ML20210D851

Semiannual Radioactive Effluent Release Rept,Jan-June 1986

ML20210D851
Person / Time
Site:	Hatch
Issue date:	06/30/1986
From:	Gucwa L GEORGIA POWER CO.
To:	Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
0692C, 692C, SL-1174, NUDOCS 8609220021
Download: ML20210D851 (58)

Text

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GEORGIA POWER COMPANY PLANT E.

I.

HATCH UNITS NO. 1 & 2 SEMI-ANNUAL REPORT t

RADIOACTIVE EFFLUENT RELEASE REPORT January 1,

1986 - June 30, 1986 i

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i 8609220021 860630 PDR ADOCK 05000322 R

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PLANT E.

I.

HATCH SEMIANNUAL REPORT RADIOACTIVE EFFLUENT RELEASE REFORT SECTION T I'P L E PA,GJ 1.

LIQUYD EFFLUENTS 1

1.1 REGULATORY L,IMITS

]

1.2 MAXIMUM PERMISSIBLE CONCENTRATYONS 5

1.3 MCASUEEMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY 5

1.4 LIQUID EFFLUENT RELEASE DATA 7

1.5 RADIOLOGICAL IMPACT ON MAN DUE TO LIQUID RELEASES 9

2 GASEOUS EFFLUENTS 18 2.1 REGULATORY LIMITS 12 2.2 MEASUREMENTS A14D A P PRO X IM A'fIONS OF TOTAL RADIOACTIVITY 24 2.3 GASLOUS EFFLUENT RELEASE DATA 29 i

2.4 RADIOLOGICAL IMPACT DUE TO GASEOUS RELEASES 30 3

SOLID WASTE 47 3.1 REGULATORY SPECIFICATIONS 41 3.2 SOLID WASTE DATA 47 4

CII ANGCS TO THE PT. ANT HATCH ODCM 50

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PL4HT E.

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HATCH SEMIAGNUAL REPORT RADIOACTIVE EFFLUENT RELEASE SEPORT JANLE LIST _OF TABLES PAGE 1-1

'rECHNICAL SPECIFICATION TABLE 3.14.1~1 RADIOACTIV3 LIQUID EFFL')ENT MONITOPING IN9TRUMENTATION 3

l - 2.s LIQUIC EFFLUENTS - SUMMATIGN 3F UMIT 1

.10 ALL K3 LEASES 1-26 LIQUIC JFfLUENTS - SUMMiTION OF

• - UNIT 3

,11 ALL RELEASES 1-3a LIQUIC EFFLUENTG - UNIT 1 12 1-3b LIQUID EFFLUENTS - UNIT 2 13 1-4a INDIVIDVAL 00SSS DUE TO LIQUIL RELEASES - UGIT 1 15 l-46 INDIVIDUAL DOSES DUE TO LIQUID RELEASES - UNIT 2 16 LIQUID 1-5 LOWER T.IMITS OF DCTECTION SAMPLE ANALYSE.S 17 2-1 TECHNICAL SPECIFICATION TABLE 3.14.2-1 RADIOACTIVE GASEOUS EFFLUENT MOIIITORING TNSTRUMFNTATION 20 l

l 2 '- 2 a GAGEOUS EFFLU6NTS - SUMMATION OF l

ALL RELEASES - UNIT 3 31 l

i-2b GASEOUS EFFLUENTS - SUMMATION OF 32 ALL RELEASES - IINIT 2 2-2c GASEOUS EFFLUENTS - SUMMATION OF l

ALL RELEASES SITE 33 2-3a GASEOUS EFFLUENTS ELEVATED R3 LEASES - UNIT 1 34

+

i 2-3b GASEOUS EFFLUENTS ELEVATED I

RELEASES - UNIT 2 36 l

2-3c GASEOUS EFFLUENTS - ELEVATED RELEASES SITE 37 i

1 TABLES LI_ST OF TABLES

.? AC E 2-4a GASEOUS BFF.LUENTS - GROUND-UNIT 1 38 LEVEL RELCASES 2-4b GASSOUS EFFLUENTS GROUND s-LEVEL RELEASES UNIT 2 39 2-4c GA3EOUS EFFLUENTS - CMOUND -

t LE'JEL RELEASBS SITE 40 2-5 CASEQUS EFFLUENTS - DOSE SITE 41 RATES 2-Ga AIR DOSES DUE TO NOBLE GASES -

UNIT 1 42 2-6b AIR DOSES DUE TO NOBLE GASES UNIT 2 43 2-7a IUDIVIDUAL DOSES DUE TO RADIOIODINE, TRITIUM, AND PARTICULATES IN GASEOUS RELEASES - UNIT 1 44 2-7b INDIVIDUAL DOSES DUE TO RADIOIODINE, THITIUM, AND FARTICULATES IN GASEOUS UNIT 2 45 RELEASES 22-8 LOWER LIMITS OF DETECTION GASECUS SAMPLE ANALYSES 46 3-la,b SOLID WASTE AND IRRADIATED FUEL SHIPMENTS 48,45 i

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RADIOACTIVE EFFLUENT RELEASE REPORT

/l' LIQUID EFFLUENTS 1.1.

REGULATORY LIMITS 1.

The Technical Specifications presented in this section are for Unit 1.

Requirements far Unit 2 are the same as Unit 1; however, the Technical Specification nuabers are not the same.

TECHNICAL SPECIFICATIONS 3.14.1 The radioactive liquid effluent monitoring instrumentation ch3nnels shown in table 3.14.1-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.15.1 are not exceeded.

The 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 UNRESTRICTED AREAS (figure 3.15-1) chall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II ~(column 2) for radionuclides other than r

dissolved or entrained noble gases.

For dissolved or entrained noble gases, the concentration shall be limited to 2.0E-04 uCi/ml total activity.

1 3.15.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each reactor unit, from the site (figure 3.15-1) shall be limited to:

i a.

During any calendar quarter to less than or equal to 1.5 mrem tc 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.

l 3.15.1.3 The liquid radwaste treatment system, as described in the ODCM, shall be used to reduce the radioactive materials in liquid wastes prior to their l

discharge when the projected doses due to the liquid l

effluent per Unit from the site (figure 3.15-1) when projected over the calendar quarter would exceed 0.18 mrem to the total body or 0.62 mrem to any organ.

(

~

3.15.1.4(a)

The contents within any outside temporary tank shall be limited to $10 curies, excluding tritium and dissolved or entrained noble gases.

(a)

An outside temporary tank is not surrounded by liners, dikes, or walla 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 Ci, 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 !

_ _ _ _,_..._ ___ _ -~. _ - _.

TABLE l-1 TECHNICAL SPECIFICATION TABLE 3.14.1-1 (SHEET 1 of 2)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument OPERABLE Applicability ACTION 1.

Gross Radioactivity Monitors Providing Automatic Termina-tion of Release Liquid Radwaste Effluent Line 1

(a) 100 2.

Gross Radioactivity Monitors not Providing Automatic Termination of Release Service Water System Effluent Line 1

(b) 101

3. Flowrate Measure-ment Devices **

Liquid Radwaste Effluent Line 1

(a) 102 Discharge Canal 1

(b)

(a) 102 l

4.

Service Water 1

At all times 103 j

System to closed j

Cooling Water System Differential Pressure

• • Pump curves may be utilized to estimate flow; in such cases, ACTION statement 102 is not required.

I (a)

Whenever the radwaste discharge valves are not locked closed.

(b)

Whenever the service water system presture is below i

the closed cooling water system pressure or et P Indication is not available.

_3_

i

TADLE l-1 (CONTINUED)

TECHNICAL SPECIFICATION TABLE 3.14.1-1 (SHEET 1 of 2)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATIONS ACTION ~10 0 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be continued, provided that prior to initiating a releases a.

At least two independent samples are analyzed in accordanne with Specification 4.15.1.1.1.

r 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 remains 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 Channelu OPERABLE requirement, effluent releases via this pathway may 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 uCi/ml.

If the channel remains inoperable for over 30 days, an explanation of the circumstances shall be included in the next semi-annual effluent release report.

ACTION 102 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided 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 /> during actual releases.

If the channel remains inoperable for over 30 days, an explanation of the circumstances shall be included in the next semi-annual effluent release report.

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 1.2 MAXIMUM PERMISSIBLE CONCENTRATIONS I

The MPC values used in determining allowable 3

liquid radwaste release rates and concentrations for principal gamma emitters, I-131, tritium, Sr-89, Sr-90 and Fe-55 are taken from 10CFR Part 20, Appendix B, Table II, Column 2.

For dissolved or entrained noble gases in liquid radwaste, the MPC is taken from Technical Specification 3.15.1.1 (Unit 1) and 3.11.1.1 (Unit

2) as 2.0E-04.uci/ml.

For gross alpha in liquid radwaste, the MPC is taken from 10CFR Part 20, Appendix B, Note 2.b as 3.0E-08 uCi/ml.

Further, for all the above radionuclides or categories of radioactivity, the overall MPC fraction is determined in accordance with 10 CFR Part 20, Appendix B,

Note 1.

The method whereby the MPC fraction is used to determine release rates and liquid radwaste effluent radiation monitor setpoints is described in Section 1.3 of this report.

1.3 MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY Prior to release of any tank containing liquid radwaste, and following the required recirculation, samples are collected and analyzed in accordance with Technical Specification Tables 4.15.1-1 (Unit 1) and 4.11.1-1 (Unit 2).

A sample from each tank planned for release is analyzed for principal gamma emitters, I-131, and dissolved and entrained noble gases by gamma spectrometry.

Monthly and quarterly composites are prepared for analysis by extracting aliquots from each sample taken from tanks which are released.

Liquid radwaste sample analyses are performed as follows:

Measurement Frequency Method l

1.

Gamma Isotopic Each Batch Gamma spectroscopy with computerized data reduction 2.

Dissolved or Each Batch Gamma spectroscopy Entrained with computerized Noble Gases data reduction l

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Mynnuronent Frcquency M a t h e 'd

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3.

Tritium Monthly Distilkation and

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Composite liquid,scintill,ation

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counting t

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Gross Alpha Monthly Gas Flow Itdp'o'rtional Composite c o u n t i n'g' ~

5.

Sr-89 c.id Sr-90 Quarterly Chemical separation and Composite gae, flow propo'rtional g

counting 4

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Fe-55 Quarterly gChemical separadion and Composite

'l o w esergy phc[tdn

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3 Gamma isotopic measu'rements a ret performed in-house in

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the radiochemistry labs using germanium! spectrometry.

Three germanium detectors are avdilable:\\ a 10%

li efficient Ge (Li) detector and two 15% ef ficient intrinsic germanium detectors,'with(2.D'TWNM

' [\\ N, resolution and housed in 4 inch-thlck lead shields.

