HL-2371, Plant Ei Hatch Units 1 & 2 Semiannual Radioactive Effluent Release Rept,Jan-June 1992
| ML20127C754 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/30/1992 |
| From: | Beckman J GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| HL-2371, NUDOCS 9209100094 | |
| Download: ML20127C754 (95) | |
Text
{{#Wiki_filter:cwoga twer cgewy 4 0 l%t*.f 4( t h (t.Ptf"i IdFkAdf l\n* o"-ce 144 n-95 4 8 Um .tgt' a. tu.nen a . ,O I Telepw.e ;m P 7 7 72 's u n eu n,.m. .,,. Georgia Power vcen.s w ra a '# H 4z t .T h f "! 't_+ . ! HL-2371 0038Ci August 28, 1992 U.S. Nuclear Regulatory Commission ATTil: Document Control Desk Washington, D.C. 20555 PLAliT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERATING LICEllSES DPR-57, NPF-5 SEMIANNUAL RADICACTIVE EFFLUENT RELfASE REPORT 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 Unit I and Unit 2 Semiannual Radioactive Effluent Release Report. This report covers the period of January 1, 1992 through June 30, 1992. Should you have any questions, piease advise. Sincerely, 0h Y J. T. Beckham, Jr. SRM/cr Enclosure cc: ( ee next page.) A 9 1
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Georgia Power d U.S. Nuclear Regulatory Commission August 28, 1992 Page Two cc: Georgia Powfr Comnany Mr. H. L. Sumner, General Manager - Nuclear Plant Mr. H. C. Nichols, Manager - Environmental Laboratory NORMS U.S. Nuclear Reaulatory Commission. Washinoton. D.L Mr. K. Jabbour, Licensing Project Manager - Hatch
.V& Nuclear Reaulatory Commission. Reaion 11 Mr. S. D. Ebneter, Regional Administrator Mr. L. D. Wert, Senior Resident Inspector - Hatch American Nuclear. insurers Mr. M. Marugg i
State of Georaig Mr. J. L. Setser, Department of Natural Resources 003864 70t3tb :-
l , GEORGIA POWER COMPANY PLOT E.1. KAToi UNITS NO.1 6 2 SD1!AWUAL RADIOACTIVE EFFLUENT RELFASE REPORT January 1,1992 - June 30,1992 I e i O-M O >
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PIANT E. I. RATO{ SEMIANNUAL RADI0 ACTIVE ('] LJ EFFLUENT RELEASE REPORT SECTION TITLE PAGE
- 1. LIQUID EFFLUENTS 1 1.1 REGULATORY L1 HITS 1 "
1.2 *iAXIMJM PERMISSIBLE CONCENTRATIONS 5 1.3 MEASUREMETTS AND APPROXIMATIONS OF 5 TOTAL RADIOACTIVITY _ l.4 LIQUID EFFLUFNT RELEASE DATA 7 1.5 RADIOLOGICAL IMPACT ON MAN DUE TO 9 LIQUID RELEASES 2 GASEOUS EFFLUENTS 19 a 2.1 REGULATORY LIMITS 19 2.2 MEASUREMENTS AND APPROXIMATIONS OF 25 TOTAL RADI0 ACTIVITY 2.3 GASEOUS EFFLUENT RELEASE DATA 30 2.4 RADIOLOGICAL IMPACT DUE TO GASEOUS 31 RELEASES 3 SOLYD WASTE L1 _ 3.1 REGULATORY REQUIREMENTS L7 3.2 SOLID WASTE PATA L7 4 CHANGES TO THE PLANT HATCH ODCM 50 AND PCP APP. A ODCM OIANGE DOCUMEVTATION 51 APP..B REVISED PAGES FROM ODCM 52 O
4 I PLANT E. I . HATCil SEMIANNUAL RADI0 ACTIVE EFFLUENT RELFASE REPORT TABLE LIST OF TABLES PAGE 1-1 TEONICAL SPECIFICATION TABLE 3.14.1-1 3 RADIOACTIVE LIQUID EFFLUENT ADNITORING INSTRU?tENTATION 1-2a LIQUID EFFLUENTS - St#MATION OF 10 ALL RELEASES - UNIT 1 1-2b LIQUID EFFLUEVTS - StBMATION OF 11 ALL RELEASES - UNIT 2 1-2c LIQUID EFFLUENTS - SmMATIONS OF 12 ALL RELEASES - SITE 1-33 13 LIQUID EFFLUENTS - UNIT 1 I-3b LIQUID EFFLUINTS - UNIT 2 0 1-3c 15 LIQUID EFILUENTS - SITE 1-4a 16' INDIVIDUAL DDSES DUE 10 LIQUID RELEASES - UNIT 1 1-4b INDIVIDUAL DOSES DUE TO LIQUID IT RELEASES - UNIT 2 1-5 LOWER LIMITS OF DETECTION - LIQUID 16 ) SAMPLE ANALYSES 2-1 TEONICAL SPECIFICATION TABLE 21 3.14.2-1 RADI0 ACTIVE GASEOUS EFFLUEST MON 11DRING INSTRUMFNTATION 2-2a GASEOUS SFFLUENTS - SU)MATION OF 32 ALL RELEASES UNIT 1 2-2b GASEOUS EFFLUENTS - SU5MATION OF 33 ALL RELEASES - UNIT 2 2-2c GASb0US EFFLUENTS - SGMATION OF 3k ALL RELEASES - SITE
?-3a GASEOUS EFFLUENTS - ELEVATED 35 RELFASES - UNIT 1 0 2-3b GASBOUS EFFLUENTS - ELEVATED RELEASES - UNIT 2 36 2-3c GASB00S EFFLUENTS - ELEVATED 37 RELEASES - SITE 1
i TABLES LIST OF TABLES PAGE 2-4a GASFOUS EFFLUENTS - GROUND- 35 LEVEL RELFASES - UNIT 1 2-4b GASEOUS EFFLUENTS - GROUND - 29 LEVEL RELEASES - UNIT 2 2-4c GASEOUS EFFLUENTS - GROUNT - Lo LEVEL RELEASES - SITE 2-C GASEOUS EFFLUENTS - DOSE L1 RATES - SITE 2-6a AIR DOSES DUE TO NOBLE GASES - L2 UNIT 1 2-6b AIR DOSES DUE TO NOBLE GASES - L3 UNIT 2 2-7a INDIVIDUAL DOSES DUE TO RADICIODINE, LL TRITIUM, AND PARTICULATES IN GASEOUS RELEASES - UNIT 1 2-7b INDIVIDUAL DOSES DUE TO RADI0 IODINE, 'S TRITIUM, ANT PARTICULATES IN GASEOUS (~] RELEASES - UNIT 2 28 LOWER LIMITS OF DETECTION - GASEOUS M SAMPLE ANALYSES 3-la,b SOLID WASTE AND IRRADIATED FUEL L8,L9 SHIPMENTS
_ -_ __ - . _ _ _ _ _ _ _ _ __ __ __.m . _ _ . n V, RADIOACTIVE EFFLUNT RELEASE REPORT 1 LIQUID EFFLocNfS 1.1. 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. 1 TECINICAL SPECIFICATIONS 3.14.1 The radioactive liquid effluent mor.itoring
. instrumentation channels shown in table 3.14.1-1 shall he OPERABLE with their alam/ trip setpoints set to ensure that the limits of Specification 3.15.1 are not exceeded. The alarm / trip setpoints of these 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 tNkESTRICTED AREAS (figure 3.15-11 shall be limited to the
-b'U.
concentrations specified in 10 CFR Part 20, A?pendix B, Table 11.(column 2) for radionuclides other i hn dissolved or entrained noble gases. For dissolved or entrained noble gases the concentration shall be limited ~to 2.0E-04 uCi/ml total activity. 3.15.1.2 The dose or dose commitment to a MINBER OF
'IllE PUBLIC from radioactive materials in liquid effluents 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.
3.15.1.3 ' The: liquid radwaste treatment system, as L described in the ODCM, shall .be used to reduce the radioactive materials in liquid wastes prior to their g discharge when the projected doses due to the liquid i effluent- per Unit from the site (figure 3.15-1) when p d projected over the calendar quarter would exceed 0.18 mrem to the total-body or 0.62 mrea to any organ. l-L -l-L L
_ _ .... - _ . _ . . - _ ~_ _ _ . - ._. _ ._ . . . . ._--_ .
- c. .
3.15.1. 4( a) The coatents within any outside temporary tank shall be limited to less than or equal
- /'~' to 10 curies, excluding tritium and dissolved or \ fentrained noble 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. 9.1. 9_ states in part: "The Radioactive Effluent !
Release Report shall. include (on a quarterly basis) unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents _that were in' excess of 1.Cl, excluding - dissolved and entrained-gases-and tritium for liquid
-effluents, or those 'in excess c' * ~ ~ Cl of noble gases or 0.02 Ci of radiciodines for g6 releases".
( w) t I W W 4* = -+ ae 4 -M -tM.- Tu -w 7-r es e - rPw
f TABLE 1-1 TECHNICAL SPECIFICATION TABLE 3.14.1-1 (SHEET 1 of 2) O RADIOACTIVE LICOID EFFLUE?E MOf!ITORING INSTRUMEtEATION Minimum Channels Instrument OPERARLE ADolicab111ty ACTION
- 1. Gross Radioactivity Monitors Providino Automatic Termina-tion of Release Liquid Radwaste -
Effluent Line 1 ta) 100
- 2. Gross Radioactivlty Monitors not Providing Automatic Termination of Release Service Water System Effluent Line 1 (b) 101
- 3. Flowrate Measure-O ment Devices **
Liquid 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
** Pump curves ray be utilized to estimate flow: in such cases, ACTION statecent 1021s not required. ,
(a) Whenever the radwaste discharge valves are not locked closed. (b) Whenever the service water system pressure is bolcw the closed cooling water system pressure or differential pressure indication is not available.
TABLE 1-1 (CO!TTINUPD) ! TECEICAL SPECIFICATION TABLE 3.14.1-1 (6 Herr 2 of 2 ) (} L.'J PADIOACTIVE LItJJID EFFLUE!7T MCUITOPI!C INS"'PUvE!!TATTON TABtEfDTATIO!j$ ACTICU 100 - With the number of channels OPERABLE less than required by the Mininum Channels OPERABLE requirement, effluent releases may be continued, provided that prior to initiating a release:
- a. At least two indeoendent samples are analyzed in accordance with Specification 4.15.1.1.1.
\ _
- b. At least two technically cua11 fled individuals independently verify the release rate calculations and discharge valving.
Otherwise, suspend release of radioactive effluents via this pathway. If the channel rerains inoperable for over 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 OPEPABLE less than
,m required by the Minimum Channels OPERABLE requirement, I i effluent releases via this pathway riay continue, provided
\- / that once per shift grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a Lower Limit of 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 ef fluent release report.
ACTION 102 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requireinent, effluent releases via this pathway may continue, provided the flowrate is estimated at least or.ce per 4 hours during actual releases. If the channel remains inoperable for over 30 days, an explanation of the circunstances shall be included in the next semi-annual effluent release report. ACTION 103 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, assure that the service water system effluent system monitor is OPERABLE. i o w _L_
/'s 1.2 MAXIMJM PERMISSIBLE CONCENTRATIONS ')
The MPC values used la determining allowable liquid radwaste release rates and concentrations for principal gamma emitters, 1-131, tritium, 3r-89, Sr-00 and Fe-55 are taken f rom 10CFR Part 20, Appendix B, Table II, Column 2. For dissolved or entrained noble gases in liquid radwaste, the MPC is taken f rom Technical Specification 3.15.1.1 (Unit 1) and 3.11.1.1 (Unit
- 2) as 2.0E-04 uCi/ml.
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 detemined in accordance with 10 CFR Part 20, Appendix E, Note 1. The method whereby the MPC fraction is used to determine release rates and liquid radwaste eff1 Lent radiation monitor setpoints is described f3 in Section 1.3 of this report. U 1.3 MEASURDfENTS AND APPROXIMATIONS OF TOTAL RADI0 ACTIVITY 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 rrom 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 frcm each sample taken from tanks which are released. Liquid radwaste sample analyses are performed as follows: Measurement Frecuency Method
- 1. Gamma Isotopic Each Batch Ganuna spectroscopy with computeri:ed data reduction
,, 2. Dissolved or Each Batch Gamma spectroscopy
[} (/ Ent rained with computerized Noble Gases data reduction e m
)
i Measurement Frequency Method O
'O 3. Tritium _ Monthly- Distillation and Composite li @id 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
_6. Fe-55 Quarterly Chemical separation and Composite low energy photon detector. . Gamma isotopic measurements are performed in-house in the radiochemistry lab using gemanium spectrometry. Three gemanium detectors are available: a 20% efficient and two 151 efficient intrinsic gemanium detectors, with 2.0 FMN 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 N 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 il 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 efficiency, 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, St-89, Sr-90, and Fe-55 are used along with the corresponding MPC values to determ.i.ne an MPC fraction for the tank planned for release. This MPC fraction >
!s then used, with appropriate safety factors, along g
with the exp uted dilution stream flow to calculate a ' j~ maximum pennissible release rate ano a liquid effluent monitor setpoint. .The monitor setpoint is calculated t'o assure that the limits of Technical Specifications 3.15.1.1 (Unit 1) or 3.11.1.1 (Unit 2) are not exceeded. 6-
A monitor reading in excess of the calculated setpoint therefore results in an automatic (m termination of the liquid radwaste discharge. () . Liauld effluent discharge is also automatically terminated if the dilution stream flow rate falls below the dllution flow rate used in the setpoint calculations and established as a setpoint on the ! dilution stream flow monitor. 4 Radionuclide concentrations, safety factors, j dilution stream flow rate, and liquid effluent radiation monitor calibration f actor are entered into the computer and a prerelease printout is generated. If the release is not permissible appropriate warnings will be included on the prerelease printout. If the release is permissibleitisapprovedbytheChemistry T.ie pertinent information is Foreman on duty. transferred manually from the prerelease printout i to a one-page release permit which is forwarded to Radwaste Operations. When the release is completed the release permit is returned from Radwaste Operations with ectual release data included. These data arc input to the computer and a post release printout is generated. The
. pest release printout contains actua! release
' rates, actual releasa concentrations and quantities, actual dilution flow, and calculated
's>) doses to an individual.
1.4
- LIQUID EFFLUENi KEL. EASE 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 Unit 2, and Table 1-3c for the site.
The values for-the four categories of Tables 1-2a and 1-lb, and .1-2c are calculated and the Tables . completed as_follows: - '14 - Fission and activation products - Thd total release values (not including tritium, gases, ard alpha) are comprised of the sum of the mearured individual radionuclide activities. This sun.is for'each batch released *.o the river for the respective aThare were no monitors out. of service greater thaa.30 days and no unplanned releases.
.q Q)
.h
quarter. Percent of applicable limit is I determined from a mixed nuclide MPC fraction ! f' - calculation. The average concentration for : C each nuclide over all released batches is divided by the corresponding individual '4PC value. The sum over all nuclides of the C/MPC ratios times 100 is the percent of applicable limit for effluent releases during the quarter.
- 2. Tritium - The measured tritium concentrations in the monthly compuite samples are used to calculate the f etal release and average diluted concentration during each period. Average diluted concentration divided by 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 gemanium spectroscopy on a one liter sample f rom each liquid radwaste batch. The average concentration of dissolved or er. trained 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 uC1/ml) to determine the MPC fraction. The result x100 is the percent of applicable limit for noble gases in liquid effluent releases during the i quarter. Radiolsotopes of iodine in any ' fom are _also determined during the isotopic analysia for each batch; therefore, a separate analysis for possible gcseous forms is not perfomed because it would not provide additional infomation. 4 Gross alpha radioactivity - The measured gross alpha concentrations in the monthly composite samples are used tc calculate the total release of alpha radioactivity.
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8-
_ _ _ . _ . . _ ___ __ _ _ _ __________.__.._..._.__._._.m _ _ . . i Ot'n er data pertinent to batch releases of radioactive liquid effluent from both units are s~ as follows: Number of batch releases: 519 Total time period for batch releases: 59,255 minutes Maximum time period for a batch release: 345.0 minutes Average time period for batch releases: 114.2 minutes-Minimum time period for a batch release: 1.0 minutes Average stream flow during periods of release of 11guld effluent into a flowing stream: 13.561 CFS 15 . RADIOLOGICAL IMPACT ON MAN LUE TO L1 QUID RELEASES Doses to an individual, due to radioactivity-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 0ffsite Dose Calculation Manual. As required by the above Technical Specifications, doses were h
x/ calculated separately for Unit 1 and Unit 2. Results are presented in Table 1-4a for Unit 1 and Table 1-4b for Unit 2. LO i.
TA BLE 1-2a E. I. HATCH NUCLEAR PLANT - UNIT 1 -(' - SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1992 _t LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES Unit Quartsr 1 Quarter 2 Est. Total
--------_-_.------.--_----____..--.-__------.......--__(.).. Error %
A. Fissisn and activation roducta
. Total release Ci 4.16E-02 2.05E-02 4.70E+01
- 2. Average-di:.uted concentratton during p Lod uCi/ml 1.04E-07 2.91E-08
- 3. app cable i t % 5.41E-01 1.61E-01 B. Tritium
- 1. Total release Ci 2.57E+00 3.89E+00 3.70E+01
- 2. Average di:.uted concentrat:.on during Lod uCi/ml 6.44E-06 5.53E-06
- 3. % ap cable 11 t t 2.15E-01 1.84E-01 C. Disso ved and entra ned gases
- 1. Total release Ci 6.96E-04 7.91E-04 1.00E+02
- 2. Average di:.uted concentratnon uring Lod uC1/cl 1.74E-09 1.12E-09
, =3. ap cable li t % 8.71E-04 -5.62E-04 D. Gross Al a ,
radioact ity
- 1. Total release Ci 0.00E+00 0.00E+00 1.20E+02 E. Volume of waste (prior to dilution) liters 1.92E+06 2.89E+06 1.00E+01 F. V lume of d lution water used liters 4.00E+0B 7,04E+08 1.60E+02 O
410-u . - .--. . - _ . _ . _ . u .. __ _. __ . . _ . .-
. ~_ . - . _ -._.-._- -- - .. .. _, _ _ _ . . . . . _ . . .
