ML17309A456

Semiannual Radioactive Effluent Release Rept for Jul-Dec 1990

ML17309A456
Person / Time
Site:	Ginna
Issue date:	12/31/1990
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9103070258
Download: ML17309A456 (66)

Text

ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEM I

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

FACIL:50-244 Robert Emmet Gihna Nuclear Plant, Unit 1, Rochester G

AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C.

Rochester Gas 6 Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION DOCKET' 05000244

SUBJECT:

"Semi a

Ra 'ctive Effluent Release Rept for Jul-Dec 199

'i W/910301 r.

DISTRIBUTION CODE:

IE48D COPIES RECEIVED:LTR ENCL SIZE:

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TITLE: 50.36a(a) (2) Semiannual Effluent Release Reports NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).

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RECIPIENT ID CODE/NAME PDl-3 LA JOHNSON,A INTERNAL: AEOD/DSP/TPAB NRR/DREP/PRPB11 RES BROOKS i B RGN1 FILE 02 EXTERNAL: BNL TICHLER,J03 NRC PDR COPIES LTTR ENCL 3

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TOTAL NUMBER OF COPIES REQUIRED:

LTTR 18 ENCL 18

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~ kl~ ON ~ IA ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. 14649-0001 VEaERRORE AREA CODE 7lO 546.2700 March 1, 1991 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

Semiannual Radioactive Effluent Release Report R.E.

Ginna Nuclear Power Plant Docket No. 50-244

Dear Sirs:

This Semiannual Radioactive Effluent Release Report is being submitted in accordance with the requirements of Technical Specification Section 6.9.1.4.

Very truly yours, Robert C. Mecredy Vice President 9i03070258 901231 PDR ADOCK 05000244 R

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Copies to:

Mr. T. Martin Regional Administrator U.S. Nuclear Regulatory Commission Region 1

475 Allendale Road King of Prussia, PA 19406 Resident NRC Inspector Ginna Station Ms.

Donna Ross New York State Energy Office Empire State Plaza

Albany, NY 12223 American Nuclear Insurers MAELU The Exchange Suite 245 270 Farmington Avenue Farmington, CT 06032 ATTN:

Winthrop Hayes Central Records, Ginna Station Category 2.22.2

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SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT R. E.

GINNA NUCLEAR PLANT ROCHESTER GAS AND ELECTRIC DOCKET NO. 50-244 JULY DECEMBER 1 1 9 9 0 9103070258

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TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 SUPPLEMENTAL INFORMATION 2 '

REGULATORY LIMITS 2.2 MAXIMUMPERMISSIBLE CONCENTRATIONS 2.3 RELEASE RATE LIMITS 2.4 MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY 2.5 BATCH RELEASES 2.6 ABNORMAL RELEASES 3.0

SUMMARY

OF GASEOUS RADIOACTIVE EFFLUENTS 4.0

SUMMARY

OF LIQUID RADIOACTIVE EFFLUENTS 5.0 SOLID WASTE

)

6.0 LOWER LIMIT OF DETECTION 7.0 RADIOLOGICAL IMPACT 8.0 METEOROLOGICAL DATA 9.0 LAND USE CENSUS CHANGES 10.0 ANNUAL TABULATION OF PERSONNEL EXPOSURE 11.0 LEAK TEST OF SEALED SOURCES 12.0 CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL 13.0 CHANGES TO THE PROCESS CONTROL PROGRAM 14.0 MAJOR CHANGES TO RADWASTE TREATMENT SYSTEMS

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LIST OF TABLES Table Table Table Table Table Table Table 1A Gaseous Effluents Summation of all Releases 1B Gaseous Effluents

.Continuous and Batch Releases 2A Liquid Effluents

Summation of all Releases 2B Liquid Effluents Continuous and Batch Releases f

3 Solid Waste and Irradiated Fuel Shipments 4

Release Permits Not Meeting LLD Requirements 5A Radiation Dose to Nearest Individual Receptor from Gaseous Releases Table Table 5B Radiation Dose to Nearest Individual Receptor from Liquid Releases 6A Number of Personnel and Man-Rem by Work and Job Function Table 6B Standard Report of Personnel Whole Body Exposure

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1 ~ 0 INTRODUCTION This Semiannual Radioactive Effluent Release Report is for Rochester Gas and Electric Corporation R.E.