,A one-liter liquid radwaste sample'is-poured into a l-Marinelli beaker in preparation for a 2000-3000 seconJb-count.

A peak search of the resulting gamma I'a y "

spectrum is performed by,the co'mputer system.

Ener'gy and net count data for all,significant peaks, are '

determined, and quantitative reduction or LLD calculations are performed for the n uc lid e s ' s p'ec i f i e d in Table Notation e of Technic,al Specijfication Tables g

4.15.1-1 (Unit 1) and 4.11.1-1}134, (Unit 2) :

Mc-54, E-59,

?,

Mo-99, Cs-Cs-137, Ce-141 and

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Co-58, Co-60, 2n-65, Ce-144.

The quantita tive aa'le'ula tions', include M*

corrections for counting tine, decay t ime', [s ample

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volume, sample geometry, detActor e f f icie ncy, baseline counts, and branching ratio.) iLD calcula'tions, 3

l including the above corrections, are made based on$the counts in two standard deviations of'the caseline e

count at the location on} the' spectrdm where a peak Ufor that radionuclide wo'ufd he, located if present.

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• 1 s

s The radionuclide concentre,tions dgtermined 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 a l p'h a,

Sr-89, Sr-90, and Fe-55 are used along with the corresponding MPC values t'o determine an MPC< fraction for the tang planned for release.

• T h i, s MPC fraction is then used, with appropriate safety factors, along with the expected dilution,r,tream (loa'to calculate a maximum permissible release rate and a liquid

• effluent monitor setpoint.

The monitor setpoint is calculated T

to assure that the limits of,Techn'ical Specifications' 3.15.1.1 (Unit 1) or 3.11. F. l * ( U,n i t 2) are not 5

exceeded.

i,

A conitor roading in excosa of the calculated setpoint therefore results in an automatic termination of the liquid radwaste discharge.

Liquid effluent discharge is also automatically terminated if the dilution stream flow rate falls y

below the dilution flow rate used in the c e tpo i n t.

calculations and established as a setpoint on the (is dilution stream flow monitor.

Radionuclide concentrations, safety factors, dilution stream flow rate, and liquid effluent radiation monitor calibration factor are entered into the computer and a prerelease printout is generated.

If the release is not permissible appropriate warnings will be included on the prerelease printout.

If the release is permissible it is approved by the Chemistry Foreman on duty.

The pertinent information is transferred manually from the prerelease printout

l i

to a one-page release permit which is forwarded

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to Radwaste Operations.

When the release is g( i completed the release permit is returned from Radwaste Operations with actual release data included. These data are input to the computer k

and a postrelease printout is generated.

The postrelease printout contains actual release rates, actual release concentrations and quantities, actual dilution flow, and calculated doses to an individual.

+

1.4 LIQUID EFFLUENT RELEASE DATA Regulatory Guide 1.21 Tables 2A and 2B are found in this report as Table 1-2a for Unit 1 and Table p

1-2b for Unit 2; and Table 1-3a for Unit 1 and 1-3b for Unit 2.

The values for the four categories of Tables 1-2a and 1-2b are calculated and the Tables completed as follows:

t 1.

Fission and activation products - The total l

release values (not including tritium, gases, and alpha) are comprised of the sum of the measured individual radionuclide activities.

This sum is for each batch s

released to the river for the respective t

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4 quarter.

Porcent of applicable limit is

-determined from a mixed nuclide MPC fraction calculation.

The average concentration for each nuclide over all released batches is divided by the corresponding' individual MPC value.

The sum over all nuclides of the i

Ci/MPCi ratios times 100 is the percent of applicable limit for effluent releases during the quarter.

2.

Tritium - The measured tritium F

concentrations in the monthly ' composite samples are used to calculate the total 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 germanium spectroscopy on a one liter sample

-from each liquid radwaste batch.

The average concentration of dissolved or entrained noble gases for all released batches is divided by the MPC value stated in Technical Specifications 3.15.1.1 and 3.11.1.1 (2.0E-04 uCi/ml) to determine the MPC fraction.

The result x100 is the percent of applicable limit for noble gases in liquid effluent releases during the quarter.

Radioisotopes of iodine in any form are also determined during the isotopic analysis for each batch; therefore, a

~

separate analysis for possible gaseous forms is not performed because it would not provide additional information.

l 4.

Gross alpha radioactivity - The measured l

gross alpha concentrations in the monthly composite samples are used to calculate the total release of alpha radioactivity.

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Other data pertinent to batch releases of radioactive liquid effluent from both units is as follows:

Number of batch releases:

764 Total time period for batch releases:

93,577 minutes Maximum time period for a batch release:

290 minutes Average time period for batch releases:

123 minutes Minimum time period for a batch release 5.0 minutes Average stream flow during periods of release of liquid effluent into a flowing stream:

6069 CFS 1.5 RADIOLOGICAL IMPACT ON MAN DUE TO LIQUID 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 1 and Unit 2.

Results are presented in Table 1-4a for Unit 1 and Table 1-4b for Unit 2.

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TABLE 1-2a E.

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HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL EFFLUENT RELEASE REPORT 1986 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES Est. Total Unit Quarter 1 Quarter 2 Error %

A.

Fission &

Activation Products 1.

Total release (not including Ci 1.73E-01 4.80E-02 4.6E+01 H-3,

gases, alpha) 2.

Average diluted concentration uCi/ml 1.32E-07 4.03E-08 during period 3.

% of applicable 7.90E-01 4.00E-01 limit B. Tritium 4

1.

Total release Ci 3.99E+00 3.45E+00 3.7E+01 2.

Average diluted uCi/ml 3.05E-06 2.90E-06 concentration during period 3.

t of applicable 1.00E-01 9.66E-02 limit C.

Dissolved and Entrained Gases 1.

Total release Ci 2.81E-03 4.40E-04 1.0E+02 2.

Average diluted uCi/ml 2.15E-09 3.70E-10 concentration during period 3.

% of applicable 1.08E-03 1.85E-04 limit D.

Gross alpha radioactivity 1.

Total release Ci 5.76E-07 8.43E-06 1.2E+02 E.

Volume of waste (prior to dilution) liters 7.13E+06 6.06E+06 1.0E+01 F.

Volume of dilution water used liters 1.31E+09 1.19E+09 1.6E+02

TADLE'l-2b E.

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HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL EFFLUENT RELEASE REPORT 1986 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES Est. Total Unit Quarter 1 Quarter 2 Error %

A.

Fission &

Activation Products 1.

Total release (not including Ci 1.84E-02 4.85E-02 4.7E+01 H-3,

gases, alpha) 2.

Average diluted concentration uCi/ml 1.73E-08 3.91E-08 during period 3.

% of applicable 1.23E+00 3.16E+00 limit B. Tritium 1.

Total release C1 3.81E+00 1.90E+00 3.7E+01 2.

Average diluted uCi/ml 3.64E-06 1.53E-06 concentration during period 3.

% of applicable 1.20E-01 5.llE-02 limit l

C.

Dissolved and Entrained Gases 1.

Total release Ci 4.40E-02 3.02E-02 1.0E+02 2.

Average diluted uCi/ml 4.19E-08 2.44E-08 concentration during period 3.

% of applicable 2.00E-02 1.22E-02 limit D.

Gross alpha radioactivity 1.

Total release Ci 1.32E-06 9.31E-07 1.2E+02 E.

Volume of waste (prior to i

dilution) liters 4.59E+06 3.54E+06 1.0E+01 I

F.

Volume of dilution water used liters 1.05E+09 1.24E+09 1.6E+02 l

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TABLE l-3a E.

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HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL LIQUID EFFLUENTS RELEASE REPORT 1986*

Continuous Mode **

Batch Mode Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 H-3 Ci 3.99E+00 3.45E+00 i

Na-24 Ci 8.39E-06 1.23E-03 Cr-51 Ci 3.52E-04 0.00E+00 Mn-54 Ci 7.23E-03 1.79E-03 Mn-56 Ci 2.01E-06 4.51E-05 Fe-55 Ci 2.48E-03 0.00E+00 Fe-59 Ci 3.91E-04 1.32E-05 Co-58 Ci 1.96E-03 1.82E-04 Co-60 Ci 1.56E-02 4.56E-03 Zn-65 Ci 4.99E-02 1.48E-02 As-76 Ci 0.00E+00 1.74E-05 s

Sr-89 Ci 0.00E+00 0.00E+00 Sr-90 Ci 0.00E+00 0.00E+00 Sr-91 Ci 0.00E+00 0.00E+00 Sr-92 Ci 1.83E-05 3.28E-06 Y-91m Ci 0.00E+00 2.36E-05 Zr-95 Ci 1.37E-05 0.00E+00 Nb-95 Ci 6.05E-05 6.32E-06 Tc-99m Ci 1.46E-04 2.21E-04 I-130 Ci 0.00E+00 0.00E+00 I-131 Ci 7.19E-04 7.36E-04 I-132 Ci 0.00E+00 1.89E-06 I-133 Ci 1.62E-04 3.57E-04 I-134 Ci 9.09E-04 0.00E+00 I-135 Ci 0.00E+00 1.19E-05 Cs-134 Ci 3.15E-02 6.74E-03 Cs-136 Ci 5.94E-06 1.22E-06 Cs-137 Ci 6.112-02 1.73E-02 Ba-140 Ci 0.00E+00 0.00E+00 Np-239 Ci 0.00E+00 0.00E+00 l

Totals Ci 1.73E-01 4.80E-02 i

j Kr-85 Ci

.1.93E-03 1.97E-04 i

Xe-133 Ci 6.36E-04 9.64E-05 Xe-135m Ci 0.00E+00 0.00E400 Xe-135 Ci 6.89E-05 1.43E-04 Ar-41 Ci 1.78E-04 3.34E-06 Totals Ci 2.81E-03 4.40E-04 i

Gross Alpha Ci 5.76E-07 8.43E-06

• Zeros in this table indicate that no radioactivity was present 4

above detectable levels.

See Table 1-5 for typical lower limits of detection for liquid sample analyses.

• • There are no continuous mode radioactive liquid release pathways at Plant Hatch... _

TABLE l-3b UNIT 2 E.

I.

