ML20029A684
| ML20029A684 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/31/1990 |
| From: | Hairston W GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| HL-1477, NUDOCS 9103010221 | |
| Download: ML20029A684 (75) | |
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w ntu,,aun m HL-1477 February 27, 1991 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555 PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50 321, 50 366 OPERATING LICENSES DPR 57, GPF-5 SEMIANNVAL RAD 10ACTlYE EFFLVENT RELEASE REPORT Gentlemen:
In accordance with the provisions of Plant Hatch technical Specifications Section 6.9.1.8 and 6.9.1.9, Georgia Power Company (GPC) is providing six copies of the Plant Hatch Units 1 and 2 Semiannual Radioactive Effluent Release Report. 1his report covers the period of July 1, 1990 through December 31, 1990.
Should you have any questions please advise.
Sincerel ,
, , b- Al W. G. Hairsto 111 WGH,Ill/DMH:ked Enclosure i
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l U. S. Nuclear Regulatory Commission i Page 2 i February 27. 1991 cc: Georai B_ Engr _(sapay Mr. H. L. Sumner, General Manager - Plant Hatch Mr. J. D. Heidt, Manager Engineering and Licensing Hatch Dr. W. R. Woodall, Manager invironmental Affairs Mr. C. L. Whatley, Manager Nuclear Insurance and Risk Management NORMS U. S. Nglcit Etnyht9fy CommissicacMhiu2LODuDi Mr. K. N. Jabbour, Licensing Project Manager - Hatch
!)A.$. Nglear Regula.12ty_f.qmmjl110n. RfiLI.on. Il Mr. S. D. Ibneter, Regional Administrator Mr. L. D. Wert, Senior Resident Inspector Hatch American Nue1 Elf 181V.Lt:I.li Mr. M. Marugg S. Late __of Georcia Mr. J. Setser, DNR
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! GEORGIA KMER CCNPAN'l 1 1
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- UNITS t10. 1 & 2 l 1
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SEMIAN!ETAL RADIOACTIVE I EFFLUEffT RELEASE REPORT
' July 1,1990 - December 31, 1990 i
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PLAtC E. I. HA7Qi SEMIANNUAL RADIOACTIVE
EFFLUEt(T RELEASE REPORT SECTIOt1 TITLF PACE
- 1. LIOUID EFFLUEtRS 1 1.1 REGULA70RY LIMITS 1 1.2 MAXIMUM PERMISSIBLE ODNCEtTTRATIONS 5 1.3 MEASUPEMEtITS AND APPROXIMATIQ4S OF 70TAL RADIOACTIVITY S 1.4 LIQUID EFFLUE2TT RELEASE DATA 7 1.5 PADIOIAGICAL IMPACT ON MAN DUE TO LIQUID RELEASES 9 2 GASEQUS EFFLUEtTTS 19 2.1 REGULATORY LIMITS 19 2.2 MEASUREMEtITS AND APPROXIMATIOf1S OF
'IOTAL RADIOACTIVITY 25 2.3 GASEOUS EFFLUElfr RELEASE DATA 30 2.4 RADIOIDGICAL IMPACT DUE 70 GASEOUS RELEASES 31 3 SOLID WAfrTE 47 3.1 REGULA70RY REQUIREMEtCS 47 3.2 SOLID WASTE DATA 47 4 SUPPLEMEtRARY INFORMATION FOR TFE PLATIT HATCH SEMIANNUAL RADIOLOGICAL EFFLUEtTP RELEASE REPORT FOR THE FIRST SIX MotfrHS OF 1990 50 5 METEOROLOGY 51 5.1 METDOROIOGICAL DATA 51
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I PLA!C E. I. HATCH +
SEMIAtNUAL RADIOACTIVE o I h EFFLUEtC RELEASE REPORT i
TABLE LI!TT OF TABLFS PAGE ,
1-1 TECHNICAL SPECIFICATION TABLE 3.14.1-1 RADIOACTIVE LIQUID EFFLUDTP M3N!'IORING INSTRUMDTTATION 3 1-2a LIOUID EFFLUDTPS - SUMMATION OF ALL RELEASES - UNIT 1 10 1-2b LIQUID EFFLUDTTS - SUMMATION OF ALL RELEASES - UNIT 2 11 1-2c LIOUID EFFWDTTS - SUMMATIONS OF ALL RELEASES - SITE 12 1-3a LIOUID EFFLUEtITS - UNIT 1 13
. 1-3b -LIOUID EFFLUE!TPS - UNIT 2 14 1-3c LIOUID EFFLUD7rS - SITE 15 1-4a INDIVIDUAL DOSES DUE TO LIQUID RELEASES - UNIT 1 16 1-4b INDIVIDUAL DOSES DUE 10 LIOUID RELEASES - UNIT 2 17 1-5. IOWER LIMITS OF DETECTION - LIOUID SAMPLE ANALYS1 10 2-1 TECHNICAL SPECIFICATION TABLO 3.14.2-1 RADIOACTIVE GASE(yJS EFFLUE!E MONI'IORItJG INSTRUMDTTATION 21 l 2-2a. GASEOUS EFFLUEtTPS - SUMMATION OF ALL RELEASES - UNIT 1 32 2-2b GASEOUS EFFLUDTTS - SUMMATION OF 33 ALL RELEASES - UNIT 2 2-2c GASEOUS EFFLUDITS - SUMMATION 07 ALL RELEASES - SITE 34 l 2-3a GASEOUS EFFLUDTTS - ELFVATED RELEASES - UNIT 1 35 1
2-3b GASEOUS EFFLUDTTS - ELEVATED j RELEASES - UNIT 2 36 l 2-3C GASEOUS EFFLUDITS - ELEVATED RELEASES - SITE 37 I
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l o l TABLES LIST Of TABLES Et&[
2-4a GASEOUS EFFLUENTS - GROUNDLEVEL RELEASES - UNIT 1 38 2-4b GA$EOUS EFFLUENTS - GROUNDLEVEL RELEASES - UNIT 2 9 2-4c _ GASEOUS EFFLUENTS - GROUNDLEVEL RELEASES - SITE 40 2-5 GASEOUS EFFLUENTS - DOSE RATES -
SITE 41 2-6a AIR DOSES DUE TO NDBLE GASES -
UNIT 1 42 2-6b - AIR DOSES DUE TO NOBLE GASES -
UNIT 2 43 2-7a INDIVIDUAL DOSES DUE TO RAD 1010 DINE, TRITIUM, AND PARTICULATES IN GASEOUS RELEASES - UNIT 1 44 2-7b INDIVIDUAL DOSES DUE TO RAD 1010 DINE.
TRITIUM, AND PARTICULATES IN GASEOUS RELEASES - UNIT 2 45 2-8 LOWER LIMITS Of DETECTION - GASEOUS SAMPLE ANALYSES 46 3-la,b SOLID WASTE AND 1RRADIATED FUEL SHIPMENTS 48-49 O
RADIOACTIVE EFFLUfWT RELEASE REPORT 1 LIQUID EFFLUEVTS 1.1. REGULA'IORY LIMITS
- 1. The Technical Specifications presented in this section are for Unit 1. Requirements for Unit 2 are the same as Unit 1r however, the Technical Specification numbers are not the same.
TECHNICAL SPECIFICATIONS 3.14.1 The radioactive 11guld effluent monitorina instrumentation channels shown in table 3.14.1-1 shall be OPERABLE with their alarm / trio setpoints set to ensure that the limits of Specification 3.15.1 are not exceeded. The alarm / trip setpoints of these channels shall be de'. ermined 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 e any time from the site to UNRES'IRICTED AREAS (floure 3.15-1) shall be limited to the t
.(~' concentrations specified in 10 CFR Part 20, Appendix B, t ( .
Table II (column 2) for radionuelldes other than l dissolved or entrained noble cases. For dissolved or entrained noble gases, the concentration shall be limited to 2.0E-04 uC1/nd total activity.
3.15.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in 11guld 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.
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- 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 11guld radwaste treatment system, as described in the ODCM, shall be used to reduce the radioactive materials in 11guld wastes prior to their discharge when the projected doses due to the 11guld effluent per Unit from the site (figure 3.15-1) when projected over the calendar quarter would exceed 0.18 lg< mrem to the total body or 0.62 mrem to any orcan.
3.15.1.4(a) The contents within any outside temporary tank shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or O entrained 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) unplarmed releases frm the site to unrestricted areas of radioactive mterials in gaseous and 11guld effluents that were in excess of 1 C2, excluding dissolved and entrained cases and tritium for 11guld effluents, or those in excess of 150 C2 of noble gases or.0.02 C1 of radiolodines for caseous releases".
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TE0!NICAL SPECIP! CATION TABLE 3.14.1-1 (SHELT 1 of 2) *
() . RADIOACTIVE LIQUID EFFLUttfr WNITORItJG INSTRUMDTTATIOtJ Minimum Channels Instrument OPERARI.P ADolicabi h ty ACTION l
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- 1. Gross Radioactivity Monitors Providina l Automatic Termina-i tion of Release l
i Liquid Radwaste Effluent Line 1 (a) 100 l
- 2. Gross Radioactivity Monitors not Providing Automatic Termination of Release Service Water System Effluent l Line 1 (b) 101
- 3. Flovrate Measure-
[h ment Devices **
-d 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 l
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- Pump curves may be utilized to estimate flows in such uses, ACTION statement 102 is not required.
(a) Whenever the radwaste discharge valves are not locked j closed.
l (b) Whenever the service water system pressure is below ,
the closed cooling water system pressure or l l
differential pressure indication is not available. '
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TABLE 1-1 (CoffrINOED)
TECHNICAL SPECIFICATIOt1 '
TABLE 3.14.1-1 (SHEET 2 of 2)
RADIOACTIVE LIOUID EFFLUEt7r MONI%3 RING INSTRUMEtTTATION v
TABLE tDTATIONS ACTION 100 - With the number of channels OPERABLE less than i
required by the Minimum Channels OPERABLE requirement, effluent releases msy be continued, provided that prior to initiating a release
- a. At least two Independent samples are analyzed in accordance with Specification 4.15.1.1.1.
- b. At least two technically qualified Individuals independently verify the release rate calculations and discharge valving.
Otherwise, suspend release of radioactive effluents via this pathway. If the channel remins 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 OPERABLE less than -
required by the Minimum Channels OPERABLE requ1rement, effluent releases via this pathway may continue, provided p,- that once per shift grab samples are collected and analyzed Q for gross radioactivity (
Detectionofatleast10'getaorgamma)ataLowerLimitof uC1/ml . If the channel remains inoperable for over 30 days, an explanation of the circumstances shall be Ir.cluded in the next riemi-annual ef fluent release repor+..
ACTION l')2 - With the number of channels OPERABLE less than rehred by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided the flowrate is est1msted at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. If the channel remains anoperable for.
over 30 days, an explanation of the circumstances 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.
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I 1.2 MAXIMUM PERMISSIBLE CONCINI' RATIONS
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[\ /-') The MPC values used in determining allowable 11guld radwaste release rates and concentrations i
for principal ganna emitters, I-131, tritium, St-89, Sr-90 and Fe-55 are taken from 10CFR Part 20, Appendix B, Table II, Column 2.