Ginna plant and is submitted in accordance with the require-ments of Technical Specification Section" 6.9.1.4.

The report covers the period from July',

1990 through December 31, 1990.

This report includes a

summary of the quantities of radioactive gaseous and liquid effluents and solid waste released from 'the plant presented in the format outlined in appendix B

of Regulatory

,Guide 1.21, Revision 1, June 1974.

All gaseous and liquid effluents discharged during this reporting period were in'ompliance with the limits of the R.E.

Ginna Technical Specifications.

2.0 2.1 SUPPLEMENTAL INFORMATION Re ulator Limits The Technical Specification limits applicable to release of radioactive material in liquid and gaseous effluents are:

2.1.1 Fission and Activation Gases The instantaneous dose rate, as calculated in the

ODCM, due to noble gases released in gaseous effluents from the site shall be limited to a release rate which would yield

< 500 mrem/yr to the total body and < 3000 mrem/yr to the skin if allowed to continue for a full year.

The air dose, as calculated in the ODCM, due to noble gases released in gaseous effluents from the site shall be limited to the following:

(i)

During any calendar quarter to < 10 mrad for gamma radiation and to < 20 mrad for beta radiation.

2 '

Radioiodine Tritium and Particulates The instantaneous dose rate, as calculated in the

ODCM, due to

'radioactive materials released in gaseous effluents from the site as radioiodines, radioactive materials in particulate

form, and radionuclides other than noble gases with half-lives greater than 8

days shall be limited to a release rate which would yield <

1500 mrem/yr to any organ if allowed to continue for a full year.

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2. 1.3 The dose to an individual, as calculated in the

ODCM, from radioiodine, radioactive materials in particulate form and radionuclides other than noble gases with'alf-lives greater than eight days released with gaseous effluents from the site shall be limited to the following:

(i)

During any calendar quarter to

< 7.5 mrem to any organ.

(ii) During any calendar year to < 15 mrem to any organ.

Li id Effluents The release of radioactive liquid effluents shall be such that the concentration in the circulating water discharge does not exceed the limits specified in accordance with Appendix B,

Table II, Column 2

and notes thereto of 10CFR20.

For dissolved or entrained noble gases the total activity due to dissolved or entrained noble gases shall not exceed 2 E-4 uCi/ml.

The dose or dose commitment to an individual as calcu-lated in the ODCM from radioactive materials in liquid effluents released to unrestricted areas shall be limited:

2.2 (i)

During any calendar quarter to

< 1.5 mrem to the total body and to < 5 mrem to any organ, and (ii) During any calendar year to <

3 mrem to the total body and to < 10 mrem to any organ.

Maximum Permissible Concentrations MPC 2.2.1 2.2.2 For gaseous effluents,, 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.

For liquid effluents, the maximum permissible concen-tration values specified in 10CFR20, Appendix B, Table II, column 2 are used to calculate release rates and permissible concentrations at the unrestricted area boundary.

A value of 2E-04 uCi/ml is used as the MPC for dissolved and entrained noble gases in liquid effluents.

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Release Rate Limits 2.4 The release rate limits for fission and activation gases from the R.E.

Ginna plant are not based on the average energy of the radionuclide mixture in gaseous effluents; therefore, this value is not applicable.

However, the average energy of the radionuclide mixture was 0.207 Mev.

Measurements and A

roximations of Total Radioactivit Gamma spectroscopy was the primary analysis method used to determine the radionuclide composition and concen-tration of gaseous and liquid effluents.

Composite samples were analyzed for Sr-89, Sr-90 and Fe-55 by a contract laboratory.

