 1.73E-06 2.75E-05 No Data 2.27E-05 Te-127 1.10E-07 3.95E-00 2.38E-08 8.15E-08 4.48E-07 No Data 8.68E-06 Te-129m 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 No Data 5.79E-05 Te-129 3.14E-08 1.18E-08 7.65E-09 2.41E-08 1.32E-07 No Data 2.37E-08 Te-131m 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 No Data 8.40E-05 Te-131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 No Data 2.79E-09 l l l l 2-37 Rev. 8, 1/94
Hatch ODCM Table 2-7 (contd). Adult Ingestion Dose Factors s Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 No Data 7.71E-05 I-130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 No Data 1.92E-06 I-131 4.16E-06 5.95E-06 3.41E-06 1.95E-03 1.02E-05 No Data 1.57E-06 I-132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 No Data 1.02E-07 I-133 1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31E-06 No Data 2.22E-06 I-134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 No Data 2.51E-10 I-135 4.43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 No Data 1.31E-06 Cs-134 6.22E-05 1.48E-04 1.21E-04 No Data 4.79E-05 1.59E-05 2.59E-06 Cs-136 6.51E-06 2.57E-05 1.85E-05 No Data 1.43E-05 1.96E-06 2.92E-06 Cs-137 7.97E-05 1.09E-04 7.14E-05 No Data 3.70E-05 1.23E-05 2.11E-06 Cs-138 5.52E-08 1.09E-07 5.40E-08 No Data 8.01E-08 7.91E-09 4.65E-13 Ba-139 9.70E-08 6.91E-11 2.84E-09 No Data 6.46E-11 3.92E-11 1.72E-07 Ba-140 2.03E-05 2.55E-08 1.33E-06 No Data 8.67E-09 1.46E-08 4.18E-05 i Ba-141 4.71E-08 3.56E-11 1.59E-09 No Data 3.31E-11 2.02E-11 2.22E-17 i Ba-142 2.13E-08 2.19E-11 1.34E-09 No Data 1.85E-11 1.24E-11 3.00E-26 La-140 2.50E-09 1.26E-09 3.33E-10 No Data No Data No Data 9.25E-05 La-142 1.28E-10 5.82E-11 1.45E-11 No Data No Data No Data 4.25E-07 Ce-141 9.36E-09 6.33E-09 7.18E-10 No Data 2.94E-09 No Data 2.42E-05 Ce-143 1.65E-09 1.22E-06 1.35E-10 No Data 5.37E-10 No Data 4.56E-05 Ce-144 4.88E-07 2.04E-07 2.62E-08 No Data 1.21E-07 No Data 1.65E-04 Pr-143 9.20E-09 3.69E-09 4.56E-10 No Data 2.13E-09 No Data 4.03E-05 Pr-144 3.01E-11 1.25E-11 1.53E-12 No Data 7.05E-12 No Data 4.33E-18 Nd-147 6.29E-09 7.27E-09 4.35E-10 No Data 4.25E-09 No Data 3.49E-05 W-187 1.03E-07 8.61E-08 3.01E-08 No Data No Data No Data 2.82E-05 Np-239 1.19E-09 1.17E-10 6.45E-11 No Data 3.65E-10 No Data 2.40E-05 2-38 Rev. 8, 1/94
Hatch ODCM Taole 2-8. Site-Related Ingestion Dose Factors, Ag Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI ! l H-3 0.00 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 C-14 3.13E+04 6.26E+03 6.26E+03 6.26E+03 6.26E+03 6.26E+03 6.26E+03 Na-24 8.86E+01 8.86E+01 8.86E+01 8.86E+01 8.86E+01 8.86E+01 8.86E+01 P-32 1.10E+07 6.84E+05 4.25E+05 0.00 0.00 0.00 1.24E+06 Cr-51 0.00 0.00 9.32E-01 5.575-01 2.05E-01 1.24E+00 2.34E+02 Mn-54 0.00 9.72E+02 1.85E+02 0.00 2.89E+02 0.00 2.985+03 Mn-56 0.00 3.87E-02 6.86E-03 0.00 4.91E-02 0.00 1.23E+00 Fe-55 3.95E+01 2.73E+01 6.36E+00 0.00 0.00 1.52E+01 1.56E+01 l Fe-59 6.14E+01 1.44E+02 5.53E+01 0.00 0.00 4.03E+01 4.81E+02 f 0.00 0.00 6.09E+03 Co-58 0.00 3.00E+02 6.73E+02 0.00 Co-60 0.00 8.71E+02 1.92E+03 0.00 0.00 0.00 1.64E+04 Ni-63 3.11E+04 2.16E+03 1.04E+03 0.00 0.00 0.00 4.50E+02 Ni-65 1.72E-01 2.23E-02 1.02E-02 0.00 0.00 0.00 5.66E-01 Cu-64 0.00 2.37E+00 1.11E+00 0.00 5.97E+00 0.00 2.02E+02 Zn-65 3.35E+03 1.07E+04 4.82E+03 0.00 7.13E+03 0.00 6.72E+03 Zn-69 1.14E-07 2.19E-07 1.52E-08 0.00 1.42E-07 0.00 3.28E-08 Br-83 0.00 0.00 3.83E-02 0.00 0.00 0.00 5.52E-02 Br-84 0.00 0.00 1.22E-12 0.00 0.00 0.00 9.61E-18 l Br-85 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Rb-86 0.00 9.73E+04 4.53E+04 0.00 0.00 0.00 1.92E+04 Rb-88 0.00 1.29E-22 6.82E-23 0.00 0.00 0.00 1.78E-33 Rb-89 0.00 1.61E-26 1.14E-26 0.00 0.00 0.00 0.00 Sr-89 2.76E+03 0.00 7.93E+01 0.00 0.00 0.00 4.43E+02 f l Sr-90 6.90E+04 0.00 1.69E+04 0.00 0.00 0.00 1.99E+03 Sr-91 8.95E+00 0.00 3.62E-01 0.00 0.00 0.00 4.26E+01 Sr-92 4.22E-02 0.00 1.83E-03 0.00 0.00 0.00 8.36E-01 Y-90 4.44E-01 0.00 1.19E-02 0.00 0.00 0.00 4.71E+03 j Y-91m 1.04E-11 0.00 4.01E-13 0.00 0.00 0.00 3.04E-11 Y-91 8.34E+00 0.00 2.23E-01 0.00 0.00 0.00 4.59E+03 Y-92 4.60E-04 0.00 1.35E-05 0.00 0.00 0.00 8.06E+00 j Y-93 3.09E-02 0.00 8.53E-04 0.00 0.00 0.00 9.80E+02 Zr-95 1.37E+01 4.39E+00 2.97E+00 0.00 6.88E+00 0.00 1.39E+04 Zr-97 2.86E-01 5.76E-02 2.63E-02 0.00 8.70E-02 0.00 1.78E+04 Nb-95 5. 99 E -01 3.33E-01 1.79E-01 0.00 3.29E-01 0.00 2.02E+03 Mo-99 0.00 1.44E+03 2.75E+02 0.00 3.27E+03 0.00 3.35E+03 Tc-99m 5.59E-04 1.58E-03 2.01E-02 0.00 2.40E-02 7.75E-04 9.36E-01 All values are in (mrem mL)/(h pCi). They are calculated using equation (2.14 ) , and data from Table 2-5, Table 2-6, and Table 2-7. When "No Data" is shown for a radionuclide-organ combination in Table 2-7, A g factors in this table are presented as zero. 2-39 Rev. 8, 1/94
l 1 Hatch ODCM Table 2-8 (contd). Site-Related Ingestion Dose Factors, A g Liver T. Body Thyroid Kidney Lung CI-LLI Nuclide Bone 0.00 7.03E-32 2.00E-33 0.00 Te-101 2.71E-33 3.91E-33 3.83E-32 0.00 8.62E-01 0.00 7.64E+00 0.00 2.34E+02 Ru-103 2.OOE+00 0.00 1.58E-03 0.00 5.17E-02 0.00 2.45E+00 Ru-105 4.00E-03 0.00 3.83E+00 0.00 5.84E+01 0.00 1.96E+03 Ru-106 3.02E+01 1.83E+00 1.33E+00 8.72E-01 0.00 5.63E+00 0.00 2.11E+02 Rh-105 1.24E+02 7.34E+01 0.00 2.43E+02 0.00 5.05E+04 Ag-110m 1.34E+02 6.65E+00 1.25E-01 2.63E+00 1.61E-02 0.00 5.16E+00 1.88E+02 Sb-124 5.33E+00 5.743-02 1.07E+00 4.74E-03 0.00 5.57E+02 4.71E+01 Sb-125 9.19E+02 3.40E+02 7.63E+02 1.03E+04 0.00 1.01E+04 Te-125m 2.54E+03 2.30E+03 7.85E+02 1.65E+03 2.62E+04 0.00 2.16E+04 Te-127m 6.44E+03 6.38E+00 3.85E+00 1.32E+01 7.24E+01 0.00 1.40E+03 Te-127 1.78E+01 4.02E+03 1.71E+03 3.71E+03 4.50E+04 0.00 5.43E+04 Te-129m 1.08E+04 6.68E-06 4.33E-06 1.36E-05 7.47E-05 0.00 1.34E-05 Te-129 1.78E-05 4.65E+02 3.88E+02 7.37E+02 4.71E+03 0.00 4.62E+04 Te-131m 9.51E+02 3.61E-17 2.73E-17 7.10E-17 3.78E-16 0.00 1.22E-17 Te-131 8.64E-17 1.26E+03 1.18E+03 1.39E+03 1.22E+04 0.00 5.97E+04 Te-132 1.95E+03 5.98E+01 2.36E+01 5.06E+03 9.32E+01 0.00 5.14E+01 I-130 2.03E+01 5.62E+02 3.22E+02 1.84E+05 9.63E+02 0.00 1.48E+02 I-131 3.93E+02 I-132 1.51E-02 4.04E-02 1.41E-02 1.41E+00 6.43E-02 0.00 7.59E-03 6.57E+01 1.14E+02 3.48E+01 1.68E+04 1.99E+02 0.00 1.03E+02 T 3
'M 6.26E-08 1.70E-07 6.09E-08 2.95E-06 2.71E-07 0.00 1.48E-10 3.68E+00 9.64E+00 3.56E+00 6.36E+02 1.55E+01 0.00 1.09E+01 I-135 Cs-134 8.63E+04 2.05E+05 1.68E+05 0.00 6.64E+04 2.21E+04 3.59E+03 Cs-136 8.58E+03 3.39E+04 2.44E+04 0.00 1.88E+04 2.58E+03 3.85E+03 Cs-137 1.11E+05 1.51E+05 9.91E+04 0.00 5.14E+04 1.71E+04 2.93E+03 Cs-138 2.64E-12 5.22E-12 2.59E-12 0.00 3.84E-12 3.79E-13 2.23E-17 Ba-139 7.05E-06 5.03E-09 2.07E-07 0.00 4.70E-09 2.85E-09 1.25E-05 Ba-140 2.30E+02 2.89E-01 1.51E+01 0.00 9.83E-02 1.66E-01 4.74E+02 Ba-141 1.06E-24 8.00E-28 3.57E-26 0.00 7.44E-28 4.54E-28 4.99E-34 0.00 0.00 0.00 0.00 0.00 0.00 l Ba-142 0.00 9.89E-02 4.99E-02 1.32E-02 0.00 0.00 0.00 3.66E+03 l La-140 2.19E-07 9.96E-08 2.48E-08 0.00 0.00 0.00 7.27E-04 La-142 l 1.84E+00 1.25E+00 1.41E-01 0.00 5.79E-01 0.00 4.76E+03 j Ce-141 Ce-143 2.00E-01 1.48E+02 1.64E-02 0.00 6.52E-02 0.00 5.54E+03 9.79E+01 4.09E+01 5.26E+00 0.00 2.43E+01 0.00 3.31E+04 Ce-144 Pr-143 5.23E-01 2.10E-01 2.59E-02 0.00 1.21E-01 0.00 2.29E+03 1.48E-28 6.14E-29 7.51E-30 0.00 3.46E-29 0.00 2.13E-35 Pr-144 6.50E-01 7.52E-01 4.50E-02 0.00 4.39E-01 0.00 3.61E+03 Nd-147 W-187 1.47E+02 1.23E+02 4.30E+01 0.00 0.00 0.00 4.03E+04 Np-239 2.12E-02 2.09E-03 1.15E-03 0.00 6.51E-03 0.00 4.28E+02 l
2-40 Rev. 8, 1/94 t l -
Hatch ODCM 2.5 LIQUID EFFLUENT DOSE PROJECTIONS 2.5.1 Thirty-One Day Dose Proiections In order to meet the requirements for operation of the LIQUID RADWASTE TREATMENT SYSTEM (see Section 2.1.4), dose projections must be made at least once each 31 days; this applieu during periods in which a discharge to UNRESTRICTED AREAS of liquid effluents containing radioactive materials occurs or is expected. Projected 31-day doses to individuals due to liquid ef fluents may be determined j l as follows: l 0 - x 31 + 0, (2.U) 7 9 , - , l where D 7p = the projected dose to the total body or organ 7, for the next 31 days of liquid releases. D 7g = the cumulative dose to the total body or organ T, for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration. l t = the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues into the next 1 quarter). D ra = the anticipated done contribution to the total body or any organ 7, due to any planned activities during the next 31-day period, if those activities will result in liquid releases that are in addition to routine liquid effluents. If only routine liquid effluents are anticipated, D 7, may be set to zero. 2.5.2 Dose Proiections for Specific Releases Dose projections may be performed for a particular release by performing a pre-release dose calculation assuming that the planned release will proceed as anticipated. For individual dose projections due to liquid releases, follow the methodology of Section 2.4, using sample analysis results for the source to be released, and parameter values expected to exist during the release period. 2-41 Rev. 8, 1/94 i i
l i Hatch ODCM 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS l l The following symbolic terms are used in the presentation of ' liquid af fluent calculations in the sub-sections above. Section of Definition Initial Use l Iggm ) l Ap = the adjustment factor used in calculating the l l effluent monitor setpoint for liquid release pathway l p the ratio of the assured dilution to the required dilution [unitiess). 2.3.2.2 l l ADF = the assured dilation factor for a planned release 2.3.2.2 f [unitiess). AF p = the dilution allocation factor for liquid release pathway p [unitless). 2.3.2.2 i the site-related adult ingestion dose commitment PSr = factor, for the total body or for any organ r, due to identified radionuclide i [ (mrem mL) / ( ha uci) ) . The values of A are ILsted in Table 2-8. 2.4.1 7 B.= 3 the crop to soil concentration factor applicable to radionuclide i, [ (pci/kg garden vegetation) / (pci/kg soil)). 2.4.3 j i l BFj = the bioaccumulation factor for radionuclide i for freshwater fish [(pci/kg)/(pCi/L} }. Values are listed in Table 2-6, 2.4.2 i l c= the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the effluent ! line, prior to dilution and subsequent release [pci/mL). 2.3.2.1 cp = the calculated effluent radioactivity monitor setpoint for liquid release pathway p [pci/mL). 2.3.2.2 C, = the gross concentration of alpha emitters in the liquid waste as measured in the applicable composite j sample [pci/mL). 2.3.2.2 I 2-42 Rev. 8, 1/94 l
Hatch ODCM Section of Definition Initial Use Term l the Ef fluent concentration Limit stated in 10 CFR 20, i CECL = Appendix B, Table 2, Column 2 [yci/mL). 2.3.2.1 Cg = the concentration of Fe-55 in the liquid waste as ; measured in the applicable composite sample l (yci/mL). 2.3.2.2 C = the concentration of gamera emitter g in the liquid E waste as measured by gamma ray spectroscopy performed on the applicable pre-release waste sample (pci/mL). 2.3.2.2 Cj = the measured concentration of radionuclide i in a sample of liquid effluent (pci/mL). 2.3.2.2 C,3 = the average concentration of radionuclide i in undiluted liquid effluent during time period 1 (pci/mL). 2.4.1 C g= the measured concentration of radionuclide i in l release pathway r that is contributing to radio-activity in the dilution stream [uci/mL). 2.3.2.2 l. 1 1 , 1 C3 = the concentration of strontium radioisotope n (Sr-89 l I or Sr-90) in the liquid waste as measured in the applicable composite sample (yci/mL). 2.3.2.2 the concentration of H-3 in the liquid waste as l Ct= measured in the applicable composite sample (pci/mL). 2.3.2.2 CF;y = the concentration factor for radionuclide i in irrigated garden vegetation [(pci/kg)/(pci/L)). 2.4.2 D, = the dilution factor from the near field of the discharge structure to the potable water intake location [unitiess). 2.4.2 l 2-43 Rev. 8, 1/94 _. _ _ - _ _ . _ _ ___ ~ - - , . _- , . , , , , - _ _.
l Hatch ODCM Section of i nefinition Initial Use j leIm ( the cumulative dose commitment to the total body or l Dy = to any organ t, due to radioactivity in liquid effluents released during a given time period 2.4.1 (arem). D ra = the anticipated does contribution to the total body or any organ t, due to any planned activities during 2.5.1 the next 31-day period (mrom). Dy = the cumulative dose to the total body or organ 7, for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release 2.5.1 under consideration (mrem). D rp = the projected dose to the total body or organ 7, for the next 31 days of liquid releases (mrem). 2.5.1 l DFir = the dose conversion factor for radionuclide i for adults, in organ T [ mrem /pci). Values are listed in 2.4.2 Table 2-7. l ECL, = ti.e liquid Effluent Concentration Limit for radio-nuclide i from 10 CFR Part 20, Appendix B, Table 2, Column 2 (pci/mL). 2.3.2.2 l f= the effluent flowrate at the location of the radio- l activity monitor (qpm). 2.3.2.1 l f,p = the anticipated actual discharge flowrate for a planned release from liquid release pathway p 2.3.2.2 (9Pm). fg = the fraction of the year that garden vegetation is 2.4.3 l irrigated (unitiess). f mp = the maximum permissible effluent discharge flowrate for release pathway p (gpm). 2.3.2.2 2-44 Rev. 8, 1/94 _
Match ODCM-Section of Term Definition Initial Use l f f r
= the effluent discharge flowrate of release pathway r g9p ), 2.3.2.2 l
ft= the average undiluted liquid waste flowrate actually observed during the period of a liquid release 2.4.1 l [9Pm). l F= the dilution stream flowrate which can be assured prior to the release point to the UNRESTRICTED AREA [gpm). 2.3.2.1 the entire assured dilution flowrate for the plant Fd= site during the release period [gpa). 2.3.2.2 ; F dp the dilution flowrate allocated to release pathway p
=
2.3.2.2 19Pm). i i Fg = the near-field average dilution factor in the t I receiving water of the UNRESTRICTED AREA l [unitless). 2.4.1 ; Fg= the average dilution stream flowrate actually observed during the period of a liquid release [gpm]. 2.4.1 I = the average irrigation rate during the growing season [L/(m2 *h)). 2.4.3 l I Ly = the water content of leafy garden vegetation edible i parts [L/kg). 2.4.3 I M= the additional river dilution factor from the near field of the discharge struc;ure for the plant site to the point of irrigation dater usage [unitless). 2.4.3 l P= the effective surface density of soil [kg/m 2). 2.4.3 I L l l 2-45 Rev. 8, 1/94
l l l t l 4 Hatch ODCM i Section of Definition Initial Use Term r= the fraction of irrigation-deposited. activity retained on the edible portions of leafy garden i 2.4.3 l vegetation. 1 the required dilution factort the minimum ratio by l l RDF = j which liquid ef fluent must be diluted before reaching l the UNRESTRICTED AREA, in order to ensure that the limits of Section 2.1.2 are not exceeded l 2.3.2.2 l (unitiess). l RDF y = the RDF for a liquid release due only to its concen-tration of gamma-emitting radionuelides (unitless). 2.3.2.2 l l RDF ny = the RDF for a liquid release due only to its concen-I tration of non-g amma-emitt ing radionuclides 2.3.2.2 (unitiess). J ' l selected to compensate for .I SF = the safety factor statistical fluctuations and errors of measurement 2.3.2.2 (unitiess). t = the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration. 2.5.1
- tb= the period of long-term buildup of activity in soil 1 2.4.3
[h). t, = the period of leafy garden vegetation exposure during the growing season (h). 2.4.3 t ! the transit time from release to receptor for fish tg = consumption th). 2.4.2 the time between harvest of garden vegetation and th= human consumption [h). 2.4.3 tw = the transit time f rom release to receptor for potable water consumption [h). 2.4.2 2-46 Rev. 8, 1/94
Hatch ODCM Section of Definiti2D Initial Use Term TF = the tolerance f actor selected to allow flexibility in , the establishment of a practical monitor setpoint accommodate effluent releases at which could concentrations higher than the ECL values stated in 10 CFR 20, Appendix B, Table 2, Column 2 (unitiess); 3.3.2.2 the tolerance factor must not excwed a value of 10. the adult rate of fish consumption (kg/y). 2.4.2 Uf= Uy = the adult consumption rate for. irrigated garden 2.4.2 vegetation (kg/y). Uw = the adult drinking water consumption rate applicable 2.4.2 to the plant site (L/y). Y, = the areal density (agricultural productivity) of leafy garden vegetation (kg/m2 ). 2.4.3 l Z= the applicacle dilution factor for the receiving l l water body, in the near field of the discharge structure, during the pariod of radic< activity release 2.4.1 (unitless). Atg = the length of time period 1, over which C,g and Fg are averaged for liquid releases (h). 2.4.1 the ef fective removal rate f or activity deposited on 1Ei = crop leavee [ h- I ) . 2.4.3 lj= the decay constant for radionuclide i ( h'I ) . 2.4.2 ! A w
= the rate constant for removal of activity from plant 2.4.;
leaves by weathering (h-I). 2-47 Rev. 8, 1/94~ l l _ - ,_. _ . _. ,, -~ _, . - - - - . . . . .
r l Haten OCCM CHAPTER 3 , GASEOUS EFFLUENTS I LIMITS OF OPERATION l 3.1 l l The following Limits of operation Laplement requirements established by Technical . ; Specifications section 6.0. Terms printed in all capital letters are defined in Chapter 10. 3.1.1 Gaseous Effluent Monitorina Instrumentation Control In accordance with Technical Specification 6.18(1), the radioactive gaseous l ef fluent monitoring instrumentation channels shown an Table 3-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Section 3.1.2.a are not exceeded. The alarm / trip setpoints Of these rhannels snall be determined in accordance with Section 3.3. 3.1.1.1 Applicability These limits apply as shown in Table 3-1. l 3.1.1.2 Actions Witn a radicactive gaseous ef fluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive gaseous efflue..ts monitored by the affected ! channel, declare the channel inoperable, or restore the setpoint to a value that will ensure that the limits of Section 3.1.2.a are met. With less than the minimum number of radioactive gaseous ef fluent monitoring instrumentation enannels OPERABLE, take the ACTION shown in Table 3-1. When the ACTION statement or other requirements of this control cannot be met, steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified CONDITION may be made if, as a minimum, the requirements of the ACTION statement are satisfieo. 3.1.1.3 Surveillance Requirements Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL 3-1 Rev. 8, 1/94
Hatch ODCM CAL;BRATION, and CF.ANNEL FUNCTICNAL TEST cperat i:ns at tne frequencies sn wn .n Table 3-2. 2.1.1.4 Basis l t The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section 3.3 to ensure that the alarm / trip will occur prior to exceeding the limits of Section 3.1.2.a. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendi.x A to 10 CFR Part 50. i l l r l l l . l l l I I 3-2 Rev. 8, 1/94
Mat;h ODCM Tacle 3'.. Rad &Oactive GaseOwn Effluent M:nitOrang Instrumentation Minimum Channels Applica-Instrument OPERABLE bility ACTION
- 1. heactor Building Vent Stack Monitoring System (Each Unit)
- a. Noble Gas Activity Monitor c 1 (a) 105
- b. Iodine Sampler Cartridge 1 (a) 107
- c. Particulate Sampler Filter 1 (a) 107
- d. Effluent System Flowrate Measurement Device 1 (a) 104
- e. Sampler Flowrate Measurement Device 1 (a) 104
- 2. Recombiner Building '/entilati0n M0nitoring System
- a. Noble Gas Activity Monitor C 1 (a) 105
- b. Iodine Sampler Cartridge ] (a) 107
- c. Particulate Sampler Filter 1 (a) 107
- d. Sampler Flowrate Monitor 1 (a) 104
- 3. Main Stack Monitoring System
- a. Noble Gas Activity Monitor C 1 (a) 105
- b. Iodine Sampler Cartridge 1 (a) 107
- c. Particulate Sampler Filter 1 (a) 107
- d. Effluent System Flowrate Measurement Device 1 (a) 104
- e. Sampler Flowrate Measurement Device 1 (a) l 104
- 4. Condenser Offgas Pretreatment Monitor (Each Unit)
- a. Noble Gas Activity Moniter 1 (b) 108 3-3 Rev. 8, 1/94
I Hetch ODCM Notatten fcr Tac;e 3-1 Tacie 3-1 (contd). l
- a. During radioactive releases via this pathway.
- b. During operation of the main condenser air ejector.
of responding to a MINIMUM DETECTABLE
- c. Monitor must be capajlepCi/mL.
CONCENTRATION of 1 x 10 ACTION 104 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirernent, effluent releases via this pathway may continue provided the flowrate is estimated at least once per 4 hours. l If the number of channels OPERABLE remains less than required by the minimum channels OPERABLE requirement for over 30 days, an explanation of the circumstances shall be included in the next Annual Radioactive Effluent Release Report. ACTION 105 - With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement, effluent releases via this pathway samples are may continue provided grab samples are taken daily and these analyzed for gross activity within 24 hours. With the number of main stack monitoring system channels OPERABLE less than required by the minimum channels OPERABLE requirement, immediately suspend drywell purge. If the number of channels OPERABLE remains less than required by the minimum channels OPERABLE requirement f or over 30 days, an explanation of the circumstances shall be included in the next Annual Radioactive ! Effluent Release Report. ACTION 107 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway j provided samples are continuously collected with auxiliary
- may continue, equipment for periods on the order of 7 days and analyzed within 48 hours
( after the end of the sampling period. If the number of channels OPERABLE remains less than required by the minimum channels OPERABLE requirement f or over 30 days, an explanation of the circumstances shall be included in the next Annual Radioactive Ef fluent Release Report. ACTION 108 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 7I hours, provided:
- a. The offgas treatment system is not bypassed; and
- b. The offgas post-treatment monitor (Dll-K615) or the main stack monitor (D11-K600) is OPERABLE.
Otherwise, be in at least HOT STANDBY within 12 hcurs. If the number of channels OPERABLE remains less than required by the minimum channels OPERABLE requirement for over 30 days, an explanation of the circumstances shall be included in the next Annual Radioactive Ef fluent Release Report. { i l 3-4 Rev. 8, 1/94 l l l
Matet ODCM Gaseous Effluent P:nitoring : n s t r ure r.t a t io n Tacle 3-2. Radioactive Surveillance Requirements CHANNEL CHANNEL CHANNEL CALIBPA- FUNCTIONAL CHECK SOURCE CHECK TION TEST Instrument
- 1. Reactor Building Vent Stack Monitoring System (Each Unit)
- a. Noble Gas Activity c D, M R Q Monitor
- b. Iodine Sampler NA W 'd NA NA Cartridge
- c. Particulate Sampler NA W *d NA NA Filter
- d. Effluent System Flowrate Measuring Device D NA E O
- e. Sampler Flowrate D, NA R Q Measuring Device
- 2. Recombiner Building Ventilation Monitoring System
- a. Noble Gas Activity g ge Da g Monitor
- b. Iodine Sampler NA W"'d NA NA Cartridge
- c. Particulate Sampler NA W 'd NA NA Filter
- d. Sampler Flowrate Measuring Device Da NA R Q
- 3. Main Stack Monitoring System
- a. Noble Gas Activity g gs-Da g Monitor
- b. Iodine Sampler Cartridge d'd NA NA NA
- c. Particulate Sampler W*
- d NA NA NA Filter
- d. Effluent Flowrate D NA R Q Monitor
- e. Sarnpler Flowrate D, NA R O Monitor
- 4. Condenser Offgas Pretreatment Monitor (Each Unit)
- a. Noble Gas Activity Db C
M R 9 Monitor 3-5 Rev. 8, 1/94
Ha*.cn CDCM Tatie 3-2 (centd). Notati:n fer Taole 3-2
- a. Requirement applies during releases via this pathway.
- b. Requirement applies during operation of the main condenser air ejector.
- c. In addition to the basic functions of a CRANNEL FUNCTIONAL TEST (Section 10.2), the CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
(1) Instrument indicates measured levels above the alarm / trip setpoint. (2) Circuit failure occurs. (3) Instrument indicates a downscale failure,
- d. The CHANNEL CHECK shall censist of verif ying sampler flow and the presence of the collection device (i.e., particulate filter or charcoal cartridge, etc.) at the weekly changeout.
i l 4 I I l l 1 3-6 Fev. 8, 1/94
1 Hatch O?Qi 3.1.2 Gaseous Effluent Dose Pate Contr:. In accordance with Technical Specifications 6.18(3) and 6.18(7), the licensee shall conduct operations-so that the dose rates due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE , BOUNDARY (see Figure 10-1) are limited as follows:
- a. For noble gases: Less than or equal to a dose rate of 500 mrom/y to the total body and less than or equal to a dose rate of 3000 mrom/y to the skin, and ,
- b. For Iodine-131, Iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem /y to any organ.
3.1.2.1 Applicability This limit applies at all times. 3.1.2.2 Actions With a dose rate due to radioactive materia; released in gaseous effluents exceeding the limit stated in Section 3.1.2, immediately decrease the release rate to within the stated limit. When the ACTION statement or other requirements of this centrol cannot be met, steps need not be taken to change the Oparational Mode of the unit. Entry into an 0;.erational Mode or other specified CONDITION may be made if, as a minimum, the requiren.ents of the ACTION statement are satisfied. 3.1.2.3 Surveillance Requirements ; l l The dose rates due to radioactive materials in areas at or beyond the SITE BOUNDARY due to releases of gaseous ef fluents shall be determined to be within the above limits, . in accordance with the methods and procedures in Section 3.4.1, by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in rable 3-3. 3.1.2.4 Basis This control is provided to ensure that caseous effluent dose rates will be maintained within the limits that historically have provided reasonable assurance that radioactive material discharged in gasecas effluents will not result in a 3-7 Rev. 8, 1/94
-~__ - . , , . . ,
l l Hat:- ODCM i stce 3:se t a MEMBER OF THE PUBLIC in ar. UNRESTRICTE: AREA, eitner -athan cr : the SITE BOUNDARY, exceeding the limits specifted in Appendix I of 10 CFP Part 50, while allowing operational flexibility for effluent releases. For MIMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the j MEMBER OF THE PUBLIC will be suf ficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. l The does rate limit for Iodine-131, Iodine-133, tritium, and radionuclides in ! particulate form with half-lives greater than 8 days specifically applies to dose rates to a child via the inhalation path.ay. This control applies to the release of gaseous ef fluents from all reactors at the site. l l l I l 3-8 Rev. 3, 1/94 [ ___j
i 9 Hatch 00CM f I Table 3-3. Radica:tave Gase:as Waste Samp1.ng and Ana.fs:s Program Sampling and Analysis Requiremente a )
}
MINIMUM ! DETECTABLE Minimum Type of CONCMtTRA-Gaseous Release Sampling Analysis Activity TION (MDC) Type FREQUENCY FREQUENCY Analysis (pci/mL) PRINCIPAL GAMMA 1 E-4 ) MC gc EMITTERS 4 Grab Sample H-3 1 E-6 Wd I-131 1 E-12 Environmental Charcoal or i CONTINUOUS C Silver I-133 1 E-10 d Release Points Zeolite
- 1. Main Stack Sample M ^' ^"^
Build ng nt CONTINUOUSc Particulate Ej"T 1 E-11 (Each Unit) Sample M
- 3. Recombine c OS Building Vent CONTINUOUSe Gross Alpha 1 E-11 p e Sample Q
C Sr-89, Sr-90 1 E-11 CONTINUOUSe p S ample 3-9 pey, g, ljg4 l 1
Match ODCM Taole 3-3 (::ntc). Notaticn for Tacle 3-3 p-.,.
- a. Terms printed in all capital letters are defined in Chapter 10. When unusual circumstances result in a MINIMUM DETECTABLE CONCENTRATION higher than requited, the reasons shall be documented in the next Annual Radioactive Effluent Release Report,
- b. The Recombiner Building Vent serves Unit 1. Sample analysis results and l
associated source terms must be assigned to Unit 1 for the purpose of ; relaase accountability and dose calculations.
- c. Sampli: g and analyses for PRINCIPAL GAMMA EMITTERS shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of l
the RATED THERMAL POWER within a one-hour period. The more frequent l sampling and analysis requirement applies only if analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant and the Main Stack Noble Gas Activity Monitor reading have both increased by a factor of 3.
- d. Sampling shall be performed weekly, and analyses completed within 48 hours i of changing (or after removal from sampler). Sampling shall also be l performed once per 24 hours for 7 days following each shutdown, startup, I or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within 3 a one-hour period, with analyses completed within 48 hours of changing. I i
When samples collected for 24 hours are analyzed, the corresponding MINIMUM DETECTABLE CONCENTRATIONS may be increased by a f actor of 10. The more frequent sampling and analysis requirement applies only if analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant and the Main Stack Noble Gas Activity Monitor reading have both increased by a factor of 3. I
- e. The ratio of the sample flowrate to the sampled stream flowrate shall be l known for the time period covered by each dose or dose rate calculation j made in accordance with controls specified in Sections 3.1.2, 3.1.3, and 3.1.4.
l l l 3-10 Rev. 8, 1/94 i l i I
i Hatch OCCM 3.1.3 casecus E f f ' u. e r, A.r Ocse C o n t r e '. In accordance with Technical Specifications 6.18(5) and 6.18 ( 8 ) , the air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 10-1) shall be limited to the following: I Less than or equal to 5 mrad f or gamma ! a. During any calendar quarter: radiation and less than or equal to 10 mrad for beta radiation, and
- b. During any calendar year: Less than or equal to 10 mrad for gamma i
radiation and less than or equal to 20 mrad for beta radiation. 3.1.3.1 Applicability This ltmit applies at all times. l 3.1.3.2 Actions With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s); defines the corrective actions that have been taken to reduce the releases; and defines the proposed corrective actions to be taken to assure that subsequent releases of radioactive noble gases in gaseous effluents will be in compliance with the limits of Section 3.1.3. . l l When the ACTION statement or other requirements of this control cannot be met, steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified CONDITION may be made if, as a minimum, the requirements of the ACTION statement are satisfied. 3.1.3.3 Surveillance Requirements Cumulative air dose contributions from noble gas radionuclides released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.2 at least once per 31 days. 3-11 Rev. 8, 1/94
Hatch ODCM I 3.1.3.4 Basts This control is provided to implement the requirements of Sections II.8, III.A and IV. A of Appendix I, 10 CFR Part 50. Section 3.1.3 implements the guides set forth in Section II.B of Appendix I. The ACTION statements in Section 3.1.3.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I, assuring that the releases of radioactive material in gaseous ef fluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance requirements in Section 3.1.3.3 implement the requirements in Section III. A of Appendix I, which require that conformance with the guides of Appendix I be shown by calculational Procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in Section 3.4.2 for calculating the doses due to the actual releases of noble gases in gaseous ef fluents are consistent with the methodology provided in Regulatory Guide 1.109 l (Reference 3), and Regulatory Guide 1.111 (Reference 5). The equations in Section 3.4.2 provided for determining the air doses at the SITE BOUNDARY are based upon the historical annual average atmospheric conditions. l i t l l l I 3~12 Pev. 8, 1/94 I i
-- .. .- . ~ . ---- _ .. . - .
