ML20246L801

Semiannual Radioactive Effluent Release Rept,Jan-June 1989

ML20246L801
Person / Time
Site:	Callaway
Issue date:	06/30/1989
From:	UNION ELECTRIC CO.
To:
Shared Package
ML20246J712	List: ML20246J707 ML20246L801 ML20246L822
References
NUDOCS 8909070046
Download: ML20246L801 (20)

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TABLE OF CONTENTS C

i LliO' INTRODUCTION q

l 2.0' SUPPLEMENTAL INFORMATION  !

-2.1 Regulatory Limits. l l

2.2 Maximum Permissible Concentrations

t 2.3' Average Energy 2.4 Measurements and Approximations of Total Radioactivity 2.5 Batch Releases 2.6 Abnormal Releases 3.0

SUMMARY

OF GASEOUS RADI0 ACTIVE EFFLUENTS

'4.0

SUMMARY

OF LIQUID RADI0 ACTIVE EFFLUENTS 5.0 SOLID WASTES SHIPMENTS 6.0 RELATED INFORMATION 6.1 Unplanned Releases 6.2 Changes to_the Process Control Program 6.3 Changes to the Offsite Dose Calculation Manual 6.4 Major Changes to Radwaste Treatment Systems 6.5 Land Use Census Changes 6.6 Inoperability of Effluent Monitoring Instrumentation Table 1A' Semiannual Summation of Gaseous Releases Table IB Semiannual Airborne Continuous and Batch Releases Table 2A Semiannual Summation of Liquid Releases Table 2B Semiannual Liquid Continuous and Batch Releases Table 3 Solid Vaste and Irradiated Fuel Shipments

' Attachment 1 Offsite Dose Calculation Manual Revision 6

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1.0 INTRODUCTION

This Semiannual Radioactive Effluent Release Report is for Union Electric Company's Callaway Plant and is submitted in accordance with the requirements of Technical Specification 6.9.1.7. The report covers the period fro.a January 1,1989 through June 30, 1989.

This' report includes a summary of the quantities of ' radioactive liquid and gaseous effluents and solid waste released from . the plant. The information is presented in accordance with the format outlined in Appendix B ' of Regulatory Guide 1.21, Revision 1. June 1974.

All liquid and gaseous effluents discharged during this reporting period were in compliance with the limits of the Callaway Plant Technical Specifications.

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2 . ') SUPPLEMENTAL INFORMATION

-2.1 Regulatory Limits ISpecified as follows are the technical specification limits applica-ble to'the release of radioactive material in liquid and gaseous effluents.

2.1.1 Fission and Activation Gases (Noble Gases)

The dose rate due to radioactive noble gases released in gaseous.

effluents from the site to areas at and beyond the site boundary shall be limited to less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin.

The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the site boundary shall be limited-to the following:

a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation and,

-b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

2.1.2 Radiciodine, Tritium, and Particulate The dose rate due to Iodine 131- and 133, tritium and all radionuclides in particulate form with half lives greater than eight (8) days released in r,aseous effluents from the site to areas at and '

beyond the site boundary shall be limited to less than or equal to 1500 mrem /yr to any organ.

The dose to a member of the public from Iodine 131 and 133, tritium,-

and all radionuclides in particulate form with half-lives greater than eight (8) days in gaseous effluents released to areas at and beyond the site boundary shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and,

b. During any calendar year: Less than or equal to 15 mrem to any organ.

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2.1.3 Liquid. Effluents The concentration of radioactive material released in liquid efflu-ents to unrestricted areas shall be limited to the concentrations

,epecified in 10 CFR Part 20 Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For.

dissolved or entrained noble gases, the concentration shall be 1 limited to 2.0E-04 microcuries/ml total activity.

The dose or' dose commitment to an Individual from radioactive materials in' liquid effluents released to unrestricted areas shall be limited:

.a. During any eat endar quarter to less than or equal to 1.5 mrem to the tm . body and less than or equal to 5 mrem to any organ, and

b. During any calendar year to less than or equal to 3 mrem to the whole body and to less than or equal to 10 mrem to any organ.