Tritium and alpha analysis were done using liquid scintillation and gas flow proportional counting respectively.

The total radioactivity in effluent releases was deter-mined from the measured concentration of each radio-nuclide present and the total volume of effluents released.

2 '

Batch Releases 2 ~ 5 ~ 1

~Li uid 1.

2 ~

3 ~

4 ~

5 ~

6.

Number of batch release:

Total time period for batch releases:

Maximum time period for a batch release:

Average time period for batch releases:

Minimum time period for a batch release:

Average stream flow (LPM) during periods of release effluent into a flowing stream:

2'5 E+02 5.01 E+04 min 1.49 E+04 min 2.05 E+02 min 1.5 E+01 min 1.29 E+06 LPM 2.5.2 Gaseous 1 ~

2 ~

3 ~

4 ~

5 ~

Number of batch releases:

Total time period for batch releases:

Maximum time period for a batch release:

Average time period for batch releases:

Minimum time period for a batch release:

1.4E+01 1.03E+04 min 2.88E+03 min 7.35E+02min 2.48E+02 min

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Abnormal Releases There were no abnormal releases of liquid or gaseous effluents during the reporting period.

3.0

SUMMARY

OF GASEOUS RADIOACTIVE EFFLUENTS The quantities of radioactive material released in gaseous effluents are summarized in tables lA and 1B.

All releases were considered to be elevated releases.

4.0

SUMMARY

OF LIQUID RADIOACTIVE EFFLUENTS The quantities of radioactive material released in liquid effluents are summarized in tables 2A and 2B.

5.0 SOLID WASTES The quantities of radioactive material released in shipments of solid waste transported from the site during the reporting period are summarized in table 3.

Principal nuclides were determined by gamma spectroscopy and non-gamma emitters were calculated from scaling factors determined by an independent laboratory from representative samples of that wa'ste type.

6.0 LOWER LIMIT OF DETECTION NOT MET One or more gamma emitting radionuclides did not meet the required lower limit of detection for 4 liquid releases.

These are listed by release number in table 4 ~

7.0 RADIOLOGICAL IMPACT An assessment of doses to the maximally exposed indivi-dual from gaseous and liquid effluents was performed for locations representing the maximum dose.

In all cases, doses were well below Technical Specification limits.

Doses were assessed based upon actual meteorological conditions considering the noble gas

exposure, inhala-

tion, ground plane and ingestion pathways.

The ingestion pathways considered were the produce, vege-

table, goat '

milk, cow' milk and meat pathway.

The results of this assessment are presented in Tables 5A and 5B.

8 '

METEOROLOGICAL DATA The annual summary of hourly meteorological data collected during 1990 is not included with this report, but can be made available at the R.E.

Ginna Plant as allowed by Technical Specifications.

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9.0 LAND, USE CHANGES There were no changes in critical receptor location for dose calculations during the reporting period.

10.0 ANNUAL TABULATION OF PERSONNEL EXPOSURE The annual tabulation of the number of station, utility and other personnel receiving exposures greater than 100 mrem/yr and their associated man-rem exposure according to work and job function required by Technical Specification 6.9.2.2 and 10CFR20.407 is included as Tables 6A and 6B.

11.0 12

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14.0 LEAK TEST OF SEALED SOURCES No sealed sources were found to be leaking when smeared by both wet and dry smears.

CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM)

There were no changes to the ODCM during the report period.

CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)

The PCP was updated with corrected drawings showing the addition of the demineralization system added, during the last reporting period.

A copy of the updated PCP is attached.

MAJOR CHANGES TO RADWASTE TREATMENT SYSTEMS There were no major changes to the Radwaste Treatment Systems during the reporting period.

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Table 1A EFFLUENT AND WASTE DISPOSAL SEMIANIGJAL REPORT GASEOUS EFFIIEZTS SVHMATION OF ALL RELEASES July December, 1990 Unit [Quarter )Quarter )Est. Total(

Error ~

A.

Fission 6 activation gases 1.

Total. release 2.