HATCH NUCLEAR PLANT SEMIANNUAL LIQUID EFFLUENTS RELEASE REPORT 1986*

l Page 1 of 2 Continuous Mode **

Batch Mode Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 H-3 Ci 3.81E+00 1.90E+00

.Na-24 Ci 3.57E-04 7.83E-03 Cr-51 Ci 8.04E-06 6.74E-04 Mn-54 Ci 1.87E-04 5.14E-04 Mn-56 Ci 3.02E-06 3.99E-05 Fe-55 Ci 1.28E-03 0.00E+00 Fe-59 Ci 0.00E+00 0.00E+00 Co-58 Ci 9.70E-05 2.82E-04 Co-60 Ci 8.45E-04 2.21E-03 Ni-65 Ci 0.00E+00 5.llE-06 Zn-65 Ci 2.53E-03 2.59E-03 As-76 Ci 0.00E+00 8.27E-06 Rb-88 Ci 0.00E+00 1.81E-03 Sr-89 Ci 0.00E+00 0.00E+00 Sr-90 Ci 0.00E+00 0.00E+00 Sr-91 Ci 0.00E+00 0.00E+00 Sr-92 Ci 0.00E+00 0.00E+00 Y-91m Ci 6.41E-06 7.24E-04 Zr-95 Ci 0.00E+00 0.00E+00 Nb-95

.Ci 5.80E-07 0.00E+00 Mo-99 Ci 0.00E+00 7.71E-05 Tc-99m Ci 2.06E-04 1.96E-03 Te-131 Ci 1.83E-04 0.00E+00 I-131 Ci 3.08E-03 8.91E-03 I-130 Ci 0.00E+00 0.00E+00 I-132 Ci 4.19E-05 3.82E-04 I-133 Ci' 2.07E-03 7.20E-03 I-134 Ci 2.83E-06 1.llE-05 I-135 Ci 3.51E-04 1.84E-03 Cs-134 Ci 2.40E-03 3.44E-03 Cs-136 Ci 4.32E-05 1.llE-04 Cs-137 Ci 4.37E-03 6.67E-03 Cs-138 Ci 0.00E+00 4.30E-05 Ba-139 Ci 2.67E-04 0.00E+00 Ba-140 Ci 0.00E+00 0.00E+00 Ce-141 Ci 5.67E-06 0.00E+00 Np-239 Ci 5.08E-05 1.14E-03 Totals Ci 1.84E-02 4.85E-02 i

I

TABLE 1-3b E.

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HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL LIQUID EFFLUENTS RELEASE REPORT 1986*

Page 2 of 2 Continuous Mode **

Batch Mode Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 Kr-85m Ci 8.28E-06 0.00E+00 Kr-85 Ci 4.38E-03 7.34E-04 Xe-131m Ci 1.llE-03 0.00E+00 Xe-133m Ci 3.05E-04 2.61E-04 Xe-133 Ci 9.59E-03 7.95E-03 Xe-135m Ci 1.18E-04 1.62E-03 Xe-135 Ci 2.85E-02 1.96E-02 Ar-41 Ci 0.00E+00 3.22E-06 Totals Ci 4.40E-02 3.02E-02 Gross Alpha Ci 1.32E-06 9.31E-07

• Zeros in this table indicate that no radioactivity was present above detectable levels.

See Table 1-5 for typical lower limits of detection for liquid sample analyses.

• • There are no continuous mode radioactive liquid release pathways at Plant Hatch.

1.

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TABLE l-4a E.

I. HATCH NUCLEAR PLANT UNIT 1 SEMIANNUAL EFFLUENT RELEASE REPORT 1986 INDIVIDUAL DOSES DUE TO LIQUID-RELEASES Cumulative Dose Per Quarter Organ Tech Units Quarter 4 of Quarter 4 of Spec 1

Tech 2

Tech Limit Spec Spec Limit Limit

- Bone 5.0 mrem /qtr 3.53E-01 7.06E+00 1.02E-01 2.04E+00 Liver 5.0 mrem /qtr 5.96E-01 1.19E+01 1.66E-01 3.32E+00 T. Body 1.5 mrem /qtr 4.25E-01 2.83E+01 1.17E-01 7.80E+00 Thyroid 5.0 arem/qtr 7.00E-03 1.40E-01 5.42E-03 1.08E-01 Kidney 5.0 mrem /qtr 2.05E-01 4.10E+00 5.77E-02 1.15E+00 Lung 5.0 mrem /qtr 6.37E-02 1.27E+00 1.77E-02 3.54E-01 GI-LLI 5.0 mrem /qtr 3.35E-02 6.70E-01 1.03E-02 2.06E-01 Cumulative Dose Per Year Organ Tech Unit Quarters

% of Tech Spec' 1 & 2 Spec Limit Limit Bone 10.0 mrem /yr 4.55E-01 4.55E+00 Liver 10.0 mrem /yr 7.62E-01 7.62E+00 Total Body 3.0 mrem /yr 5.12E-01 1.81E+01 Thyroid 10.0 mrem /yr 1.24E-02 1.24E-01 Kidney 10.0 mrem /yr 2.63E-01 2.63E+00 Lung 10.0 mrem /yr 8.14E-02 8.14E-01 GI-LLI 10.0 mrem /yr 4.38E-02 4.38E-01 e

i i

4 i

4 TADLE l-4b E.

I.

HATCH NUCLEAR PLANT UNIT 2 SEMIANNUAL EFFLUENT RELEASE REPORT 1986 INDIVIDUAL-DOSES DUE TO LIQUID RELEASES Cumulative Dose-Per Quarter Organ Tech Units Quarter 4 of Quarter 4 of Spec 1

Tech 2

Tech Limit Spec Spec Limit Limit Bone 5.0 mrem /qtr 2.45E-02 4.90E-01 3.71E-02 7.42E-01 Liver 5.0 mrem /qtr 4.15E-02 8.30E-01 6.21E-02 1.24E+00 T. Body 1.5 mrem /qtr 2.98E-02 1.99E+00 4.46E-02 2.97E+00 Thyroid 5.0 mrem /qtr 2.llE-02 4.22E-01 6.24E-02 1.25E+00 Kidney 5.0 arem/qtr 1.42E-02 2.84E-01 2.135-02 4.26E-01 Lung 5.0 mrem /qtr 4.50E-03 9.00E-02 6.77E-03 1.35E-01 GI-LLI 5.0 mrem /qtr 1.95E-03 3.90E-02 3.31E-03 6.62E-02 Cumulative Dose Per Year Organ Tech Unit Quarters

% of Tech Spec 1 & 2 Spec Limit Limit Bone 10.0 mrem /yr G.16E-02 6.16E-01 Liver 10.0 mrem /yr 1.04E-01 1.04E+00 Total Body 3.0 mrem /yr 7.44E-02 2.48E+00 Thyroid 10.0 mrem /yr 8.35E-02 8.35E-01 Kidney 10.0 mrem /yr 3.55E-02 3.55E-01 Lung 10.0 mrem /yr 1.13E-02 1.13E-01 GI-LLI 10.0 mrem /yr 5.26E-03 5.26E-02 J

i J

l l

I i

l l,

,, - _ _. -, _ - c.._.____ - _ - - -... _ _ _. - _..,, -.. _ -. _ _ _ _ _, _. - - _ -

-m.,____,.._.___....._.-..-

TABLE 1-5 LOWER LIMITS OF DETECTION - LIQUID SAMPLE ANALYSES The values in this table represent apriori lower limits of detection (LLD) which are typically achieved in laboratory analyses of liquid radwaste samples.

RADIONUCLIDE LLD UNITS Mn-54 5.38E-08 uCi/ml Fe-59 7.78E-08 Co-58 4.67E-08 Co-60 4.78E-08 Zn-65 1.31E-07 Mo-99 5.10E-07 Cs-134 7.18E-08 Cs-137 6.05E-08 Ce-141 1.41E-07 Ce-144 6.30E-07 I-131 6.51E-08 Xe-135 8.45E-08 Fe-55 8.00E-07 Sr-89 2.30E-08 Sr-90 7.67E-09 H-3 5.00E-07,

2 GASEOUS EFFLUENTS 2.1 REGULATORY LIMITS The Technical Specifications presented in this section are for Unit 1.

Requirements for Unit 2 are the same as for Unit 1; however, the Technical Specification numbers are not the same.

TECHNICAL SPECIFICATIONS 3.14.2 The radioactive gaseous effluent monitoring instrumentation channels shown in table 3.14.2-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.15.2.1(a) are not exceeded.

The alarm / trip setpoints of these channels shall be determined in accordance with the 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 UNRESTRICTED AREAS (figure 3.15-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

1 a.

The dose rate limit for noble gases shall be $ 500 mrem / year to the total body and 53000 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 radionuclides other than noble gases with half-lives greater than 8 days shall be $ 1500 mrem / year to any l

organ.

3.15.2.2 The ait dose 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:

I a.

During any calendar quarter, to $ 5 mrad for gamma radiation and 5 10 mead for beta radiation.

b.

During any calendar year, to 5 10 mrad for gamma radiation and 5 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 form with half-lives greater than 8 days in gaseous effluents released to UNRESTRICTED AREAS (figure 3.15-1) from each reactor unit shall be limited to the following:

. ~ -_. - -...- -.-,

a.

During any calendar quarter to 1 7.5 mrem to any organ, b.

During any calendar year to 6 15 mrem to any organ.

3.15.2.4 The GASEOUS RADWASTE TREATMENT 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 GASEOUS RADWASTE TREATMENT SYSTEM operability shall be demonstrated by administrative-controls which assure that the offgas treatment system is not bypassed.

3.15.2.5 The annual (calendar year) dose or dose commitment' to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem.

(With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifications 3.15.1.2(a), 3.15.1.2(b),

3.15.2.2.(a), 3.15.2.2(b), 3.15.2.3(a), or 3.15.2.3(b), calculations shall be made including direct radiation 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 54 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 system pretreatment monitor station shall be limited to $ 240,000 uCi/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 gaseous and liquid effluents that were in excess of 1 Ci, excluding dissolved and entrained gases and tritium for 11guld effluents, or those in excess of 150 Ci of noble gases or 0.02 Ci of radiolodines for gaseous releases."

TABLE 2-1 TABLE 3.~14.2-1 (SilEET 1 OF 4)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum ChanneIs Instrument OPERABLE Applicability Pa ra me te r ACTION 1.

Main Condenser Ofrgas Treatment System Explosive Cas Monitoring System Hydrogen Monitor (1)

7. Hydrogen 106 2.

Reactor Building Vent Stack Monitoring System a.

Noble Gas Activity Monitor (1)

Radioactivity Rate 105 Measurement +

b.

lodine Sampler Cartridge (1)

Verify Presence of 107 Ca rt ri dge I

c.

Particulate Sampler Filter (1) y Verify Presence or 107 i

filter d.

Erfluent System flowrate Measurement Device (1)

System Flowrate 104 Measurement e.

Sampler Flowrate Measurement Device (1)

Sampler flowrate 104 Measurement 3.

RecomDiner Building Ventila tion Moni toring Systes a.

Noble Cas Activity Monitor (1)

Radioactivity Rate 105 Mea su remen t +

b.

Iodine Sampler Ca rtridge (1)

Yerify Presence of 107 Cartridge c.

