Text

Semiannual Radioactive Effluent Release Rept,Jan-June 1988
	ML20153C259
Person / Time
Site:	Fort Saint Vrain
Issue date:	06/30/1988
From:	Robert Williams; PUBLIC SERVICE CO. OF COLORADO
To:	;
References
	P-88314, NUDOCS 8809010060
	Download: ML20153C259 (51)
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SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT January - June 1988 Public Service Company of Colorado Fort St. Vrein Nuclear Generating Station August 1988 8809010060 800630 PDR ADOCK 05000267 R PNU EEK

!l/

~P-88314 August 25, 1988

'This report summarizes the radiological effluent released from the-

~ Fort St. Vrain Nuclear Generating Station for _the period of January through June, 1988. This information is provided pursuant to the i requirementsofSections7.5.1.e,8.1.1.g.8),8.1.2.d,and8.2.1.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

-l 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 cetectable activity (MDA). This results in reporting upper limit values that are in excess of true activities, r

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l P-88314 -3 ' August 25, 1988 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 count, above the system background, that will be detected with a 95% probability of being correct and only a 5% probability of falsely concluding that a blank observation represents a real signal. The LLO values specified in our Technical Specifications are as follows:

Liquid Principle Gamma Emitters 5.00E-07 pCi/ml

! Dissolved Noble Gases 1.00E-05 pC1/ml Tritium 1.00E-05 vCi/ml Iodine-131 1.00E-06 pCi/ml Gross Alpha 1.00E-07 uti/mi '

Strontium-89,90(Composite) 5.00E-08 vCi/ml Gaseous '

I Principle Gamma Emitters 1.00E-04 pCi/cc 3 (Gas)

Principle Gamma Emitters 1.00E-11 pCi/cc '

(Particulate) i Tritium (Gas) 1.00E-06 vCi/cc i

Iodine-131(Charcoal) 1.00E-12 uCi/cc I GrossAlpha(Particulate) 1.00E-11 pCi/cc ;

Strontium-89,90(Particulate) 1.00E-11 uti/cc Gross-Beta (Particulate) 1.00E-11 vC1/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" valces were not included in the total values for the pathway.

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5 P-88314 August 25, 1988 Tne 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 IC, the continuous release mode values are not reported because this release pathway is the same-as the batch mode. All other blanks on Tables 1C and 2B are due to the fact that no LLO 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 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 iscontainedinsection4a)oftheSupplementalInformation Section. ;

On February 10, 1988, an unplanned radioactive gas release was

, detected at Reactor Building exhaust stack monitors RT-73437-1 and RT-73437-2. Restrictions in the core support floor vent line filters and strainer M-6311 caused a pressure backup in the vent line and i resulted in activity being released form the Reactor Building exhaust stack via the core support floor vent relief valve V-6389. Total activity released was calculated to be 4.26E-01 Ci of noble gas and ,

4.41E-04 Ci of particulates. A description of this event can be '

found in Public Service Company of Colorado Non-Emergency Event Notification 88-006 and follow-up notifications88-007 and 88-009, ;

submitted to the Nuclear Regulatory Commission. It was recorded in Public Service Company memo, PPS-88-1659, to the Plant Operations Review Committee. !

An off-site electrical power disturbance on April 4, 1988, caused fluctuations in tne grid frequency to plant equipment resulting in an '

unstable condition. Based on potential loss of control, the decision was made to manually scram the reactor. During the coc!1own, the !

plant experienced an unplanned radioactive gas release from the core support floor due to flow restrictions in the vent system. A total cf 1.91E+00 Ci of activity was released. Notification of this event was made to the Nuclear Reoulatory Commission as recorded in Public Service Company Non-EP' .cncy "vent Notifications88-011 and 88-012, and described in Public Servi.e Company memo, PPS-88-1654, to the Plant Operations Review Committee. This event was reported to the :

, Nuclear Regulatory Commission in Licensee Event Report 88-004 :

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P-88314 August 25, 1988 An expansion joint on a circulating water pump failed on April 7, 1988, resulting in the Icss of circulating water, the cooling tower and turbine vacuum. The reactor was manually scrammed.