For dicsolved or entrained noble cases in liquid radwaste, the MPC is taken from Technical Specification 3.15.1.1 (Unit 1) ard 3.11.1.1 (Unit
- 2) as 2.0E-04 uC1/ml .
For gross alpha in 11guld radwaste, the MPC 2s taken from 10CFR Part 20, Appendix B, Note 2.b as 3.0E-03 uC1/ml.
Further, for all the above radionuclides or categories of radioactivity, the overall MPC fraction is determined in accordance with 10 CPR Part 20, Appendix B, Note 1.
The method whereby the MPC f racticn is used to determine release rates and 11guld radwaste effluent radiation monitor setpoints is described in Section 1.3 of this report.
( 1.3 MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY Prior to releace of any tank containing liquid radwaste, and following the regul:ed recirculation, samples are collected and analyzed in accordance with Technical Specification Tables 4.15.1-1 (Unit 1) and 4.11.1-1 (Unit 2). A sample from each tank planned for release is analyzed for principal gamma emitters, I-131, and dissolved and entrained noble casca by gamma srectrometry.
Monthly and quarterly composites are prepared for analysis by extracting aliquots from each sample taken from tanks which are released. Ligurd radwaste samric analyses are performed as follows:
Measurement Frequency Method
- 1. Gamma Isotopic Each Batch Gamma spectroscopy with computerized data reduction
- 2. Dissolved or Each Batch Gamma spectroscopy EnU 31ned with computerized f-sg Noble Gases data reduction
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1 Mxsurement Preauency, Method (3 3. Tritium Monthly Distillation and l C/ Composite 11auid scintillation countina
- 4. Gross Alpha Monthly Gas flow proportional Composite countina 1
- 5. Sr-89 and Sr-90 Ouarterly Chemical separation and composite cas flow proportional I
countana
- 6. Fe-55 Ouarterly Chemical separation and Comoosite low enercy photon detector.
Gamm isotopic measurements are performed in-house in the radiochemistry lab usino aermanium spectrometry.
Three germanium detectors are available: a 20%
efficient and two 15% efficient intrinsic aermnium detectors, with 2.0 fvHM resolution and housed in 4 2nch-thick lead shields. A one-liter 11guld radwaste sample is poured into a Marinella beaker in preparation for a 3000 second count. A peak search of the resulting camm ray spectrum is performed by the p computer system. Energy and net count data for all significant peaks are determined, and quantitacave (v) reduction or LLD calculations are performed for the l
nuclides specified in Table Notation e of Technical Specification Tables 4.15.1-1 (Unit 1) and 4.11.1-1
!~ (Unit 2): Mn-54, Fe-59, co-58, co-60, 2n-65, Mo-99, t
Cs-134, Cs-137, Ce-141 and Ce-144 The quantitative calculations include corrections for counting time, decay time, sample volume, sample geometry, detector l efficiency, baseline counts, and branching ratto. LLD
- l. calculations, includina 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 radionuellde would be located if
! present.
The radionuclide concentrations determined by camma spectroscoolc analysis of a sample taken from a tank
! planned for release and the most current sample analysis results available for tritium, cross alpha, Sr-89, Sr-90, . and Fe-55 are used alono with the corresponding MPC values to determine an MPC fraction for the tank planned for release. This MPC fraction l 1s then'used, with appropriate safety factors, alona with the expected dilution stream flow to calculate a 1 maximum permissible release rate and a liquid effluent I (_)
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monitor setpoint. The monitor setpoint is calculated to assure that the limits of Technical Specifications
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l 3.15.1.1 (Unit 1) or'3.11.1.1 (Unit 2) are not l exceeded. !
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A monitor reading in excess of.the calculated setpoint therefore results in an automatic term 2 nation of the 11guld radwaste discharoe.
rs Liquid effluent discharae is also autonatically I
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terminated af the dilution stream flow rate falls j below the dilution flow rate used in the setpoint calculations and established as a setpoint on the dilution stream flow monitor.
Radlonuclidt concentrations, safety factors, I dilution stream flow rate, and 11guld effluent 1
<;adiation monitor calibration factor are entered into the computer and a prerelease printout is 1 generated. If the release 18 not termissible appropriate Warnings will be included on the prerelease prantout. If the release is permissible it is approved by the Chemistry Foreman on duty. The pertinent infornation is transferred manually from the prerelease printout I to a one-page release termit which as forwarded to Radwaste Operations. When the release is completed the release permit is returned from Radwaste Operations with actual release data included. These data are input to the computer and a post release printout is generated. The post release trintout contains actual release rates, actual release concentrations and quantitles, actual dilution flow, and calculated
_s doses to an individual, v
1.4 LIQUID EFFLUEtc RELEASE DATA Reculatory Guide 1.21 Tables 2A and 28 are found in this report. as Table 1-2a for Unit 1, Table 1-2b for Unit 2 and Tab.?e 1-2c for the siter 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 1-2b, and 1-2c are calculated and the Tables completed as follows:
- 1. Fission and activation products - The total release values (not including tritium, cases, and alpha) are comprised of the sum of the measured individual radionue12de activities. This sum 1s -for each batch released to the river for the respective O
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L ti quarter. Percent of applicable limit is determined from a mixed nuclide MPC fraction
/~T- calculation. The average concentration for i) each nuellde over all released batches is divided by the corresponding individual MPC value. The sum over all nuelldes of the C/MPC ratios times 100 is the percent of apo11 cable limit for effluent releases during the quarter.
- 2. Tritium - The measured tritium concentrations in the monthly composite samples are used to calculate the total release and average diluted concentration during each period. Averace diluted concentration divided by the MPC limit, 3.00-03 uC1/ml, is converted to percent to otve the percent of applicable limit.
- 3. Dissolved and entrained cases -
Concentrations of dissolved and entrained gases in liquid effluents are measured by germanium spectroscopy on a one liter samole from each 11guld radwaste batch. The average concentration of dissolved or entrained noble gases for all released batches is div2ded by the MPC value stated
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In Technical Specifications 3.15.1.1 and '
i 3.11.1.1 ( 2.0E-04 uC1/ml) to determine the
\ MPC fraction. The result x100 la the percent of applicable limit for noble cases in liquid effluent releases during the quarter. Radioisotopes of lodine in any form are also determined durinq the isotopic analysis for each batch: therefore, a separate analysis for possible gaseous forms is not performed because it would not provide additional Information.
- 4. Gross alpha radioactivity - The measured gross alpha concentrations in the monthly l composite samples are used to calculate the '
total release of alpha radioactivity.
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Other data pertinent to batch releases of radioactive 11guld effluent from both units are as follows:
(V) !! umber of batch rele:ses: 558
,. Total time period for batch releases: 62,641 minutes Maximum time period for a batch release: 244.0 minutes Average time period for batch releases: 112.3 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: 2,932 CPS 1.5 RADIOLOGICAL IMPACT ON MAN DUE 'IO LIOUID 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 Offsite Dose Calculation Manual. As required by the above Technical Specifications, doses were calculated separately for Unit I and Unit 2.
Results are presented in Table 1-4a for Unit 1 j] and Table 1-4b for Unit 2.
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i TABLE 1-2a E. I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990 4
O LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES Unit Quartet- 3 Quarter 4 Est. Total Error (%)
A. Fis ion and act vation pro ucts -
- 1. Total relense Ci 3.82E-02 2.08E-02 4.70E+01
- 2. Average di..uted concentrat:.on dur:.no cer:.od uCi/ml 3.16E-08 2.62E-08 -
- 3. 4-oc Aph11 cable limat % 1.32E+00 9.20E-01 B. Tritium
-1. Total release C1 4.38E+00 3.50E+00 3.70E+01
- 2. Average di:.uted concentrat:.on ur: uCi/ml 3.62E-06 4.42E-06 ,
- 3. oc n$ph$[c;,od able '
imat- % 1.21E-01 1.47E-01 C. Disso:.ved and ses entra:.n$d
- 1. Tota r$$aase Ci. 2.07E-02 4.04E-03 1.00E+02
- 2. Average di:.uted
- concentrat;.on durj
- 3. 4o d dpnah$riod 1 cable uCi/ml 1.71E-08 5.09E-09
-limat % 8.57E-03 2.55E-03 D. Gross Alpha radioact1vity
- 1. Total release Ci- 0.00E+00 0.00E+00 1.20E+02 E.Volymeofwaste bS[u$[onf liters- 4.94E+06 3'.58E+06 1.00E+01 F.: Volume of dilution water used liters. 1.21E+09 7.92E+08 1.60E+02 O
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TABLE 1-2b E. I. IIATCH NUCLEAR PLANT - UNIT 2 SEMIAUNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES
.U Unit Quarter 3 Quarter 4 Est. Total Error (%)
A. Fis ion and act vation pro ucts
- 1. Total relepse Ci 5.01E-03 1.60E-02 4.70E+01
- 2. Average di-uted concentrat:.on during per:.od uCi/ml 6.04E-09 3.35E-08
- 3. % of applicable limit % 2.77E-01 2.93E-01 B. Tritium
- 1. Total release Ci 2.06E+00 1.62E+00 3.70E+01
- 2. Average diluted concentrat:.on durina cerhod uCJ/ml 2.48E-06 3.41E-06
- 3. %od 6phlicable lim $t % 8.27E-02 1.14E-01 C. Dissolved and entrained gases
- 1. Total release Ci 3.53E-03 6.01E-04 1.00E+02
- 2. Average diluted concentrat: .og
([]) 3. 1"sh"!psfluable u 1/ml 4.25E-09 1.26E-09 limat % 2.13E-03 6.31E-04 D. Gross Alpha radioactivity
- 1. Total relhase Ci 0.00E+00 0.00E+00 1.20E+02 E. Volvme of waste (Drlor to dilution) liters 3.71E+06 1.94E+06 1.00E+01 F. Volume of dilution water used liters 8.30E+08 4.76E+08 1.60E+02 O
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TABLE 1-2c E. I. HATCH NUCLEAR PLANT - SITE SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990
('e) LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES Unit Quarter 3 Quarter 4 Est. Total Error (%)
A. Fis ion and act vation pro ucts
- 1. Total release C1 4.33E-02 3.67E-02 4.70E+01
- 2. Average di:,uted concentrat:.on
- 3. furina per: odof 6pplicable uCi/ml 2.12E-08 2.90E-08
.init % 8.93E-01 6.85E-01 B. Tritium
- 1. Total release Ci 6.44E+00 5.12E+00 3.70E+01
- 2. Average di:.uted concentrat:.on during 3er:.od uCi/ml 3.16E-06 4.04E-06
- 3. %of Apallcable '
limit % 1.05E-01 1.35E-01 C. Dissolved and entrained gases
- 1. Total release C1 2.43E-02 4.64E-03 1.00E+02
- 2. Average di: uted f- concentrat:.on d
6p licable uCi/mi i .no ner:.od 1.19E-08 3.66E-09 3..%ur:C o limat % 5.95E-03 1.83E-03 D. Gross Aloha radioactivity
- 1. Total relbase Ci 0.00E+00 0. 00E4 00 1.20E+02 E. Volume of waste (prior to dilution) liters 8.65E+06 5.51E+06 1.00E+01 F. Volume of dilution water used liters 2.04E+09 1.27E+09 1.60E+02
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TABLE 1-34*
E. I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990 LlQUID EFFLUENTS
('N) v Continuous Mode ** Eatch Mode Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 H-3 Ci 4.38E+00 3.50E+00 rission and activation products Na-24 C; 7.47E-04 6.66E-04 Cr-51 C; 3.41E-05 3.23E-04 Mn-54 C: 5.99E-04 3.56E-04 Mn-56 C 2.73E-05 0.00E400 Fe-59 C: 1.02E-05 1.13E-05 Co-58 C; 1.02E-04 3.42E-04 Co-60 C: 3.81E-03 2.53E-03 Zn-65 C: 5.47E-03 3.11E-03 As-76 C: 9.68E-05 3.39E-05 Sr-89 C; , 3.61E-04 2.25E-04 Sr-91 C; 5.89E-05 0.00E+00 Sr-92 C; 7.30E-06 0.00E400 Y-91m C: 6.34E-05 0.00E+00 Nb-95 C; 0.00E+00 1.74E-06 Nb"97 Ch 8.85E-05 0.00E+00 Mo-99 C: 4.48E-OS 0.00E+00 Tc-99m C: 1.820-04 9.57E-05 Sb-125 C: 3.43E-06 0.00E+00 I-131 C: 2.11E-03 1.54E-03 I-132 C: 5.4tE-04 2.05E-06 I-133 C; 6.54E-03 1.32E-03 r~' I-134 I-135 C;
C; 4.61E-05 4.72E-03 2.82E-04 1.09E-04
\
Cs-134 C: 2.41E-03 1.74E-03 Cs-137 C; 9.36E-03 7.72E-03 Ba-140 Ch 3.72E-05 0.00E+00 La-14 0 C: 1.2SE-05 1.01E-04 Ce-141 C: 0.00E+00 6.36E-06 Ce-144 C; 1.13E-05 0.00E+00 Np-239 C; 7.42E-04 2.65E-04 Total C; 3.82E-0? 2.08E-02 Dissolved and entrained gases Xe-133m C 0.00E+00 3.47E-06 Xe-133 C: 2.12E-03 8.62E-04 Xe-135m C: 7.87E-03 2.33E-04 Xe-135 C: 1.08E-02 2.94E-03 Ar-41 C: 8.05E-07 0.00E400 Total C;, 2.07E-02 4.04E-03 Gr-Alpha Ci 0.00E+00 0.00E+00 1
- Zeroes i t. tis table pndicate that no radioactivity was I resent a ove detectab..e levels Sowerlimtsofdetect;,onforliquidsamp,,eanalyses.ypical See Table 1-5 for t '
l
- There are no continuous' modo radioactive liquid release pathways at Plant Hatch.