Particulate Sampier filter (1)

Verify Presence of 107 Filter d.

Sampler flowrate Measutement (1)

Device Sampler Flowrate 104 Measurement

TABLE 2-1 TABLE 3.14.2-1 SHEET (2 OF 4)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 3

Minimum Instrument Channels

}

OPERA 8LE Acolicability Pa ramete r

- ACTION t

4.

Main Stack Monitoring System t

]

Noble Gas Activity Monitor (1) a.

Radioactivity Rate 105 Measurement +

b.

Iodino Sampler Cartridge (1) j Verify Presence of 107 Ca rt ridge 4

1 c.

Particulate Sampler Filter (1) i yerify fresence or 107 Filter d.

Errluent System flowrate Measuring Devices (1)

System Flowrate 104 Mea suremen t e.

Sampler Flowrate Measuring (1)

)

b Device Sampler Flowrate 104 4

Y Measurement-i 5.

Condenser Orrgas Pretreatment l

Monitor 4

j Moble cas Activity Monitor (1)

Radioactivity Rate 108 1

Measurement L

1 i

i 5

-O

-T y

=

q-}

v

-W=

y at g

ww '

p~~

v V

TABLE 2-1 TABLE 3.14.2-1 (SHEET 3 CF 4)

RADIOACTIVE GASEOUS EFFLUENT MONITCRING INSTRUMENTATION Table Notations

+ Monitor must be capable of responding to a Lower Limit of Detection of 1 x 10-' pCi/ml.

• During releases via this pathway.

• • During main condenser offgas treatment systen 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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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 105 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided s'

grab samples are taken daily and 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 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 106 - With the number 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 (b)

Using a temporary hydrogen analyser installed in the offgas system line downstream of the recombiner, 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 />.

l TABLE 2-1 TABLE 3.14.2-1 (SHEET 4 OF 4)

RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION Table _ Notations (Continued)

I If the number of channels OPERABLE remains less than required by the Minimum Channels OPERABLE requirement for t

over 30 days, an explanation of the circumstances shall be included in the next semi-annual effluent release. report.

ACTION 107 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided samples are continuously collected with auxiliary sampling 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 and of the sampling period.

N 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.

4 ACTION 108 - With the number of channels OPERABLE less than

• [

required by the Minimum Channels OPERABLE requirement, i

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:

The offgas system is not bypassed, and a.

b.

The offgas post-treatment monitor (D11-K615) or the main stack monitor (Dil-K600) is OPERABLE.

4 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 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.

f f

i' j

1 i

L 1

4 i

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22 MEASUREMENT AND APPROXIMATIONS OF TOTA L. R4D IC.\\CTIV!TY Waste. gas re le a.se at Plang ha ton is confined to four paths: 'nain stack (also called the of.fges ven t), Unit 1 teactor bailding vect; Unit '2 reactor building vent,'and the recoa'ei n e r building.vept.

Each o.f thcoe four patts is continuously monstered for gaseoua radioactivity. -Sach is equipped witn an integraticg-type cample coilection device for collecting particulates and iodines.

Sample collect. ion la in accordance With Technical Specification Thbles 4 15.2-1 (Unit 1) and 4 11.2-1 (Unit 2).

Unless required more-frequently under certain circurstances specified in Table Hotationo to the abovementioned tables, samples are collect.ed as follows:

l.

Noble

(

gas samplea are col.lected by grab sampling konthly.

2.

Tritium samples are collected by grab

)

sampling monthly.

i 3.

Radioiodine samples are collected by pulling the sample stream through a charcoal cartridge over a 7-day pexiod.

4.

Particulates are collected by pulling tho cample stream through a particulate filter

{

cver a 7-day period.

5.

The 7-day particulate filters above are I

analyzed for gross alpha activity.

6.

Quarterly composite samples are prepared from the particulate filters collected over the previous quarter and the qtiarterly composite sample is analyzed for Sr-$9 and Se-90.

k Sample analyses results and release flow rates from the four release points form the. basis for calculating released quantities of 1

radionuclide-npecific radioactivity, dose rates i

associated with gaseous releases, and cumulative i

doses for the current quarter and year.

This task is normally performed with computer assistance,

=

i i

I

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i 1

j I

~_

The noble gas grab sample analysis results are used along with maximum expected release flow rates from each of-the four vents to calculate monitor setpoints for the gaseous effluent monitors serving the four release points, to es6ure that the limits of Technical specifications 3.15.2.1.a (Unit 1) or 3.ll.2.1.a (Unit 2) are not exceeded.

Calculation of monitor setpoints is described in the Plant Hatch ODCH.

Mith each release period released radioactivity, dose rates, and cumulative doses are calculated.

Cumulative dose results are tabulated along with percent of Techn.ical Specification limits (3.15.2.2 and 3,15.2.3 JUnit 1); 3.11.2.2 and 3.11.2.3 (Unit 2) for each release, for the current quatter and year.

After each calendar quarter (13 weeks) a summary of waste gas releases from the four vents is coepiled for pre,paration of the Semiannual Effluent Relaase Report required by Technical Opecifications 6 9.1.8 and 6.9.1.9 and described in RFC Regulatory Guide 1.21.

The methods for determining released quantities of radioactivity, dose r.a t.e s and cumulative doces are as to11 owns 1.

F15 STOW AND ACTIVATION GAG The radionuclide-specific released radioactivity la determined from sample analyses results collected an described above and average release flow rates over the period represented by the collected s a y. p l e,

Instantaneous dose rates due to noble gases and due to radiciodines, tritium, and particulates are calculated (with computer assistance).

Calculated dose rates are compared to the dose rate limits specified in 3.15.2.1.a (Unit 1) and 3.11.2.1.a (Unit 2) for noble gases; and i

3.15.2.1.b (Unit 1) and 3.11.2.1.b (Unit 2) for radiciodine, tritium, and particulates.

Dose rato calculation methodology is presented in the Plpnt Hatch ODCH.

I i -

Eeta and gamma air deses due to noble gasen a.re calculated for the location in the unrestricted area with the potential for ths highest exposure due to gaseous releases.

Air doses -are calculated for each release period and cuma10tive totals are kept for each unit for the curreht calendar quarter and ye.ar.

Cumulative air doses are compared to the done limits specified in Technical Specifications 3.15,2.2 (Unit 1) and 3.11.2.2 (Unit 2).

Current percent of technical specification limits are shown on the printout for each release period.

Air dose calculation methodology is presented in the Plant Hatch ODCM.

2.

RAD 10 IODINE, TRITIUM, AND PARTICULATE RELEAS.SS Released guantities of radioiodines ace determined from the weekly samples and release flow rates for the four release points.

i Radioiodine concentrations are determined by gamma spectroscopy.

Released quantities of particulates are determined from tne weekly (filter) samples and release flow rates for the four release points.

Gamma spectroscopy is used to quantify concentrations of principal gamma emitters.

After each calendar month the particulate filters from each vent are combined, fused, and strontium separation is performed.

Since sample flows and vent flows are almost constant over each monthly period the filters from each vent can be dissolved together.

Decay corrections are made back to the middle of the quarterly collection f

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 from each vent by passing the sample atream from a cold e

trap immersed in liquid nitrogen or an acetone and ice mixture.

The grams of wa ter vapor /ctibic foot gas is measured upstream of the cold trap in order to alleviate the difficulties in determining water vapor collection efficiencico.

The tritium camples are analyzed by an independent laboratory and results are fuenished In uci/nl of water.

The tritium concentration in water in converted to tritium concentration in air and this value is input into the composite file of the computer to be used in release, doce rate, and individual dose calculations.

i

\\

Dose rates due to radioiodine, tritium, and j

particulates are calculated for a hypothetical child, exposed to the inhalation pathway, at the location in the unrestricted area where the potential dose rate is expected to be the highest.

cose rates are calculated for each release point, for each release period, and the i

total dose rate from all four release points are compared to the dose cate limits specified in Technical Specifications 3 15.2.1.b (Unit 1) or 3.ll.2.1.b ( U r. i t 2).

Individual doses due to radioicdine, tritium, and particulates are calculated ior the critical receptor, which for Plant Eatch is an infant exposed to the grass-cow-milk, 13halation, and ground-plane pathways.

Individuol doses are calculated for occh release period ar,d curulative totals are kept for eacn unit for the current calendar quarter and year.

Cumulative Individual dosen are compared to the dose limits specified in Technical Specifications 3.15.2.3 (Unit 1) and 3.11.2.? (Unit 2).

Current percent of technical specification liraitu are chown on tme printout for each release period.

3.

GROSS ALPHA RELEASE The gross alpha release to computed each month by counting the particulate filtere each week for gross alpha activity in a proportional counter.

The four or five weeks' nunbers are then recorded on a data sheet and the settvsty is summed at the er.d at the month.

This concentration is input to t3e composite file of the computer and is used for celease calculations.

4.

ERROR ESTIMATES Regulatory G u.id e 1.21 requires that estimated i

total error lu analysis technigaea be reported.

These est.imates are required for the total fission and activation gas release, total I-131 release, total pa rticulates with balf-lives

[

greater than 8-day release, and total tritium J

release.

I i

l s

l

. q L _

j

h

=. c _

. The total or maximum error associated with the effluent measurement will include the cumulative i

errors resulting from the total operation of sampling and measurement.

Because it may be very difficult to. assign error terms for each parameter affectitig the final measurement, 4

detailed statistical evaluation of error are not suggested.

The objective should be to obtain an

+

overall estimate of the error associated with measurements of radioactive materials released in liquid and gaseous effluents and solid waste."

Estirste1 errots are based on errors in counting eq u ippe r.t calibration, counting statistics,-vent flow rates, v24t sample flow rates, non-steady release raten, chemical yield factors, and sample losses for such items as charcoal cartridges.

1 (1)

Fission and Activat. ion Total Release was I

calculated from sample analysis results and release point flow rates.

l Statistical Error 60%

C'unting Equipment Calibration 104 o

Vent Flow Rates 10%

Non-Steady Release Rates 204 loot (2)

I-131 Release was calculated from each weekly sample:

Statistical Error 60%

j Counting Equipment Calibration 104 l

Vent Flow Rates 10%

i Vent Sample Flow Rates lot Non-Steady Release Rotes 10%

Losses From Charcoal Cartridge 104 110%

i j.

(3)

Pa r ticulates wi'.h nalf-lives greater than 8 days

)

releace was calculated from sample analysis j

results and release point flow rates.

l r

j_

Statistical Error at LLD concentration 604 Counti ng Equipment Calbration 10%

i Vent Flow Rates 10%

[

Vent Sample Flow Rates 10%

Non-Steady Release Rates 104 1004 i

i 4

I i

i

.2b

(4)

Total Tritium Ralease was dominated by the reactor building vent tritium release; hence, the larger statistical errors of the off-gas vent and recombiner building vent tritium releases do not affect the error in the total tritium release Water Vapor in Sample Stream Determination 20%

Vent Flow Rates 104 Counting Calibration and Statistics 10%

Non-Steady Release 50%

904 2.3 GASEOUS EFFLUENT RELEASE DATA Regulatory Guide 1.21 Tables IA, 18, 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 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 rate in uCi/second for each category.