2.1.4 Uranium Fuel Cycle Sources 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.

2.2 Maximum Permissible Concentrations 2.2.1 The maximum permissible concentration. values specified in 10CFR20, Appendix B, Table II, Column 2 are used to calculate release rates and permissible concentrations of liquid radioactive effluents at the unrestricted area boundary. A value of 2.0E-4 microcuries/mi is used as the MPC for dissolved and entrained noble gases in liquid effluents.

2.2.2 For gaseous ef fluents, maximum permissible concentrations are not directly used in release rate calculations since the applicable limits are stated in terms of dose rate at the unrestricted area boundary.

2.3 Average Energy This is not applicable to the Callaway Plant's radiological effluent technical specifications.

2.4 Measurements and Approximations of Total Radioactivity The quantification of radioactivity in liquid and gaseous effluenta was accomplished by performing the sampling and radiological analy-sis of effluents in accordance with the requirements of Table 4.11-1 and Table 4.11-2 of the Callaway Plant Technical 3

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Specifications (See NUREG-1058, " Technical Specifications, Callaway Plant, Unit'No. 1" (October, 1984)).

Gamma spectroscopy was the primary analysis technique used to i determine the radionuclides composition and concentration of liquid and gaseous effluents. Composite samples were analyzed for Sr-89, Sr-90, and Fe-55 by an independent laboratory. Tritium and alpha 9 were measured for both liquid and . gaseous effluents using liquid scintillation- counting and gas flow proportional counting tech-niques, respectively.

The total radioactivity in effluent releases was determined from the measured concentrations of each radionuclides present and the total volume of effluents discharged. Gross beta or gamma radioactivity measurement techniques were not utilized to approximate the total radioactivity in effluents.

2.5 Batch Releases 2.5.1 Liquid Number of batch releases: 120 Total time period for batch releases: 74,456 minutes Maximum time period for a batch release: 985 minutes Average time period for batch releases: 621 minutes Minimum time period for a batch release: 1 minute Average stream flow during periods of release of effluent into a flowing stream: 48,197 cfs*

• Ref: Letter, L. A. Waite (US Geological Survey) to T. M. Baxter (Union Electric Co.) dated July 6, 1989 2.5.2 Gaseous Total for the Non-Outage Reporting Period Related Number of batch releases 42 37 Total time period for batch releases 79,336 minutes 5,956 minutes Maximum time period for a batch release 54,037 minutes 740 minutes Average time period for batch releases 1,889 minutes 160 minutes Minimum time period for a batch release 36 minutes 36 minutes 2.6 Abnormal Releases 2.6.1 Liquid Number of releases: 0 Total Activity released: 0 4

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- 3 2.6.2 Gaseous

-2.6.2.1 Nuaber of releases: 1 2.6.2.2 . Total Activity released: 3.02E-4Ci

'3.0

SUMMARY

OF GASEOUS RADI0 ACTIVE EFFLUENTS '

3.1 The quantities of radioactive material released in gaseous effluents are summerized in Table 1A and IB. Note that for this reporting period no gaseous effluents were considered as elevated releases.

4.0

SUMMARY

OF LIQUID RADIOACTIVE EFFLUENTS 4.1 The quantities of radioactive material released in liquid effluents are ,

summarized in Table 2A and 2B. i 5.0 SOLID WASTES 5.1 The quantities of radioact?ve material released in shipments of solid ,

waste and irradiated fuel t ansported from the site during the reporting period are summarized in Table 3. The activity and fractional abundance of each nuclide was determined for each waste type by an independent i laboratory based upon radiochemical analysis of samples of that waste  !

type. The curie amount of each nuclide listed in Table 3 was determined  !

as the product of the fractional abundance and the total curies shipped.

Those nuclides which comprise at 1 cast 1% of the total activity for a j particular waste type are presented in Table 3. Additionally, as noted ,

in the " Solid Waste Disposition" section of Table 3, one shipment was  ;

released with eventual disposal at the Barnwell, S.C. disposal facility.  ;

This shipment was consigned initially to a vaste processor for super- I compaction. Shipment for disposal of the supercompacted waste will be j made during.the 2nd half of 1989.