TABLE 1-2b r~ E. I. HATCH NUCLEAR PLANT - UNIT 2
-( SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1992 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES o . .....................--............-_.---...... ............
Unit Quarter 1 Quarter 2 E;t. Total Error (%) A.- Fission and activaticn products
- 1. Total release Ci 2.09E-01 2.57E-01 4.70E+01
- 2. Average di:.uted concentrathon during pernod uCi/ml 2.72E-07 2.84E-07
- 3. % of applicable limit % 6.15E+00 1.36E+00 B. Tritium 3.70EF01
- 1. Total release Ci 5.12E+00 5.64E+00 2.-Average di:.uted concentratnon during Dernod uCi/ml 6.64E-06 6.21E-06
- 3. % of aphlicable limit % 2.21E-01 2.07E-01 C. Dissolved and entrained gases
- 1. Total release Ci 2.51E-02 8.68E-02 1.00E+02
- 2. Average diluted
.. congentrathon
(' duryng pernod uCi/ml 3.26E-08 9.55E-08
- 3. % og applicable limat % 1.63E-02 4.78E-02 D. Gross Aloha radioactivity
- 1. Total release Ci 0.00E+00 0.00E+00 1.20E+02 E. Volume of waste (prior to dilution) liters 3.31E+06 3.96E+06 1.00E+01 F. Volume of dilution water used liters 7.71E+08 9.08E+08 1.60E+02 t
i O l i l l-v y v v1y my-r~ ' -w - - - * - -
i e TABLE 1-2c E. I. HATCH NUCLEAR PLANT - SITE f r~'h SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1992 Ll LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES _________________________7___________________________________ E __________________________________________________((g{,j}{,, A. Fis ion and act vation pro ucts
- 1. Total release Ci 2.51E-01 2.78E-01 4.70E+01
- 2. Averege di:.uted concentratnon during per;od uCi/ml 2.14E-07 1.7?E-07 '
- 3. % of applicable limit % 4.23E+00 8.37E-01 B. Tritium
- 1. Total release Ci 7.69E+00 9.53E+00 3. 7 0E4 01
- 2. Average di:.uted concentrat: on during cer;.od uCi/ml 6.57E-06 5.91E-06
- 3. t of ap licable limit 4 2.19E-01 1.97E-01 C. Dissolved and entrained gases
- 1. Total release ci 2.58E-02 8.75E-02 1.00E+02
- 2. Average di:.uted
("T (_,/ concentrat:.on during ner:od uCi/ml 2.21E-08 5.43E-08
- 3. % of apblicable limit % 1.10E-02 2.72E-02 D. GrossAlgha i o$ aire ase Ci 0.00E+00 0.00E+00 1.20E+02 E. Volume of waste (prior to dilution) liters 5.22Ec06 6.85E+06 1.00E+01 F. Volume of dilution water used liters 1.17E+09 1.61E+09 1.60E+02
__ , _ _ ~ _ _
. . . _ . _ n ., . ,
TABLE 1-3a* E. I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1992 (' LIQUID EFFLUENTS Continuous Mode ** Batch Mode Nuclides Reltased Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 H-3 Ci 2.57E+00 3.89E+00 Fission and activation products: Na-24 C: 4.74E-05 5.78E-05 Cr-51 C: 1.89E-04 4.65E-04 Mn-54 C: 8.63E-04 6.16E-04 Fe-55 C: 1.49E-03 0.00E+00 Fe-59 C: 1.30E-05 0.00E+00 Co-58 C:, 2.82E-04 7.33E-05 Co-60 C: 6.81E-03 6.56E-03 Zn-66 C; 6.51E-03 5.06E-03 As-76 C: 2.74E-03 9.40E-04 Sr-89 C;. 0.00E+00 3.17E-04 Sr-92 C: 0.00E+00 7.24E-07 Y-91m C: 1.76E-03 1.96E-04 Nb-95 C:, 5.10E-06 3.20E-07 Hb-97 C: 8.91E-04 1.41E-04 Tc-99m C: 8.76E-05 4.56E-05 Sb-125 C: 2.85E-05 4.27E-05 I-131 C;, 1.40E-04 8.17E-05 I-133 C: 1.43E-04 9.91E-05 Cs-134 C;. 2.15E-03 6.40E-04 Cs-137 C;, 1.71E-02 5.06E-03 Cs-138 C:, 4.26E-05 1.26E-05 Ba-140 C: 8.55E-05 7.12E-06 ( La-140 C: 1.56E-04 5.17E-05 Co-144 C:, 2.22E-05 0.00E+00 Np-239 C: 6.41E-05 3.47E-05 Zn-69m C: 1.17E-05 1.55E-06 Total C: 4.16E-02 2.05E-02 Dissolved and entrained games: Xe-131m C: 3.57E-04 0.00E+00 Xe-133 C; 3.03E-05 1.93E-04 Xe-135 C: 3.07E-04 5.98E-04 Ar-41 C: 1.95E-06 0.00E+00 Total CL 6.96E-04 7.91E-04 Gr-klpha Ci 0.00E+00 0.00E+00
- Zeroes i this table indicate that no radioactivity was for typical
{ gresentaovedetectab{elevels,ower 1 m ts of detect on f or liqugee Tab {e 1-5d samp e analyses
**There are no continuous mode radioactive liquid release pathwais at Plant Hatch.
(O _)
TABLE 1-3b* E. I. HATCH NUCLEAR-PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE RrPORT 1992
. LIQUID EFFLUENTS Continuous Mode **
Batch Mode Nuclides < Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 H-3 Ci 5.12E+00 5.64E+00 Fission and activation products: Na-24 C:, 2.23E-02 1.28E-02 Cr-51 C: 1.60E-02 1.19E-03 Mn-54 C: 1.48E-03 1.79E-04 Mn-56 C: 8.28E-05 4.18E-05 Co-58 C: 1.40E-03 2.80E-04 Co-60 C: 4.05E-03 1.45E-03 Zn-65 C: 1.03E-02 6.54E-03 As-76 C: 9.70E-03 7.17E-04 Sr-89 C: 3.64E-04 9.11E-04 Sr-90 C: 1.89E-05 0.00E+00 Sr-91 C: 1.16E-03 1.66E-03 Sr-92 CL 2.16E-04 4.70E-04 Y-91m C: 1.03E-v3 4.09E-05 Zr-95 C; O.00E+00 4.92E-06 Nb-95 CL 0.00E+00 7.22E-06 Nb-97 CL 4.96E-03 1.57E-05 Mo-99 C: 2.40E-03 1.92E-04 Tc-99m C: 1.65E-02 3.94E-03 Ru-103 CL 0.00E+00 2.92E-06 Sb-125 C: 0.00E+00 3.71E-06 Te-131 C: 0.00E+00 2.17E-05 I-131 C: 1.02E-02 9.93E-04 I-132 C: 9.96E-04 5.24E-04 1/6^) I-133 C: 8.94L-03 4.58E-03 (_/ I-134 C: 2.72E-05 5.60E-05 I-135 C: 5.12E-03 3.89E-03 Cs-134 C: 2.49E-03 1.82E-03 Cs-136 C: 1.70E-04 1.93E-04 Cs-137 CL 3.89E-03 2.37E-03 Cs-138 C1 0.00EA00- 1.32E-05 Ba-139 Ci 0.00E+00 1.15E-04 Ba-140 Ca 3.40E-04 6.04E-04 La-140 C: 2.24E-04 S.10E-04 Ce-141 C: 0.00E+00 3.01E-06 N -239 Ch 8.47E 2.11E-01 Z -69m C: 4.41E-04 5.25E-05 S -124 C: 0.00E+00 2.99E-05 Total C: 2.09E-01 2.57E-01 Dissolved and entrained gases: Xe-131m C: 3.06E-04 5.36E-04 Xe-133m C:, 1.33E-04 1.28E-04 Xe-133 C:, 4.90E-03 1.08E-02 Xe-135m C: 9.68E-03 7.51E-03 Xe-135 C: 1.01E-02 6.77E-02 Ar-41 C: 0.00E+00 6.37E-06 Total. C: 2.51E-02 8.68E-02 Gr-Alpha Ci 0.00E+00 0.00E+00
- Zeroes in this table indicate that no radioactivity was poverli 's resentaovedetectab{e1ertsofdetectonfo.liqugdsampleanalysesee Table 1-5 for typical g7 V ** here are no continuous rode radioactive liquid re ease pathways at Plant hatch
_1L_
TABLE 1-3c* l E. I. HATCH NUCLEAR PLANT - SITE I SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1992 LIQUID EFFLUENTS Continuous Mode ** Batch Mode Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 >
.......__________.._________________..___.._______..__..._.4 H-3 Ci 7.69E+00 9.53E+00
-Fission and activation products:
Na-24 C; 2.24E-02 1.29E-02 Cr-51 C: 1.61E-02 1.66E-03 Mn-54 C: 2.35E-03 7.94E-04 Mn-56 C: 8.28E-05 4.18E-05 Fe-55 C: 1.49E-03 0.00E+00 Fe-59 C: 1.30E-05 0.00E+00 Co-58 C: 1.68E-03 3.54E-04 Co-60 C: 1.09E-02 8.01E-03 Zn-65 C; 1.68E-02 1.16E-02 As-76 C:. 1.24E-02 1.66E-03 Sr-89 C: 3.64E-04 1.23E-03 Sr-9u C: 1.89E-05 0.00E+00 Sr-91 C: 1.16E-03 1.66E-03 Sr-92 C: 2.16E-04 4.71E-04 Y-91m 2: 2.79E-03 2.37E-04 Zr-95 C 0.00E+00 4.92E-06 Nb-95 C; 5.10E-06 7.54E-06 Nb-97 C: 5.85E-03 1.56E-04 Mo-99 C: 2.40E-03 1.92E-04 Tc-99m C:. 1.66E-02 3.99E-03 Ru-103 C; 0.00E+00 2.92E-06 Sb-125 C: 2.85E-05 4.64E-05 Te-131 C: 0.00E+00 2.17E-05
./ I-131 C: 1.03E-02 1.07E-03 I-132 C: 9.96E-04 5.24E-04 I-133 C: 9.09E-03 4.68E-03 I-134 C 2.72E-05 5.60E-05 I-135 C 5.12E-03 3.89E-03 Cs-134 C: 4.64E-03 2.46E-03 Cs-136 C: 1.70E-04 1.93E-04 Cs-137 C: 2.10E-02 7.43E-03 Cs-138 C:. 4.26E-05 2.58E-05 Ba-139 C4 0.00E+00 1.15E-04 Ba-140 C: 4.25E-04 6.11E-04 La-140 C: 3.80E-04 5.62E-04 Ce-141 C: 0.00E+00 3.01E-06 Ce-144 C: 2.22E-05 0.00E+00 Np-239 C: 8.48E-02 2.11E-01 Zn-69m C: 4.53E-04 5.40E-05 Sb-124 C: 0.00E+00 2.99E-05 Total C: 2.51E-01 2.78E-01 Dissolved and entrained gases:
Xe-131m C: 6.63E-04 5.36E-04 Xe-133m C: 1.33E-04 1.28E-04
- Xe-133 C; 4.93E-03 1.10E-02 Xe-135m-- C 9.68E-03 7.51E-03
.Xe-135 C: 1.04E-02 6.83E-02 Ar-41 C: 1.95E-06 6.37E-06 Total C;, 2.b8E-02 8.75E-C2 Gr-Alpha Ci 0.00E+00 0.00E+00
- Zeroes in this table indicate that no radioactivity wag
- s' resent above detectab e levels See Table 1-5 for t Sower limits of detect on for liquid sample analyses.ypical
**There are no continuous mode radioactive liquid release pathways at Plant Hatch.
r ( TABLE 1-4a
.x E. I. HATCH NUCLEAR PLANT - Unit 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1992 INDIVIDUAL DOSES DUE TO LIQUID RELEASES i
Cumulative Doses per Quarter Organ Tech Unit Quarter 1 % of Quarter 2 . of Spec Tech .ech Limit Spec ace" Limit Limit Bone 5.0 mrem 7.46E-02 1.49E+00 3.79E-02 7.57E-01 Liver 5.0 Drem 1.10E-01 2.20E*00 5.75E-02 1.15E+00 TEody 1.5 mrem .7.47E-02 4.98E+00 3.93E-02 2.62E+00 Thyroid .5.0 7arem 9,61E-04 1.92E-02 1.37E-03 2.74E-02 Ki8ney 5.0 mrum 3 .,7 9 E- 0 2 7.58E-01 2.00E-02 4.00E-01
. Lung 5.0 mrem 1.21E-02 2.42E-01 6.23E-03 1.25E-01 GILLI 5.0 mrom 7.67E 03 1 53E 01 7.74E 03 1.55E 01 Cumulacive Doses This Year Organ Tech Unit Quarters % of Spec 1& 2 Tech Limit Speg Limit (y ....____
Bone 10.0 mrem 1.12E-01 1.12E+00
\/ Liver 10.0 mrem 1.67E-01 1.67E+00 TBody 3.0 mrem 1.14E-01 3.80E+00 Thyroid '10.0 mrem 2.33E-03 2.33E-02 Kianey 10.0 mrem 5.79E-02 f,.79E-01 Lung 10.0 mrom 1.83E-02 1.83E-01
-GILLI 10.0 mrem 1.54E ____..___________..________..______
1.54E 01 ___ ._________ ..__________-02 O
-16
-~,,...M
n - . . - . _ _ . ,_ _ .. I i TABLE 1-4b i E. I. HATCH NUCLEAR PLANT - Unit 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1992 INDIVIDUAL DOSES DUE TO LIQUID RELEASES Cumulative Doses per Quarter
% of Organ Tech Unit Qrarter 1 % of Quarter 2 Speg Tech Tech Limit Spec Spe Limit Limit
__ ___...__..._--_...___ ..____--__._____--________.______._- 1.67E-02 3.35E-01 Bone 5.0 mrem 2.69E-02 5.38E-01 Liver 5.0 mrem 4.82E-02 9.64E-01 3.06E-02 6.12E-01 TBody 1.5 mrem 3.41E-02 2.27E+00 2.19E-02 1.46E+00 - Thyroid 5.0 mrom 8.50E-02 1.70E+00 9.79E-C3 1.96E-01 Kianey 5.0 mrem 1.81E-02 3.62E-01 1.12E-02 2.24E-01 Lung 5.0 mrem 4.87E-03 9.74E-02 3.14E-03 6.2fE-02 mrem 1.74E-01 6.93E-03 1.394 ___
.........________ .....___-01 GILLI 5.0 8.71E-03 Cumulative Doses This Year
.....___________..._..___~____________ _____......_______ ...
Organ Tech Unit Quarters % of Spec 1 &2 Tech Llnit Spec Limit Bone 10.0 mrem 4.36E-02 4.36E-01 Liver 10.0 mrem 7.80E-02 7.88E-01 TBody 3.0 mrom 5.60E-02 1.87E+00 9 Thyroid 10.0 mrem 9.48E-02 9.48E-01 Kidney 10.0 mrom 2.93E-02 2.93E-01 Lung 10.0 mrem 8.01E-03 8.01E-02 GILLI 10.0 mrem 1.56E 02 1.56E-01 l
TABLE 1-5 LOh'ER LIMITS OF DETECTION - LIQUID SVtPLE ANALYSES l rh (v) The "alues in this table represent apriori lower limits ) of detection (LLD) khich are typically achieved in laboratory analyws of liquid radvaste samples. l RADIONUCLIDE LLD UNITS Mn-54 T.'I8E-08 iiCICiiil Fe-59 7.78E-08 Co-53 4.67E-08 Co.60 4.78E-08
- n-65 1.31E-07 Mo-99 5.10E-07*
Cs-134 7.18E-08 Cs-It? 6.05E-08 Ce-141 1.41E-07 Ce-144 6.30E-07* I-131 6.5iE-08 Xe-135 8.45E-08 Fe-55 2,00E-05 St-89 5.00E-08 Sr-90 5.00E-08 H-3 1.00E-05 G *In accordance with Technical Specification Tables 4.15.1-1 (Unit 1) and (, / ' 4.11.1-1 (Unit 2), Table Notation b, the pemissible Lower Limit of Detection may be increased inversely proportional to the magnitude of the gamma yield. However, the LLD detemined 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). Yf l x,./ t e _18 l
2 : GASEOUS EFFLUENTS 2.1- REGULATORY LIMITS-O The Technical Specifications presented in this section are for Unit 1. Requirs :.ents for Unit 2 are the same as for Unit 1; however, the Technical Specification numbe s are not the same. TECHNICAL SPECIFICATIONS 1 3.lt.. 7.m radioactive gaseous ef fluent monitoring inunr- .ation channels show.. In table 3.14.2-1 shall be APL . Ab with their alam/ trip setpoints set to ensure that the limits of Specification 3.15.2.1 (a) are not exceeded. The alarm / trip setpoints of these channels shall be detemined in accordance with the ODCM. (Technical Specification Table-3.14.2-1 is included in this section as Table 2-l'. 3.15.2.1 The dose rate at any time in the UNRESTRICTED AREAS (figure 3.15-1) due to radioactive materials released in gaseous effluents 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
- n. and less than or equal to 3000 mrem / year to the skin,
- b. The dose rate limit for I-131, I-133, tritium, and for all radioactive materials in particulate fom and radionuclides-other than noble gases with half 'ives greater than 8 days-shall be less thar. or equal to 1500 mrem / year to any organ.