Core support floor vent relief valve (V-6389) lifted due to excess pressure and discharged unpurified heiium into the reactor plant exhaust stack giving a high activity indication on Reactor Building exhaust stack monitor RT-7324-1. A total of 2.69E-01 Ci of activity was calculated to have been released. Notification of this event was made to the Nuclear Regulatory Commission as recorded in Public Service Company Non-Emergency Event Notification 88-014, and reported to the Plant Operations Review Committee in Public Service Company memo, PPS-88-1655. This event was reported to the Nuclear Regulatory Commission in Licensee Event Report 88-006.

On April 26, 1988, the Colorado Department of Health was notified of an unplanned radioactive gas release due to the loss of liquid nitrogen to the low temperature absorber in the helium purification train. A total of 4.40E-3 Ci of activity was calculated to have been released based upon noble gas monitor RT-31193 readings and the duration of the release. Notification of this avent was made to the Nuclear Regulatory Commission as recorded on Non-Emergency Event Notification 88-017.

An unplanned radioactive gas release occurred on May 6, 1988, following an automatic reactor scram caused by the malfunction of an analog averaging instrument in the overall plant control system, A reduction in helium purge flow to the circulator shafts allowed primary coolant to enter the buffer helium and bearing water systens and the Reactor Building. Rapid depressurization of the Prestressed Concrete Reactor Vessel caused overheating of the low temperature

, absorber and offgassing to the purified helium header. Based upon RT-73437-2 readings and the event duration, it was calculated that 3.52E-03 Ci of activity was released. Notification of this event was made to the Nuclear Regulatory Commission as recorded in Non-Emergency Event Notification 88-020 and reported to the Plant Operations Review Committee in Public Service Company memo PPS-88-1995. The event was reported to the Nuclear Regulatory Commission in Licensee Event Report 88-009.

There were no unplanned radioactive liquid waste releases made from the Liquid Waste System 62 or the Reactor Building Sump during this reporting period.

4 P-88314 August 25, 1988 ,

There. were repeated sporadic incidents during this reporting period of tritium activity from secondary coolant, System 31, being released into the service water cooling tower, System 42, through the pressure relief valve. The Design Change Action Request to reroute the relief discharge line to the. Turbine Building Sump that was reported in the last Semi-Annual Effluent Release Report, P-88086, is still in progress. Using System 42 analysis and estimated makeup water values, it-is calculated at 5.29E-01 Ci of tritium was released by this pathway.

There were no radioactive solid-waste shipments for disposal made during this reporting period.

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

during this reporting period. L L

There were no changes to the Process Control Program (SUSMAP-3), ,

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

Changes to the Offsite Dose Calculation Manual (ODCM),SUSMAP-2 reported in the Semi-Annual Radioactive Effluent. Release Report dated 1 February 29, 1988, were finalized in SUSMAp-2, Issue 16, effective '

date May 3, 1988. There were no new changes to the ODCM, SUSMAP-2 durir.g this reporting period.

There were no radioactive gas or radioactive liquid release monitors i out of service for a period of greater than 30 days during this reporting period.

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 40 CFR Part 190 for the entire year. Therefore, the radiation doses are not included in this !

report.

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

Radiological Environmental Monitoring Program.

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P-88314 August 25, 1988 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT =

SUPPLEMENTAL-INFORMATION=

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Facility: Fort-St. Vrain Nuclear 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 m % sed from the plant is maintained as low as reasonably' achi.tvable and in any event within the limits of 10CFR20 and in accordance with 10CFR50.

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

.. 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 particulates 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:-

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 particulates with half-lives greater than eight days, less than or equal to 1500 millirems per year to any organ.

E' P-88314 -August 25, 1988 The dose or dos commitment' t'o '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 niillirems 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. AverageJrfggy ,

The average energy (E-BAR) of the radionuclide mixture in release of fission and aettratien sases is not calculated nor used at this facility.

4. Measurements and_Apprsximation 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 an iodine monitor. In i.he event of high activity in the release header, control functions are initiated which divert the gas to the gas waste holdup system.

