ML20246J693

Semiannual Radioactive Effluent Release Rept for Jan-June 89089

ML20246J693
Person / Time
Site:	Fort Saint Vrain
Issue date:	06/30/1989
From:	Crawford A PUBLIC SERVICE CO. OF COLORADO
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
P-89311, NUDOCS 8909050241
Download: ML20246J693 (34)

Text

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Public Service'  ; % "l 1 ,. .

P.O. Box 840 -

Denver CO 80201 0840 A. Clegg Crawford Vice President Nuclear Operations August 15, 1989-Fort St. Vrain Unit No. 1 P-89311 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washing +sn, DC 20555 Docket No. 50-267

SUBJECT:

Semi-Annual Radioactive Effluent Release Report Gentlemen:

Attached please find the Semi-Annual Radioactive Effluent Release Report for the Fort St. Vrain Nuclear Generating Station.

This report covers the period January 1,1989 through June 30, 1989, and is submitted pursuant to Section 7.5.1.e of the Fort St. Vrain Technical Specifications.

Please contact Mr. M. H. Holmes at (303) 480-6960 if you have any questions regarding this report.

Sincerely, A?yb A. C. Crawford, Vice President, Nuclear Operations Fort St. Vrain Nuclear Generating Station ACC:VHF/bw Attachments 8909050241 690630 PDR ADOCK0500g7

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.P-89311 August 15, 1989

'cc w/ attachments:

Regional Administrator, Region IV U. S. Nuclear Regulatory Commission Attn: Mr. T. F. Westerman, Chief Projects Section B 611 Ryan Plaza Drive, Suite'1000 Arlington, Texas. 76011 Mr. Robert Farrell Senior Resident Inspector NRC Office Fort St. Vrain Dottie Sherman, ANI Library American Nuclear Insurers The Exchange, Suite 245 270 Farmington Avenue Farmington, Connecticut 06032 l

Mr. Robert Quillen, Director Radiation Control Division Colorado Department of Health 4210 E. lith Ave.

Denver, CO 80220 Mr. Mike Dolphin, Site Manager General f.tomic Company NED Site Fort St. Vrain Mr. Milt Lammering, Chief Regional Representative, Radiation Program U. S. Environmental Protection Agency Region VIII 999 18th Street, Suite 500 Denver, CO 80202-2405 Mr. Farrell D. Hobbs Environmental Manag2 ment Rockwell International Building 250, P.O. Box 464 Golden, Co 80402 L_ _ __ _ _ _ _ _

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V P-89311 August 15, 1989 Dr. James E. Johnson Radiology & Radiation Biology Dept.

135 BRB Colorado State University Fort Collins, CO 80523 t-1 i

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SEMI-ANNUAL RAv10 ACTIVE EFFLUENT.

RELEASE REPORT January - June 1989 Public Service Company of Colorado Fort St. Vrain Nuclear Generating Station August 1989 i

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P-89311 August 15, 1989 1.0

SUMMARY

This report summarizes the radiological effluent released from the Fort St. Vrain Nuclear Generating Station for the period of January through. June, 1989. This information is provided pursuant to the requirements of Sections 7.5.1.e, 8.1.1.g.8, 8.1.2.d, e and j, 8.1.3.e and f, and 8.2.1.h.1 of the Fort St. Vrain Technical Specifications.

This report follows the reporting format recommended by Regulatory Guide 1.21 as well as the requirements of the aforementioned sections of our Technical Specifications.

The following tables with a supplemental information section are included with this report:

Table Description 1A Gaseous Effluents - Summation of All Releases 1C Gaseous Effluents - Ground-Level Releases 2A Liquid Effluents - Summation of All Releases 2B Liquid Effluents 3 Solid Waste and Irradiated Fuel Shipments 4A Hourly Meteorological Data Please note that Table IB (of Regulatory Guide 1.21) has been omitted from this report because all of our gaseous effluents are assumed to be ground-level releases as opposed to being elevated releases.

Fort St. Vrain Technical Specifications apply exclusively to the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions.

This list .does not mean that only these nuclides are considered.

Other gamma emitting nuclides that are identifiable, together with those of the above nuclides, are analyzed and included in this report.

Sample activities that are less than the detection capabilities of our equipment are entered in this report using the value resulting from the calculation of the lower limit of detection (LLD) or minimum detectable activity (MDA). This results in reporting upper limit values that are in excess of true activities.