However, the applicable Technical Specification limits are not in terms of release rate in uCi/second but in terms of dose rate in mrem / year, as presented in Technical Specifications 3.15.2.1 (Unit 1) and 3.11.2.1 (Unit 2).

Noble gases are limited an specified in 3.15.2.1.a and 3.ll.2.1.a.

The other three categories (tritium, radioiodines, and particulates) are limited as a group as specified in 3.15.2.1.b and 3.ll.2.1.b.

Further the limits specified in Technical Specifications 3.15.2.1 and 3.11.2.1 are site limits, not unit limits.

Dose rates due to noble gas releases and due to radioiodine, 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.

Limits 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.

, ]

Limits for cumulative individual doses, due to radiolodine, tritium, and particulates, are specified in Technical Specifications 3.15.2.3 (Unit 1) and 3.11.2.3 (Unit 2). 'These limits are also unit limits.

Cumulative individual doses are presented in Tables 2-7a and 2-7b, with percent of technical specification limits.

2.4 RADIOLOGICAL IMPACT DUE TO GASEOUS RELEASES Dose cates 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 radioiodine, 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 il and 3/4.11.2.2 (Unit

2). These results are presented in Tables 2-6a and 2-6b.

Cumulative doses to an individual duc to radioiodine, 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-Ja and 2-7b.

Dose rates and doses were calculated using the methodology presented in the plant Hatch Offsite Dose Calculation Manual.

TABLE 2-2a E.

I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES E.

I.

Hatch Unit Quarter Quarter Est. Total Nuclear Plant 1

2 Error 4 A.

Fission &

Activation Gases 1.

Total Ci 1.85E+03 2.70E+03 1.00E+02 Release 2.

Average uCi/sec 2.38E+02 3.43E+02 Release Rate For Period

• 3.

4 of Tech n

Spec Limit B.

Iodines 1.

Total Ci 9.262-05 5.98E-04 1.10E+02 Iodine-131 2.

Average uCi/sec 1.19E-05 7.61E-05 Release

(

Rate For Period

• 3.

t of Tech 4

Spec Limit C.

Particulates 1.

Particulates Ci 3.00E-04 4.21E-04 1.00E+02 with half-lives 8

days l

2. Average uCi/sec 3.86E-05 5.35E-05 Release Rate For Period j'

3.

t of Tech 4

Spec Limit 4.

Gross Alpha Ci 1.33C-06 3.28E-06 l

Radioactiv*

l ity l

D.

Tritium

1. Total Ci 1.66E+00 1.85E+00 9.00E+01 Release 2.

Average uct/sec 2.13E-01 2. 3 5 0 - t' i Release Rate For Period

• 3.

t of Tech Spec Limit

• Technical Specification limita are in termo of dose rate (mrem /yr) and dose Imrem).

See Tablec 2-5. 1-6a, 2+oh, t

2-7a, and 2-7b.

i '

- = - - - - - - - - -. - -

J

TABLE 2-2b E.

I.

HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES E.

I.

datch Unit Quarter Quarter Est. Total Nuclear Plant 1

2 Error %

A.

Fission &

Activation Geses 1.

Total Ci 3.83E+03 4.94E+03 1.00E+02 Release 2.

Average uCi/sec 4.93E+02 6.28E+02 Release Rate For Feriod

• 3.

% of Tech Spec Limit B.

Iodines 1.

Total Ci 7.34E-03 4.16E-03 1.10E+02 Iodine-131 2.

Average uCi/sec 9.44E-04 5.29E-04 Release Rate For Period

• 3.

% of Tech Spec Limit C.

Particulates 1.

Particulates Ci 4.06E-04 5.77E-04 1.00E+02 with half-livec 8

days 2.

Average uCi/sec 5.22E-05 7.34E-05 Release Rate For Period

• 3.

% of Tech Spec Limit 4.

Gross Alpha Ci 1.03E-06 2.32E-07 Radioactiv-ity D.

Tritium 1.

Total C4 3.29E+00 4.73E+00 9.00E+01 delease 2.

Average uCi/sec 4.23E-01 6.02E-01 RcJease aate Ymc Period

• 3.

4 of Tech Spec Limit

• Technical Specification limits are in terms of dose rate 4 m r e n./y r ) and done (mrem).

See Tables 2-5, 2-6a, 2-6b, 2-7a, and 2-70.

TASL3 2-2c E.

I. dATCH NUCLEAR PLANT - SITE SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES E.

I.

Hatch Unit Quarter Quarter Est. Total Nuclear Plant 1

2 Error 4 A.

Fission &

Activation Gases 1.

Total Ci 5.69E+03 7.64E+03 1.00E+02 Release 2.

Average uCi/sec 7.32E+02 9.72E+02 Relear,e Rate For Period

• 3.

4 of Tech Spec Limit B.

Iodines 1.

Total Ci 7.43E-03 4.76E-03 1.10E+02 Iodine-131 2.

Average uCi/sec 9.56E-04 6.05E-04 Release Rate For Period

• 3.

t of Tech t

Spec Limit C.

Particulates 1.

Particulates Ci 7.06E-04 9.98E-04 1.00E+02 with half-lives 8

days 2.

Average uC1/sec 9.08E-05 1.27E-04

.s t

Release i

Rate For Period

• 3.

t of Tech t

Spec Limit 4.

Gross Alpha C1 2.35E-06 3.51E-06 Radioactiv-ity D.

Tritium 1.

Total Ci 4.95E+00 6.58E+00 9.00E+01 I

Release 2.

Average uCi/sec 6.37E-01 8.37E-01 Release Rate For Period

• 3.

4 of Tech t

Spec Limit

• Technical Specification limits are in terms of dose rate (mrem /yr) and dose (mrem).

See Tables 2-5, 2-6a, 2-6b, 2-7a, and 2-7b.

l

,1*

2

• l'g

'9 I

-TABLE 2-3a

/

\\

E.

I.

HATCH NUCLEAR PLANT UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUEN5 RELEASE REPORT IN86 ik ELEVAIED RELEASE *

/

GASEOUS EFFLUENTS Page'1 of 2

>r Continuoud Mode Batch Mode ** d l

Nuclides i

[

Released Unit Quarter'l***0uarter 2 Oun'rter 1 Quarter 2 t

1.

Fission d

i

~

j Gases

\\

t c

1(

a

- 1 Kr-85m C1 0.00E+00

,6.15E+'J0 /

\\

,/

Kr-85 Ci 0.00E+03 0.00E+00 (\\

j, I;

s i,%

' 'j)

Kr-87 Ci 0.00E+00 7.70E+00 Kr-88 Ci 0.00E+00 8.05E+00[.

5

+4 t

Xe-133 Ci 0.00E+00 4.22E+02 t

-1 Xe-135 Ci 0.00E+00-5.55E+01 3

}

/*,

j Xe-135m Ci 0.00E+00 3.35E+02 f' h Xe-138 Ci 0.00E+00 1.64E+01 I

Xe-131m Ci 0.00E+00 1.09C+01 g

Xe-133m Ci 0.00E+00 5.55E+00 Xe-137 Ci 0.00E+00 0.00E+00 1

Ar-41 0.00E+00 1.02E-01 L

f

-4 TOTAL FOR

((,

i PERIOD Ci 0.00E+00 5,66E+02 (k

2.

Iodines I-131-Ci 0.00EF00 5.05E-04 I-133 Ci 0.00E+00 1.81E-04 I-135 Ci 0.00E+00 1.65E-05 TOTAL FOR i

PERIOD Ci 0.00E+00's 7.03E-04 f

• I 16 ;

e' I

4 0-

/

'2

}*

f

/

f,s f

I

,' i t

i y

}

t r

l

/,

l i

\\

4

-34.'

s t

- - -.. - -, -. - ~ -

---

+*;*r

-~-----r+

~ ~ - - ' - - - ~ * - ' - ' ~ - ' * - -~-

b* ' '

rf

) { n..

J' ' #

TABLE 2-3a i#

l[A [f j E.

I. HATCH NUCLEAR PLANT - UNIT 1 l

[)

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS ELEVATED RELEASE

• 4 P d

1. J Page 2 of 2 b r 3, ?

' '/$

J-Continuous Mode Batch Mode **

!4 Nuc1 ides 4i aeleased Unit Quarter 1***0uarter 2 Otarter 1 Quarter 2

?

(

3.

Particu-Ia t e_s_

t, Cr-51 Ci 0.00E+00 0.00E+00 Mn-54 Ci 0.00E+00 0.00E+00 Co-58 Ci 0.00E+00 0.00E+00 3

Co-60 Ci 0.00E+00 0.00E+00

[ zn-65 Ci 0.00E+00 0.00E+00 t

,Si-89 Ci 0.00E+00 9.05E-05 Sr-90 Ci 0.00E+00 5.20E-07 Nb-95 Ci 0.00E+00 0.00E+00 i

i Cs-134 Ci 0.00E+00 0.00E+00 l

Cs-137 Ci 0.00E+00 0.00E+00

gb,{,,

Ba-140-Ci 0.00E+00 1.41E-05 i

La-140 Ci 0.00E+00 7.4SE-06 TOTAL FOR PERIOD Ci 0.00E+00

-1.13E-04

• zeroes in this table indicate that no radioactivity was present above detectable levels.

See Table 2-8 for. typical lower limits of detection for gaseous sample analyses.

• • There are no batch mode radioactive gaseous release pathways at 1

Plant Hatch.

• • • Unit I was not operating during the first quarter; therefore, there were no Unit I releases via the Main Stack.

e 4

~

N I

I lf,h ov i

i

+k l

f) i

.,c y

, ~..., _ _ _ _ - -.. _ - - - - - - - _.. _ -

TABLE 2-3b E.

I.

HATCH NUCLEAR PLANT UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS - ELEVATED RELEASE

• Continuous Mode Batch Mode **

Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter ?

1.

Fission Gases Kr-85 Ci 0.00E+00 0.00E+00 Kr-85m Ci 5.81E+01 3.76E+01 Kr-87 Ci 4.02E+01 1.84E+01 Kr-88 Ci 6.27E+01 4.25E+01 Xe-133 Ci 2.53E+03 1.54E+03 Xe-135 Ci 6.35E+02 1.10E+03 Xe-135m Ci 1.44E+01 1.46E+02 Xe-138 Ci 6.53E+01 2.00E+01 Xe-131m Ci 2.09E+01 1.70E+01 Xe-133m Ci 4.75E+01 3.41E+01 Xe-137 Ci 0.00E+00 0.00E+00 Ar-41 Ci 1.01E+01 3.37E+00 TOTAL FOR PERIOD Ci 3.48E+03 2.962+03 2.