Avera e release rate for eriod Ci 4.30E+Ol 3 ~ 46E+Ol 7.0 E+00 uCi sec 5.41E+00 4.35E+00 3.

Percent of technical s ecification limit 0

8.59E-04 6.90E-04 B.

Iodines 1.

Total iodine-131 2.

Avera e release rate for eriod Ci 7.44E-05 7.24E-05 2.6 E+01 uCi sec 9.36E-05 9.11E-05 3.

Percent of technical s ecification limit 4

2.06E-Ol 2.00E-Ol C.

Particulates 1.

Particulates with half-lives > 8 da s Ci 8.35E-06 1.21E-05 3.0 E+01 2.

Avera e release rate for eriod uCi sec 1.05E-06 1.52E-06 Percent of technical s ecification limit 4

7.90E-05 1.14E-04 Gross al ha radioactivit Ci

7. 04E-08

1. 19E-07 D.

Tritium 1.

Total release 2.

Avera e release rate for eriod 3.

Percent of technical s ecification limit E.

Carbon-14 1.

Total release 2.

Avera e release rate for eriod 3.

Percent of technical s ecification limit Ci 3.07E+01 4.41E+01 3.2 E+00 uCi sec 3.86E+00 5.55E+00 4.54E-04 6.53E-04 Ci 7.38E-01 1.01E+00 3.0 E+01 uCi sec 9.28E-02 1.27E-Ol 4.94E-06 6.75E-06

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Table lB EFFLUENT AND WASTE DISPOSAL SEMIANNlJAL REPORT GASEOUS EFFLUENTS - ELEVATED RELEASE July December, 1990 CONTINUOUS MODE BATCH MODE Nuclides Released 1.

Fission gases k

ton-85 Unit uarter arter uarter arter 3

4 Ci 1.06E+00 1.77E-Ol 1.30E+00 1.59E+00 k

ton-85m k

ton-87 k

ton-88 Ci 2.05E-02 2.41E-02 Ci 1.95E-02 2.31E-02 Ci 2.88E-02 3.54E-02 1.21E-03 xenon-133 Ci 3.61E+Ol 2.31E+01 7.86E-03 4.52E+00 xenon-135 xenon-135m xenon-138 Others s ecif ar on-41 xenon-131m xenon-133m Ci 3.36E+00 4.32E+00 Ci 1.82E-Ol 1.53E-01 Ci 6.84E-02 6.09E-02 Ci Ci 2.53E-01 2.06E-01 Ci 4.08E-01 1.64E-01 Ci 2.01E-01 4.05E-02 3.28E-02 1.63E-01 1.10E-02 Total for eriod 2.

Iodines iodine-131 Ci 4.17E+Ol 2.83E+01 1.31E+00 6.32E+00 Ci 4.38E-05 5.37E-05 4.58E-08 3.96E-07 iodine-133 iodine-135 Ci 3.06E-05 1.83E-05 Ci 4.58E-08 Total for eriod 3.

Particulates strontium-89 strontium-90 cesium-134 cesium-137 barium-lanthanum-140 Others s ecif unidentified Ci 7.44E-05 7.20E-05 4.58E-08 4.42E-07 Ci.

Ci Ci ci Ci 8.35E-06 1.21E-05 Note:

Isotopes for which no value is given were not identified in applicable releases.

• Sample sent out for analysis but results not yet received.

Data for identified isotopes will be included with next semi-annual report for January June, 1991.

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Table 2A EFFIiUENZ AND KQi~ DISPOSAL SEMIANKJAL REPORP ZZQUID EFEIIUWlS SUMMATION OF ALL R1K2WSES July - December, 1990 I Unit IQuarter IQuarter IEst.TotalI I2.

Average diluted concentration dur iod'i 3.

Percent of a licable limit B.

Tritium l.

Total release A.