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R P-89311 August 15, 1989 l' The lower. limit of detection (LLD), for the purpose of this report, is defined as the smallest concentration of radioactive material in a sample that will yield a net cotnt, above the system background, that

, will be detected with a 95% probability of being correct and' only a L 5% probability of falsely concluding that a blank observation represents a real signal. The LLD values specified in our Technical Specifications are as follows:

, Liquid Principle Gamma Emitters 5.00E-07 pCi/mi Dissolved Noble Gases 1.00E-05 pCi/ml Tritium 1.00E-05 pCi/mi l Iodine-131 1.00E-06 pCi/mi Gross Alpha 1.00E-07 pCi/mi Strontium-89,90(Composite) 5.00E-08 pCi/ml Gaseous Principle Gamma Emitters 1.00E-04 pCi/cc (Gas)

Principle Gamma Emitters 1.00E-11 pCi/cc (Particulate)

Tritium (Gas) 1.00E-06 pCi/cc Iodine-131(Charcoal) 1 00E-12 pCi/cc Gross Alpha (Particulate) 1.00E-11 pCi/cc Strontium-89,90(Particulate) 1.00E-11 pCi/cc i

Gross-Beta (Particulate) 1.00E-11 pCi/cc Where applicable, we have listed "less-than" values for those nuclides listed specifically in our Technical Specifications. These "less-than" values were calculated using the observed LLD values and the total volume of the media. The "less-than" values were not included in the total values for the pathway.

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P-89311 August 15, 1989 The percent of. Technical Specification limit on Table 1A.is blank in some cases because'this value could not be calculated from data which were at _or_below the minimum detectable activity. On Table 1C, the continuous release mode values are not reported because this release pathway is the same as the batch mode. All other blanks on Tables IC and 28 occur because no LLD values for these nuclides are required to be.. calculated per Technical Specifications.

There has been some confusion in the past as to the total volume of g

water used for dilution of radioactive liquid effluent. All average diluted concentrations are based on the activity at the unrestricted area. Although this effluent could eventually reach one of two rivers (St. Vrain Creek and. South Platte River) which converge approximately one and one half miles downstream -of the plant, no further dilutions were assumed. Additional discussion on river flow is contained in section 4a of the Supplemental Information Section.

On March 12, 1989 a planned release of primary coolant was made from the' Prestressed Concrete Reactor Vessel (PCRV) as part of an

' intentional- evacuation process in conjunction with the removal of moisture from the core. The release was made through the : Reactor Building . Exhaust LStack and was monitored throughout the entire release. Total activity release was calculated to be 2.58 E-03 C1.

Although the release was both planned and monitored, it was reported as an abnormal release to fully document the activity released to the environs.

A second planned and monitored release of primary coolant was made through-the Reactor Building Exhaust Stack on April 29, 1989. The release was made to depressurize the PCRV in preparation for the removal of 2 control rods for repair. Total activity released was

' 7. 0 E-05 Ci. The release was reported as an abnormal release to document fully the activity released to the environs.

On June 28, 1989, it was discovered that an unplanned gas waste release had occurred from'the 1A gas waste surge tank. This release started when plant personnel intended to initiate a release from the IB' gas waste surge tank. An error in the valve lineup caused the wrong tank to be discharged. The release was terminated when the error was discovered and the IA tank was sampled and analyzed for activity levels. Total activity released to the environs was calculated to be 2.51 E-02 C1. This event was reported to the Nuclear Regulatory Commission in Licensee Event Report 89-12.

There were no unplanned radioactive liquid waste releases made from the Liquid Waste System 62 or the reactor building sump during this reporting period.

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P-89311 August 15, 1989 The following discussion correlates specific points mentioned in Fort St. Vrain Technical Specification 7.5.le to the contents of the Semi-annual Effluent Report.

Three shipments of radioactive waste were made this reporting period.

1 One shipment consisted. of dry active waste, processed by a vendor, and shipped from their facility to the disposal site. Dry active waste consisted of contaminated miscellaneous waste, both compacted and noncompacted. This shipment was determined to be Low -Specific Activity, Class A (unstable) wad.e, and was packaged in strong, tight, or LSA containers. The principal radionuclides and curie content for dry active waste was estimated by the WASTETRAK computer code. WASTETRAK calculated the concentration of gamma emitting l 'i sotopes based on the nuclide distribution determined from representative samples as part of the 10CFR61 Waste Classification program, and from the measured radiation level on the container.