P-88314 August 25, 1988 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 of this is directed through high efficiency particulate filters (HEPA) and activated charcoal beds prior to the l 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 radioiodine released I during that period is calculated based on the integrated flow during the collection period,

c. Particulate l

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

l d. Liquid Effluent 1

All liquid effluent discharged from the site reaches the unrestricted area at the Goosequill Ditch. From that point i the effluent can be diverted to the St. Vrain Creek via the :

St. Vrain Slough, or more commonly diverted to the I l Goosequill Pond which is approximately one mile North of the l

plant site. Outfall from the Goosequill Fond reaches the South Platte River. Both rivers converge approximately one and one half miles from the plant site, The average stream l flow reported in section Sa. of this supplemental report is l a summation of both rivers and was received and tabulated from data provided by the Colora h Department of Natural ,

j Resources in Greeley, Cnicrado. 1 I

P-88314 August 25, 1988 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 radionuclide 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 alpha, 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 system. The Reactor Building Sump can hold 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 filters and demineralizers. This system is designed to collect and process contaminated waste water resulting from reactor operations.

Prior to each release, duplicate samples are cuantitatively analyzed for their radioactive constituents. These analyses include gross beta, gross alpha, tritium and gamma spectral analyses. The results of these analyses and other analyses

, as dictated by the gross beta results are used to determine the maximun release rate from the site. The liquid effluent is diluted with cooling tower blowdown which flows at a minimum of 1100 gallons per minute. The resulting mixture is sampled during the release period to confirm compliance 1 with regulatory limits, i The liquid effluent froa 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 I accordance with established procedures.

Composite samples from the batch releases, and continuous releases are analyzed monthly for Sr-89, Sr-90 and S-35.

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

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P-88314 August 25, 1988

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.

S. Batch Releases

a. Liquid l Number of Bate.h Releases l 73 l l l l lTetal Time Period for l l l

l Batch Releases l 6.95E+02 HOURS l I l 1 l Maximum Time Period for l l la Batch Release l 8.60E+01 HOURS l l l l l Average Time Period for l l la Batch Release l 9.52E+00 HOURS l 1 I I l Minimum Time Period for l l la Batch Release l 2.20E+00 HOURS l

' I I l l Average Stream Flow During l l l Period of Effluent Releast i 3.86E+05 GALLONS PER l lInto a Flowing Stream l MINUTE I l _l l 1

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, P-88314 August 25, 1988 L

b. Gaseous

l Number of Batch Releases l 139 l l l l l Total Time Period for l l la Batch Release 1 6.34E+02 HOURS l l l l l l Maximum Time Period for l l lt. Batch Release l 1.45E+01 HOURS l i l I l l l Average Time Period for l l l la Batch Release 1 4.56P00 HOURS l l l l I i

l Minimum Time Period for l l l la Batch Release l 1.47E+00 HOURS l I I i l

6. Abnormal Releases '

a. Liquid l Number of Releases l 1 l

> l l l

' l Total Activity Released l 5.29E-01 CURIES l l l l

b. Gaseous l Number of Releases l 4 l !

! I I l Total Activity Released i 2.61E+00 CORIES l ,

I l l !

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TABLE 1A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1988 GASEOUS EFFLUENT-SUMMATION OF ALL RELEASES l Unit l Quarter l Quarter i Est. Total l l l 1 l 2 l Error % l l l l l l A. Fission and activation products

1. Total release l Ci l 1.10E+02 l 8.84E+01 1 1.73E+01 l l 1 l I l

2. Average release rate for luc 1/secl 1.40E+01 l 1.12E+01 l period l l l l I I I I

3. Percent of technicel I % l 1.32E+00 l 9.0pr.01 l specification limit l l l l B. Iodine

1. Total iodine-131 l Ci l <5.42E-06 i <5.24E-06 l l l 1 l l I

2. Average release rate for luc 1/secl <6.90E-07 l <6.90E-07 l period l l l l l 1 I l

3. Percent of technical l % l l l specification limit l l l l C. Particulates

1. Particulates with half-lives l Ci i <2.84E-07 i <1.95E-07 l l

> 8 days l l l l l 1 l l 1

2. Average release rate for luc 1/ sect <3.61E-08 i <2.48E-08'l period l l l l l l l l

3. Percent of technical l % l l l specification limit l l l 1 I

l_ l l l Gross alpha radioactivity l Ci

4. l <5.07E-08 i <4.43E-08 l 1 I I I

0. Tritium l

l

1. Total release l Ci i 1.33E+00 1 1.93E+00 l 1.73E+01 l I I I I

2. Average release rate for luc 1/secl 1.69E-bTl l 2.45E-01 1 j period l l l l I I l 1

3. Percent of tech. spec. I % l 2.82E-03 l 4.09E-03 l limit l l l l j I

TABLE IC EFFLUENTANDVASTEDISPOSALSEMIANNUALREPORT(1988)