- \
s)m !
l
TABLE 1-3be E. I HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990 LIQUID EFFLUENTS
[h
\_J Continuous Mode **
Batch Mode lides eased Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 H-3 C1 2.06t+00 1.62E+00 rission and activation products:
Na-24 C: 3.26E-04 3.93E-04 Cr- 1 C: 1.74E-05 5.00E-05 Mn g4 s C: 4.92E-0$ 1.90E-05 Mn-56 C; 5.74E-07 7.51E-07 Co-58 C; 2.16E-05 2.73E-05 Co-60 C; , 3.00E-04 1.83E-04 Zn-(5 Ca 2.25E-04 1.91E-04 As-76 C: 1.77E-05 2.22E-05 Y-91m C: 9.17E-06 9.57E-06 Nb-95 C: 1.43E-06 0.00E+00 Hb-97 C: 2.40E-05 6.38E-04 Tc-99m C: 2.60E-05 1.27E-05 I-131 C; 2.36E-04 8.780-05 I-132 C: 1.01E-04 0.00E+00 1-133 C; 1.21E-03 2.05E-04 I-134 C: 8.75E-07 1.93E-06 I-135 Ch 5.75E-04 0.00E+00 Cs-134 C: 4.30E-04 2.74E-03 Cs-137 C: 1.44E-03 1.14E-02 Cs-138 C: 0.00E+00 1.12E-05 Total C; 5.01E-03 1.60E-02
(~N Dissolved and entrained gases:
s) Kr-85 C: , 6.51E-05 0.00E+00 Xe-133 C: , 2.61E-04 1.46E-04 Xe-135m C: 9.69E-04 1.09E-04 Xe-135 C; 2.24E-03 3.46E-04 Total C: 3.53E-03 6.01E-04 Gr-Alpha ci 0.00E+00 0.00E+00
- Zeroes i this table that no rapioactivity was resent aaove power ts ofdetectabnnd%
detect:,onefor cat aeve 11gugd s ee Tab _e 1-5 samp:,e analysesfor typical reJherearenocontinuousmoderadioactiveliquid case pathwayn at Plant Hatch O
LJ
_ 3 1. _
TABLE 1-3c0 E. I. HATCH NUCLEAR PLANT - SITE SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990 LIQUID EFFLUENTS O Continuous Mode **
Batch Mode Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 H-3 Ci 6.44E+00 5.12E+00 Fission and activation products:
Na-24 C: 1.07E-03 1.06E-03 Cr-51 C: 5.17E-05 3.73E-04 Mn-54 C: 6.4BE-04 3.75E-04 Mn-56 C: 2.79E-05 7.51E-07 Fe-59 C; ,
1.02E-05 1.13E-05 Co-58 C: 1.24E-04 3.69E-04 Co-60 C 4.11E-03 2.72E-03 Zn-65 C; 5.70E-03 3.30E-03 As-76 C: 1.14E-04 5.60E-05 l Sr-89 C: 3.61E-04 2.25E-04 Sr-91 C: 5.89E-05 0.00E+00 Sr-92 C: , 7.30E-06 0.00E+00 Y-91m C: 7.26E-05 9.57E-06 Nb-95 C: 1.43E-06 1.74E-06 Hb-97 C: ,
1.12E-04 6.38E-04 Mo-99 C 4.48E-05 0.00E+00 Tc-99m C: 2.08E-04 1.08E-04 Sb-125 C:, 3.43E-06 0.00E+00 1-131 C: , 2.35E-03 1.63E-03 1-132 C: 6.46E-04 2.05E-06 I-133 C: 7.75E-03 1.52E-03 I-334 C: 4.70E-05 2.83E-04 I-135 C; 5.29E-03 1.09E-04 Cs-134 C: 2.84E-03 4.48E-03
(
Cs-137 C: 1.00E-02 1.91E-02 Cs-138 C: 0.00E+00 1.12E-05 Ba-140 C: 3.72E-05 0.00E+00 La-140 C:- 1.25E-05 1.01E-04 Ce-141 C: 0.00E+00 6.36E-06 Ce-144 C: 1.13E-05 0.00E+00 Np-239 C: 7.42E-04 2.65E-04 Total C: 4.33E-02 3.67E-02 Dissolved and entrained gases:
Kr-85 C 6.51E-05 0.00E+00 Xe-133m C 0.00E+00 3.47E-06 Xe-133 C 2.38E-03 1.01E-03 Xe-135m C: 8.84E-03 3.4?E-04 Xe-135 C: ,
1.30E-02 3.28E-03 Ar-41 C 8.05E-07 0.00E+00 Total C 2.43E-02 4.64E-03 Gr-Alpha Ci 0.00E+00 0.00E+00
- Zeroes i this table indicat that no radioactivity was resent a ove de See Table 1-5 for t Sowerlimts.of3ectablelovesetection for 11guid sample analyses.ypical reJherearenocontinuousmoderadioactiveliquid ease pathways at Plant Hatch O
1 i
TABLE 1-4a '
/' ' E. I. HATCH NUCLEAR PLANT - Unit 1
(' SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 1940 INDIVIDUAL DOSES DUE TO LIQUID RELEASES I
Cumulative Doses per Quarter Organ Tech Unit Quarter 3 % of Quarter 4 % of Tech S{eg Tech
$m$t L$mSt Bone 5.0 mrem 4.18E-02 8.37E-01 2.95E-02 5.89E-01 Liver 5.0 mrem 6.75E-02 1.35E+00 4.52E-02 9.04E-01 TBody 1.5 mrem 4.70E-02 3.13E+00 3.12E-02 2.08E+00 Thyroid 5.0 mrem 5.66E-02 1.13E+00 8.86E-03 1.77E-01 Kianey 5.0 mrem 2.47E-02 4.94E-01 1.55E-02 3.10E-01 Lung 5.0 mrem 8.64E-03 1.73E-01 4.95E-03 9 00E-02 GILLI 5.0 mrem 7.30E-03 1.46E-01 2.78E-03 5.55E-02 Cumulative Doses This Year Organ Tech Unit uarters % of Spec ,2,3,& 4 Tech L1mit Spec Limit r'~ Bone 10.0 mrem 2.54E-01 2.54E+00
( Liver 10.0 mrem 4.30E-01 4.30E+00 TDodv 3.0 mrem 2.99E-01 9.97E+00 Thyrbid 10.0 mrem 1.36E-01 1.36E+00 Kianey 10.0 mrem 1.60E-01 1.60E+00 Lung 10.0 mrem 5.26E-02 5.26E-01 GILLI 10.0 mrem 5.92E-02 5.92E-01 O
- . _ _ . . - _ . _ . _ _ . . .__ __._._ _- ~ . _ _ . . _ _
i t
TABLE 1-4b E. I HATCH NUC J.AR PLANT - Unit 2 SEMIANNUAL RADIOACTIVE EULUENT RELEASE REPORT 1990 INDIVIDUAL DOSES DU3 TO LIQUID RELEASES ES=21*!h!.9o!!!.E!E.92!E!!E. ................. ............. :
Organ Tech Unit Quarter 3 % of Quarter 4 % of [
Sg Tech Tech -
b$m$t b$mSt B no 5.0 arem - 1.68E-0 5.10E-02 1.02E+00
&W Tvrbid 1:8 5.0 in:"
mram I.38 -03:f8 1
9:!!:81:8}1 1:*8E:81 1:ttE:88 2.44 6.79E-04 36E-K aney 5.0 mrem 7 .092-03 22E-02 1.42 -01 2.622-02 1
- 5. 24t 02- 01 Luno 5.0 mrem 3.93E-03 7.85E-02 8.70E-03 1.74E-01 i GILLI
- 5. 0 - -mrom- 3.10E-03 6.20E-02 1.64E-03 3.27E-02 Cumulative
..... ... .. ..Doses ..This Year..... . ......... ... ..... . ..... ....
Organ- Tech- Unit uarters % of .
Spec ,2,3,& 4 Tech
- L1mit Spec >
Limit ,
Bone 10.0 mrom 8.52E-02 8.52E-01
.\ Liver 10.0 mrom 1.382-01 1.38E+00 !
.TBody Thyroid 3.0 mram 9.75E-02 3.25E+00 '
10.0 mram 2.71E-02 2.71E-01 Ki8ney 10.0- arem S.04E-02 Luna 10.0 mrom 1.96E-02 5 96E-0104E-01 1.
GILLI
........n 10.0 mram 1.46E-02 1.46E-01 s
.2 O
, . . . ,.-.u. .- -- ..-,..--..a;.-...-_--.-.--...-..-_ . . - . _ . - . - - - , - .
TABLE 1-5 LOWER LIMITS O' DE'TEct10N .- LIOUID SAMPLE ANALYSES
\h The values in this table represent aorlori lower limits of detection (LLD) which are typically achieved in laboratory analyses of 11guld radwaste samples.