After applying the appropriate scaling factor to obtain the concentration for difficult to measure nuclides, WASTETRAK calculated a total curie content for the container.

The total volume and curie quantity for dry active waste are given in Table 3.

The other two shipments consisted of irradiated materials. Both shipments were composed of three control rod absorber strings and one instrumented control rod string generated during control rod drive

, orifice valve assembly refurbishment. Principal radionuclides and

' curie content for irradiated components were estimated by an activation analysis performed by a vendor. The vendor estimated the accuracy of the activation analysis to be 20%. It was determined that waste in both shipments exceeded a Type A quantity of radioactive . material and both were classified as Class C (stable) waste. The waste was packaged in a strong, tight container and

. shipped to the disposal facility in a Type B container.

l Total volumes and curie quantities for irradiated components are I

given in Table 3.

For all shipments made this reporting period, no solidification agents or absorbent material were employed.

There were no major changes made to the radioactive waste systems during this reporting period.

P-89311 August 15, 1989 There were no changes to the Process Control Program (SUSMAP-3),

Issue 2, effective date November 13, 1984, during this reporting period.

A change to the Offsite Dose Calcult' ion Manual (0DCM), SUSMAP-2, Issue 16, effective date May 3,1988 was nade per PDR 89-0258 dated March 30, 1989. The change involved a sample location description for the Radiological Environmental Monitoring Program milk sampling site R-8. A new sampling site location was added when the previous dairy went out of business.

Gaseous effluent activity monitors, or their associated recorders, found to be inoperable for more than thirty days ending during the period of January 1, 1989 to June 30, 1989, are required to be reported per Environmental Limiting Conditions for Operation (ELCO) 8.1.1.a.

The Auxillary Cooling Method (ACM) powered backup reactor building exhaust stack monitors RT/RIS-4801, RT/RIS-4082, and RT/RIS-4803, were out of service for periods ranging from over six months to over ten months (differs for each of the three channels). The particulate and iodine channels (-4801 and -4802) were returned to service on April 26, 1989, and the noble gas channel (-4803) was returned to service June 12, 1989. Despite the assignment of a technician to these monitors fulltime during most of the period of nonoperability, progress was slow for a variety of reasons:

The monitors are obsolete and are no longer supported by the manufacturer. They had numerous "known bugs" which the manufacturer corrected on later models but left undocumented and lack of written calibration procedures which had to be developed all contributed to the long repair time. There were numerous failures of the monitors, many of which are suspected of being due to excessive ambient temperatures over a long period of time. Action is being taken to replace these monitors with more recent model s and to address the temperature problem. The primary stack monitors were relied upon during the period the ACM backup monitors were inoperable.

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h P-8931'l- . August 15, 1989 Gaseous effluent flow measurement instruments are also required l to be reported if inoperable for over thirty days, per ELCO 8,1.1.g.8. Of these instruments, only FT-6351, the radioactive gas waste surge tank to the reactor building exhaust flow transmitter had a maintenance request- outstanding for over thirty days. The transmitter response was found to be non-linear and could not pass the surveillance calibration test.

The Non-Conformance Report (NCR)88-156 was dispositioned to use this equipment "as is" temporarily with a calibration curve posted for operators to determine the actual flow. The instrument was in this condition for about seven months for undetermined reasons, even though parts were soon available. It was repaired and verified to be in calibration on January 16, 1989. By use of the posted calibration curve, the instrument was consider-d operable.

Liquid effluent activity monitors inoperable for more than thirty days are required to be reported per ELCO 8.1.2.d. and ELCO 8.1.3.e One of the two radioactive liquid waste effluent monitors, RT/RIS-6213, was inoperable due to being found out of calibration, reading low, for a period of 41. days from date of discovery. This was during a period of time just before plant start-up when several monitors were awaiting repair. Lack of technician manpower qualified to repair radiation monitors was the primary cause of the longer repair time. The compensatory actions required by ELC0 8.1.2.d during liquid waste releases are routinely performed regardless of monitor operability.

Samples were taken every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during reactor building sump releases, as required by ELCO 8.1.3.e. and analyzed to determine activity levels.

The recorder associated with the liquid effluent activity monitors, reportable under ELC0 8.1.2.e and ELCO 8.1.3.f, was not out of service over thirty days.