6.0 RELATED INFORMATION 6.1 Unplanned Releases  !

Unplanned reloases are inadvertent or accidental releases of radioactive material, or eleases of radioactive material via normal pathways without a release pernit or proper authorization, or without proper sampling and q analysis, or releases which are conducted in such a manner as to result j in significant deviation from the requirements of the release permit.

There was one unplanned release during the reporting period.

On April 11, 1989, the Auxiliary Boiler was sampled and found to be contaminated. Co-60, Co-58 and Mn-54 were identified by isotopic analy-sis. Plant operations was notified immediately to route drains and l boiler blowdown to liquid radwaste. Attempts were made to blowdown the i boiler to remove the contamination, however, very low levels of activity still remained.. An investigation was made to determine the source of this contamination. It appears that due to activity identified in the Radwaste Aux. Steam Recovery Tank, the contamination originated from a radwaste system. However, a specific source could not be identified.

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The boiler operation was continued throughout Refuel III. During normal operation stream is continuously vented to atmosphere to prevent tripping the boiler when at low load. As steam is needed to operate plant equipment, such as the radwaste evaporators, the automatic vent is closed and no steam is released. An engineering evaluation request was initiated to assess system operation with low level radioactivity present.

To assure that no Tech Spec limits had been approached, a conservative

! dose calculation was performed for a member of the public (Nearest

[ Resident) and the maximum dose rate was calculated at the Site Boundary.

l The dose to a Member of the Public from this release was calculated to be 6.78E-04 mrem (Nearest Resident). Technical Specification 3.11.2.3 states:

"The dose to a Member of the Public from lodine-131 and 133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents rolessed, from cach unit, to areas at and beyond the Site Boundary shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ and _ _

h. During any calendar year: Less than or equal to 15 mrems to any, organ.

The total dose calculated is a very small fraction of a percent of Tech Spec Limits. .In addition, the annual dose to a member of the public is well below the environmental dose limits of 40 CFR 190.

The dose rate calculated at the Site Boundary using the highest release rates was 1.58E-03 mrem /yr. Technical Specification 3.11.2.1 states:

" Tie dose rate due to radioactive material released in gaseous. effluents from the site to areas at and beyond the Site Boundary shall be Ilmited to the following:

b. For lodine-131 and 133, for tritium and for all radionuclides in particulate form with half-lives greater than 8 days:

Less than or equal to 1500 mrems/yr to any organ.

The dose rate calculated is a very small fraction of a percent of our Tech Spec limits.

A 10 CFR 50.59 safety evaluatica was performed to evaluate system operation in accordance with IE Bulletin No. 80-10. The safety evaluation concluded there was no possibility of an accident which is different than any previously evaluated in the FSAR, an unreviewed safety question did not exist and the event did not present a potential safety hazard to plant personnel or the general public. Although the auxiliary boiler was operated for a period of approximately 32 days with low levels of activity present and steam vented automatically during periods of low load, this release did not adversely affect or endanger the health or safety of the general public.

This event and a Safety Evaluation is documented by Request for Resolution (RFR) 06487 Rev. A.

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6.2 Changes to the Process Control Program Revision 7 of the Callaway Plant Process Control Program was reviewed and approved for issue by the <nt site Review Committee on July 2, 1987. . This revision was inadvertently omitted in the July - December, 1987 Semi-Annual Effluent Release Report. There were no changes to the Process Control Prcgram during the period January 1 - June 30, 1989.

An explanation of the changes incorporated into Revision 7 and their impact on the Process Control program follows.

Section 8.0, Radioactive Waste Processing Using Contracted Vendor Ser-vices, was revised to provide specific guidance on what type of process-ing ) packaging of wet radioactive waste by vendor solidification and/or dewatering) may be provided by vendors to satisfy the surveillance

. requirements of Technical Specification 3/4.11.3.c. Additionally, reference to 10CFR61 disposal requirements was included as a specific example of conditions that may warrant exercising contracted vendor services for the processing of Callaway Plant wastes.