GASEOUS EFFLUENTS--GROUND-LEVEL RELEASE CONTINUOUS MODE BATCH MODE l Released l Unit l Quarter 1 l Quarter 2 I Quarter 11 Quarter 2 l l 1 1 I .l. I I Fission gases l

Ikrypton-85 l Ci l l l <4.63E+00 l 1.44E+00 l l l l l 1 l l t

l krypton-85m l Ci l l l 2.93E-01 l 5.54E-01 l l l l l l 1 l l Ik ypton-87 ll Ci l l I 3.23E-02 1 5.06E-02 l l l l 1 I I I ikryoton-88 l C1 I I I 2.73FDTI 5.56E-01 l l l l l l ~ l l lxenon-133 i Ci l l l 1.06E T I 8.10E+01 l l l l l l l lxenon-135 i Ci l l l 2.04E+00~ll 3.12E+00 I I l- I l i l I lxenon-135m l~ Ci I l l l 9.62E-03 I l __ l l .-- -- l l I lxenon-138 l C1 l l l <2.07E-01_l1 <2.87E-01 l l 1 - I l i lxenon-133m l Ci I I I 7.42E-01 1I 5.84E-01l I l I I I I I l I 1

lxenon-131m l C1 l 1 l 4.68E-01 1 1.05E+00 l I I I I~~ I I I l argon-41 l Ci i l l l 8.89E-03 I I I I I I ITotal for period

Il Ci_II I i 1.10E+02 l~ 8.84E+01 l 1 I I I I I I l

Iodines 1

l liodine-131 l C1 I l i <1.23E-09 l <1.71E-09 I I I I I I I l liodine-133 I Ci i l l l l l l 1 I I I I Iiodine-135 i Ci l i l l I I I I I I I I ITotal for period I Ci l 1 1 0.00E+00 1 0.00E+00 I I I I I I I I l

l l

Total values do not include "<" data l

1

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

O i

Jn

$f A Page'2-E %

TABLE 1C(Continued)

Particulates -

.4

-l strontium-89 - l Ci .l. -l. l l .I

- l- l l l l l l

l. strontium-90. .l- Ci l- l: l l l.

t

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

l cesium-134 l- C1 l .I I I . l:

I l I I I I I

l cesium-137 l Ci l . l_ l l- l.

, I I I I I I l-

".--' 'lbarium-lanthanum-140

[ Ct l l l l -l 1 l- -l i I I I l' a

3 e

I l

~

t 1

1

-- j I

I I

l i

Total values do not include "<" data '

ll

- + - - - , ,, . - ~.. , # - ,, ,,.m.,,v,,. -,- ,. m-. . - - - _ , , - - . - -,.,.x . , ,,,,--._ -- _,,,c --..---,-,-,.,,r.,v.,_, ,

TABLE 2A

~ EFFLUENT AND WASTE DISPOSAL SEMIANNVAL REPORT 1988 LIQUID EFFLUENT-SUMMATION OF ALL RELEASES l Nuclide l Units l Quarter i Quarter l Est. Total-l l l l 1 l 2 I Error. %- l

1. I I I _l - 'l

~A. Fission and activation products-t

1. .Tetal release lCs-137 l Ci l 0.00E+00 1 2.36E-06 l 1.73E+01 l l l l l l l

2. Average diluted juC1/ml l 0.00E+00 l 4.33E-12 l concentration l l l l l l l 1

3. Percent of .

l % l l 2.16E-05 l applicable limit l l l l

1. Total release lCo-60 l Ci l 0.00E+00 l 1.15E-06 l 1.73E+01 l l l l l l l - ,

2. Average diluted luci/ml l 0.00E+00 l 2.11E-12.l concentration l l l l l 1 1 I

3. Percent of l % l l 7.03E-06 l applicable limit l l l l B. Tritium i

1. Total release lH-3 l C1 l A.34E+01 1 5.59E+01 1 1.73E+01 l l l 1 I I i

2. Average diluted luci/mi l 7.95E-05 l 1.02E-04.I, concentration l l l l l l I 1

3. Percent of l % { 2.65E+00 l 3.40E+00 I applicable limit l l l _, l C. Dissolved and entrained gases

1. Total release lXe-133 I ci l 9.05E+03 l 1.63E-03 l 1.73E+01 l l 1 1 I I I

2. Average diluted luci/mi l 1.66E-08 l 2.99E-09 l concentration l l l l l l l l

3. Percent of I % l 8.30E-03 l 1.50E-03 l applicable limit l l l l

1. Total release IXe-133m l Ci l 6.92E-00 1 0.00E+00 l 1.73E+01 l I I I I I I

2. Average diluted luCi/mi l 1.27E-10 1 0.00E+00 l concentracion l l l l l l 1 1

3. Percent of I % l 6.35E-05 l l applicable limit l l l l

, - . , . . . , . - -r - - - - - , - - , ~- -n-

TABLE 2A(Continue'd)