Hatch ODCM 3.1.4 Cc .t rcl on Gaseems E f fluent Dose t! a Me*:er :f the Puc.;c f In accordance with Technical Specifications 6.18(5) and 6.18(9), the dose to a MEMBER OF THE PUBLIC from I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 10-1) shall be limited to the followings
- a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ, and During any calendar year: Less than or equal to 15 mrem to any organ.
b. 3.1.4.1 Applicability ints limit applies at all times. 3.1.4.2 Actions With the calculated dose from the release of I-131, I-133, tritium, or radio-nuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limie.s, prepare and.suomit to the Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) f or exceeding the limit; defines the corrective actions that have been taken to reduce the releases of radioiodines and radionuclides in particulate f orm with half-lives greater than 8 days in gaseous effluents; and defines proposed corrective actions to assure f that subsequent releases will be in compliance with the limits stated in Section 3.1.4. When the ACTION scatement or other requirements of snas ccr. trol cannot be met, l steps need not be taken to change the Operational Mode of sne unit Entry into an Operational Mode or other specified CONDITION may be made if, as a minimum, the requirements of the ACTION statement are satisfied. 3.1.4.3 Surveillance Requirements Cumulative organ dose contributions to a MEMSER OF THE PUBLIC f rom I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days released in gaseous ef fluents f rom each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.3 at least once per 31 days. 3-13 Rev. 8, 1/94
_ . _ _ _ _ . . _ .. _ m.. - l Hat:P COCM 3.1.4.4 Basts This control is provided to implement the requirements of Section II.C, I!!. A and IV. A of Appendix I, 10 CFR Part 50. The limits stated in Section 3.1.4 are the guides set forth in Section II.C of Appendix I. The ACTION statements in Section 3.1.4.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous af fluents to UNRESTRICTED AREAS will be kept f "as low as is reasonably achievable. " The calculational methods specified in the Surveillance Requirements of Section 3.1.4.3 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 exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to l be substantially underestimated. The calculaticnal metheds in Section 3.4.3 f or calculating the doses due to the actual releases of the subject materials are consistent with the methodology provided in Regulatory G;;de 1.109 (Reference 3), and Regulatory Guide 1.111 (Reference 5). These equations provide for determining the actual doses based upon the historical annual average atmospheric conditions. The release specifications for radiciodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these i calculations were: 1) individual inhalation of aircorne radionuclides, 2) deposition of radionuclides onto green leafy garden vegetation with subsequent l consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of r.an. i I i I l 3-14 Rev. 8, 1/94 1 j
I l l , Hatch ODCM l
- ntrt; l
3.1.5 Gasecus Fai aste Tre st-ent 5.ste-In accordance with Technical Specification 6.18(6), the CASEOUS RADWASTE TREATHENT SYSTEM as described in Section 3.2 shall be in cperation. 3.1.5.1 Applicability Whenever the main condenser air ejector is in operation. 3.1.5.2 Actions With gaseous radwaste f rom the main condenser air e;ector system being discharged without treatment for more than 7 days, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a special Report which includes the following Information:
- a. Identification of the inoperable equipment or subsystem and the reason for j inoperability, l
l ! b. Action (s) taken to restore the inoperable equipment to OPERABLE status, and l I c. Summary descriptton of action (s) taken to prevent a recurrence. I f When the ACTION statement or etner requirements of this control cannot be met, steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified CONDITION may be made if, as a minimum, the requirements of the ACTION statement are s a*. i s f ied . 3.1.5.3 Surveillance Requirements The CASEOU$ RADWASTE TREATMENT SYSTEM shall be demonstrated to be OPERABLE by administrative controls which ensare that the offgas treatment system is not bypassed. 3.1.5.4 Basis The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM ensures that the system l will be available f or use whenever gaseces ef f luents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used, when specified, provides reasonable assurance that the l ' releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This control implements tne requirements of 10 CFR l ' 3-15 Fev. 8, 1/94
Hatch OCCM A: 10 CFR Part 50, and tne Part 50. 36a, Genera. Oesign Criter.:n 60 Of Appentx , The design objectives given in Section II.D of Appendix : .to 10 CFR Part 50. specified limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives set forth in f Section II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents. l This control applies to the release of radioactive materials in gaseous ef fluents f rom each unit at the site. For units with shared radwaste systems, the gaseous ef fluents f rom the shared system are proportioned among the units sharing that system. Systems 3.1.6 Maior Chances to Gaseous Radioactive Waste Treatment ( Licensee initiated MAJOR CHANGES TO GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEMS
- a. Shall be reported to the Nuclear Regulatory Commission in the Annual Radioactive Ef fluents Release Report for the period in which the change l was implemented, in accordance with Section 7.2.2.7.
i
- b. Shall become ef fective upon review and approval by the Plant Review Board, i
l l l l 1 l l 1 ! l l l 3-16 Rev. 8, 1/94 l t 1
I Hntch ODCM 3.2 GASEOUS RADWASTE TREATMENT SiSTEM At Plant Hatch, there are four points where radioactivity normally is released to the atmosphere in gaseous discharges. These four release pathways ares the Unit 1 and Unit 2 reactor building vent stacks; the Unit I recombiner building vent; and the main stack, which serves both units. In addition, releases may be made from any of the building exhaust augmented ventilation systems that have been included in Table 3-1, Table 3-2, and Table 3-3. The main stack serves as the discharge point from the following release sources from each unit: e Mechanical vacuum pumps; e offgas treatment system (see Figure 3-1); e Gland seal exhaust; and
- Standby gas treatment system (through which drywell purges are discharged).
In addition, the waste gas treatment building ventilation also discharges through the main stack. Each reactor building vent stack serves as the discharge point for the following release sources of its respective unit: e Reactor building; e Refueling floor ventilation; i e Turbine building; and e Radwaste building. The Unit 1 recombiner building vent disenarges directly to the atmosphere. Releases from all of the above discharge pcznts except the main stack are considered to be ground-level releases; releases from the main stack are considered to be elevated releases. Cnapter 8 discusses the calculation of 1 atmospheric dispersion parameters using One grcund-level and elevated models. All release pathways are considered to 'e CCNT:NUOUS c (as opposed to BATCH) in nature. t i I 3-;7 Fev. 8, 1/94 l _ . _ . . m -.- . . . , _ . , , , .- . . - _ . .. _ . . _ , , _ , . , . _ , . . . - . _ , _ . . . . . _ . _ _ . . . . . . , _ _ . , , _ . - . .
Ma: a. COOV F ROM CONDENSE R V 187 STAGE EXCTOR sst stAoE mTERCONDENsER 2NO BT AGE EJECTOR 2NO STAGE INTE RCONDENSE R h =
~M.J 340 ST AGE EACTOR OFFGASREMEATER j V
l CATALYTIC RECOMelNE R
, Y I' ) O FI CAS CONDE NSE R R ADI A TION y MONITOR O MATER SEPARATOR
' l HOLDUP LINE V
COOLER CONO EN$E R
' )
I
)9 MOtSTU R E ST RIP SEPARATOR HEATER ;
I , 1 Y CHARCOAL CH A RC OAL FILTE RS l FILTERS lePLACE5) ($ PL ACES) ,...A R ADIATION
} MONITOR 1
Y FROM
$TANOGY l
GAS AFTgm R ADI ATION TRg ATwt WT FILTgm teONITOR d SYST E M s . tau d , ROM 1 OLAND l l SEAL SYSTEM l Figure 3-1. Schematic Diagram of the Concenser Offgas Treatment System 3-18 Rev. 8, .!94 l - - _ - _ - _ _ - - _ - _ - _ _ _ _ . - - _ ~ ---
Hatch ODCM 3.3 GASEOUS EFFLUENT MONITO?. SETPOINTS 3.3.1 General Provisions Recardina Noble Gas Monitor Setpoints Noble gas radioactivity monitor setpoints calculated in accordance with the methodology presented in this section are intended to ensure that the limits of Section 3.1.2.a are not exceeded. They will be regarded as upper bounds for the actual . high alarm setpoints. That is, a lower high alarm setpoint may be established or retained on the monitor, if desired. Intermediate level setpoints should be established at an appropriate level to give suf ficient warning prior to reaching the high alarm setpoint. If no release is planned for a given pathway, or if there is no detectable activity in the gaseous stream being evaluated for release, the setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should a significant inadvertent release Occur. As established in Section 3.1.1, gaseous ef fluent monitor setpoints are required f only for the noble gas monitors on the release streams listed above. However, Section 3.3.6 discusses setpoint methodologies for particulate and iodine monitors on an information-only basis. [ Note: Section 3.3.3 is included for section numbering compatibility with the ODCMs of the other Southern Company nuclear power plants. This section is l i not required by the existing release source and discharge point I configuration of Plant Hatch. l l l l l 1 l l f l t 3-19 Rev. S, 1/94 i I {
I Hat:P. ODCM l Me r:0:logies Table 3-4. Applicacility of caseous Men.tcr Se p .- Final Release Pathways with no Monitored Source Streams Setpoint Method: Section 3.3.2 Release Type: CONTINUOUS l Main Stack Release Elevation Elevated l Monitor: Dll-K600 A and B l Maximum Flowrate: 20,000 cfm (9.44 E+06 mL/s) l Unit 1 or Unit 2 Reactor Buildino Vent Ground-level Release Elevation: Monitor: D11-K619 A and B / 2Dll-K636 A and B Maximum Flowrate: 300,000 cfm (1.42 E+08 mL/s) l Unit 1 Recombiner Buildino Vent l Release Elevation: Ground-level ( D11-P003 A and B l Monitor: I l Maximum Flowrate: 500 cfm (2.36 E-05 mL s; Buildina Exhaust Auomented Ventilation The systems in this category are not currently vented to the atmosphere.
)
Final Release Pathways with One or More Monitored Source Streams I l Plant Hatch currently has no release pathways in this category. 1 I l 1 l (i? Q),3 Values for Use in setpoint calculations 8.37 x 10 4 s/m 3 Ground-Level Releases: :ENE Sector] Elevated Releases: 4.10 x 10-8 s/m3 ENE Sector: l 3-20 Rev. 8, 1/94
_ . ~ ~ , _. . . _ - ~__-.- - - - .-- - - _ - - - _ f Hat e.- CDCM
,- 3.3.2 Setooint for the Final Noble Gas Monitor on Each Release Pathway =
3.3.2.1 Overview of Method caseous ef fluent radioactivity monitors are intended to alarm prior to' exceeding Therefore, their alarm setpoints are established the limits of Section 3.1.2.a. i to ensure compliance with the following equations c = the lesser of (3*1) AG SF = X
- Rx. ,
where: e= the setpoint, in uCi/mL, of the radioactivity monitor measuring the concentration of radioactivity in the effluent line prier te release. The setpoint represents a concentration whien, if exceeded, could result in dose rates exceeding the limits of Section 3.1.2.a at or beyond the SITE BOUNDARY. AG = an administrative allocation f actor applied to divide the release l limit among all the gaseous release pathways at the site. ; SF = the safety factor selected to compensate for statistical fluctuations and errors of measurement. X= the noble gas concentration for the release under consideration. I the ratio of the dose rate limit for tne total body, 500 mrem /y, j Rt = . to the dose rate to the tota' body for the conditions cf the release under consideration. the ratio of the dose rate limit f or the skin, 3000 mrem /y, to the Rk= dose rate to the skin for the conditions of the release under consideration. Equation (3.1) shows the relationships of the critical parameters that determine the setpoint. However, in order to apply the methodology presented in the equation to a mixture of noble gas radionuclides, radionuclide-specific concentrations and dose factors must be taken into account under conditions of maximum flowrate for the release point and annual average meteorology. 3-21 Rev. 8, 1/94
Hatch ODCM ine radioactivity The basic setpoint etnod presented celow is a p p '. ; ; s o l e t For monitors monitor nearest the point of release for the release pathway. measuring the radioactivity in source streams that merge with other streams prior to subsequent monitoring and release, the modifications presented in Section I l 3.3.3 must be applied. < ( 3.3.2.2 Setpoint calculation Steps gt;ep 1: Determine the concentration, X,y, of each noble gas radionuclide i in the gaseous stream v being considered for release, in accordance with the sampling and analysis requirements of Section 3.1.2. Then sum these concentrations to determine the total noble gas concentration, I X;y.
.i Steo 2: Determine R , the ratio of the dose rate limit for the total body, t
500 mrem /y, to the total body dose rate due to noble gases detected in the release under consideration, as follows,
- a. for release pathways for which the release elevation is ground-level:
=
500 R, T O) sb b [K
- O s-)
i (3 2) (X
- b. for release pathways f or which the release elevation is '
l i elevated;
=
500 R, (3 3) [ [Vi
- On]
a where: 500 = the dose rate limit for the total body, 500 mrem /y. (176)vb = the highest annual average relative concentration at the SITE BOUNDARY for the discharge point of release pathway v. Table 3-4 f includes an indication of what release elevation is applicable to each release pathway; release elevation determines the appropriate j value of (i]6)vb-3-22 Rev. 8, 1/94
Hatch ODCM gas it , = :ne total-Dcdy dose fart:r :.c :: ; sr.- 3 : .ss.:ns fr:m no .e radionuclide 1, in (mrem /y)/(9C1,.3), from Table 3-5. Q.= the release rate of noble gas radionucitde i from the release 3 pathway under consideration, in uCi/s, calculated as the product of Xjy and f,y, where: X,y = the concentration of noble gas radionuclide i for the particular release, in uCi/mL. f p.
= the maximum anticipated flowrate for release pathway v during the period of the release under consideration, in mL/s.
Vj = the elevated finite-plume total cody : se factor due to gamma emissions from noble gas radionuci de i an ef fluents released f rom the main stack, in (mrem /y)/(uCi/s), fr:= Table 3-6. Sten 3: Determine kR , the ratio of the dose rate ;;mit for the skin, 3000 mrem /y, to the skin dose rate due to n cle gases detected in the release under consideration, as follows,
- a. for release pathways for whlen tne release elevation is ground-level:
3000 p ,
- l l M,) 2,i l (3 4) 1 (27D ) vb [ '(E r > ,
i i
- c. for release pathways for wnt:r tne release elevation is elevated:
=
3000 RL ( *$} ( i [ L , ( X7D ) vb
- I 13 ; *s 9 s t >I 8
l l where: 1 3000 = the dose rate limit for the skin, 3000 mrem /y. l l 3-23 Rev. 8, 1/94 l l l l l
I Hatch Og,gh tne ektn.d se fa::or cue t: cet; e.- - e s s ;;n s f r:r. neo.e gas radto-
=
L, nuclide i, in (mrem /y)/ (uci/m 3
), from Table 3-5.
Mj = the air dose factor due to gamma emissions from noble gas radio-nuclide i, in (mrad /y) / (pci/m ), from Table 3-5. 3 1.1 = the factor to convert air dose in arad to skin dose in area. Bj = the elevated finite-plume air dose factor due to gamma emissions from noble gas radionuclide i in effluents released from the main stack, in (mrad /y)/(gC1/s), from Table 3-6. All other terms were defined previously. Steo 4: Determine the maximum noble gas radioactivity monitor setpoint con-centration. Based on the values determined in previous steps, the radioactivity monitor l setpoint for the planned release is calculated to ensure that the limits of Section 3.1.2.a will not be exceeded. Because the radioactivity monitor responds primarily to radiation f rom noble gas radionuclides, the monitor setpoint c oy (in pCi/mL) is based on the concentration of all noble gases in the waste stream, as ! follows: i AG,.= ST [ Xg R, I (3) i e m. = the lesser of AG,
- ST [ X ,,.
- R x.
I where: e nv = the calculated setpoint, in uCi/mL, for the noble gas monitor serving gaseous release pathway v. AG y = the administrative allocation factor for gaseous release pathway v, applied to divide the release limit among all the gaseous release pathways at the site. The allocation factor may be ; l assigned any value between 0 and 1, under the condition that the l sum of the allocation factors for all simultaneously-active final release pathways at the entire plant site does not exceed 1. l l 3-24 Rev. 8, 1/94 l i I
.. _. -- - -. .- -__.--~:.. . ~ - . _ . . .. . ._
Hatch ODCM Alternative metncas for ceterm.nat.:n of AG s are presented .n Section 3.3.4. SF = the safety factor selected to compensate for statistical fluctuations and errors of measurement. The value for the safety f actor must be between 0 and 1. A value of 0.5 is reasonable for gaseous releases; a more precise value may be developed if desired. X,y = the measured concentration of noble gas radionuclide i in gaseous stream v, as defined in Step 1, in pCi/mL. The values of g and Rk to be used in the calculation are those which were determined in Steps 2 and 3 above, l Step 5: Determine whether the release is permissible, as follows: If c nv 2 I X,y , the release is permissible. However, if c nv is within about i 10 percent of I Xiy , it may be impractical to use this value of env' I This situation indicates that measured concentrations are approaching values which would cause the limits of Section 3.1.2.a to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to adjust the allocation of the limits among the active release points. The setpoint calculatlons (steps 1-4) must then be repeated with parameters that reflect the modified conditions. If e nv < I X,y, the release may not be made as planned. Consider the 3 l alternatives discussed in the paragraph above, and calculate a new l i setpoint based on the results of tne actions taken, i i
- i l 3.3.2.3 Use of the calculated Setpoint l
l I ( The setpoint calculated above is in the units pCi/mL. The monitor actually measures a count rate that includes background, so that the calculated setpoint must be converted accordingly: c [, - ( c,n. E s. ) + B ,. (3.7) where: I 3-25 Rev. 8, 1/94 j l \
~ . _ . - - - -
.- - - -~_ .
t I r l Hatch ODCM f f c[ = tne menator setpc...; as a ccent r e. t e . Ey = the monitor calibration factor, in count rate /(pct /mL). Monitor data for conversion between count rate and ca? ."ation concentration may include operational data obtained from l determining the monitor response to ef fluent stream concentrations 1 measured by sample analysis. l the monitor . background count rate. In all cases, monitor By = background must be controlled so that the monitor is capable of l responding to concentrations in the range of the setpoint value. l contributions to the monitor background may include any or all of the following f actors a.~tbient background radiation, plant-related radiation levels at the monitor location (which may change between shutdown and power conditions), and i ternal background due to contamination of the monitor's sample chamber. The count rate units for en,E, and B iny equation (3.7 ) must be the same, cpm g or cps. l 3.3.3 Setooints for Noble cas Monitors on Ef fluent source Streams The listing n Table 3-4 shows that Plant Hatch currently has no gaseous release pathways that meet the following criteria: a setpoint as required for them under the effluent controls of this ODCM; and they are monitored prior to merging with cther streams, and passing a f:nal radioactivity mon: tor. This section, which presents a setpoint methodology for such monitors,1s included in the Plant Hatch ODCM for compatibility with the ODCMs of the other Southern Company nuclear power plants. 3.3.3.1 Setpoint of the Monitor on the Source Stream I Step 1: Determine the concentration X,s of each noble gas radionuclide i in source stream s (in , ./mL) according to the results of its required sample analyses ( see Section 3.1. 2 ] . Step 7: Determinet r , the ratio of the dose rate Izmit for the total body, 500 mrem /y, to the tctal body dose rate due to ncble gases detected in the source stream under consideration. Use the X 3 values and the maximum anticipated source stream flow rate f u in equation l 1 3-26 Rev. 8, 1/94 I J I i
l l 2 Hatch ODCM (3.3) the release :s elevates, t: j d (3.2) (or :n equatten .f 1 + determine the total body dose rate for the source stream. " substituting rt fCf $' > i ! 1 The SITE BOUNDARY relative dispersion value used in Steps 2 and 3 for the l source stream is the same as the (X76)sb that applies to the respective 1 merged stream. This is because the (176) value is determined by the 4 meteorology of the plant site and the physical attributes of the release point, and is unaffected by whether or not a given source stream is i j l f
- operating.
1 !, for the skin, 3000 Steo 3: Determine rk, the ratio of the dose rate limit mrem /y, to the skin dose rate due to noble gases detected in the f' source stream under consideratten. Use the X g values and the l 4 maximum anticipated source stream flow rate f,, in equation (3.4) + l (or in equation (3.5) if the release is elevated) to determine the I skin dose rate for the source stream, substituting rk f0f E'k i i l Steo 4: Determine the maximum noble gas radioactivit'. monitor setpoint con-centration, as follows: A, a AG,' SF [ X j, rg i
- ( '8) i c ns
- the lesser nf AG, SF [ X ,,
- rg i
i i j where: N c as = the calculated setpoint (in CL/mL) for the noole gas monitor 8 serving gaseous source stream s. 4
, AG, = the administrative allocation factor applied to gaseous source stream s. For a given final release point v, the sum of all the AG s values for source streams contributing to the final release point must not exceed the release point's allocation factor AGy. ! Xj, = the measured concentration of noble gas radionuclide i in gaseous source stream s, as defined in Step 1, in sci /mL.
i 1
' 3-27 Rev. 8, 1/94 1
5 '
Hatch ODCM The values of rt and rg t ce used in tne ca.:..at.:n are incse wn.cn were i det e rmined in Steps 2 and 3 above. The safety factor, SF, was defined previously. Step 5: Determine whether the release is permissible, as follows: However, if c m is within about If c ,a EX;,, the release is permissible. i 10 percent of IXis, it may be imprattical to use this value of c g. I that measured concentrations are ! This situation indicates approaching values which would cause the limits of Section 3.1.2.a to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to l adjust the allocation of the limits among the active release points. The setpoint calculations tsteps 1-4) must then be rcpeated with parameters that reflect the modified conditions. i l If c o3 < EXis , the release may not be made as planned. Consider the 1 alternatives discussed in the paragraph above, and calculate a new l setpoint based on the results of the actions taken. 3.3.3.2 Effeet on the Setpoint of the Monitor on the Merged Stream Before beginning a release from a monitored source stream, a setpoint must be determined for the source stream monitor as presented in Section 3.3.3.1. In addition, whether or not the source stream has its own effluent monitor, the previously-determined maximum allowable setpoint for the downstream final monitor l on the merged stream must be redetermined. This is acccrrplished by repeating the steps of Section 3.3.2, with the following modifications. Modification 1: The new maximum anticipated flowrcte of the merged stream is the sum of the old merged stream maximum flowrate, and the maximum flowrate of the source stream being considered for release. (favInew (felW
- f as (3 9)
Modification 2: The new concentration of noble gas radionuclide i in the merged stream includes both the contribution of the 3-28 Rev. 8, 1/94 l l I I
u- ~ s --.w... . . ~ . . ., -.a a. .m., nu. m . . , - . - . - . . . - . - - - . -- J a 4 Hatch OOCM 1 merged s' ream w;:i; cut e s;. :e st re am, ad tne scar:e l a stream being considered.for re' ease. . I = (fav)ou ' (xiv> ou + f as *x is (3.10) (xjy)new (f v)new a 1 i I AG ( 3,3.4 Determination of Allocation Far: tors. When simultaneous gaseous releases are conducted, an administrative allocation facter must be applied to divide the release limit among the active gaseous 7 i release pathways. This is to assur6 that the dose rate limit for areas at and e beyond the SITE BOUNDARY (see section 3.1.2) will n .t be exceeded by simultaneous l releases. The allocation factor for any pathway eay be assioned any value between 0 and 1, under the following two conditions:
- 1. The sum of the allocation factors for all simultaneously-active f2na; release paths at the plant sito may not exceed 1.
The sum of the allocation factors for all simultaneously-active source streams merging into a given final release pathway may not excead the allocation factor of that final release pathway. Any of the following three methods may be used to assign the allocation f actors , j to the active gaseous release pathways: ,
- 1. For ease of implementation, AGy may- be equal fer all release pathways:
AG y = 1 (3.11) N where: N= the number of simultaneously active gaseous release pathways.
- 2. AGy for a given release pathway may be selected based on an estimate of the portion of the total SITE BOUNDARY dose rate (from all simultaneous releases) that is contributed by the release pathway. During periods when a given building or release pathway is not subject to gaseous radioactive releases, it may be assigned an allocation factor of zero.
- 3. AG y for a given release pathway may be selected based on a calculation of the portion of the total SITE BOUNDARY dose rate that is contributed by the release pathway, as follows, 3-29 Rev. 8, 1/94
Hatch ODCM
- a. fcr ground-leve. re'. ease pcants:
(27 l it [t K 0sl s 4 i
=
A G ,. l N (3.12) f [ (Vi Dis) * [ (I7U)rb [IKi Ost) i r=1 s I i l b. for the elevated release point (main stack): i f [{Yi Qis )
=
A G '. N (3.13) ) E (Vi i Qs)
* [ { (XID)rb [ {K; Q,,) l i r=1 ' i l l
l l where: 1 1 (X/Q)yg = the annual average SITE BOUNDARY relative conceitration 1 applicable to the gaseous release pathway v for which the 3 allocation factor is being determined, in s/m . K3 = the total-body dose factor due to gamma emissions from noble gas radionuclide i, in (mrem /y)/(uc i/m 3 ), from Table 3-5. Q,y = the release rate of noble gas radienuclide : from release pathway
/, in uCi/s, caltulated as the product of X 3y and f,y, where X,y = the concentration of noble gas radionuclide i applicable to the gaseous release pathway v for which the allocation f actor is being determined, in uCi/mL. ,
1 f,y = the discharge flowrate applicable to gaseous release pathway v for which the allocation factor is being determined, in mL/s. Note: As applied in equations (3.12) and (3.13), Q,y is restricted to ground-level release pathways. V = the elevats d finite-plume total body dose factor doe to gamma emirsions from noble gas radionuclide i in ef fluents released f rom the main stack, in (crem/y)/(yci/s), from Table 3-6. 3-30 Rev. 8, 1/94
Hatch 00CM Q3 = ne release rate of ner.e gas ca:::.ac..:e ; from the main sta:4, in uCi/s, calculated as the product of the X n. and f,s values J specific to the main stack. (X/Q)4 = the annual average SITE BOUNDARY relative concentration applicable to active gaseous release pathway r, in s/m 3. Qir = the value of Q,y applicable to active release pathway r, in yci/s. N= the number of simultanNusly active gaseous release pathways (including pathway v that is of interest). NOTE: Although equations (3.11), (3.12), and (3.13) are written to l illustrate the assignment of the allocation factors for final release pathways, they may also te ased to assign allocation f actors to the source streams that merge .nto a given final release pathway. 1 3.3.5 Setooints f or Noble Gas Monitors with S ecial Recuirements Unit 1 Condenser Of foas Pretreatment Moniter Monitor: 1D11-K601 and 1D11-K602 Unit 2 Condenser Offaas Pretreatment Monitor Monitor: 2D11-K601 and 2D11-K602 For the purpcse of implementing Section 3.1.1, the alarm setpoint level for these noble gas monitors will be calculated as follows: 5 2 . 4 0 x 10
,nco ,
g CO , ' , CU (3 l4 where: 2.40 x 105 = the release rate limit for pretreatment condenser offgas as specified in Technical Specifications 3.15.2.7 (Unit 1) and 3.11.2.7 (Unit 2), in uCi/s. e nco = the reading of the condenser offgas pretreatment monitor at the alarm setpoint, in mR/h. E co
= the calibration factor for the condenser offgas pretreatment monitor, in (uci/s) per (cfm mR/h).
3-31 Rev. 8, 1/94
l Hatch COCM f ,.g = ne c:ndenser offgas f ' c w r a *. e , . . - rfi. 3.3.6 Setooints for Particulatt and Iodine Monitors In accordance with Section 5.1.1 of NRC NUREG-0133 (Reference 1), the effluent controls of Section 3.1.1 do not require that the ODCM establish setpoint calculation methods for particulate and iodine monitors, i i i I l \ l i l l 3-32 Rev. 8, 1/94
Hatch ODCM i 3.4 GA3E005 EFFLUFNT CO.*F ANCE CA'CULAT;cN3 4 3.4.1 Dose Rates at and Beyond the Site Boundary 4 T
- Because the dose rate limits for areas at and beyond the SITE specified in j
Section 3.1.2 are site limits applicable at any instant in time, the summations 1 extend over all simultaneously active gaseous final release pathways at the planc i site. Table 3-4 identifies the gaseous final release pathways at the plant site, l 1 l and indicates the (176)vb I# (57DIsb) value for each. 3.4.1.1 Dose Rates Due to Noble Cases 1 j For the purpose of implementing the controls of Section 3.1.2.a, the dose rates due to noble gas radionuclides in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows: 4 For total body dose rates:
+
I DR g = { (Tfp ) g.b b (E l O ') it b lYi Oss ) (3.15) V i
- I I
i I For skin dose rates: 4 DRg e { f(xjg)sh b {(E l
- 1 15 ) Oss) 4 v( i i (3.16) 1' l
{ (X/Q)sh Li + 1.13; O j, j i 2 l where: i DRt = the total body dose rate at the time of the release, in mrem /y. 'l DRk= the skin dose rate at the time of the release, in mrem /y. Q.= the release rate of noble gas radionuclide 1, in pCi/s, equal to 3 i j the product of f tv and Xiy, where: f,= n the actual average flowrate for release pathway v during 1
' the period of the release, in mL/s.
j 3-33 Rev. 8, 1/94 J k i r e -.,-. e , , , , , . ->-m-- - . - , , -
Hat r. 00cM
- 3. 6, nc defan.tten :: t ;, , ,
Note: For equattens (J.153 a r.: . and the summations over v, are restricted to ground-level release pathways. Q;3 = the release rate of noble gas radionuclide i from the main stack, in yci/s, equal to the product of the f tv and Xgy values specific to ' the main stack. (This definition applies to both equations (3.15) and (3.16).) l (X/Q)sb = the value of (X/Q)vb for the main stack; that is, the highest annual average relative concentratien at the SITE BOUNDARY, for the main stack, in s/m3 . Table 3-4 includes the value of (x76)sb' i All other terms were defined previously. 3.4.1.2 Dose Rates Due to Iodine-131, Iodine-133 Tritium, and Radionuclides in Particulate Form with Half-Lives Greater than 8 Days For the purpose of implementing the controls of Section 3.1.2.b, the dose rates due to Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form , with half-lives greater than 8 days, in areas at or beyond the SITE BOUNDARY, due l to releases of gaseous effluents, shall be calculated as follows: ( 'l } DR g = [v (X7D) tb { Pjo O , + ( RT ) sb b #oO i ss
- i 8 l
where: DRo = the dose rate to organ o at the time of the release, in mrem /y. P;g = the site-specific dose factor for radionuclide i and organ o, in (mrem /y)/(pci/m3 ). Since the dose rate limits specified in Section 3.1.2.b apply only to the child age group exposed to the inhalation for pathway, the values of Pjo may be obtained f rom Table 3-9, " R,; Inhalation Pathway, Child Age Group. " Q{y= the release rate of radionuclide i f rom gaseous release pathway v, l in pCi/s. For the purpose of implementing the controls of Section 3.1.2.b, only I-131, I-133, tritium, and all radionuclides in 3-34 Rev. 8, 1/94
Hatch ODCM partteulate form w;tn nalf-lives greater inan 5 days s r. : a . : ce included in this calculation. All other terms were defined previously. 3.4.2 Noble Gas Air Dose at or Beyond Site Boundary For the purpose of implementing the controls of Section 3.1.3, air doses in areas at or beyond the SITE BOUNDARY due to releases of noble gases from each unit shall be calculated as follows (adapted from Reference 1, page 28, by including , i only long-term releases): l l l l Dy - 3.17 x 10'8 {' (m)g { (N, . dy l v i (3.18) (MIsb b (Ni* isI i i l l l l i . I D)
= 3.17 x 10 ~8 !{ WOg { Ly . dy lv i (3.19)
[ fB; 6,3: i where: 3.17 x 10-8 = a units conversion f acter: 1 y : ; 3 . '. 5 x 10 7 s). Dg = the air dose due to beta emissions from noble gas radionuclides, in mrad. Dy = the air dose due to gamma emissions from noble gas radionuclides, in mrad. Nj = the air dose factor due to beta emissions from noble gas radio-3 nuclide i, in (mrad /y)/(pci/m ), from Table 3-5. 3-35 pev, 8, 1/94
. - . . . - . . _ . . , . _ _ . . . , . _ . , _ _ . _ _ ~ . _ _ . . _ ,,..- . .