These changes were made in response to an identified deficiency in the program during an audit by Union Electric Quality Assurance of Callaway Plant Process Control Program implementation. Previous revisions of the Process Control program had included reference to safety-related cor,-

tracts-for handling of vendor services. Contracted services were deter-mined not to require safety related designation provided oversight by Union Electric management was performed during on-site vendor service activitics.

These changes did not reduce the overall confermance of the program to the requirements governing its intended use. A copy of revision 7 of the

_Callaway Plant Process Control Program is retained in plant files and is avoilable for review.

6.3 Changes to the Offsite Dose Calculation Manual Revision 6 of the Callaway Plant ODCM was issued May 19, 1989. In addition to the following overview of the modifications that were incor-porated into the ODCM, a complete copy of the ODCM (Rev. 6) is included as attachment 1.

A listing of the changes incorporated into Rev. 6 include the following:

1. The calculation of the maximum effluent flow rate (fmax) and the monitor setpoint (c) was revised to utilizt a common safety factor in an inverse relationship. Formerly, the adjustment factor utilized in the setpoint calculations to prevent spurious alarms was related to the number of concurrent discharges and had no effect on the effluent flow rate calculation. This enhancement ensures that as the setpoint is increased by the safety factor in order to prevent spurious alarms, the effluent flow rate is decreased to compensate. The revision to equation 2.6 ensures that the monitor will respond properly to the particular discharge. A description of the methodology for calculating a liquid monitor setpoint for very low levels of gamma emitting nuclides was added to this revision.

For discharges which are below the monitor's background, the 7

setpoint is established'at a value of twice' background. In order to ensure adequate monitor sensitivity, background is limited in' accordance with ANSI N13.10-1974 such that a change of IE-7 pCi/ml will be detected.

Page 3: Definition of e was revised to better describe the

. relationship between setpoint and' effluent / dilution flow.

Pages 5-10: The setpoint calculations were revised to be more accurate; based on radionuclides distribution, yields, cali-bration curves and other monitor parameters. The variables necessary for this change are defined in these pages of the ODCM.

Page 9: The maximum allowable background level, calculated using ANSI N13.10-1974 as the basis, is provided.

Page 10: Equation.2.6 providea the methodology to correct the monitor response for each release based on the energy and concentration of each gamma emitter in the effluent.

2. Tables 1, 4 and 5 were revised to add dose factors for additional nuclides. The calculation of these dose factors is presented in HPCI 89-02. Table 2 was revised to add the bioaccumulation factor for Eu-154, which is documented by calculation ZZ-57.

Page.16-17: Table 1, InFestion Dase Commitment Factor (Ag) for Adult Age Group; Eu-154 was added.

Page 18: Table 2, Bioaccumulation Factor (BFg ) used in the Absence of Site specific data, Eu was added Page 33-35: Table 4. Dose Parameter (Pi) for Radionuclides other than Noble Gases; Eu-154 was added.

Page 42 - 59: Table 5, Pathway Dose Factors (R4 ) for radionuclides other than Noble Gases; Sn-113 and Eu-154 were added to Inhalation Pathway; Be-7, Co-57, Br-82, Cd-109, Sn-113, Sb-124, Sb-125, Eu-154 and Hf-181 were added to the Ground Plane Pathway; Sn-113 and Eu-154 were added to the Meat Pathway; Sn-113 and Eu-154 were adoed to the Grass-Cow-Milk Pathway; Be-7, Co-57, Br-82, Cd-109, Sn-113 Sb-124, Sb-125, Eu-154 and Hf-181 were added to the Grass-Goat-Milk Pathway and Sn-113 and Eu-154 were added to the Vegetation Pathway.

3. The gaseous monitor setpoint calculation methodology was revised to discuss the use of the setpoint required by Specification 4.9.4.2 during core alterations. In addition, this value will a190 be utilized as the setpoint in the event that there is no detectable noble gas activity in the containment atmosphere sample. Although the basis for this setpoint is a fuel handling accident, it is applicable at times not involving fuel movement due to the relatively low activity assumed to be released. Calculation HPCI 87-04 assumes a release of less than 3.2E-5 of the activity that is 8

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.h normally considered to be released-in a fuel handling, accident.. The. -l dose' rate at the site boundary-is therefore: -)

i Calculate'd . .

l Concentration Release Rate -Dose Rate Factor. < Dose Rate ~ .