Iodines I-131 Ci 6.20E-03

3. ole-03 I-133 Ci 1.25E-03 1.62E-03 I-135 Ci 1.31E-04 5.82E-04 TOTAL FOR PERIOD Ci 7.58E-03 5.21E-03 3.

Particu-lates Cr-51 Ci 0.00E+00 0.00E+00 Mn-54 Ci 0.00E+00 0.00E+00 Co-58 Ci 0.00E+00 0.00E+00 Co-60 Ci 0.00E+00 0.00E+00 Zn-65 Ci 0.00E+00 0.00E+00 Sr-89 Ci 7.34E-05 1.81E-04 Sr-90 Ci 1.10E-07 1.04E-06 Nb-95 Ci 0.00E+00 0.00E+00 Cs-134 Ci 0.00E+00 0.00E+00 Cs-137 Ci 0.00E+00 0.00E+00 Ba-140 Ci 3.83E-05 2.07E-05 La-140 Ci

3. ole-05 9.85E-06 TOTAL FOR PERIOD Ci 1.42E-04 2.13E-04

• Zeroes in this table indicate that no radioactivity was present above detectable levels.

See Table 2-8 for typical lower limits of detection for gaseous sample analyses.

• • There are no batch mode radioactive gaseous release pathways at Plant Hatch.

TADLE 2-3c E.

I. HATCH NUCLEAR PLANT - SITE SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS - ELEVATED RELEASE

• Continuous ~ Mode Batch Mode **

Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 1.

Fission Gases i

Kr-85 Ci 0.00E+00 0.00E+00 Kr-85m Ci 5.81E+01 4.37E+01 Kr-87 Ci 4.02E+01 2.61E+01 i

Kr-88 Ci 6.27E+01 5.05E+01 Xe-133 C1 2.53E+03 1.96E+03 Xe-135 Ci 6.35E+02 1.15E+03 Xe-135m Ci 1.44E+01 1.79E+02 Xe-138 Ci 6.53E+01 3.63E+01 Xe-131m Ci 2.09E+01 2.78E+01 Xe-133m Ci 4.75E+01 3.96E+01 Xe-137 Ci 0.00E+00 0.00E+00 Ar-41 Ci 1.01E+01 3.47E+00 TOTAL FOR PERIOD Ci 3.48E+03 3.52E+03 4

2.

Iodines I-131 Ci 6.20E-03 3.52E-03 I-133 Ci 1.25E-03 1.80E-03 I-135 Ci 1.31E-04 5.99E-04 TOTAL FOR PERIOD Ci 7.58E-03 5.92E-03 3.

Particu-l lates Cr-51 Ci 0.00E+00 0.00E+00 Mn-54 Ci 0.00E+00 0.00E+00 f

Co-58 Ci 0.00E+00 0.00E+00 Co-60 Ci 0.00E+00 0.00E+00 Zn-65 Ci 0.00E+00 0.00E+00 Sr-89 Ci 7.34E-05 2.72E-04 Sr-90 Ci 1.10E-07 1.56E-06 Nb-95 Ci 0.00E+00 0.00E+00 Cs-134 Ci 0.00E+00 0.00E+00 Cs-137 Ci 0.00E+00 0.00E+00 Ba-140 Ci 3.83E-05 3.48E-05 La-140 Ci 3.01E-05 1.73E-05 l

TOTAL FOR PERIOD Ci 1.42E-04 3.26E-04 l

• Zeroes in this table indicate that no radioactivity was present above detectable levels.

See Table 2-8 for typical lower limits of detection for gaseous sample analyses.

l

• • There are no batch mode radioactive gaseous release pathways at

(

Plant Hatch.

i.., _... -

TABLE 2-4a E.

I. HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS GROUND-LEVEL RELEASES

• Continuous Mode Batch Mode **

Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 1.

Fission Gases Kr-85 Ci 0.00E+00 4.52E+02 Kr-85m Ci 0.00E+00 0.00E+00 Kr-87 Ci 0.00E+00 0.00E+00 Kr-88 Ci 0.00E+00 0.00E+00 Xe-133 Ci 1.85E+03 1.68E+03 Xe-135 Ci 0.00E+00 8.34E-01 Xe-135m Ci 0.00E+00 0.00E+00 Xe-138 Ci 0.00E+00 0.00E+00 Xe-131m Ci 0.00E+00 3.11E-05 Xe-133m Ci 0.00E+00 0.00E+00 Xe-137 Ci 0.00E+00 0.00E+00 Ar-41 Ci 0.00E+00 0.00E+00 TOTAL FOR PERIOD C1 1.85E+03 2.13E+03 2.

Iodines I-131 Ci 9.26E-05 9.29E-05 I-133 Ci 1.19E-05 4.95E-05 I-135 Ci 0.00E+00 0.00E+00 TOTAL FOR PERIOD Ci 1.05E-04 1.42E-04 3.

Particu-lates Cr-51 Ci 0.00E+00 3.69E-06 Mn-54 Ci 1.36E-05 8.93E-06 Co-58 Ci 1.46E-06 0.00E+00 Co-60 Ci 9.82E-05 6.71E-05 Zn-65 Ci 1.13E-04 7.91E-05 Sr-89 C1 1.79E-05 9.16E-05 Sr-90 Ci 0.00E+00 1.59E-06 Nb-95 Ci 0.00E+00 0.00E+00 Cs-134 Ci 5.98E-06 2.47E-06 Cs-137 Ci 5.00E-05 5.38E-05 Ba-140 Ci 0.00E+00 7.85E-09 La-140 Ci 0.00E+00 0.00E+00 TOTAL FOR PERIOD Ci 3.00E-04 3.08E-04

• Zeroes in this table indicate that no radioactivity was present above detectable levels.

See Table 2-8 for typical lower limits of detection for gaseous sample analyses.

• • There are no batch mode radioactive gaseous release pathways at Plant Hatch.

_3g_

TABLE 2-4b UNIT 2 E.

I.

HATCH NUCLEAR PLANT SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES

• Continuous Mode Batch Mode **

Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 l.

Fission Gases Kr-85 Ci 0.00E+00 0.00E+00 Kr-85m Ci 0.00E+00 0.00E+00 Kr-87 Ci 0.00E+00 0.00E+00 Kr-88 Ci 0.00E+00 0.00E+00 s-Xe-133 Ci 2.17E+02 1.92E+03 l

Xe-135 Ci 1.37E+00 5.66E+01 Xe-135m Ci 0.00E+00

'0.00E+00 Xe-138 Ci 0.00E+00 0.00E+00 Xe-131m Ci 0.00E+00 0.00E+00 Xe-133m Ci 0.00E+00 0.00E+00 Xe-137 Ci 0.00E+00 0.00E+00 Ar-41 Ci 0.00E+00 0.00E+00

-TOTAL FOR PERIOD Ci 3.54E+02 1.98E+03 j

2.

Iodines I-131 Ci 1.14E-03 1.15E-03 I-133 Ci 5.69E-04 4.41E-04 I-135 Ci 9.01E-05 4.55E-05 TOTAL FOR PERIOD Ci 1.80E-03 1.64E-03 l

3.

Particu-lates Cr-51 Ci 0.00E+00 0.00E+00 Mn-54 Ci 0.00E+00 0.00E+00 Co-58 Ci 0.00E+00 0.00E+00 Co-60 Ci

3. ole-06 0.00E+00, 2n-65 Ci 2.16E-06 0.00E+00 Sr-89 Ci 2.57E-04 3.61E-04 Sr-90 Ci 2.80E-07 6.24E-07 Nb-95 Ci 2.24E-07 1.33E-06 Cs-134 Ci 0.00E+00 0.00E+00 Cs-137 Ci 1.02E-06 7.36E-07 Ba-140 Ci 0.00E+00 0.00E+00 La-140 Ci 0.00E+00 0.00E+00 TOTAL FOR PERIOD Ci 2.64E-04 3.64E-04

• Zeroes in this table indicate that no radioactivity was present above detectable levels.

See Table 2-8 for typical lower limits of detection for gaseous sample analyses.

• • There are no batch mode radioactive gaseous release pathways at Plant Hatch. -.

TABLE 2-4c SITE i

E.

I.

HATCH NUCLEAR PLANT SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES

• Continuous Mode Batch Mode **

Nuclides Released Unit Quarter 1 Qur.rter 2 Quarter 1 Quarter 2 1.

Fission Gases Kr-85 Ci 0.00E+00 4.52E+02 Kr-85m Ci 0.00E+00 0.00E+00 Kr-87 Ci 0.00E+00 0.00E+00 Kr-88 Ci 0.00E+00 0.00E+00 Xe-133 Ci 2.07E+03 3.60E+03 Xe-135-Ci 1.37E+02 5.74E+01 Xe-135m Ci 0.00E+00 0.00E+00 Xe-138 Ci 0.00E+00 0.00E+00 Xe-131m Ci 0.00E+00 3.11E-05 Xe-133m Ci 0.00E+00 0.00E+00 Xe-137 Ci 0.00E+00 0.00E+00 Ar-41 Ci 0.00E+00 0.00E+00 TOTAL FOR PERIOD Ci 2.21E+03 4.llE+03 2.

Iodines I-131 Ci 1.23E-03 1.24E-03 I-133 Ci 5.81E-04 4.91E-04 I-135 Ci 9.01E-05 4.55E-05 TOTAL FOR PERIOD Ci 1.90E-03 1.78E-03 3.

Particu-lates Cr-51 Ci 0.00E+00 3.69E-06 Mn-54 Ci 1.36E-05 8.93E-06 Co-58 Ci 1.46E-06 0.00E+00 i

Co-60 Ci 1.01E-04 6.71E-05 Zn-65 Ci 1.15E-04 7.91E-05 Sr-89 Ci 2.75E-04 4.53E-04 Sr-90 Ci 2.80E-07 2.21E-06 Nb-95 Ci 2.24E-07 1.33E-06 Cs-134 Ci 5.98E-06 2.47E-06 Cs-137 Ci 5.10E-05 5.45E-05 Ba-140 Ci 0.00E+00 7.85E-09 La-140 Ci 0.00E+00 0.00E+00 TOTAL FOR PERIOD Ci 5.64E-04 6.72E-04

• Zeroes in this table indicate that no radioactivity was present above detectable levels.

See Table 2-8 for typical lower limits of detection for gaseous sample analyses.

• • There are no batch mode radioactive gaseous release pathways at Plant Hatch....