Although not necessarily required to be reported by a literal reading of ELCO 8.1.2.d, the liquid effluent flow monitoring equipment was inoperable for about nine months until January 24, {

1989, due to FE-4101-1 being broken with no parts available, j During this time a jumper was installed under Temporary Configuration Report (TCR) 88-09-01 to bypass the control function of FS-4101 to allow liquid effluent releases to be made. Flow was measured during these releases by use of a Parshall flume in the Gooscquill Ditch as allowed in the Technical Specifications.

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P-89311 August 15, 1989 I

The liquid waste effluent flow recorder FR-6215 was found to be i l out of calibration on May 16, 1989 and despite recalibration, has continued to disagree with the tank level method of flow  !

measurement. Two reasons have been identified and corrective ,

action initiated, but a final calibration check ~had not been done by the' end of the reporting period, June 30, 1989, resulting in 46 days out of calibration during this reporting period. FR-6215 is not required to be operable by'FSV Technical Specifications. The' flow was verified by tank level 1 measurement.

ELC0 8.1.2.j is interpreted to require either of the Gas Waste Compressor Cooling Water Activity Monitors to be reported if inoperable over thirty days, as well as the associated recorder.

l One of the two monitors, RT/RIS-46E11, was inoperable: due to being found out of calibration, reading low for 31 days (from date of di scove ry) . Difficulty procuring a replacement photomultiplier Labe was the cause of the delay. Several other I inoperable radiatiori monitors needed the same parts at the same l time, just before p' tant startup following a lengthy outage. ELCO 8.1.2.j requires only one of the two monitors to be operable; at least one was operable at all times.

The recorder was not inoperable during the reporting period.

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The Radioactive Effluent Release Report, submitted within 60 days after January 1 of each year, will include an assessment of radiation doses to show conformance with 40CFR Part 190 for the entire year. Therefore, radiation doses are not included in this report.

There were no major changes made to the radioactive waste systems during this report.

An annual land use census is performed as part of the Fort St.

Vrain Radiological Environmental Monitoring Program. Changes made to environmental sampling locations as a result of the annual land use census are reported in the annual Radiological Environmental Monitoring Program report.

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P-89311 August 15, 1989 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT SUPPLEMENTAL INFORMATION

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Facility: Fort St. Vrain fluclear Generating Station Licensee: Public Service Company of Colorado

1. Regulatory Limits Results of radioactivity analyses of gaseous and liquid effluent are used in accordance with the methodology and parameters listed

~in the Offsite Dose Calculation Manual (SUSMAP-2) to assure that the concentrations.at the point of release are maintained within the limits set forth in the Technical Specifications. These limits will ensure that the quantity of radioactive effluent released from the plant is maintained as low as reasonably achievable and in anyz event within the limits of 10CFR20 and in

-accordanca with 10CFR50.

The air dose due to noble gases released in gaseous effluent at the unrestricted area is limited to:

a. 5 millirads gamma and 10 millirads beta during any calendar quarter, and,

b. 10 millirads gamma and 20 millirads beta during any calendar year.

The dose to a member of the public due to I-131, tritium and radioactive particulate with half-lives longer than eight days in gaseous effluent is limited to:

a. 7.5 millirems to any organ during any calendar quarter, and,

b. 15 millirems to any organ during any calendar year.

The dose rate due to radioactive gaseous effluent is limited to:

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, a. For noble gases. less than or equal to 500 millirems per year to the total body and less than or equal to 3000 millirems per year to the skin, and,

b. For I-131, tritium and radioactive particulate with half-lives greater than eight days, less than or equal to 1500 millirems per year to any organ.

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P-89311- August 15, 1989 The dose or dose commitment to a member of the public from radioactive material in liquid effluent released to unrestricted areas are limited as follows:

a. During any calendar quarter to less than or equal to 1.5 millirem to the total body and to less than or equal to 5 millirems to any organ, and,

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

2. Maximum Permissible Concentrations

-All Maximum Permissible Concentration (MPC) values used in determining allowable release rates from the gas waste holdup system and the liquid waste system are those listed in Table II, Columns 1 and 2 respectively, of Appendix B to 10CFR20. In addition, .for the MPC of dissolved noble gases in liquid effluent, the value of 2.00E-04 microcuries per ' milliliter was used.

3. Average Energy The average energy (E-BAR) of the radionuclides mixture in release of fission and activation gases is nct calculated nor used at this facility.