RADICtFJCLIDE LLD UNITS Mn-54 5.38E-08 ucl/ml Fe-59 7.78E-08 Co-58 4.67E-08 Co-60 4.78E-08 n-65 1.31E-07 Mo-99 5.10E-07*
Cs-134 7.18E-08 Cs-137 6.05E-08 Ce-141 1.41E-07 Ce-144 6.30E-07*
I-131 6.51E-08 X e-135 8.45E-08 Fe-55 2.00E-06 St-89 5.00E-08 St-90 5.00E-08 !
H-3 1.30E-05 ,
g 'In accordance with Technical Specification Tables 4.15.2-1 (Unit 1) and )
't 4.11.1-1 (Unit 2), Table Notation b, the permissible Lower Limit of V) i Detection my be increased inversely proportional to the monitude of the ganrna yield. However, the LLD determined in this m nner must not exceed 10 percent of the Maximum Permissible Concentrattors (MPC) value spectfled in 10CFR20, Appendix B, Table II (Column 2) .
G
-1e-
h 2 - GASEOUS EFFLUENTS 2.1 REGULA'IORY LIMITS The Technical Specifications presented in this section ,
are for Unit 1. Requirements for Unit 2 are the same :
as for Unit 11 however, the Technical Soecification numbers are not the same.
TECHNICAL SPECIFICATIONS 3.14.2 The radmactive caseous effluent monitoring instrumentation channels shown in table 3.14.2-1 shall be <
OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.15.2.1(a) are not exceeded.
The alarWtrip setpoints of these channels shall be determined in accordance with the ODCM. Technical Specification Table 3.14.2-1 is included in this section as Table 2- 1. )
3.15.2.1 The dose rate at any time in the UtaESTRICTED AREAS (figure 3.15-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the follcuing values:
- a. The dose rate limit for noblo gases shall be less than or equal to 500 mrem / year to the total body and less than or equal to 3000 mrem / year to the skin.
- b. The dose rate limit for I-131, I-133, tritium, and for all radioactive materials in particulate form and radio-oc11 des other than noble cases w2th half-lives greater than 8 days shall be less than or equal to 1500 mrem / year to any organ.
3.15.2.2 The air dose in UtBES'IRICTED 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 lese than or equal to 10 mrad for beta radiation.
l-
- b. During any calendar year, to less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
I' 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 form with half-lives greater than 8 days in gaseous effluents released to UtaESTRICTED AREAS (figure 3.15-1) from each reactor unit shall be limited to the following:
o 19 l
- a. During any calendar quarter to less than or equal to 7.5 mrem to any oronn.
() b. During any calendar year to less than or equal to 15 mrem to any organ. .
3.15.2.4 The GASDNS RADWASTE TREATMETR SYSTEM as described in the ODCM shall be in operation. (This Technical Specification applies whenever the main condenser air ejector system is in operation.)
4.15.2.4 GASDIS RADWASTE TREATME!C SYSTEM operability shall be demonstrated by administrative controls which assure that the offgas treatment system is not bypassed.
3.15.2.5 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle soitrees shall be limited to less than or equal to 25 mrem to the total bod / or any organ, except the thyrosd, which shall be limited to less than or equal to 75 neem.
(With the calculated doses from the release of radioactive materials in 11guld or caseous effluents exceeding twice the limits of Specifications 3.15.1.2(a), 3.15.1.2(b),
l'_s). 3.15.2.2.(c), 3.15 . 2. 2 ( b ) , 3.15.2. 3( a ), or
's_/ 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 r.aln condenser offgas treatment system shall be limited to less than or equal to 4 percent by volume.
3.15.2.7 The gross ganna radioacti. . rate of the noble gases Xe-133, Xe-135, Xe-138, Kr-85m, Kr-87, ani Kr-88 neasured at the main condenser evacuation system pretreatrent monitor station shall be limited to less than or equal to 240,000 uC1/second.
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 caseous and 11guld effluents that were in excess of 1 C1, excluding dissolved and entrained gases and tritium for O 11guld effluents, or those in excess of 150 C1 of noble gases or 0.02 C1 of radiolodines for caseous releases."
= ~ a . . .
.,m
,. .g g
< - r( j) _-
t%)e .
\
1Antt 3.14.z-t ( Si!I E I I of bl
- MAR!OMIlyJ_cAstous Linute!, nore t ton , no Inst RUMf NI AMOff O Minimum Charine I s-applicabiIiey Pgame te r AC) 10^4 e
t'Ai r ingnt OPE kAelt 1 M. tin Cueusesiser OtTges T reatment System l'xplosive Cas Monitoring System layd rogeti Honitor- (1) ** *. Ifydrogen 106
- 2. hector liteilding Ve sack Monitoring System
- a. Note s e Ga s Ac t i v i ty Monitor {1)
- Hadioactivity Hate 10'p
, tn Mea sterement +
td .
tn
- b. todinc Sampler Cartridge (1)
- Verify Presence or 10T 8
Cartridge y
{_l p a
Par t senlate Sampler filter (1)
- Verif Presence or 107 c.
i
' Filter
- d. Iffluent System flowsate
- System I tovrate f ois ,
Measiesement Device (1)
Measterement
- e. Sampler flowrate Heasesrement
- Samp f er i f owrate T O's Device (I) taeasurement
- 3. Recombisn'r Stei tdistg Verit s iatiors Hon i tori swJ System a, Noble Gas Activity Monitor (1)
- Rad, -ttivity Rate IOS
.Measniement +
Sixtine Sampler cartridge (1)
- Verirv Fresence or tot b.
ca rt r i4]r' Pa rt acasla to Sis *Ap l e #~ l i l t e s' (8) Vecify Presence of 101 C.
- Vilter S.emp l e r i i ow rit te Mea s.t6 s tre ess t (1)
- Samp t e r i inv ra te IO's d.
()ev a a et 54easurement * .
e k,
= b O ___ 5 5 _
.c*'%
_ ts :.
,.-~. k.
. v I ABI i 3. Iis.2- 1 (SIILET 2.OF 4) f(AlliOACTIVf CAS[Opi III(U[ D gfgIORING t GIRUME N T A T iOff Minimiem Cisa nvie i s .
' trJsirassen[ OP[RABIf ' App tj cabi lily Pa ramete r ACTION *
- 4. M_ein Stack Monitoring System
- a. Noble Gas Activity Monitor. (1) Mdienctivity Rate -105
% ses6ement +
- b. todine Sampler Cartri(Ige (1)
- Verify Presence tf' 107 Cartridge .
Particulate Sampler Filter
- Verify Presence of 107
- c. '(1) 3 Fiiter >
N
- d. Errtuent System i towrate in Meastaring Devites (1)- System flowrote 108 y Mea ses rement s Sampler i Iowra te Heasaaring
- Sampler T f owra te 102: m
(( c.
Devece
-(1)
Mensterement n o
- 5. Curidereser crrgas Pret reatment M. 3 Monitor ,~
No!> te Gas Act ivi ty Hossi tor ==. Radioactivity Rate 108
{1)
Mea ster emen t
.. c. . . . . .. . .
. .n. ..
. . . , . . .n_. . . . , , _ ..
TABLE 2-1 (CONT'r)
TW L)
TABLE 3.14.2-1 (SHEET 3 0F 4)
RADICACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Table Notations U
- Monitor must be capable of responding to a Lower Limit of Detection of 1 x 10-' uCi/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 patnway may continue, provided the flowrate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
If 3.he number of channels OPERABLE reme. ins 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 105 - With tne 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 ar.d analyzed daily for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With the number of main stack monitoring system channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, without delay-suspend drywell purge.
If the number of c' c is OPERABLE remains less than required by the M :w . Channels OPERABLE requirement for over 30 days, an Lapianation of the circumstances shall be included in-the next semi-annual effluent release report.
ACTION 106 - With the numoer of channels OPERABLE less than required:by the Minimum Channels OPERABLE requirement, operation of.the main condenser offgas treatment system may continue _provided:
(a) Gas samples are collected once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and analyzed within the ensuing 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or f- s (b) Using a temporary hydrogen analyzer installed in the
_( ) offgas system line cowns,trean of the recomoiner, hydrogen concentration readings are taken and logged -
every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
TA II 0-1 (conh'L) l c_
, e-'
N ,)s TABLE 3.14.2-1 (SHEET 4 0F 4),
RADICACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION l'
Table Notations (Continued)
I 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 effluer,t release report.
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 sempling equipment for periods on the order of 7 days and analyzed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 circun, stances shall be l included in the next semi-annual effluent release recort.
i T'i ACTION 108 - With the number of channels OPERABLE less than
- ( ,/
' required by the Minimum Channels OPERAB.E recuirement, release to the environment may continue for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided:
- a. The offgas system is not bypassed, and
- b. The offgas post-treatment monitor (011-K615) or the main stack monitor (D11-K600) is OPERABLE.
Otherwise, be in at least HOT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
.If the number of channels OPERABLP remains less than required by the Minimum Channels OPERABLE requirement for E
i over 30 days, an explanation of the circumstances shall be included in the next semi-annual effluent release report.
L:
l' M
s) 2h.
I
, t J
1
' j J
, "e .2.2-.- MEASUREMENP AND APPROXIMATIONS OF KyrAL RADIOACTIVITY
~ Waste gas release-at Plant Hatch is confined to ,
4 - four paths: main stack (also called the offgas '
( vent)~, Unit 1; reactor building vent: Unit 2 reactor building vent, and the recombiner- i building vent. -Each of these four paths is continuously monitored for gaseous radioactivity. Each is equipped w; W an integrating-type samle' collection device for collecting particulates and lodines. Samole collection-is in'accordance with Technical '!
Specification Tables 4.15.2-1 (Unit 1) and
.[
4.11.2-1.(Unit 2). .Unless required more g frequently under-certain circumstances specified in Table Notations to the above mentioned tables, samples are collected as follows:,
m 4'
- 1. Noble gas samles are collected by Arab sampling monthly.
- 2. Tritium.samles are collected by grab sampling monthly.
- 3. Radioiodine samles are collected by pulling the sample stream through a charcoal :
v ;tridge over a 7-day petlod. ,
4.- 1Partic'21r?,W are' collected by pulling the' :i t_ sample stream through a particulate filter =
b ' '
. -over a_.7-day period. .[
a 5 ~- . The:7-day particulate; filters above are
'X : analyzed for gross , alpha act2vaty.. l y
4-
.6. Quarterly composite samples are prepared- a from the particulate filters collected-over-1 f
, the previous quarter and the quarterly .
-composite sample'is analyzed for Sr-89 and-ay Sr-90.
= Sample analyses ~results and release flow rates:
-from the four: release points. form the. basis for- .
calculating released quantitles of- 1' radionuclide-specific radioactivity, dose rates associated with gaseous releases, and cumulative ,
c '. Jdoses for. the current quarter and l year. This .
l
_ task is normally performed-with computer .a
- 1 assistance.
g if._
o e s+ - , as . h--r,-- .,,n, ,- .,,,,,,e.,-.-n--e-- , , , , , e s , ,n~--- ,--.n,.+- .---- -- a
The noble gas grab sample analysis results are used along with maximum expected release flow rates from each of the four vents to calculate f(Y monitor setpoints for the gaseous ef fluent monitors serving the four release points, to
(,,/
assure that the limits of Technical Specifications 3.15.2.1.a (Unit 1) or 3.11.2.1.a (Unit 2) are not exceeded. Calculation of monitor setpoints is described in the Plant Hatch ODCM.