TABLE 2-5 E.

I.

HATCH NUCLEAR PLANT - SITE SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 GASEOUS EFFLUENTS - DOSE RATES Dose Rates Due to Noble Gases Organ Tech Unit Quarter 4 of Quarter

% of Spec 1

Tech 2

Tech Limit Spec Spec Limit Limit Total 500 mrem /yr 7.77E-01 1.55E-01 1.04E+00 2.08E-01 Body Skin 3000 mrem /yr 1.80E+00 6.00E-02 2.97E+00 9.90E-02 Dose Rates Due to Radioiodine, Tritium, and Particulates Organ Tech Unit Quarter

% of Quarter

% of Spec 1

Tech 2

Tech Limit Spec Spec Limit Limit Bone 1500 mrem /yr 2.82E-04 1.88E-05 5.40E-04 3.60E-05 Liver 1500 mrem /yr 4.94E-03 3.29E-04 6.26E-03 4.17E-04 Total 1500 mrem /yr Body 4.88E-03 3.25E-04 6.21E-03 4.14E-04 Thyroid 1500 mrem /yr 2.56E-02 1.71E-03 2.60E-02 1.73E-03 Kidney 1500 mrem /yr 4.95E-03 3.30E-04 6.26E-03 4.17E-04 Lung 1500 mrem /yr 6.10E-03 4.07E-04 7.52E-03

5. ole-04 GI-LLI 1500 mrem /yr 4.88E-03 3.25E-04 6.22E-03 4.15E-04._ __

TABLE 2-6a E.

I.

HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 AIR DOSES DUE TO NOBLE GAS RELEASES Cumulative Doses Per Quarter Type

. Tech T7. i t Quarter 4 of Quarter 4 of of Spec 1

Tech 2

Tech Radi-Limit Spec Spec j

ation Limit Limit 1

Gamma 5.0 mrad 1.41E-01 2.82E+00 1.31E-01 2.62E+00 i

Beta 10.0 mrad 4.19E-01 4.19E+00 5.74E-01 5.74E+00 Cumulative Doses Per Year Type of Tech Unit Quarters

% of Tech Radiation Spec 1 & 2 Spec Limit Limit Gamma 10.0 mrad 2.72E-01 2.72E+00 Beta 20.0 mrad 9.93E-01 4.97E+00 f

4 1

9 f

i l

i - -

4 TABLE 2-6b E.

I.

HATCH NUCLEAR PLANT - UNIT 2 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986

)

AIR DOSES DUE.TO NOBLE GAS RELEASES Cumulative Doses Per-Quarter Type Tech Unit Quarter 4 of Quarter 4 of of Spec 1

Tech 2

Tech Radi-Limit Spec Spec ation Limit Limit Gamma 5.0 mrad 8.05E-02 1.61E+00 1.78E-01 3.56E+00 Beta 10.0 mrad 1.39E-01 1.39E+00 4.81E-01 4.81E+00 Cumulative Doses Per Year Type of Tech Unit Quarters 4 of Tech Radiation Spec 1 & 2 Spec Limit Limit Gamma 10.0 mrad 2.59E-01 2.59E+00 Beta 20.0 mrad 6.20E-01 3.10E+00 B

d 4.. -.., _ _ _ _

., _ - -. _ _ = _ - - _ _, _ _ _ - _ - -. -.

r TABLE 2-7a E.

I.

HATCH NUCLEAR PLANT - UNIT 1 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 INDIVIDUAL DOSES DUE TO RADIOIODINE, TRITIUM, AND PARTICULATES IN GASEOUS RELEASES Cumulative Dose Per Quarter Organ Tech Unit Quarter

% of Quarter

% of Spec 1

Tech 2

Tech Limit Spec Spec Limit Limit Bone 7.5 mrem 3.938-04 5.24E-03 4.75E-04 6.33E-03 Liver 7.5 mrem 6.19E-04 8.25E-03 5.81E-04 7.75E-03 Tot. Body 7.5 mrem 3.49E-04 4.65E-03 3.17E-04 4.23E-03 Thyroid 7.5 mrem 6.19E-03 8.25E-02 1.79E-02 2.39E-01 Kidney 7.5 mrem 4.04E-04 5.39E-03 3.91E-04 5.21E-03 Lung 7.5 mrem 2.98E-04 3.97E-03 2.63E-04 3.51E-03 GI-LLI 7.5 mrem 3.77E-04 5.03E-03 3.12E-04 4.16E-03 Cumulative Dose Per Year Organ Tech Unit Quarters

% of Tech Spec 1 & 2 Spec Limit Limit Bone 15.0 mrem 8.68E-04 5.79E-03 Liver 15.0 mrem 1.20E-03 8.00E-03 Total Body 15.0 mram 6.66E-04 4.44E-03 Thyroid 15.0 mrem 2.412-02 1.61E-01 Kidney 15.0 mrem 7.95E-04 S.30E-03 Lung 15.0 mrem 5.61E-04 3.74E-03 GI-LLI 15.0 mrem 6.89E-04 4.59E-03 i.

TABLE 2-76 E.

I.

HATCH NUCLEAR PLANT - UNIT 2 1-SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 1986 INDIVIDUAL DOSES DUE TO RADIOIODINE, TRITIUM, AND PARTICULATES IN GASEOUS RELEASES l

Cumulative Dose Per Quarter Organ Tech Unit Quarter 4 of Quarter 4 of Spec 1

Tech 2

Tech Limit Spec Spec Limit Limit Bone 7.5 mrem 7.94E-04 1.06E-02 7.16E-04 9.55E-03 Liver 7.5 mrem 8.63E-04 1.15E-02 7.13E-04 9.51E-03 Tot. Body 7.5 mrem 4.95E-04 6.60E-03 4.~75E-04 6.33E-03 Thyroid 7.5 mrem 2.17E-01 2.89E+00 1.44E-01 1.92E+00 Kidney 7.5 mrem 9.70E-04 1.29E-02 7.85E-04 1.05E-02 Lung 7.5 prem 2.09E-04 2.79E-03 2.86E-04 3.81E-03 GI-LLI 7.5 mrem 2.29E-04 3.05E-03 2.95E-04 3.93E-03 Cumulative Dose Per Year Organ Tech Unit Quarters

% of Tech Spec 1 & 2 Spec Limit Limit Bone 15.0 mrem 1.~ 5 1 E - 0 3 1.01E-02 Liver 15.0 mrem 1.58E-03 1.05E-02 Total Body 15.0 mrem 9.70E-04 6.47E-03 Thyroid 15.0 mrem 3.61E-01 2.41E+00 Kidney 15.0 mrem 1.76E-03 1.17E Lung 15.0 mrem 4.95E-04 3.30E-03 GI-LLI 15.0 mrem 5.24E-04 3.49E-03 l

l i

I l

TABLE 2-8 LOWER LIMITS OF DETECTION - GASEOUS SAMPLE ANALYSES The values in this table represent apriori lower limits of detection (LLD) which are typically achieved in laboratory analyses of gaseous radwaste samples.

RADIONUCLIDE LLD UNITS Kr-87 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 I-131 4.37E-14 I-133 6.16E-13 Mn-54 2.78E-14 Fe-59 4.62E-14 Co-58 2.46E-14 Co-60 2.88E-14 Zn-65 7.51E-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 Sr-89 1.17E-14 Sr-90 3.82E-15 H-3 5.82E-12

3.

SOLID WASTE 3.1 REGULATORY REQUIREMENTS The Technical Specifications presented in this section are for Unit 1.

Requirements for Unit 2 are the same as for Unit it however, the Technical Specification numbers are not the same.

TECHNICAL SPECIFICATIONS 3.15.3.1 The solid radwaste system shall be used in accordance with the PROCESS CONTROL PROGRAM to provide for the SOLIDIFICATION of wet solid wastes and for the SOLIDIFICATION and packaging of other radioactive wastes, as required, to ensure the meeting of the requirements of 10 CFR Part 20 and of 10 CFR Part 71 prior to shipment of radioactive wastes from the site.

6.9.1.9 states in part:

The Radioactive Effluent Release Report shall include the following information for each type of solid waste shipped offsite during the report periods a.

Container volume b.

Total curie quantity (specify whether determined by measurement or estimate) c.

Principal radionuclides (specify whether determined by measurement or estimate) d.

Type of waste, e.g.,

spent resin, compacted dry waste, evaporator bottoms e.

Type of container, e.g.,

LSA, type A, type B,

large quantity f.

Solidification agent, e.g.,

cement.

3.2 SOLID WASTE DATA Regulatory guide 1.21 Table 3 is found in this report as Table 3-la. and 3-lb.

Table 3-la EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR) January thru June 1986 SOLIO WASTE AND IRRADIATED FUEL SHIPMENTS FOR UNIT I & II SOLIO WASTE. SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

I I

l' l

l l

1. Type of waste l

UNIT l 6 month lEst. Totall l

l l period i ERROR % l l

a. Spent resins, filter studges, evaporator l m3 120.75 E+1 1 l

l bottoms etc.

I Ci l 4.8 7 E+2 l 1.0 E+1 l l

b. Dry compressible waste, contaminated l

m3 141.92 E+1 1 l

l eoulp. etc.

I Ci ll4.4 7 E+01 1.0 E-1 l l

c. Irradiated components, control rods, l

m3 IV.

E I

l l

etc.

I Ci 10 E

I f.

E l

l

d. Other (describe) Oily trash-speedi-dry mi4 m3

'10.18 E+1 l l

l equip. etc.

Solidified Oil, CRD Filters l Ci 3.42 E+01 1.0 E-1 l 2.

Estimate of major nuclide composition (by type of waste) l ISOTOPE l

PERCENT l

CURIES l

l 1

l l

l a.

Zn-65 1

41.05 l

200.22 l

[

Cs-137 l

12.92 1

63.00 l

l Co-60 l

12.55 l

61.20 l

l All Others l

33.48 l

163.27 l

l b.

Zn-65 1

51.96 l

7.52

[

l Co-60 1

27.43 l

3.97 l

l Cs-137 1

6.86 l

0.993 l l

All Others l

13.75 l

1.99 l

l c.

None Shipped this period

\\

l\\

l\\

l l

\\

\\

\\

l N

l N

l l

\\

\\

\\

l N

l

\\

l l

\\

\\

\\

l

\\l N

l d.

Co-60 l

55.38

'l 1.893 l l

Zn-65 1

20.33 l

0.695 l l

.Nb-95 l

9.04 1

0.309 l l

Zr-95 1

4.94 l

0.169 l l

All Others l

10.30 l

0.352 l 3.

Solid Waste Disposition Number of Shipments Mode of Transportation Destination 39 Cask Barnwell

't H"

knml 4.

IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shipments Mode of Transportation Destination 0

N/A N/A

m

_._m_

Y January 1 - June 30 EFFLUENT. AND WASTE DISPOSAL SEMIANNUAL REPORT (YEAR) '1986 SOLID. WASTE AND IRRADIATED FUEL SMIPMENT.

FOR UNIT I & II TABLE 3-13' 5

i TYPE OF CURIE PRINCIPLE BURIAL NUMBER GF VOLOME OF'

. TYPE 8

WASTE QUANTITY NUCLIDES CONTAINER CONTAINERS BACE CON -

SMIPPING f

i 3 CONTAINER DETERMINED DETERMINATION DESCRIPTION SNfrPEC-TAINFk Ft DEWATERED MEASURED MEASURED Sn-65 CARBON STEEL 9.

200 LSA, TYPE A I

RESIN 193.6021' Cs-137,Co-60 LINER

_16 206

(

DENATERED MEASURED MEASURED sn-65 NIGN INTEGRITY 1

195 LSA, TYPE A

[

RESIN 182.8255 Cs-137,Co-60 COSTAINER(Polyt

_7 193 1

DEWATERED MEASURED MEASURED En-65 MIGN INTEGRIst 2

119 LSA, TYPE B f

RESIN 110.9418 Cs-137,Co-60 CONTAINER (Poly)

__,c i

i DEWATERED MEASURED MEASURED En-65 FIBERGLASS 2

19 LSA, TYPE A i

i RESIN 0.3231 Cs-137,Co-60 REINFORCED VESSEL 2_ Pres.Ves)21' DRY ACTIVE ESTIMATED ESTIMATED 3n-65 B-25 Bos 134 92

,LSA, STRONG

[

4 4

WASTE (DAW) 14.4733 Cs-137,Co-60 (Steel)

TIGHT CONTAINER

. t FAVA ESTIMATED ESTIMATEDEn-65 Carbon Steel 2

183 LSA,STC I

CSL 0.0338 Cs-137,Co-60 Liner (6'x6')

+

)

DAW ESTIMATED ESTIMATED B-25 NA 137 LSA,STC g

OVERPACK(Steel)

]

DAW ESTIMATED ESTIMATED B-25 MA 134-LSA,STC OVERPACK(Steel) i DAW ESTIMATED ESTIMATED 3n-65 s-25 9

132 LSA,STC Cs-137,Co-60 OVERPACK(Steel)

}

h DAW ESTIMATED ESTIMATED S-25 NA 126 LSA,STC j)

OVERPACE(Steel)

- r DAW ESTIMATED ESTIMATED B-144 NA 248 LSA,STC 4

BOX (Steel) l DAW ESTIMATED ESTIMATED B-88 NA 108 LSA.STC BOI(Steel)

DAW ESTIMATED ESTIMATED 8-85 NA 109 LSA,STC j

BOX (Steel)

DAW ESTIMATED ESTIMATED.

B-87-6 NA 118 LSA,STC I

3 BOX (Steel)

]

DAW ESTIMATED ESTIMATED P"X STRONG TIGNT NA.

115 LSA,STC l

Cc'7AINER(Steel)

DAW ESTIMATED ESTIMATED OVERPACK NA 222 LSA,STC

)

CONTAINER (Steel)

+

j DAW ESTIMATED ESTIMATED OVERPACE NA 375 LSA,STC 4

CONTAlWER(Steel) i DAW ESTIMATED ESTIMATED DRUM (NIC)

NA 40.2 LSA,STC

}

OVERPACE(Polv) j

'CRD ESTIMATED ESTIMATEDEr-93 DRUM (HIC) 1-38.4 LSA. TYPE A

{

FILTERS 3.1925 Co-60,3n-65,Mb-95 OVERPACK(Poly) l IRRADIATED ESTIMATED MEASURED CARBON STEEL NA 26.2 FSV-1 TYPE S l

COMPONENTS LINER CONTAINER l

• • OTHER ESTIMATED ESTIMATEDEr-95 DOT 17. N (S t e el) 474

7. 5,

LSA, 7A (011) 0.2260 Co-60,En-65,Nb-95 DRUM TYPE A i

• Solidified with cement j

• • 011 was either solidified with cement or absorbent material.

4 i

1 4.

CHANGES TO THE PLANT HATCH ODCM Technical Specification 6.9.1.8 requires in par ~t that changes to the Plant Hatch Offsite Dose Calculation Manual (ODCM) be reported to the Commission in the next Semiannual Effluent Release Report.

On January 30, 1986, the Plant Review Board (PRB) approved two changes to the Plant Hatch Offuite Dese Calculation Manual (ODCM).

These changes are presented in subsections 4.1 and 4.2 along with copies of the affected pages of the Plant Hatch ODCM.

4.1 CHANGE TO PAGE 1.1-8 OF THE PLANT HATCH ODCM Change Note 2 to step 5 of section 1.1.1, page 1.1-8, by deleting the statement "If DF<1, A=

(1/DF)."

Rationale for change:

This is a superfluous condition wnich was inadvertently left in Note 2 following review of the Draft ODCM.

The determination of the value of A as described in Step 5 (prior to Note 2) is adequate and sufficient to assure that LRW monitcr setpoints are determined to ensure that the limits of Technical Specifications 3.15.1.1 8( Un i t 1) and 3.11.1.1 (Unit 2) will not be exceeded.

Effect on dose calculations.and setpoint determinations This change has no effect on doue calcuations.

This change does not affect the basic methodology for determining LRW monitor setpoints.

However, it clarifies the methodology presented by removing a superfluous condition.

(The affected page is attached.)

S ATTACHMENT TO SUBSECTIGN 4.1 s

NOTE 1: The calculated setpoint concentration, c, establishes the base value for the monitor setpoint. However, in establishing the actual monitor setpoint for a particular monitor, background radiation icvels must be considered. Normally, the actual monitor setpoint includes the calculated setp'oint value plus background. Background levels must be controlled such that radioactivity levels in the effluent stream being monitored can be accurately assessed at or below the calculated setpoint value.

NOTE 2: If Of < 1, * - (1/DFrf As stated earlier, if DF = 0, the f

I radiation monitor setpoint should be established as close to background as practicable to prevent spurious alarms and yet alarm should an inadvertent radioactive release occur.

If calculated setpoint values are near actual concentrations planned for-release, it may b,e impractical to set the monitor alarm based on this value.

In this case a new setpoint may be calculated by decreasing the effluent

(

flow, increasing the dilution flow, or by decreasing { C t by further processing of the liquid radwaste planned for release, and by following the methodology presented in Steps 2, 3, and 4.

Within the limits of the conditions stated above, monitor setpoints for liquid radwaste effluent radiation monitors may be determined as follows:

Lionid Radwaste Effluent Radiation Monitor 1011-N007 (Unit 1) or 2011-N007 (Unit 2)

Perform Step 2, solving Equation 3 for DF using the appropriate values in the concentration term from the sample analyses for the particular waste tank batch to be discharged. Then perform Steps 3, 4, and 5 to determine the monitor setpoint.

(s HATCH 00CM, REV 1 5/11/84 1.1-8,

4.2 CHANGE TO PAGE 3.0-4 OF THE PLANT HATCH ODCM Change the Table Notation to Table 3.0-1, page 3.0-4, by deleting the phrase "O.$ miles for" in'the description of the location of Station 172.

)

Rotionale for changer The actual location of Station 172 is approximately 2.3 miles for riverwater, sediment, and clanc.

The r6ference to the location of 0.5 miles for sediment and clans is erroneous.

Ef f ect on dose calculations and setpoint determinations:

This ch&nge has no effect on dose calculations bor setpoint determinations.

(The affected page is attached.)

t i

ATTACHMENT TO SUBSECTION 4.2 4

TABLE 3.0-1 (SHEET 2 0F 2)

RADIOLOGICAL ENVIRONMENTAL SAMPLING LOCATIONS I

Location Distance Sample Number Descriptive Location Direction (miles)

Tvoe

• 301 Toombs Central School N

8.2 0

304 State prison ENE 11.3 AD 304 State prison ENE 10.8 M

309 Baxley substation S

10.0 ADV 311 Johnson 8rothers SW 9.1 M

• Sample Types:

~

A - Airborne radioactivity D - Direct radiation i

M - Milk f

R - River (fish or clams, shoreline sediment, and surface water) i-Y a Vegetation

• • Station 170 is located at approximately 0.8 miles for riverwater,1.1 miles for sediment and clams, and 0.9 miles for fish.

skap sediment and clams, and 1.7 miles for fish. Station 172 is located taken can be rather precisely defined.The location frcm which riverwater locations for clams have to be extended over a wide area to obtain aOften, sufficient quantity; even then the quantity may not be sufficient.

water adds to the difficulty in obtaining clam samples; high water might High also make an otherwise suitable location for sediment sampling unavailahic.

A stretch of the river on the order of a mile or so is generally needed to obtain adequate fish samples.

above represent approximations of the locations about which the catches areT takens 4

J 0687n HATCH 03CM, REV 1 S/11/84 3.0-4 1

l. _____ _.~._,.

Gergia P:wer Comram 333 Pm1mont Actua At!4n's Gwg'a 30300 T2kchona 404 $266526 Madeg /dt ess:

Ibst OfM.st Box 4545 Attanta Gergia 30202 g

-Qeomia Power tre wfghfrc sys!cm L. I Gucwa Man %et Nuc'*;er Safety anc' Ucer sing SL-1174 0692C Septeraber 2,1986 U. S. Nuclear Regulatcry Ccamission

REFERENCE:

Office of Inspection and Enforcement RII: JNG Region II - Suite 2900 50-321/50-366 101 Marietta Street, !N Semi. Annual Atlanta, Georgia 30323 Effluent Report ATTEllTION: Dr. J. Helson Grace Gentlemen:

Pursuant to the Edwin I. Hatch Nuclear Plant Units 1 and 2 Technical Specifications 6.9.1.8 and 6.9.1.9, respectively, Operating Licenses DPR-57 and NPF-5, please find enclcssd two copies of the Semi-Annual Radioactive Effluent Release Report for January 1,1986, through June 30, 1986.

If you have questions or concerns, please contact this office at any time.

Sincerely, WI 0 -L f

L. T. Gucwa LTG/lc Enclosure c: Georgia Power Company U. S. Nuclear Regulatory Commission Mr. J. P. O'Reilly Mr. P. Holmes-Ray, Senior Resident Mr. J. T. Beckham, Jr.

Inspector-Hatch Mr. H. C. Nix, Jr.

Office of Inspection and Enforcement (6)

Mr. D. B. Cochran Mr. W. R. Woodall G0-NORMS State of Georgia American Nuclear Insurers Mr. J. Setser, DNR Mr. L. Cross O?

y{

p

... : ~

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