3.15.2.2 The air se in' UNRESTRICTED AREAS (figure 3.15-1) due to noble gases released in gaseous effluents from each reactor unit shall be limited to the following:
- 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.
- 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.
3.15.2.3 The dose to any organ of a MEMBER OF THE PUbLIC from I-131, I-133, tritium , and all radionuclides in particulate fom with half-lives greater than 8 days in gaseous-effluents released to UNRESTRICTED AREAS (figure i - 3.15-1) from each reactor unit'shall be limited to the ! O- following:. l: V 5
-19_
i
i l
- a. During any calendar quarter to less than or
, ~'T equal to 7.5 mrem to any organ, i
~'
- b. During any calendar year to less than or equal to 15 mrem to any organ.
3.15.2.1 The GASEOUS RADWASTE TREATMENT SYSTEM as described in the ODCM shall be in operation. (This T-chnical Specification applies whenever the main condenser air ejector system is in operation.) 4.15.2.4 GASEOUS RADWASTE TREANENT 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 bO BER 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 liniited to less than or equal to 75 mrem.
(With the calculated doses from the release of radioactive materials in liquid or gaseous
,_s effluents exceeding twice the limits of I ') 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 contributions from the reactor units and from outside storage tanks to determine whether the above 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 condenser offgas treatment system shall be limited to less than or equal to 4 percent by volume. 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 rystem pretreatment monitor station shall be limited to less than or equal to 240,000 uti/second. 6.9.1.9 states in part:
"The Radioactive Effluent Release Report shall include (on a quarterly basis) unplanned relaases from the site to unrestricted areas of radioactive materials in gaseous and liquid
<w effluents that were in excess of 1 C1, excluding
(') ~ dissolved and entrained gases and tritium for liquid effluents, or those in excess of 150 Ci of noble gases or 0.02 Ci of radioiodines for gaseous releases." 1
hmuummummmme e- e e l Alit [ 3.14.?-1 ( Siff f f I of 4) H A910Act iV E CASI OUS E F f i bt til _ MONI 10f0 NGO NS_IILUMEN T AT t 01 H it e i ntem Crianne I s Paramej e_r f,ql10ff OPtHABLE Apgjj cab ( l i e y g g gem I
- 1. H.ein Condenser Offgas treatment ;
Systee Explosive Cas Monitoring Systeen 106 i , ** ', itydrogen l flydrogen Honitor (l)
- 2. Reactor fini nding vent Stack Hooltoring Syst is
- Hadioactivity Rate 10S
- a. f6oble Ca s Act ivi ty Monitor (!) He s su reaten t + lQ f *1
- Verif'y Frese;nce of IG?
8 b. lodissa Sampler Cartridge {I) Cartridge QJ r< > H
- Vertry Presence or 107 ^
i Pa rt ectst ato Sample r f il ter gIJ
- c. Filter h is
- d. Errinent System f lowrate
- Systems flowrote 1 086 %
Han sas t omen t Device (1) Me n s te re me n t. Y ( o fois % e, Sampler i t owrote Heasurement
- Sampler f lowrate Devico (3) Measurement y v
- 3. Itocombiner Dui tding Ventilation Monitoring System
- Radioactivity Rate 105
- a. Nob f u Cas Act ivi ty Honitor (1) Heasurement +
- Verity Presence or for
- b. todsne Sampler cartridge it) Cartridge
- Verify Presence or 101
- c. Particulate Sampler Iitter (1) filter Napier F lowrnte 10's Samp l e r f l ow ra t e Mea su v emen t (4) Measurement
- d. .
De v i e:e 1 . L - . _ . . _ . _ _ , N
m .. -, -- - , _ Wi, .
- Q '
'f,
- . e TAets:-3.14.2-1T(SinEETl2 OF 4)
- RADIOACf IVE CA'J,018s JJJ8 'JENT MOef f TORING " f NSTRt#tf MI AT f ON Hinbaum ChanneIs' ACitON
. OPERA 81(: . 90licability Parameter ins t rtament
.4. Main Stack Monitoring System
* * . Radioactivity Rate! 105'
- a. ' Noble Gas Activity Monitor -
(1)L Measursement + Verify Presence or - -107
- b. lodine Sampf er CartrlIfge i '(1)
- Cartridge. .
- Verify Presence or 107 ' ,1 -
- c. Particulate Sampler N iter (1) Filter ~g;
.e
' tn
- d. Cf riuent System flowrate
- System Ficrsrate 104 ~ y Measuring Devices (1) Measurement e
.ga I * ' sampler flowrote lots -
y e. Spapier F I'owrate Hanssarihg (1) Measurement - ,tn i Device Condenser Orrgas Protreatment- # S. Monitor ro
*** dadioactivity Rate 108 O Nuts t e Ga s Ac t i v i ty Nois i t o r -' ( l') Measurement %
e s i
- " ' - i .- , , . _ _ ,
tat ;_ (Cheet 3 or 1.) pf Qf
*ABLE
. 3,14.2-1 (SHEET 3 0F 4)
RADICAGLIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Table Notations
+ Monitor must be capable of responding to a Lower Limit of Detection of 1 x 10-' pCi/ml .
*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 OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided the flowrate is estimated at least once per 4 hours.
If the number of channels OPERABLE remains less .han required by the Minimum Channels OPERABLE requirement for , ("'} over_30 days, an explanation of the circumstances shall be x/ included in the next semi-annual effluent release repor*.. ACTION-105 - With the number of channels OPERABLE less than
. required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided grab samples are taken daily and analyzed daily for gross activity within 24 hours. . With the number of main stack monitoring system channels 0PERABLE less.than required by the_ Minimum Channals OPERABLE requirement, without delay suspend drywell purge.
If the number of channels OPERABLE remains less than required by the Minimum Channels OPERABLE requirement for over 30 days, an erple: Ttion of the circumstances shall-be included in the next .;.ni-annual effluent release report. ACTION 106 - With the number of channels OPERABLE less than required by the Miqimum Chanrels OPERABLE requirement, operation of the main condenser offgas treatment system may 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 offgas system line downs,treaa of the retomoiner, ('~) N-- hydrogen concentration readings are taken and logged every 4 hours.
. ~ ,m 2A2LE 2-1(Sheet h or L) '
- (v)-
TABLE 3.14.2-1 (SHEET 4 0F 4) RACTUACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Table Notations (Continued) If the number of channels OPEI 2BLE 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 recort, ACTION 107 - With the number of channels OPERABLE less than 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 after 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.
( ,) ACTION 108 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, release lto 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 stack monitor (011-K600) is OPERABLE. ' Otherytise, 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 effluent release recort.
.k
-2L-
? 2.~ 2 ifEA,91RFMNT dND APPROXIMATIONS OF -TOTAL RGI0 ACTIVITY Waste gas release at Plant Hatch'is confined to
[V'Y _four paths: main stack (also called the offgas vent), Unit 1 reactor -building vent;- Uni t 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-t>Te sample collection device for collecting particulates and lodines._ 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, p-if ^~
- 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
; f rom -the particulate filters collected over the previous. quarter and the quarterly.
composite sample is analyzed for Sr-89 and Sr-00. . Sample analyses results and release flow rates from the four release points form the basis for calculating released'quantitles of-radionuclide-specific radioactivity, dose rates associated witi gaseous releases, :and cumulative ~ doses for: the current quarter and year. This task is normally performed with computer assistance.
-25
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 pointt, to assure that- the limits of Technical Spcifications 3.15.2.1.a (Unit 1) or 3.11.2.1.a (Unit 2) are not exceeded. Calculation of snonitor setpoints is described in the Plant llatch ODOt. With each release period released radioactivity, dose rates, and cumulative doses are calculated. Cumulative dose results are tabulr.ted along with percent of Technical Specification limits ( 3.15.2.2 and 3.15. 2.3 (Unit 1): 3.11.2.2 and _ 3.11.2 3 (Unit 2) for each release, for the current quarter and year. After each calendar quarter (13 weeks) a sunnary of waste gas releases from the four vents is compiled for preparation of the Semiannual Effluent Release Report required by Technical Specifications 6.9.1.8 and 6.9.1.9 and described in NRC Regulatory Guide 1.21. The methods for determining released quantities of radioactivity, dose rates and cumulative doses are as O 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.
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 (bait 2) for noble gases; and 3.15.2.1.b (Unit 1) and 3.11.2.1.b (Unit 21 for radiolodine, tritium, and particulates. Dose rate calculation methodology is presented in the Plant Hatch 0D01. O
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 gaseous releases. Air doses are calculated for each release period and cumulative totals are kept _for each unit for the current c&ndar quarter and year. Cumulative air doses are compared to the dose limits specified in Technical Specifications 3.15.2.2 (Unit 1) and 3.11.2.2 (Unit 2). Current percent of technical specificatien limits are shown on the printout for each release period. Air dose calculation methodology is presented in the Plant Hatch ODCM.
- 2. RADIGIODINE, TRITIUM, AND PARTICULATE RELEASTfi Released quantities of radiciodines are determined from the weekly samples and release
- flow rates for the four release points.
Radiciodine concentrations are determined by gamma spectroscopy. Released quantities of particulates are determined from the weekly (filter) saples and release flow rates for the four release points. Gr i spectroscopy is used to quantify c mtrations of principal gamma emitters. 3r each calendar quarter the particulate
- i. tters f rom each vent are combined, fused, and struntium separation is performed. Since sample flows and vent flows are almost constant over each quarterly period the filters 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. Tritium samples are obtained monthly f rom each vent by aassing the sample stream through a cold t rap. Tae grams of water vapor / cubic foot gas is measured upstream of the cold trap in order to ~ alleviate the difficulties in determining water vapor collection efficiencies. The tritium samples are analyzed by an independent laboratory and results are furnished 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 O be used in release, dose rate, and individual dose calculations.
-27_
--_____-___-________________________________________________________________-__________-_-________-___-__-a
m
'l
# Dose rates due to radiolodine, tritium, and n particulates are calculated for a hypothetical 1" )- child, exposed to the inhalation pathway, at the
- _ - location in the unrestricted area where the potential dose rate is expected to be the highest. - Dose rates are calculated for each rs). case point, for each release period. and the total dose rate ' f rom all four release points are compared to the dose rate limits specified in Technical Specifications 3.15.2.1.b (Unit 1) or 3.11.2.1.b (Uni t 2) .
Individual doses due to radioiodine, tritium, and particulates 'are calculated for the critical receptor,- which is described in the Plant Hatch ODCM. ' Individual- doses are calculated for each release- period and ctenulative totals are kept for each _ unit- for the current calendar quarter and year. _ Cumulative individ"al doses are compared to the dose limits specified in Technical 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.
'3 GROSS ALPHA RELEASE O.- -
-V 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 sheet and the activity is summed at the end of the conth. The summed activity. is then divided by the total monthly volume-to determine the concentration.~ This concentration Is- input ta the composite file of the computer and is used for release calculations.
4 ERROR ESTIMATES' Regulatory Guide 1.21 requires that estimated total error in analysis techniques be reported. These estimates are required for the total fission and activation ges release, total I-131.
. release, total particulates with half-lives greater than 8-day release, and total tritium release.
-J .. , "The. total or maximum error associated with the effluent measurement will include the cumulative errors resulting from he total operation of J /; sampling and measuremen+.. Because it may be very (f difficult to assign error tenns for each pa ameter affecting the final measurement,
- w. ailed statistical evaluation of- error are not su.wested. The objective should be to obtain an overall estimate of the error associated with measurements'of radioactive materials released in
--liquid and gaseous effluents and solid waste."
Estimated errors are based on errors in counting equipment calibration, counting statistics, vent flow ratas, vent sample flow rates, non-steady release rates chemical yield factors, and sample losses for such items as charcoal cartridges. (11 Fission and Activation Total Release was calculated from sample analysis results and release point flow rates. Statistical Error 60% Counting Equipment Calibration lot Vent Flow Retes 10% Non-Steady' Release Rates -20% 15DT p, (2) .I-131 Release was calculated from each weekly ( j sample: Statistical Error 60% Counting Equipment Calibration 101 Vent Flow Rates 10% Vent Sample Flow Rates 10% Non-Steady Release Rates 10% Losses From Charcoal Cartridge 10% ITOT (3) Particulates with half-lives greater than 8 days release- was calculated f rom sample analysis results and release point flow rates. Statistica1' Error at LLD -concentration 60% Counting Equipment Calibration lot,
-Vent Flow Rates 10%
Vent Sample-Flow Rates 10% Non-Steady Release Rates 101 1UDT rm w
-w.
[ -(4) Total Tritium Release .eas dominated by the . reactor building vent' tritium release; hence, the m larger statistical errors'of the off-gas vent and ' V recombiner building ventf tritium releases do not affect the error in the total tritium release: Water Vapor in Sample Stream Determination 201-VE6t Flow Rates ' . 10%- Counting Calibration and Statistics- 10% Non-Steady Release 501, VOT s 2.3
- GASEOUS EFFLIIEVT RELEASE DATA Regulatory Guide 1.21 Tables '1 A, IB, and 1C 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 1, Table .3-3b for Unit 2 and Table 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.
To complete Tables 2-2a-c, total release for each of the four categories -(fission and activation gases; iodines; .particulates;; and tritium) was divided by the
. number of seconds in the quarter to obtain a release 3 rate in uCi/second for_each category.
d--- However,: the _ applicable Technical Specification limits are not in tems-of release rate in uCi/second but in terms of dose : rate in mrem / year, as presented in Technical Specifications 3.15.2.1 (Unit Il and l3.11.2.1,(Unit 2). Noble gases are limited ,as specified in 3.15.?.1.a and 3.11.2.1.a.. The other
~
three categories -(tritium, radiciodines, 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 ana
. 3.11.2.1 are -site . limits, not unit limits. Dose rates
.due to noble gas: releases and.due to.radiolodine, tritium and particulates are presented-in Table 2-5 along with percent'of technical specification limits.
Gross alpha radioactivity is reported. in Tables 2-2a, 2-2b, and 2-2c.as curies released in each quarter.
; Limi_ts for cumulative beta and gamma air doses, due to noble gases, are specified in Technical = Specifications 3.15. 2.2 (Unit 1) .- and . 3.11.'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.
*There were no monitors -out of service greater than 30 days and no unplanned releases.
t
, ,- - r e
n- ? T '[_\ (
^ ._
'~
i* , Limits.for' cum'lative u individual doses,-due to
* .' radiciodine, tritium, Land particulates, are specified M: 'in Technical: Specifications 3.15.2.3 (Unit 1) and-if -3.11.2.3 (Unit 2). These limits are also unit
~
111mits'. ' Cumulative-individual doses are presented in i
; Tables ~ 2-7a and 7b, _ with percent of ' technical
- specificatlon-limits.
12.4 - RADIOLOGICAL i lMPACT DUE TO GASEOUS RELEASES Dose' rates-due to noble gas releaces were calculated for'the' site-in accordance.with Technical
- Specifications' 3/1.15.2.1.a (Unit 1) and 3/4.11.2.1.a (UnitL2). Results' are presented in Table ~ 2-5. Dose-rates due -to radiciodine, tritium,' and particulates in gaseous releases were calculated in accordance with
~
Technical 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.
'Cumulatiie air doses due-to. noble gas releases were
- calculated. for' each unit' in accordance with Technical
' Specification 3/4.15.2.2 '(Unit 1) and 3/4.11.2.2 (Unit 2). These results:are presented in Table 2-6a -for Unit 1 and Table 2-6b' for -Unit 2. .
TCumulative doses to an indiv3 dual due-to radioiodine, tritium, , and particulates were calculated for each N(xJ % : 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 1 and L Table .2-7b for Unit 2.
- Dose rates' and doses were ' calculated using the
-methodology presentedzin the Plant Hatch Offsite Dose LCalculation Manual, p
l-Ib . 4 l
TABLE}~2a SEMIANNUAL ftADIO C E F LEKS PORT 1992 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES Unit Quarter 1 Quarter 2 Est. To al
. . . . . . . . . . . . . . .. . . . . . . . . . . . .. . . . . . .. . . . . . . . . . . . . .. ' EEe E. b.
o!$ hi
$$"'I$tal Ci 5.35E+01 9.43E+01 1.00E+02
- 2. !' UCi/sec C.39E+00 1.20Et01 Re$eaka a Rat For
*3. f SpecL$kh it n
ci 6.53E-03 2.17I-03 1.10E+02
- 2. g uC1/sec 7.79E-04 2.76E-04 L:.otrat
$$$$$tes ci 6.27E-04 1.34E-03
/
' w$thhalf- 1.00E+02 gives>8
- 2. Sage uC1/sec 7 48E-05 1.71E-04 R li ?r
*3.k"of ch %
4 bS$s-S Radfonct$vkty C1 2.50E-06 2.72E-07 - D. Tritium 1: Reisak Lnl . ucfhec !:1!!!8i ?:lil!8i '- "+ d I!*rfoP" peh $$ t
*TachnicalSgecification fgggm/g gdesa (mren) limits are in .orms 2-5, of dose
. See Tables 2-6a, 2-6gate v
O
-.-_-- _ _ _ _ - _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~
' TABLE 2-2b O. SauxxNot%de?%"EmeMTEidsihr usi GASEOUS EFFLUENTS - SUMMATION O' ALL RELEASES Unit Quarter 1 Quarter 2 Rat. Total
..........................._*EEo!.n L.