.Nuclide (pCi/cc) (pCi/sec) (mr/yr) per'(pCi/m 8)' (mr/yr) '

7. - .'

.'Ke-131m 2. 9 6 E-5..- 2.79E2- 9.15El 2.55E-2 )

Xe-133m 5.38E-5 5.07E2 2.51E2 1.27E j 1Xe-133' '.4.70E-3 4.43E4 .2.94E2 1.30E1-Xe-105- -3.84E-6 3.62E1 1.81E3 o.55E-2.

, Kr-85.: 1.79E-4 1.69E3 1.61El 2;72E-2 j TOTAL DOSE RATE = 1.32E1 mrem /yr-Therefore, this setpoint controls offsite. doses to a value much' m

less'than'the limit of 500 mrem /yr.

Page 9: If the liquid effluent sample.contains no= detect-- =l able gamma activity or if-the value.of (E(Cg) SF) is  !

less than the liquid monitor background, the sefpo+ int will.- l be set at twice the liquid moniter background.  ;

l Page 24: When the gaseous sample contains no detectable noble gas activity, the gas monitor setpoint will be set-at a value of less . than or equal to SE-3 pursuant to-Technical Specification 4.9.4.2.

Page 24: During core alternations, the setpoint for the containment purge monitors will be set at a value of less than or equal to SE-3, pursuant to Tech Spec 4.9.4.2.

4.- Minimum holdup requirements for the Waste Gas Decay Tanks were added..

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-Page 60: Normally, waste gases -will be retained for at 'I least 60 dayq prior to discharge.

5. The dispersion parameters given in Tables 9 and 10 were revised' in accordance with the calculations performed for FSAR Change .j Notice 88-42 and documented by calculation ZZ-67. The dis- -l persion parameters for f arming areas within the Site Boundary j were derived from calculation ZZ-67 by sutuning the area-weighted dispersion parameters to form a " composite" dispersion parameter representative of the time spent and activities  ;

performed within the Site Boundary.

Page 85-89: Assumptions made, methodology utilized and "

new table entries. generated are presented on these pages.

6. In addition, several other changes were incorporated into this j revision of the ODCM. They are as follows:  !

-l Page 4: Guidance in the event that an alarm / trip cetpoint j is reached was provided to ensure Technical Specification l 3.11.1.1 limits were not exceeded.

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Page 5: Location'of sampling type and frequency for

' liquid monitors was added.

Page 5: Setpoint control for the Turbine Building Drain is defined as a function of background.

2. The definition of At1 was clarified ~ to state that it corre-sponds to the duration of release vice other time intervals (e.g. one year).

Page 82: The long-term (annual average) dispersion esti-mates now utilize the variable trajectory plume segment atmospheric model MESODIF-Il and the straight-line gaussian dispersion model X0QD0Q rather than .the fluctuating plume model PUFF and the straight-line dispersion model.

6.4 Major Changes to Radwaste Treatment Systems There were no major changes to the. Liquid, Gaseous or Solid Radwaste Treatment Systems during the reporting period. However, one change was implemented which will significantly reduce the activity levels in the Reactor Coolant System (RCS). This change involved the approval to use finer mesh RCS filters to remove corrosion' products

.from the RCS. This change was documented in Callaway Modification Package 88-1018 and was reviewed and approved by the Onsite Review Committee (ORC) on 2-15-89 (Ohc meeting 833s).

-The Chemical Volume Control System Reactor Coolant Filter (FBG06) is designed to collect resin fines and particulate from the letdown stream. Installing the smaller micron filters will reduce the amount of corrosion products in the RCS. To accommodate higher differential pressure (DP) across the filters, the scaling c1 the DP indicator will also be changed. A control room alarm for filter high differential pressure has been added to assure the operator is promptly alerted when a filter requires changeout.