C. Dissolved'and entrained

, gases'(Continued)

1. ' Total release lXe-135 l C1 -l 1.93E-05 l 2.37E-06 l 1.73E+01 l-I l l l l l

2. Average diluted luCi/mi l 3.54E-11 l 4.34E-12 l concentration l l l l l l l l

3. Percent of l % i 1.77E-05 1 2.17E-06 l -,.

applicable limit l _,1 l l

1. Total release ;Xe-131m l Ci l 2.42E-05 l 4.81E-06 l -1.73E+01 l l l l l. I l

2. Average diluted luci/mi l 4.44E-11 l 8.82E-12 l concentration l l l l l l l l

3. Percent of I % l 2.22E-05 l 4.41E-06 l applicable limit l l l l D. G,oss alpha radioactivity

1. Total release l Ci l<9.73E-05 l 5.02E-04 l 1.73E+01 l l- 1 I l l E. Volume of waste released (priortodilution) l Liters l 9.45E+07 l 8.32E+07 l 1.00E+01 l l l l l l F. Volume of dilution water used during release l Liters l S.46E+08 l 5.46E+08 l 1.00E+01 l l l l l l .

)

> l 1

l 1

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TABLE 2B EFFLUENT ANP WASTE DTSPOSAL SEMfANNUAL REPORT (1988)

, LIQUID EFFLUENTS.

CONTINU0US MODE -BATCH MODE lNuclides Released- l Unit l Quarter -1 l Quarter 2 I Quarter _-1 l Quarter 2 l l l 1 1 I I l-l strontium-89 I C1 .I <2.58E-04 l <2.14E-04 l <4.64E-06 l <5.16E-06 l l l l l' I l l l strontium-90 l Ci l <2.65E-04 l <2.21E-04.l <4.73E-06 l <5.35E-06 l l 1 I I I I I l cesium-134 i C1 l <6.13E-03 l <S.32E-03 l <8.40E-05-l_-<1.56E-04 I I I I I I I I I cesium-137 I Ci l <6.82E-03 l <5.93E-03 l <9.36E-05 I <1.51E-04 I I I I I l l I I iodine-131 l C1 1 <5.51E-03 l <4.78E-03 l <7.55E-05 l <1.40E-04_l '

l l l l l 1 -l I= cobalt-58 i C1 l <5.88E-03 l <5.09E-03 l <8.05E-05 l <1.49E-04 i i i l-- 1 I I I i cobalt-60 i C1 1 <5.64E-03 l <4.89E-03 l <7.74E-05 l <1.24E-04 l l 1 1 I I I I l iron-59 l C1 l <1.22E-02 l <1.06E-02 1.<1.67E-04 l <3.10E-04 l '

l I I I I l l l zinc-65 l C1 l <1.47E-02 I <1.28E-02 l <2.02E-04 l <3.75E-04 l 1 1 1 1 l l manganese-54 i C1 l <6.70E-03 l <5.82E-03 l <9.20E-05 l il <1.70E-0 ,

I _l I i ! -l l l cnromium-51 l Ci l l l l l l l l l l l. I I zirconium-niobium-95 l C1 l l l l l l l l l l l l l molybdenum-99 l C1 l <4.53E-02 l <3.93E-02 1 <6.22E-04 l <1.15E-03 I I l l 1 I i I l technetium-99m l Ci l l l l l l l l l. I I l ,

I barium-lanthanum-140 l Ci l l l l l l 1 I I I I l 1 l l cerium-141 l Ci l <9.78E-03 l <8.48E-03 l <1.34E-04 l <2.49E-04 l l 1 I I I I I l-I tritium l Ci l 6.75E+00 l 6.40E+00 l 3.67E+01 l 4.95E+01 1 1 I I I I I I l sulfur-35 l Ci l <1.91E-02 1 <1.79'-02 1 <2.76E 04 l <5.70E-04 I I I I I i i i I

< l TotalforperiodTabove)*l C1 l 6.75E+00 l 6.40E+00 l 3.67E+01 l 4.95E+01 I i i l l I I I o

Total values do not include "<" data 4 l

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

.,. r- , .

c ,

-f

-_Page 2 TABLE 28-(Continued)

/

.l. xenon-133 l .-Ci- l <1.98E-02 l <1.72E-02_l <2.43E-04 1.<2.06E-03.l- .

l %

l .l. l _l .l l.