A. Fis ion and
,ct$vation al Ci 6.27E+01
$**I$g$a*
- 2. Ne uCi/sec 7.48E+00 2.02E+02 1.00E+02
- 2. 571 +01 hkk $r
*3. k"o! !
specL$$kt % f,$ o ci !.46E-03 2.64E-03 1.10E+02
- 2. gg uCi/sec 6.52E-04 3.36E-04 '
$h$ob pac-[$$t f*"j!!g"lej[h..Ci 2.07E.O. 4. 1E-04 1.00m+02 O give.>e
- 2. e uCi/ cec 3.66E-05 6.24E-05 Ratg or
*3. h*of Tecg 4
- 4. h Ra$sak$ha1oactSvity C1 2.99E-06
- 2.96E-07 .
oN1 v.83 00 6.23E+0 9.00E+01 2.kvarafge uCffsec 17h++00 7.92E-0 Rk Fr
*3. k*o hoch Spec Linit 4
- Tech g m7 g 1 cal g gSgecification ota (mram). limits Seeare in terms Tables of dose 2-5. 2-6a , 2-6gate O
-33_
i I
(~ s TABLE 2-2c ('"j E. I. HATCH NUC 4 M PL7 - SITE dEMIANNUAL RADIOACTIVE EFFwENT RE' l dASEOUS EFFLUENTu - SUKHATION 0, ' AG REuEASESASE REPORT 1992 Unit Qvarter 1 Quarter 2 Est. To
. . . . . . . . . . . . . . . . .. . . . . . . . . . . . . .. . . . . . . .. . . . . . . ... . .E E1 A. Fission and Activation l
$"*h$tal C1 1.16E+02 2.97E+02 1.00E+02 eltase 2.- verage uCi/sec 1.392+01 Releaie 3.'77E+01 -
Rat For Per od
*3. 4o Tech 4 l spec Limit i I
B. Iodines
- 1. "otal s.1 1.20E-0) 4.81E-03 odine-131 1.10E+02
- 2. ggagg uC1/ soc 1.43E-03 6.12E-04 Rat For '
er od
*3. o T 4 pec LIa:h mit C.
r't 1.rticulatps Dn Particu2ates Ci 9.34E-04 1.84E-03
' ') v:.th half- 1.00E+02 p,ves>8
- 2. gfage uC1/sec 1.11E-04 2.33E-04 Mat Fcr Per od
*3. 4o T Spac L! t %
- 4. Grops A ha C1 S.49E-06 S.68E-C7 Radioact vity D. Tritium
- 1. Total Ci 1.25Et01 1.20E+01 9.00E+01
- 2. Average uC1/seo 1.SOE+00 1.52L+00 Release Rat For Per od
*3. 4o Tech %
Spec Limit
- Technical S
{nrem/ggg angacification doae (nrem) .limits See are in terms Ta las of dose 2-5, 2-6a, rate 2-6b,
/~T
\_/
-34
, _. . _ _ . - ~-- _ _ _ . ~_ . . . . . _ . . .. _ . - - _ - - ._
qp n
'y. ;*
- w s
^
TABLE O-3a-
' ~
4EMIANNUAL l'c E GASEQUSEFFLUhT8:- ATED;tE53 PhRT1992 LEASES *- Co t - 84toh Mode **
.... .. ........ . .n...in..uous.... M. ode ........
uclides .._ ...
-..eleased.... ... _.. ... Quarter 1- Qu rte Unit .............a....r.... 2 Quarte . r.1. -............Quarte r 2 1.-Fi.ssion."G.ases.. . . ... ..
Kr-85m. CL 5. 9 E- 01'. 5. 3E-01
-Kr-47 . C;, 1. 0E+00 4. 4E+0p e- .C;, '
.. OR+ 9EN a 2
e-l.33
. ,352 ' C: . : . 08E+00 01' 1 . OF% t e' C. ,
p .85E+00 4.061450, e-!.35
,38 C:. a.2S
..........E+01'
.. ..... . ..... .5.38E+01-Tot l' !"or- . .... ....... . .. . ..._...
Per .. .od Ci: 3.9 8.
.. . . .. .....8E+01-
.. . ... 2.6E+01 .
2. Iodines. .. . .. . .. _-.. .. .. ..... . ...... .. 4 I-:L 31 : . C: , 6.35 -03 2.06E-03 I- . 3 3 --
. I-:.
- C; 2.57 -03 4.12R-03.
. ..35;.... . C: . _ .1.74E-03.. -......5 708-03 Total For' .. ..._ .. . .. ....
-Period. C1 1.07 1.19E-02 f ..
.. ..... .... .. ...E-02... . _ . .. .. _...__. ...
~
- 3. Par
.ticulates Sr-89 . C; ,
l.69E .19E-05
'Br-9 Cs Cl . .35E .75E-08 Ba- 0 C:
C;
. .302 .482-08 La- 0
.38 -os__ .74E-05 C: . . 45 -0$. .54E-04, b '
. . . ..Othe.l _ . r. _ ..C. .; ..... 3..S -07J...... ..4 8 E. .0 7..-- . . . ... .. ..
J otal For eriod C1 1.18E-04 y ......... ...._-.......... ......2.4 4E-04
*2eroes in thisdetectab..e table ;.ndicate that no radioactivity was Aower - . above presentimits - of detect:.on levels for ga. seous, See Table sample2-8 for ty analyses.pical.
pathways
**There are at noPlan bgtch Hatch. mode radioactive gaseous release l-a f - -
g
-3s.
t a _'~.
-. . ,-. ,- . . - . . - - . . . , ., i. .J... c- .. _ , - - . . . -
(d' TABLE 2-3b E. I. MATCH NUCLEAR P - UNIT 2 SEMIANNUAL RADI0 ACTIVE EFTLUENT ASE MPORT 1992 CASEOUS ETTLUENTS - ELEVATED LEASES *
.................................o..............h.. Continuous M de Mode **
Batc Nuclides Released Unit Quarter 1 Qua . . . ..2. . .Quarte
..................................rter . . . . . .r. . . . .Qua 1
. . r t e r. 2 1
Fission.. Gases Kr-35m CL 5.99E-01 5.03 -01 Kr-87 CL 1.50E+00 4.24 +00 e-13 3- C: 1.50E+00 2.99E+0p e-135m C:. 1.07E+01 1.70E+0. e-135 CL 3.795+00 4.06E+00
. e-138
_-_......... ....CL2.11E+01 5.3BE+01 Total For Period
......___.......__.... Ci 2E+01 3.9.. .... 8. 26E+01
- 2. Iodines 131 l. 5.07E-03 2.06g-03
-133 : . 2.36E-03 4.12E-03 Tot
....._.....a.........
I-135 For C 1.72E-03
.._.. 70E _03S ..._ .....____...... .......
Per d Ci 9.14E-03 1.19E-02
. O.
- 3. Particulates Sr-89 C:, 1.6FE-05 1.19E-05 Sr-90 CL .T J
.40g-08 6.75E-0 Cs-137 CL .-08 5.4BE-0 Ba=140 CL .35E-05 7.74E-0 La-140 CL .60E-05 1.54E-04 Ot
.___ _.h.er1 ............. Ca 35E-07 6.48E-07 Tota For Peri d
_........ ....................... C1 1.17E-04 4E...04 2.4
- Zeroes i thi table ndicate that no radioacti ity was er6ser etectab a levels. See Table 2-8 or ty lover .t a ovoim ts o detect on for qsseous sample ana yaea.pical
**There are no batch mode radioactive gaseous release pathways at Plant Hatch.
O 36.
1
- q '
. r'y SEMIANNUALiAbio hE hFTABLE}-3c .5S PORT 1992 GASIOUS EFFLUENTS - - ELEVATED RELEASES
- Conti ous
.............c.h
. . . . . . . . . . . . . . . . . . .. . . . . . . . n u.. . . . .M. o d e Nuclides Bat Mode **
Re. leas ed.......U. nit Qu a rte.r.1 ....arter Qu 2 Qua ..............r
.............rter 1 Quarte ..
2
.1... F..ission Gases r-85m CL 1.20 +00 1.01 +00 r-87 Cn 3.00 +00 8.48 +00 e-:.3 3 - C:. 3.01 +0 5.99 +0 Xe-:.35m CL 2 3.40C+0 Xe = h,3 8 35 C; , 7 .64E+0 15E+0 8.11E+
Xe- Cn 4.26
.......................E.+.01. ....1.08t+0 02 otal For Ci 7.90 eriod
................... ._..!+01 ............ 1.65E+
02........................
- 2. Iodines I-131 Cd 1.142-02 12E-03 I-133 Ch 4,93E-03 A.24E-03 8.
3.46
.. .I_135. _ . . . . . . . C. .;. .......E.-03._... 1.14E-02
- ot 1 For 1.98E-02 er od
.... ..............................E-02 C1 2.38
- 3. Particulates Sr.89 C 3.34E-05 2.37E-05 Sr 90 C: , 1.26E-07 1.35E 07
'Cs.137 CL 1.86E-07 1.10E-07 Ba-140 CL 6.73E-05 1.55E.04 La.140 C:. 1.32E-04 3.08E-04
... ...th.eO
___................6 r1 CL 1.27E-0 1.30E-06 Tot 1 F
-Per od or 2.35E-04
.................._........... C1 4.88E-04 ..................__....__....
*2eroes 1 this table ity was nresent a ove detectabl.eindicate levels. that esanoTable radioactig;.or 2-8 typical lower lim to of deteot;.on for gaseous sample ana..yses.
**There ar pathvays aI Solan$ b tch mods radioactive gaseous release Hatch.
/
v) . 37
l i f* LJ TABLE 2- a ' E. I. MATCH NUC PLANT _- UNIT 1 SEMIANNUAL PJ.DIOACTIVE EFF NT RELEASE REPdC 1992 l CASEOUS EFFLUENTS = OROUND-LEVEL RELEASES
- l l
Conti uous
. . . . . . . . . . n. . . . .M. .od.e............. Batch Mode ** .
Nuclides Released._..... ._....U.n.it Qua.rter.1 .Q.uart.e..r
.... .... . .. 2...Q.u arter Quarter 2 1 ......
1.
. F..ission
... .cases Xe-133 Ci 1.082- 2 00E+00 Xe=135 C1 1.37E+ 1 1.
0 172+01 Total For Pe
.r.Iod Ci 1.37E+01
... ... .......... ...... .. 17E+011 .. ..._ . .... ... ....
2.I.odines
-131 C:, 1.77E-04 1.09E-04
-133 C; C:,
3.56E-04 8.80E-04 I-135
...0 00Et00 4.262-04 ot 1 For er od_ .... .. ...i C 5.33E-04
.. ...... . .1.41E-03
- 3. Particulates
/N ...........................___...__ ...... ..................
V Cr-51 Cl. 6.26E-06 1.60 -04 Mn-54 C: , 4.09E-06 8 -06 Co-58 C; . .03E-05 2 .96E-05 74 Co.60 C;. .75E-05 3.41E-05 Zn.6$ C: C.,
.12 -04 6 48E-04 Sr-89 , 9.86 -06 7.74E-06 Sr-90 C; 0.00 +00 7.10E-08 Mo-99 C:., 0.00 +00 8.27E-05 Cs-137 C; , 5.95 -07 0.00E+00 Ba-140 C: , P,.17 -05 4.68E-05 La-140 C; ,.61 -05 7.73E-05 otherl C;, :. 34E-06
. ... ........... 5.63.E._06 . .... . .... __... ..
Tot 1 For
- Per od Ci 5.09E-04 1.10E-03
*2aroes in this table present above detectab.$ndicate .e levels. that Seeno radioactivity Table 2-8 for va$ ty cal lower limits of detect; on for geseous sample analysee.p
' **There are nE batch mode radioactive gaseous release pathways at P. ant Match. O N.)
- p. ,
()f_ TABLE L-4b . E. I. MTCH WCLEAR PLAC - UNIT 2 SEMIANNUAL RADICACTIVE EFFLUENT R3 LEASE REPORT 1992 GASEQUS EFFLUENTS - GROUND-2/EL RELEASES
- Cont Betch Mo
. . . . .. .. . . . . . . . . . . . . .. . . . . .i n.u ous M od e........................de**
.ee .....
Nuclides Unit Quarter ......ter ReleasM...........................1 Quar
........................r 2 Quarter 1 Quarte ... 2 1
Fission.. Cases e-131m CL 1.28E-0 0.00 +00 e-133 -Ca 0.00E+0 8.27 +01
. . . . .Xe-13
. . . 5.. . . .C : . 2.35E+0 3.72 +01 Tota For 1.20E+02
........................E..+.01 Peri d C1 2.35
- 2. Iodines I-131 C: 3.98E-04 5.82E-04 I-133 C 9.20E-04 3.76E-03 I-135 C;. 3.43E-04 6.63E-0J
............................................u.................
Total For Peri od ........C1.. .. 1. 6 6E..-0
.... 3 1.10E 02 l
- 3. Particulates ;
(~) s# .................... .......... ............................. : Co-58 Cf . . 7E-06 .00E+00 ! Co-60 CL . 1Fe0* . 00E+00 l C: . SE-05 0.00E+00
'In-65 r-89 C:.. .32E 1.87E-0$ '
Sr-90 C . 0.00E+00 3.68E-08 Mo-99 C:, 1 16E-04
. 0.00E+ 0 Cs-137 C:, 4.63E-07 1.05E- 6 Li-140 C:, 2.00E-06 7. SE- S La-140 Ca 4.72E-06 1 Other1 C. 2.23E-05 3 . 4E-043E-05 Total For Period..........Ci .........E.
1.90
.04 2.47E-04
*Zoroes i nresent a ove detectab ndicate this-table e levels.t hat SeenoTable radioactivityfor was-lower lim ts of detect on for gaseous sample2-8 ty analyses.pical
**There are no batch node radioactive gaseous release-pathways at Plant Hatch.
D-0
-39
. . . _ . . . ._- . . . . _ _ . . . . . _ . . _ _ _ _ _ ._ . ..-_4
-i
%.L i :. gz ,
-TAB -4c I k .
~ - SEMIANNUAL EA 10 CbE- E $5 LUEN PORT GASEOUS EPPLUENDS - ROUND- RELEASES *'1992 -1 Continuous Mode- Batch Mode **
-Nuclides Released..
...U. nit Quart.er 1.......Q.uart..er.'2 ....
... .. ....u..rter -Qa .. ..
1..Q..ua..r.ter.. 2 . 1.. .Fiss.
. .ion..Ga.se.s e-131a C; l,.28 - O. OE+ 0 . .. .... .....
e-133 C ,
, . 08 - 8. 7 E + 1' C, ,
.. ..e-13..S= .. ...... ..J.72 +.... .. . 4. SE+ 1 otal For" ~
eriod Ci
..... 3.72.E+01'. ........ 1.32E+02 2.Todines-
-131- - Of .
5.76E- 4 6.91 -04
-133 C; :. 38E- 3 4
-135- -03 C;. 3.43E- 4 7 64 Tota 1 For--
. . 05 -03. .. _... ..
Peri d
. ..__ ........-Ci. .
2.19E._03_..
... . 1 24E...02
~
'.3 Particul'ates
... n-54 ......... ........ 6.26 .......... '. 60 ._.......................
;[ '
r-51
. J :.
- o .09 -
-0
-0 Co-5 C:. ,
.16 - [p.7o.90 0 Co-6 .[t.05 -
n
.4.',-0 Ir0 '4. -
r ,.
- 3. ~ fp ) -
>E- 5
, 0. + 112 - 7 Mc- 9= .. :1. -4 ..h YE- SE
! Cs-137 : C;. '
- c. -06 j, . 052= 6 ,
-Ba-140 - C:.
ht i 1:.d.7-05 ! :81
..17g-04
__. _.. m._...t...LL _...............%....:_81..... 8: LTotal'For _.. ..__.... _.._
. ..Period.. .. .. .TCi .. ...
6.99 ..E...04. .1.35E..-03
.-*2eroes-1 thi table b e~ levels
- uresent a ove etectab. .ndicatethatnoradioactiityva! See Table 2-8 or ty cal-lower lin ts.o detect;,on for ga.-seous sample ana ypes.p pathways:at Plant Hatch.**There are no batch mode radionotive gaseous release
~
l i L vr*- p . . , .e r g r- - e i---r -- + . - --y
O Todio 2-5
~
E TC NUC PLAN - ITE SEMIANNUAL ftA 10 CTI E EFF NT AS REPORT 1992 QASEOUS EFTLUENTS - DOSE RATE Dose Rt.tes Due to Noble Cases Organ Tech Unit Quarter 1 of och Quarter 2 4 of t Dec Tech Lh Bod 500 mram
.. .......... . .it...... ........ SpecL.. uit..
. k..i n.y 3000 nrem r 7.14E-02 1. 1.30E-01 2.60E-02
........1.54E-01. . . .. 5.43E-02 r 1
...2E-03 2.42E-01 ..941E-03 h3ffD"3ff.33I.33b"$b3b3!bE!i.Ifbbb"#L.".3$.$"bb33kf3I*.....
Organ Tech Unit Quarter 1 % of g jech Quarter 2 % of Tech Bone 1500 mre r 9.21E-05 6.14E-06 ". 88 -04
.....Lk $t. . .... .. .., ...$$ft . ...