Hatch 00CM ts, the higr.est t i' D ) sb = the value of (iS D , s b fer :ne ma;n sta:<; tnat annual average relative concentratton at the SITE BOUNDARY, for the main stack, in s/m 3. Table 3-4 includes the value of (576)sb' K, = the air dose factor due to gamma emissions from noble gas radio-3 nuclide i, in (mrad /y)/ (pci/m ), from Table 3-5. f the elevated finite-plume air dose factor due to gamma emissions Bi= l from noble gas radionuclide i in ef fluente released from the main l stack, in (mrad /y)/(uci/s), from Table 3-6. I 1 Q,y = the cumulative release of noble gas radionuclide i from non- l elevated release pathway v, in pCi, during the period of interest. 63 = the value of 6g for the main stack; that is, the cumulative release of noble gas radionuclide i f rom the main stack, in uCi, during the period of interest. and all other terms are as defined above. Because the air dose limit is on a per-reactor-unit casis, the summations extend over all gaseous final release pathways (other than the main stack, which has its own term) for a given unit. For a release pathway discharging materials originating in both reactor units, the activity discharged f rom the release point may be apportioned to the two units in any reasonable manner, provided that all activity released via the particular shared release pathway is apportioned to one or the other unit. l The gaseous final release pathways at the plant site, and the (2/DJsb for each, are identified in Table 3-4. l 3-36 nev. 8, 1/94
Hntch ODCM Table 3-5. Dese Factors f or Exposare to a Se . 'nf antte Cloud of Ncc;e Gases y - Body (K) $ - Skin (L) y - Air (M) $ - Air (N) Nuclide (mrem /y) per (mrem /y) per (mrad /y) per (mrad /y) per (yci/m 3) (pci/m 3) (pci/m 3) ( Ci/m3) 7.56 E-02 0.00 E+00 1.93 E+01 2.88 E+02 Kr-83m 1.17 E+03 1.46 E+03 1.23 E+03 1.97 E+03 Kr-85m 1.61 E+01 1.34 E+03 1.72 E+01 1.95 E+03 Kr-85 5.92 E+03 9.73 E+03 6.17 E+03 1.03 E+04 Kr-87 Kr-88 1.47 E+04 2.37 E+03 1.52 E+04 2.93 E+03 1.66 E+04 1.01 E+04 1.73 E+04 1.06 E+04 Kr-89 1.56 E+04 7.29 E+03 1.63 E+04 7.83 E+03 Kr-90 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 Xe-131m 2.51 E+02 9.94 E+02 3.27 E+02 1.48 E+03 Xe-133m 2.94 E+02 3.06 E+02 3.53 E+02 1.05 E+03 Xe-133 Xe-135m 3.12 E+03 7.11 E+02 3.36 E+03 7.39 E+02 Xe-135 1.81 E+03 1.86 E+03 1.92 E+03 2.46 E+03 Xe-137 1.42 E+03 1.22 E+04 1.51 E+03 1.27 E+04 Xe-138 8.83 E+03 4.13 E+03 9.21 E+03 4.75 E+03 Ar-41 8.84 E+03 2.69 E+03 9.30 E+03 3.28 E+03 All values in this table were obtained from Reference 3 (Table B-1), with units converted. j l l I I l l i l
)
3-37 Rev. 8, 1/94
I l Hatch ODCM Table 3-6. E 'se Factors for Exposure t Oirect Radaatton from Nocle Cases in an Elevated Finite Plume y - Total Body (V) y - Air (8) Nuclide (mrom/y) per (uci/s) (mrad /y) per (uci/s) Kr-83m 0.00 E-00 0.00 E-00 l Kr-85m 8.25 E-05 8.69 E-05 i Kr-85 1.26 E-06 1.35 E-06 Kr-87 4.40 E-04 4.59 E-04 I Kr-88 1.09 E-03 1.13 E-03 Kr-89 9.44 E-04 9.87 E-04 i Kr-90 7.00 E-04 7.38 E-04 Xe-131m 1.68 E-06 1.76 E-06 Xe-133m 1.29 E-05 1.37 E-05 Xe-133 1.38 E-05 1.43 E-05 Xe-135m 2.42 E-04 2.59 E-04 Xe-135 1.33 E-04 1.42 E-04 Xe-137 9.55 E-05 1.02 E-04 Xe-138 6.16 E-04 6.44 E-04 Ar-41 7.34 E-04 7.72 E-04 Values are as reported in Reference 24. They were calculated in accordance with Reference 1 (Section 5.2.1) and Ref erence 3 ( Appendix F), using the meteorological joint frequency distributions presented in Reference 14. All values in this table are f or the Site Boundary (1545 m) in the ENE sector. 3-38 Rev. 8, 1/94
~~.w..a, n,.n,- . __ - ...~..a + w.-- w,- -n.. ,,. _.a.u - - . --. r l
r Hatch ODQ l l l 3.4.3 Dose to a Member of the Public at or Beyond Site Boundarv ) l The dose received by an individual due to gaseous releases from each reactor l unit, to areas at or beyond the SITE BOUNDARY, depends on the individual's location, age group, and exposure pathways. The MEMBER OF THE PUBLIC expected to receive the highest dose in the plant vicinity is referred to as the controlling receptor. The dosissetrically-significant attributes of the currently-defined controlling receptor are presented in Table 3-7. Doses to a MEMBER OF THE PUBLIC due to gaseous releases of I-131, I-133, tritium, and all radionuclides in particulate form from each unit shall be calculated as follows (equation adapted from Reference 1, page 29, by considering only long-term releases): Ra jp; [ l ygp
- N,?y (3.30)
Dja' " 3.17 x 10 [ v p l where Dj, = the dose to organ j of an individual in age group a, due to gaseous j releases of I-131, 1-133, tritium, and all radionuclides in in mrem. particulate f orm with half-lives greater than 8 days, 7 3.17 x 10-8 = a units conversion f actor: 1 y/(3.15 x 10 m). radionuclide 1, R aipj = the site-specific dose factor for age group a, exposure pathway p, and organ j. For the purpose of implementing the controls of Secticn 3.1.4, the expcsure pathways applicable to calculating the dose to the currently-defined controlling receptor are included in Table 3-7; values of R,,pj for each exposure pathway and radionuclide applicable to calculations of dose to the controlling receptor are listed in Table 3-8 through Table 3-10. A detailed discussion of the methods and parameters used for 9. calculating Rjj ap for the plant site is presented in Chapter That information may be used for recalculating the Rapj ; values if the underlying parameters change, or for calculating Raipj values 3-39 Rev. 8, 1/94 i I
- _ _ - . . - _ . . _ , ~ . . -. _ _ -
__.. ~ - . - - - - .- - . - . . - . - . . - . ~ . . - . - . - - - _ - . Hatch OCCM i l for special rad.:nuclides ant age groups when perterming the l assessments discussed in Section 3.4.4 below. Wy ;p = the annual average relative dispersion or deposition at the location of the controlling receptor, for release pathway v, as appropriate to exposure pathway p and radionuclide i. I For all tritium pathways, and for the inhalation of any radio- , 1 nuclide Wy ;p is (X76)yp, the annual average relative dispersion factor for release pathway v, at.the location of the controlling receptor (s/m ) . 3 For the ground-plane exposure pathway, and for all ingestion-related pathways for radionuclides other than tritium: Wy ;p is (67D)yp, the annual average relative deposition factor for release pathway v, at the location of the controlling
~
receptor (m-2). Values of (x i)sp and (676)sp for use in j calculating the dose to the currently-defined controlling receptor are included in Table 3-7. l' 6ly= the cumulative release of radionuclide i from release pathway v, during the period of interest (pCi). For the purpose of j 3.1.4, only I-131, I-133, implementing the controls of Section l J tritium, and all radionuclides in particulate form with half-lives l greater than 8 days should be included in this calculation. In any l dose assessment using the methods of this sub-section, only radio-nuclides detectable above background in their respective s amples should be included in the calculation. Because the MEMBER OF THE PUBLIC dose limit is on a per-unit basis, the summations extend over all gaseous final release pathways for a given unic. For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may se apportioned between the two units in any reasonable manner, provided that all activity released from the plant site is apportioned to one or the other unit. The gaseous final release pathways at the plant site, and the release elevation for each, are identified in Table 3-4. 3-40 Rev. 8, 1/94
Hatch ODCM l l !ac;e 3-7. Attributes of tne Controlling Receptor The locations of members of the public in the vicinity of the plant site, and the exposure pathways associated with those locations, were determined in the Annual Land Use census (Reference 12). Dispersion and deposition values were calculated based on site meteorological data collected for the period 1984 through 1986 (Reference 16). Based on an analysis of this information, the current controlling receptor for the HNP site is described as follows. Sectors wsw Distance: 1.2 miles 1 l l Ace Group: Child Exposure Pathways Ground plane Inhalation Garden vegetation Discersion Factors ( X76) g.p: 3.18 x 10 4 s/m 3 Ground-Level: Elevated: 6.53 x 10-8 s/m 3 Deposition Factors (D76)yp: Ground-Level: 8.80 x 10-9 m'2
*2 Elevated: 1.37 x 10-9 m Elevated Plume Dose Factors +
V V, 3 Radionuclide Radionuclide (mrem /y)/(uci/s) (mrem /y)/(uCi/s) l Kr-85m 6.92 E-05 Xe-133m 1.07 E-05 Kr-85 9.95 E-07 Xe-133 1.23 E-05 Kr-87 3.36 E-04 Xe-135m 1.90 E-04 Kr-88 8.23 E-04 Xe-135 1.09 E-04 Kr-89 7.20 E-04 Xe-137 7.54 E-05 Kr-90 5.39 E-04 Xe-138 4.71 E-04 Xe-131m 1.42 E-06 Ar-41 5.59 E-04
+ These values were calculated using the methods and data described in Reference 29. They are necessary when performing calculations for the purpose of demonstrating compliance with the limits of section 5.1.
3-41 Rev. 8, 1/94 j
l Hatch OCCM 3.4.4 Ocse Calculations te Sucecrt Other Require-e-te Case 1: Under Technical Specification 6.6.1, a radiological impact assess-ment may be required to support evaluation of a reportable event. Dose calculations may be perf ormed using the equations in Section 3.4.3, with the substitution of the dispersion and deposition parameters [(X/Q) and using the appropriate and (D/Q)) for the period covered by the report, Methods for pathway dose factors (Raipj) for the receptor of interest. calculating (X/Q) and (D/Q) from meteorological data are presented iii Chapter 8. The values of R aipj presented in Table 3-8 through Table 3-10 are applicable only to the currently-defined controlling receptor, so that when dose calculations must be per f ornied for a different receptor, R,jg values Methods and applicable to that receptor must first be calculated. parameters for calculating Raipj f or radionuclides and age groups other than
- 9. When those required in Section 3.4.3 are presented in Chapter calculating Rjj ap for evaluation of an event, pathway and usage factors specific to the receptor involved in the event may be used in place of the values in Chapter 9, if the specific values are known.
l t case 2: A dose calculation is required to evaluate the results of the Land ' Use Census, under the provisions of Section 4.1.2. In the event that the Land Use Census reveals that exposure pathways have changed at previously-identified locations, or if new locations are identified, it may be necessary to calculate doses at two or more locations to determine which should be designated as the controlling receptor. Such dose calculations may be performed using the equations in Section 3.4.3, with the substitution cf the annual average dispersion and l I deposition values [ ( X'/D) and (I)~/ Q) ] for the locations of interest, and using the appropriate pathway dose factors ( R,,g ) for the receptors of interest. Methods for calculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8. The values of Raipj preser.ted in Table 3-8 through Table 3-10 are applicable only to the currently-defined controlling receptor, so that when dose calculations must be perf ormed for a dif f erent l receptor, R,jp ; values applicable to that receptor must first be calculated. Methods and parameters for calculating R ain for radionuclides and age l 3-42 Rev. 8, 1/94 i l
.~, . . . , . .
l . . . .
l Hater ODCM I groups otner than those required in Sect.:- 3.4,3 are presented an C.. apter 9. Case 3: Under Section 5.2, a dose calculation to required to support determination of total dose to a receptor of age. group other than that currently defined as the controlling receptor. Dose calculations shall be performed using the equations in Section 3.4.3, using the . dispersion and deposition parameters defined in Table 3-7 for the controlling receptor, but substituting the appropriate pathway dose factors (Rupj) for the receptor age group of interest. The values of R,;p; presented in Table 3-8 through Table 3-10 are applicable r only to the currently-defined controlling receptor, so that when dose l calculations must be performed for a different receptor age group, R ag values applicable to that receptor must first be calculated. Methods and parameters for calculating R,,p) for radionuclides and age groups other than those required in Section 3.4.3 are presented in chapter 9. l l t i l l i l l . 3-43 Rev. 5, 1/94
Hatch OOCM Table 3-8. Ru n, for Ground Plane Pathway. A.. A;e Groups T. Body Skin __Nuclide H-3 0.00 0.00 l C-14 0.00 0.00 P-32 0.00 0.00 Cr-51 4.66E+06 5.51E+06 Mn-54 1.39E+09 1.63E+09 Fe-55 0.00 0.00 Fe-59 2.73E+08 3.21E+08 Co-58 3.79E+08 4.44E+08 Co-60 2.15E+10 2.53E+10 l Ni-63 0.00 0.00 Zn-65 7.47E+08 8.59E+08 Rb-86 8.99E+06 1.03E+07 Sr-89 2.16E+04 2.51E+04 Sr-90 0.00 0.00 Y-91 1.07E+06 1.21E+06 Zr-95 2.45E+08 2.84E+08 Nb-95 1.37E+08 1.61E+08 Ru-103 1.08E+08 1.26E+08 Ru-106 4.22E+08 5.07E+08 Ag-110m 3.44E+09 4.01E+09 Sb-124 5.98E+08 6.90E*08 Sb-125 2.34E+09 2.64E+09 I Te-125m 1.55E+06 2.13E+06 Te-127m 9.16E+04 1.08E+05 Te-129m 1.98E+07 2.31E+07 I-131 1.72E+07 2.09E+07 I-133 2.45E+06 2.98E+06 Cs-134 6.86E+09 8.00E+09 Cs-136 1.51E+08 1.71E+08 Cs-137 1.03E+10 1.20E+10 l l Ba-140 2.05E+07 2.35E+07 l 1.54E+07 ' Ce-141 1.37E+07 Ce-144 6.95E+07 8.04E+07 Pr-143 0.00 0.00 Nd-147 8.39E+06 1.01E+07
- 1. Units are m .2 (mrem /yr) / (pci/s) .
- 2. The values in the Total Body column also apply to the Bone, Liver, Thyroid, Kidney, Lung, and GI-LLI organs.
- 3. This table also supports the calculations of Section 6.2.
l l 3-44 Rev. 8, 1/94 l l l
Hatch ODCM Taole 3-9. R,g for Inhalation Patnway. Chaid Age Group Liver T. Body Thyroid Kidney Lung GI-LLI Nuclide Bone H-3 0.00 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 C-14 3.59E+04 6.73E+03 6.73E+03 6.73E+03 6.73E+03 6.73E+03 6.73E+03 0.00 0.00 4.22E+04 P-32 2.60E+06 1.14E+05 9.88E+04 0.00 Cr-51 0.00 0.00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 Mn-54 0.00 4.29E+04 9.51E+03 0.00 1.00E+04 1.58E+06 2.29E+04 0.00 1.11E+05 2.87E+03 Fe-55 4.74E+04 2.52E+04 7.77E+03 0.00 0.00 1.27E+06 7.07E+04 Fe-59 2.07E+04 3.34E+04 1.67E+04 0.00 0.00 1.77E+03 3.16E+03 0.00 0.00 1.11E+06 3.44E+04 Co-58 0.00 1.31E+04 2.26E+04 0.00 0.00 7.07E+06 9.62E+04 Co-60 0.00 2.75E+05 6.33E+03 Ni-63 8.21E+05 4.63E+04 2.80E+04 0.00 Zn-65 4.26E+04 1.13E+05 7.03E+04 0.00 7.14E+04 9.95E+05 1.63E+04 0.00 0.00 7.99E+03 Rb-86 0.00 1.98E+05 1.14E+05 0.00 5.99E+05 0.00 1.72E+04 0.00 0.00 2.16E+06 1.67E+05 Sr-89 1.01E+08 0.00 6.44E+06 0.00 0.00 1.48E+07 3.43E+05 Sr-90 9.14E+05 0.00 2.44E+04 0.00 0.00 2.63E+06 1.84E+05 Y-91 Zr-95 1.90E+05 4.18E+04 3.70E+04 0.00 5.96E+04 2.23E+06 6.11E+04 Nb-95 2.35E+04 9.18E+03 6.55E+03 0.00 8.62E+03 6.14E+05 3.70E+04 Ru-103 2.79E+03 0.00 1.07E+03 0.00 7.03E+03 6.62E+05 4.48E+04 Ru-106 1.36E+05 0.00 1.69E+04 0.00 1.84E+05 1.43E+07 4.29E+05 Ag-110m 1.69E+04 1.14E+04 9.14E+03 0.00 2.12E+04 5.48E+06 1.00E+05 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 6.73E+03 2.33E+03 9.14E+02 1.92E+03 0.00 4.77E+05 3.38E+04 Te-127m 2.49E+04 8.55E+03 3.02E+03 6.07E+03 6.36E+04 1.48E+06 7.14E+04 Te-129m 1.92E+04 6.85E+03 3.04E+03 6.33E+03 5.03E+04 1.76E+06 1.82E+05 I-131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.8BE+04 0.00 2.84E+03 3.85E+06 3.38E+04 0.00 5.48E+03 I-133 1.66E+04 2.03E+04 7.70E+03 Cs-134 6.51E+05 1.01E+06 2.25E+05 0.00 3.30E+05 1.21E+05 3.85E+03 Cs-136 6.51E+04 1.71E+05 1.16E+05 0.00 9.55E+04 1.45E+04 4.18E+03 Cs-137 9.07E+05 8.25E+05 1.28E+05 0.00 2.82E+05 1.04E+05 3.62E+03 Ba-140 7.40E+04 6.48E+01 4.33E+03 0.00 2.11E+01 1.74E+06 1.02E+05 Ce-141 3.92E+04 1.95E+04 2.90E+03 0.00 8.55E+03 5.44E+05 5.66E+04 Ce-144 6.77E+06 2.12E+06 3.61E+05 0.00 1.17E+06 1.20E+07 3.89E+05 Pr-143 1.85E+04 5.55E+03 9.14E+02 0.00 3.00E+03 4.33E+05 9.73E+04 Nd-147 1.0BE+04 8.73E+03 6.81E+02 0.00 4.81E+03 3.28E+05 8.21E+04 I l l 1. Units are (mrem /yr)/ (pci/m 3
) for all radionuclides.
- 2. This table also supports the calculations of Section 6.2.
3-45 Rev. 8, 1/94
l l Hatch ODCM l ! Table 3-10. P , for Garden Vegetation Patn.af, Cnt;d Age Group l T. Body Thyroid Kidney Lung GI-LLI Nuclide Bone Liver 4.01E+03 4.01E+03 4.01E+03 4.01E+03 H-3 0.00 4.01E+03 4.01E+03 1.78E+08 C-14 8.89E+08 1.78E+08 1.78E+08 1.78E+08 1.78E+08 1.78E+08 0.00 0.00 0.00 9.31E+07 P-32 3.37E+09 1.58E+08 1.30E+0B 0.00 1.17E+05 6.50E+04 1.78E+04 1.19E+05 6.21E+06 Cr-51 0.00 0.00 1.86E+08 0.00 5.58E+08 Mn-54 0.00 6.65E+08 1.77E+08 0.00 0.00 2.40E+08 7.87E+07 Fe-55 8.01E+08 4.25E+08 1.32E+08 0.00 0.00 1.86E+08 6.70E+08 Fe-59 3.98E+08 6.43E+08 3.20E+08 0.00 0.00 3.76E+08 Co-58 0.00 6.44E+07 1.97E+08 0.00 0.00 0.00 2.10E+09 co-60 0.00 3.78E+08 1.12E+09 0.00 0.00 0.00 0.00 1.42E+08 Ni-63 3.95E+10 2.11E+09 1.34E+09 0.00 1.36E+09 0.00 3.80E+08 Zn-65 8.13E+08 2.16E+09 1.35E+09 0.00 4.52E+08 2.78E+08 0.00 0.00 0.00 2.91E+07 Rb-86 0.00 1.03E+09 0.00 0.00 0.00 1.39E+09 Sr-89 3.60E+10 0.00 3.15E+11 0.00 0.00 0.00 1.67E+10 Sr-90 1.24E+12 0.00 4.99E+05 0.00 0.00 0.00 2.48E+09 Y-91 1.86E+07 0.00 1.21E+06 0.00 8.85E+08 Zr-95 3.86E+06 8.48E+05 7.55E+05 4.10E+05 1.60E+05 1.14E+05 0.00 1.50E+05 0.00 2.96E+08 Hb-95 1.53E+07 0.00 5.90E+06 0.00 3.86E+07 0.00 3.97E+08 I Ru-103 7.45E+08 0.00 9.30E+07 0.00 1.01E+09 0.00 1.16E+10 Ru-106 2.17E+07 1.73E+07 0.00 4.04E+07 0.00 2.58E+09 Ag-110m 3.21E+07 l Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 l ! sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 3.51E+08 9.50E+07 4.67E+07 9.84E+07 0.00 0.00 3.38E+08 Te-125m 1.32E+09 3.56E+08 1.57E+08 3.16E+08 3.77E+09 0.00 1.07E+09 Te-127m 8.41E+08 2.35E+08 1.31E+08 2.71E+08 2.47E+09 0.00 1.03E+09 Te-129m I-131 1.43E+08 1.44E+08 8.17E+07 4.75E+10 2.36E+08 0.00 1.28E+07 I-133 3.53E+06 4.37E+06 1.65E+06 8.11E+08 7.28E+06 0.00 1.76E+06 Cs-134 1.60E+10 2.63E+10 5.55E+09 0.00 8.15E+09 2.93E+09 1.42E+08 co-136 8.24E+07 2.27E+08 1.47E+08 0.00 1.21E+08 1.80E+07 7.96E+06 Cs-137 2.39E+10 2.29E+10 3.38E+09 0.00 7.46E+09 2.68E+09 1.43E+08 Ba-140 2.77E+08 2.42E+05 1.61E+07 0.00 7.89E+04 1.45E+05 1.40E+08 Ce-141 6.56E+05 3.27E+05 4.86E+04 0.00 1.43E+05 0.00 4.08E+08 Ce-144 1.27E+08 3.98E+07 6.78E+06 0.00 2.21E+07 0.00 1.04E+10 Pr-143 1.46E+05 4.37E+04 7.23E+03 0.00 2.37E+04 0.00 1.57E+08 Nd-147 7.15E+04 5.79E+04 4.48E+03 0.00 3.18E+04 0.00 9.17E+07 l Units are (mrem /yr)/(uc i/m3 ) for tritium, and m *2 (mrem /yr)/(pci/s) for all other radionuclides. l l 3-46 Rev. 8, 1/94 i
. _ . _ . , . . _ . .~ - - . . . _ - - ..- - - .
Haten ODCM 3.5 GASEOUS EFF'.UENT DOSE PRC.*ECT!ONS 3.5.1 Thirtv-One Day Dose Proiections Because continuous operation of the gaseous radwaste treatment system is required (see Section 3.1.5), routine 31-day dose projections are not required for l effluent control compliance at Plant Hatch. However, whenever it is desired to perform such projections, projected 31-day air doses and doses to individuals due to gaseous af fluents may be determined as follows: For air doses:
' Dg('
Ogp
- x 31 +
Oga
'3 (3.21)
<p ,
D yp *
- 31
- D ya l
For individual doses: l f p 5 I* I 1 D og = x 31 + D on l where: for Dgp = the projected air dose due to beta emissions f rom noble gases, the next 31 days of gaseous releases. Dgg
= the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.
Dg, a the anticipated air dose due to beta emissions from nobit gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous ef fluents. If only routine l gaseous effluents are anticipated, Dg, may be set to zero, i l D = the projected air dose due to gamma emissions f rom noble gases f or l 9 the next 31 days of gaseous releases. l 1 3-47 Rev. 8, 1/94 , i ! i l 1
Hatch ODCM the cumulative air dose due te g a.-c.a em;sseens from noble gas Dp = releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration. D,= the anticipated air dose'due to gamma emissions from noble gas y releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous ef fluents. If only routine gaseous effluents are anticipated, D y, may be set to zero. D og = the projected dose to the total bod / or organ o, due to releases o f I- 131, I-133, tritium, and particulates f or the next 31 days of gaseous releases. Dg = the cumulative dose :o the total body or organ o, due to releases of I-131, I-133, tritium, and particulates that have occurred in l the elapred portion of the current quarter, plus the release under consideratien. D,= o the anticipated dose to the total body or organ o, due to releases of I-131, I-133, tritium, and particulates, contributed by any planned activities during the next 31-day period, if those i activities will result in gaseous releases that are in addition to ? routine gaseous effluents. If only routine gaseous effluents are anticipated, D o, may be set to zero. t = the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under , consideration (even if the release continues into the next quarter). i 3.5.2 Dose Proiections for Specific Releases l l l Dose projections may be performed f or a particular release by performing a pre-t release dose calculation assuming that the planned release will proceed as I anticipated. For air dose and individual dose projections due to gaseous effluent releases, follow the methodology of Section 3.4, using sample analysis results for the gaseous stream to be released, and parameter values expected to exist dusing the release period. 3-48 Rev. 8, 1/94
H a t :-- OOCM 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS Sect.:n of Inttaa'. Use Definition Term AG = the administrative allocation factor for gaseous streams, applied to divide the gaseous release limit 3.3.2.1 among all the release pathways [unitiess). AG, = the administrative allocation factor for gaseous source stream s, applied to divide the gaseous limit among all the release pathways release 3.3.3 [unitiess}. AG y = the administrative allocation factor for gaseous release pathway v, applied to divide the gaseous i limit among all the release pathways release 3.3.2.2 [unitiess). Bi= the elevated finite plume air dose factor due to gamma emissions from noble gas radionuclide i in the released from the main stack effluents 6 3.3.2.2 [(mrad /y)/(yci/s)]. e= the setpoint of the radioactivity monitor measuring , the concentration of radioactivity in the effluent 3.3.2.1 line prior to release [uci/mL}. ! e nco = the reading of the condenser offgas pretreatment monitor at the alarm setpoint ( rrJt/ h ] . 3.3.5 l e ns = the calculated noble gas effluent monitor setpoint l 3.3.3 for gaseous source stream s [pC1/mL). . c nv = the calculated noble gas effluent monitor setpoint for release pathway v [pci/mL]. 3.3.2.2 Dj, = the dose to organ j of an individual in age group a, due to gaseous releases of I-131, I-133, tritium, and radionuclides in particulate form with half-lives 3.4.3 l greater than 8 days (mrem). l 3-49 Rev. Si 1/94 , L l i L -. . -
i H+tch ODCM 5e ct .:r. :: In;t.a'. .'s e Term Definition D,= o the anticipated dose to organ o due to releases of non-noble-gas radionuclides, contributed by any [ planned activities during the next 31-day period l 3.5.1 , f [ mrem). 1 l the cumulative dose to organ o due releases of non- ! Doe = noble-gas radionuclides that have occurred in the l l elapsed portion of the current quarter, plus the 3.5.1 release under consideration [ mrem). D op = the projected dose to organ o due the next 31 days of releases of non-noble-gas radionuclides gaseous 3.5.* (mrem). Dg = the air dose due to beta emissions from noble gas 3.4.1 radionuclides [ mrad). the anticipated air dose due to beta emissions from DJa = noble gas releases, contributed by any planned activities during the next 31-day period (mrad). 3.5.1 Dgg
= the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release 3.5.;
under consideration [ mrad) . Djp = the projected air dose due to beta emissions from noble gases, for the next 31 days of gase:as releases 3.5.. [ mrad].
= the air dose due to gamma emissions from noble gas D) 3.4.1 radionuclides [ mrad) .
D), = the anticipated air dose due to gamma emissions f rom noble gas releases, contributed by any planned activities during the next 31-day period [ mrad]. 3.5.~ 3-50 Rev. 8, 1/94
Hetch 00CM I Secta:n Of Definitton Inittal Use Term D yg = the cumulative air dose due to gamma emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release 3.5.1 under consideration (mrad). l D p = the projected air dose due to gamma emissions from noble gases, for the next. 31 days of gaseous releases 3.5.1 [ mrad]. (6/D)s.p = the annual average relative deposition factor for release pathway v, at the location , of the controlling 3.4.3 receptor, from Table 3-7 (m). i DRk= the skin dose rate at the time of the release 3.4.1.1 (mrem /y). DRg = the dose rate to organ o at the time of the release 3.4.1.2 (mrem /y). DRt = the total body dose rate.at the time of the release Ji4.1.1 (mrem /y). E co = the calibration factor for the condenser offgas pretreatment monitor ((uci/s) per (cfm mR/h)). 3.3.5 f p. = the maximum anticipated actual discharge flewrate for l release pathway v during the period cf the planned 3.3.2.0 release (mL/s).
= the condenser offgas flowrate (cfm). 3.3.5 f
en f = the maximum anticipated actual discharge flowrate for u gaseous source stream e during the period of the planned release (mL/s). 3.3.3 K, = the total body dose factor due to gamma emissions l from noble gas radionuclide i, from Table 3-5 l 3 3.3.2.2 ((mrem /y)/( C1/m )}. 3-51 Rev. 8, 1/94 ( l
~. - . - ~ - _ . . . . - . - - -- - ~ . - _ - - - _ _ . . . - - . ~ - - - ..
t Hatch QQQ3 Section of Definition Initial Use Term Lj = the skin dose f actor due to beta emissions f rom noble radionuclide i, from Table 3-5 gas 3.3.2.2 ((mrem /y)/(pci/m3 )] . M; = the air dose f actor due to gamma emissions from noble radionuclide i, from Table 3-5 gas 3 3.4.2 ( (mrad /y)/(yci/m ) j . , N= the number of simultaneously active gaseous release 3.3.4 pathways (unitlessj. N, = the air dose factor due to beta emissions from noble radionuclide 1, from Table 3-5 gas , 3 3.4.2 [(mrad /y)/(yCL/m )). l F,o = the site-specific dose factor for radionuclide 1 (I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days) l and organ o. The values of Pjo are equal to the site-presented in Table 3-9 specific R,,g values [ (mrem /y) / (pci/m 3) ) . 3.4.1.2 i Q,y = the release rate of noble gas radionuclide i from l release pathway v during the period of interest (yci/s). 3.3.2.2 i Qig= the release rate of radionucitde 1 (*-131, I-133, ! tritium, and radionuclides in particulate form with half-lives greater than 8 days) from gaseous release I pathway v during the period of interest [pci/s). 3.4.1.2 6= the cumulative release of noble gas radionuclide i 3 from the main- stack during the period of interest f 3.4.2 [pci). l-l l 6n, = the cumulative release of noble gas radionuclide i l from release pathway v during the period of interest 3.4.2 [yci). 3-52 Rev. 8, 1/94 l
l Hatch ODCM Sectton cf Definttton lp_t t t a l Usg Term the cumulative release of non-noble-gas radionuclide 6lv= i from release pathway v, during the period of 3.4.3 interest (pci]. the site-specific dose f actor for age group a, radio-Raig = nuclide i, exposure pathway p, and organ j. Values and units of R,ig for each exposure pathway, age and radionuclide that may arise in group, calculations for implementing Section 3.1.4 are 3.4.3 listed in Table 3-8 through Table 3-10. Rk= the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to noble gases in f 3.3.2.1 l the release under consideration (unitless). l l Rt = the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases 3.3.2.1 in the release under consideration (unitiess). rk = the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to ncele gases in 3.3.3.1 the source stream under consideration (unitiess) . rg = the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the source stream under consideration l 3.3.3.1 [unitiess). 1 SF = the safety factor used in gasecus setpoint calculations to compensate for statistical 3.3.2.2 fluctuat.ons i and errors of measurement (unitless). l the number of whole or partial days elapsed in the l t = current qua.ter, including the period of the release 3.5.. under consideration. V,= the elevated finite plume total body dose f actor due to gamma emissions f rom noble gas radionuclide i in l ' 3-53 Rev. 8, 1/94 l
i i l Hatch ODCM Section of Definition Initial Use Term i the effluents released from the main stack [(mrem /y)/(pCi/s)). 3.3.2.2 Wy ;p a the annual average relative dispersion [ti?6)yp) or at the location of the deposition ( (D7D)yp] controlling receptor, for release pathway v ,- as l appropriate to ex posure pathway p and radio-3.4.3 nuclide 1. X= the noble gas concentration for the release un'ier consideration (yci/mL). 3.3.2.2 X = the concentration of radionuclide i applicable to tr active gaseous release pathway r (gC1/mL). 3.3.4 X,3 = the measured concentration of rr.dionuclide i in gaseous source stream s (pci/mL). 3.3.3 i 1 I X,y = the measured concentration of radienuclide i in gaseous stream v [uci/mL). 3.3.2.2 (X/Q) = the highest relative concentration at any point at or beyond the SITE BOUNDARY (s/m3 ). 3.3.2.1 (57D)rb = the annual average SITE BOUNDARY relatave concen-l tration applicable to active gaseous release pathway ) 3.3.4 r (s/m3). (X7D)sb = the highest annual average relative concentration at the SITE BOUNDARY for the main stack, from Table 3-4 3.4.2 (s/m3 ). (176)y3 = the highest annual average relative concentration at the SITE BOUNDARY f or the o..scharge point of release pathway v, from Table 3-4 (s/m3 ). 3.3.2.2 3-54 Rev. 8, 1/94
l Haten 03CM Se tion of De f i n it <.cn Initta! Use Term (IGD),p = annual average relative dispersion f actor for release pathway v, at the location of the controlling 3 3.4.3 receptor, fram Table 3-7 [s/m ). l 6 l 1 I l ! i l \ l l i l I l l i { 3-55 Rev. 8, 1/94 l
i l l Hot:- OCCM l CHAPTER 4 l , RADIOLOGICAL EtWIPONMENTAL MONITORING PROGRAM ( l i 4.1 LIMITS OF OPERATION , I Thus, a single The following limits are the same for both units at the site. program including monitoring, land use survey, and quality assurance serves both units. ! l Radiolocical Environmental Monitorina 4.1.1 ) 6.19(1), the Radiological l In accordance with Technical Specification Environmental Monitoring Program (REMP) shall be conducted as specified in I Table 4-1. ; 4.1.1.1 Applicability l This control applies at all times. i Actions I 4.1.1.2 4.1.1.2.1 With the REMP not being conducted as specified in Table 4-1, submit ; (NRC), in the Annual Radiological to the Nuclear Regulatory Commission Environmental Surveillance Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. Deviations from the required sampling schedule are permitted if specimens are i unobtainable due to hazardous conditions, unavailability, inclement weather, If deviations are due to equipment equipment malfunction, or other just reasons. malf unction, ef forts shall be made to complete corrective action prior to the end of the next sampling period. 4.1.1.2.2 With the confirmedI measured level of radioactivity as a result of l plant effluents in an environmental sampling medium specified in Table 4-1 exceeding the reporting levels of Table 4-2 when averaged over any calendar quarter, submit within 30 days a Special Report to the NRC pursuant to Technical Specification 6.9.2. The Special Report shall identify the caustu s ) for exceeding the limit (s) and define the corrective action (s) to be taken to reduce radioactive ef fluents so that the potential annual dose to a MEMBER OF THE PUBLIC 1 Defined as confirmed oy reanalysis of the original sample, or analysis of a duplicate or new sample, as appropriate. The results of the confirm-atory analysis shall be completed at the earliest time consistent with the analysis. 4-1 Rev. s, 1/94
( Hatch ODCM 3, 3.. 3, and 3.;.4. Tne l 1s less than tne calendar year lamits of Sect a:ns 2. . l methodology and parameters used to estimate the potentral annual dose to a KEMBER l l OF THE PUBLIC shall be indicated in the Special Report. When more than one of the radionuclides in Table 4-2 are detected in the sampling medium, this report shall be submitted ifs concentration (1) , concentration (2l , , , , , g,n reporting level (1) reporting level (2) When radionuclides other than those in Table 4-2 are detected and are the result of plant effluents, this Special Report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar This Special Report is year limits stated in Sections 2.1.3, 3.1.3, and 3.1.4. the result of plant not required if the measured level of radioactivity was not effluents; however, in such an event, the condition shall be described in the The levels of naturally-Annual Radiological Environmental Surveillance Report. occurring radionuclides which are not included in the plant's ef fluent releases l l need not be reported. l 4.1.1.2.3 If adequate samples of milk, or during the growing season, grass l l or leafy vegetation, can no longer be cctained from one or more of the sample l t 4-1, or if the availability is frequently or locations required by Table l to identif y specific locations f or j persistently wanting, ef f orts shall be made: ' obtaining suitable replacement samples; and to add any replacement locations to the REMP given in the ODCM within 30 days. The specific locations from which Pursuant to Technical samples became unavailable may be deleted from the REMP. Specification 6.17, documentation shall be submitted in the next Annual Radioactive Effluent Release Report for the change (s) in the ODCM, including revised figure (s) and table (s) reflecting the changes to the location (s), with supporting information identif ying the cause of the unavailability of samples and justifying the selection of any new location (s). When the ACTION statement or other requirements of thic control cannot be met, Entry into steps need not be taken to change the Operational Mode of the unit. as a minimum, an Operational Mode or other specified CONDITION may be made if, l the requirements of the ACTION statement are satisfied. I I i I l i t 4-2 Rev. B, 1/94
_ m __ - _ . _ _ _ - _ - Hatch COCM r Sarvetilance Peq. re.ents 4.1.1.3 l The REMP samples shall be collected pursuant to Table 4-1 from the locations described in Section 4.2, and shall be analyzed pursuant to the requirements of Table 4-1 and Tat,le 4-3. Requiteo detect ion capabilities f or thermoluminescent dosimeters used for environmental ueasurea.ents shall be in accordance with the recommendations of Regulatory Guide 4.13. Program changes may be initiated based j on operational experience. Analyses shall be performed in such a manner that the stated MINIMUM DETECTABLE CONCENTRATIONS (MDCs) will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizcs, the presence of interfering radionuclidee, or other uncontrollable circumstances may render these In such cases, the contributing factors will be identified HDCs unachievable. and described in the Annual Radiological Environmental 9urveillance Report. 4.1.1.4 Basis l The REMP regaired by this control provides representative measurements of radiation and of radioactxve materials in those exposure pathways, and f or tacse radionuclides, which lead to the highest potential radiation exposures of MEhBERS OF THE PUBLIC resulting from the plant operation. The REMP implements Section IV. B. 2, Appendix I, 10 CFR 50, and thereby supplements the radiologicsl ef fluent l monitoring program by measuring concentrations of radioactive materials and levels of radiation, whfab may then be compared with those expected on the basis of the ef fluent measurements and modeling of the environmental exposure pathways. The detection capabilities required by Table 4-3 are within state-of-the-art f or routine environmental measurements in industrial taboratories. l t i l r I l
\
4-3 Rev. 8, 1/94 l l
i a a g ki Approximate Exposure Pathway " umber of Sr.mpling and Type f Analysis and Frequency and/or Sample Sample Collection Frequency Locctions a
- 1. AIRBORNE l y
Radiciodine canister. I-131 weekly. a 6 Continuous cperation Radiolodines and of sampler with sample *f Particulates collection weekly. Particulate sampler. Analyze for *' g gross beta radioactivity not less than a 24 hours following filter change and *d analyze for I-131 weekly. Perform gamma isotopic analysis on affected sample when gross beta activity is 10 $< times the yearly mean of control . samples. Composite (by location) for n gamma isotopic analysis quarterly. '3 s 2 DIRECT RADIATION a
- 2.
- 37 Quarterly. Gamma dose quarterly. .
#. Direct Radiation 1 o
- 3. INGESTION u 4
h Bi-weekly. Gamma isotopic and I-131 analyses bi-Milk weekly. n. n Gamma isotopic analysis on edible o Fish or clams C 2 Semi-annually. portions semi-annually. y Grass or Leafy 3 Monthly during crowing Gamma isotopic analysis monthly.d " Vegetation season.