Through a gradual reduction of the filter micron rating from a 30 micron nominal (49 absolute) filter to a 0.45 micron absolute filter, the activity in the RCS will be reduced thereby reducing exposure to personnel - during maintenance activities. This should also result in a significant reduction in liquid effluent activity.

The results of the nuclear safety evaluation state that the components added by the change will have no impact on the Safety Design Basis concerning the design and fabrication codes required by Regulatory Guide 1.29. The smaller mesh filters used for this change will meet the original Westinghouse specifications. The alarm setpoint of 25 paid will allow sufficient time for operator action to valve out the filter before the vendor's design differential pressure limit of 75 psid is reached. As a result, no unreviewed safety question exists for this change.

This change will substantially reduce the activity levels of the RCS and therefore reduce the potential for radiation dose to plant personnel and the public.

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l' 6.5 Land Use Census Changes There were no changes in critical receptor locations for dose calcu-lations during the reporting period.

6.6 Inoperability of Effluent Monitoring Instrumentation All effluent monitoring instrumentation was OPERABLE within the limits specified by Specifications 3.3.3.9 and 3.3.3.10 during the reporting period.

6.7 Jnstances of Liquid Holdup Tanks or Waste Gas Decay Tanks Exceeding Technical Specification Limits All liquid tanks and waste gas decay tanks were within the limits of Specifications 3.11.1.4 and 3.11.2.6 during'the reporting period.

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TABLE 1A SEMIANNUAL SUMMATION OF GASEOUS RELEASES ALL AIRP.ORNE EFFLUENTS QUARTERS 1 AND 2, 1989 UNIT QUARTER 1 QUARTER 2 EST TOTAL TYPE OF EFFLUENT ERROR %

A. FISSION AND ACTIVATION GASES

1. 'IOTAL RELEASE CURIES 6.57E+1 2.30E+2 20

2. AVERAGE RELEASE RATE FOR PERIOD UCI/SEC 8.45 2.93E+1

3. PERCENT OF TECH SPEC LIMIT  % N/A N/A B. RADIOIODINES

1. TOTAL 10 DINE-131 CURIES 1.68E-6 9.69E-5 23

2. AVERAGE RELEASE RATE FOP.FiRIOD UCI/SEC 2.17E-7 1.23E-5

3. PERCENT OF TECH SPEC LIMIT  % N/A N/A C. PARTICULATE

1. PARTICULATE (HALF-LIVES > 8 DAYS) CURIES 4.85E-7 1.19E-5 30

2. AVERAGE RELEASE RATE FOR PERIOD UCI/SEC 6.23E-8 1.52E-6

3. PERCENT OF TECH SPEC LIMIT  % N/A N/A

4. GROSS ALPHA RADI0 ACTIVITY CURIES 3.01E-7 1.77E-6 D. TRITIUM

1. . TOTAL RELEASE CURIES 2.68 1.25E+1 14

2. -AVERAGE RELEASE RATE FOR PERIOD UCI/SEC 3.46E-1 1.58

3. PERCENT OF TECH SPEC LIMIT  % N/A N/A PAGE 1 of 1

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TABLE IB SEMIANNUAL AIRBORNE CONTINUOUS AND BATCH RELEASES GROUND LEVEL RELEASES

' FISSION CASES, 10 DINES, AND PARTICULATE QUARTERS 1 AND 2, 1989 CONTINUOUS RELEASES BATCH RELEASES NUCLIDE UNIT QUARTER 1 QUARTER 2 QUARTER 1 QUARTER 2

1. FISSION GASES Kr-85M CURIES 3.54E-1 5.69E-1 9.94E-3 4.50E-3 Kr-85 CURIES 0 0 1.67E0 2.44E0 Kr-87 CURIES 9.07E-2 0 0 3.17E-4 Kr-88 CURIES 6.90E-1 1.84E-1 1.20E-4 4.41E-3 Xe-131M CURIES 0 3.16E-1 2.90E-2 1.00E-1 Xe-133M CURIES 0 0 4.04E-3 .1.15E0 Xe-133 CURIES 5.45E+1 6.94E+1 2.82E0 1.46E+2 Xe-135M CURIES 9.99E-2 0 0 0 Xe-135 CURIES 4.51E0 8.46E0 9.25E-2 1.04E0 Xe-138 CURIES 0 0 0 0 Ar-41 CURIES 0 0 7.52E-1 3.45E-2 TOTAL FOR PERIOD CURIES 6.03E+1 7.89E+1 5.37E0 1.51E+2