Liver TBodv 1800 tre r 1.33E-02 8.87E-04 !. 36 -02 1500 mre
- 1. 5 -05
- 9. 5 -04 T yr61d 1500 mram rr 2 1 -02 04 1.35E-02 8 K onay 1500 mrom r 1.33E-02 8.83E . -02 1.90E-63 4.48E-02 2 .99E-03 97E-04 Lun 1500 mrem r 1.40E-02 8.88E-04 1.36E-02 9.09E-04 O GIL
------------------r -------- ---------
1500 mrem 1.32E-02 9.32E-04 1.47E-02 8.83E-04 1.35E-02 8.98E- 9.79E-04
---------------- --04 -
O
.u.
_ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . - - - - . - - - - - - - - - - - - - ~ - _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - _ ^
1 ~ t
- ) '
1 TABLE 2-6a b -- l E. I.-HATCH NUCLEAR-PLANT - UNIT 1
'( SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1992 AIR DOSES DUE TO NOBLE GAS RELEI.SES
~
L Cumulative Doses per Quarter , T pe Tech Unit Quarter 1 % of Quarter 2 % of'
- o Spec -Tech Tech L Radi-- Limit Spec Speg i ation Limit Limit l Gamma 5.0 mrad 7.54E-03 1.51E-01 7.26E-03 1.45E l Beta 10.0- mrad 9.12E 03 9.12E 8.04E 03. 8.04E 02 L ________-._________________ --___ _ --_-02 Cumulative Doses This Year l T pe Tech Unit Quarters % of o SpeQ 1 &' 2 Tech o Radi- Limit' Spec l ation L1mit L
b$bb5~~ 5 bib mr$d 5Ikkb-b2 5Ik8E b5 Bota 20.0 mrad 1.72E . 8.58E-02 ___________________-_--_--_-02 _____--___________________--___--- ' (R : r
'S t
l t I I-l- &s
*4 #
ll? 73 . . TABLE 2-6b
~>
t E. I. HATCH NUCLEAR PLANT - UNIT 2
-(f; SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1992 AIR DOSES.DUE TO NOBLE GAS RELEASES -
i Cumulative Doses per Quarter T pe Tech Unit Quarter 1 % of- Quarter 2 % of o Spec Tech Tech Radi- Limit Speg Speg ation. Limit Limit. Gamma 5.0 mrad 1.26E-02 2.51E-01 2.80E-02 5.60E-01 Beta 10.0 mrau 1.55E-02 1.55E-01 4.77E 02 4.77E 01
. Cumulative Doses This' Year.
T pe Tech Unit Quarters % of o Speg 1& 2 Tech
'Radi- Limit Speg ation Limit Gamma- 10.0 mrad 4,05E-02 4.05E-01' Deta 20.0 mrad 6.33E 02 3.16E-01 V
f*
-L3-
- f. ..
/"'
- d. TABLE 2-7a-E. I b'TCH NUCMAR PLANT - Unit 1 SEMIANNUAL RADIOACTIVE tFFLUENT RELEASE REPORT 1992 INDIVIDUAL DOSES DUE TO RDIOIODINE RITIUM,
. AND PMTICUMTES IN GAGEOUS RK ES Eu"31*tiv*9!!!!.E!E.9"*E!!E.............................._..
Organ Tech Unit Quarter 1 Soec of Quarter 2 % of Limit och Tech Dec Spec ___......__.......imit......._......._... Limit ....... B no 7.5 mram 5.42E=04 7.23E-03 L ver 7.5 mrem 6.53E-04 8.71E-03 T odv 7.5 mren 1 1 77E-03 2.36E-02 3.54E-03 4.72E-02 Thyrbid 66E-03 2.21E-02 3.39E-03 4.53E-02 7.5 mren 1.78 -02 3. 9.78E-03 1.30E Xianey 7.5 mren 1.70 2. -02 3.40E-03 Lune 25 mram 1.53 -03 2. 4.53E-02
-CILLI 45 mre -02 3.22E-Op 4.29E-02
.____........ ......m. _......___.2....... .... __...... 29E-02 1.53 -03 -02 3.22E-0., 4 Cunu
......lative Doses T.his Year Organ Tech -Unit Quarters
- (?) . t$15t 2'2 .o{
......__.........._......_........$$$t ......__......___ ...__..
Bone J- 15.0 mren 1.20 -03 7.97E-03 Liver N
' .0 mrom 5.31 - 3 T Mbid T vr
.0 0
mram mrom 5.05 - 3 54 37 -02
-02 i 2.75 - 1.84 -01 K Bney !. .0 mrom 5.10 - -3.40 -02 Lung 15.0 mrom- 4.75 -03 3
. GIEI 15.0- area 4.75E-03 17
_. __....._........... .........3 16E-02...E-02 4 l l 1 o _LL. l l _-- - ~ g
W
-s' r .
f -
~
TABLE 2-7b E - - .HATCH NUC 2A't P
--I - Unit 2 SEMIANNUAL kADICACTIVE EF7W:NT ASE MPORT 1992
?O lLADICIOD NE TRITIUM, INDIVIDUALDOSESDU"IhNGASEQUS AND PARTICUMTE ELfASES Cumulative D.oses per
......._ ...... Quarter..................................
Organ Tech Unit Quarter 1 % of Quarter 2 % of ggeg Tech Tech
-............. .............. 2__ _ b............... b.. >
-B ne 7. 5: arem 3.55E-04 -4.73E 3.05E-04 4.07E-03 L ver' 7.5 nrem .99E-03 6.65E-02 3.09E-03 4 12E-02 T odv 7.5 -arem .96E-03 6.62E-02 3.07E-03 4.10E-02 Thyrbid' -7.5- mres .22E P2 2.95E-01 1.85E-02 2.47E-01 Ki8ney -7.5 nrem 5.01E-03 6.69E-02 3.12E-03 4.16E-02 Lung 7.5- nrem 4.93E-03 6.57E-02 3.05E-03 4.07E-02
.......................4.94E-03 GILLI 7.5- mram 6.59E-02 3.06E-03 4.082 02 Cunulative Doses This Year
- Organ Tech Unit Quarters % of
='
-; e Sgmgt 1&2 Tech L spec Limit
.............n....... ____......_............................
i Bone 15.0 arem -6.60E-04 4.40E-03 Liver 15.0 Erem 8.0BE-03 5.39E-02 15.0 mrom 8.04E-03 '5.36E-02 TBody'd 15*0 nrem 4.07E-02 2.71E-01 ThEroiKI ney 15.0- mrem '8.14E-03 5.42E-02 Lung 15 mram 7.98E 5.32 GILLI ..
...... 15 .00 ,mrom 8.00E-03.............__......
5.33b02 02 l: I-Q
-w.
. _ _ _ , - . _ - . _ _ . - . ~ . _ _ . _ . _ _ . - . _ _ _ _ - _ _ _ . _ . _ _ _ . _ . _ _ _ . _ _ _ .
,o ;
> {
TABLE 2-8 ! "= LOWER LIMITS OF DETECTION - GASEOUS SAMPLE ANALYSES O- The values in this table represent apriori lower limits of I detection-(LLD) which are typically achieved in labotatory j analyses of gaseous radwaste samples. l RADIONUCLIDE LLD UNITS i Kr 1.31E-07 uCi/ml Kr-88 2.10E-07 Xe-133 1.62E-07 Xe-133m 6.07E-08
;? Xe-135 5.77E-08 Xe 138- 2.85E-06 -
!*'31 4.37E-14 1 13i 6.16E-13 401-54 -2.78E-14
-Fe-59 4.62E-14 Co-58 2.46E-14 Co 2.88E-14 2n-65 7.51E Mo-99 6.02E-13 Cs-134 3.64E-14 Cs-137- 2.88E-14 x Cc-141 4.945-14 i Cc-144 2.02E-13
- Sr-89 1.00E-11 Sr 1.00E 11 - 3 1.00E-06 H
bO' i L6-
-p
i e 3. SOLID WASTE 3.1 REGULATORY REQJIREMENTS The Technical Specifications presented in th)s section are for Unit 1. Requirements for Unit 2 are the same as for Unit 1; however, the TWmical Specification ntabers are not the same. TECilNICAL SPECIFICATIONS 3.15.3.1 The n11d radwane system shall be used in accordance wit. 1e PROCESS COVfROL PROGRAM to provi:le 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 alpment of radioactive wastes f rom the site. 6.9.1.9 states in part: The Radioactive Effluent Release Report shall include the following information for ea h 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 racionuclides (specify whether ,
determined by messurement or estimate)
- d. Type of waste, e.g. , spent resin, compacted dry waste, evaporator bottoms
- e. Type of container, e.g. , LSA, type A, type B, large quantity
- f. Solidification agent, e.g., cement.
3.2 SOLID WASTE DATA I egulatory guide 1.21 Table 3 is found in this report as Table 3-la, and 3-lb. l 0
!.7-J
o EFFLUENT AND WASTE DISPOSAL SD11 ANNUAL REPORT 1992 January 1,1992 - June 30,1992 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS FOR UNITS 1 AND 11 TABLE 3-1A FORM TITLE: REG GUIDE 121 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL AEPORT OF SOLID WASTE AND IRAADIATED FUEL SHIPMENTS PERIOD COVERED FROM 1/1/92 TO 6/30/92 FOR UNIT- 1&2 A SOLID WASTE SHIPPED OFr, J FOR BURIAL OR DISPOSAL (Not irradiated fuel)
)ITT'IFE 1
OF W ASTE I UNIT j 6 month ,Est. Total l i period !ERAOR%I a Spent resins, fdter sludges. LM3 ;118E+02 ,J 6pparator bot 1 oms, etc. i. C4 i3 86E+02 l 1.0 E + 01 b Dry compressible easte, i M3 :4 ?E.01 ! contaminated eau'p etc . C i1.JE+00 L1.0E+01;
- c. trradiated componentt, control g M3 ! !
j l rods.Recire Pump Internals - Ce t j ' d Control Rod Dnve Filters LM3 !10BE+00 i ' Ci 12.11 E.01 2.0E + 01
,e. Other (desetibe) _M3 i !_y i equip etc i C4 ! l !
- 2. Estimate of major nuchde composition (by type of waste) .
I ISOTOPE iPERCENT CURIES ]
'A: Zn-65 I533E+01 2.06E+02 j i Fe-55 l 1.24 E.01 4.76 E +01 j183E+01 7.05EWI
[Co-60 i _ 11.36 E +00 526E+00, C4-137 I 1.47 E +01 5.65E + 01
]THER 13.33E+01
' 5.59E-01 ],
'B: Zn-65
! Fe-55 !103E+01 %73FW1 i Co-60 I 3.57E+11 ! 5.98E-01 i 8.79E+00 i 147E-01 hs-137 i t.19E+01 i1.70E-01 1 OTHER i i C: Zn-65 Fe-55 i Co-60 ! i C5-137 i i OTHER I
.D: Zn-6s 1199E+00 :4.19 E-01 i Fe-55 I 4 77E+01 I1.01E+01
! 4.77E+01 1.01 E+01
[Co-60 iO00E+00 0.00E+ 00
! Cs-137
. OTHER i 2 53E+00 5 35E-01 I:E Zn-65 i Fe-55 _
! Co-60
[Cs-137 i OT T7 I 3 Solid Waste Disposition Number c1htmems Mode of Transpor1ggpn Qegnpjgn 49 TRACTOR TRAILER BARNWELL B IRRADIATED FUEL -iHIPPMENTS (Disposition) Numtser of Sh'2 : M_ ode of Transpqrtation Destination
!!one "cne :;one
_ _j
p -~ EFFLUINT AND WASTE DISPOSAL SDi!MNUAL RUMT 1992 January 1,1992 tune 30,1992 SOLID WASTE AND IRRADI ATED FUEL SillPMEVfS
- Vf; FOR (WITS I AND II TAf1LE 3-IB TORM TITLE: UNIT 1 AttD 2 TECH SPEC 6.9.1.9 ETTLUENT AND WASTE
_DISEQhAL.JZ2ilAlfNUAL REPORT OF SOLID _ WASTE ANDJRRADI ATl2 D FUEL SHIEl(ENTS PERIOD COVERED TRCH: 1 / I / 92 Mt 6 f 30f 92 FOR UNIT: 1&2 TY PE Ct?RIE P9 fMCIP11 DL1LAl Nt'Whl1 Wtt1ME TYPE 3(R,jptf1 CAT 1oM OF Qt;ANTTTY NtK17Pta CONTAINf3 (Th MINF18 (WI AIN FJ $HIPMt NT A(Y tfT WAFTE 14713 MINED [IE713 MIN E D tt$CRlFTION SHi>M.D CUBIC FT (bMTAtht2 DEWAT11ED 1ei ImeS. C4*0 HKJH I 132 4 TYPE 3 f t3INI Fs33 IN1TCRfiY CASK NiA MTASt*ED MEARRfD CONTAINil 10-142
,,~x l ItWATTRED 20$ Zaes.Cc60 HmH Jo 202.1 TYTT A RE31NS Cel37. 7e53 INTEGRTTY CA$K N/A MEASLEED MEAst*ID CTWTAINER 14-210 DAW I 68 Za63.Ca60 FTRONO 18 93 3110NG (L4Y) Cal 37. Fm13 TICiff Ticett N/A (ACT!VD MFAll'RFD MEAll1 LED CONTAINER CVNTAINf33 (W ASTT) 8 23 DOKE. 3 23 90XE3 CRD 21.f Zed 5. Coe0 HIGH l M1 TYPE 5 F1.1135 Fe33 IVTTORfiY CASK N/A MT.Ast: RED MEAst*ID COWT AIN[3 CN$1-)$$
+ CONT AINS TOTAL VOLUME SHIPPED FROM VENDORS As WELL AS PLANI"
/"T Y
_L9
t 4 CilB'GES TO TiiE PLWf IMTG1 ODC*i f PCP O Technical Specifications 6.9.1.8 and 6.9.1.9 requite in part that changes to the Plant llatch Offsite Dose Calculation Manual (ODCAO be reported to the Commission in the next SemiariiiGal Effluent Release Report. On February 27, 1992, the PRB recommended approval of changes to the Plant flatch OlX31. The changes which have occurred in the Radiological Environmental Monitoring Program (REM 10, consist of changes in the Meterological Section and ninor editorial changes. Documentation of the PRB review and approval of these changes is presented in Appendix A of thi s report. Appendix B of this report contains the nodified pages of the Plant llatch ODCNi discussed in Appendix A. These modifications to the Plant 11atch ODCM are -identified as Revision 6. There were no changes to the Process Control Pro;;ran. s O I O
-50
__ . _ . . . _ _ . . _ . _ _ . ~ . . . _ _ . . . . - _ _ . _ . _ _ _ _ . _ _ _ _ _ _ . _ . . - - - _ _ _ _ .
> l l
-j 1
Georgia Power Company l Plant.E. I.liatch ! Semi annual Report
-Plant Radioactive Effluent Releases '
January 1 through June .30,1992 Appendix A r i t
. ODCM Change Documentation I
+
b e i i k r t d I b 1 y
.l h
p 4 lw r O , l? $ F t 1 :: l
-51 1-o
.- . ~...-.4..- _.~... .. ,_ _.. .. _-._ _ .-_.. .., __.. ._..~.. - .,_.- -,~.._ -. _..+
2
CE0hefA POWER COMPANY l PAGE 1 Of_J F uder E . I . RATCH
'TORM TITLE: j s DOctgenfT CHANGE REQUEST Change Affects ( ) Unit One ( ) Unit Two ()<3 Units One and Two REQUEST NO: 92 Of" REVISION NO: O kl REQUEST DATEL/ >2 9 2.
C __________________________________________ L E Request Icitia+eds (>O On-Site, ( ) Off Site By: C/yde M Da'on A R Document (s) Affectedt OWM/11 ta/aaaAL of N Flumu . D 7 n 3% d4 7/ pc
& :. A . -les SAcM - Y/* 6.DGd - //IF: / ' '
C O M P. _ , _ L CI N G E 7.nirironmental Impact Statement Required: ( ) YES ()<) NO (>Q YES ( ) NO Described in FSAR: Unreviewed Sciety Question: ( ) YES (>4 NO No. Attachments 3 [$d%1SE l_Q2-07-92. b / &l/7/94 ht.ve da l 2 ho/8 L MANAGER VSC DATE
/ NSC MEb5ER DATE NSC SUPV./ COORDINATOR 'DdE PRB Recommends Change ( YES ( ) No f 84/ $ U 'fd PRS MEETING NO. DATE P H no, give reason _
! R . m Request: f' Approved. ( ) Disapproved ud e ** M 4 1 /[U DATE
'r_ 2__
L Q1-NP H disapproved, give reason n LU 1 B02.13 00AC-REG-003-OS ENG-0015 Rev. 3 c l' u ;
. . _ . _ _ _ _ _ _ _ _ _ . _ _. . . _ . _ _ . _ ~ . _ . _ _ _ . ...._> _ _ _
CEORGIA POWER COMPAFY PACE 2 OT 2 Ff1WT E.I. KATCE
, KRM TITLE: ,
DOC 1 MENT PRANCE RE00EST , DoCR Not #1-00 ._ REVISION Not (> . _ _ _ Date Revissa Doctaments Received on Site 1 h e~' - 6 'Y/6 '//96 l
/
N PRB Secretary Etified By _
< Date MM
/ /
70 y C Change Implemented Under Amendment / Revision Numbert V. d L E (Attach any correspondence concerning this change to original DoCR) A R ___________________________________________ S NSC Dispositions ( Reconnend Approval (As Requested) A F ( ) Recoamend Approval (With Conneents) E T ( ) Recosusend Disapproval Y II disapproval recosmsended, give reason (s): ' A N 4 D Change Dispositions (/) APPROVED (As Requested) ( ) APPh0VED (With Comments) ( ) DISAPPROVED ' C OO $ w1_ 4 l2
- l9 L N _
DATE P MANAGEUNSC L II-disapproved give ressant I A N C Followu,o action initiated II disapproved above E ( ) VOID this Dr.% and issue new onet New DoCR Not ( ) REVISE this doc 2 New Revision Not and Date Revisedt ( ) OTHER: MANAGER NSC DATE l l 302.13 00AC-RIG-003-OS ENG-0015 Rev. 3 -
Page 1 of 4 i GIORGIA POWER C(EPANY
- f. PLANT E. 1. HATCH FORM TITLE:
ATTACHMENT 1 SAFE 1Y EVALUATION ( ) ONE : ( ) TWO : ( X ) ONE AND TWO O UNIT NUMBER: DOCUMENT NUMBER: DOCUMENT TYPE: 92-0/f DOCUMENT CHANGE REQUEST (DoCR) M vim . TEVISION NURR E DOICURGT TITLE (DRICABLE): 0 ODCH REVISION A. SAFETY SYSTEM APPLICABILITY (THIS SECTION IS NOT REOUIRED TO BE C FOR PROCEDURES): The iocument to which this evaluation applies represents:
- 1. ( X ) YES , ( ) NO A change to safety-related equipment or document?