- 4. WATERBORNE 2 Composite" sample Gamma isotopic analysis inonthly.
surface collected monthly. Composite ( tay location) for tritium analysis quarterly. w e
< 2 Semiannually Gamma isotopic analysis yearly. g Sediment <
O CD 7 i g s o e X t.
I l I i W - % Mi s ,
.tr-i Approximate Exposure Pathway Number of Sampling and Type of Analysis and Frequency g Sample Collection Frequency o and/or Sample Locations a rt I River Water collected 1-131 analysis on each sample when bi- 3 Drinking Water *E From each of weekly collections are required.
the one to near the intake will be a composite sample; Gross beta and gamma isotopic analyses i three nearest on each sample; composite (by water supplies the finished water will be a grab sample. location) for tritium quarterly. , which could be > affected by HNP These samples will be l collected monthly $ discharge: one o, l sample of river unless the calculated o water near the dose due to * [ i intake and one consumption of the <j sample of water is greater than - l finished water. 1 mrem / year; then the j m collections will be g bi-weekly. The r
, ! collection may revert 8 - ;}
- to monthly should the ,
y calculated doses it become less than 1 a,, mrem / year. w e X u N y
)
4h T M O 4 a
% z e e 4
O (* 8 x 2 _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ -s
1 m U TABLE NOTATIONS and in Figure 4-1 through Figure 4 -5. ,
- a. sample locations are shown in Table 4-4, -
- b. Up to three In sampling locations within 5 miles and in different sectors will be used as addition, one or more control locations beyond 10 miles will be used.
2 o available. o Clams will be sampled if a
- c. Commercially or recreationally important fish may be sampled. 7 difficultles are encountered in obtaining sufficient fish samples.
- d. If gamma isotopic analysis is not sensitive enough to meet the required MINIMUM DETECTABLE x CONCENTRATION (MDC), a separate analysis for I-131 may be performeds at intervals not exceeding a h
- e. Composite samples shall be collected by collecting an aliquot o e-few hours. O o
f. If it is found that river water downstream of HNP is used for drinking, water samples will E be collected and analyzed as specified herein. ?. g
- g. A survey shall be conducted annually at least 50 river miles downstream of HNP to identify a those who use the Altamaha River water for drinking. e W"
An e O 9 m 3 r' W t-5 fl p F* 19 O M t-3 O
"U M
M w z
< d O
3 I O H O x 2 e 4
l l 1!1 I* o7 O0O"E 0 f,o3, f, h a e n<rg n3mn0>e s2 $0,,,,ae 'tO<e b' o.Oo"<.=<
'k E s u .on )*L,c m$3e l
s t y s
) i f
an t x e eo we Li 2 3 3 y t + + + ra E E E a g w ote k 1 1 2 h s g/ t se iC a p aVp r G ( r e t a w g n
) 2 i 0 1 1 2
+ k k /L + + + .
+
l E E E E E n d i i i r e s MpC 3 6 7 3 4 d u ( o e n b f y l I a e m v ) L e t s e / L e i w 3 3 g h 4 4
+
4
+
4
+
4
+ + +
l p cp n , + i s g E E E E E E E m i a 0 t F /k 3 1 3 1 2 1 2 s 2 r o i r p cp e f o e ( t R a e w u l r g a o n v i ee ) 1 1 k n a nt s3 m 1 ra - + + i . , ol e/ E E E rd s bu asi de t rcCpc ii 9 1 2 s s r u i At ( o x r f e e a b P e y uy a l a w am v h t L a 1/ p
) I' 41
- 3 2 2 2 2 1 1 2 2
- rL 4 3 2 0 + + + + 1C p r
+ + + + + + + +
e/ + E E E E E E E E E E E E e ti E R t ac Wp 1 4 1 3 3 4 7 2 3 5 2 1 F4 a ( 2 C0 w
+
- 0E g 4 n 3 i - e k hf n t o i r se d s iu - i l o - s sa n y iv h f _ l a n 4 9 5 8 5 0 6 6 9 5 5 5 9 1 3 4 3 1 7 3 4 1 0 0 1 4 1 Ta I A 5
- - - - - - 1 - - - -
3 -
- n e o o nZ r b - s s a a
_ H M F c C Z N I C C B L . . a b za<* ,- w%@4 . aI"
l' l \lili 2N O0O1 y .ge a mo ye :y9 3 oeNn*<$oe oO3 e3
)
>on
- 4. uf S. e P t , 2 2 ng + +
ek) E E he m/y ci iir 5 8 ih dcd . ht ep 1 1 w S( n si k as ee pd i n r o, rl o ytk)ig 1
+
1
+
1
+
ec hu s fa a/t E E E tn s e tie Ooi . a L ecw 6 6 8 d . e d r gp G e( .a d s
)
v 1 r e u C . . s D M 0e1 1h u e (
. b t 7 e n b y n 1 1 1 ohn a o )
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Hatch ODCM 4.1.2 1._and Use Censas In accordance with Technical Specification 6.19(2), a land use census shall be conducted and shall identify the locations of the followings the nearest MILK ANIMAL I and the nearest permanent residence in each of the 16 meteorological sectors within a distance of 5 miles; and all MILK ANIMALS within a distance of 3 miles. , 4.1.2.1 Applicability This control applies at all times. 4.1.2.2 Actions I i 4.1.2.2.1 With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than values :;rrently being calculated in accordance with Section 3.4.3, identify the new location (s) in the next Annual Radioactive Effluent Release Report. i 4.1.2.2.2 With a land use census identifying a location (s) which yields a l ! calculated dc,ae or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in ' accordance with Section 4.1.1, add the new location (s) to the REMP within 30 days if samples are available. The sampling location, excluding control station location (s), having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from the REMP if new sampling locations are added. Pursuant to Technical Specification 6.17 sucmit in the next Annual ' Radioactive Effluent Release Report any change (s) in the ODCM, including the revised figure (s) and table (s) reflecting any new location (s) and inf ormation i i supporting the change (s). When the ACTION statement or other requirements of th;s control cannot be met, steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified CONDITION may be made if, as a minimum, the requirements of the ACTION statement are satisfied, l l 1 Defined as a cow or goat that is producing milk f or human consumption. l 4-9 Rev. S, 1/94 1
l I Hatch COCM 4.1.2.3 Surveillance Pequirements The land use census shall be conducted annually, using that information which will provide good results, such as'a door-to-door census, a visual census from l automobile or aircraf t, concultation with local agriculture authorities, or some l l combination of these awthods, as feasible. Results of the land use census shall l be included in the Annual Radiological Environmental Surveillance Report. 4.1.2.4 Basis This control' is provided to ensure that changes in the use of UNRESTRICTED AREAS - are identified and that modifications to the REMP are made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I.to 10 CFR Part 50. l l 4-10 Rev. 8, 1/94
. - - .. .. . . . __ . ~ _ . .- .~. .- -. - _
Hatch 00CM 4.1.3 Interlaboratory Comoarison Pro: ram In accordance with Technical Specification 6.19 ( 3 ), analyses shall ce performed on radioactive materials supplied as part of an Interlaboratory Comparison Analyses are required to be Program which has been approved by the NRC. performed only in cases in which the sample type and analysis t.re the same as the sample type and analysis included in Table 4-1. 4.1.3.1 Applicability This control applies at all times. 4.1.3.2 Actions 4.1.3, report the
!ith analyses not being performed as required by Section corrective actions taken to prevent a recurrence in tne Annual Radiological Environmental Surveillance Report.
When the ACTION statement or other requirements of this :ontrol cannot be met, Entry into steps need not be taken to change the Operational Mode of cne unit. if, as a minimum, an Operational Mode or other specified CONDITION may be made the requirements of the ACTION statement are satisfied. 4.1.3.3 Surveillance Requirements Either a summary of the results obtained as part of the requ; red Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Surveillance Report, or participants in the EPA cross-check program shall provide the EPA program code designation for the plant in the Annual Radiological Environmental Surveillance Report. 4.1.3.4 Basis The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent che.xs on the precision and accuracy of the measurements of radioactive material :n environmental sample matrices tre performed as part of the quality assurance program f or environmental monitoring, in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2, Appendix I, 10 CFR 50. 4-11 Rev. 8, 1/94
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l Hatch ODCM 4.2 RADIOLOGICAL ENVIRONMENTAL MONITOR:!:G LOCAT!ONS Table 4-4, and Figure 4-1 through Figure 4-5 specif y the locations at which the measurements and samples are taken for the REMP required by Section 4.1.1. I I l r 4-12 gey, 8, 1/94 (
Hatch ODEM Table 4-4. Radiological Eny tror.mer.t al Monitoring *.ocations Distance Sampip Location Descriptive Location Direction Type Number (miles) WNW 0.8 D 064 Roadside park N 1.9 D 101 Inner ring NNE 2.5 D 102 Inner ring Inner ring NE 1.8 AD 103 l Inner ring ENE 1.6 D 104 E 3.7 D 105 Inner ring Inner ring ESE 1.1 DV 106 107 Inner ring SE 1.2 AD SSE 1.6 D i 108 Inner ring D 109 Inner ring S 0.9 f Inner ring SSW 1.0 D 110 111 Inner ring SW 0.9 D Inner ring WSW 1.0 ADV 112 113 Inner ring W 1.1 D Inner ring WNW 1.2 D 114 Inner ring NW 1.1 D 115 Inner ring NNW 1.6 AD 116
** R 170 Upriver WNW
** R 172 Downriver E Outer ring N 5.0 D 201 Outer ring NNE 4.9 D 202 203 Outer ring NE 5.0 D Outer ring ENE 5.0 0 204 Outer ring E 7.2 D 205 206 Outer ring ESE 4.8 D Outer ring SE 4.3 D 207 208 Outer ring SSE 4.8 D 209 Outer ring S 4.4 D Outer ring SSW 4.3 D 210 4-13 Rev. 8, 1/94
1 l l l Hatch ODCM I Tao.e 4-4 ( entd,. Radtolog; cal EnvLrenmental Monitoring Locations l l Distance S ampip Location Number Descriptive Location Direction (miles) Type SW 4.7 D 211 Outer ring WSW 4.4 D 212 Outer ring W 4.3 D 213 Outer ring Outer ring WNW 5.4 D 214 bnf 4.4 D 215 Outer ring Outer ring NNW 4.8 D 216 N 8.0 D 301 Toombs Central School State Prison ENE 11.2 AD 304 ENE 10.3 H 304 State Prison 309 Baxley substation S 10.0 AD 416 Emergency News Center NNW 21.0 DV
- Sample Types:
A - Airborne Radioactivity D - Direct radiation M - Milk R - River (fish or clams, shcreline sediment, and surface water) V - Vegetation
** Station 170 is located at approximately 0.6 river miles upstream of the intake structure for river water, 1.1 river miles for sediment and clams, and 1.5 river miles for fish.
l Station 172 is located at approximately 3.0 river miles downstream of the discharge structure for river water, sediment, and clams, and 1.7 l river miles for fish. The location f rom which river water and sediment may be taken can be rather precisely defined. Often, the sampling locations for clams have to be extended over a wide area to obtain a suf ficient quantity. High water adds to the difficulty in obtaining clam samples; high water might also make an otherwise suitable location for sediment sampling unavailable. A stretch of the river on the order of a few miles or so is generally needed to cbtain adequate fish samples. The mile locations given above represent approximations of the locations about which the catches are taken. 4-14 Rev. 8, 1/94
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~ 4-19 Rev. 8, 1/94
- - - , - . ~ - . . . - - - -. . - - -
1 Hatch ODCM 1 CHAPTER 5 TOTAL DOSE DETERMINATIONS 5.1 LIMIT OF OPERATION 1 In accordance with Technical Specification 6.18(10), the dose or dose coassitment ] d to any MEMBER OF THE PUBLIC over a calendar year, due to releases of radio-1
' activity and to radiation from uranium fuel cycle sources, shall be limited to ' less than or equal to 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem.
1 5.1.1 Aeolicability 1 I This limit applies at all times. 5.1.2 Actions With the calculated doses f rom the release of radioactive materials' in liquid or i gaseous ef fluents exceeding twice the limits of Section 2.1.3, 3.1. 3, or 3.1. 4, calculations shall be made according to Section 5.2 methods to determine whether ) j the limits of Section 5.1 have been exceeded. If these limits have been j exceeded, prepare and submit a Special Report to the Nuclear Regulatory Commission, pursuant to Technical Specification 6.9.2, within 30 days, which l
~
defines the corrective actions to be taken_to reduce subsequent releases to l l prevent recurrence of exceeding the limits of Section 5.1 and includes .e i schedule for achieving conformance with the limits of Section 5.1. This Special Re port , as defined in 10 CFR 20.2203, shall also include an analysis which estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC f rom uranium l f uel cycle sources (including all ef fluent patnways and direct radiation) for the + calendar year that includes the release (s) covered by this report. This Special Report shall also describe the levels of radiation and concentrations of I radioactive material involved, and the cause of the exposure levels or J j concent;ations. If the estimated dose (s) exceeds the limits of Section 5.1, and ' if the release condition resulting in violation of the provisions of 40 CFR 190 has not already been corrected, the Special Report shall include a request for } variance in accordance with the provisions of 40 CFR 190 and including the specified inf ormation of 40 CFR 190.11(b) . Submittal of the report is considered a timely request, and a variance is granted until staf f action on the request is
~
complete. 7 5 When the ACTION statement or other requirements of this control cannot be met, i i steps need not be taken to change the Operational Mode of the unit. Entry into 1 l I 5-1 Rev. 8, 1/94 a i i -_
Hatch ODCM T an Operational Mode or other specified CONDITION may be made if, as a man .um, the requirements of the ACTION statement are satisfied. 5.1.3 Surveillance Reauirements Cumulative dose contributions f rom liquid and gaseous ef fluents and f rom direct This requirement radiation shall be determined in accordance with Section 5.2. i is applicable only under the conditions set forth above in Section 5.1.2. 5.1.4 Basis This control is provided to meet the dose limitations and reporting requirements of 40 CFR 190. The control requires the preparation and submittal of a Special Report whenever the calculated doses f rom plant radioactive ef fluents exceed the limits of Section 5.1. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a HEMBER OF THE PUBLIC will exceed the the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I and if direct radiation doses from the units, such as direct exposure f rom outside storage tanks, are kept small. The Special Report will describe a course of action which should result in the limitation of dose to a MEMBER OF THE PUBLIC for a calendar year to within the 40 CFD 190 limits. For the purposes of the Special Report, it may.be assumed that the dose commitment to the KEMBER OF THE PUBLIC from other uranium f uel cycle sources is negligible with the exception that dose contributions from other uranium fuel f l cycle facilities at the same site or within a radius of 5 miles must be If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the f considered. ' requirements of 40 CFR 190, the Special Report with a request for variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.2203(a)(4), is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the '.imits of 40 CFR 190 and does not apply in any way to tne An requirements for dose limitation addressed in other sections of this ODCM. individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle. 5-2 Rev. 8, 1/94
Hatch ODCM 5.2 DEMONSTRATION OF COMPLIANCE There are no other uranium fuel _ cycle facilities within 5 miles of the plant site. Therefore, for the purpose of demonstrating compitance with the limits of PUBLIC in the vicinity of the Section 5.1, the total dose to a MEMBER OF THE follows: plant site due to uranium fuel cycle sources shall be determined as Dn = Dg, + Do + Dp + Dy (5.1) where: in the total dose or dose commitment to the total body or organ k, Dn = mrem. the dose to the same organ due to radicactivity discharged f rom the DL= plant site in liquid effluents, calculated in accordance with Section 2.4.1, in mrem. the dose to the same organ due to non-ncble-gas radionuclides Do = discharged from the plant site in gaseous ef fluents, calculated for the controlling receptor in accordance with Section 3.4.3, in mrem. Op = the direct radiation dose to the whole body of an individual at the controlling receptor location,. due to radioactive materials retained within the plant site, in mrem. Values of direct radiation dose may be determined by measurement, calculation, or a combination of the two. the controlling Dy = the external whole body dose to an individual at recepter location, due to gamma ray e: .ssier.s f rom noble gas radio-nuclides discharged from the plant s;te in gaseous effluents, in is calculated as follows (equation adapted f rom Reference mrem. DN 1, page 22, by re-casting in cumulative dose form):
~8 + { (V,
- 6j,) (5.2)
Dy = 3.17 x 10 {v (y/~d ) ,p { (Kj - @j,.)
> i r where:
5-3 Rev. 8, 1/94
- _ . .. .. .-.. .... . _ . -- _ . ~ -. . - .- ..~ - . - . . . - -
1 i i Haten ODCM 1 y/i3.15 1 I s1 ' 3.17 x 10-6 = a unit s convers.:n factor: 6,y = the cumulative release of noble gas radionuclide a from non-elevated release pathway v (pCL), during the period of interest. the l 6;, = the value of 6;y for the main stack; that 19, cumulative release of noble gas radionuclide i from the 1 main stack (pci), during the period of interect, 3 y= the total-body dose factor due to gamma emissions from noble gas radionuclide i (mrem /y)/(uci/m 3 ), from I Table 3-5. the elevated f inite-plume total-body dose f actor at. the 1 d Vg= controlling receptor location, due to gamma emissions from noble gas radionuclide i in ef fluents released f rom the main stack (mrem /y)/(u c i/s), from Table 3-7. J
~
(d D)yp = annual average relative dispersion factor for release pathway v, at the location of the controlling receptor, from Table 3-7 (s/m3 ). As defined above, DL and DG are for different age groups, while DD and Dg are not age group specific. When a more precise determination of Dn is desired, and those values values of DL and DG may be calculated for all four age groups, used in equation (5.1) to determine age group specific valuer of Dn; the largest value of Dn for any age group may then be compared to the limits of Section 5.1. l l j l i l i l l
\
5-4 Rev. 8, 1/94 l r
,w.
i Hntch ODCM f CHAPTER 6 TO POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE
- THEIR ACTIVITIES INSIDE THE SITE BOUNDARY 6.1 REQUIREMENT FOR CALCULATION To support the reporting requirements of Section 7.2.2.3, an assessment of the radiation doses f rom radioactive liquid and gaseous effluents to MEMBERS OF THE (Figure 10-1) shall be PUBLIC due to their activities inside the SITE BOUNDARY at least once per calendar year.
performed as specified in Section 6.2, 6.2 CALCULATIONAL METHOD For the purpose of performing the calculations required in Section 6.1, the dose the to a member of the public inside the SITE BOUNDARY shall be determined at l The dose to locations, and for the receptor age groups, defined in Table 6-1. much a receptor at any one of-the defined locations shall be determined as follows: (6.1) D gg
=
{D3+DS
- Dp } = F o where:
Dg = the total dose to the total body or organ k, in mrem. the dose to the same organ due to Jnhalatton of non-noble-gas DA= radionuclides discharged f rom tne plant s;te n gaseous ef fluents, calculated in accordance with Section 3.4.3, in mrem. The (i}6) value to be used is given for each receptor location in Table 6-1; depleted (i~/D) values may be used in calculations for non-noble-gas radionuclides. the dose to the same organ due to around plano deposition of non-D3= noble-gas radionuclides discharged from the plant site in gaseous effluents, calculated in accordance with Section 3.4.3, in mrem. The (6~/6) value to be used is given for each receptor location in Table 6-1. 6-1 Rev. 8, 1/94
Hatch OCCM Dp = the external whole :::dy dose due :: gama ray emissions f rcrn nocle gas radionuclides discharged from the plant site in gaseous effluents, calculated using equation (5.2), in mrem. The values of (X/6) and V; that are to be used are given for each receptor location in Table 6-1. F, = the occupancy f actor for the given location, which is the f raction of the year that one individual MEMBER OF THE PUBLIC is assumed to be present at the receptor location [unitiess). Values of F o for each receptor location are included in Table 6-1. I i 5 2 6-2 Rev. 8, 1/94
Hatch QQEB Table 6-1. Attributes of Member of t r.e Put11: Receptor Locations Inside tne SITE BotfNDARY Locations Roadside Park, WNW at 1182 meters ' Ace Group Child i Occucancy Factors 2.28x 10 4 (based on 2 hours per year) 4 Dispersion and Deposition Parameters: Parameter Ground-Level Elevated Undepleted (X/D), s/m 3 7.83 E-6 2.42 E-8 Depleted (i/6), s/m 3 7.00 E-6 2.37 E-8 (6/D), m*2 2.01 E-S 1.29 E-9 I 1 Elevated Plume Dose Factors: 1 V, V; Radionuclide Radionuclide (mrem /y)/(pCi/s) (mrem /y)/fuci/s) 8.39 E-05 Xe-133m 1.32 E-05 Kr-85m 1.31 E-06 Xe-133 1.37 E-05 Kr-85 4.60 E-04 Xe-135m 2.51 E-04 Kr-87 l Kr-88 1.14 E-03 Xe-135 1.37 E-04 9.89 E-04 Xe-137 9.91 E-05 = Kr-89 7.32 E-04 Xe-138 6.45 E-04 Kr-90 1.70 E-06 Ar-41 7.69 E-04 Xe-131m Y
- Values from Reference 16.
+ See footnotes to Table 3-6.
l i I 6-3 Rev. 8, 1/94 l l
i Hatch ODCM P.:" ;c Receptor Lc:at;ons Tacle 6-1 (contd). Attricutes of Menmer af the . Inside the SITE 80UNuARY mR Location Camping Area, WNW at 1274 meters hoe Group Child occupanev Factor: 5.48 x 10-3 (based on 48 hours per year) QJsoersion and Deposition Parameters: i l Ground-Level Elevated Parameter 2.38 E-8 Undepleted (i/D), s/m 3 7.03 E-6 3 6.27 E-6 2.33 E-8 Depleted ( X76 ) , s/m (676), n(2 1.80 E-8 1.21 E-9 Elevated Plume Dose Factors: Ya V: i Radionuclide Radionuclide ! (mrem /y)/(uci/s) (mrem /y)/(uci/s) 7.84 E-05 Xe-133m 1.24 E-05 Kr-85m 1.22 E-06 Xe-133 1.28 E-05 Kr-85 4.28 E-04 Xe-135m 2.34 E-04 Kr-87 1.06 E-03 Xe-135 1.27 E-04 Kr-88 9.19 E-04 Xe-137 9.23 E-05 Kr-89 6.80 E-04 Xe-138 5.99 E-04 Kr-90 I Ar-41 7.14 E-04 Xe-131m 1.59 E-06 i i l
\
- Values from Reference 16.
+ See footnotes to Table 3-6.
l l 6-4
- Rev. 8, 1/94 1
4 i
1 l l Hatch OQQ,fj Tacle 6-1 (contd). Attributes of Member of the Patlic Receptor Locations Inside the SITE BOUNDARY l l l l Locations Recreation Area, SSE at 1030 meters Ace Grouos Child Occupancy Facter: 2.37 x 10~2 (based on 208 hours per year) Dispersion and Deoosition Paramete : Ground-Level Elevated Parameter Undepleted (176), s/m 3 6.42 E-6 3.30 E-8 ; 5.73 E-6 3.21 E-8 Depleted (i?6), s/m 3 I (D7D), m *2 2.36 E-8 1.56 E-9
-m
' 7 I Elevated Plume Dose Factors: j I
Yi Yi Radionuclide Radionuclide (mrem /y)/(sc i/s) ! (mrem /y)/(uci/s) l 7.21 E-05 Xe-133m 1.14 E-05 Kr-85m 1.13 E-06 Xe-133 1.17 E-05 Kr-85 3.99 E-04 Xe-135m 2.17 E-04 Kr-87 9.90 E-04 Xe-135 1.18 E-04 Kr-88 8.57 E-04 Xe-137 8.57 E-05 ( j Kr-89 1 6.34 E-C' Xe-138 5.58 E-04 Kr-90 6.66 E-04 Xe-131m 1.46 E-06 Ar-41 i
- Values from Reference 16.
+ See footnotes to Table 3-6.
l l l t 6-5 Rev. 8, 1/94 f i h
l Hatch ODCM Table 6-1 (contd). Attr;oates of Member :f tne P.O.i: Receptor
- orations .
Inside the SITE 80UNO ARY l l Location: Visitors Center, WSW at 694 meters Ace Group Child Occupancy Factors 4.57 x 10 ** (based on 4 hours per year) l
\
l Dispersion and Deposition Parametere: l Parameter Ground-Level Elevated i Undepleted (2/D), s/m3 1.87 E-5 5.00 E-8 , I 1 4.97 E-8 Depleted (i76), s/m 3 1.72 E-5 (D76), n(2 5.47 E-B 2.26 E-9 l I I I Elevated Plume Dose Factors: V, I V, Radionuclide f Radionuclide (mrem /y)/(uci/s) l (mrem /y)/(uci/s)
\
1.47 E-04 Xe-133m 2.34 E-05 Kr-85m Kr-85 2.34 E-06 Xe-133 2.36 E-05 8.27 E-04 Xe-135m 4.49 E-04 Kr-87 2.06 E-03 Xe-135 2.42 E-04 Kr-88 Kr-89 1.78 E-03 Xe-137 1.77 E-04 Kr-90 1.31 E-03 Xe-138 1.16 E-03 Xe-131m 2.98 E-06 Ar-41 1.38 E-03 l 1 I
- Values from Reference 16.
+ See footnotes to Table 3-6.
l i 6-6 Rev. 8, 1/94
Hatch 09_qn CHAPTEP - REPORTS 1 7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT 7.1.1 Recu irement for Reoort s 6.9.1.6, the Annual Radiological In accordance with Technical Specification Environmental Surveillance Report covering the REMP activities during the (A single previous calendar year shall be submitted before May 1 of each year. report fulfills the requirements for both units. ) The material provided shall l be consistant with the objectives outlined in Section 4.1 and Section 7.1.2 of the ODCM, and in Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part ! 50. 7.1.2 Pecort Contents The materials specified in the following sub-sections snail be included in-each Annual Radiological Environmental Surveillance Report: 7.1.2.1 Data f l The report shall include summarized and tabulated results of all REMP samples l required by Table 4-1 taken during the report period, in a format similar to that l contained in Table 3 of the Radiological Assessment Branch Technical Position (Reference 18); the results for any additional samples shall also be reported. In the event that some results are not available for inclusion with the report, the report shall be submitted nottng and explaining the reasons for the missing results; the missing data shall be submitted as soon as possible in a supplementary report. The results for naturally-occurring radionuclides not included in plant effluents need not be reported. 7.1.2.2 Evaluations Interpretations and analyses of trends of the results shall be included in the report, including tne following: (as appropriate) comparisons with pre-studies, operational controls, and previous environmental operational surveillance reports; and an assessment of any observed impacts of the plant operation on the environment. If the measured level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 4-2 is not the result of plant effluents, the condition shall be described as required by Section 4.1.1.2.2. 7-1 Rev. 8, 1/94
Natch ODCM 1 7.1.2.3 Programmatic !aformation i Also to be included in each report are the following a summary description of the REMP; a map (s) of all sampling locations keyed to a table giving distances f l and directions f rom the main stack; the results of land use censuses required by 1 Section 4.1.2; and the result' of licensee participation in the Interlaboratory Comparison Program required 1.; Section 4.1.3. (The report shall include either a sumnary of the results obtained as part of the required Interlaboratory Comparison Program or, for licensees participating in the EPA cross-check program, the EPA program code designations for the plant.) 4 7.1.2.4 Descriptions of Program Deviations be included in eacn Discussions of deviations f rom the established program must report, as follows: 4-1, a 7.1.2.4.1 If the REMP is not conducted as required in Table description of the reasons for not conducting the program as required, and the plans for preventing a recurrence, must be included in the report. If the KDCs required by Table 4-3 are not achieved, the 7.1.2.4.2 contributing factors must be identified and described in the report. If Interlaboratory comparison Program analyses are not y formed 7.1.2.4.3 j as required by Section 4.1.3, the corrective actions taken to prevent a recurrence must be included in the report. l l l l 4 l 7-2 Rev. 8, 1/94
.. . .. -~ -. -
l l Hntch ODCM 7.I ANNUAL RADIOACTIVE EFFLUENT RELEASE RE?CRT I 7.2.1 Recuirement f or Re po r.1 l In accordance witt. Technical Specification 6.9.1.8, the Annual Radioactive Ef fluent Peltase Report cavering the operation of the units during the previous calendar year of operation shall be sabmitted before May 1 of each year. (A , However, the submittal shall single submittal may be made for Units 1 and 2. l specify the releases of radioactive material in liquid and gaseous effluents from ' each unit and solid radioactive waste frem the site.) The report shall include a summary of the quantities of radioactive liquid and gaseous ef fluents and solid waste released from tha units. The material provided shall be consistent with t the objectives outlined throughout this ODCM and the Process Control Program l (PCP) and in conformance with 10 CFR Part 50.3Ca and Section IV.B.1 of Appendix I to 10 CFR Part 50. t i 7.2.2 Report Contents The materials specified in the following sub-sections shall be included in each l Annual Radioactive Effluent Release Report: ( 7.2.2.1 Quantities of Radioactive Materials Released l The report shall include a summary of the quantities of radioactive liquid and l l gaseous effluents and solid waste released from the units as outlined in NRC Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Ef fluents f rom Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, l with liquid and gaseous ef fluent data summarized on a quarterly basis and solid radioactive waste data summarized on a semiannual basis following the format of Appendix B thereof. The report shall include documentation of quantities of radioactive materials in unplanned releases of gaseous and liquid ef fluents f rom l the site to UNRESTRICTED AREAS, tabulated either by quarter or by event, provided that: such liquid releases exceeded 1 C1, excluding tritium and dissolved or entrained noble gases; or such gaseous releases exceeded 150 Ci of noble gases or 0.02 Ci of radioiodines. For gamma emitters released in liquid and gaseous effluents, in addition to the principal gamma emitters for which MDCs are , specifically established in Table 2-3 and Table 3-3, other peaks which are measurable and identifiable also shall be identified and reported. l l 7-3 Rev. 8, 1/94 y rye- e--- -
-e
Hatch ODCM 7.2.2.2 Meteorological Data I f The report shall include an annual summary of hourly meteoro'1gical data collected over the previous year. This annual summary may be either in the form , l of an hour-by-hour listing of wind speed, wind direction, and atmospheric { stability, and precipitation (if measured) on magn < tic tape; or in the form of joint f requency distributions of wind speed, wind direction, and atmospheric f stability. In lieu of submission with the Annual Radioactive Effluent Release i j i Report, the licensee has the option of retaining this summary of required I meteorological data on site in a file that shall be provided to the NRC upon request. 7.2.2.3 Dose Assessments The report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released frem each unit during the previous calendar year. Historical annual average meteorology or the meteorological conditions concurrent with the time of release of radioactive l materials in gaseous effluents (as determined by sampling frequency and This i measurement) shall be used f or determining the gaseous pathway dose. assessment of radiation doses shall be performed in accordance with Sections 2.4, 3.1.3, 3.1.4, 3.4.2, 3.4.3, 5.1, and 5.2. 2.1.3, J
- If a determination is required by Section 5.1.2, the report shall also include an assessment of radiation doses to the likely most expcsed MEMBER OF THE PUBLIC t
f rom reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation; this dose assessment must be performed in accordance with Chapter 5. The report shall also include an assessment of the radiation doses f rom radioactive 11gurd and gaseous ef fluents to MEMBERS OF THE PUBLIC due to their a c'. iv it ie s inside the SITE BOUNDARY (Figure 10-1) during the report period; thic assessment must be performed in accordance with Chapter 6. 7.2.2.4 Solid Radwaste Data For each type of solid waste shipped offsite during the report period, the , 1 following information shall be included:
- a. Container volume,
- b. Total curie quantity (specify whether determined by measurement of estimate),
7-4 Rev. 8, 1/94
Hatch ODQb
- c. Principal radionuclides (specify whethor determined my measurement or !
estimate), evaporator
- d. Type of waste (e.g., spent resin, compacted dry waste, I
! bottocs), Type of container (e.g., LSA, Type A, Type B, Large Quantity), and e. Solidification agent (e.g., cement, urea formaldehyde.) f. f f 7.2.2.5 Licensee Initiated Document Changes I l l Licensee initiated changes phall be submitted to the Nuclear Regulatory Commission as a part of or concurrent with the Annual Radioactive Effluent Such changes to Release Report for the period in which any changes were made. This the ODCM shall be submitted pursuant to Technical Specification 6.17. requirement includes: locations in the radiological 7.2.2.5.1 Any changes to the sampling environmental monitoring program, including any changes made pursuant to Section 4.1.1.2.3. Documentatior of changes made pursuant to Section 4.1.1.2.3 shall include supporting inf ormation identifying the'cause of the unavailability of F samples.
,cluding Any changes to dose calculation locations or pathways, r
7.2.2.5.2 any changes made pursuant to Section 4.1.2.2.2. 7.2.2.6 Descriptions of Program Deviations Discussions of deviations f rom the established pregram shall be included in each report, as follows: l 7.2.2.6.1 The report shall include deviations from Minimum Detectable l concentration (MDC) requirements included in Table 3-3. 7.2.2.6.2 The report shall include deviations from the liquid and gaseous ef fluent monitoring instrumentation operability requirements included in Sections 2.1.1 and 3.1.1, respectively. The report shall include an explanation as to why the inoperability of the liquid or gaseous effluent monitoring instrumentation l was not corrected within the specified time requirement. (This requirement does not include the Service Water System to closed Cooling Water System Dif ferential Pressure channel.) l 7.2.2.6.3 The report shall include notification if the contents within any outside temporary tank exceed the limits of Technical Specification 3.15.1.4 (Unit 1) or Technical Specification 3.11.1.4 (Unit 2). 7-5 Rev. 8, 1/94
( Hatch ODCM i I 7.2.2.7 Major Changes to Radica:tive Waste Trea: rent Systems As required by Sections 2.1.5 and 3.1.6, licensee initiated MAJOR CHANGES TO i RADIOACTIVE WASTE TREATMENT SYSTEMS (liquid and gaseous) shall be reported to the l Nuclear Regulatory Commission in the Annual Radioactive Ef fluents Release Report l covering the period in which the change was reviewed and accepted for ' implementation.I The discussion of each change shall containt i
- a. A sununary of the evaluation that led to the determination that the 50.59; change could be made in accordance with 10 CFR Part i
' b. Suf ficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; i l ' c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;
- d. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous ef fluents that dif f er f rom those previously predicted'in the license application and amendments thereto; t
i e. An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that dif f er f rom those previously estimated in the license application and amendments thereto;
- f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents, to the actual releases for the period prior to when the changes are to be made;
- g. An estimate of the exposure to plant operat;ng pers:nne'. as a result of the change; and
- h. Documentation of the fact that the change was reviewed and found l l
acceptable by the Plant Review Board. I 1 In lieu of inclusion in the Annual Radioactive Effluents Release Report, this same information may be submitted as part of the annual FSAR update. I I 7-6 Rev. 8, 1/94 I l I. I i _ -. . _ _ . _ , . . . , , , _ _ . - ~ _ _ . _ _ . . _ . _ _ , , . _ _ . _ _ . . . . . . _ - . . _ . . .m.-__
Hatch ODCM 7.3 MONTHLY OPERATING REPORT This ODCM establishes no requirements pertaining to the Monthly Operating Report. 7.4 SPECIAL REPORTS Special reports shall be submitted to the Nuclear Regulatory Commission in accordance with Technical Specification 6.9.2, as required by Sections 2.1.3.2, 2.1.4.2, 3.1.3.2, 3.1.4.2, 3.1.5.2, 4.1.1.2.2, and 5.1.2. i l 7-7 pev. 8, 1f94
I l I Match ODCM l 1 CHAPTEF 5 METEOROLOGICAL MODELS The models presented in this chapter are those which were used to compute the specific values of meteorology-related parameters that are ref erenced throughout this ODCM. These models should also be used whenever it is necessary to calculate values of these parameters for new locations of interest. Note la When calculating values of annual average parameters for new locations, use the joint f requency meteorological data presented in Reference 28. Those are the data which were used to compute the specific values of meteorology-related parameters that are referenced throughout this ODCM. Note 2: Although Plant Hatch has no mixed-mode releases, the sections on mixed-mode calculations (8.1.3 and 8.2.3) are in:'. .2ded to preserve section number compatibility with the ODCMs of the other Southern Company nuclear power plante. 8.1 ATMOSPHERIC DISPERSION Atmospheric dispersion may be calculated using the appropriate f orm . of the I sector-averaged Gaussian model. Gaseous release elevations may be considered to ! be either at ground-level, elevated, or mixed-mode. Facility _ release elevations ) for each gaseous release point are as indicated in Table 3-4. f 8.1.1 Ground-Level Releaseg Relative concentration calculations for ground-level releases, or for the ground-level portion of mixed-mode releases, shall be made as follows: 2.032 5 K r /A (X/0)g = { (8.1) N r uj Eg where: (X/Q)G = the ground-level sector-averaged relative concentration for a given wind direction (sector) and distance (s/m ) 3. 2.032 = (2/n)U2 divided by the width in radians of a 22,5' sector, which is 0.3927 radians. l 8-1 Rev. 8, 1/94 l
l Hatch ODCM 3 = the plume depletacn fa:ter fer a.. ras.:n cl. des :t .r e r : nan .:ble gases at a distance r snown in Figure 8-3. For noole gases, the depletion f actor is unity. If an undepleted relative concentration l is desired, the depletion factor is unity. Only depletion by deposition is considered since depletion by radioactive decay would be of little significance at the distances considered. I l Kr = the terrain recirculation factor corresponding to a distance r, taken from Figure 8-2. i np= y the number of hours that wind of wind speed class j is directed into the given sector during the time atmospheric stability category k existed. N= the total hours of valid meteorological data recorded throughout the period of interest for all sectors, wind speed classes, and stability categories. uj = the wind speed (mid-point of wind speed class j) at ground level (m/s). r= the distance from release point to location of interest (m). the vertical standard deviation of the plume concentration Ed= distribution considering the initial dispersion within the building wake, calculated as follows: 7 3 2 'l 2
, d*E (8.2)
Eg = the lesser of: op 1 v 3 ;cg, the vertical standard deviation of the plume concentration cd = distribution (m) for a given distance and stability category k as I shown in Figure 8-1. The stability category is determined by the l vertical temperature gradient AT/Az ('C/100 m). n= 3.1416 i I b= the maximum height of adjacent plant structure (47 m). 1 8-2 Rev. 8, 1/94 i
.V '
Hatch 00CM B.1.2 Elevated Peleases Pelative dispersion calculations for elevated releases, or for .he elevated l ! portion of mixed-mode releases, shall be made as follows: I l l , , 1 l
-h 2.032 K r Ok Djk exp 2 gg,33 (X/0)E "
" ' b .