With each release period released radioactivity, dose rates, and cumulative doses are calculated.
Cumulative dose results are tabulated along with percent of Technical Specification limits
( 3.15.2.2 and 3.15.2.3 (Unit 1): 3,11.2.2 and 3.11.2.3 (Unit 2) for each release, for the current quarter and year.
After each calendar quarter (13 weeks) a summary 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 quantitles of radioactivity, dose rates and cumulative doses are as o follows:
Lt 1. FISSION ANu ACTIVATION GAS The radionuclide-specific released radioactivity is determined from sample analyses results collected as described above and averaqe release flow rates over the period represented by the collected sample.
Instantaneous dose rates due to noble gases and due to radiolodines, tritium, and particulates are calculated (with computer assistance).
Calculated dose rates are compared to the dose rate limits specified in 3.15.2.1.a (Unit 1) and 3.11.2.1.a (Unit 2) for noble gasest and 3.15.2.1.b (Unit 1) and 3.11.2.1.b (Unit 2) for radiolodine, tritium, and particulates. Dose rate calculation methodology is presented in the Plant Hatch 00CM.
p.
v
- . . .. . . - -_ - . - . - . . ~. . . - . .
's y if
' Beta- and gamma air doses due to noble gases are icalculated for the location in the unrestricted area with the potential ~for~the highest exposure :r due to gaseous releases.= Air doses are
'A calculated for.each release period and cumulative-
--totals are kept for,each unit for the current calendar quarter and year. Cumulative air doses are-compared to the dose limits spectfled In ,
Technical. Specifications 3.15.2.2 (Unit it ard i 3.11.2.2 (Unit 21._ Current percent of techn1;al specification limits'are shown on the printout for each release period. Air dose calculation methodology Is presented in the Plant Hatch ODCM.
1
- 2. RADIOIODINE, TRITIUM, AND PARTICULATE RELEASES Released quantitlesTof radiolodines are determined from the weekly samples and release flow rates for the four release points.
" Radioiodine concentrations are determined by gama spectroscopy.
Released quantities of particulates are -
e determined from the weekly (filter) samples and release flow rates for the four release points.
Gama spectroscopy is used to quantify.
concentrations of principal gamma emitters, a After each calendar' quarter the particulate ,
y = filters lfrom_each vent are combined, fused, and ;
- f strontium separation.1s performed. Since sample 39 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 B
collection period. Where significant.St-89 or.
Sr-90.is'not-detected, LLD's are calculated. q Strontium concentrations are input-to the '
composite file of the computer to be-used in j release,. dose rate and individual dose calculations.
Tritium samples-'are obtained montnly from each
% vent by passing:the sample stream through a_ cold i
[i ' ~ trap. The grams.of water vapor / cubic foot gas is measured upstream of the-cold trap in order to-U alleviate'the difficulties in determining water vapor collection efficiencies.' The tritium W samples are analyzed by an independent laboratory j '
and results are furnished in uC1/ml of water.
The tritium' concentration' in. Water is converted to tritium concentration in air.and this value is input into the composite file of the computer-to be used in. release, dose rate, and individual i (O -
dose calculations.
f %.)
1 1 4 . . . . _ ~ . . ~ _ _ -. _ _ . _ _ _ ._
. . . - ~ . . - - ~ - . . - - - - _ - - - - . - - . , - - - - - .
5iC
.--n Dose rates due'to radioiodine, tritium, and
- particulates are calculated for a hypothetical >
b]?
i
< child, exposed.to the inhalation oathway, at.the~ - ' t location in the unrestricted-area where the-potential-dose rate is expected to be the highest; Dose rates are calculated for each release point, for each release period, and the'
?> ,
total dose rate.from all four release points are-compared.to the dose rate limits specified in.
Technical Specifications 3.15.2.1.b (Unit 1) or 3.11.2.1.b-(Unit 21.
-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 cumulative totals are kept for each unit-for-the current: calendar quarter and year. Cumulative' Individual 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-1 1 for each release period.
h J3. GROSS ALPHA RELEASE i The gross alpha release is computed each month by J : counting the particulate filters each week forL j '
$: gross; alpha activity in a scintillation counter.
'L The four.or five weeks' numbers are then recorded on a data sheet and the activity:is summed'at'the end of the month.' - This concentration ~is ' input. to
-the composite file of the computer and'is used .
for-release calculations. [
4.. ERROR-ESTIMATES G Reculatory Guide 1.21 r,equires that estimated j'
,m -total; error =in. analysis techniques be reported.
These-estimates are required for the total N fission-andiactivation'qas release, total I-131
-release,! total'particulates With half-lives f
. greater than 8-day release, and ~ total tritium -
. release. s 1 1 4
U. - --
i
. , _ . _ . -_ _ _ , , . . .a _ , -,_._,.-._a
s "The total or maximum error associated with the effluent measurement will include the cumulative errors resultinq from the total operation of
(~) sampling and measurement. Because it nay be very
\/ difficult to assion error terms for each parameter affecting the final measurement, detailed statistical-evaluation of error are not suggested. The objective should be to obtain an overall estimace of the error associated with measurements of radioactive materials released in 11guld and gaseous effluents and solid waste."
Estinated errors are based on errors in countinq equipment calibration, counting statistics, vent flow rates, vent sample flow rates, non-steady release rates, chemical yield factors, and sample losses for such items as charcoal cartridaes.
(1) F1ssion and Activation Total Release was calculated from sample analysis results and release point flow rates.
Statistical Error 60%
Countina Equinient ca1ibration 104-Vent Flow Rates 10%
Non-Steady Release Rates 20%
100%
(2) I-131 Release was calculated from each weekly
( sample:
v Statistical Error 60%
Counting Equipment Calibration 10%
Vent Flow Rates 10%
Vent Sample Flow Rates 10%
Non-Steady Release Rates 10%
Losses From Charcoal Cartridge 10%
110%
(3) Particulates with half-lives areater than 8 days release was calculated from sample analysis results and release point flow rates, Statistical Error at LLD concentration 60%
Counting Equipment Calibration 10%
Vent Flow Rates '10%
Vent Sample Flow Rates 10%
Non-Steady Release Rates 10%
100%
/~'
()) .
(4)..-Total Tritium Reledse wis dominated by the .
reactor buildinq vent tritium releases hence, the a
' larger statistical errors of the off-gas vent and
~
4
, recombiner building vent tritium releases do not affect the error in the total tritium release:
Water Vapor in Sanole Stream Determination 20%
l Vent Plow Rates 10% l Counting Calibration and Statistics Non-Steady Release 10% l 50%
WI 2.3 : GASEOUS EFFLUENT RELEASE DATA Regulatory Guide 1.21-Tables lA, la, and 1C are found '
in this report as Tables 2-2a-c, 2-3a-c, and 2-4a-c.
Data are presented on a quarterly basis as required by i Regulatory Guide 1.21.
- To complete Tables 2-2a-c, total release for each of the four categories (fission and activation gasest iodinesr particulated; and tritium) was divided by the number of seconds in the quarter to obtain a release :
ratefan uC1/second for-each category.- i However, the applicable -Technical Specification limits - ~f are not<1n. terms'of release rate in uct/second but in !
4 terms of.: dose rate in mreW year, as presented in 1
-Technical Specifications 3.15.2.1 (Unit.1) and 3.11.2.1.(Unit.2);,! Noble gases are, limited as
~
specified int 3.15.2.1 a- and 3.11.2.1.a. The other
, three categories (tritium, radiolodines, and.
particulates) are~11mited as:a group as specified in
- 3.15.2;1.b and 3.11.2.1.b. Further-the limits ,
^
ispecified in Technical Spacifications 3.15.2.1 and
=
3.11'.2.1 are site limits, not unit limits. ' Dose rates duelto noble gas releases and due to radiciodine,- :
- tritium, and particulates are-presented in-Table 2-5 4
along with percent of' technical-specification limits.--
~ Gross alpha radioactivity is: reported in Tables 2-2a, 2b,.and 2-2c as curies released in each quarter. -
Limits for cumulative beta anti gamma air : doses, due to h noble gases, - are -specified .in Technical. Specifications '
- 3.1512.2 (Unit 1) and 3.11.2.2 (Unit-2). lThese limits !
are unit limits. Cumulative air. doses are presented
!in Tables-'2-6a_and 2-6b, along with percent of technical specification limits.
f
.'--c.' -
- * -w--- . , + ..,.,--,i-:r, ,me , . = , , ,+- y ,,-y, wm-- y +,-a
5
- 1 Limits for cumulative individual doses, due to radiolodine, tritium, and particulates, are specified in Technical Specifications 3.15.2.3 (Unit 1) and 9e s
) 3.11.2.3 (Unit 2). These limits are also unit li mits. Cumulative individual doses are presented in Tables 2-7a and 2-7b, with percent of technical specification limits.
2.4 - RADIOLOGICAL IMPACT DUE '!O GASEDUS RELEASES Dose rates due to noble gas releases were calculated for the site in accordance with Technical Specifications 3/4.15.2.1.a (Unit 1) and 3/4.11.2.1.a (Unit 2). Results are presented in Table 2-5. Dose rates due to radiolodine, 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.
Cumulative 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 Tables 2-6a and 2-6b.
Cumulative doses to an Individual due to radiolodine,
(~'y tritium , and particulates were calculated for each
\_,/ unit in accordance with Technical Specifications 3/4.15.2.3 (Unit 1) and 3/4.11.2.3 (Unit 2). These results are presented in Tables 2-7a and 2-7b.
Dose rates and' doses were calculated using the methodology presented in the Plant Hatch Offsite Dose Calculation Manual, 1
0
. _ , . .. . _ .. _ _ ._ . . . _ . ~. . . _ _
u TABLE 2-2a E. I. HATCH NUCLEAR PLANT - UNIT l' rx SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT :1990 SUMMATION OF ALL RELEASES
-:(] GASEOUS EFFLUENTS Unit Quarter-3 Quarter 4 Est. Total Error (%) 4-A. Fis ion an' Actfvation
, Gases
- 1. Total Ci 5.03E+01 5.07E+01 1.00E+02 Release
- 2. Average uC1/sec 6.40E+00 6.52E+00 Release Rate For Period
- 3. % of Tech- %
- Spec Limit B. Iodines
- 1. Total Ci 1.83E-04 8.02E-04 1.10E+02 Iodine-131
- 2. Average Release--
uCi/sec '2.32E-05 1.03E-04 Rate For.
P
- 3.--%eriod of Tech' %
Spec Limit C. Particu:.ates
- 1. Part;,culates Ci 2.56E-04 2.'35E-04 1.00E+02
( w;.th half-
- 1.ves > 8 days
- 2. Avera
'Releagee- uci/sec 3.26E-05 3.02E-05 Rate For Period
- 3. % of-Tech % i Spec Limit
.4. Gross Alpha C11 5.61E-06 3.99E-06 Radioactivity t,
-D.- Tritium.