BASIS FOR ANSWER: This DoCR proposes changes to the 00CH.
- 2. ( X ) YES ; ( ) NO A change which could impact safety-related equipment or document?
BASIS FOR ANSWER: See A.1 above.
- 3. ) YES ; ( X ) NO A cnange to a system, structure, or component
( which handles / controls radiation hazards? BASIS FOR ANSWER: This DoCR does not involve plant design or operational change. . eek . .Am B. 10 CFR 50.59 APPLICABILITY: The document to which this evaluation applies represents:
- 1. ( ) YES ; ( X ) NO A change to the plant as described in the FSAR (i.e., will this change require a revision to some portion of the FSAR)?
BASIS FOR ANSWER: The reference to the Radiological Ef fluents Monitoring Program in the FSAR, which is actually described in the Technical Specifications, is not affected by these changes.
- 2. ) YES ; ( X ) N0 A change to procedures described in the FSAR
( (i.e., is the document a skfety related procedure?)? BASIS FOR ANSWER: This is not a procedure revision. O
................................................................................ 10AC-MGR-010-05 MGR-0020 Rev. 1 N/A
Page 2 of 4 t GEORGIA POWER COMPANY PLANT E. 1. HATCH ATTACHMENT 1 ) ' VORM TITLE: SAFETY EV%UATION q g ................................................................................ B. 10 CFR 50.59 APPLICABILITY (CONTINUEQli
- 3. ( ) YE r; ( X ) NO A Test or Experiment not described in the FSAR which affects plant safety?
BASIS FOR ANSWER: This is not a test or experiment.
- 4. ( ) YES ; ( X ) NO A change to the Technical Spect'ications and/or Environmental Technical Specifications ir.corporated in the operating license?
BASIS FOR ANSWER: This DoCR propows changes to the 00CM only. 1E the answer to 211 the questions in section A and B are 'HO", Skip section C. If the answer to any question in section A or B is "YES", complete section C. C. UNREVIEWED SAFETY OVESTION CRITERIAi p 1. ( ) YES ; ( X ) NO Does the proposed activity increase the probability of occurrence of an accident ( previously evaluated in the FSAR? BASIS FOR ANSWER: The plant will continue to operate as den-and analyzed in the FSAR. The proposed changes affect the 00CM only. They do not decrease the effectiveness of the 00CM, are in compliance with applicable regulatory documents, and are in agreement with the Technical Specifications. (Refer to Attachment 2 for supporting information for these conclusions.) Therefore, this does not increase probabilities or consequences of an FSAR analyzed accident or equipment malfunction, nor does it create a different type.
- 2. ( ) YES ; ( X ) NO Doss the proposed activity increase the consequences of an accident previously evaluated in the FSAR?
BASIS FOR ANSWER: See C.l. b) v .
................................................................................ 10AC-MGR-010-05 MGR-0020 Rev. 1 N/A
GEORGIA POWEF. COMPANY Page 3 of 4
, . PLANT E. I. HATCH A 77Ach!m T / '
FORM TITLE: SAFETY EVALUATIDN C. pHREVIEVED SAFETY OVESTION CRITERIA (CONTINUEDl:
- 3. ( ) YET ; ( X ) NO Does the proposed activity increase the probsbility of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR7 BASIS FOR ANSWER: See C.1.
- 4. ( ) YES ; ( X ) NO Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR?
BASIS FOR ANSWER: See C.I.
- 5. ( ) YES ; ( X ) NO Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the FSAR7 BASIS FOR ANSWER: See C.l.
- 6. ) YES ; ( X ) NO Does the proposed activity create the probability O (
of a malfunction of equipment important to safety of a different type than any previously evaluated in the FSAR? BASIS FOR ANSWER: See C.l.
- 7. ( ) YES ; ( X ) NO Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification?
BASIS FOR ANSWER: See C.1. if the answer to AD.Y of the questions in section C is "YES', an unreviewed safety question is indicated. Approval froh, the NRC is required before the document can be used; reN r to subsection 8.3.2 for guidance on exceptions to this. O MGR-0020 Rev. 1 N/A 10AC-MGR-010-05
Page 4 of 4 GEORGIA POWER COMPANY } ATTACHMENT 1 l , . PLANT E. 1. HATCH FORM TIT!.E: SAFETY EVALUATION I e f ............................................ ................................... D. TECHNICAL SPECIFICATIQKLGONSIDERATIONSi
- 1. ( ) YET; ( X ) NO Does this document require a change to the Technical Specifications and/or the Environmental Technical Specifications?
BASIS FOR ANSWER: This is a change to the ODCM only.
- 2. E a change to the Technical Specifications and/or the Environmental Technical Specifications is required:
- a. Record associated Document Change Request (DoCR) Number:
- b. ) YES ; ( ) NO Does the amendment have to be implemented
( pr_ig.t to use of the document?
- c. E the answer to D.2.b was "YES", specify controls on document use:
- d. E the answer to D.2.b was "NO", tell why it is permissible to use it prior to implementation of the amendment:
PREPARER'S SIGNATURE: h Cl-/ /92-C/peh-7v2 REVIEVER'S SIGNATURE: YM 7- /m/d / 74-APPROVER'S SIGNATURE: M M,, 2 / 20 / 9Z m p i i t
................................................................................. 10AC-MGR-010-05 MGR-0020 Rev. 1 N/A
- . - - - . - - . _.~ .- -..---
h-4 L DocR 92-Oh. ATTACWENT 2 EXPLANATION OF CHANGES
- 1. Page 1: These are editorial changes to the Lble of Contants.
- 2. Page iv: Figure 3 J-3, ' Location of Additional Control Station for Milk,'
' is being deleted becaese milk is no longer available at location 3 T. The requirements of Unit 1 Te hnical Specification (TS) 3.16.1.c allow in part for the deletion of locations from which samples become unavailable. Efforts made to find a replacement location were unsuccessful due to the lack of other milk animal locations in the area. The TS requirement in footnote b of Table 3.16.1-1 states that up to 3 sampling locations within 5 miles will be used as available, and that 1 or more control locations beyond 10 miles will be used. Therefore, this requirement continues to be met by location 304, which is 10.8 miles away.
- 3. Page 2.2-47: The value of the annual ground level deposition parameter (0/Q) is actually 8.80 E-9, but because of a clerical error it '
is currently given as 9 (0 E-9. Because this value is a direct multiplier in the equation for organ dose gi'.en on page 2.2-5 of the 00CM, the use of the incorrect value has yielded , a higher calculated dose than would actually have occurred. This change is purely administrative.
- 4. Page 2.3-2: The existing reference to Table E.4-8 is being changed to Table 1 of Referent * .'3, which is being added. Reference 23 incorporates updated meteorological tables, into the 00CM, that '
contain values developed from data collected from the new Plant Hatch meteorological tower. Since the Hatch Semiannual Ra e Effluent Release Reports have included data co e rom the new tower since its completion, the pru:ous reference in the 00CM has not affected information contained in the reports.
- 5. Page 2.3-3: This change incorporates adjustments to elevated release height h to account for site specific local terrain elevations, thereby yielding more accurate dose calculations.
- 6. Page 2.3-5: The dnse equation buildup factor (Bg) equation and the air coefficient are being changed to match the equation given in Reaulatory Guide 1.109.- Revision 1 pages 1.109-73, figure F-4, l
and 1.109.74, Figure F-11. This Regulatory Guide information i is the correct information acceptable to the NRC mand connitted to by Plant Hatch.
- 7. Page 2.3-6: Refer to explanation for adding Reference 23, item 3 above.
- 8. Page 2.4-3: Refer to explanatien for height adjustments, item 4 above.
LO . _,---,,,,w,-,r,wv-.
, =w r, ww rw -
-y--w-v,g~cm-5--Smr er ,
Attachment 2
,Page Teo 9 Pages 3.0-1, 3.0-3, 3.0-4, 3.0-9, 3.0-10:
O for the deletion of the milk locations, refer to the explanation of item 1 above. ! tem 1 addresses the deletion of location 316 due to unavailability of milk, which is the sane reason for location 216 being deleted from page 3.0-3. The Tech. Spec. allowances explained in Ites 1 apply to location 216 as well. In addition, the distance to 1ccation 304 given on page 3.0-4 has been reduced fors 10.8 to 10.3 miles. This is a conservative change that measuras from the plant to the pasture center, rather than to the milking parlor.
- 10.
References:
Item 13 in the References section is being replaced by anothe' source of information, item 23 is being added, and is explained in item 3 above. O M1 l 0
. E * $ . . 97 - Of'
> '~
. . $ ' 2 0. 3 FA8LE OF CONTENTS
'l LO IG 1.0 L10010 EFFLUENTS 1. 0.1 51 LIQUID EFFLUENT MONITOR SETPOINTS 1.1 -1 1.1.1 Liquid Radweste Ef fluent Radiation Monitors 1.1 -1 l l 1.1.2 Plant Service Water Monitors 1.1-9
/ 1.2 DOSE CALCULATION FOR LIQUID EFFLUENTS 1. 2-1
- 1. 3-1 1.3 00SE PROJECTIONS FOR LIQUI 3 EFFLUENTS /
1.3.1 Monthly Dose Projections 1. 3 -1 1,3.2 Dose Projections for Specific Releases y 1.3-2
/
.4 DEFINITIONS OF LIQUID EFFLUENT TERMS 1.4-1
.5 LIQUID RA0 WASTE TREATMENT 3YSTEMS 1.5-1
.6 MIXING '0F LIQUID WASTE TANKS 1. 6-1 7.0 2,0-1
/ GASEOUS EFFLUENTS
/2.1 GASEOUS EFFLUENT MONITOR SETPOINTS 2 .1 -1 r<~
i
./2.1.1 ' Unit 1 Practor Building vent St 4, 2.1 -1 Unit 2 geactor Building vent ack, Unit 1 Recombiner Building V nt,- an Building Exhaust Augmented'Y tilati
/p.1.2 Main-Stack- 2.1 .
/2.1.3 _ Determination of Allocation Factor, AG L' 6 Y2.2 6ASEOUS EFFLUENT DOSE RATE AND DOSE CALCULATIONS 2.2-1
/2.2.1 _ Unrestricted Area Boundary Dose Rates 2.2-1 2.2.1.a Dose Rates Due to Noble'sases 2.2-1
/2.2.1.b Dose Rates Due to Radiciodines, 2.2-2 Tritium, and Particulates
. 2.2 Unrestricted Area Air Oose and Dose 2.2 to Individual ,
-I2.2.a Air Oose -in Unrestricted Area - 2.2-4
.7 2.2.b Dose to-An-Individual in _ 2.2-5 :
Unrestricted Area y 1 2.2.2.c Dose Calculations to Support _2.2-r ' Other Specific Technical 7 Specifications // HATCH 00CM,' REY 5 12/89: i .
' " "'e'D-e W c'- M w'saf'Nw MW zT U D M4 T W-r er -Pe tPMW7'wuP e w-r w 9 -wr N e'w ei'ime werNecr'-=DNmwvN=-=v_ ts-+wN1r wm Tr==wa r - 41e 9e ew--s hebTE perwmww' ut f'*
t at \ ', t
, t.
k117 0F FIGURES b_ Title noe
' F i o u re_ .
Exampic Calibration Curve for Liquid Effluent Monitor 1.1-10 1.1 -1 ! U cuid_Radwaste Treatment System (Unit 1) 1.5-2
- 1. 5 -1 1.5-3 Liquio dadwaste Treatment System (Unit 2) l 1~. 5 -2 2.3-7
,6 2 ,3 -1 Vertical Standard Duiation of Material in a Plume (oz) 2.3-8 ;
2.3-2 Open Terrain Recirculation Factor ' Plume Oepletion Ef f ect for Ground-Level hieases 2.3-9
- 2. 3 -2 . 2.3-10 2.3-4 Plume Depletion Ef f ect f or Greater Than 100-m Releases
_ Relative Deposition for Ground-Level Releases 2.3-11
'2.3-5 2.3-17 2.3 Relative Deposition for Greater Than 100-m Reletses Condenser Offgas Treatment System 2.5 T
- 2. 5 -1 .
Radiological Environmental Sampling Location Map 3.0-5 3 . 0 -1_ ($1te Periphery) 3.0-6 3 . 0 -2 -ram ological Environmental Sampling Location Map
~(Beyond. Site Vicinity) -3.0 ^
2.0 2 L :: tie- ef 'dditir :! Centre! Stat':- fr 3.0-3.0-f Location (,f Additional Control Station for TLDt 3 and Yegetation P O c.-.w..._.,.,,,,,,,,,,,,,
- A E...-. Of _ _
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e F-O 12/89 iv ! HATCH-00CM 'REV 5 l C, Sb -
, CQCR No. 9e-or TABLE 2.2-12
! ' Ti ^. C h' A ,,)., i ? .' ),
I J
^ "
.. . _ _31 C F M/f FONTROLLING RECEPTOR
. (g) ' ' -
. -.._. ..- - [Ter Support Section 2.2.2.b)
The location and exposure pathways associated with the controlling receptor are determined based on information obtained during the Annual Land Use d Survey.* Tispersion and deposit on values were calculated f or the location of the controlling receptor based on the HNP site meteorological data collected for the period 1984 through 1986.** l Location: W5W at 1.2 mies ! Age Group: Child Exposure Pathways: Inhalation, grcund-plane, and garden vegetation Dispersion and Deposition Values for This Location: Ground-Level SeC l (x/Q') = 3.18E-6 [ n pC ( 8t
.l.. ,
2 P (0/0')vp (>= +-90E-A m Elevated e sec (x/Q')sp = 6.53E-B , i "a i (0/0')sp = 1.37E-9 m
- Reference 22. l
(~T } . ** Refere.1ce 21. t w/ HATCH 00CM, REV 5 12/89 2.2-47
]CCR No. 92 . _ cr (T ACHMENT Not 3 . %. 1.n ain recirculation factor corresponains to a cistance r > ri d d OF . raken f1 om Figure 2.3-2.
= the number of hours the wind of wind soeed class j is directed njk ilt,o the given sector during the time atmospneric stability q e4ory k g These values may be obtained f rom Table ca _
- C . t 0 o f R e f e renc e ,lf.N
/ 23j.s uv
- N = the total hours of valid meteorological data recorded for all se pdARs, and stebility categories f rom nd s
Table E--4-t of Ref erenc e ,r. T w
/
23 ~ r = the distance from the point of release to the receptor lo stion, in meters,
= the wind speed (mid-point of wind speed class j) at ground ujk level (m sec-1) during atmospheric statellity k.
= the vertical standard deviation of the plume concentration T Ezk distribution considering the initial dispersion within the (3) ouilding wake.
(a'
- b'/2v) #*
= the lesser of or
# ( z)
= the vertical standard deviation of the plume concentration ok z
distribution (meters) for a given distance and stab 1.ity category k as shown in Figure 2.3-1. The stability category is determined by the vertical temperature gradient aT/az ('C/100 m).
< = 3.1416.
D = the maximi.n height of the adjacent s,lant structure (47 m), g Mt.TCH 00CH, REY l 5/11/84 2.3-2
D O G N o.-- 92 CI
' l t . - ATTACH.GT No. 3 2.3.1.b Elevated Releases SHEET 8 OF f-/T O X/0 = the sector-averaged annual average relative concentration at any distance in the given sector for radionuclides other than noble gases.
= 2.032 K r y ak nikeX (-h 3 tk')
jk NUjr azk where:
= the plume deDietion factor teren from Figure 2.3-4 F e en-6k .
elevated release, this fattar is stability dependent.
. h = e Wese aleau hoV+, :n <nJeo . = hs - h. -
~
g /1 W.T'. % h = the height of the main stack (120 m).