# 2k, jk Uj *:k i
where: (X/Q)E = the elevated release sector-averaged re'ative concentration for a given wind direction (sector) and distance (s/m 3 ). 6k= the plume depletion factor for all radionuclides other than noble gases at a distance r for elevated releases, as shown in i Figure 8-4, Figure 8-5, and Figure 8-6 For an elevated release, t ' this f actor is stability dependent. For noble gases, the depletion factor is unity. If an undepleted relative concentration is desired, the depletion factor is unity. Only depletion by l deposition is considered since depletion ' yc rac: ' active decay would be of little significance at the distances considered. the number of hours that wind of wind speed clase ; is directed nk= j into the given sector during the time atmospherie stability category k existed. 1
= the wind speed (mid-pcint of wind speed class ;) at the effectzve u) release height h (m/s).
h= the effective height of the release (m), which is calculated as follows:I 1 h=h v -h p (8.4) I Effective release height may be further ad;usted for plume rise in accordance with Section E.4.3.2 of Appendix E to Reference 7. 8-3 Rev. d, 1/94
i l i Hatch ODCM l 1 hy = the height of the release point (m), which is the height of the main stack, 120 m. , hg= the maximum terrain height between the release point and the point 4 of interest (m), from Figure 2.3-12 of Reference 8. 4 All other symbols are as previously defined in Section 8.1.1. i 1 8.1.3 Mixed-Mode Releast_s j Relative dispersion calculations for mixed-mode releases shall be made as follows: 4 i (X/0)y = ( 1 -E) * (X/0)E
- E*(X/0)G (8.5) i where:
1 (X/Q),y = the mixed-mode release sector-averaged relative concentration f or a given wind direction (sector) and distance (s/m )3 . E= the fraction of hours during which releases are considered as 4' ground-level releases, calculated as follows: .I t* W I 1.0 for _# 51.0
")
W# W 2.58 - 1.58 - - f or 1. 0 < _# s 1. 5 r "j - ") E = (8.6)
- W# W 0.3 - 0.06 - __ t o r 1. 5 < _# s 5.0
, Uj. Uj W
0 for # > 5. 0
")
All other symbols are as previously defined. L
, 8-4 Rev. 8, 1/94
Hat: h ODC.M S,2 RELA!!VE DEPOSITION Plume depletion may be calculated using the appropriate form of the sector-Gaseous release elevations may be considered to be averaged Gaussian model. Facility release elevations f or I either at ground-level, elevated, or mixed-mode. in Table 3-4. each gaseous release points are as indicated 8.2.1 Ground-Level Releases Relative deposition calculations for ground-level releases, or for the ground-level portion of mixed-mode releases, shall be made as follows: 2.55 O # K, (8.7) (D/0)G
=
N r [ Ak 3 where: (D/Q)G = the ground-level sector-averaged relative -3 deposition for a given i wind direction (sector) and distance (m ') . radians in a 22.5' sector 2.55 = the inverse of the numrer of [= (2 n/16)-I]. taken from Figure 8-7 for D, = the deposition rate at d; stance r, h ground-level releases (m_1). i into the sector nk = the nunber of hours in which the wind is directed of interest, and during which stability category k exists. All other symbols are as defined previously in Section S.1. 8.2.2 Elevated Releases Relative deposition calculations for elevated releases, or for the elevated portion of mixed-mode releases, shall be made as follows: 2.55 K [ lok Dek) (8.8) (D/0)E 3 - A l l where: i l 8-5 Rev. 8, 1/94 f l
, . _.. _ _ __ _. _ . _ . _ . . _ . _ . . . . _ . . _ .. . . . _ . . _ . . . _ _ _ _ _ _ . = . . . . _ Haten ODCM
- r a
= the elevated-plame sector-averaged re'.a n ve deposit:or.
(D/Q)E given wind direction (sector) and distance (m'2). r, taken from the elevated plume deposition rate at distance Dd= Figure 8-8, Figure 8-9, or Figure 8-10, as appropriate to the plume ef fective release height h defined in section 8.1.2, for stability l class k (m-I) . 1 All other symbols are as defined previously. 8.2.3 Mixed-Mode Peleases i Relative deposition calculations for mixed-mode releases shall be made as j follows:
= (1 -E) * (D/0)g + E
- p/0)G (8.9)
(D/0)3f . where sector-averaged relative deposition for l (D/Q)y = the mixed-mode release a given wind direction (sector) and distance (m-2), l ! E= the fraction of hours during which releases are considered as l l ground-level releases, defined :.n Secti0n 8.1. 3. l All other symbols are as previously defined. I i l l 8-6 gey, 8, 1/94 I i
-- -g-
" y4 , ,
Hatch ODCM l B.3 ELEVATED PLUME DOSE FACTCRS Certain gaseous ef fluent dose calculations recurre the use of the elevated-plume These i noble gas dose parameters B, or V,, which are first defined in Section 3.4. parameters are calculated as follows: l Djk
- Sei a y a, E,
- I ,k ( r ) (8.10) g { ")
0 r N j,k,e i
~
' (8.11) 1.1 K [ lE ' 'I ' ' -
e y '. r N j.k.t "j l where: K= a numerical constant representing the aggregated numerical 4 constants and unit conversions, 2.1 x 10 A, , = the number of photons in energy group a emitted per transformation of radionuclide i (number / decay), p a, = the air energy absorption coefficient for photons in energy group e (m-I). E, = the photon energy assigned to energy group e (MeV).
= the gdimensionless) result of integrating the emissicn and Ike ( r ) over the entire spatia; attenuation of photons of energy group e, activity distrtbution of a plume that has spread under atmospheric stability classification k, for a dose receptor at downwind distance r (see below for calculational method).
fcr 1.1 = the average ratio of the photon energy absorption coef ficient This tissue to that of air over the energy range of interest. ratio converts air dose (rad) to dose equivalent (rem). 8-7 Rev. 8, 1/94
l_ I i Match ODCM i a!, = the tissue energy sceception :ceiter:ent for pnotens an energy group e (cm2 /g), d= the tissue density thickness taken to represent , the depth at which total body dose is received (5.0 g/cm*). All other symbole are as previously defined. the dose For a sector-averaged plume model like that described in Section 8.1, integral Ike(r) i= calculated as follows:
==
A x G( 2, k, r) u A(p g, R) x L dL dz gg,gg; Ikg(r) = - l l B (pg g, pa , R) g 2 *% 0 0 where: L= the upwind or downwind distance from the differential volume element of the plume to the dose receptor point, of the t z= the vertical distance f rom the dif ferential volume element plume to the dose receptor point. R= the total distance from the differential volume element of the plume to the dose receptor point s R = L'+r-Bg = the air dose buildup f actor through a thickness R of air l pa.g ug B = 1 + ) {p g =R f pag in air,.for energy group p, = the total photon attenuation ccefficient e (m*I). G= the function describing the vertical distributton of activity in a plume that has travelled downwind a distance r f rom the point of emission, at an ef fective height h, under stability classification k:
- 8-8 Rev. 8, 1/94 j
l
Hatch 00CM G = exp -(r-h)- ,
,,p l , ( r +
2 h )- 2 2 og , 2 og A= the attenuation and geometric loss factor for photons in energy group s, for the distance R from the dif f erential plume volume element to the dose receptor point: exp {-u, R] R 2n L dL dz = the volume of the differential plume element. (When the 2n is f actored out into the constants in equations (8.10), (8.11), and (8.12), only L dL dz is left.) All other symbols are as previously defined. l A derivation of the model describing the gamma dose rate f rom an elevated finite Numerical methods for evaluating plume is found in Chapter 7 of Reference 19.
- 3. Details of the f
the dose integral are found in Appendix F of Reference numerical methods used there may be found in Reference 20. l t l l l 8-9 Rev. 8, 1/94 l l l
Haten ODCd Table 8-1. Terrain Elevatacn Above Plant Site Grade l 4 This table intentionally left blank. l l l 4 1 1 8-10 pey, 'f, if94
= = * . - + , , , _
. - - . - - . . . . - . . - - - - -. . . - - - - . ~ . _ -. . ~ ~ - . - . , . - - - - .... -- . .. . . . ,
i 1
%tch 00Cw, l
I l l
., 1 1000 'l I I t I )i J t
' /~ I I
i < ! /
/ >
/i ! /
/ I I / 6 L-h ,/
) } ,/
/ / / / -
/ / ,/ /
4
/ # '
/ '
' /
x 100 ,
. / <
i a , i- .m x , s I
.r / / / l t ./
/
/
r ,. / / /! .' i sr i/ / 7 A/Ii L ' - I i w- - g
~
/ '/ / of/ j/ l i M l i ,
/// /?/ V ,-1 M' \ \ l p /' 7
/ / /
// ,!/ j G
l l l i , in
,71 /
^
js: y' ;i , l i
/ /! / if i i
.. 4
/ /i / / .
/ V / l l l !
/l / I l l! l l l l f l l ll l l J
l\ l 10 100 0.1 1.0 PLUME TRAVEL OtSTANCE IKILOMETERS) 1 Range of Vertical Range of Vertical Category Temperature Gradient Temperature Gradient (*C/100 m) ('F/100 ft) A AT/AZ < -1.9 l AT/A* < -1.0 B -1.9 s AT/a2 < -1.7 -1.0 s AT/AZ < -0.9 C -1.7 s AT/AZ < -1.5 ) -0.9 s AT/AZ < -0.8 D -1.5 s AT/A* < -0.5 l -0.8 s AT/AZ < -0.3 E -0.5 s AT/AZ < 1.5 -0.3 s ST/AZ < 0.8 F 1.5 s AT/AZ < 4.0 0.8 s ST/AZ < 2.2 G 4.0 s AT/AZ l 2.2 s AT/AZ This graph is reproduced from Reference E (Figure 1). I F gure 8-1. Vertical Standard Deviation of Mater;al in a Plume (o g) 8-11 Rev. 8, 1/94 _ . _ = . , . _ . . _ - _ _ , .._.,;...._.._- _._._.__-.s
I il l
'aten ODCM d
l l i
)
l l 1 l l 1 l 10
. iii !ii
, , i ,i , , i .i i iiii1 l i . I itii l 1 I I iiIii i iilII 4
1 I il i i l !! \l i I i lIlll l \ \\l l{ e ! k lkk C U j
,N ' l
%s 2 .
i l FN I iiliI , m l g t U . . , ,ii. e i iii, 4i E . .. , , , i, ii g i i gi i u i i i ii,ei i ; i i i iiii 9 I I i Ilill l i l l1litj } l l l{j} { i I I I \IIll I ll!II\kl l l ll jpg
< t t r
ltj
< 3 3 ilm l j ,i lu ,,
0.1 100 1.0 to 0.1 otsTAsoct (Estoutis ms This graph is reproduced from Reference 4. Figure 8-2. Terrain Recirculation Facter (Kg) 8-12 Rev. 8, 1/94
Mater. ODCM
- r. , e i.
i ' l , l l l l- ' il l !! I
$ l o, N , m ,Illh
_! ~,., , um M i iin n
! ,, "! !? N I! ! ,, llH!l' 'I!!!!I IM !!! ! ,, .!!' l .
A i n% i !
- lh !i lho \
i!il!
! i.
I j ',
.i t
!. i 32 g,
I' i !I' .t l l i ii l t i.l' i j ' :l 10.0 100.0 200.0 11 1.: PLUME TRAVEL DtSTANCE (KILOMETER $1 This graph is reprcouced from Reference 5 (T.gure 2). Figure 8-3. Plume Depletten Effect for Ground Level Releases t I 8-13 Rev. 8, 1/94 l I
unten COCM l l 1 I l l 3 ,,
,s
! h N ~EL,.KT TE,1 '
1
,.s Obs I l'! X\
I ! I ,., ! l l 1111 E os $ k h ! l i ,, i l ul l !iul ! l !,s I l lil ll i i ll l l NN1 i ,, ! !l! l l l: i I !\N ll! l l i !! t Il li
\
lli il i i i! I! l l ll i i ! !illi i ll !j i 100 3- 2004 j 1.0 10.0 31 PLUME TRAVEL DLSTANCE IKtLOMETEMS) l l l l This graph is reproduced from Reference 5 (Figure 3). Figure 8-4. Plume Depletion Effect for 30-Meter Beleases 8-14 g,y, g, 1794
Maten 00CM i l
'O i l sTASLt-.
a, !ll! . l iIi iI EUTA Lj W i . I
,g, I,,
ilIN i i I ! l I!!
- i UNSTABLE ,N l
! ! ! '*8# i, i l %%
l ' ' '
J
$ ca y < &
og
! li ll l .
- lIl N i 1 1 A C 04 g_3 3
' ! : N!l t , ; I! : . l ll !! \l 0.1 t . i mi i
' ll !
i i i ii i ! It
!, !, l !!
,. , I, f l. ',
'0 toa 100.0 200.0 PLUME TRAVEL DISTANCE (KILOMETERS)
I I This graph is reproduced from Reference 5 (Tigure 4). Figure 8-5. Plume Depletion Effect for 60-Meter Releases , j 8-15 Rev. 8, 1/94 l i
Mate. ODCM 1.0
% _% NEUTRAL (D) oy
, ,, l I I i \ l
! ., i i !! Hll l l r3M'3NfN\
iI l STABLE (E,F.C) l h g l
; l l NO DEPLETION i \
< (FRACTidN REMAINik = 12 h \
8 6 04 i ll l 1
' \
I . I I r
,a .I i 0.2
' i li '
i li I \ O.1 l 0.1 1 .0 102 1ces zoos PLUME TRAVEL OtSTANCE (KILOMETERS 1 l l i l l This graph is reproduced from Beference 5 (Figure 5). ] Plume Depletion Effect for 100-Meter Releases ! rigure 8-6. 8-16 Rev. 8, 1/94 l
- - - - - - - c . - .n -~ ,-, ,- r-v.a. -a,.m .e ,-------v,emy on,--
._. - ~. . . _ . - -
Hatch ODCM 104 : l l I 104 \ N i l - x i l ' h i 1
' 4 i s. 1
- ' I (
! l \ . I I l e i i ll N t l l l E % !
C l l \ l 3
\ s%
- (
$ 10-5 l
i \. ~
.b ! ! .
. , i i i o
- a. i i ! i ix I i i i X
! c i
! l l l \ l'N 5a I I l-l li i
! l \ t N
l s W 10 4 \ i f I l' I e ;
, i, , ,
i
- t
( ,
' I i i
' l l
! I ll
- I IIl l l i l lllIl !
lllll l l 1- i 10-7 0.1 1.0 10.0 100.0 200.0 PLUME TRAVEL OlST ANCE (KILOMETERS) This graph is reproduced f rorn Ref erence 5 (Figure 6). Figure 8-7. Relative Deposition f or Ground-Level Releases 8-17 Rev. 8, 1/94
t watch 00Cw l l Id f i UNSTABLE (A,B.C) l I - - - - l 104 '
- m 2 1 %
w r v s / 't ), w -s f 'h i ( l w [ \\ NEUTRAL ! NNs s 4 % > z f NEUTRAL (D)
\\g h Ig4 Q STABLE g3 / % v f
,m x
O I rY (
- a. / 2
> / h N E '
s'N \ w / /
$ / f X s i s* / / N %
h
- h STABLE (E,F.G) (
104 l' N l l ',
' I l
I 1 I ! I l
' 4 I I T f i /
/
10-7 0.1 1.0 10.0 100D 2004 l PLUME TRAVEL DIST ANCE (KILOMETERS) , I 1 1 r This graph is reproduced frcm Reference 5 (Figure 7). Figure B-8. Relative Deposition for 30-Meter Releases 8-18 gev. 8, . 94
1 l Match O Q j 1 104
; f e UNSTABLE (AA,C)
[ \ !
/ N A I
i f NEUTRAL (D)
% qh I
\
1(y5 / j , ,
/ A,s ,
e ,/ , , * . i i NA j p / , ,
/ "If' a
! If ! I ii i \1% %_ UNSTABLE l z i I / I ! I I il i K X l
$ l l[ l ll NEUTRAL \ \
$ /I \ \
104 '
- i. 1 .
1 , i s . i ,
, i ! , ! i i . i ;
o -1 . ! if ! i . i ! J
$ t i ti /
o il i I I . y If ' I ! l I ll / a
! ! !l lll f {
$ Ii l I 1
'l l; l- '
i I l l STABLE (E.F.G) 10-7 ' l 1 i I , c r , i i i! I ii1 i+:i i I
!! ' '!! I
! l ! I!it ! iI !
i 1 I i l!!II ! i! I I/ 1l l 4 l l
' l lll l I i llllI'! ! YII l l ll
. 30-8 i I i l !i ll [ A 0.1 1.0 10.0 100.0 200.0 PLUME TRAVEL DISTANCE (KILOMETERS) i 1 l 1 1 This graph is reproduced from Reference 5 (Figure 8). + rigure 8-9. Relattve Deposition for 60-Meter Releases 8-19 Rev. 8, 1/94 l a.
- ,,,,,,...,w.. . . . ~ . , . , , ,
l ! I Hatch 00CM i l i l l l 104 i i t t I i UNSTABLE (A,B,C) i .l
/ \ N l l
\
N 10-5 ; A m l r' %
- i i f i < * >
w ! ii E . i e f i NN6 I i l' y i
,f
/
i.,, f NEUTRAL (D) , i x4 ; , y / i ! Iiij j/ i i i
! 61 I +\ ih e / i i Itil I/ i i .l l11 -
N i l'N t i / I ll i / l ll ll : IlX h N h l ,
}
j Q 10-6 , , ! ,
- i o ,
*.. t u .
.I l 8s I I
i i !F i e i e i ii ! .i i i I i t i
$ / t i Il I i . tit i .
l y / l I l If llj l I ! lll ! t i i ll f l E m . I Il/Illll I I IIl l STABLE (E,F,G) l !
! 1 :ll Sf ! !
l NO DEPLETION l l l 10-7 , ,
!. I I I
1 1 1 I) . . i. . ! i )
. !/ i . . .. . , i .ii . , ,
i i If 1 t l!!Ii I I !!IIi , I i j j f ; i i ll
' I ii'l- i!!! 4 I k !l l l I ! i !I' ! lI
!! l l ,g i liili ! ii '
l! l 0.1 1.0 10.0 100.0 200.0 PLUME TRAVEL DISTANCE (KILOMETERS) t This graph is reproduced from Peference 5 (Figure 9). Figure 8-10. Relative Deposition for 100-Meter (or Greater) Releases B-20 Rev. 8, 1/94
Hatch CDCM CHAPTER 9 METHODS AND PARAMETERS FOR CALCULATICN OF GASEOUS EFFLUENT PATHWAY DOSE FACTORS, R ,,p) i 9.1 INHALATION PATHWAY FACTOR 3 ! For the inhalation pathway, R,3p; in (mrem /y) por (pCi/m ) is calculated as
)
follows (Reference 1, Section 5.3.1.1): l l l R aipj
- Kg (BR)a * ( DIA) asj (9 1) l i
where: l the units conversion factor: ICD pCi/pC1. Kg = 3 the breathing rate of receptor age group a, in m /y, f rom (BR)a = l Table 9-5. I t l the inhalation dose factor for receptor age group a, (DFA),y = l radionuclide i, and organ j, in mrem /oCi, f rorr Table 9-7 l l through Table 9-10. l l 6 t I 9-1 Rev. 8, 1/94 f
Hatch ODCM GROUND PLANE PATHWAY FACTOR 9.1 1 For the ground plane external exposure pathway, R,, p) in (m2 mrem /y) per (uci/s) is calculated as follows (Reference 1, Section 5.J.l.2):
-kj t
~* (9.2)
R aipj " K1*K2* (88f) * (UTG)ij
- 3, I '
l I where: the units conversion factor: 106 pCi/yCi. Kg = K the units conversion factor: !!760 h/y. 7= i (SHF) = the shielding factor due to structure (dimensionless). 1 i The value used for (SHF) is 0.7, from (Reference 3, Table E-15). (DFG)g = the ground plane dose f actor for radionuclide 1 and organ 3 j, in (mrem /h) per ( pC i /m- ) , from Table 9-15. Dose factors are the same for all age groups, and those for the total body also apply to all organs other than skin. i, = the radioactive decay constant for radionuclide i, in s -I. Values of 1, used in ef fluent calculations should be l l based on decay data f rom a recognized and current source, ! such as Reference 26. t = the exposure time, in s. Tne value used for t is 4.73 = 10 0 s (= 15 y), from (Reference 1, Section l 5.3.1.2). I t 1 9-2 Rev. 8, 1/94
i Hatch ODCM 9.3 GARDEN VEGETATION PATHWAY FACTOR ) For radionuclides other than trit tum in the garden vegetation consumption Rjj 2 (uc i/s) is calculated as follows pathway, ap in (m ,aree/y) per (Reference 1, Section 5. 3.-l . 5 ) : p ,,PJ.
. gg. *
(DTL )ajj yv g x; ,1,) (9.3)
-Xi 'L (Ual IL*
- Ua5 '8 * ^ s 'hv )
l where: the units conversion factor: 106 pCi/uci. Kg = r= the f raction of deposited activity retained on the edible f parts of garden vegetation (dimensionless). The value f 1.0 for radiotodines and 0.2 for I used for r is particulates, from (Reference 3, Table E-1). Yy = the areal density (agricultcral productivity) of growing leafy garden vegetation, in kg/m2 , from Table 9-1. 1, = the radioactive decay constant for radionuclide i, in s -I. Values of 1, used in ef fluent calculations should be based on decay data f rom a recognized and current scurce, such as Reference 26. l = the rate constant for remova'. cf activity on leaf and w 1 plant surfaces by weathering, in s'I , from Table 9-1. f f the ingestion dose factor for receptor age group a, (DFL),9 = radionuclide i, and organ ), in mrem /pci, f rom Table 9-11 through Table 9-14. U,L = the consumption rate of f resh leaf y garden. vegetation by a receptor in age group a, in kg/y, from Table 9-5. I Rev. 6, 1/94 9-3 l i , ,,,
. . = _ - .. .. - . . - . -
Hatch ODCM U,5 = the consumpta:n rate Of stere: garden vegetatt n cy a receptor in age group a, in kg/y, from Table 9-5. ft= the fraction of the annual intake of fresh leafy garden vegetation that is grown locally (dimensionless), f t .m Table 9-1. f g
= the fraction of the annual intake of stored garden vegetation that is grown locally (dimensionless), from Table 9-1.
the average time between harvest of fresh leafy garden tt= vegetation and its consumption, in s, from Table 9-1, thy = the average time between harvest of stored garden vegetation and its consumption, in s, from Table 9-1. ,
)
For tritium in the garden vegetation consumption pathway, R in (mrem /y) J Section 5.3.1.5), per (uci/m3 ) is calculated as follows (Reference 1, I based on the concentration in air rather than deposition onto the ground: i K3 = (DFL)a;; *{Ual IL
- UaS f g' (9 4) \
R aspj " Kg
- 0 'S * { H$h l I
t l l where: l the units conversion factor: *0 3 g/kg. K3= H= the absolute humidity of atmospheric air, in g/m ,3 from Table 9-1. 1 0.75 = the f raction of the mass of total garden vegetation that is water (dimensionless), c l ! 0.5 = the ratio of the specific activity of tritium in garden l vegetation water to that in atmospheric water l (dimensionless), and other parameters are as defined above. 9-4 Rev. 8, 1/94 i y -
.y w <.-c----w. --,..-,....w.-.- - - , , , , . e - . . - . _y-, mmc +,,,- _.- ,-,..visory-, c. w.w e,c, r--%w-,vv$
Hatch ODCM Table 9-1. Miscellaneous Par ameters f or tr.e Garder. Veget ation Patr. ay The f ollowing parameter values are for use in calculating R ,p) g for the garden vegetation pathway only. The terms themselves are defined in Section 9.3. Value Reference Parameter 2.0 kg/m 2 Ref. 3, Table E-15 Yy 5.73 x 10'I s *I Ref. 1, page 33 1, (14-day half-life) 1.0 Ref. 1, page 36 ft f O.76 Ref, 1, page 33 E tt 8.6 x 104 s Ref. 3, Tacle E-15 (1 day) t bv 5.18 x 106 s Ref. 3, Table E-15 (60 days) H 8 g/m 3 Ref, 3 i f 1 l 9-5 Rev. 8, 1/94
Wa:ch CDCM 9.4 GRASS-COW-MI' K PATHWAY FACTO? For radionuclides other than tritium in the grass-cow-milk pathway, Ra ,g in (m2 mrem /y) per (pci/s) is calculated as follows (Reference 1, Section 5.3.1.3): R aipj
- Kg *
- Oy
- Uap *Imi * (DFL)ay (9.5) e i lhm .
r fp i s, (1 - fp f,) Yp Ys L t l l where: the units conversion factor: 106 pCi/ sci. ( Kg = r= the f raction of deposited activity retained en the edibia parts of vegetation (dimensionless). The value used for r is 1.0 for radiciodines and 0.2 for particulates, from (Reference 3, Taole E-1) . l I the radioactive decay constant for radionuclide 1, in l 1, = l s -I. Values of 1, used in ef fluent calculations should be based on decay data f rom a recognized and current source, such as Reference 26. l.= the rate constant for removal of activity on leaf and w plant surfa:es by weathering, in s -I, frem Tacle 9-2. the cow's consumption rate et feed, in kg/d, from Qp = i Table 9-2. l the consumption rate of cow milk by a receptor in age U,p = group a, in L/y, from Table 9-5. F uu = the stable element transfer coefficient applicable to radionuclide i, for cow's milk, in d/L, from Table 9-6. 9-6 Rev. 8, 1/94 l 1
Hatch ODCM
= the Angestico dose facter for receptor age group a, (C F1.) q radionuclide i, and organ 3, in mrem /pci, f rom Table 9-11 through Table 9-14.
f = the fraction of th- year that the cow is on pasture p (dimensionless), from Table 9-2. f, = the fraction of the cow's feed that is pasture grass while the cow is on pasture (dimensionless), from Table 9-2. Y = the areal density (agricultural productivity) of growing p pasture feed grass, in kg/m2 , from Table 9-2. Y 3
= the areal density (agricultural productivity) of growing stored feed, in kg/m2 , from Table 9-2.
thm = the transport time from harvest of stored feed to its consumption by the cow, in s, from Table 9-2. tf= the transport time from consumption of feed by the cow, to consumption of milk by the receptor, in s, from Table 9-2 For tritium in the grass-cow-milk patnway, R in (mrem /y) per (uCi/m 3) is calculated as follows (Reference 1, Section 5.3.1.5), based on the concentration i'n air rather than deposition onto the ground: R a ,p, = Kg < K3 Of C ap f ,,a - TDT aq :.M , (9.6)
- where
i K3= the units cor. version factor: 103 g/kg. H= the absolute humidity of atmospherte air, in g/m 3 , from Table 9-2. 9-7 Rev. 8, 1/94
f l i Hat " ODCM j 0.75 = the fra::2cn of t r.e tr.a s a 2: ::a, vegetat.:r. : a: is l water (dimenstonless). 0.5 = the ratio of the specific activity of tritium in vegetation water to that in atmospheric water (dimensionless). and other parameters are as defined above. l i l l ? l l
\
l l l l t 9-8 Fev. S, 1/94 i
Hat :r. CDC1
, Tacle 9-2. Miscellaneous Parametero for the ~.; r a s s - C:w-M a .' 4 Patnway The following parameter values are for use in calculating R,jj p for the grass-cow-milk pathway only. The terms themselves are defined in Section 9.4.
i . Parameter Value Reference 4 1, 5.73 x 10~7 s'I Ref. 1, page 33 (14-day half-life) Qp 50 kg/d Ref. 3, Table E-3 f 1.0 Ref. 1, page 33 p i f, 1.0 Ref. 1, page 33 j i , Y O.7 kg/m' Ref. 3, Table E-15
, p Y3 2.0 kg/m 2 Ref. 3, Table E-15 a
thm 7.78 = 106 s Ref. 3, Table E-15 (90 days) tg 1.73 x 105 s Ref. 3, Table E-15 (2 days) H 8 g/m 3 Ref. 3 i i J i 9-9 Rev. 8, 1/94
4 Hatch 00CM 9.5 G RA S S-GO AT-M I L K PATHWAY FACTOR For radionuclides other than tritiutn in *.he grass-goat-milk pathway, R,,p) in (m2
- mram/y) per (pci/s) is calculated as folleas (Reference 1, Section 5.3.1.3):
Rapj; = Kg * *Of*Uap *Tmi ' (DTL ) aij (Ai+Aw) f (9.7)
-Ai /6n lf p f,
(1 - f p I) s e , - Ai if Y Ys
. p l
where: Kg = the units conversion factor: 106 pC1/uC1. ! r= the f raction of deposited activity retained on the edible parts of vegetation (dimensionless) . The value used for r is 1.0 for radiciodines and 0.2 for particulates, from (Reference 3, Table E-1). 1 1, = the radioactive decay constar.: for radionuclide i, in s'I - Values of 1, used in ef f'ue-: calculations should be based on decay data f rom a reccgnized and current source, such as Reference 26. 1, = the rate constant for remcval Of activity on leaf and plant surfaces by weathering, .n s -I, from Table 9-3. Op = the goat's consumption rate of feed, in kg/d, from Table 9-3. U,p = the consumption rate of goat milk by a receptor in age group a, in L/y, from Table 9-5. Fg= g the stable element transfer coefficient applicable to radionuclide i, for goat's milk, in d/L, from Table 9-6. 9-10 Rev. 8, 1/94
~. -_ __ __
Hat:- ODCM i :se f ?. r *. O r ';r recept 1 age gr +F a, t r.e :.ngest. n j (DFL)q = radionuclide t, and organ ), in mrem /pci, f rom Tacle 9-11 i through Table 9-14. f = the fraction of the year that the goat is on pasture p (dimensionless), from Table 9-3. l f, = the fraction of the goat's feed that is pasture grass 1 f while the goat is on pasture (dimensionless), from l ' Table 9-3. i ' Y p
= the areal density (agricultura' productivity) of growing pasture feed grass, in kg/m , from Table 9-3.
Y = the areal density (agr;;ultura. productivityi of 7:c.;ng 3 stored feed, in kg/m 2, from Tacle 9-3. tg = the transport time from harvest of stored feed to its l i l consumption by the goat, in s, from Table 9-3. l tr = the transport time from consumption of feed by the goat, to consumption of milk by the receptor, in s, from Table 9-3. 1 3 For tritium in the grass-goat-milk pa t h.'a y , R i r. (mrem /y) per i_C;/m ) ( is calculated as follows (Reference 1, Secti:n 5.3.1.5), based en the concentration in air rather than deposition onto the ground: Ra,p; Kg K3 Of U 4 F ,,,, - s :F: > ay ' : . ? S * !' C ' 5 5, (9.8) H l l where: [ the units conversion f actor: 103 g/kg, K3= i H= the absolute humidity of atmospheric air, in g/m 3, from Table 9-3. 9-11 Rev. 5, 1/94 t i
Haten 00C.w. 0.75 = the fraett:n of :ne Aas :f :::ai vegetation r. a t .s water (dtmenstonless). 0.5 = the ratio of the specific activity of tritium in vegetation water to that in atmospheric water (dimensionless), and other parameters are as defined above. i l 9-12 Rev. 8, 1/94
Hatch ODCM I Taole 9-3. Miscellaneous P a r a.~.e '. e r s for the Crass-00at-Mala Pathway 1 The following parameter values are for use in calculating R gj p for the grass-goat-milk pathway only. The terms themselves are defined in Section 9.5. Parameter Value Reference 1, 5.73 x 10*7 s'I Ref. 1, page 33 4 (14-day half-life) 6 kg/d Ref. 3, Table E-3 Qp fp 1.0 Ref. 1, page 33 l I f, 1.0 Ref. 1, page 33 Y C.? kg/m 2 Ref. 3, Table E-15 p Y3 2.0 kg/m 2 Ref. 3, Table E-15 thm 7.78 x 106 s Ref. 3, Table E-15 (90 days) tf 1.73 = 105 s Ref. 3, Table E-15 (2 days) H 8 g/m 3 Ref. 3 1 l i 1 l l l 9-13 Rev. 8, 1/94
Hate.- aDCM 3.6 G RA S S -COW-ME AT PATHWAY FACTOR For radionuclides other than tritium in the grass-cow-meat pathway, R,qy in (m2* mrem /y) per (pCi/s) is calculated as follows (Reference 1, section 5.3.1.4): i i l R ap j;
=
Kg * *Of*Uap
- I,f * (DFL)ay
-k; thm (9.9) f p f, (1 - I p f,) e
. . .e - Aj rf i Yp Ys l
i j where: Kg = the units conversion factor: 106 pCi/uct. r= the f raction of deposited activity retained on the edible parts of vegetation (dimensionless) . The value used for r is 1.0 for radiciodines and 0.2 for particulates, from (Reference 3, Table E-1). 1, = the radioactive decay constant for radionuclide 1, in s *I. Values of 1 used in ef fluent calculations should be 3 based on decay data f rem a recogn; zed and current source, such as Peference 26. i l Ig = the rate ccnstant for removal of activity on leaf and plant surfa:es by weate.er.ng, : r. s -l, from Table 9-4. Op = the cow's consumption rate of feed, in kg/d, from Table 9-4. U,p = the consumption rate of meat by a receptor in age group a, in kg/y, from Table 9-5. Fg = the stable element transfer coefficient applicable to radionuelade 1, for meat, in d/kg, from Table 9-6. 9-14 Rev. 8, 1/94
Hatch ODCM a. (DFL)q = ine ingest;;n c:se fa;::r fr receptor age gr:sp radtonuclide i, and organ ;, in mrem /pci, f rom Table 9-11 through Table 9-14. f = the fraction of the year that the cow is on pasture p (dimensionless), from Table 9-4. f,= the fraction of the cow's feed that is pasture grass while the cow is on pasture (dimensionless), from Table 9-4. Yp = the areal density (agri:ultural productivity) of growing pasture feed grass, in kg/m , from Table 9-4. Y, = the areal density (agr::ultura'. productivity) of growing stored feed, in kg/m 2, from Table 9-4. , i l tg = the transport time fr:m harvest of stored feed to its consumption by the cow, in s, from Table 9-4. tr = the transport time from consumption of feed by the cow, I to consumption of meat by the receptor, in s, from i l Table 9-4. For tritium in the grass-cow-meat pathway, R in (mrem /y) per (uC1/m3) is calculated as follows (Reference 1, section 5.3.1.4), based on the l concentration in air rather than deposition onto the ground:
,^5 R agp, =
F} K3*Of'Uap ~ ,; * :T:>ay 3.75 lLl.H (9.10) where: the units conversion factor: 103 g/kg. K3= 3 H= the absolute humidity of atmospheric air, in g/m , from Table 9-4. 9-15 Rev. 8, 1/94
Hatch CDCM 0.75 = the f r a r t .c r. of :ne rass of : :: .s . e g e t a: t er. :na: .s water (dt. tensionless). 0.5 = the ratio of the s pecific activity of tritaum in vegetation water to that in atmospheric water (dimensionless). and other parameters are as defined above. l 9-16 Rev. 8, 1/94
Haten ODCM Tac'e
. 9-4. M;scellanecus Para.Teters f:r *.ne Grass-C:w-Mea; Path.ay The following parameter values are for use in calculating R,3py for the grass-cow-meat pathway only. The terms themselves are defined in Section 9.6.