.l- - xenon-135 l_ C1 'l <5.24E-03 l1<4.55E-03 l <8.34E-05 l <1.16E-04.['.

l .

l- 1 I _ l i I i

c I

b i

Total values do not include "<" data

!~

l

- . . _ - _ , _ . -._.-_,.__....___-,_._-...,_.,,...__..,,,_,..,.__...,_._.,_....a

. TABLE 3 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1988)

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_

l l l Period l Error. % l l l l _l I

'l l 1 .I I l .a. Spent resins, filter I m' l 0.00E+00 l l l sludges, evaporator l Ci l 0.00E+00 l- 0 l-l bottoms, etc. l l l l l l l l l l l l 1 1 I b. Dry compressible waste, l m' l 0 00E+00 l l l contaminated equip. etc. l Ci- l 0.00E+00 l 0 l l_ l I l l I I I I l l c. Irradiated components, I m' l 0.00E+00 l l l control rods, etc. l Ci l 0.00E+00 l 0 l l I I I I I I I I I 1 d. Other(describe) lm 8 l 0.00E+00 l l l l Ci l 0.00E+00 1 0 l l l l l l

2. Estimate of major nuclide composition (by type of waste)

a. PROCESS WASTE: NONE I % l l 1 % 1 I l _% I I

b. ORY ACTIVE WASTE. I % i l I % l l l % l l i

c. IRRADIATED COMP.: ,

I % l l l l % l l ;

I % l I i

d. OTHER: I % i l l % 1~ l I % l. l

3. Solid Waste Disposition Number of Shipments Mode of Transportation Destination NONE i B. IRRADIATED FUEL SHIPMENTS (Disposition) i Number of Shipments Mode of Transportation Destination NONE

y Table 4A Hourly Meteorological Data

Table 4A Hourly Heteorologicci Data

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h Public Service ~ P.O. Box B40 Denver, CO 80201 0840 August 25, 1988 R.O. WILUAMW JR.
Fort St. Vrain VICE PRESIDtNT Unit No. 1 NUCLEAR OPERATIONS P-88314 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Docket No. 50-267
~~~~SUBJECT:~~~~
Semi-Ar.nual Radioactive Effluent Release Report Gentlemen:
Please find the attached Semi-Annual Radioactive Effluent Release Report for the Fort St. Vrain Nuclear Generating Station.
This report covers the period January 1,1988 through June 30, 1988, 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 hats any ques +. ions regarding this report.
Very truly yours, w
f &
R. O. Williar: , Jr.
Vice President, Nuclear Operations R0W:VHF/skd Attachment pN I \
. , , . , . . = . ,
)
E I
P-88314 August 25, 1988 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, TX- 76011 3
Mr. Robert Farrell 1 Senior Resident Inspector Fort St. Vrain Station Building 104 Dottie Sherman, ANI Library American Nuclear Insurers The Exchange, Suite 245, '
270 Farmington Avenue Farmington, CT 06032 Mr. Albert J. Hazle, Director !
Radiation Control Division i Colorado Department of Health 0 er, 8 b 4
Mr. Mike Dolphin, Site Manager -
General Atomic Company <
Fort St. Vrain <
NED-Site ;
Mr. Milt Lammering, Chief Regional Representative, Radiation Program V. S. Environmental Protection Agency ,
Region VIII '
999 18th Street, Suite 500 Denver, CO 80202-2405 Mr. Farrel D. Hobbs i Environmental Management Rockwell International Building T"452-B P. O. Box 464 Golden, CO 80401
P-88314 August 25, 1988 Dr. James E. Johnson Radiology & Radiation Biology Dept.
135 BRB Colorado State University Fort Collins, CO 80523 1
)
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ML20153C259

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