4. Measurements and Approximation of Total Radioactivity

a. Fission and Activation Gases Batch releases from the gas waste holdup system are performed after sampling and analyses for noble gases and tritium. These analytical results are used along with atmospheric dilution factors to determine the allowable release rate. Gas is released on a continuous basis through a gas waste header which is monitored by a noble gas monitor and a particulate / iodine monitor. In the event of high activity in the release header, control functions are initiated which divert the gas to the gas waste holdup system.

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0; P-89311 August 15, 1989 All radioactive gases are released to the reactor building exhaust ventilation system which has a flow rate of approximately 30,000 cubic feet per minute. The full flow L of this is directed through high efficiency particulate filters (HEPA) and activated charcoal beds prior to the release to the environment.

Downstream of the activated charcoal beds, the gas stream radioactivity is continuously monitored and recorded' by noble gas monitors, particulate monitors, and iodine monitors,

b. Iodine For gaseous iodine, the reactor building exhaust . ventilation is monitored and recorded on a continuous basis. The iodine cartridges used in these monitors are removed after one week of service and quantitatively analyzed on a gamma spectroscopy system. The quantity of radiciodine released during that period is calculated based on the integrated flow during the collection period.

c. Particulate As in. the case of the iodine discussed in b. above, a particulate . filter is removed and analyzed each week. Gross beta analysis as well as gamma spectral analysis is performed to identify and quantify any radionuclides. The quantities of any radionuclides on this filter with half-lives greater than eight days would similarly be correlated to total flow during the collection period.

d. Liquid Effluent All liquid effluent discharged from the site reaches the unrestricted area at the Goosequill Ditch. From that point the effluent can be diverted to the St. Vrain Creek via the St. Vrain Slough, or more commonly diverttd to the Goosequill Pond which is approximately one mile North of the plant site. Outfall from the Goosequill Pond reaches the South Platte River. Both rivers converge approximately one and one half miles from the plant site. The average stream i flow reported in section Sa. of this supplemental report is ]

a summation of both rivers and was -eceived and tabulated i frcm cata provided by the Colorado Department of Natural Resources in Greeley, Colorado.

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, P-89311 August 15, 1989

' Liquid effluent is released from the site using both a continuous and batch mode. The continuous mode (automatic discharge mode) is used on the Turbine Building Sump effluent where the only-expected radionuclides is tritium.

This discharge path utilizes a continuous sampler and an

-aliquot of this composite sampler is taken three times per week .and analyzed for gross beta, gross a',pha, tritium and gamma emitters. Total flow integrators enable us. to calculate the total activity released via this pathway based on composite sample results. Discharge from the Turbine Building Sump is made directly to the unrestricted area with no dilution.

The batch release mode is used on the reactor building sump effluent and the liquid waste processing systein. The reactor building sump can '. ol d several hundred thousand gallons of waste water from various sources which could be-contaminated. The liquid waste system consists of two 2000-gallon receivers, one 2000 gallon monitoring tank and associated fiiters and demineralizers. This system is designed to coilect and process contaminated waste water resulting from reactor operations.

Prior to each release, duplicate samples are quantitativrJ/

analyzed for their radioactive constituents. These anal..

include gross beta, gross alpha, tritium and gamma spect.al analyses. .The results of these analyses and other analyses as dictated'by the gross beta results are used to determine the maximum release rate from the site. The liquid effluent is diluted with cooling tower blowdown which flows at a minimur.. of 1100 gallons per minute. The resulting mixture is sampled during the release period to confirm compliance with regulatory limits.

1 The liquid effluent from the batch release mode is monitored continuously by redundant gamma activity monitors.

All tank level indicating devices, flow monitoring and recording devices, and radiation monitor equipment are calibrated and maintained at scheduled intervals in accordance with established procedures. l Composite samples from the batch releases, and continuous l- re' eases are analyzed monthly for Sr-89, Sr-90 and S-35.

All sample results are conservatively decay-corrected to the start of the composite period. j l

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P-89311 ' August 15, 1989

e. Overall Error The overall error associated with determining the total activity released from the site for both gaseous and liquid effluent is estimated to be 17.3 percent. This'value is the square root of the sum of squares of counting statistics, associated calibration errors, sampling errors and tank volume estimates, each considered to be plus or minus 10 percent.