Parameter Value Reference 1, 5.73 x 10*7 s'I Ref. 1, page 33 (14-day half-life) Op 50 kg/d Ref. 3, Table E-3 f 1.0 Ref. 1, page 33 p f 3 1.0 Ref. 1, page 33 Y p 0.7 kg/m 2 Ref. 3, Table E-15 Ys 2.0 kg/m 2 Ref. 3, Table E-15 thm 7.78 x 106 s Ref. 3, Table E-15 (90 days) tf 1.73 x 106 s Ref. 3, Table E-15 (20 days) H 8 g/m 3 g,f, 3 L l l 9-17 gey, e, 1/94
Hatch ODCM Tacle 9-5. Individua; Usage Fa: tors Receptor Age Group Usage Factor Infant Child Teenager Adult Milk Consumption Rate, U,P 33o 33o 400 310 (L/y) Meat Consumption Rate, U,P 41 65 O 110 (kg/y) Fresh Leafy Garden Vegetation Consumption Rate, U,L 0 26 42 64 (kg/y) Stored Garden Vegetation Consumption Rate, Us a 0 520 630 520 ; (kg/y) I Breathing Rate, (BR), 1400 3700 8000 8000 (m3 fy) I i All values are from Reference 3, Table E-5. 9-18 Rev. 8, 1/94
Haten ODCM able 9-6. Stable Element Transfer Data Cow Milk Goat Milk Meat Element *
- pm (d/L) Fm (d/L)* Fg (d/kg)
H 1.0 E-02 1.7 E-01 1.2 E-02 C 1.2 E-02 1.0 E-01 3.1 E-02 Na 4.0 E-02 4.0 E-02 3.0 E-02 P 2.5 E-02 2.5 E-01 4.6 E-02 Cr 2.2 E-03 2.2 E-03 2.4 E-03 Mn 2.5 E-04 2.5 E-04 8.0 E-04 Fe 1.2 E-03 1.3 E-04 4.0 E-02 Co 1.0 E-03 1.0 E-03 1.3 E-02 Ni 6.7 E-03 6.7 E-03 5.3 E-02 Cu 1.4 E-02 1.3 E-02 8.0 E-03 Zn 3.9 E-C2 3.9 E-02 3.0 E-02 Br 5.0 E-02 5.0 E-02 2.6 E-02 Rb 3.0 E-02 3.0 E-02 3.1 E-02 Sr 8.0 E-04 1.4 E-02 6.0 E-04 Y 1.0 E-05 1.0 E-05 4.6 E-03 Zr 5.0 E-06 5.0 E-06 3.4 E-02 Nb 2.5 E-03 2.5 E-03 2.8 E-01 Mo 7.5 E-03 7.5 E-03 8.0 E-03 Tc 2.5 E-02 2.5 E-02 4.0 E-01 Ru 1.0 E-06 1.0 E-06 4.0 E-01 Rh 1.0 E-02 1.0 E-02 1.5 E-03 Ag 5.0 E-02 5.0 E-02 1.7 E-02 Sb 1.5 E-03 1.5 E-03 4.0 E-03 Te 1.0 E-03 1.0 E-03 7.7 E-02 I 6.0 E-03 6.0 E-02 2.9 E-03 Cs 1.2 E-02 3.0 E-01 4.0 E-03 Ba 4.0 E-04 4.0 E-04 3.2 E-03 La 5.0 E-06 5.0 E-06 2.0 E-04 Ce 1.0 E-04 1.0 E-04 1.2 E-03 Pr 5.0 E-:6 5.0 E-06 4.7 E-03 Nd 5.0 E-06 5.0 E-06 3.3 E-03 W 5.0 E-04 5.0 E-04 1.3 E-03 Np 5.0 E-06 5.0 E-06 2.0 E-04 i i
* - Values from Reference 3 (Table E-1) except as follows:
Reference 2 (Table C-5) for Br and Sb.
+ - Values from Reference 3, Table E-2 fer H, C, P, Fe, I Cu, Sr, I, and Cs in goat milk, and Table E-1 for all other elements in cow milk, except as follows:
Reference 2 (Table C-5) for Br and Sb in cow milk. l 9-19 Rev. 8, 1/94 I i I
Haten ODCM
*able 9-7. Inhalation Dose Fac*:rs for tne Infant Age Group l Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 4.62E-07 4.62E-07 4.62E-07 4.62E-07 4.62E-07 4.62E-07 C-14 1.89E-05 3.79E-06 3.79E-06 3.79E-06 3.79E-06 3.79E-06 3.79E-06 Na-24 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 i
i P-32 1.45E-03 8.03E-05 5.53E-05 No Data No Data No Data 1.15E-05 Cr-51 No Data No Data 6.39E-08 4.11E-08 9.45E-09 9.17E-06 2.55E-07 Mn-54 No Data 1.81E-05 3.56E-06 No Data 3.56E-06 7.14E-04 5.04E-06 ! Mn-56 No Data 1.10E-09 1.58E-10 No Data 7.86E-10 8.95E-06 5.12E-05 i Fe-55 1.41E-05 8.39E-06 2.38E-06 No Data No Data 6.21E-05 7.82E-07 Fe-59 3.69E-06 1.68E-05 6.77E-06 No Data No Data "' . 2 5 E - 0 4 1.E-05 I Co-58 No Data 8.71E-07 1.30E-06 No Data No Data 5.55E-04 7.95E-06 l l Co-60 No Data 5.73E-06 8.41E-06 No Data No Data 3.22E-03 2.28E-05 Ni-63 2.42E-04 1.46E-05 8.29E-06 No Data No Data 1.49E-04 1.73E-06 Ni-65 1.71E-09 2.03E-10 8.79E-11 No Data No Data 5.80E-06 3.58E-05 Cu-64 No Data 1.34E-09 5.53E-10 No Cata 2.84E-09 6.64E-06 1.07E-05 Zn-65 1.38E-05 4.47E-05 2.22E-05 No Data 2.32E-05 4.62E-04 3.67E-05 l Zn-69 3,85E-11 6.91E-11 5.13E-12 No Data 2.87E-11 1.05E-06 9.44E-06 Br-83 No Data No Data 2.72E-07 No Data No Data No Data No Data Br-84 No Data No Data 2.86E-07 No Data No Data No Data No Data Br-85 No Data No Data 1.46E-08 No Data No Data No Data No Data
- Rb-86 No Data 1.36E-04 6.30E-05 No Data No Data No Data 2.17E-06 l
l Rb-88 No Data 3.98E-07 2.05E-07 No Data No Data No Data 2.42E-07 Rb-89 No Data 2.29E-07 1.47E-07 No Data No Data No Data 4.87E-08 Sr-89 2.84E-04 No Data 8.15E-06 No Data No Data 1.45E-03 4.57E-05 j Sr-90 2.92E-02 No Data 1.8SE-03 No Data No Data 8.03E-03 9.36E-05 , 1 Sr-91 6.83E-08 No Data 2.47E-09 No Data No Data 3.76E-05 5.24E-05 l All values are in (mrem /pci inhaled). They are obtained from l Reference 3 (Table E-10). Neither Reference 2 nor Reference 3 l I j contains data for Rh-105, Sb-124, or Sb- 125. l 9-20 Rev. 8, 1/94 l l I
Hat:t CDCM
!aole 9-7 (contd). Inhalata n Dose Factors for tne : .f an Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 7.50E-09 No Data 2.79E-10 No Data No Data 1.70E-05 1.00E-04 Y-90 2.35E-06 No Data 6.30E-08 No Data No Data 1.92E-04 7.43E-05 Y-91m 2.91E-10 No Data 9.90E-12 No Data No Data 1.99E-06 1.68E-06 Y-91 4.20E-04 No Data 1.12E-05 No Data No Data 1.75E-03 5.02E-05 Y-92 1.17E-08 No Data 3.29E-10 No Data No Data 1.75E-05 9.04E-05 Y-93 1.07E-07 No Data 2.91E-09 No Data No Data 5.46E-05 1.19E-04 l
i Zr-95 8.24E-05 1.99E-05 1.45E-05 No Data 2.22E-05 1.25E-03 1.55E-05 Zr-97 1.07E-07 1.83E-08 8.36E-09 No Data 1.85E-08 7.88E-05 1.00E-04 t l Nb-95 1.12E-05 4.59E-06 2.70E-06 No Data 3.37E-C6 3.42E-04 9.05E-06 Mo-99 No Data 1.18E-07 2.31E-08 No Data 1.89E-07 9.63E-05 3.48E-05 Tc-99m 9.98E-13 2.06E-12 2.66E-11 No Data 2.22E-11 5.79E-07 1.45E-06 Tc-101 4.65E-14 5.88E-14 5.80E-13 No Data 6.99E-13 4.17E-07 6.03E-07 Ru-103 1.44EW 6 No Data 4.85E-07 No Data 3.03E-06 3.94E-04 1.15E-05 Ru-105 8.74E-10 No Data 2.93E-10 No Data 6.42E-10 1.12E-05 3.46E-05 Ru-106 6.20E-05 No Data 7.77E-06 No Data 7.61E-05 8.26E-03 1.17E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 7.13E-06 5.16E-06 3.57E-06 No Data 7.80E-26 2.62E-03 2.36E-C5 Sb-124 No Data No Data No Data No Data No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 3.40E-06 1.42E-06 4.70E-07 1.16E-06 No Data 3.19E-04 9.22E-06 Te-127m 1.19E-05 4.93E-06 1.48E-06 3.4BE-06 2.6BE-05 9.37E-04 1.95E-25 Te-127 1.59E-09 6.81E-10 3.49E-10 1.32E-09 3.47E-09 7.39E-06 1.74E-05 Te-129m 1.01E-05 4.35E-06 1.59E-06 3.91E-06 2.27E-05 1.20E-03 4.93E-05 Te-129 5.63E-11 2.48E-11 1.34E-11 4.82E-11 1.25E-10 2.14E-C6 1.88E-05 Te-131m 7.62E-08 3.93E-08 2.59E-08 6.38E-08 1.89E-07 1.42E-04 8.51E-05 Te-131 1.24E-11 5.87E-12 3.57E-12 1.13E-11 2.85E-11 1.47E-06 5.87E-06 9-21 Rev. E, 1/94
i Haten ODCM l Tacle 9-7 teentd). Inhalation Dose Fa:: cts fcr tne :nfant Age Gr:ap Bone Liver T. Body Thyroid Kidney Lung CI-LLI Nuclide i Te-132 2.66E-07 1.69E-07 1.26E-07 1.99E-07 7.39E-07 2.43E-04 3.15E-05 , i I-130 4.54E-06 9.91E-06 3.98E-06 1.14E-03 1.09E-05 No Data 1.42E-06 ; I 2.71E-05 3.17E-05 1.40E-05 1.06E-02 3.70E-05 No Data 7.56E-07 I-131 1.21E-06 2.53E-06 8.99E-07 1.21E-04 2.82E-06 No Data 1.36E-06 I-132 9.46E-06 1.37E-05 4.00E-06 2.54E-03 1.60E-05 No Data 1.54E-06 I-133 I-134 6.5BE-07 1.34E-06 4.75E-07 3.18E-05 1.49E-06 No Data 9.21E-07 I-135 2.76E-06 5.43E-06 1.98E-06 4.97E-04 6.05E-06 No Data 1.31E-06 Cs-134 2.83E-04 5.02E-04 5.32E-05 No Data 1.36E-04 5.69E-05 9.53E-07 Cs-136 3.45E-05 9.61E-05 3.78E-05 No Data 4.03E-05 8.40E-06 1.02E-06 Cs-137 3.92E-04 4.37E-04 3.25E-05 No Data 1.23E-04 5.09E-05 9.53E-07 Cs-138 3.61E-07 5.58E-07 2.84E-07 No Data 2.93E-07 4.67E-08 6.26E-07 Ba-139 1.06E-09 7.03E-13 3.07E-11 No Data 4.23E-13 4.25E-06 3.64E-05 Ba-140 4.00E-05 4.00E-08 2.07E-06 No Data 9.59E-09 1.14E-03 2.74E-05 Ba-141 1.12E-10 7.70E-14 3.55E-12 No Data 4.64E-14 2.12E-06 3.39E-06 Ba-142 2.84E-11 2.36E-14 1.40E-12 No Data 1.36E-14 1.11E-06 4.95E-07 La-140 3.61E-07 1.43E-07 3.68E-08 No Data No Data 1.20E-04 6.06E-05 La-142 7.36E-10 2.69E-10 6.46E-11 No Data No Data 5.87E-06 4.25E-05 Ce-141 1.98E-05 1.19E-05 1.42E-06 No Data 3.75E-06 3.69E-04 1.54E-05 Ce-143 2.09E-07 1.38E-07 1.58E-08 No Data 4.03E-08 8.30E-05 3.55E-05 Ce-144 2.28E-03 8.65E-04 1.26E-04 No Data 3.84E-04 7.03E-03 1.06E-04 Pr-143 1.00E-05 3.74E-06 4.99E-07 No Data 1.41E-06 3.09E-04 2.66E-05 Pr-144 3.42E-11 1.32E-11 1.72E-12 No Data 4.80E-12 1.15E-06 3.06E-06 Nd-147 5.67E-06 5.81E-06 3.57E-07 No Data 2.25E-06 2.30E-04 2.23E-05 W-187 9.26E-09 6.44E-09 2.23E-09 No Data No Data 2.83E-05 2.54E-05 Np-239 2.65E-07 2.37E-08 1.34E-08 No Data 4.73E-08 4.2SE-05 1.78E-05 9-22 Rev. 8, 1/94
l l I Hatch ODCM j Table 9-8. Inhalation Dose Ts:: Ors for tne Dn.'d Age Gr:;p Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI , l H-3 No Data 3.04E-07 3.04E-07 3.04E-07 3.04E-07 3.04E-07 3.04E-07 C-14 9.70E-06 1.82E-06 1.82E-06 1.82E-06 1.82E-06 1.82E-06 1.82E-06 , Na-24 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-06 P-32 7.04E-04 3.09E-05 2.67E-05 No Data No Data No Data 1.14E-05 Cr-51 No Data No Data 4.17E-08 2.31E-08 6.57E-09 4.59E-06 2.93E-07 Mn-54 No Data 1.16E-05 2.57E-06 No Data 2.71E-06 4.26E-04 6.19E-06 i Mn-56 No Data 4.48E-10 8.43E-11 No Data 4.52E-10 3.55E-06 3.33E-05 Fe-55 1.28E-05 6.80E-06 2.10E-06 No Data No Data 3.00E-05 7.75E-07 Fe-59 5.59E-06 9.04E-06 4.51E-06 No Cata i No Data 3.43E-04 1.91E-05 Co-58 No Data 4.79E-07 8.55E-07 No Data No Data 2.99E-04 9.29E-06 Co-60 No Data 3.55E-06 6.12E-06 No Data No Data 1.91E-03 2.60E-05 Ni-63 2.22E-04 1.25E-05 7.56E-06 No Data No Data 7.43E-05 1.71E-06 Ni-65 8.08E-10 7.99E-11 4.44E-11 No Data No Data 2.21E-06 2.27E-05 Cu-64 No Data 5.39E-10 2.90E-10 No Data 1.63E-09 2.59E-06 9.92E-06 Zn-65 1.15E-05 3.06E-05 1.90E-05 No Data 1.93E-05 2.69E-04 4.41E-06 Zn-69 1.81E-11 2.61E-11 2.41E-12 No Data 1.58E-11 3.84E-07 2.75E-06 Br-83 No Data No Data 1.2BE-07 No Data No Data No Data No Data Br-84 No Data No Data 1.48E-07 No Data No Data No Data No Data Br-85 No Data No Data 6.24E-09 No Data No Data No Data No Data Rb-86 No Data 5.36E-05 3.09E-05 No Data No Data No Data 2.16E-06 RD-88 No Data 1.52E-07 9.90E-09 No Data j No Data No Data 4.66E-09 Rb-89 No Data 9.33E-08 7.83E-08 No Data No Data No Data 5.11E-10 Sr-89 1.62E-04 No Data 4.66E-06 No Data No Data 5.83E-04 4.52E-05 Sr-90 2.73E-02 No Data 1.74E-03 No Data No Data 3.99E-03 9.28E-05 Sr-91 3.28E-08 No Data 1.24E-09 No Data No Data 1.44E-05 4.70E-05 All values are in (mrem /pci inhaled). They are obtained from Reference 3 (Table E-9). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb- 12 5, 9-23 Rev. 8, 1/94
Hatch 00CM Tacle 9-S (contdi. Inhalation Dese Factors f:r tne tid Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 3.54E-09 No Data 1.42E-10 No Data No Data 6.49E-06 6.55E-05 Y-90 1.11E-06 No Data 2.99E-08 No Data No Data 7.07E-05 7.24E-05 Y-91m 1.37E-10 No Data 4.98E-12 No Data No Data 7.60E-07 4.64E-07 Y-91 2.47E-04 No Data 6.59E-06 No Data No Data 7.10E-04 4.97E-05 Y-92 5.50E-09 No Data 1.57E-10 No Data No Data 6.46E-06 6.46E-05 Y-93 5.04E-08 No Data 1.3EE-09 No Data No Data 2.01E-05 1.05E-04 Zr-95 5.13E-05 1.13E-05 1.00E-05 No Data 1.61E-05 6.03E-04 1.65E-05 Zr-97 5.07E-08 7.34E-09 4.32E-09 No Data . 25E-05 3.06E-05 9.49E-05 Nb-95 6.35E-06 2.48E-06 1.77E-06 No Data 2.33E-06 1.66E-04 1.00E-05 Mo-99 No Data 4.66E-08 1.15E-08 No Data 1. 06E ! 3. 66E-05 3.42E-05 Tc-99m 4.81E-13 9.41E-13 1.56E-11 No Data 1.37E-11 2.57E-07 1.30E-06 Tc-101 2.19E-14 2.30E-14 2.91E-13 No Data 3.92E-13 1.58E-07 4.41E-09 Ru-103 7.55E-07 No Data 2.90E-07 No Data 1.90E-06 1.79E-04 1.21E-05 Ru-105 4.13E-10 No Data 1.50E-10 No Data 3.63E-10 4.30E-06 2.69E-05 Ru-106 3.68E-05 No Data 4.57E-06 No Data 4.97E-05 3.87E-03 1.16E-04 i Rh-105 No Data No Data No Data No Data No Data ' No Data No Data Ag-110m 4.56E-06 3.08E-06 2.47E-06 No Data 5.74E-06 1.48E-03 2.71E-05 Sb-124 No Data No Data No Data No Data No Data No Data No Data i sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 1.82E-06 6.29E-07 2.47E-07 5.20E-0~ N: Data 1.29E-04 9.13E-06 Te-127m 6.72E-06 2.31E-06 8.16E-07 1.64E-06 1.72E-Oil4.00E-04 1.93E-05 Te-127 7.49E-10 2.57E-10 1.65E-10 5.30E-10 1.91E-09 2.71E-06 1.52E-05 Te-129m 5.19E-06 1.85E-06 8.22E-07 1.71E-06 1.36E-05 4.76E-04 4.91E-05 Te-129 2.64E-11 9.45E-12 6.44E-12 1.93E-11 6.94E-11 7.93E-07 6.89E-06 l Te-131m 3.63E-08 1.60E-08 1.37E-08 2.64E-0B 1. 0 8E-: ~ li s . 5 6E-08.32E-05 5 l 1 Te-131 5.87E-12 2.28E-12 1.78E-12 4.59E-12 1.59E-11 5.55E-07 3.60E-07 i l I l 9-24 Rev. 8, 1/94
Hatch ODCM Tacle 9-8 (0:ntd). Inhalation Dose Fa:tcrs for tne Chi.d Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI l l Te-132 1.30E-07 7.36E-08 7.12E-08 8.58E-08 4.79E-O' l.02E-04 3.72E-05 l I-130 2.21E-06 4.43E-06 2.28E-06 4.99E-04 6.61E-06 No Data 1.38E-06 I-131 1.30E-05 1.30E-05 7.37E-06 4.39E-03 2.13E-05 No Data 7.68E-07 I-132 5.72E-07 1.10E-06 5.07E-07 5.23E-05 1.69E-06 No Data 8.65E-07 I-133 4.48E-06 5.49E-06 2.08E-06 1.04E-03 9.13E-06 No Data 1.48E-06 I-134 3.17E-07 5.84E-07 2.69E-07 1.37E-05 8.92E-07 No Data 2.58E-07 I-135 1.33E-06 2.36E-06 1.12E-06 2.14E-04 3.62E-06 No Data 1.20E-06 Cs-134 1.76E-04 2.74E-C4 6.07E-05 No Data 8.93E-05 3.27E-05 1.04E-06 Cs-136 1.76E-05 4.62E-05 3.14E-05 No Data 2.58E-05 3.93E-06 1.13E-06 Cs-137 2.45E-04 2.23E-04 3.47E-05 No Data 7.63E-05 2.81E-05 9.78E-07 Cs-138 1.71E-07 2.27E-07 1.50E-07 No Data 1.68E-07 1.84E-08 7.29E-08 Ba-139 4.98E-10 2.66E-13 1.45E-11 No Data 2.33E-13 1.56E-06 1.56E-05 Ba-140 2.00E-05 1.75E-08 1.17E-06 No Data 5.71E-09 4.71E-04 2.75E-05 Ba-141 5.29E-11 2.95E-14 1.72E-12 No Data 2.56E-14 7.89E-07 7.44E-08 Ba-142 1.35E-11 9.73E-15 7.54E-13 No Data 7.87E-15 4.44E-07 7.41E-10 La-140 1. 4E-07 6.08E-08 2.04E-08 No Data No Data 4.94E-05 6.10E-05 ! La-142 3.50E-10 1.11E-10 3.49E-11 No Data No Data 2.35E-06 2.05E-05 Ce-141 1.06E-05 5.28E-06 7.83E-07 No Data 2.31E-06 1.47E-04 1.53E-05 Ce-143 9.89E-08 5.37E-08 7.77E-09 No Data 2.26E-08 3.12E-05 3.44E-05 l 1 Ce-144 1.83E-03 5.72E-04 9.7?E-05 No Data 3.17E-04 3.23E-03 1.05E-04 ) l Pr-143 4.99E-06 1.50E-06 2.47E-07 No Data 8.11E-07 1.17E-04 2.63E-05 Pr-144 1.61E-11 4.99E-12 8.10E-13 No Data 2.64E-12 4.23E-07 5.32E-08 Nd-147 2.92E-06 2.36E-06 1.84E-07 No Data 1.30E-06 8.87E-05 2.22E-05 l l W-187 4.41E-09 2.61E-09 1.17E-09 No Data No Data 1.11E-05 2.46E-05 Np-239 1.26E-07 9.04E-09 6.35E-09 No Data 2.63E-08 1.57E-05 1.73E-05 i l l l l l
- 9-25 Rev. 8, 1/94
l l l i Hatch ODCM Table 9-9. Inhalation D:ce Fart rs fer tne Teenager Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.59E-07 1.59E-07 1.59E-07 1.59E-07 1.59E-07 1.59E-07 l C-14 3.25E-06 6.09E-07 6.09E-07 6.09E-07 6.09E-07 6.09E-07 6.09E-07 Ma-24 1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 P-32 2.36E-04 1.37E-08 8.95E-06 No Data No Data No Data 1.16E-05 Cr-51 No Data No Data 1.69E-08 9.37E-09 3.84E-09 2.62E-06 3.75E-07 Mn-54 No Data 6.39E-06 1.05E-06 No Data 1.59E-06 2.48E-04 8.35E-06 Mn-56 No Data 2.12E-10 3.15E-11 No Data 2.24E-10 1.90E-06 7.18E-06 Fe-55 4.18E-06 2.98E-06 6.93E-07 No Data No Data 1.55E-05 7.99E-07 Fe-59 1.99E-06 4.62E-06 1.79E-06 No Data No Data 1.91E-04 2.23E-05 Co-58 No Data 2.59E-07 3.47E-07 No Data No Data 1.68E-04 1.19E-05 Co-60 No Data 1.89E-06 2.48E-06 No Data No Data 1.09E-03 3.24E-05 Ni-63 7.25E-05 5.43E-06 2.47E-06 No Data No Data 3.84E-05 1.77E-06 Ni-65 2.73E-10 3.66E-11 1.59E-11 No Data No Data 1.17E-06 4.59E-06 Cu-64 No Data 2.54E-10 1.06E-10 No Data 8.01E-10 1.39E-06 7.68E-06 Zn-65 4.82E-06 1.67E-05 7.80E-06 No Data 1.08E-C5 1.55E-04 5.83E-06 Zn-69 6.04E-12 1.15E-11 8.07E-13 No Data 7.53E-12 1.98E-07 3.56E-08 Br-83 No Data No Data 4.30E-08 No Data No Data No Data No Data Br-84 No Data No Data 5.41E-08 No Data No Data No Data No Data I l Br-85 No Data No Data 2.29E-09 No Data No Data No Data No Data ] Rb-86 No Data 2.38E-05 1.05E-05 No Data No Data No Data 2.21E-06 Rb-88 No Data 6.82E-08 2.40E-08 No Data N Data No Data 3.65E-15 i l Rb-89 No Data 4.40E-08 2.91E-08 No Data No Data No Data 4.22E-17 Sr-89 5.43E-05 No Data 1.56E-06 No Data No Data 3.02E-04 4.64E-05 i Sr-90 1.35E-02 No Data 8.35E-04 No Data No Data 2.06E-03 9.56E-05 Sr-91 1.10E-08 No Data 4.39E-10 No Data No Data 7.59E-06 3.24E-05 All values are in (mrem /pC1 inhaled). They are obtained from Reference 3 (Table E-8). Neither Reference 2 nor Reference 3 contains data for Rh-105, S b- 12 4 , or Sb-125. 9-26 Rev. 8, 1/94
Haten ODCM Tacle 9-9 (contd). Inhalatten Dose Factore fcr tne 7eenager Age Cr .p b-""**""*" Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 1.19E-09 No Data 5.08E-11 No Data No Data 3.43E-06 1.49E-05 Y-90 3.73E-07 No Data 1.00E-08 No Data No Data 3.66E-05 6.99E-05 Y-91m 4.63E-11 No Data 1.77E-12 No Data No Data 4.00E-07 3.77E-09 Y-91 8.26E-05 No Data 2.21E-06 No Data No Data 3.67E-04 5.11E-05 Y-92 1.84E-09 No Data 5.36E-11 No Data No Data 3.35E-06 2.06E-05 Y-93 1.69E-08 No Data 4.65E-10 No Data No Data 1.04E-05 7.24E-05 Zr-95 1.82E-05 5.731:-06 3.94E-06 No Data 6 42E-06 3.36E-04 1.86E-05 Zr-97 1.72E-08 3.40E-09 1.57E-09 No Data 5.15E-09 1.62E-05 7.88E-05 Nb-95 2.32E-06 1.29E-06 7.08E-07 No Data 1.25E-06 9.39E-05 1.21E-05 Mo-99 Ns data 2.11E-08 4.03E-09 No Dsta 5.14E-08 1.92E-05 3.36E-05 Tc-99m 1. 'i 3 E- 13 4.83E-13 6.24E-12 No Data 7.20E-12 1.44E-07 7.66E-07 Tc-101 7.40E-15 1.05E-14 1.03E-13 No Data 1.90E-13 8.34E-08 1.09E-16 Ru-103 2.63E-07 No Data 1.17E-07 No Data 9.29E-07 9.79E-05 1.36E-05 Ru-105 1.40t-10 No Data 5.42E-11 No Data 1.76E-10 2.27E-06 1.13E-05 Ru-106 1.23E-05 No Data 1.55E-06 No Data 2.38E-05 2.01E-03 1.20E-04 j Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 1.73E-06 1.64E-06 9.99E-07 No Data 3.13E-06 8.44E-04 3.41E-05 l Sb-124 No Data N .) Data No Data No Data No Data No Data No Data Sb-125 No Data No Data No Data N3 Data No Data No Data No Data Te-125m 6.10E-07 2.80E-07 8.34E-08 1.75E-07 No Data 6.70E-05 9.38E-06 Te-127m 2.25E-06 1.02E-06 2.73E-07 5.48E-07 8.17E-06 2.07E-04 1.99E-05 Te-127 2.51E-10 1.14E-10 5.52E-11 1.77E-10 9.10E-10 1.40E-06 1.01E-05 Te-129m 1.74E-06 8.23E-07 2.61E-07 5.72E-07 6.49E-06 2.47E-04 5.06E-05 Te-129 8.87E-12 4.22E-12 2.20E-12 6.48E-12 3.32E-11 4.12E-07 2.02E-07 Te-131m 1.23E-08 7.51E-09 5.03E-09 9.06E-09 5.49E-08 2.97E-05 7.76E-05 Te-131 1.97E-12 1.04E-12 6.30E-13 1.55E-12 7.72E-12 2.92E-07 1.89E-09 9-27 Rev. 8, 1/94
i i Haten 00CM Table 9-9 (contd). Inhalation Dose Factors f or tne Teenager Age Group , t l I Nuclide Bone Liver T. Body Thyroid Kidney Lung CI-LLI I Te-132 4.50E-08 3 . 6 3 "l-0 8 2.74E-08 3.07E-08 2.44E-07 5.61E-05 5.79E-05 I-130 7.80E-07 2.24E-06 8.96E-07 1.86E-04 3.44E-06 No Data 1.14E-06 I-131 4.43E-06 6.14E-06 3.30E-06 1.83E-03 1.05E-05 No Data 8.11E-07 l I-132 1.99E-07 5.47E-07 1.97E-07 1.89E-05 8.65E-07 No Data 1.59E-07 I-133 1.52E-06 2.56E-06 7.78E-07 3.65E-04 4.49E-06 No Data 1.29E-06 I I-134 1.11E-07 2.90E-07 1.05E-07 4.94E-06 4.58E-07 No Data 2.55E-09 I-135 4.62E-07 1.18E-06 4.36E-07 7.76E-05 1.86E-06 No Data 8.69E-07 Cs-134 6.28E-05 1.41E-04 6.86E-05 No Data 4.69E-05 1.83E-05 1.22E-06 Cs-136 6.44E-06 2.42E-05 1.71E-05 No Data 1.38E-05 2.22E-06 1.36E-06 Cs-137 8.38E-05 1.06E-04 3.89E-05 No Data 3.80E-05 1.51E-05 1.06E-06 Cs-138 5.82E-08 1.07E-07 5.5BE-08 No Data 8.2BE-08 9.84E-09 3.38E-11 1 Ba-139 1.67E-10 1.18E-13 4.87E-12 No Data 1.11E-13 8.08E-07 8.06E-07 Ba-140 6.84E-06 8.38E-09 4.40E-07 No Data 2.85E-09 2.54E-04 2.86E-05 l Ba-141 1.78E-11 1.32E-14 5.93E-13 No Data 1.23E-14 4.11E-07 9.33E-14 l Ba-142 4.62E-12 4.63E-15 2.84E-13 No Data 3.92E-15 2.39E-07 5.99E-20 La-140 5.99E-08 2.95E-08 7.82E-09 No Data No Data 2.68E-05 6.09E-05 La-142 1.20E-10 5.31E-11 1.32E-11 No Data No Data 1.27E-06 1.50E-06 Ce-141 3.55E-06 2.37E-06 2.71E-07 No Data 1.11E-06 7.67E-05 1.58E-05 Ce-143 3.32E-08 2.42E-08 2.70E-09 No Data 1.08E-08 1.63E-05 3.19E-05 Ce-144 6.11E-04 2.53E-04 3.28E-05 No Data 1.51E-04 1.67E-03 1. 0BE- 04 l l Pr-143 1.67E-06 6.64E-07 8.2BE-08 No Data 3.86E-07 6.04E-05 2.67E-05 Pr-144 5.37E-12 2.20E-12 2.72E-13 No Data 1.26E-12 2.19E-07 2.94E-14 Nd-147 9.83E-07 1.07E-06 6.41E-08 No Data 6.28E-07 4.65E-05 2.28E-05 W-187 1.50E-09 1.22E-09 4.29E-10 No Data No Data 5.92E-06 2.21E-05 Np-239 4.23E-08 3.99E-09 2.21E-09 No Data 1.25E-08 8.11E-06 1.65E-05 9-28 Rev. 8, 1/94
l Hatch ODC'i 7a:1e 9-10. Inhalation Dose Factors f:r One Adait Age Gr:up l Bone Liver T. Body Thyroid Kidney Lung GI-LLI Nuclide l I H-3 No Data 1.58E-07 1.5BE-07 1.58E-07 1.58E-07 1.58E-07 1.53E-07 i C-14 2.27E-06 4.26E-07 4.26E-07 4.26E-07 4.26E-07 4.26E-07 4.26E-07 Na-24 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 P-32 1.65E-04 9.64E-06 6.26E-06 No Data No Data No Data 1.08E-05 l Cr-51 No Data No Data 1.25E-08 7.44E-09 2.85E-09 1.80E-06 4.15E-07 Mn-54 No Data 4.95E-06 7.87E-07 No Data 1.23E-06 1.75E-04 9.67E-06 Mn-56 No Data 1.55E-10 2.29E-11 No Data 1.63E-10 1.18E-06 2.53E-06 Fe-55 3.07E-06 2.12E-06 4.93E-07 No Data No Data 9.01E-06 7.54E-07 2.35E-05 Fe-59 1.47E-06 3.47E-06 1.32E-06 No Data No Data 1.27E-04 Co-58 No Data 1.98E-07 2.59E-07 No Data No Data 1.16E-04 1.33E-05 Co-60 No Data 1.44E-06 1.85E-06 No Data No Data 7.46E-04 3.56E-05 Ni-63 5.40E-05 3.93E-06 1.81E-06 No Data No Data 2.23E-05 1.67E-06 Ni-65 1.92E-10 2.62E-11 1.14E-11 No Data No Data 7.00E-07 1.54E-06 Cu-64 No Data 1.83E-10 7.69E-11 No Data 5.78E-10 8.48E-07 6.12E-06 Zn-65 4.05E-06 1.29E-05 5.82E-06 No Data 8.62E-06 1.08E-04 6.68E-06 Zn-69 4.23E-12 8.14B-12 5.65E-13 No Data 5.27E-12 1.15E-07 2.04E-09 Br-83 No Data No Data 3.01E-08 No Data No Data No Data 2.90E-08 Br-84 No Data No Data 3.91E-08 No Data No Data No Data 2.05E-13 Br-85 No Data No Data 1.60E-09 No Data No Data No Data No Data Rb-86 No Data 1.69E-05 7.37E-06 No Data No Data No Data 2.00E-06 Rb-88 No Data 4.84E-08 2.41E-08 No Data No Data No Data 4.18E-19 Rb-89 No Data 3.20E-08 2.12E-08 No Data No Data No Data 1.16E-21 l Sr-89 3.80E-05 No Data 1.09E-06 No Data No Data 1.75E-04 4.37E-05 ! Sr-90 1.24E-02 No Data 7.62E-04 No Data No Data 1.20E-03 9.02E-05 Sr-91 7.74E-09 No Data 3.13E-10 No Data No Data 4.56E-06 2.3)E-05 All values are in (mrem /pci inhaled). They are obtained from Ref erence 3 (Table E-7), except as f ollows: Reference 2 (Table C-1) f or Rh- 10 5, Sb-124, and Sb-125, 9-29 Rev. 8, 1/94 l ,_ _ _ _ . _ , __ _ , _
Hatch 00CM Tacle 9-10 ( cor.t d ) . Inhalat;:n Dose Facters fer the Ac;1t Age Group l I Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 8.43E-10 No Data 3.64E-11 No Data No Data 2.06E-06 5.38E-06 Y-90 2.61E-07 No Data 7.01E-09 No Data No Data 2.12E-05 6.32E-05 f Y-91m 3.26E-11 No Data 1.27E-12 No Data No Data 2.40E-07 1.66E-10 Y-91 5.78E-05 No Data 1.55E-06 No Data No Data 2.13E-04 4.81E-05 Y-92 1.29E-09 No Data 3.77E-11 No Data No Data 1.96E-06 9.19E-06 Y-93 1.18E-08 No Data 3.26E-10 No Data No Data 6.06E-06 5.27E-05 ! l Zr-95 1.34E-05 4.30E-06 2.91E-06 No Data 6.77E-06 2.21E-04 1.88E-05 i Zr-97 1.21E-08 2.45E-09 1.13E-09 No Data 3.71E-09 9.84E-06 6.54E-05 l 1 Nb-95 1.76E-06 9.77E-07 5.26E-07 No Data 9.67E-07 6.31E-05 1.30E-05 )
\
l 3.64E-08 1.14E-05 3.10E-05 I l Mo-99 No Data 1.51E-08 2.87E-09 No Data Tc-99m 1.29E-13 3.64E-13 4.63E-12 No Data 5.52E-12 9.55E-08 5.20E-07 Tc-101 5.22E-15 7.52E-15 7.38E-14 No Data 1.35E-13 4.99E-08 1.36E-21 Ru-103 1.91E-07 No Data 8.23E-08 No Data 7.29E-07 6.31E-05 1.38E-05 Ru-105 9.88E-il No Data 3.89E-11 No Data 1.27E-10 1.37E-06 6.02E-06 ( Ru-106 8.64E-06 No Data 1.09E-06 No Data 1.67E-05 ?.17E-03 1.14E-04 Rh-105 9.24E-10 6.73E-10 4.43E-10 No Data 2.86E-09 2.41E-06 1.09E-05 Ag-110m 1.35E-06 1.25E-06 7.43E-07 No Data 2.46E-06 5.79E-04 3.78E-05 Sb-124 3.90E-06 7.36E-08 1.55E-06 9.44E-09 No Data 3.10E-04 5.08E-05 Sb-125 8.26E-06 8.91E-08 1.66E-06 7.34E-09 No Data 2.75E-04 1.26E-05 j Te-125m 4.27E-07 1. 9 8E -0 7 5.84E-08 1.31E-07 1.55E-06 3.92E-05 8.83E-06 Te-127m 1.58E-06 7.21E-07 1.96E-07 4.1'E-07
. 5.72E-06 1.20E-04 1.87E-05 Te-127 1.75E-10 8.03E-11 3.87E-11 1.32E-10 6.37E-1C 8.14E-07 7.17E-06 Te-129m 1.22E-06 5.84E-07 1.98E-07 4.30E-07 4.57E-06 1.45E-04 4.79E-05 Te-129 6.22E-12 2.99E-12 1.55E-12 4.87E-12 2.34E-11 2.42E-07 1.96E-08 Te-131m 8.74E-09 5.45E-09 3.63E-09 6.88E-09 3.86E-08 1.82E-05 6.95E-05 Te-131 -12 7.44E-13 4.49E-13 1.17E-12 5.46E-12 1.74E-07 2.30E-09 9-30 Rev. 8, 1/94
__, , __. - _ __, _ _ _. . ~ . , _
Hatch ODCM I Table 9-10 (contd). Inhalation Dose factors for the Ada.t Age Group Bone Liver T. Body Thyroid Kidney Lung GI-LLI Nuclide Te-132 3.25E-08 2.69E-08 2.02E-08 2.37E-08 1.82E-07 3.60E-05 6.J7E-05 5.72E-07 1.68E-06 6.60E-07 1.42E-04 2.61E-06 No Data 9.61E-07 I-130 3.15E-06 4.47E-06 2.56E-06 1.49E-03 7.66E-06 No Data 7.85E-07 I-131 1.45E-07 4.07E-07 1.45E-07 1.43E-05 6.48E-07 No Data 5.08E-08 I-132 1.08E-06 1.85E-06 5.65E-07 2.69E-04 3.23E-06 No Data 't.11E-06 I-133
! 8.05E-08 2.16E-07 7.69E-08 3.73E-06 3.44E-07 No Data 1.26E-10 I-134 l
3.35E-07 8.73E-07 3.21E-07 5.60E-05 1.39E-06 No Data 6.56E-07 I-135 Cs-134 4.66E-05 1.06E-04 9.10E-05 No Data 3.59E-05 1.22E-05 1.30E-06 Cs-136 4.88E-06 1.83E-05 1.38E-05 No Data 1.07E-05 1.50E-06 1.46E-06 Ca-137 5.98E-05 7.76E-05 5.35E-05 No Data 2.78E-05 9.40E-06 1.05E-06 Cs-138 4.14E-08 7.76E-08 4.05E-08 No Data 6.00E-08 6.07E-09 2.33E-13 Ba-139 1.17E-10 8.32E-14 3.42E-12 No Data 7.78E-14 4.70E-07 1.12E-07 Ba-140 4.88E-06 6.13E-09 3.21E-07 No Data 2.09E-09 1.59E-04 2.73E-05 Ba-141 1.25E-11 9.41E-15 4.20E-13 No Data 8.75E-15 2.42E-07 1.45E-17 Ba-142 3.29E-12 3.38E-15 2.07E-13 No Data 2.86E-15 1.49E-07 1.96E-26 La-140 4.30E-08 2.17E-08 5.73E-09 Nc Data No Data 1.70E-05 5.73E-05 La-142 8.54E-11 3.88E-11 9.65E-12 No Data No Data 7.91E-07 2.64E-07 Ce-141 2.49E-06 1.69E-06 1.91E-07 No Data 7.83E-07 4.52E-05 1.50E-05 Ce-143 2.33E-08 1.72E-08 1.91E-09 No Data 7.60E-09 9.97E-06 2.83E-05 Ce-144 4.29E-04 1.79E-04 2.30E-05 No Data 1.06E-04 9.72E-04 1.02E-04 i Pr-143 1.17E-06 4.69E-07 5.80E-08 No Data 2.70E-07 3.51E-05 2.50E-05 Pr-144 3.76E-12 1.56E-12 1.91E-13 No Data 8.81E-13 1.27E-07 2.69E-18 Nd 147 6.59E-07 7.62E-07 4.56E-08 No Data 4.45E-07 2.76E-05 2.16E-05 i l W-187 1.06E-09 8.85E-10 3.10E-10 No Data No Data 3.63E-06 1.94E-05 Np-239 2.87E-08 2.82E-09 1.55E-09 No Data 8.75E-09 4.70E-06 1.49E-05 9-31 Rev. 8, 1/94 e-- - ,-
-+-se-
l Hatch ODCM Table 9-11. Ingestion Dose Factors for ne :nfant Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 3.08E-07 3.08E-07 3.08E-07 3.08E-07 3.08E-07 3.08E-07 C-14 2.37E-05 5.06E-06 5.06E-06 5.06E-06 5.06E-06 5.06E-06 5.06E-06 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 I Na-24 1.01E-05 l' P-32 1.70E-03 1.00E-04 6.59E-05 No Data No Data No Data 2.30E-05 Cr-51 No Data No Data 1.41E-08 9.20E-09 2.01E-09 1.79E-08 4.11E-07 Mn-54 No Data 1.99E-05 4.51E-06 No Data 4.41E-06 No Data 7.31E-06 l l Mn-56 No Data 8.18E-07 1.41E-07 No Data 7.03E-07 No Data 7.43E-05 l Fe-55 1.39E-05 8.98E-06 2.40E-06 No Data No Data 4.39E-06 1.14E-06 i i Fe-59 3.0SE-05 5.38E-05 2.12E-05 No Data No Data 1.59E-05 2.57E-05 Co-58 No Data 3.60E-06 8.98E-06 No Data No Data No Data 8.97E-06 Co-60 No Data 1.08E-05 2.55E-05 No Data NC Data No Data 2.57E-05 Ni-63 6.34E-04 3.92E-05 2.20E-05 No Data No Data No Data 1.95E-06 Ni-65 4.70E-06 5.32E-07 2.42E-07 No Data No Data No Data 4.05E-05 Cu-64 No Data 6.09E-07 2.82E-07 No Data 1.03E-06 No Data 1.25E-05 1.84E-05 6.31E-05 2.91E-05 Zn-65 No Data 3.06E-05 No Data 5.33E-05 l Zn-69 9.33E-08 1.68E-07 1.25E-08 No Data 6.98E-08 No Data 1.37E-05 l Br-83 No Data No Data 3.63E-07 No Data No Data No Data No Data Br-84 No Data No Data 3.82E-07 No Data No Data No Data No Data Br-85 No Data No Data 1.94E-08 No Data No Data No Data No Data Rb-86 No Data 1.70E-04 8.40E-05 No Data No Data No Data 4.35E-06 Rb-88 No Data 4.98E-07 2.73E-07 No Data No Data No Data 4.85E-07 Rb-89 No Data 2.86E-07 1.97E-07 No Data No Data No Data 9.74E-08 Sr-89 2.51E-03 No Data 7.20E-05 No Data No Data No Data 5.16E-05 Sr-90 1.85E-02 No Data 4.71E-03 No Data NO Data No Data 2.31E-04 Sr-91 5.00E-05 No Data 1.81E-06 No Data No Data No Data 5.92E-05 All values are in (mrem /pci ingested). They are obtained from Reference 3 (Table E-14). Neither Reference 2 nor Reference 3 contains data for Rh-105, sb-124, or Sb-12 5. 9-32 Rev. 8, 1/94