~ /Wfd #nn 5 b& hC = tew. tin hwoir a.+ few ha: d sduest , m nierers (Cb%.ed &cn n = fegate tne number 2.h is et' AtterM of hours the winct S) ff o wind speed class j is directed _ y
' Thto the given sector during the time atmosiiheMr staYW catgorykexisted. These values may be obtained from Table a- . m
( g of Reference. ' n) = NU -w the wind speed (mid-point of wind speed class j) at the he.Jht Ujk of release h (m sec-1) luring ~ atmospheric stability k. N = the total hours of valid meteorological cata recorded for all sectors, wind speed classes, and stability categories from Table E.4-7 of Reference 5. The remaining ,yabols are the same as those previously defined. When considering the direct ganrna radiation f rom an elevated finite plume, the constants Bi and Vi defined in Section 2.1.2 for each identified noble gas radionuclide are calculated using the following: Bj = K,I I I hik Ati wa Et I f rikt mead /vr rjki N Uj vCi/sec
. J HATCH 00CM, REV i 5/11/84 2.3-3
. fl;ft : ErEUn e 12l<%C hetp,tr twy he {w H,er ac(jk tul i+r p/a..,
i :u in acewla. a es h
~ ,e t k,, E.4. 3 2 a ' f.p p c., Q g
- S Caese.,a 5
DOCR No. 9z - cf ATT.S.CF 6:NT .W.3 _'?!iT 84 in[ star en=niiviied integral is used in representing the surrnation of the dose rates from all the points (the concentration within a differential volume element) of a distributad source (pivme). Fcr the sector-averaged Saussi6n model, the integral is given uy: 1 Sq[u.va,(L8 +Z8)1/8] G(z) exo (-u (La + ga)t/s] L dL dz 28'8og )J L3 +Z 8 where: 8 = the buildup factor. 9
.~-
(+ ' g'1/kulla+Z)1/8 8
'g)fceks
- x t': .1~- n%
u
= the &br' stricr; tion toef ficient f or air (m*1) .
k = (u va)/Va* L = the horizontal distance from the receptor to the differential volume element. _n z
= the vertical distance f rom the receptor to the differential volume element.
G(z) = a futiction relating the variation of radionuclide concentration with the height for a given differential volume element. 8
= exo (- (z-h)a/2eg 8) + exp (-(z+h) 8 /2az]
2, L dL dz = the dif f erential volume element of the integral. (The 2e is subsecuently factored out of the integral, leaving the L dL dz.) O HATCH 00CM, REV 1 5/11/04 2.3-5 - - - - - _ - _ _ - _ - _ - - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ --------_..-.-m..m . _,_ __ _ _ , _ _ _ _ _ _ _ _
~
l
,')C G ?Q. 92. 6f n ,..(...
. . w ".v. .:;T s ,t:o. >
+m /7 4 e a ne ne u describing the samma dose rate from a finite elevated plume is found in Chapter 7 of Ref erence 10. Numerical methods for evaluatimy the dose integral are found in Appendix F of Reference 3 in the forfn of two computer code listings. The first listing performs the integrativn using a Gauss-Legendre quadrature; the other utilizes a
-Newton-Cotes formulation. Details of these two methods may be found in Reference 9.
2.3.2 Relative Deposition Relative deposition per unit area is given by: 0/0 = the sector-averaged annual average deposition at any distance for a given sector. 2.55 Krgk on I N (m-8) k r where: O-2.55 = the inverse of the number of radians in a 22.5' sector
~
(2w/16)-1 Kr = the terrain recirculation- f actor described in the previous section. Og = the deposition rate at a given d3 stance taken from Figure 2.3-5 f or ground-level releases and Figure 2.3-6 for e.levated releases, nk
= the number of hours the wind is directed into the sector of interest 'or a ground-level release or for an elevated release
$4ng wtrk(CTiine stadrtTTtV categpey-t-existyThese values
_7 may be f ound in Tables 4.,-T and +:4-tt of Ref erence,F; " ( 23
/ 7 ,
N- = the total hours of. valid meteorological data) [hw ' 7ab/4> n / M 2 Cf !?c {enots 2 3. '
- 0681n
-HATCH 00CM, REV 1.' 5/11/84 2.3-6 g
i OCO. :o. 92__.pf I
- U,' O :'.2f*II .N O. 3 'ection of Definition Initial Use Term .
_A r - 7gg v (0/Q')$p = the annual average deposition parameter for the 2.2.2.D location of controlling (critical) receptor for racin stack releases. I Et
= the energy assigned to energy group t (Hey). 2.3.1 r _. s ,. y ., ,
tLlYC L Ye vel (~ h = the +%+vete+ release height (meters) = h., h'
- 2. 3. b
\, h, Ine ndyar =r the. mw n o mck L t 2.o mmJ L .t . i he = lerro r hdy nl at IC&L*s cd I'Ttitsr Intfu>) 3I
( l{ kt
= g results o umerica,1QtegraQover the spatial distribution of an elevated finite plume.
, G 2,L 4 = the plume depletion f actor for all radionuclides 2.3.1 other than noble gases at distance r.
ok
= the plume depletion factor for all radionuclides 2.3.1 other than noble gases at a distance r for the appropriate stability class k (radiciodines and particulates).
K = a constant associated with the Ki and Yi calcu- 2.3.1 lation representing a cornbination of constants and unit conversions. Kj = the total body dose f actor due to gar.ma emissions 2.1.1 froes radionuclide 1 (mrem /yr per vC1/ms) f rom Table 2.1-1. l
,y.
HATCH 00CM, REV 3 8/87 2.4-3 l I 1
1._ 6 1)'A .i .. . ' .!. .") i . .. , o .g sc
... . . c .;. 3 i' v:: _g1 a #r _,
O 3.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i Sampling locations, as required by RETS 4.16.1 (Unit 1),3r3_ described in Table 3.0-1 and shown on maps in Figures 3.0-1 through 3.0-{.
\ >s
\ - N,_,,/
There are no known drinking water users downstream of Plant Hatch. Ther"* ore, the LLO f or I-131 in water need not be as stringent as that for milk. The survey of milk animals is based on the requirement in Appendix ! to 10 CFR Part 50 that the licenses
- Identify changes in the use of unrestricted areas (e.g., for agricultural purposes) to permit modifications in monitoring programs for evaluating doses to individuals from principal patt< ways of exposure,' The consumption of milk from animals grazing on contaminated pasture and the consumption of vegetation contaminated by airborne radiciodine are major potential sources of exposure. Samples from milk animals are considered a better indicator of radioiodine in the environment than vegetation. Because sufficient milk samples frequently are not available within 5 miles, vegetation samples will be collecteo also.
Grass is available almost year-round, whereas leafy vegetation is available only for 8 months of the year at best. The sampling stations for gra n are located near the site boundary in two sectors with high of fsite D/Q values where it might be practical to establish a vegetation plot. The highest of fsite 0/Q tor each individual sector occurs approximately at the site boundary. Although either fish or clam samples may be collected f rom the river, fish samples are praf erred, because the maximum dose comitment to a member of the public as a result of liquid effluents is through the fish consumption pathway. { , Q}.) ( l I HATCH 00CM, REV 5 12/89 3.0-1
,0C0 Mc. 9z.of ,
AIIAC .'in'.NT i .O. 3
- nir fio cejk r p - -
Q TABLE 3.0-1 (SHEET 1 0F 2) ~
' RADIOLOGICAL ENVIRONMENTAL SANPLING LOCATIONS ;
Distacce Sample Location Tvoe
- Number Descriotive location Direction (silet)
D64- Roadside part WNW 0.8 0 101 Inner ring N 1.9 0 102 Inner ring NME 2.5 0 103 Inner ring- NE 1.8 AD ' 104 Inner ring ENE 1.6 0 105 Inner ring E 3.7 0 l 106 Inner ring ESE 1.1 DV 107- Inner ring SE 1.2 AD 100- Inner ring SSE 1.6 0 10g Inner ring S 3.9 0 110. Inner ring SSW 1.0 0 111 Inner ring SW 0.9- 0
- 112- Inner ring WSW !.0 ADV 113- Inner ring W 1.1 0 114 Inner ring WNW 1.2 0 JO .
Inner ring NW 1.1 0 115' o. 116 Inner ring NNW 1.6 AD l
)
170 Upriver WNW R
~~='1s=
R 172 Downriver E 201. Outer ring N 5.0 0 202 Outer ring NNE 4.9 0 203 Outer rific NE 5.0 0 1 204 Outer ring ENE .5.0 0 205 Outer ring E 7.2 0 206- Outer ring ESE 4.8 0 l . 207 Outer ring SE 4.3 0 o
, 208 Outer ring SSE -4.8 9 20g outer ring .S 4.4 0 210- Outer ring SSW 4.3 0 211- Outer ring SW- 4.7 0 l
212 Outer ring WSW 4.4 0 m 213~ Outer ring; 'W 4.3 0 214 Outer ring ~ WNW 5.4 0 215 Outer ring- NW 4.4 0- l 21 6 Outer ring- NNW 4.8 0
-4M- O':-t '; 70= -NNW- -+re- --M- l
*See Sheet 2 of 2 for. footnote.
l .
**See Sheet 2 of 2 for footnote. -
- )~
NATCH 00CM,LREY 3 8/87 3.0-3
- l. ,
1 l pcq :.w. n . or ; y TABLE 3.0-1 (SHEET 2 0F 2) l M d ..'i.2.i
.'e'O. !
] Ug CF [/f" RADIOLOG : CAL ENVIRONMENTAL SAMPLING LOCATIONS
\J' Location Distance Sample Numeer Descriptive location Direction (miles) Tvoe -
301 Toombs Central School N 8.0 0 1 304 State prison ENE 11.2 AD 304 State prison ENE 10.f3 M i 309 Baxley substation S 10.0 A0 l 4%- '
. . , . . . . . . . . , - -MW- :4 ,4- --n--
416 Emergency News Center NNW 21.0 DV i l
" Sample Types:
A - Airborne radioactivity 0 - Direct radiation (l k/ M - Milk R - River (fish or clams, shoreline sediment, and surf ace water) V - Vegetation
-m.
" Station 170 is located at approximately 0.6 river miles upstream of the l
. intake structure for rivervater,1.1 river miles for sediment and clams, and 1.5 river miles for fish.
Station 172 is located at approximately 3.0 river miles downstream of the discharge structure for riverwater, sediment and clams, and 1.7 river miles for fish, rte location f rom which riverwater 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 sufficient 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 f ew miles or so is generally needed to obtain adequate fish samples. The mile locations given above represent approximations of the locations about which the catches are taken. O)- t 0687n HATCH 00CM, REV 5 12/09 3. 0 -4 l 1 l' t.-_.
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LOCATION OF ADDITIONAL
/' CONTROL STATION FOR MILX
\ eDmN I. HArCH GeorgiaPower = vci.tx a r u a r - u nit i FIGURE 3.0-3
"" 3.0-9 l
n e ~ s 9jL . Of
' ' h.,~-w .. r ..
i, 7T,.C.'.....EN, T iJo. 5 l 4:..:n
;- /3f z er w pd I l
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.(
l HATCH 00CM REV 5 12/89 LOCATION OF ADDITIONAL CONTROL
'"l"A*^= STATION FORMYf(ETATION GeorgiaPower A (FIGURE 3.0-f3 )
,v -
_ 3.0- W Q
._ _ _ _ _ _ . . _ _ _ _ . . _ . _ _ . _ . _ . . _ _ . . _ . _ . _ - _ . _ _ .__m___ > _ _
~'
l ,
' ,w . , .G ? . :. ?2 .Of '
~* ~. I :. 5 REFERENCES l '> =' ' ' dMW
< u Ar
- 1. J. S. Boegli..R. R. Bellamy, W. L. Britz, and R. L. Waterfield. .
' Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants,' NUREG-0133, October 1978. I
- 2. 'CaEulation of Annual Doses to Man f rom Routine Releases of Reactor Ef fluents for the Purpose of Evaluating Compliance with 10 CFR 50.
Appendix 1,' U. S. NRC 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, Appendit 1
- U. S. NRC Requiatory Guide 1.109. Rev.1, October 1977. ,
4 HNP-2 Environmental Report - Operatinu License Stage, Georgia Power Company, July 1975, (including Appendix A. 'Edwin 1. Hatch Plant-Unit 1 , Preoperational Environmental Surveillance Report Number 1,* submitted March 1,1975). S. Edwin 1. Hatch Nuclear Plant-Unit 1 Final Saf ety Analviis RfDort, Georgia Power Company.
- 6. Edwin 1. Hatch Neelear Plant-Unit 2 Final Saf ety Analysis Recort, Georgia Power Company.
- 7. ' Methods- for Estimating Atmospheric Transport and Dispersion of Gaseous Ef fluents in Routine Releases f rom Light-Water-Cooled Reactors '
!!. 5. NRC Regulatory Guide 1.111, March 1976.
- 8. ' Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Ffil" _y in Routine Releases f rom Light-Water-Cooled Reactors,'
O. S. NRC Reaulatory Guide 1.111, Rev.1, July 1977.
- 9. Handbook of Mathematical Function! with Formulas.-Graohs and Mathematical Tables, edited by Abramowitz, M. and Stegun, I. A.,
National Bureau of Standards, U. S. Department of- Comerce,1965. H., U. S. Department
-10. Meteoroloav and Atomic Ener.qy, edited by Slade, D.
of _ Commerce, July 1968. :
- 11.
- Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for-the Purpose of implementing Appendix 1,'
O. S. NRC Requiatory Guide 1.113, Rev.1, April 1977.
- 12. Michols. M. C and Holder S. D., Plant Edwin I. Hatch Units 1 and 2 Thermal- Plume Model Verification, Gaorgia Power Company Environmental Af f airs Center, March 1981.
- 13. !!!::t c' it:t: :
'- "' Ont H:t:5 '.'-d d 'd::t: Unh". C ;- W 4 :-
e___.... m - _ .. sasa Se /Y cr 1. 5// lhall k IfCt-ML Erewrs Geug,k ?.n.r
$lmf@ty , (ChNt- ll, iISC.
f HATCH 00CM,-REV 1 5/11/84
i
,m-. ,.
m..s
* ~
L . Cf 31FERENCES (Continued) /, {. .
,,T l g, J
- niiT $5 CF !M
'. Hatt.h Nuclear kl ant land Use Survey - fiET, G60f a evwer udgany,
{%)h February 1987. . 15. Letter to Georgia, Pner Company f rom Pickard Lowe, and Garrir.k. Inc., Wast.imston, D.C., May 11, 1987.
- 16. Letter to Georgia Power Company f r{'a Pickard. Lowe, and Garrick. Inc.,
Washington, D.C. , June 3,1987.
- 17. Letter to Georgia -Power Company f rom Pickard Lowe, and Garrick. Inc.,
Washington, D.C., June 11, 1987,
- 18. Intern,1,l_ Hemorandum ,trqe nto 0.M. Hopper, Georgia Power Company, June 9
- 19. Letter to Geot _ . ..,0any from Ouantum Technology. Inc. Marietta, Georgia, June .,67.
- 20. Fu;*_ Nuclear Plant Land use survey, Georgia 2r 3r Company, N P sar 1987. ,
- 21. g,12r to Georgia Power Company f rom Pickard, lowe, and Garrick. Inc.,
W4snington. D.r.. November 30, 1587.
- 22. -$tch Nuclear Plant Land use Survey, Georgia Power Company, November 16, 1988, and December Jr 1988.
9 Z3 Q% m ZC _.un; s t%& Mixy34 m m J H. Qa,,> % d. Pj.y 4
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~
F b c i sH 00CH, REV 5 12/89
Georgia Power Company Plant E. I. Hatch Semi annual Report Plant Radioactive Effluent Releases January 1 through June 30, 1992 Appendix B Revised Pages From ODCs1 ) 1 6 s 1 O
. - _ _ -~_
i 7-s, lABLE OF CONTENTS d Page
.l.0 L10U10 EFFLUENT 1 1.0-1 1.1 Li, D EFFLUENT MONITOR SETPOINTS 1.1-1 1.1.1 Liquid Radwaste Effluent Radt; tion Monitors 1,1-1 1.1.2 Plant Service Water Monitors 1.1-9 c
1.2 DOSE CALCULATION FOR LIQUID EFFLUENTS 1.2-1 1.3 DOSE PROJECTIONS FOR LIQUID EFFLUENTS 1.3-1 1.3.1 Monthly Dose Projections 1,3-1 1.3.2 Dose Projections for specific Releases 1.3-2 1.4 OEFINITIONS OF LIQUID EFFLUENT TERMS 1.4-1 1.5 LIQUID RADWASTE TREATMENT SYSTEMS 1.5-1 1.6 MIXING 0F LIQUID WAST TANKS 1.6-1
~
2.0 GASEQUS EFFLUENTS 2.0-1 2.1 CA;r.0VS EFFLUENT MONITOR SETPOINTS 2.1-1 r) (x,/ 2.1.1 Unit 1 Reactor Building Vent Stack, Va.it 2 Reactor Building Vent Stack, 2.1-1 l Unit 1 Recombiner Building Vent, and Building Exhaust Augmented Ventilation 2.1.2 Main Stack 2.1-4 2.1.3 Determination of Allocation Factor, AG 2.1-6 2.2 GASE0US EFFLUENT DOSE RATE AND DOSE CALCULATIONS 2.2-1 - 2.2.1 Unrestricted Area Boundarv Dose Rate 2.2-1 2.P.I.a Dose Rates Due to Noble Gases % 2.2.1.b Dose Rates Due to Radiciodines, 2.. Tritium, and Particulates 2.2.2 Unrestricted Area Air Dose and Dose 2.2-4 to Individual 2.2.2.a Air Dose in Unrestricted Area 2. 4 2.2.2.b Dose to An Individual in .. 2-5 Unrestricted Area 2.2.2.c Dose Calculations to Support 2.2-17 Other Specific Technical Specifications
,/~x 4-HATCH ODCM, REV 6 4/92 i
.. .. 1 1
LIST OF FIGURES l E1oure Title Paae 1.1-1 Emple Calibration Curve for Liquid Effluent Monitor - ' 1-10
- 1. 5- 1. Liquid Radwaste Treatment System (Unit 1) 1.5-2 1.5-2 Liquid Radwaste Treatment System (Unit 2) 1.5-3 2.3-1 Vertical Standard Deviation of Material in a Plume (s ) 2.3-7 2.3-2 Open Terrain Re. .rculation Factor i 2.3-8 2.3-3 Plume Depletion Effect for Ground-Level Releases 2.3-9 2.3-4 Plume Depletion Effect for Greater Than 100-m Releases 2.3-10 2.3-5 Relative Deposition for Ground-Level Releases 2.3-11 2.3-6 Polative Deposition for Greater Than 100-m Raleases 2.3-12
-2.5-1 Condenser Offgas Treatment System 2.5-2 3.0-1 Radiological Environmental Sampling Location Map 3_0-5 (Site Periphery) 3.0-2 Radiological Enviror, mental Sampling Location Map 3.0-6
-(Beyond the Site Vicinity) 3.0-3 Location of Additional Control Station for TLDs 3.0-9. l and Vegetation-O k
u LHATCif ODC!4, REV 6 4/92 iv W v - . -_ , ,- - , . . . . .