5. Batch Releases

a. Liquid l Number of Batch Releases l 111 l 1 1 I l Total Time Period for l . l l Batch Releases 1 9.21E+02 HOURS l l l 1

-lMaximum Time Period for l l la Batch Release l 2.82E+01 HOURS l l l l

.l Average T'me Period for l l la Batch elease [ 8.29E+00 HOURS l l l l l Minimum Time Period for l l la Batch Release l 3.68E+00 HOURS l l l l l Average Stream Flow During l l l Period of Effluent Release l 2.75E+05 GALLONS PER l lInto a Flowing Stream l MINUTE I I I I

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P-89311 August 15, 1989

b. Gasecus l Number of Batch Releases. l 171 l l l l l Total Time Period for l l la Batch Release I. 5.75E+02 HOURS l l l 1.

l Maximum Time Period for l l la Batch Release l 1.51E+01 HOURS l 1 l l l Average Time Period for l l la Batch Rclease l 3.36E+00 HOURS l 1 l  !

l Minimum Time Period for l- l la Batch Release l 3.33E-01 HOURS l l __I I

6. Abnormal Releases

a. Liquid l Number of Releases l 0 l l 1 l l Total Activity Released l N/A l l l 1

b. Gaseous INumber of Releases l 3 l l l 1 l Total Activity Released l 2.78E-02 CURIES l l l l.

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a TABLE 1A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1989 GASEOUS EFFLUENT-SUMMATION OF ALL RELEASES l Unit l Quarter l Quarter l Est. Total l l l 1 l 2 l Error. % l l l l l I

f. Fission m d activation products

1. Totai release l Ci l 4.38E-01 l 1.04E+02 l 1.73E+00 l l- I l l l

2. Average release rate for juC1/seci 5.63E-02 l 1.32E+01 l period l l l l l l l l

3. Percent of technical l  % l 2.26E-02 l 1.16E+00 l specification limit l l l l B. Iodine

1. Total iodine-131 l Ci l <3.75E-06 l <3.79E-06 l l l l l l l

2. Average release rate for luc 1/seci <4.82E-07 l <4.82E-07 l period l l l l l l l l

3. Percent of technical l  % l l l l specification limit l l l l C. Particulate l 1. Particulate with half-lives l Ci l <1.31E-07 l <1.37E-0/ l l l > 8 days l l l l l l l l l

2. Average release rate for luCi/secl <1.69E-08 l <1.74E-08 l period l l l l l 1 l I

3. Percent of technical l  % l l  !

specification limit l l l l l l l l

4. Gross alpha radioactivity l Ci l <3.56E-08 l <3.53E-08 l l __ l - -- l l D. Tritium I

1. Total release l Ci l 6.55E-02 l 1.63E+00 l 1.73E+00 l l 1 I I I I l 2. Average release rate for juC1/ sect 8.42E-03 l 2.07E-01 l L period l l l l l l l l

3. Percent of tech. spec. l  % l 1.40E-04 l 3.46E-03 l limit l l l l I

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TABLE IC EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1989)

GASEOUS EFFLUENTS--GROUND-LEVEL RELEASE CONTINU0US MODE BATCH MODE lRaleased j Unit l Quarter 1 l- Quarter E l Qui ter 1 l Quarter 2 l 1 1 1  ! L I I Fission gases l krypton-85 ' l Ci l l l <6.76E+00 l 1.51E+0C l l l l 1 l l l l krypton-85m j C1 l l l 4.87E-02 l 8.81E-01 l 1 l l l l l l l Lrypton-87 l Ci l l l <5.52E-02 l 6.65E-02 l 1 I I I l -- - - 1 I l krypton-88 l C1 l l l 4.49E-02 l 8.28E-01 l ~

l l l l l l 1 l xenon-133 l Ci l l l 1.24E-01 l 9.42E+01-l l l__ l l l-- 1 I lxanon-135 l C1 l l l 2.20E-01 l 5.06E+0 l l l l l 1 l __  !

l xenon-135m l Ci l l l l l 1 _ l. I l _. I I -

1 l xenon-138 l C1 l l l <3.22E-01 l <3.46E-01 {

l I I I I l_ l l xenon-133m l Ci l l l <1.966-01 l 9.77E-01 l l 1._. l i i 1 I lxenon-131m 1 C1 l l l l 6.53E-01 l l l l l l l l l argon-41 l C1 l l l l l l ~l l , ._ I I I I l Total for period

• l Ci l l. l 4.38E-01 l 1.04E+02 l l l l l l l .-1 Iodines liodine-131 l Ci l l l <1.31E-09 j <1.41E-09 l l_ l l I l -- 1 I l iodine-133 l Ci l l l l l l _. l___ l- 1 I I I l iodine-135 l Ci j l l l l l 1 1 I I I I

'l Total for period l Ci l l l 0.00E-01 l 0.00E-01 l

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• Total values do not include "<" data i

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9 Page 2 TABLE 1C(Continued) .