Hatch 00CM Taole 9-11 (contd). Ingestion Oose Factors f:r tne :nfant Age Group Nuclide Sone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 1.92E-05 No Data 7.13E-07 No Data No Data No Data 2.07E-04 Y-90 8.69E-08 No Data 2.33E-09 No Data No Data No Data 1.20E-04 Y-91m 8.10E-10 No Data 2.76E-11 No Data No Data No Data 2.70E-06 Y-91 1.13E-06 No Data 3.01E-08 No Data No Data No Data 8.10E-05 Y-92 7.65E-09 No Data 2.15E-10 No Data No Data No Data 1.46E-04 Y-93 2.43E-08 No Data 6.62E-10 No Data No Data No Data 1.92E-04 Zr-95 2.06E-07 5.02E-08 3.56E-08 No Data 5.41E-08 No Data 2.50E-05 Zr-97 1.48E-08 2.54E-09 1.16E-09 No Data 2.56E-09 No Data 1.62E-04 Nb-95 4.20E-08 1.73E-08 1.00E-08 No Data 1.24E-08 No Data 1.46E-05 Mo-99 No Data 3.40E-05 6.63E-06 No Data 5.08E-05 No Data 1.12E-05 Tc-99m 1.92E-09 3.96E-09 5.10E-08 No Data 4.26E-08 2.07E-09 1.15E-06 i Tc-101 2.27E-09 2.86E-09 2.83E-08 No Data 3.40E-08 1.56E-09 4.86E-07 i ! Ru-103 1.48E-06 No Data 4.95E-07 No Data 3.08E-06 No Data 1.80E-05 Ru-105 1.36E-07 No Data 4.58E-08 No Data 1.00E-06 No Data 5.41E-05 Ru-106 2.41E-05 No Data 3.01E-06 No Data 2.85E-05 No Data 1.83E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 9.96E-07 7.27E-07 4.81E-07 No Data 1.04E-06 No Data 3.77E-05 Sb-124 No Data No Data No Data No Data No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 2.33E-05 7.79E-06 3.15E-06 7.84E-06 No Data No Data 1.11E-05 Te-127m 5.85E-05 1.94E-05 7.0SE-06 1.69E-05 1.44E-04 No Data 2.36E-05 I I Te-127 00E-06 3.35E-07 2.15E-07 8.14E-07 2.44E-06 No Data 2.10E-05 Te-129m 1.00E-04 3.43E-05 1.54E-05 3.84E-05 2.50E-04 No Data 5.97E-05 Te-129 2.84E-07 9.79E-08 6.63E-OB 2.38E-07 7.07E-07 No Data 2.27E-05 Te-131m 1.52E-05 6.12E-06 5.05E-06 1.24E-05 4.21E-05 No Data 1.03E-04 Te-131 1.76E-07 6.50E-08 4.94E-08 1.57E-07 4.50E-07 No Data 7.11E-06 9-33 Rev. 8, 1/94
I Hatch ODCM l Table 9-11 (contd). Ingest:cn Dcoe Factors for One '.fant Age Group l Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 2.08E-05 1.03E-05 9.61E-06 1.52E-05 6.44E-05 No Data 3.81E-05 I-130 6.00E-06 1.32E-05 5.30E-06 1.48E-03 1.45E-05 No Data 2.83E-06 I-131 3.59E-05 4.23E-05 1.86E-05 1.39E-02 4.94E-05 No Data 1.51E-06 I-132 1.66E-06 3.37E-06 1.20E-06 1.58E-04 3.76E-06 No Data 2.73E-06 i l l I-133 1.25E-05 1.82E-05 5.33E-06 3.31E-03 2.14E-05 No Data 3.08E-06 l l I-134 8.69E-07 1.78E-06 6.33E-07 4.15E-05 1.99E-06 No Data 1.84E-06 ! I-135 3.64E-06 7.24E-06 2.64E-06 6.49E-04 8.07E-06 No Data 2.62E-06 i Cs-134 3.77E-04 7.03E-04 7.10E-05 No Data 1.81E-C4 7.42E-05 1.91E-06 i Cs-136 4.59E-05 1.35E-04 5.04E-05 No Data 5.38E-05 1.10E-05 2.05E-06 ! Cs-137 5.22E-04 6.11E-04 4.33E-05 No Data 1.64E-04 6.64E-05 1.91E-06 Cs-138 4.81E-07 7.82E-07 3.79E-07 No Data 3.90E-07 6.09E-08 1.25E-06 f Ba-139 8.81E-07 5.84E-10 2.55E-08 No Data 3.51E-10 3.54E-10 5.58E-05 l Ba-140 1.71E-04 1.71E-07 8.81E-06 No Data 4.06E-08 1.05E-07 4.20E-05 Ba-141 4.25E-07 2.91E-10 1.34E-08 No Data 1.75E~'.0 1.77E-10 5.19E-06 Ba-142 1.84E-07 1.53E-10 9.06E-09 No Data 8.81E-11 9.26E-11 7.59E-07 La-140 2.11E-08 8.32E-09 2.14E-09 No Data No Data No Data 9.77E-05 La-142 1.10E-09 4.04E-10 9.67E-11 No Data No Data No Data 6.86E-05 l Ce-141 7.87E-08 4.80E-08 5.65E-09 No Data 1.48E-08 No Data 2.48E-05 Ce-143 1.48E-08 9.82E-06 1.12E-09 No Date 2.86E-09 No Data 5.73E-05 Ce-144 2.98E-06 1.22E-06 1.67E-07 No Data 4.93E-07 No Data 1.71E-04 l Pr-143 8.13E-08 3.04E-08 4.03E-09 No Data 1.13E-08 No Data 4.29E-05 ! Pr-144 2.74E-10 1.06E-10 1.38E-11 No Data 3.84E-11 No Data 4.93E-06 Nd-147 5.53E-08 5.68E-08 3.48E-09 No Data 2.19E-08 No Data 3.60E-05 W-187 9.03E-07 6.28E-07 2.17E-07 No Data No Data No Data 3.69E-05 Np-239 1.11E-08 9.93E-10 5.61E-10 No Data 1.98E-09 No Data 2.87E-05 l l 1 9-34 Rev. 8, 1/94
I Hatch ODCM Table 9-12. Ingestion Dose Factors for the Cntid Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 2.03E-07 2.03E-07 2.03E-O' 2.03E-07 2.03E-07 2.03E-07 C-14 1.21E-05 2.42E-06 2.42E-06 2.42E-06 2.42E-06 2.42E-06 2.42E-06 1 Na-24 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 I P-32 8.25E-04 3.86E-05 3.18E-05 No Data No Data No Data 2.28E-05 I Cr-51 No Data No Data 8.90E-09 4.94E-09 1.35E-09 9.02E-09 4.72E-07 Mn-54 No Data 1.07E-05 2.85E-06 No Data 3.00E-06 No Data 8.98E-06 Mn-56 No Data 3.34E-07 7.54E-08 No Data 4.04E-07 No Data 4.84E-05 Fe-55 1.15E-05 6.10E-06 1.89E-06 No Data No Data 3.45E-06 1.13E-06 Fe-59 1.65E-05 2.67E-05 1.33E-05 No Data No Data 7.74E-06 2.78E-05 Co-58 No Data 1.80E-06 5.51E-06 No Data No Data No Data 1.05E-05 Co-60 No Data 5.29E-06 1.56E-05 No Data No Data No Data 2.93E-05 Ni-63 5.38E-04 2.88E-05 1.83E-05 No Data No Data No Data 1.94E-06 Ni-65 2.22E-06 2.09E-07 1.22E-07 No Data No Data No Data 2.56E-05 i Cu-64 No Data 2.45E-07 1.48E-07 No Data 5.92E-07 No Data 1.15E-05 i Zn-65 1.37E-05 3.65E-05 2.27E-05 No Data 2.3CE-05 No Data 6.41E-06 Zn-69 4.38E-08 6.33E-08 5.85E-09 No Data 3.84E-08 No Data 3.99E-06 Br-83 No Data No Data 1.71E-07 No Data No Data No Data No Data Br-84 No Data No Data 1.98E-07 No Data No Data No Data No Data Br-85 No Data No Data 9.12E-09 No Data No Data No Data No Data l Rb-86 No Data 6.70E-05 4.12E-05 No Data No Data No Data 4.31E-06 Rb-88 No Data 1.90E-07 1.32E-07 No Data No Data No Data 9.32E-09 l Rb-89 No Data 1.17E-07 1.04E-07 No Data No Data No Data 1.02E-09 j Sr-89 1.32E-03 No Data 3.77E-05 No Data No Data No Data 5.11E-05 Sr-90 1.70E-02 No Data 4.31E-03 No Data No Data No Data 2.29E-04 Sr-91 2.40E-05 No Data 9.06E-07 No Data No Data No Data 5.30E-05 All values are in (mrem /pci ingested). They are obtained from Reference 3 (Table E-13). Neither Reference 2 nor Reference 3 i contains data for Rh-105, Sb-124, or Sb-125. , l l l l 9-35 Rev. 8, 1/94 l l i
Hatch 00CM Tat;e 9-12 (contd). Ingestion Dose Factore for tne Dhild Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LL1 Sr-92 9.03E-06 No Data 3.62E-07 No Data No Data No Data 1.71E-04 Y-90 4.11E-08 No Data 1.10E-09 No Data No Data No Data 1.17E-04 Y-91m 3.82E-10 No Data 1.39E-11 No Data N2 Data No Data 7.48E-07 Y-91 6.02E-07 No Data 1.61E-08 No Data No Data No Data 8.02E-05 Y-92 3.60E-09 No Data 1.03E-10 No Data No Data No Data 1.04E-04 Y-93 1.14E-08 No Data 3.13E-10 No Data No Data No Data 1.70E-04 Ir-95 1.16E-07 2.55E-08 2.27E-08 No Data 3.65E-08 No Data 2.66E-05 Zr-97 6.99E-09 1.01E-09 5.96E-10 No Data 1.45E-09 No Data 1.53E-04 Nb-95 2.25E-08 8.76E-09 6.26E-09 No Data 8.23E-09 No Data 1.62E-05 Mo-99 No Data 1.33E-05 3.29E-06 No Data 2.84E-05 No Data 1.10E-05 l Tc-99m 9.23E-10 1.81E-09 3.00E-08 No Data 2.63E-08 9.19E-10 1.03E-06 Tc-101 1.07E-09 1.12E-09 1.42E-08 No Data 1.91E-08 5.92E-10 3.56E-09 Ru-103 7.31E-07 No Data 2.81E-07 No Data 1.84E-06 No Data 1.89E-05 Ru-105 6.45E-08 No Data 2.34E-08 No Data 5.67E-07 No Data 4.21E-05 Ru-106 1.17E-C5 No Data 1.46E-06 No Data 1.58E-05 No Data 1.72E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 5.39E-07 3.64E-07 2.91E-07 No Data 6.78E-07 No Data 4.33E-05 l Sb-124 No Data No Data No Data No Data No Data No Data No Data sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 1.14E-05 3.09E-06 1.52E-06 3.20E-06 No Data No Data 1.10E-05 Te-127m 2.89E-05 7.78E-06 3.43E-06 6.91E-06 8.24E-05 No Data 2.34E-05 Te-127 4.71E-07 1.27E-07 1.01E-07 3.26E-07 1.34E-06 No Data 1.84E-05 l Te-129m 4.87E-05 1.36E-05 7.56E-06 1.57E-05 1.43E-04 No Data 5.94E-05 Te-129 1.34E-07 3.74E-08 3.18E-08 9.56E-08 3.92E-07 No Data 8.34E-06 Te-131m 7.20E-06 2.49E-06 2.65E-06 5.12E-06 2.41E-05 No Data 1.01E-04 Te-131 8.30E-08 2.53E-08 2.47E-08 6.35E-08 2.51E-07 No Data 4.36E-07 l l l 9-36 Rev. 8, 1/94
l l [ Hatch ODCM l Table 9-12 (contd). Ingestion Dose Fact:rs for tne Cn.1d Age Group Bone Liver T. Body Thyroid Kidney Lung GI-LLI Nuclide Te-132 1.01E-05 4.47E-06 5.40E-06 6.51E-06 4.15E-05 No Data 4.50E-05 I-130 2.92E-06 5.90E-06 3.04E-06 6.50E-04 8.82E-06 No Data 2.76E-06 I-131 1.72E-05 1.73E-05 9.83E-06 5.72E-03 2.84E-05 No Data 1.54E-06 I-132 8.00E-07 1.47E-06 6.76E-07 6.82E-05 2.25E-06 No Data 1.73E-06 I-133 5.92E-06 7.32E-06 2.77E-06 1.36E-03 1.22E-05 No Data 2.95E-06 1-134 4.19E-07 7.78E-07 3.58E-07 1.79E-05 1.19E-06 No Data 5.16E-07 I-135 1.75E-06 3.15E-06 1.49E-06 2.79E-04 4.83E-06 No Data 2.40E-06 Cs-134 2.34E-04 3.84E-04 8.10E-05 No Data 1.19E-04 4.27E-05 2.07E-06 Cs-136 2.35E-05 6.46E-05 4.18E-05 No Data 3.44E-05 5.13E-06 2.27E-06 Cs-137 3.27E-04 3.13E-04 4.62E-05 No Data 1.02E-04 3.67E-05 1.96E-06 Cs-138 2.28E-07 3.17E-07 2.01E-07 No Data 2.23E-07 2.40E-08 1.46E-07 Ba-139 4.14E-07 2.21E-10 1.20E-08 No Data 1.93E-10 1.30E-10 2.39E-05 Ba-140 8.31E-05 7.29E-08 4.85E-06 No Data 2.37E-08 4.34E-08 4.21E-05 Ba-141 2.00E-07 1.12E-10 6.51E-09 No Data 9.69E-11 6.58E-10 1.14E-07 Ba-142 8.74E-08 6.29E-11 4.88E-09 No Data 5.09E-11 3.70E-11 1.14E-09 La-140 1.01E-08 3.53E-09 1.19E-09 No Data No Data No Data 9.84E-05 La-142 5.24E-10 1.67E-10 5.23E-11 No Data No Data No Data 3.31E-05 I ce-141 3.97E-08 1.98E-08 2.94E-09 No Data 8.68E-09 No Data 2.47E-05 Ce-143 6.99E-09 3.79E-06 5.49E-10 No Data 1.59E-09 No Data 5.55E-05 Ce-144 2.08E-06 6.52E-07 1.11E-07 No Data 3 . 612 - No Data 1.70E-04 Pr-143 3.93E-08 1.18E-08 1.95E-09 No Data 6.39E-09 No Data 4.24E-05 Pr-144 1.29E-10 3.99E-11 6.49E-12 No Data 2.11E-11 No Data 8.59E-08 i Nd-147 2.79E-0E l ; 26E-08 1.75E-09 No Data 1.24E-08 No Data 3.58E-05 W-187 4.29E-07 2.54E-07 1.14E-07 No Data No Data No Data 3.57E-05 l Np-239 5.2bE-09 3.77E-10 2.65E-10 No Data 1.09E-09 No Data 2.79E-05 t 9-37 Rev. 8, 1/94 l l
i Hatch ODCM l l *acle 9-13. Ingestion Dose Ea:t:rs for the Teenager Age Grr.;; ! 1 Nuclide Bone Liver T. Body Thyroid Kidney 1.u ng GI-LLI l H-3 No Data 1.06E-07 1.06E-07 1.06E-07 1.06E-07 1.06E-07 1.06E-07 1 1 1 C-14 4.06E-06 8.12E-07 8.12E-07 8.12E-07 8.12E-07 8.12E-07 8.12E-07 ; i i Na-24 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 l l P-32 2.76E-04 1.71E-05 1.07E-05 No Data No Data No Data 2.32E-05 Cr-51 No Data No Data 3.60E-09 2.00E-09 7.89E-10 5.14E-09 6.05E-07 l Mn-54 No Data 5.90E-06 1.17E-06 No Data 1.76E-06 N: Data 1.21E-05 l l
- Mn-56 No Data 1.58E-07 2.81E-08 No Data 2.00E-07 Nc Data 1.04E-05 l
Fe-55 3.78E-06 2.68E-06 6.25E-07 No Data No Data 1.*0E-06 1.16E-06 Fe-59 5.87E-06 1.37E-C5 5.29E-06 No Data No Data 4.32E-06 3.24E-05 l l l Co-58 No Data 9.72E-07 2.24E-06 No Data No Data No Data 1.34E-05 l Co-60 No Data 2.81E-06 6.33E-06 No Data No Data No Data 3.66E-05 l f Ni-63 1.77E-04 1.25E-05 6.00E-06 No Data No Data NO Data 1.99E-06 l Ni-65 7.49E-07 9.57E-08 4.36E-08 No Data No Data N Data 5.19E-06 l Cu-64 No Data 1.15E-07 5.41E-08 No Data 2.91E-07 N: Data 8.92E-06 ( l Zn-65 5.76E-06 2.00E-05 9.33E-06 No Data 1.28E-05 NO Data 8.47E-06 I Zn-69 1.47E-08 2.80E-08 1.96E-09 No Data 1.83E-08 Nc Data 5.16E-08 Br-83 No Data No Data 5.74E-08 No Data No Data NO Data No Data l Br-84 No Data No Data 7.22E-08 No Data No Data No Data No Data Br-85 No Data No Data 3.05E-09 No Data No Data No Data No Data Rb-86 No Data 2.98E-05 1.40E-05 No Data No Data N Data 4.41E-06 Rb-88 No Data 8.52E-OS 4.54E-C8 No Data No Data *: Data
. 7.30E-15 Rb-89 No Data 5.50E-C8 3.89E-08 No Data No Data NO Data 8.43E-17 Sr-89 4.40E-04 No Data 1.26E-05 No Data No Data No Data 5.24E-05 Sr-90 8.30E-03 Ne Data 2.05E-03 No Data No Data No Data 2.33E-04 Sr-91 8.07E-06 No Data 3.21E-07 No Data No Data No Data 3.66E-05 i
All values are in (mrem /pci ingested) . They are Octatned from Reference 3 (Table E-12). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb-125. 9-38 Rev. 8, 1/94 i
i i Hater ODCM l Tacle 9-13 ( Ontd). Ingestion Oose Facters f or the Tecnager Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 3.05E-06 No Data 1.30E-07 No Data No Data No Data 7.77E-05 Y-90 1.37E-08 No Data 3.69E-10 No Data No Data No Data 1.13E-04 l Y-91m 1.29E-10 No Data 4.93E-12 No Data No Data No Data 6.09E-09 Y-91 2.01E-07 No Data 5.39E-09 No Data No Data No Data 8.24E-05 l Y-92 1.21E-09 No Data 3.50E-11 No Data No Data No Data 3.32E-05 i l Y-93 3.83E-09 No Data 1.05E-10 No Data No Data No Data 1.17E-04 Zr-95 4.12E-08 1.30E-08 8.94E-09 No Data 1.91E-08 No Data 3.00E-05 l l Zr-97 2.37E-09 4.69E-10 2.16E-10 No Data 7.11E-10 No Data 1.27E-04 l Nb-95 8.22E-09 4.56E-09 2.51E-09 No Data 4.42E-09 No Data 1.95E-05 l Mo-99 No Data 6.03E-06 1.15E-06 No Data 1.38E-05 No Data 1.08E-05 Tc-99m 3.32E-10 9.26E-10 1.20E-08 No Data 1.38E-08 5.14E-10 6.08E-07 Te-101 3.60E-10 5.12E-10 5.03E-09 No Data 9.26E-09 3.12E-10 8.75E-17 Ru-103 2.55E-07 No Data 1.09E-07 No Data 8.99E-07 No Data 2.13E-05 Ru-105 2.18E-08 No Data 8.46E-09 No Data 2.75E-07 No Data 1.76E-05 i Ru-106 3.92E-06 No Data 4.94E-07 No Data 7.56E-06 No Data 1.88E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 2.05E-07 1.94E-07 1.18E-07 No Data 3.70E-07 No Data 5.45E-05 Sb-124 No Data No Data No Data No Data No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Cata Te-125m 3.83E-06 1.38E-06 5.12E-07 1.07E-06 No Data No Data 1.13E-05 1 l Te-127m 9.67E-06 3.43E-06 1.15E-06 2. 30E-06 l 3.92E-05 No Data 2.41E-05 Te-127 1.58E-07 5.60E-08 3.40E-08 1.09E-01 6.40E-07 No Data 1.22E-05 Te-129m 1.63E-05 6.05E-06 2.58E-06 5.26E-06 6.82E-05 No Data 6.12E-05 Te-129 4.48E-08 1.67E-08 1.09E-08 3.20E-08 1.88E-07 No Data 2.45E-07 Te-131m 2.44E-06 1.17E-06 9.76E-07 1.76E-06 1.22E-05 No Data 9.39E-05 Te-131 2.79E-08 1.15E-08 8.72E-09 2.15E-08 1.22E-07 No Data 2.29E-09 i l l l l l l l 9-39 Rev. 8, 1/94 l i l l _ ___ _ . _ _. _ - __a
Hatch ODCM Table 9-13 (contd). .ngestion Dose Factors for the Teenager Age Group i Nuclide Bone Liver T.8ody Thyroid Kidney Lung GI-LLI ! Te-132 3.49E-06 2.21E-06 2.08E-06 2.33E-06 2.12E-05 No Data 7.00E-05 l I-130 1.03E-06 2.98E-06 1.19E-06 2.43E-04 4.59E-06 No Data 2.29E-06 1 I-131 5.85E-06 8.19E-06 4.40E-06 2.39E-03 1.41E-05 No Data 1.62E-06 I-132 2.79E-07 7.30F-07 2.62E-07 2.46E-05 1.15E-06 No Data 3.18E-07 l I-133 2.01E-06 3.41E-06 1.04E-06 4.76E-04 5.98E-06 No Data 2.58E-06 j I-134 1.46E-07 3.87E-07 1.39E-07 6.45E-06 6.10E-07 No Data 5.10E-09 ! I-135 6.10E-07 1.57E-06 5.82E-07 1.01E-04 2.48E-06 No Data 1.74E-06 ! Cs-134 8.37E-05 1.97E-04 9.14E-05 No Data 6.26E-05 2.39E-05 2.45E-06 Cs-136 8.59E-06 3.38E-05 2.27E-05 No Data 1.84E-05 2.90E-06 2.72E-06 l Cs-137 1.12E-04 1.49E-04 5.19E-05 No Data 5.07E-05 1.97E-05 2.12E-06 Cs-138 7.76E-08 1.49E-07 7.45E-08 No Data 1.10E-07 1.28E-08 6.76E-11 1 Ba-139 1.39E-07 9.78E-11 4.05E-09 No Data 9.22E-11 6.74E-11 1.24E-06 i Ba-140 2.84E-05 3.48E-08 1.83E-06 No Data 1.18E-08 2.34E-08 4.38E-05 l l Ba-141 6.71E-08 5.01E-11 2.24E-09 No Data 4.65E-11' 3.43E-11 1.43E-13 l Ba-142 2.99E-08 2.99E-11 1.84E-09 No Data 2.53E-11 1.99E-11 9.18E-20 La-140 3.48E-09 1.71E-09 4.55E-10 No Data No Data No Data 9.82E-05 La-142 1.79E-10 7.95E-11 1.98E-11 No Data No Data No Data 2.42E-06 Ce-141 1.33E-08 8.88E-09 1.02E-09 No Data 4.18E-09 No Data 2.54E-05 Ce-143 2.35E-09 1.71E-06 1.91E-10 No Data 7.67E-10 No Data 5.14E-05 j Ce-144 6.96E-07 2.88E-07 3.74E-08 No Data 1.72E-07 No Data . 75E-04 Pr-143 1.31E-08 5.23E-09 6.52E-10 No Data 3.04E-09 No Data 4.31E-05 Pr-144 4.30E-11 1.76E-11 2.18E-12 No Data 1.01E-11 No Data 4.74E-14 Nd-147 9.38E-09 1.02E-08 6.11E-10 No Data 5.99E-09 No Data 3.68E-05 W-187 1.46E-07 1.19E-07 4.17E-08 No Data No Data No Data 3.22E-05 ) l Np-239 1.76E-09 1.66E-10 9.22E-11 No Data 5.21E-10 No Data 2.67E-05 9-40 Rev. 8, 1/94 l 1
l Hatch ODCM Table 9-14. Ingest.or. Dose Factors for the AdMt Age Group i Bone Liver T. Body Thyroid Kidney Lung CI-LLI Nuclide H-3 No Data 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 C-14 2.84E-06 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 i P-32 1.93E-04 1. 2 0E-0 5 7.46E-06 No Data No Data No Data 2.17E-05 j Cr-51 No Data No Data 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 I Mn-54 No Data 4.57E-06 8.72E-07 No Data 1.36E-06 No Data 1.40E-05 1.15E-07 2.04E-08 No Data 1.46E-07 No Data 3.67E-06 Mn-56 No Data Fe-55 2.75E-06 1.90E-06 4.43E-07 No Data No Data 1.06E-06 1.09E-06 Fe-59 4.34E-06 1.02E-05 3.91E-06 No Data No Data 2.85E-06 3.40E-05 j Co-58 No Data 7.45E-07 1.67E-06 No Data No Data No Data 1.51E-05 Co-60 No Data 2.14E-06 4.72E-06 No Data No Data No Data 4.02E-05 Ni-63 1.30E-04 9.01E-06 4.36E-06 No Data No Data No Data 1.88E-06 Ni-65 5.28E-07 6.86E-08 3.13E-08 No Data No Data No Data 1.74E-06 i Cu-64 No Data 8.33E-08 3.91E-08 No Data 2.10E-07 No Data 7.10E-06 Zn-65 4.84E-06 1.54E-05 6.96E-06 No Data 1.03E-05 No Data 9.70E-06 Zn-69 1.03E-08 1.97E-08 1.37E-09 No Data 1.28E-08 No Data 2.96E-09 5.79E-08 Br-83 No Data No Data 4.02E-08 No Data No Data No Data Br-84 No Data No Data 5.21E-08 No Data No Data No Data 4.09E-13 Br-85 No Data No Data 2.14E-09 No Data No Data No Data No Data Rb-86 No Data 2.11E-05 9.83E-06 No Data No Data No Data 4.16E-06 Rb-88 No Data 6.05E-08 3.21E-08 No Data No Data No Data 8.36E-19 l Rb-89 No Data 4.01E-08 2.82E-08 No Data No Data No Data 2.33E-21 l 4.94E-05 l Sr-89 3.08E-04 No Data 8.84E-06 No Data No Data No Data Sr-90 7.58E-03 No Data 1.86E-03 No Data No Data No Data 2.19E-04 Sr-91 5.67E-06 No Data 2.29E-07 No Data No Data No Data 2.70E-05 All values are in (mrem /pc1 ingested). They are obtained from Reference 3 (Table E- 11 ) , except as follows: Reference 2 (Table A-3) for Rh-105, Sb-124, and Sb-12 5. 9-41 Rev. 8, 1/94 l
l Hate- ODCM Tacle 9-14 (c:ntd). Ingest.cn Deze Factors for the Adult Age Group j Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI l l Sr-92 2.15E-06 No Data 9.30E-08 No Data No Data No Data 4.26E-05 Y-90 9.62E-09 No Data 2.58E-10 No Data No Data No Data 1.02E-04 Y-91m 9.09E-11 No Data 3.52E-12 No Data No Data No Data 2.67E-10 Y-91 1.41E-07 No Data 3.77E-09 No Data No Data No Data 7.76E-05 i Y-92 8.45E-10 No Data 2.47E-11 No Data No Data No Data 1.48E-05 ; Y-93 2.68E-09 No Data 7.40E-11 No Data No Data No Data 8.50E-05 Zr-95 3.04E-08 9.75E-09 6.60E-09 No Data 1.53E-08 No Data 3.09E-05 Zr-97 1.68E-09 3.39E-10 1.55E-10 No Data 5.12E-10 No Data 1.05E-04 Nb-95 6.22E-09 3.46E-09 1.86E-09 Ho Data 3.42E-09 No Data 2.10E-05 l Mo-99 No Data 4.31E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06 l Tc-99m 2.47E-10 6.98E-10 8.89E-09 No Data 1.06E-08 3.42E-10 4.13E-07 Tc-101 2.54E-10 3.66E-10 3.59E-09 No Data 6.59E-09 1.87E-10 1.10E-21 Ru-103 1.85E-07 No Data 7.97E-08 No Data 7.06E-07 No Data 2.16E-05 Ru-105 1.54E-08 No Data 6.08E-09 No Data 1.99E-07 No Data 9.42E-06 l Ru-106 2.75E-06 No Data 3.48E-07 No Data 5.31E-06 No Data 1.78E-04 Rh-105 1.22E-07 8.86E-08 5.83E-08 No Data 3.76E-07 No Data 1.41E-05 Ag-110m 1.60E-07 1.48E-07 8.79E-08 No Data 2.91E-07 No Data 6.04E-05 Sb-124 2.81E-06 5.30E-08 1.11E-06 6.79E-09 No Data 2.18E-06 7.95E-05 Sb-125 2.23E-06 2.40E-08 4.48E-07 1.98E-09 No Data 2.33E-04 1.97E-05 Te-125m 2.68E-06 9.71E-07 3.59E-07 8.06E-07 1.09E-05 No Data 1.07E-05 Te-127m 6.77E-06 2.42E-06 8.25E-07 1.73E-06 2.75E-05 No Data 2.27E-05 i I j Te-127 1.10E-07 3.95E-08 2.38E-08 8.15E-08 4.48E-07 No Data 8.68E-06 Te-129m 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 No Data 5.79E-05 l Te-129 3.14E-08 1.18E-08 7.65E-09 2.41E-08 1.32E-07 No Data 2.37E-08 Te-131m 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 No Data 8.40E-05 Te-131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 No Data 2.79E-09 . 9-42 Rev. 8, 1/94 l
Hatch ODCM l l Table 9-14 (contd). Ingestion Dose Facters fcr the Adult Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 No Data 7.71E-05 I-130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 No Data 1.92E-06 I-131 4.16E-06 5.95E-06 3.41E-06 1.95E-03 1.02E-05 No Data 1.57E-06 I-132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 No Data 1.02E-07 I-133 1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31E-06 No Data 2.22E-06 I-134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 No Data 2.51E-10 I-135 4.43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 No Data 1.31E-06 Cs-134 6.22E-05 1.48E-04 1.21E-04 No Data 4.79E-05 1.59E-05 2.59E-06 Cs-136 6.51E-06 2.57E-05 1.85E-05 No Data 1.43E-05 1.96E-06 2.92E-06 l Cs-137 7.97E-05 1.09E-04 7.14E-05 No Data 3.70E-05 1.23E-05 2.11E-06 Cs-138 5.52E-08 1.09E-07 5.40E-08 No Data 8.01E-08 7.91E-09 4.65E-13 Ba-139 9.70E-08 6.91E-11 2.84E-09 No Data 6.46E-11 3.92E-11 1.72E-07 Ba-140 2.03E-05 2.55E-08 1.33E-06 No Data B.67E-09 1.46E-08 4.18E-05 l Ba-141 4.71E-08 3.56E-11 1.59E-09 No Data 3.31E-11 2.02E-11 2.22E-17 l l Ba-142 2.13E-08 2.19E-11 1.34E-09 No Data 1.85E-11 1.24E-11 3.00E-26 l l La-140 2.50E-09 1.26E-09 3.33E-10 No Data No Data No Data 9.25E-05 I l l La-142 1.28E-10 5.82E-11 1.45E-11 No Data No Data No Data 4.25E-07 l Ce-141 9.36E-09 6.33E-09 7.18E-10 No Data 2.94E-09 No Data 2.42E-05 l Ce-143 1.65E-09 1.22E-06 1.35E-10 No Data 5.37E-10 No Data 4.56E-05 Ce-144 4.88E-07 2.04E-07 2.62E-08 No Data 1.21E-07 No Data 1.65E-04 Pr-143 9.20E-09 3.69E-09 4.56E-10 No Data 2.13E-09 No Data 4.03E-05 Pr-144 3.01E-11 1.25E-11 1.53E-12 No Data 7.05E-12 No Data 4.33E-18 l Nd-147 6.29E-09 7.27E-09 4.35E-10 No Data 4.25E-09 No Data 3.49E-05 W-187 1.03E-07 8.61E-08 3.01E-08 No Data No Data No Data 2.82E-05 Np-239 1.19E-09 1.17E-10 6.45E-11 No Data 3.65E-10 No Data 2.40E-05 9-43 Rev. 8, 1/94 l l_______-__ . -- - . . - - -
Hatch ODCM Table 9-15. External Dose Factors for Standang on Contaminated Ground Nuclide T. Body Skin Nuclide T. Body Skin H-3 0.00 0.00 Sr-91 7.10E-09 8.30E-09 C-14 0.00 0.00 Sr-92 9.00E-09 1.00E-08 Na-24 2.50E-08 2.90E-08 Y-90 2.20E-12 2.60E-12 P-32 0.00 0.00 Y-91m 3.80E-09 4.40E-09 Cr-51 2.20E-10 2.60E-10 Y-91 2.40E-11 2.70E-11 Mn-54 5.80E-09 6.80E-09 Y-92 1.60E-09 1.90E-09 Mn-56 1.10E-08 1.30E-08 Y-93 5.70E-10 7.80E-10 Fe-55 0.00 0.00 Zr-95 5.00E-09 5.80E-09 ! Fe-59 8.00E-09 9.40E-09 Zr-97 5.50E-09 6.40E-09 Co-58 7.00E-09 8.20E-09 Nb-95 5.10E-09 6.00E-09 l Co-60 1.70E-08 2.00E-08 Mo-99 1.90E-09 2.20E-09 Ni-63 0.00 0.00 Tc-99m 9.60E-10 1.10E-09 Ni-65 3.70E-09 4.30E-09 Tc-101 2.70E-09 3.00E-09 l l Cu-64 1.50E-09 1.70E-09 Ru-103 3.60E-09 4.20E-09 Zn-65 4.00E-09 4.60E-09 Ru-105 4.50E-09 5.10E-09 l Zn-69 0.00 0.00 Ru-106 1.50E-09 1.80E-09 Br-83 6.40E-11 9.30E-11 Rh-105 6.60E-10 7.70E-10 Br-84 1.20E-08 1.40E-08 Ag-110m 1.80E-08 2.10E-08 Br-85 0.00 0.00 Sb-124 1.30E-08 1.50E-08 Rb-86 6.30E-10 7.20E-10 Sb-125 3.10E-09 3.50E-09 Rb-88 3.50E-09 4.00E-09 Te-125m 3.50E-11 4.80E-11 Rb-89 1.50E-08 1.80E-08 Te-127m 1.10E-12 1.30E-12 Sr-89 5.60E-13 6.50E-13 Te-127 1.00E-11 1.10E-11 Sr-90 0.00 0.00 Te-129m 7.70E-10 9.00E-10 l ! All values are in (mrem /h) per (pCi/m )2 . They are obtained from Ref erence 3 (Table E-6), except as follows: Reference 2 (Table A-7 ) for Rh-105, sb-124, and Sb-125. 9-44 Rev. 8, 1/94 i r
1 Hatch ODCM Table 9-15 (contd). External Dese f actors f or Standing en contaminatec Ground l Nuclide T. Body Skin f Te-129 7.10E-10 8.40E-10 Te-131m 8.40E-09 9.90E-09 Te-131 2.20E-09 2.60E-06 Te-132 1.70E-09 2.00E-09 I-130 1.40E-08 1.70E-08 I-131 2.80E-09 3.40E-09 l I-132 1.70E-08 2.00E-08 3.70E-09 4.50E-09 ) l I-133 I-134 1.60E-08 1.90E-08 i I-135 1.20E-08 1.40E-08 l Cs-134 1.20E-08 1.40E-08 Co-136 1.50E-08 1.70E-08 l Cs-137 4.20E-09 4.90E-09 l Cs-138 2.10E-08 2.40E-08 1 Ba-139 2.40E-09 2.70E-09 Ba-140 2.10E-09 2.40E-09 Ba-141 4.30E-09 4.90E-09 Ba-142 7.90E-09 9.00E-09 La-140 1.50E-08 1.70E-08 La-142 1.50E-08 1.80E-08 i Ce-141 5.50E-10 6.20E-10 l Ce-143 2.20E-09 2.50E-09 l Ce-144 3.20E-10 3.70E-10 i ! Pr-143 0.00 0.00 Pr-144 2.00E-10 2.30E-10 Nd-147 1.00E-09 1.20E-09 W-187 3.10E-09 3.60E-09 Np-239 9.50E-10 1.10E-09 I l l l 9-45 Rev. 8, 1/94
- . - , - _ . - - ~.. .- - -. -. . . - . - . . - - . - -..--_ .-._.
i i l I Hat-- DDCM J l CHAPTER M f DEFINITIONS OF EFFLUENT CONTROL TERyj l l l The terme defined in this chapter are used in the presentation of the above chapters. These terms are shown in all capital letters to indicate that they are specifically defined. 10.1 TERMS SPECIFIC TO THE ODCM The following terms are used in the ODCM, but are not found in the Technical l Specifications: l 1 l BATCH RELEASE Prior to A BATCH RELEASE is the discharge of wastes of a discrete volume. l sampling for analyses, each liquid batch shall be isolated andassure then l method described at the 00CM to thoroughly mixed by a representative sampling. pOMPOSITE SAMPLE A COMPOSITE SAMPLE is one which contains mater;al f rom multiple waste ' releases, in which the quantity of sample is prcportional to the quantity
)
of waste discharged, and in which the method of sampling employed results in a specimen that is representative of the wastes released. Prior to f analyses, all liquid samples that are to be al;quotted for a COMPOSITE SAMPLE shall be mixed thoroughly, in order f or t..e COMPOSITE SAMPl.E to be ( representative of the effluent release.- When assessing the consequences of a waste release at the pre-release or post-release stage, the most recent available COMPOSITE SAMPLE results for the applicable release pathway may ce used. CONTINUOUS RELEASE A CONTINUOUS RELEASE is the discharge of wastes cf a non-discrete volume, l e.g., f rom a volume within a system that has an input flow during the continuous release. GASEOUS RADWASTE TREATMENT SYSTEM The GASEOUS RADWASTE TREATMENT SYSTEM is the of f gas holdup system designed l and installed to reduce radioactive gaseous offluents by collecting primary coolant system offgases from the pr mary system and providing for delay or holdup for the purpose of reducing the total radicactivity prior to release to the environment. 10-1 Rev. S, 1/94 l
. . . _ . , - _-~ - , _ , . - _, --