L TABLE 2.2-12 C0t4TROLLitlG RECEPT 0R (lo support section 2.2.2.b) The location aTd exposu.e pathways associated with the controlling receptor are determined based on information obtained during the Annual Land Use Survey.* Dispersion and deposition values were calculated for the location of the controlling receptor based on the HilP site meteorological data collected for the period 1984 through 1986.** Locaticn: WSW at 1.2 miles Age Group: Child Exposure Pathways: Inhalation, ground-plane, and garden vegetation Dicpersion and Deposition Values for This Location: Ground-Level (TTQ' ),, = 3.18E 4 **' m, (D7Q') o e 8.80E-9 e 1 ! m llevated < (f7Q ' ) ,, = 6. T 3 E -B *
- c m, _
(57Q ' ),, = 1. 3 7 E -9 m
- Reference 22.
O ** Reference 21. HATC4 C2CM, REV 6 4/92 2.2-47
,s. K, -
the terrain recirculation factor corresponding to a distance r h taken from figure 2.3-2. n3 = the number of hours the wind of wind speed class j is directed ,
--iniq the given sector during the time atmospheric stability category k existed. These values may be obtained frc, Table 1 of Reference 23.
N * - the total hours of valid meteorological data recorded for all sectors, wind speed classes, and stability categories from Table 1 of Reference 23. r - the distance from the point of release to the receptor location, in meters. up - the wind speed (mid-point of wind speed class j) at ground level (m sec") during atmospheric stability k. I,, = the vertical standard deviation of the plume concentration
- distribution considering the initial dispersion within the building wake.
.n-N,j. I (a' + b'/2n)*"
= the lesser of or-
~
[3 (a,) g
. o,, - the vertical standard des;ation of the plume concentration distribution (meters) for a given distance'and stability category k as shown in Figure 2,3-1. The stability category is-determined by the vertical temperature gr3dient AT/oz (oC/100m).
n - 3.1416. b - the maximum _ height of the adjacent plant structure (47 m). y
-s i
7 n v HATCH ODCM, REV 6 4/92 2.3-2
, a :-
' l , ' [ .' . - -_ , _ - -
e
'2.3.1.b Elevated Releases X/Q - the sector-averaged annual average relative concentration at any distance in the given sector for radionuclides other than s.
noble gases. n,, e x p ( -h '/ 26,,*) I 6'
- 2.032 A, jk NUp 6,,
where: 6, - the plume' depletion factor taken from figure 2.3-4. For an elevated release, this factor is stability dependent. h - - effective release height, in meters - h, - h,. l-NOTE: Effective release height may be further adjusted for plume rise in accordance with sect hn E.4.3.2 of Appendix E of Reference 5. h, - the height of the main stack (120 m), h,- -- terrain height at location of interest, in meters (obtained
- from figure 2.3-12 of Reference 5).
n e the number of hours the wind of wind speed class j is directed into the given sector during the time atmospheric stability
~
category k existed. These values may be obtained from Table 2 of Reference 23. Up -
-the wind speed (mid-point of trind speeo class j) at the height of release h (e sec) during atmospheric subility k.
N - the total hours of valid meteorological dcta recorded for all . sectors, wind speed classes, and stability categories from
-Table E.4-7 of Reference 5.
The remaining symbols are the same as those previously defined. When considering the direct gamma radiation from an el vated finite plume, the
. consta'nts B, and V, defined in Section 2.1.2 for each identified noble gas radionuclide are calculated using.the following:
! B, - 3 III " ^" #* ' b"'
rjki N U, pC1/ sec HATCH ODCM, REV 5 4/92 2.3-3
O [1 D' The aforementioned integral is used in representing the summation of the dose rates from all the-points (the concentration within a differential volume element) of a-d_istributed source (plume).. For the sector-averaged Gaussian model, the integral is given by: 1 i" B,[p, ,,(L' + z')"') G(z) exp (- (L' + z')"') L dl dz 2 '!"g, . , . , L' '. z' .n
. where:
B, - - 'the buildup factor.
. 1. + kp ( L ' + z ' ) "' . I
-p- -
the attenuation coefficient for air (m). l k ' = _ (p . - p, ) / , . I L - the horizontal distance from the receptor to the {" differential volume element. z = the vertual distance from the receptor to the differential volume element.
.G(z). - a function relating the variation of-radionuclide concentration with the height for a given differential volume element.
- exp [- (z-h)*/2al] + exp_ (-(z+h)*/20,')
/2n L dL dz = the differential volume element of the integral. (The 2n
, is subsequently factored out of the integral, leaving'the L'dl dz.)
Y% U
. HATCH ODCM, REV 6 -4/92 2.3-5 L
1' p
- t. *: '
A derivation of the model describing the ganma dose rate from a finite elevated plume is found in Chapter 7 of Reference 10. Numerical methods for evaluating the dose integral are found in Appendix F of Reference 3 in the form of two cuputer code listings. The first listing performs the integration using a Gauss-Legendre quadrature; the other utilizes a Newton-Cotes formulation. Details of these twc methods may be found in Reference 9.
~2.3.2 Relative Decosition Relative deposition per unit' area is given by:
0/0 - the sector-averaged annual average deposition at any distance fer a given sector.
, [ 2. 5 5 K,0, n, .
k r N 4 where: 2.55 - the inverse of the number of rauians in a 22.5o sector (277/16 )" . K. - the terrain recirculation factor described in the previous - section. 0, - the deposition rate at a given distance taken from Figure 2.3-5 for ground-level releases and Figure 2.3-6 for
. elevated releases.
n, - the number of hours the wind is directed into the sector of g interest for a ground-level release or for an elevated l' release during which time stability category k exists. These values may be found in Tables 1 and 2 of Reference 23. l 5 9 - the-total hours of valid meteorological data, from Tables 1 and 2 of-Reference 23. O
. . wn u:- =m=x a ,
HATCii C:CM, REV 6 4/92 2.3-6
/ . 3.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 1-Sampling locations, as required by PETS 4.16.1 (Unit 1), are described in Table 3.0-1 and shown on maps in Figures 3.0-1 through 3.0-3. { There are no known drinking water users downstream of Plant Match. Therefore, the LLD for 1-131 in water need not be as stringent as that for milk. The survey of milk animals is based on tha requirement in Appendix ! to 10 CFR Part 50 that the licensee " Identify changes in the use of unrestricted areas __ (e.g., for agricultural purposes) to permit modifications in monitoring programs for evaluating doses to individuals from principal pathways of n exposure." The con umption of milk from animals grazing on contaminated pasture and the consumption of vegetation contaminated by airborne radiciodine
- are major potential sources of exposure. Samples from milk animals are concidered a better indicator of radiciodine in the environment than vegetation. Because sufficient milk samples frequently are not available ,
within 5 miles, vegetation samples will be collected also.
' Grass is available almost year-rcund whereas leafy vegetation is available only for 8 months of the year at best. The sampling stations for grass are "
located near the site bcundary in ti;o sectors with high offsite D/Q values _ where it might be practical to establish a vegetation plot. The highest - offsite 0/Q for each individual sector occurs approximately at the site boundary, Although either fish or clam samples may be collected from the river, fish samples are preferred, because the maximum dose commitment to a member cf the pubile as a result of liquid effluents is through the fish consumption pathway. HATCH ODCM, REV G 1/F2 3.0-1 l
l . - TABLE 3.0-1 (SHEET 1 0F 2) Cl RAT:0 LOGICAL ENVIRONMENTAL. SAMPLING LOCATIONS Lntation - Distance Sampla Number _ Descriptive location Direction (miles) Tvoe
- 064 Roadside park WNW 0.8 D 101 Inner ring N 1.9 0 102 Inner ring NNE 2.5 D 103 :nner ring NE 1.8 AD 104 Inner ring ENE 1.6 D 105 Inner ring E 3.7 D ,
106 Inner *ing ESE 1.1 DV 107. Inner ring SE 1.2 AD 108 Inner ring SSE 1.E D 109 Inner ring S 0.9 0 110 inner ring SSW l.0 D 111 Inner ring SW 0.9 D 112 inner ring WSW l.0 ADV 113 inner ring W l.1 0 114 Inner ring WNW l.2 0 115 Inner ring NW l.1 0 Inner ring 116 NNW l.6 All 170 Upriver WNW ** R 172 Downriver ** E R 201 Outer ring N 5.0 0 202 Outer ring NNE 4.9 0 203 Outer ring NE 5.0 0 204 Outer ring ENE 5.0 D b 205 Outer ring E 7.2 0 206 Outer ring ESE 4.8 D 207 Outer ri"g SE 4.3 0 208 Outer ring SSE 4.8 0 209 Outer ring S 4.4 D 210 Outer ring SSW 4.3 0 211_ Outer ring SW 4.7 0
-212 Outer ring WSW 4.4 D 213 Outer ring W 4.3 0 214 Outer ring WNW 5.4 0 215 Outer ring NW 4.4 D 216 Outer ring NNW 4.8 0 l
~*See Sheet 2 of 2 for footnote.
**See Sheet 2 of 2 for footnote.
HATCH ODCM, REV 6 4/92 3.0-3
TABLE 3.0-1 (SHEET 2 0F 2) k) RAD!0 LOGICAL ENVIRONMENTAL SAMPLING LOCATIONS Location _ Distance Sample Number Descriotive loc; tion Direction biles) Tvoe
- 301 Toombs Cent.al School N 8.0 D 304 State prison ENE 11.2 AD 304 State prison ENE 10.0 M i 309 Baxley substation S 10.0 AD l
416 Emergency News Center NNW 21.0 DV
- Sample Types:
V A'- Airborne radioactivity D - Direct radiation M - Milk P. - River (fish or clams, shoreline sediment, and surface water) V - Vegetation t
;O G
** Station 170 is located at approximately '.5 river miles upstream of the intake structure for riverwater, 1.1 river miles for sediment and c'ams, and 1.5 river miles for fish, b Station 172 is located at approximately 3.0 river milet downstream of the _
discharge structure for'riverwater, tediment anc lams, and 1.7 river miles .. for fish. The location from which riverwater 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 sufficient quantity. High water adds to-;he 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 obtain adequate fish samples. The mile locations given above represent approximations of the locations about which the catches are taken. V(3 - o . wi :m :c:s ce: no HATCH ODCM, REV 6 4/91 3,0-4 _____-__m___.m_- m _ _ _ --______.____-_-_.____m__a._-_- _- -__-.____-__ .___._________m_
+-
i f .~ b \ U(' : J ,
~
m u.
" t r'"
i t
/ 1 .
4 N
't
~
b l 1 f .
, ,1 ......,
_(8 '81 SCALE m wLES g w .,4 . cm4wiuu. : ,k wrw O ll
'l HNP 44 4 River &
j m e .. t 4 , l. 6.asley r.-.. ) I l
.....y ;
, I l- i "ATC" ODCM, REV s /92 LOCATION Or 2^:: ;0t,;' c-c----
. ,f tonisi.watcs t > ^.
- 54 - % F 0; , ."a AND '. Er' '" AT : : ,
Let gia Poner u$
;g ~
sucaia,,tiur FIGURE 3. -3 l i
- .e s.o 3.0-9 .
L i-4
- L f REFERENCES
- 1. J. S. Boegli, R. R. Bellamy, W. L. Britz, and R. L. Waterfield,
" Preparation of Radiological Effluent Technical Specifications for Nuclear GTwer Plants," t @EG-0133, i October 1978.
- 2. " Calculation of Annual Doses to Man from Routino Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix 1," U. S. NRC Reaulatory Guide 1.109, March 1976.
- 3. " Calculation of Annual Doses to Man from Routine Releases of Rear. tor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix 1," U. S. NRC Reaulatory Guide 1.109, Rev. 1, October 1977.
- 4. HNP-2 Envfronmental Reoort - Operatina license Staae, Georgia Power Company, July 1975, (including Appendix A, "Edwia 1. Hatch Plant-Unit 1 Preoperationa' Environmentai Surveillarce Report Number 1," submitted March 1, 1975).
- 5. Edwin I. Hatch Nuclear Plant-Unit 1 Final Safety Analysis Reoort, Georgi?
Power Company,
- 6. Edwin 1. Hatch Nuclear Plant-Unit 2 Final Safety Analysis Reoort, Georgia Powr. Company.
O 7. " Methods for Esticating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," O. S. NRC Requlatory Guide 1.111, March 1976.
- 8. " Methods for Estimating Atmospheric 1ransport and Dispersion of Gaseous Effluent: in Routine Releases from Light-Water-Cooled Reactors,"
L.S. NRC Reatlatory Guide 1.111, Rev.1, July 1977.
- 9. Handbook of Mathematical Functions with Formulas. Graphs and Mathematical Tables, edited b" Abramowitz, M, and Stegun, I. A., National Bureau of Strndards, U. S. Department of Commerce,1965,
- 10. tLeteoroloav and Atomic Eneray, edited by Slade, D. H., U. S. Department of Commerce, July 1968. .
- 11. " Estimating ,quatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of implementing Appendix 1,"
,U. S. NRC Reaulatory Guide 1.113, Rev.1, April 1977.
- 12. Nichols, M. C. and Holder, S. D., Plant Edwin 1. Hatch Vrits 1 and_2 Thermal Plume Modal Verification, Georgia Power Company Environmental Affairs Center, March 1981. ,
1
- 13. Internal Mer:arandum, Bill Duvall to Howa*d Pocerl, Georgia Power Company, .
October 11, 1990. O ! SATCH 00CM, REV 6 4/92 [
PEFEPENCFJi (Continued) 1987, Georgia Power Company,
- 14. Hatch February 1987. Nuqlear Plant I,and U$e Survey -from Pick 15.
l.etter to Gecrqia Power 11, 1987 CQmpany k rd, Lowe, and Garrick, inq., Washington, D.C., May letter to Gegrqia Power Company from Fiq a kard, Lowe, and Garrigk, Inc.,
- 16. Washington, D.C., June 3, 1987.
17. Letter to Georqit 11, PQwer 1987. Company from Piq Hqpper, Georgia Power Company, Washington, D.C., June t Internal Memorgnqvm, W ,H , Qlli.nqe.r. i.tta, _Q.j ,M. . . 18 June 9, 1987. f rom Quantum Technology, Jnc . , Mar e letter to GeQroia Power Company 17, 1987
- 19. Georgia, June Georgia Power Company, 20.
H. atch Nuqlear Plant I,and Use Survey, (owe, grd Garrick, Inc., d November 1987. GeQrgia Power Company 87. from Pickar , 6 21 Letter tQ Washington, D.C., November Georgia 30,19 Power Company, November l 1 l Hatqh Nuqlear Plant l,and t)se Survey, l 22 1988, and December 20, 19C8 from J. H. Davis, Health Physics t mber 17, 1990.
- 23. I,.,etter to GeQrgia Power CQmpany[ansultant, L t
't ODCM, REV 6 4/92
~~ *% ~~m
, . y WY, 4
-s s REfPffK[$
R (Continued)
./
- 14. Hatch Nuclear Plant Land Use Survey - 1987, Gesrgia Power Company, february 1987.
- 15. Letter to Georaia Power Company from Pickard. Lowe. and Garrick, in t ,
Washington, D.C., May 11, 1987.
- 16. Letter to Georaia Power Corroany from Pickard. Lowe. and Garrick. Inc.,
Washington, D.C., June 3, 1987,
- 17. L,1 r to Georaia Power Comoany from Pickard. Lowe. and Garrick. inc.,
Wr>'.ington, D.C., June 11, 1987.
- 18. Intet nal Memorandum. W.H. 0111naer to 0.M. Hoooer, Georgia Power ".or..pany.
June 9, 1987.
- 19. lettir to Georaia Power Company from Quantum Technoloav, inCz, Marietta, Geor.^a, June 17, 1987.
- 20. Hatc' Nuclear Plant land Use Survey, Georgia Power Company, November 1987.
- 21. Letter to Georaia Power Company from Pickard. Lowe. and Carrick. Inc.,
Washington, D.C., ilovember 30, 1987. x
,3
() 22. Hatch Nuclear Plant land Use Survey, Georgia Power Company, November 16, 1988, and December 20, 1988. l
- 23. Letter to Georgia Powe- Comoany from J. H. Davis. Health Physics
[g.nsultant, lilburn, Ceurgia, September 17, 1990. s t Q t y-4
/
HATCH ODCM. P.EV E 4/92 { 1 L \ 9}}