Particulate l strontium-89 l Ci l l l l l j i I i l I- I l l strontium-90 l Ci l l l l l l l l l l l l l cesium-134 l Ci l l l l l .)

I I I I I I I l ce si um-13'i l Ci l l l l l

-l l I I I I I g l barium-lanthanum-140 l C1 l l l l l 1 I I I I I I I a

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• Total values do not include "<" data ,

-- ---- - . - - - _m____-_u-__m---_____-________m-_-______m_.._m____m.____________m__m_._______.___._______.m--_ ___.-_.________________m mm_.-- .

4 TABLE 2A' EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1989 LIQUID EFFLUENT-SUMMATION OF ALL RELEASES l

l Nuclide l Units l Quarter l Quarter l Est. Total l l l l 1 l 2 l Error. % l l l l l l l A. Fission and activation products-

1. Total. release lCs-137 l Ci l 6.05E-06 l 0.00E-01 l 1.73E+01 l l l l l l 1

2. Average diluted luci/mi l 1.12E-11 l 0.00E-01 l concentration l l l l l l l l

3. Percent of l  % l 5.60E-05 l 0.00E-01 l applicabic limit l l l l

1. Total release- lCo-60 l C1 l 0.00E-01 l 1.95E-06 l 1.73E+01 l l l l l l l

2. Average diluted luci/ml l 0.00E-01 l 3.57E-12 l concentration l l l l l l l l 3 .- Percent of l  % l 0.00E-01 l 1.19E-05 l applicable limit l l l l B. Tritium

1. Total release lH-3 l Ci l 1.69E+01 l 2.26E+01 l 1.73E+01 l l l l l l l

2. Average diluted luc 1/ml l 3.13E-05 l 4.14E-05 l concentration l l l l l l ._ l l

3. Percent of l  % l 1.04E+00 l 1.38E+00 l applicable limit l l l l C. Dissolved and entrained gases

1. Total release lXe-133 l Ci l 3.22E-05 l 3.65E-04 l 1.73E+01 -l l l 1. l l l

2. Average diluted luc 1/ml l 5.97E-11 l 6.71E-10 l concentration l l l l l l l l

3. Percent of l  % l 2.98E-0E l 3.36E-04 l applicable limit l_ l l _l D. Gross alpha radioactivity

1. Total release 1 Ci l<3.61E-05 l <4.33E-05 l 1.73E+01 l l l l 1 1 E. Volume of waste released . i 1.00E+01  ;

(prior to dilution) l Liters l 1.73E+07 l 1.88E407 l l I _1 l_. I l

n.-*-'

I I -.

1. 9'

' TABLE 2A(Continued)'

F. . Volume of dilution water used during release l Liters l 5,40E+08 l 5.46E+08 l 1.00E+01 l l' I I I I i

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

TABLE 2B EFFLUENT AND t!ASTE DISPOSAL SEMIANNUAL REPORT (1989)

L l LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE lIfuelides Released l Uni. l Quarter 1 l Quarter 2 l Quarter 1 l Quarter 2 l l l l l l ~l- 1

-l strontium-8" l C1 l <1.53E-04 l <6.64E-05 l <4.91E-05 l <7.55E-06 l l l l l l l l l strontium-90 l Ci l <3.47E-05 l <6.74E-05 l <8.78E-06 l <7.76E-06 l 1 I I I I I I l cesium-134 l Ci l <9.43E-04 l <1.08E-03 1 <2.43E-04 l <2.10E-04 l l l l l l l 1 l cesium-137 l Ci j <8.43E-04 l <9.63E-04 l <2.05E-04 l <1.88E-04 l l l l _I I I I l iodine-131 l Ci l <8.34E-04 l <9.53E-04 l <2.15E-04 l <1.86E-04 l l- 1 I I I I I l ccbalt-58 l Ci l <9.18E-04 l <1.05E-03 l <2.37E-04 l <2.05E-04 l l l l 1 _I I I l cobalt-60 I Ci l <9.86E-04 l <1.13E-03 l <2.54E-04 l <2.00E-04 l l l l 1 I I I l iron-59 l C1 l <1.78E-03 l <2.03E-03 1 <4.57E-04 l <3.95E-04 l l . I 'l .I I I I l z'nc-65 l Ci l <2.00E-03 l <2.28E-03 l <5.14E-04 l <4.45E-04 l 1 1 I I I I I l manganese-54 l Ci l <9.06E-04 l <1.04E-03 l <2.33E-04 l <2.02E-04 l l -