2. IDDINES I-131 CURIES 1.15E-6 3.35E-5 5.34E-7 6.34E-5 I-133 CURIES 8.45E-7 3.03E-6 4.05E-7 1.72E-6 I-135 CURIES 0 0 0 0 Br-82 CURIES 0 0 4.09E-7 0 TOTAL FOR PERIOD CURIES 1.99E-6 3.66E-5 1.35E-6 6.51E-5

3. PARTICULATE H-3 CURIES 1.92E0 4.85E0 7.61E-1 7.60E0 Co-58 CURIES 0 0 0 0 Co-60 CURIES 0 1.19E-5 0 0 Rb-88 CURIES 0 0 0 0 Sr-89' CURIES 0 0 0 0 Cs-134 CURIES 0 0 0 0 Cs-137 CURIES 0 0 0 0 Ce-141 CURIES 4.85E-7 0 0 0 Sr-90 CURIES 0 0 0 0 G ALPHA CURIES 3.01E-7 5.32E-7 0 1.24E-6 TOTAL FOR PERIOD CURIES 1.92E0 4.85E0 7.61E-1 7.60E0 PAGE 1 0F 1

TABLE 2A SEMIANNUAL SUMMATION OF LIQUID RELEASES ALL LIQUID EFFLUENTS QUARTERS 1 'ND 2, 1989 UNIT QUARTER 1 QUARTER 2 EST TOTAL TYPE OF EFFLUENT ERROR %

A .' FISSION AND ACTIVATION PRODUCTS

1., TOTAL RELEASE (NOT INCLUDING TRITIUM, CASES, ALPHA) CURIES 5.23E-3' 1.84E-3 20

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD UCI/ML 4.09E-9 3.99E-9

3. . PERCENT OF APPLICABLE LIMIT  % 7.47E-2 4.74E-1 B. TRITIUM

1. TOTAL RELEASE CURIES 1.37E+2 6.85+1 14

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD UCI/ML 1.07E-4 1.49E-4

.3. PERCENT OF APPLICABLE LIMIT  % 3.56E0 4.95E0 C.: DISSOLVED AND ENTRAINED CASES

1. TOTAL RELEASE CURIES 2.99E-1 6.85E-2 27

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD UCI/MI 2.34E-7 1.49E-7

3. PERCENT OF APPLICABLE LIMIT  % 1.17E-1 7.44E-2 D. GROSS ALPHA RADIOACTIVITY

1. TOTAL RELEASE CURIES 2.34E-3 3.01E-4 29 E. WASTE VOL RELEASED (PRE-DILUTION) GAL 6.065E6 4.419E6 10 F. VOLUME OF DILUTION WATER USED GAL 3.314E8 1.174E8 10 PAGE 1 0F 1 i

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l TABLE 2B SEMIANNUAL LIQUID CONTINUOUS AND BATcti kELEASES TOTALS FOR EACH NUCLIDE RELEASED QUARTER 1 AND 2, 1989 CONTINUOUS RELEASES BATCH RELEASES NUCLIDE UNIT QUARTER 1 QUARTER 2 QUARTER 1 QUARTER 2 ALL NUCLIDES H-3 CURIES 0 0 1.37E+2 6.85E+1