Hatch ODCM f LIOUID RADWASTE TM. ATMENT SYSTEM installed to A LIQUID RADWASTE TREATMENT SYSTEM is any system designed and systematic ! itquad effluents by reduce rad active materials in collection, retention, and processing through filtration, eva porat ion, This system consists of at separation and/or ion exchange treatment. l least one collection tank, one evaporator or demineralizer system, one l post-treatment tank and associated components providing for treatment flow l and functional control. } 4 MAJOR CMANGES TO 3ADIOACTIVE WASTE TREATMENT SYSTEMS For the purposes of the ODCM, MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS include the f ollowing changes to such systems: ( (1) Major changes in process equipment, components, structures, or effluent monitoring instrumentation as described in the Final Safety Analysis Report (FSAR) or as evaluated in the Nuclear Regulatory Commission staf f's Saf ety Evaluation Report (SER) (e.g., deletion of evaporators and .nstallation of demineralizers); I 1 (2) Changes in the design of radwaste treatment systems that could significantly increase quantities of ef fluents released f rom those previously considered in the FSAR and SER; (3) Changes in system design which may invalidate the accident analysis as described in the SER (e.g. , changes in tank capacity that would l alter the curies released); er (4) Changes in system design that could potentially result in a ! in occupational exposure of operating significant increase personnel (e.g., use of tempcrary equipment without adequate shielding provisions), MINIMUM DETECTABLE CONCENTRATION The MINIMUM DETECTABLE CONCENTRATION (MDC) is defined, for purposes of the controle in this ODCM, as the smallest concentration of radioactive material in a sample that will yield a net count above system background and that will be detected with 95-percent probability, with only 5-percent a blank observation represents a j probability of falsely concluding that !
"real" signal.
For a particular measurement system, which may include radiochemical separation, the MDC for a given radionuclide is determined as follows (Reference 17): 10-2 Rev. 8, 1/94
Hatch OOCM l
- I
+ 3. 2 9 Rb E
E 5 % .Es b MDC = N E V= 2.22 x lob *Y=e where MDC = the a priori MINIMUM DETECTABLE CONCENTRATION (pCi per unit mass or volume). Rb= the background counting rate, or the counting rate of a blank sample, as appropriate (counts per minute), t,
= the length of the sample counting period (minutes).
the length of the background counting period (minutes). ! tb= E= the counting efficiency (counts per disintegration) V = the sample size (units of mass or volume). 2.22 x 106 = the nu:nber of disintegrations per minute per pCi. Y= the fractional radiochemical yield, when applicable. 1= the radioactive decay constant for the given radionuclide ! (h-I), Values of 1 used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 26. I At = for effluent samples, the elapsed time between the midpoint cf sample collection and the time of counting (h); for environmental samples, the elapsed time between the end of sample collection and the time of counting (h). Typical values of E, V, Y, and at should be used in the calculation. It should be recognized that the MDC is defined as an a prior 2 (before the fact) limit representing the capability of a measurement system, and not I as an a poster 2cr: (after tne fact) lim;; for a partic.:lar measurement. f l l PRINCIPAL GAMMA EMITTERS l The PRINCIPAL GAMP.A EMITTERS for which the MINIMUM DETECTABLE CONCENTRATION (MDC) limit applies include exclusively the following radio-nuclides: e For liquid radioactive effluents: Mn-54, Fe- 5 9 , co- 5 8 , C o- 6 0, 2n-65, Mo- 9 9 , Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an MDC of 5 x 10-6 pC1/mL. l 10-3 Rev. 8, 1/94
Hatch 00CM i For gaseous radioactive effluents: in nt:.e gas releases, Fr-s?, I Kr-88, Xa-133, Xe- 13 3m , Xe-135, Xe-13B; and in part;: . ate f releases, Mn- 54, Fe- 59, co- 5 8, co- 60,
- n-6 5, Mo-99, Cs- 13 4, Cs- 13 7, Co- 141, and Co-144.
- For environmental media: The gamma emitters specifically listed in Table 4-3.
These liste do not mean that only these nuclides are to be considered. l l Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Report, the Annual Radtological Environmental Effluent Release Surveillance Report, or other applicable reportis). i f i I Rey, s, 1/94 10-4
)
F
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i Hatch ODCM 10.2 TERMS DEFINED IN THE TECHNICAL SPECIFICATIONS % The following terms are defined in the Technical Specifications, Section 1.0. Because they are used throughout the Limits of Operation sections of the ODCM, In the event of discrepancies between they are presented here for convenience. the definitions below and those in the Technical Specifications, the Technical Specification definitions shall take precedence. ACTIONISi For Unit la An ACTION shall be that part of a control that prescribes remedial measures required under designated conditions. For Unit 2: ACTIONS shall be those additional requirements specified as of the control. corollary statements to each control, and shall be part Cf_ANNEL CALIBRATION i A CHANNEL CALIBRATION shall be the u3ustment, am r.ecessary, of the channel output, such that it responds within. the required range and The accuracy to known values of the parameter which the channel monitors. CHANNEL CALIBRATION shall encompass the entire channel including the ' sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps, such that the entire channel is calibrated, f CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of ti.e channel indication and/or status with other indications and/or status derived frem independent instrument channels l measuring the same parameter. CHANNEL FUNCTIONAL TEST A CRANNEL FUNCTIONAL TEST shall bei e Analog Channels - the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including: for both units, alarm and/or trip l I functions; and for Unit 2 only, channel failure trips. e Bistable Channels - the injection of a simulated signal into the sensor to verif y OPERABILITY including alarm and/or trip f unctions. 10-5 Rev. 8, 1/94
l Hatch 00CM ES E EOUIVALENT !-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (uci/g) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 or those in Table E-7 of NRC Regulatory cuide 1.109, Revision 1, 1977. FREOUENCY NOTATION The FREQUENCY NOTATION specified for the performance of surveillance requirements shall correspond to the intervals defined below, with a l maximum allowable extension not to exceed 25% of the surveillance interval. l NOTATION FREOUENCY i i ' S (Once per shift) At least once per 12 hours. D (Daily) At least once per 24 hours. W (Weekly) At least once per 7 days. l M (Monthly) At least once per 31 days. l Q (Quarterly) At least once per 92 days. I SA (Semi-annually) At least once per 184 days. 1 R (Refueling) At least once per 18 months. S/U (Startup) Prior to each reactor startup. NA Not applicable. P (Prior) Completed prior to each release. , I I MEMBERfS) OF THE PUBLIC ! A MEMBER OF THE PUBLIC shall be an individual in a controlled area or an UNRESTRICTED AREA. However, an individual is not a MEMBER OF THE PUBLIC during any period in which the individual receives an occupac2cnal dose. This category may include persons whc use portions of the site for recreational, occupational, or Other p;rpcses not associated with the plant. MILK ANIMAL A MILK ANIMAL is a cow or goat that is producing milk for human consumption. I The italicized terms in this definition, which are not otherwise used in this ODCM, shall have the defin'.tions assigned to them by 10 CFR 20.1003. 10-6 Rev. 8, 1/94
Hatch ODCM OPERABLE for OPERABILITY 1 OPERABILITY exists when a system, subsystem, train, component or dev;ce is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its f unction ( s) are also capable of performing their related support function (s). OPERATIONAL CONDITION For Unit 1: This term is not defined. See definition of REACTOR MODE. For Unit 2: An OPERATIONAL CONDITION shall be any one inclusive combination of Mode Switch position and average reactor coolant as defined in Section 1.0 of the Unit 2 Technical temperature, Specifications. REACTOR MODE For Unit la The REACTOR MODE is established by the Mode Switch position. The four Mode Switch positions are REFUEL, SHUTDOWN, START & HOT STANDBY, and RUN; thus the four possible REACTOR MODES are: Pefuel Mode, Shutdown Mode, Start & Hot Standby Mode, and Run Mode. (See Unit 1 Technical Specifications Section 1.0 for definitions of these terms. ) For Unit 2: This term is not defined. See definition of OPERATIONAL CONDITION. RATED THERMAL POWER For Unit 1: RATED THERMAL POWER is operation at a steady state power of 2436 KWt. This is also referred to as 100 percent thermal power. 1 I For Unit 2: RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant cf 2436 MNt. SITE BOUNDARY For the purpose of effluent controls defined in the ODCM, the SITE BOUNDARY shall be as shown in Figure 1D-1. l SOURCE CHlCl A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. 10-7 Rev. E, 1/94 I y - , - _
I Hntch ODCM 1 1 i THEPMAL POWER For Unit la This term is not defined. For the purposes of effluent ! I controle in the ODCM requiring special sampling in the event of specified changes in THERMAL POWER, the detinition shall be taken to be the same as l that for Unit 2. For Unit 2: THERMAL POWER shall be the total reactor core heat transfer rate to,the reactor coolant. UNRESTRICTED AREA The UNRESTRICTED AREA shall be any area access to which is neither limited nor controlled by the licensee, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. l l I r 1 I I i l l l 10-8 Rev. 8, 1/94
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I 1 1 SOLID RADIOACTIVE WASTE PROCESS CONTROL PROGRAM FOR THE I GEORGIA POWER COMPANY EDWIN l. HATCH NUCLEAR PLANT JANUARY 1994, REVISION 4 1
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I l t i l
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_ _.. 7 l-QtSTRIBUTION LIST Cczy bh Recioient Location - i Reddick, R. G. Plant Hatch - Simulator Bldg. 1 , i Baxley, Georgia l
. l Plant Hatch - Service Bldg. 2 l Sorrell, E. C.
l I Document Control ! I
- Baxley, Georgia
' Plant Hatch Service Bldg. .3 Lewis, J. C.
Baxley, Georgia l l I Plant Hatch - Service Bldg. 4 l Arnold, Brian Baxley, Georgia l Plant Hatch - Service Bfdg. 5 Smith, Dorsey l 1 Baxley, Georgia l Plant Hatch - Service Bldg. 6 Bennett, Deryte 7 l Baxley,. Georgia Plant Hatch - Simulator Bldg. 7 l
, Tipps, S. B.
Baxley, Georgia j l ! Plant Hatch Service Bldg. 8 i . Sorrell, E. C. Baxley, Georgia j Southem Nuclear Operating Company 11 Hopper, D. M. I l Inverness Center, Bldg. 40 2 Birmingham, Alabama
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Nichols, M. C. ggia Power Company 15
-' Lovironmental Center 5131 Maner Road
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Smyrna, Georgia Southern Company Services 16 l Wehrenberg, J. A. l
- Inverness Center, Bldg. 42 Birmingham, Alabama i
Southern Company Services 18 Hempstead, J. S. 3 inverness Center, Bldg. 42
,I Birmingham, Alabama Robson, G. W. Southern Nuclear Operating Company 20 inverness Center, Bldg. 40 Birmingham, Alabama Revision 4,1/94 ww- .w en.ie.
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I i i ! EDWIN 1. HATCH NUCLEAR PLANT t l SOLID RADIOACTIVE WASTE PROCESS CONTROL PROGR.A_td
- 1. PURPOSE The processing of radioactive w3ste for disposal at a licensed radioactive waste burial site requires that the waste be appropriately analyzed, processed, and packaged, prer.enting a final waste form that is acceptable for transportation and burial at a licensed radioactive waste disposal site. The purpose of this Process Control Program is to document the radioactive waste processing methods and the quality control steps that are taken at the Edwin 1. Hatch Nuclear Plant to verify compliance with applicable regulatory requirements and, in particular, to assure an acceptable waste product meeting the applicable waste stability characteristics of 10 CFR 61.56. NRC Generic Letter 89-01 allows nuclear power plant licensees to transfer procedural details formerly contained in Technical Specification 3.15.3 (Unit 1) and 3.11.3 (Unit 2) to the Process Control Program. These procedural details, along with associated definitions and reporting requirements, are included in the Hatch Nuclear Plant Process Control l
i Program as Appendix A. At Plant Hatch, the routine processing systems that generate radioactive waste requiring offsite disposal are:
- 1. The reactor water cleanup system.
- 2. The condensate cleanup system.
- 3. The spent fuel pool cleanup system.
- 4. The liquid radwaste system.
All of these systems employ the use of either ion exchange resins or the powdered ion exchange resins as the processing medium. Spent resins are dewatered in an appropriato liner based on the waste class and burial site criteria. In addition to these procevQ syste.1s, two other types of radioactive waste that are routinely processed are the compactible and noncompactible trash (DAW) and slightly radioactively contaminated oils. 1 Revision 4,1/94
l l This Process Control Program addresses the processing of these types of waste and the measures in place to assure the generation of an acce~ptable waste product. Operating criteria for spent resin transfer to an appropriate liner and in-liner dewatering are addressed, as are ! also the criteria for the processing of DAW (compactible and noncompactible trash) and i miscellaneous contaminated liquids (including oils). l ll. REGULATORY OVERVIEW I All waste processing. packaging, and shipping are conducted in accordance with approved l f procedures to assure compliance with applicable federal, state, and burial-site requirements. Waste processed for disposat is evaluated per approved plant procedures for compliance with: l
- 1. Waste classification requirements of 10 CFR 61.55.
f 2. Waste characteristic requirements off 10 CFR 61.56.
- 3. Manifest reporting requirements of 10 CFR 20.2006.
l Waste is packaged in containers meeting or exceeding the requirements for both l l 1 ( transportation and disposal. Shipments are conducted in accordance with the requirements of 49 CFR 172-177 and 10 CFR 71. l All waste processing is ce mi in a manner consistent with the principles of ALARA. The l procedures that have . md to cover waste processing operations address appropriate radiation safet ,ch as job preplanning, radiation source shielding, and job prerequisites and materim nents so as to minimize stay times. 1
- lll. DEWATERING OF RESINS t
The processing of the liquid Ltreams by ion exchange resins (bead or powdered) results in a I waste product that is most appropriately dewatered in a suitable disposable liner (carbon steel j l or high-integrity container). Prior to transfer of the spent resin to a liner, a sample is collected 4 1 Revision 4,1/94 l 2
I and analyzed by gamma spectroscopy to quantify the radioactive material concentration. Based on this inf ormation, the waste is appropriately classified in accordance with approved 4 procedures per the criteria of 10 CFR 61.55, if a sample of the resins cannot be collected prior to transfer to a liner due to either design or operationa! limitations, a sample collected 5 after transfer or an external radiation level measurement may be used in accordance with l i approved procedures for determining the waste class. I The dewatering process is conducted in accordance with approved procedures with l j } appropriate operating parameters to assure a waste product with as little freestanding water 1 as possible but,in no case,in excess of 1 percent by volume (i.e., meeting the waste stability ' criteria of 10 CFR 61.56). The specifics cl the dewatering process vary depending upon the i type of dewatering process employed and the type of resin (bead or powdered). However, $ ' 'the common approach to dewatering is the removal of essentially all interstitial water. i Appropriate verifications (dependent upon the process method) are conducted to assure an j l acceptable waste form. i i IV. PACKAGING OF DAW (COMPACTIBLE AND NONCOMPACTIBLE TRASH) l g s All radioactively contaminated trashis appropriately packaged, surveyed, and labeled prior to shipment for disposal. Bagged wastes are opened to assure exclusion of unacceptable waste j , ! products, such as wate' and oil. Compactible trash is processed by compaction to reduce the
- volume of the waste; noncompactible materials are segregated and normally packaged separately. Af ter packaging, the waste containers are stored awaiting shipment. Container l
j integrity is verified by visual examination (post compaction and pre shipment) to assure an acceptable waste package for transportation and an acceptable waste form for disposal, i 4 , 1 . The dewatering process at Plant Hatch is a contractor supplied system. Refer to Section l i Vi for the process controls that are applicable. l Revision 4,1/94 3 2
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l V. PROCESSING OF MISCELLANEOUS LIQUIDS (Oll) Periodically, it may become necessary to process slightly radioactively contaminated liquids, , i including oil, for offsite disposal. These liquid wastes are either processed by absorption or solidification. Absorption is accomplished using at least twice as much absorbent as is i r.ecessary to completely absorb the liquid; only an absorbent approved by the burial site is used. Any solidification of liquids is conducted in accordance with approved plant procedures
- or a contractor-supplied solidification procedure that has beer'specifically developed for the 1
l solidification process. (Refer to Section VI for the use of a contractor or contractor-supplied
- process for waste processing.)
l l The solidification of liquids (including oils) utilizing the onsite mixing unit is conducted in accordance with approved operating procedures and parameters. As appropriate, test samples are conducted every tenth batch to verify the pre-established mixing ratios. Sampling is accomplished in accordance with approved plant procedures to assure that the samples are representative of the waste being processed. Any unacceptable products require a re-i evaluation of the mixing ratios; any unacceptable waste containers are identified (by visual l inspection), and appropriate actions are taken to assure that the waste as shipped meets the applicable waste stability requirements of 10 CFR 61.56. Additionally, all containers of the solidified waste are inspected prior to closure to assure a' solidified matrix absent of any freestanding water. VI. USE OF CONTRACTOR FOR WASTE PROCESSING Contractor-supplied process and/or service may be used at Plant Hatch for the processing of radioactive waste for offsite disposal. For the operation of such a process,it may be desirable to use process control measures and procedures developed by the contractor specifically for the system or process. Therefore, previously addressed process control measures for a particular type waste may be superseded by contractor-supplied measures, as appropriate. l 4 Revision 4,1/94
l l l Prior to the use of a contractor-supplied process or service for waste processing at Plant Hatch, a management review of the contractor's process control and procedures is perf ormed f to assure an operation compatible with plant operation and in accordance with regulatory f requirements. Contractor-supplied waste processing shall be perf ormed in areas with features I j and/or controls adequate to contain inadvertent spills and overfills. ; For the processing of waste that is intended to be shipped for disposal at a licensed radioactive waste burial site, additional precautions are taken to assure a final waste product that meets the ::ppropriate waste stability requirements of 10 CFR 61.56. In particular, the following items, as applicable, are to be documented by the contractor (or Plant Hatch manuals or procedures) prior to utilization for waste processing: i O A general description of the solidification or dewatering process, i including type solidification agent (if applicable), major process ) equipment and interf ace with plant equipment, types of waste that can be processed, and operating parameters. O Process control measures that provide for the verification of the generation of a suitable waste product, including items (as may be l I appropriate f or the process method), such as representative sampling, laboratory tests, and acceptacle criteria. O Specifically approved procedures for the operation of the process equipment that will assure operation within the bounds as dehneated by the process control measures. O Appropriate acceptance criteria for evaluating the acceptability of the final waste product. 1 t Waste products will be verified as meeting the criteria f or disposal prior to final closure of the container. Revision 4,1/94 5
d APPENDIX A SOLID RADWASTE DEFINITIONS, PROCED'JRAL DETAILS AND REPORTING REQUIREMENTS PER NRC GENERIC LETTER 89-01 This Appendix to the Hatch Nuclear Plant (HNP) Process Control Program (PCP) contains the definitions, procedural details, and reporting requirements pertaining to solid radioactive waste f ormerly presented in the HNP Technical Specifications. These definitions, procedural details.
! and reporting requirements have been transferred to the PCP in accordance with NRC Gencric Letter 89-01. In the body of the text of this Appendix, terms that have been specifically defined appear in all capitalletters to indicate that these terms have specific definitions. The definition of PROCESS CONTROL PROGRAM and the discussion of changes and approval of changes to the PCP are presented in HNP Technical Specifications and also are included in
- Appendix A for convenience.
A.1 DEFINITIONS A.1.1 OPERABLE - OPERABILITY l l . A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY I when it is capable of performing its specified function (s), implicit in this definition shall be the assumption that all necessary attendant instrumentation; controls, normal and emergency l electrical power sources; cooling or seal water; lubrication or other auxiliary equipment that } is required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s). i j A-1 Revision 4,1/94
l 1 l A.1.2 PROCESS CONTROL PROGRAM *
- l. l The PROCESS CONTROL PROGRAM (PCP) shall be implemented by procedures which 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 compliance with 10 CFR Parts 20,61, and 71, State regulations, burial ground requirements, i and other requirements governing the disposal of solid radioactive waste.
l A.1.3 SOLIDIFICATION SOLIDIFICATION shall be the conversion of wet radioactive wastes into a form that meets shipping and burial-ground requirements. 1 I A.2 CHANGES TO THE PROCESS CONTROL PROGRAM i l j Changes to the PCP shall meet the following requirements: l 1. They shall be documented, and records of reviews performed shall be retained as required by Technical Specification 6.10.2.o. This documentation shall contain the following:
- a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s).
- b. A determination that the change will maintain the overall l l conformance of the solidified waste _ product to existing requirements of Federal, State, or other applicable regulations.
- 2. They shall become effective after review and acceptance by the PRB and the approval of the General Mr. nager Nuclear Plant.
t A2 Revision 4.1/94
A.3 SOLID RADIOACTIVE WASTE SYSTEM A.3.1 Solid Radioactive Waste System Control The solid radwaste system shall be used in accordance with the PROCESS CONTROL PROGRAM to provide for the SOLIDIFICATION of wet solid wastes and for the SOLIDIFICATION and packaging of other radioactive wastes, as required, to ensure that they meet requirements of 10 CFR Parts 20 and 71, prior to shipment of radioactive wastes from the site. i A.3.2 Applicability f This requirement applies et all times. l A.3.3 Actions l A.3.3.1 With the requirements of 10 CFR Parts 20 and 71 not satisfied, suspend shipments of defective containers of solid rcdioactive wastes from the site. A.3.3.2 For Unit 1: When the ACTION statement or other requirements of this control cannot be met, steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified condition may be made if, as a minimum, the requirements of the ACTION statement are satisfied. For Unit 2: The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable. l A-3 Revision 4.1/94
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l A.3.4 Surveillance Requirements
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The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of wastes prior to shipment. A.3.5 Bases The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging before being shipped offsite. This control implements the requirements of 10 CFR Part 50.36(a) and General Design Cnterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in. establishing the PROCESS CONTROL PROGRAM may inc.ade, but are not limited to, waste , type, waste pH, waste / liquid / solidification agent / catalyst ratics, waste oil content, waste l principal chemical constituents, and mixing and curing times. A.4 REPORTS A.4.1 Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Release Report, submitted in accordance with Technical Specification 6.9.1.8, shall include a summary of the quantities of solid radwaste released from the units as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting ! Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light Water-Cooled Nuclear Power Plants." Revision 1. June 1974, with data summarized on a 6. month basis following the format of Appendix B thereof, Reporting requirements presented in this section address only solid udioactive wastes. For a comprehensive presentation of reporting requirements pertaining to the Annual Radioactive ; Effluent Release Report, see the Plant Hatch Offsite Oose Calculation Manual. A.4 Revision 4,1/94 i
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l For each type of solid radwaste shipped offsite during the report period, the report shall j include the following information: l l
- a. Container volume.
- b. Total curie quantity (specify whether determined by measurement or estimate).
- c. Principal radionuclides (specify whether determined by measurement or estimate).
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- d. Type of waste (such as spent resin, compacted dry waste, evaporator
- bottoms),
- e. Type of container (such as LSA, type A. type B, large quantity).
i ! f. Solidification agent (such as cement). l l Major changes to the solid radioactive waste treatment system shall be reported to the Nuclear Regulatory Commission in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed and accepted by the PRB. l i A-5 Revision 4,1/94 l l I I
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