1~ l l l .

l l chromium-51 l C1 l l l i l I. I _-l_ _I 1. _ .__ I l zirconium-niobium-95 l Ci l l l l 1 I I I I _I I I I molybdenum-99 l Ci l <5.76E-03 l <6.57E-03 l <1.48E-r3 l <1.28E-03 l l 1 I I  : I I l technetium-99m l Ci l l l l l l 1 I l- i .__l 1 l barium-lanthanum-140 l Ci l l l l l l l l _I  ! I i 1 cerium-141 l Ci l <1.33E-03 l <1.52E-03 i <3.44E-04 l <2.97E-04 I I l l l l l l l tritium l C1 l 1.48E-02 l 6.46E-01 l 1.69E+01 l 2.20E+01 l l 1 I I i 1 l l sulfur-35 l Ci l <3.51E-03 l <3.79E-03 l <9.58E-04 l <9.14E-04 l i i i l l I i l Total for period (above)*l Ci l 1.48E-02 l 6.46E-01 l 1.69E+01 l 2.20E+01 l l l 1 I I I I

• Total values do not include "<" data

TABLE 2B (Continued) l xenon-133 l C1 l <2.63E-03 l <3.00E-03 l <6.53E-04 l <8.95E-04 l 1 I I I I I I

'l xenon-135 l C1 l <7.18E-04 l <8.20E-04 l <1.84E-04 l <1.59E-05 l.

.l__ l I I I I I l

• Total values do not include "<" data n_ :__ _ _ . -

4';

TABLE 3

, EFFLUENT AND t!ASTE DISPOSAL SEM1 ANNUAL REPORT (1989)

SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

' l. 1. Type of waste. l Unit l 6-month l Est. Total l

. 1- l l Period I. Error. % 1 1 I I .I 1 1: 1 i l l l= a '. Process waste, i.e., l m' I l l l spent resins, filter l Ci l l l

l. sludges, evaporator l l l l l bottoms, etc. l l 1 l-l- l l l l l l l l 1

-l b. Dry active waste'.i.e., I m' l 2.86E+00 l 1.47E+01 l

l. dry compressible waste, 1 Ci l 1.16E-02 l l

.l contaminated equip. etc. I l l l l l l l_ __l i I I I I l c. Irradiated components, I m' l 1.71E+00 l- 2.00E+01 l

-I control rods, etc. I Ci l 1.08E+03 l 1 l l l l l l 1 1 I I

l. d. Other(describe) l m' l l l 1 l Ci l l l l __I ._ I I . I 2 ,, Estimate of major nuclide composition (by type of waste)

Type of Waste lIsotopelContentl Curies l Error % ~l i I  % l l l

a. Process waste i NONE I %i ~

. E- I _[

b. Dry Active Waste l H3 I 51.1% l 5.94E-03 13.00E+00  !

l MN54 l 15.9% l 1.85E-03 12.20E+01 l CS137 l 13.4% l 1.56E-03 12.20E+01 IAG110M i 12.2% l 1.42E-03 11.25E+01 l CS134 l 7.31% ! 8.50E-04 17.10E+01

c. Irradiated l FESS I 40.7% l 4.39E+02 12.00E+01 Components i NI63 1 27.6% i 2.98E+02 12.98E+02 l l C060 l 23.0% i 2.48E+02 12.00E+01 l l 11 3 l 8.30% l 8.96E+01 12.00E+01 l l NI59 l 0.2% 2.41E+00 12.00E+01 l l MN54 1 0.2% 2.28E+00 12.00E+01 I

d. Other 1 NONE I  % i

. E-- 1 i

3. Solid Waste Disposition Number of Shipments Mode of Transportation Destination 3 Highway Beatty, Nevada B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shipments Mode of Transportation Destination 0

p-t@

4 , ,.

1 i

'f .

Table 4A

Hourly Meteorological Data-

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