.Na-24 CURIES O O O O Cr-51 CURIES 0 0 0 6.84E-5 Mn-54 CURIES 0 0 3.57E-6 5.15E-5 Fe-55 CURIES 0 0 2.09E-3 O Fe-59 CURIES O O O O Co-58 CURIES 0 0 0 1.26E-4 Co-60 CURIES 0 0 2.74E-4 8.76E-4 Zn-65 CURIES 0 0 0 0 Rb-88 CURIES 0 0 0 0 Sr-89 CURIES 0 0 2.82E-3 0 Zr-95 CURIES 0 0 0 0 Nb-95 CURIES 0- 0 0 0 Mo-99 CURIES 0 0 0 0 Tc-99M CURIES 0 0 0 0 Ag-110m CURIES 0 0 0 0 1-131 CURIES 0 0 0 4.73E-5 I-133 CURIES 0 0 0 0 1-135 CURIES 0 0 0 0 Cs-134 CURIES 0 0 9.41E-6 1.79E-5 Cs-136 CURIES 0 0 0 0 Cs-137 CURIES 0 0 2.56E-5 5.24E-5 La-140 CURIES 0 0 0 0 Ce-141 CURIES 0 0 0 0 Ce-144 CURIES 0 0 0 0 W-187 CURIES 0 0 0 0 Kr-85 CURIES 0 0 0 0 Kr-85M CURIES 0 0 0 0 Xe-131M CURIES 0 0 1.93E-3 4.02E-4 Xe-133 CURIES 0 0 2.92E-1 6.70E-2 Xe-133M CURIES 0 0 3.61E-3 6.04E-4

.Xe-135 CURIES 0 0 1.64E-3 5.48E-4 Xe-135M CURIES 0 0 0 0 Ba-140 CURIES 0 0 0 0 Sr-90 CURIES 0 0 0 5.98E-4 G ALPHA CURIES 0 0 2.34E-3 3.01E-4 UNIDENTIFIED CURIES 0 0 0 0 TOTAL FOR PERIOD CURIES C 0 1.37E+2 6.86E+1 p

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TABLE 3 I SOLID WASTE & IRRADIATED FUEL SHIPMENTS QUARTERS 1 & 2, 1989 A. SOLID ' WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (DOES NOT INCLUDE' IRRADIATED FUEL)

TYPE OF WASTE 6-MONTH EST. TOTAL PERIOD ERROR (%)  ;

I

a. Spent resins, filter sludges 46.5 m8 125% I evaporator bottoms, etc. 385.0 Ci Cs-137 20.50% 7.88E1 Ci Cs-134 8.68% 3.34E1 Ci Fe-55 26.33% 1.01E2 Ci  ;

Co-60 26.92% 1.04E2 Ci i Ni-63 13.00% 5.00E1 Ci  ;

Mn-54 -3.23% 1.24E1 Ci I

b. . Dry compressible vaste, 17.0 m 8 125%

contaminated equipment, etc. 1.38 Ci Cr-51 2.89% 3.98E-2 Ci co-58 27.16% 3.74E-1 Ci  !

Fe-55 45.52% .6.26E-1 Ci l Nb-95 3.18% 4.39E-2 Ci Zr-95 2.34% 3.22E-2 Ci Co-60 .11.98% 1.65E-1 Ci Ni-63 3.15% 4.34E-2 Ci Mn-54 3.64% 5.01E-2 Ci ,

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c. Irradiated components, control 0 m8 rods, etc. O Ci

d. Other 0 m8 0 Ci l

~ Solid Waste Disposition l

l Number of Mode of Class of Solid Type of Shipments Transportation Destination Waste Shipped Container 4- Truck Richland, WA A LSA ,

3 Cask Barnswell, S.C. B LSA !

1* Truck Barnswell, S.C. A LSA

• Shipped to vaste processor 1st half, 1989. To be disposed of in 2nd half l 1989.

Page 1 of 2

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TABLE 3 (cont.)

Solidification Agent Cement (applicable to waste type'"A" only).

4 B. ' IRRADIATED FUEL SHIPMENTS (DISPOSITION)'

Number of Shipments ' Mode of Transportation Destination 0' N/A N/A 1

Page 2 of 2

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ATTAC11 MENT 1  :

OFFSITE DOSE CALCULATION MANUAL ,

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'I The attached is a complete copy of the Callaway- , .

Plant Offsite Dose Calculation Manual (rev. 6, May.1989) 1

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