' 1. Total- . Ci 3.81E+00 2.84E+00 9.00E+01 2.' Average Release uCi/sec 4.85E-01 3.66E-01 .;
Rate.For:
Period-
- 3. % of Techt 1%
Spec Limit t
'
- Technical Specification limits are in terms--of dose-rate
_." farem/yrg~anBdose
-7a, an . 2-7b (mrem).- See. Tables 2-5, 2-6a,-2-6b, O
-)
l
-TABLE 2-2b: l E. I. HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990 GASEOUS: EFFLUENTJ - SUMMATION OF ALL RELEASES
~O3 Unih Quarter 3 Quarter 4 Est. Total l Error (%) l 1
i A. ion and
'Fis$vation Act Gases
- 1. Total Ci 3.17E+01 4.24E+01 1.00E+02 Release
- 2. Average Release uCi/sec 4.03E+00 5.46E+00 Rate for Period
- 3. % of Tech %
Spec Limit B. _ Iodines I
- 1. Total Ci. 1.82E-04 1.32E-03 1.10E+02
- 2. Average Release-uCi/sec 2.32E-05 1.69E-04 Rate-For
- Period
-*3.-% of Toch %
-Spec Limit 1
( -C. Particulates
- 1. Particulates- Ci 1.40E-04 1.69E-04 1.00E+02 LN U:.th half -
1.ves > 8 days 2.. Average Release-uCi/sec 1.78E-05 2.17E-05 Rate For Period .
- 3.c4 of Tech %
Spec Limit ~
4.-Gross Alpha Ci 2.91E-06 3.42E-06 Radioactivity _
D.- Tritium
- 1. Total Ci 5.68E+00= 7.24E+00 9.00E+01
- 2. Average uC1/sec 7.22E-01 9.31E-01
-Release
, ' Rate For Period
- 3 '. :- % of Tech %
Spec Limit
- Technical Specification limits are in terms-of dose rate fmrem/yrg-anddose(mrem). See Pables_2-5, 2-6a, 2-6b,-
-7a, an 2-7b
,O .
u w
(l'
TABLE 2-2c
~E. I. KATCH NUCLEAR PLANT - SITE sf, SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990 +
3 GASEOUS EFFLUENTS - SUMMATION-OF ALL RELEASES Unit Quarter 3 Quarter 4 Est. Total Error (%)
A. Fis ion and Act$vation.
Gases
- 1. Total Ci 8.20E+01 9.32E+01 1.00E+02 Release
- 2. Average uCi/sec 1.04E+01 1.20E+01 Release Rate For Period
- 3. 4 of Tech %
Spec Limit B. Iodines
- 1. Total Ci 3.65E-04 2.12E-03 1.10E+02 Iodine-131
- 2. Average Release uCi/sec 4.64E-05 2.72E-04 Rate For-Period
- 3. % of Tech - %
Spec Limit
-C. Particu:.ates
- 1. Part;.culates Ci 3.96E-04 4.03E-04 1.00E+02 Oi w .th half-1:.ves > 8 days
- 2. Average Release uCi/sec 5.04E-05 5.19E-05
. Rate-For Period
. 3. % of Tech- %-
Spec Limit-
- 4. Gross Aloha- Ci 8.52E-06 7.42E-06 Radioactivity .
D. Tritium 7
- 1. Total Qi 9.49E+00 11.01E+01 9.00E+01-2.? Average Release uC1/sec 1.21E+00 1.30E+00
= Rate For
. Period
- 3'.2% of Tech %
Spec Limit j 2* Technical Specification limits are in terms of deso rate (mrem 2-7a,/yr)-and
.and 2-7b. dose (mrem). See. Tables 2-5,-2-6a, 2-6b, 1
I l
-3L_
. ~ -
TABLE 2-3a ,1 E. I.-HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990
[ GASEOUS EFFLUENTS --ELEVATED. RELEASES
- _.......___............_............_............. Continuous Mode BatchMode **
Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4
- 1. Fission Gases Kr-85m C; 8.31E-02 0.00E+00 Kr C; 4.91E-01 6.38E-01 Xe-133 C: 5.56E-02 3.14E-01 Xe-135m- C: 4.49E&OO 5.81E+00 Xe-135 C: 8.22E-01 8.74E-01 Xe C: 1.68E+01 2.05E+01
__...._-138 _..........................................___........
f Total:For Period C1 2'.28E+01 2.81E+01
.-------...................---......__-03 I-135 C: 7.86E-04 4.42E Total For:
Period Ci l40E-03 8.25E-03
- 3. Particulates.
i
-Co-60 .(1 0.00E+00 ~1.13E Zn-65 (1 -0.00E+00 3.50E-07 Sr-89 C: 11.53E 1.65E-08
-Sr-90 (1 1.15E-07 -0.00E+00 .
Hb-95 C: 4.64E-08 0.00E+001 1
!Cs-137- C: 2.75E-07 1.19E-06 Ba-140 (1. -3.50E-05? 5.43E-05 La-140 C: :7.54E-05 8.25E-05 'i Ce-141.
........ _-08 9.61E -0.00E+00=
.___...........'C; 1
-Total For Period' -aCi 1.26E-04: 1.38E-04 m ___.. --_________.. _.....____.--.. ___-_____...____......... ,
- Zeroes in this table indicate that no radioactivity was ,
presen a ove detectab e levels. See Table 2 or ty '
lower im ts-of detect on'for gaseous sample ana_yses.pical
} ' ?. **There are no batch'uode radioactive gaseous release pathways at Plant Hatch.
(( f
_ _ _ _ . D
i TABLE 2-3b E. I. MATCH NUCLEAR PLANT - UNIT'2 SEMI ANNUAL RADIOACTIVE EFFLUENT RELEASE- REPORT- 1990 GASEOUS EFFLUENTS - - ELEVATED RELEASES
- Continuous Mode Batch Mode **
Nuclides Released- Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4
- l. Fission Gases Kr-85m C: 8.31E-02 0.00E+00 Kr-87 C: 4.91E-01 8. 8 2 E .Xe-133 C: 5.56E-02 4.34E-01 Xe-135m C: 4.49E+00 6.68E+00-Xe-135 - C: 8.22E-01 1.02E+00
..........__.._C:
Xe-138 1.68E+01 2.39E+01 Total For Period Ci- 2.28E+01 3.29E+01
- 2. Iodines.
I-131- C: -9.27E-05 1.25E-03 I-133 C; 5.23E-04 7.03E-03
______-135 I C: .7.86E 04 1.10E t ___________..._____ _ ________._-02 ___.________.______ ____.
Total For Period C1 1.40E 1.93E 02
-._.........__________.._-03 ___________ ....________________....
- 3. Particulates Co-60 C:. 0.00E+00. -3.40E-07
--Zn-65 C;. 0.00E+00' 5.91E-07 Sr-891 C: 1.53E-05 1.85E-08 I Sr-90 C: 1.15E-07 0. 00E+00 : '
.Nb-95 C: 4.64E-08 0.00E+00 Cs-137 C: 2.75E-07 1.25E-06 Ba-140 C: 3.50E-05 6.10E ,
-La-140. C:, . 7.54E-05 9.42E-05 I Ce 141 C;, 9.61E 0.00E+00 '
-Total ^For Period lCi 1.26E-04 11.57E-04
- Zeroes-in this-table indicate that no radioactivity was present above detectable levels. See Table 2-8 for ty lower limits of detection for gaseous sample analyses.pical
-**There are no batch mode radioactive gaseous release q pathways at Plant Hatch. .
1
-36
TABLE 2-3c E. I. HATCH NUCLEAR PLANT - SITE
^
s SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990 i
- GASEOUS EFFLUENTS - ELEVATED RELEASES
- L)
Continuous Mode Batch Mode **
Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4
- 1. Fission Gases Kr-85m C: 1.66E-01 0.00E+00 Kr-87 C: 9.81E-01 1.52E+00 Xe-133 C: 1.11E-01 7.48E-01 Xe-135m C; 8.99E+00 1.25E+01 Xe-135 C: 1.64E+00 1.89E+00 Xe-138 C; 3.36E+01 4.44E+01 Total For Period Ci 4.55E+01 6.10E+01
-s I-133 C: 1.05E-03 1.01E-02
( ) I-135 C: 1.57E-03 1.55E-02 q j - - - - - - - . . . . . . . - _ . - - - . . . . . - - - - - - - . . . . - - - - . . . . . . - - - - - . . . . - -
Total For Period C1 2.80E-03 2.76E-02
- 3. Particulates Co-60 C; 0.00E+00 4.54E-07 Zn-65 C:. 0.00E+00 9.41E-07 Sr-89 C: - 3.05E-05 3.50E-08 Sr-90 C: 2.31E-07 0.00E+00 Nb-95 C: 9.28E-08 0.00E+00 Cs-137 C: 5.51E-07 2.44E-06 Ba-14G C: 6.99E-05 1.15E-04 La-140 C: 1.51E-04 1.77E-04 Ce-141 C: 1.92E-07 0.00E+00 Total For Period C1 2.52E-04 2.96E-04
- Zeroes in this table indicate that no radioactivity was Dresent above detectable levels. See Table 2-8 for ty lower limits of detection for gaseous sample analyses.pical n
\
/
- There are no batch mode radioactive gaseous release
(_,) pathways at Plant Hatch.
_ ~ . ._ - _ _ . _ _ _.. _ . , _ . .. _ _ _ . . . _ . _ . _ . . . _ _
TABLE 2-4a - - -
E. I. RATCH NUCLEAR PLANT - UNIT 1 D + SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990
'(J GASEOUSf EFFLUENTS GROUND-LEVEL RELEASES
- Continuous Mode Batch Mode **
. Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 1.
___ Fission Gases Kr-85m- Cj - 1.31E+00 0.00E+00 Kr-87 C: 1.29E-02 0.00E+00 Xe-133 . C: 1.84E+00 2.26E+01 Xe-135m C: 4.60E-02 0.00E+00 Xe-135 C: 2.41E+01 7.45E-03 Xe-138 C: 1.84E-01 0.00E+00 t Total Fo.
Period Ci 2.75E+01 2.26E+01
- I-133- C; 6.62E-04 5.99E-04 I 135 C: 2.16E 0.00E+00 s ______ _ __._______________-04 ........_____.__........._________._
Total For-Period Ci 9.68E
______-04 6.85E 04 3..Particulates
~ Cr-51' C; 1.26E 0.00E+00 Mn-54 C: 7.56E-06 6.27E ;
Co-58 - C: 2.35E-06 2.79E-06 Co-60 C: 6.76E-05 .4.39E-05 Zn-65 C: . 1.27E-05 2.59E-05
-Sr-89 C; 7.86E 1.23E-05
- Sr-90 C: 0.00E+00 - 8.27E-09.
Cs-137 - C; 0.00E+00 1.10E-06 Ba-140 C: 1.05E-05 0.00E+00 La 140 ~ C; 3.70E
_______ _ ___________.'8.77E-06____.. __________-06 .. _________..._________
- Total-For-
- Period ______..C1 1.30E-04 9.61E 05
- Zeroesiin this-table indicate that no radioactivity was Dresent above detectable levels. See Table 2-8:for ty
' lower limits of detection for gaseous sample analyses.pical
' - () **There are no batch mode radioactive gaseous release Lpathways-at Plant Hatch.
36-9y ..py- p y , .--,.w-----myywwveyys.-g p-+..--mm,-
. w yy..-wwy
_ . . . _ . _ ._ ..- _. _ . __ _ _ _ _ _ _ ~ _ _ _
TABLE 2-4b E. I. IIATCH NUCLEAR PLANT - UNIT 2
- SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990
( GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES
- Continuous Mode
........... _________.......__________'.__________..____ Batch Mode ** _____
Nuclides-Released'_______..__....._-
___.-.... Unit Quarter 3 Quarter 4. Quarter 3 Quarter 4
- 1. Fission Gases
____.-_-135-Xe Ci 8.97t+00 9.51E+00 Total For Period Ci 8.97E+00 9.51E+00
--___..._________________-04 __________________________.......___
Total For Period Ci 8.32E 04 4.66E
_--.____-04
(~~
\
- 3. Particul'ates Mn-54' C: 3.36E-07 0.00E+00 Zn-65 C; 8.29E-07 0.00E+00' Sr C: 1.13E-05 1.12E Cs-137 C; 3.43E-07 2.55E-07 Ba 140 C: 1.07E-06 0.00E+00 Total For Period C1 1.39E 1.15E
_________________________ __.. ______-05 _________________........
- Zeroes in this. table indicate that no radioactivity was; present above-detectable levels.- See Table 2-8 for ty i
-lower limits of detection for gaseous sample analyses.p ca l-
- There are no batch mode radioactive gaseous release pathways at Plant Hatch.
i
~.
w 1
I
1:
TABLE 2-4c E. I. RATCH NUCLEAR PLANT - SITE
,s SEMIANRUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 1990 (v ) GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES
- Continuous Mode Batch Mode **
Nuclides Released Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4
- 1. Fission Gases Kr-85m C: 1.31E+00 0.00E+00 Kr-87 C: 1.29E-02 0.00E+00 Xe-133 C: 1.84E+00 2.26E+01 Xe-135m C: 4.60E-02 0.00E+00 Xe-135 C: 3.31E+01 9.52E+00 Xe 138 C: 1.84E-01 0.00E+00 Total For Period C1 3.65E+01 3.22E+01
/~N I 135 C: 4.63E-04 0.00E+00
\v ) ___.____..... _____..................._ ..____.......____._..
Total For Period Ci 1.80E-03 1.15E-03
- 3. Particulates Cr-51 C: 1.26E-05 0.00E+00 Mn-54 C: 7.94E-06 6.27E-06 Co-58 C: 2.35E-06 2.79E-06 Co-60 C 6.76E-05 4.39E-05 Zn-65 C: 1.35E-05 2.59E-05 Sr-89 C:. 1.92E 2.36E-05 Sr-90 C: 0.00E+00 8.27E-09 Cs-137 C: 3.43E-07 1.35E-06 Ba-140 C: 1.16E-05 0.00E+00 La-140 C: 8.77E-06 3.70E 06 1 For Tot!od Per Ci 1.44E-04
_________..___________-04 1.08E
- Zeroes in this table indicate that no radioactivity was present above detectable levels. See Table 2-8 for ty lower limits of detection for gaseous sample analyses.pical
() **There are no batch mode radioactive gaseous release pathways at Plant Hatch.
t i
- 1. 0 -
r -
Table'2-5 E.'I.-HATCH NUCLEAR-PLANT - SITE E(s . SEMIANNUAL RADIOACTIVE-EFFLUENT RELEASE REPORT ~1990 G*,SEOUS EFFLUENTS-- DOSE RATES Dose Rates!Due to Noble Gases Organ: Tech Unit Quarter 3 % of -Quarter 4 % lof spec Tech Tech Limit Spec- Spec Limit Limit.
TBody- 500 ~ mram/yr '7.09E-02 1.42E-02 2.97E-02' 5.95E-03 Skin 3000 arem/yr 1.53E-01 5.09E-03 6.34E-02 2.11E ..............___ ............___...............__e ........
Dose Rates Due.to Radioicdine, Tritium, and Particulates Tech % of
~
- Organ Unit Quarter 3- -Quarter 4 % of.
-Spec Tech-Limit' Tech Spec . Spec Iimit Limit:
Bone 1500 arem r.L4.41E-05 2.94E-06" 4.57E 3.05E-06:
. Liver. 1500 . mrem r1 1.08E-02 7.20E-04 1.19E-02 7.96E-04
( '
TBody 1500L mram r- 1.08E-02 .7.19E-04 1.19E-02 7.95E-04 Thyroid-1500 imram r 1.86E-02 1~.24E-03 .1.91E-02 1.27E-03 Kianey- 1500- mram r. 1.08E-02 7.21E-04 1.20E-02 7.98E-04 Lung. 1500 mrem r '1.14E-02 7.58E-04 1.23E-02 8.23E-04 l GILLI 150S- arem r. 1.08E-02 . 7 .19 E- 0 4 - 1.19E 7.95E-04 l
'x x
_h1-I
_ . . . . .. . . _ . . - _ _ . ._~.- _ _ . _ . _ ._ .__.__..._.-....._.___m___
i TABLE 2-6a !
\ E. 'I . HATCH NUCLEAR: PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990-AIR DOSES DUE-To NOBLE GAS RELEASES cumulative 1 Doses per Quarter +
Type Tech- Unit Quarter 3 % of Quarter.4 % of
-of Spec Tech Tech Radi- Limit Spec Spec ation: Limit Limit Gamma, 5.0 mrad. 1.38E 2.76E-01 2.60E-03 5.20E-02
. Beta
. 10.0 .. . mrad 1.73E-02 1.73E-01 6.4SE . .
- 6. 4 5E- 02 i
Cumulative Doses This Year-Type . Tech = Unit Quarters % of oY Spec I,2,3,& 4 Tech Radi- Limit Spec ation- Limit -
Gamma. 10.' O ' mrad 8.67E-02 8.67E-01 4 -
O) .
-(- -
Bata
. .20.0 . .....
mrad la48E-01 ....
-7.39E-01 . . .... .
1 s
!. 2-l
. -.. , . -.- - . _ _ . . - _ ~ , , -.-
1 TABLE '2-6b E. I. MATCH NUCLEAR PLANT -- UNIT 2 (s SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE - REPORT 1990 AIR DOSES DUE TO NOBLE. GAS RELEASES-Cumulative Doses per Quarter Type Toch Unit Quarter 3 % of Quarter 4 % of oY Soec Tech Tech
'Radi-- Limit Spec Spec ation L1mit
........................____...__....__.....................'. L1mit EGamma 5.0 mrad 4.96E-03 9.92E-02 5.41E-03 1.062-01 Beta -10.0 mrad. 5.97E 03 5.97E-02 6.3BE-03 6.3BE-02 Cumulative Doses This Year Wpe Tech Unit Quarters % of J ot. Spec I,2,3,& 4 Tech Radi . Limit Speg' ation Limit Gamra 10.0- mrad 6.18E-02 -6.18E-01 Beta 20.0 mrad 1.50E-01 7.52E-01 L3 I
TABLE 2-7a
/_T/
s E. I. HATCH NUCLEAR PLANT - Unit 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990
'~' INDIVIDUAL DOSES DUE TO RADIOIODINE TRITIUM, AND PARTICULATES IN GASEOUS RELkASES Cumulative Doses per Quarter Organ Tech Unit Quarter 3 % of Quarter 4 % of Tech Spec Tech Limit SDec Spec Limit Limit Bone 7.5 mrem 5.97E-04 7.95E-03 4.84E-04 6.45E-03 Liver 7.5 mrem 2.34E-03 3.12E-02 1.77E-03 2.36E-02 TBody 7.5 mrem 2.35E-03 3.13E-02 1.76E-03 2.35E-02 Thyroid 7.5 mrem 4.45E-03 5.93E-02 5.17E-03 6.90E-02 Kianey 7.5 mrem 2.33E-03 3.10E-02 1.76E-03 2.34E-02 Lung 7.5 mrem 2.37E-03 3.15E-02 1.76E-03 2.35E-02 GILLI 7.5 mrom 2.37E 3.15E 1.77E-03 2.35E-07
....... ____._______________-03 _____.____-02 _ . _ _ _ . . . . . _ _ _ _ _ _ _ _ . . .
Cumulative Doses This Year l Organ Tech Unit Quarters % of SpeQ 1,2,3,& 4 Tech l
("'I-i V
Limit Speg L1mit t ......-__.___ ... _____.................. __........__....___
Bone 15.0 mrem 2.76E-03 1.84E-02 Liver 15.0 mrem 8.37E-03 5.58E-02 TBody 15.0 mrem 8.23E-03 5.49E-02 Thyroid 15.0 mrem 3.87E-02 2.58E-01 Kidney 15.0 mrem 8.22E-03 5.48E-02 Lung 15.0 mrem 8.11E-03 5.41E-02 GILLI 15.0 mrem 8.11E 5.41E 02 l
______._.......____________-03 __________ .._______..___.________
i q) l
- 1. h .
l
fs TABLE 2-7b
/ \ E. I. HATCH NUCLEAR PLANT - Unit 2
( ,/ SEMI ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1990 INDIVIDUAL DOSES DUE TO RADIOIODINE TRITIUM, AND PARTICULATES IN GASEOUS RELbASES Cumulative Doses per Quarter Organ Tech Unit Quarter 3 % of Quarter 4 % of Speg Tech Tech Limit Spec Spec L1mit Limit Bone 7.5 mrem 1.71E-04 2.29E-03 1.51E-04 2.01E-03 Liver 7.5 mrem 2.82E-03 3.76E-02 3.70E-03 4.94E-02 TBody 7.5 mrem 2.82E-03 3.76E-02 3.70E-03 4.93E-02 Thyroid 7.5 mrem 4.83F-03 6.44E-02 7.34E-03 1.06E-01 Kiancy 7.5 mrem 2.82E-03 3.77E-02 3.71E-03 4.95E-02 Lung 7.5 mrem 2.81E-03 3.75E-02 3.69E-03 4.92E-02 GILLI 7.5 mrem 2.82E-03 3.76E-02 3.69E-03 4.92E-02 Cumulative Doses This Year Organ Tech Unit Ouarters % of
(^T Sgey 1,2,3,& 4 Tech bt Bone 15.0 mrem 7.53E-04 5.02E-03 Liver 15.0 mrem 1.39E-02 9.28E-02 TBody 15.0 mrem 1.39E-02 9.28E-02 Thyroid 15. 4 mrem 3.02E-02 2.01E-01 Kidney 15.0 mrem 1.39E-02 9.30E-02 Lung 15.0 mrem 1.39E-02 9.24E-02 GILLI 15.0 mrem 1.39E-02 9.26E-02 l
l l
1 l
1
("% i L5_ l 1
1
TABLE 2-8 LOWER LIMITS OF DL'rECTION - GASEOUS SAMPLE ANALYSES f ~< The values in this table represent apriori lower limits of detection (LLD) which are typically achieved in laboratory fL'j analyses of gaseous radwaste samles.
RADIONUCLIDE LLD UNTTS Kr-87 1.31E-07 uC1/nd Kr-88 2.10E-07 X e-133 1.62E-07 Xe-133m 6. 07E-0 8 X e-135 5.77E-08 Xo-138 2.85E-06 I-131 4. 37E-14 I-133 6.16E-13 Mn 2.78E-14 Fe-59 4. 62E-14 Co-58 2.46E-14 Co-60 2.88E-14 Zn-65 7.51E-14 Mo-99 6.02E-13 Cs-134 3.64E-14 Cs-137 2.88E-14 Ce-141 4.94E-14 Ce-144 2.02E-13 f~'T Sr-89 1.00E-ll
() Sr-90 H-3 1.00E-11 1.00E-05 4
%d 14 6 -
- J.a.iu-' Jr,, .
4
- 3. SOLID WASTE ,
3.1 REGULA'!ORY REQUIREMEm'S
.(N s
) The Technical Specifications presented in this section are for Unit 1. Requirements for Unit 2 are the same as for Unit 1r however, the i
. Technical Specification numbers are not the same. I TECHNICAL SPECIFICATIONS 3.15.3.1 The solid radwaste system shall be used in accordance with the PROCESS CONI'ROL PROGRAM to provide for the SOLIDIFICATION of wet solid wastes and for the SOLIDIFICATION and packaging of other radioactive wastes, as required, to ensure the meeting of the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site.
6.9.1.9 states in part:
The Radioactive Effluent Release Report shall include the following informtion for each type of solid waste shipped offsite during the report period:
- a. Container volume
- b. Total curle quantity (specify whether determined by measurement or estimte)
V c. Principal radionuclides (specify whether determined by measurement or estimte)
- 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. 'So11difxcettoo agent, e.g., cement.
3.2 SOLID WASTE DATA Regulatory nuide 1.21 Table -3 is found in this t eport as Table 3-la and 3-lb.
A
_h7-
EFFLUEtTT AND WASTE DISPOSAL SEMIANNUAL REIORT 1990 July 1, 1990 - December 31, 1990 7s - SOLID WASTE AND IRRADIATED FUEL SHIPMEttrS ICR UNITS I AND II TABLE 3-1A A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fue.)
l 1. Type of waste l UNIT l 6 month lEst. Totall l l 1 period i ERROR %~l
- a. Spent resins, filter sludges, evaporator l m3 1 1.67 E 21 l bottoms, etc. I Ci l CETE 2 ' 1.0 E1 l l b. Dry compressible wste, contaminated ] m3 1 6 .99 E 2 l equip, etc.
l _
l Ci i 2.6L E O . 0 E1 l l c Irradiated components, control rods, l m3 1.1 L E 1 l l _,,,,_ l Ci 2 .97 E L 2 .0 E1 l l d. Control Rod Drive Filters l m_
3 l E l l _ l Ci l E E l l e. Other (describe) l m3 l 1 .72 E 1 l l l equip. etc. Sru r F11 tera l Ci l o 99 E O l 2.0 E1 l 2, Estimate of major nuclide composittor. (by type of waste)
-l- ISOTOPE l PERCENT l CURIES .]
l a. co 60 1 17.86 l 100.7L l l Or-137 I L.78 ! 26.97 l l "n 6c l L2.07 l 237.32 l
/7 l Fe.cc 1 10.7c I 60.6L .l Vl l b.
Other T e nt nr-Cn.60 I
l
'L.Rt 96.70 l
l 119.Lo 0.70 l
l l c n _177 1 6.cc i 0.17 l l S.6s I 51.10 l 1.36 ,1
- l. Fe rc i L.17 1 0.11 l l_ Other inntnrae I l' do 1 0.10 l
, l c. en 60 l 50.7 - 0 l 't ;260 l
.l Fe-se I 15. 91 l c .050 _ ,\
- l' Mi 63 1 4.L1 l _ . 70 l l Other Tentnren I (.01 l t1 l l I I l d. na- I l l
- l. Ma l I l l_ m i j- l l- m I ,..
I I l- 1 I _1 l -e- 'N _ c L _ l 1 co I n,e l l %_he 1 6 .1, I n 6, l l- en_6a I es.rf I- eme l l rn.cc l 90.91 l o 07 __l l nnne Tent mes _ l L co I n_Ls l
.3. Solid Waste Dispositton Number of Shipments Mode of Transportation Destination p
(> 2 Tractor & Traller Barnwell, SC D
I Cask & Van B, IRRADIA1ED FUEL SHIPMENTS (Disposition)
Number of Shipments Mode of Transportation Destination I'A I;A UA
.k3
EFFLUEfC AflD WASTE DISIOSAL SEMIAffffUAL REIOPT 1990 July 1, 1990 . Decenber 31, 1990 SOLID WASTE AfiD IRRADIATED FUEL Sf!IPMPffts fDR UtJITS I At1D II v
, TABLE 3-1B l
l' t y pt "[~ cua r t i en1Nc1 plt I eus!AL INumute or l votunt or i tvec l sol tote tcArtomt l _or lov4Ntity /l WCLibts / I cose A!Nik icoNTAINtRt.1 TACH coNTA2Nik Itatt s'nt N t / l Act ml I l watir etTtemiutot_.ottt W Nat1oN I etscnefION 1 sutPPto icyn; ntt Ut nLQoNT AINLRJ __ l l I I I I I I iDevateredlC19.P6 I Co.60,Co.137 i Poly.l!IC l 29 l 202.1 i Uhc l NA I l
Resinn I measured "n.65,Fe.55_ l fiuPac E1 210 l l I ': i I
{E*8i(CPC,C(en
{ measured l l l {ypeA LCA l l 1 ES2.. .... ... 1............J..............J.........L..............L....... 1...........J
'I Devatered 3kk.90 I Cr.51,Zn.65fPolyllIC 2 l 132.h l NLC l UA l
l Recins- i Co.60, Fe 55I NuPac E1 1h2 L L,T,ype B i ;
(CUPO) meacuredl Co.137 I LCA I I I I mencured l I I I
.........J Dry ompreus.1
_ .1.6B lCo.60,Cs.137 '..............l.........L............
i IZn.65,re.55 I (netal)
B.25 boxes 32 l 96 g lttrong i LCA ; NA l
p[ible l ctimatedi measured i l
8 l
I Itight l i!
l x N aste l I l I I I
I 1 1 g.... ......p... ....p q I
....q.........p.
l
.. ....... p.
l
.....p. .........q I
U'. I l 9 99 lMn.5L,Zn.65 l Poly-llIC i 3 l 202.1 l LCA j NA ;
Tords i ebtic. ate 0 Co.60, Fe.5'l E1 210 l 1 t ; l I Filters l- I estimated -
l IHRC Typel l I l. 1 I I l A i 1
) 1 l l 1 l l 1 .....p 7....
- Irransatep- l Co.60,Fe 55 Cteel q.l .10
. . . ... .T' * *k ,""T~ ~. k 6~~
l 57. l
'* ~ ~r*R ad l i No.-- ~ ~ ~ ~ lt ~ ~ ~ ~A'*-
l Hardware 125,70h.50 - N1 63 Containere l I 22.4 Iactive I l- '
1 TN.RAll I 1 I i l entir.atef ectimated !CNSI.355 { l lhc.l 4aterialll l l
I..... .............. . .. . .. . . . . . . J. . .. . . . . .L . . . . . . . . .L WI P . S. .l . . . . . . .. . . . J Dry 0.96 lCo.60, Co.231 Cteel i Unknown l Ur4known l Ur.known l NA l
, I Compreen-: estimatep 'n.65, Fe.5L Containere l l l l Iible i Iestimated 1 I i i l l I Wacte l I l I i 1 l (SEG to- l l l l l l l l Entnvellh l l l [ , i j CPS =Condansate Phase Separator CE0= Scientific Eco' agy Group li1C=High Integrity Container
[
]LCA= Low Specific Activity (GC'JPS= Clean Up Phase Deparator hp.
in uu i ii i r -
i
- 4. SUPPhPMlWTAPY TNitPMATION LOR THE PLA!TP HATCH SfHTAtM1AL FADIOtfGICAL 0FFLUffrr PPLEASF PPKPT TOR *!HF FIRST Six WPTrHS OF 1990.
() The purpose of this section is to acknowledae the f ailure to meet the lower limit of detection (LLD) requirement for Sr-89 when analy:Ino two 11guld composite samples representina the second quarter 1990, and to discuss the potential impact on 11guld effluent values reported in the HNP Semlannual Radioloq1 cal Ef fluent Pelease Peport for the first six months of 1990.
The lower limit of detection for St-89 in 11guld ef fluent samples spec 1 fled in Technical Specification Table 4.15.1-1 (Unit 1) and Table 4.11.1-1 (Unit 2) is SE-8 uC1/ml. In both instances in which the required lower linut of detection was not achieved, the primsry cause was inadequate sanole size.
For the Unit 1 11guld effluent samole compos 1ted durano the second quarter 1990, results of analysis for St-89 were reportel as less than 7E-8 uC1/ml. Positive composite sample results are norna11y reported to two significant figures. Therefore, St-89 could potentially have teen present in Unit 1 liquid effluent at a concentration of 6.9E-8 uC1/ml. During the second quarter 1990, 4.78E6 11ttes of 11guld effluent were released from Unit 1, Therefore, 3.3E-4 curies of Sr-89 potentially nay have been released in Unit 1 11guld effluent during the second quarter 1990.
This amount of radioactivity would not have affected the " fission and activation products
- total which was reported as 1.35E-1 curies in Table 1-2a of the Semiannual Radiological Effluent Pelease Ih
\- '
Repsct for the first six months of 1990. While the "t of applicable 11nat" would be sllchtly increased, it would remain less than one-third of one percent. Doses to an individual, reoorted in Table 1-4a of the aforemen*.ioned report, would not be increased due to this amount of Sr-89 in Unit 1 liquid ef fluent.
For the Unit 2 liquid effluent sample camposited during the second quarter 1990, results of analysis for Sr-89 were reported as less than 6E-8 uC1/ml. Therefore, Sr-89 could have been present in Unit 2 11guld effluent at a concentration of 5.9E-8 uC1/ml. During the second quarter 1990, 2.43E6 liters of liquid effluent were released from Unit 2, potentially resulting in a release of 1.43E-4 curies of Sr-89. This amount of radioactivity would have increased the
" fission and activation products" total, presented in Table 1-2b of the Senuannual Radioloq1 cal Effluent Pelease Report for the first six months of 1990, from 1.13E-2 curies to 1.14E-2 curies, which would be an increase of less than one percent. Although the "% of applicable limit" would be increased s11ohtly, it would remain approxinately one-fourth of one percent. Doses to an individual, reported in Table 1-4b of the report mentioned above, would not be increased due to this- amount of Sr-89 in Unit 2 liquid ef fluent.
In summary, althouah Sr-89 may hav . been present in liquid effluents released from Plant Hatc; during the second quarter 1990, the ef fect on reported quantitles of radioactivity released would
(~l have been minimal. Potential doses to an Individual resulting from
\m / Sr-89 in liquid effluents would have been too sna11 to affect reoorted doses.
l
.. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ J
1 5 ME'rtopotmY '
T In accordance with Technical Specif teation 6.9.1.9, the (d annual suwary of meteorological data collected at Plant Hatch over 1990 as presented in this section.
5.1 1990 Meteoroloq1eal Data Attachment 1 Joint Frequency Tables of Wind Speed and Wind Direction 10m vs Delta Temperature 60-10m.
Attachment 2 Joint Prequency Tables of Wind Speed and Wind Direction 60m vs Delta Temperature 60-10m.
Attachment 3 Joint Frequency Tables of Wind Speed and Wind Direction 200m vs Delta Temperature 100-10m.
~h (V
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