Fission and activation products Total release (not including tritium, ases al ha Ci 2.39E-02 4.66E-02 7.0 E+00 I

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uCi ml 1.40E-10 2.73E-10 2.22E-02 3.48E-02 Ci 4.27E+Ol 5.13E+01 3.2 E+00 I2 ~

Average diluted concentration dur iod I

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I uCi ml 2.50E-07 3.00E-07 3.

Percent of a licable limit 8.32E-03 1.00E-02

'issolved and entrained gases Total release Average diluted concentration dur iod Ci 3.42E-03 2.61E-02 3.2 E+Ol I

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I uCi ml 2.00E-ll 1.53E-10 3.

Percent of a licable limit D.

Gross alpha radioactivity 1.

Total release 1.00E-05 7.63E-05 Ci 5.5E-06 6.0E+01 E. Volume of waste released rior to dilution liters 2.89E+07 3.60E+07 5.0 E+00 F. Volume of dilution water used dur iod liters 1.71E+11 1.71E+ll 5.0 E+00

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i EFFLUENT AND Table 2B WASTE DISPOSAL SEMIANNUAL REPORT LIQUID EFFLUENTS July December, 1990 CONTINUOUS MODE BATCH MODE Nuclides Released Unit uart'er uarter uarter uarter strontium-89 strontium-90 cesium-134 cesium-136 cesium-137 iodine-131 iodine-133 iodine-135 cobalt-58 cobalt-60 iron-59 zinc 65 man anese-54 chromium-51

~

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tellurium-131 m

zirconium-niobium-95 mol bdenum-99 technetium-99m barium-lanthanum-140 cerium-141 Other s ecif iron-55 ruthenium-106 Cici Ci Ci Ci Ci Ci Cici Ci Ci Ci Ci Ci Ci Ci Ci Cici 2.90E-05 9.39E-06 3.51E-05 2.20E-06 6.64E-04 6.02E-03 2.82E-04 1.84E-03 1.38E-04 3.24E-04 1.74E-03 7.94E-03 5.72E-04 1.06E-03 8.74E-03 1.53E-02 1.29E-04 8.33E-05 5.88E-03 3.55E-03 1.03E-05 1.41E-05 2.29E-03 5.37E-04 2.55E-04 3.57E-05 9.59E-04 8.37E-05 2.19E-05 2.26E-04 1.91E-05 8.73E-03 1.91E-04 3.94E-04 3.12E-05 5.73E-06 6.15E-04

2. 13E-04 silver-110 m

antimon -122 Ci 7.06E-04 2.25E-04 antimon -124 antimon -125 unidentified Total for eriod above xenon-133 xenon-135 Ci Ci Ci Ci ci Ci

'.04E-04 3.61E-04 8.02E-05 8.84E-04 1.48E-03 2.30E-02 4.51E-02 2.33E-03 2.54E-02 1.09E-03 7.07E-04 NOTE:

Isotopes for which no applicable releases.

value is given were not identified in

• Sample sent out for analysis but results not yet received.

Data for

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identified isotopes will be included with next Semi-Annual Report for January June, 1991.

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Table 3

EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A.

SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel) 1.

T e of waste

)a.

Spent resins, filter sludges, eva orator bottoms etc.

)b. Dry compressible

waste, con-taminated e i etc.

(c. Irradiated components, control rods etc.

) d. Other (describe)

Unit mCi mCi mCi mCi 6-month

( Est. Total Period Error 4 7 '7E+Oll 2E+00 2.05E+02 5E+00 5.38E+01) 2E+00 3.28E-OO 5E+00

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

Fc-55 Co-60 Ni-63 Co-58 Cs-134 Sb-124 1.7E+01 1.2E+01

l. OE+01 8.8E+00 7.1E+00 4.5E+00

b. Fc-55 Cs-137 Co-58 Co-60 Ni-63 Sb-125 2.5E+01 2.3E+Ol 1.9E+Ol 1.5E+Ol 7.9E+00 5.3E+00

3. Solid Waste Disposition Number of Shi ments Mode of Trans ortation Destination 19 Highway Vehicle

Barnwell, SC B.

IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shi ments Mode of Trans ortation Destination None

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Table 4

RELEASE PERMITS NOT MEETING LLD REQUIREMENTS No.

Date Isoto es Cause 298 (7/14/90

[Fe-59,Zn-65,Cs-134,Cs-137,Ce-141 I

I 305

[7/18/90

)Fe-59, Zn-65 I

I 520

[ 12/ll/90 ( Mn-54, Co-58, Fe-59, Zn-65, Ce-141 I

I 525

[12/15/90(Ce-141 a ~

a ~

a ~

a ~

a.

Activity from other isotopes caused an increased background resulting in the LLD calculation exceeding 5E-07 uCi/ml for the listed isotopes.

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5A RADIATION DOSES TO NEAREST INDIVIDUALRECEPTOR FROM GASEOUS RELEASES IN REM 1990 QUARTER 3 Adult Teen Child Infant

/Total

/

/

/Total I

I (Total

/

/

/Total Direction Bod Skin Th roid Bod Skin Th roid Bod Skin Th roid Bod Skin Th roid N

4-3E-8 3.2E-8 4.4E-8 4.5E-8 3.2E-8 4.6E-8 4.4E-8 3.2E-8 4.4E-8 3.1E-8 3-2E-8 3-1E-8 NNE 1-2E-7 1-2E-7 1.2E-7 1.2E-7 1.2E-7 1.2E-7 1.1E-7 1.2E-7 1.2E-7 8.4E-8 1.2E-7 8-6E-8 NE 9-6E-8 7-6E-8 9.7E-8 9.7E-8 7.6E-8 9.8E-8 9.0E-8 7.6E-8 9.1E-8 6.4E-8 7.6E-8 6.5E-8 ENE 7-3E-8 5-1E-8 7.3E-8 7.4E-8 5.1E-8 7.5E-8 6.9E-8 5.1E-8 7.0E-8 4.8E-8 5.1E-8 4.9E-8 E

8.1E-7 5.0E-7 9.1E-7 1.0E-6 6.5E-7 1.1E-6 1.8E-6 1.2E-6 1.9E-6 1.7E-6 1.5E-6 1.9E-6 ESE 1.5E-6 9.7E-7 1.6E-6 2.0E-6 1.4E-6 2.1E-6 3.8E-6 3.0E-6 4.1E-6 8.2E-7 6.1E-7 1-1E-6 SE 1.4E-6 1.0E-6 1.5E-6 1.8E-6 1.4E-6 2.0E-6 3.4E-6 2.8E-6 3.6E-6 1.5E-6 1.4E-6 1.8E-6 SSE 2-6E-7 1.5E-7 3.0E-7 2.8E-7 1-5E-7 3.3E-7 3.4E-7 1.5E-7 4.2E-7 2.0E-7 1.5E-7 3.1E-7 S

3.2E-7 9.6E-8 3.7E-7 3.5E-7 9.6E-8 4.0E-7 4.5E-7 9.6E-8 5.4E-7 2.4E-7 9.6E-8 3.4E-7 SSW 9.0E-7 4.2E-7 1.0E-6 9.7E-7 4.2E-7 1.1E-6 1.2E-6 4.3E-7 1.4E-6 5.0E-7 4.2E-7 6.8E-7 SW 6.4E-7 2.5E-7 7.2E-7 7.1E-7 2.6E-7 8.0E-7 9.3E-7 2.6E-7 1.1E-6 6.3E-7 2.6E-7 8.5E-7 WSW 1.6E-6 1.1E-6 1.7E-6 2.2E-6 1.6E-6 2.3E-6 4.4E-6 3.6E-6 4.6E-6 1.3E-6 9.4E-7 1.7E-6 W

8.2E-7 3.5E-7 8.5E-7 9.9E-7 4.5E-7 1.0E-6 1.6E-6 7.9E-7 1.7E-6 7.5E-7 3.2E-7 8.8E-7 WNW 3 9E-8 6.4E-8 4.0E-8 3.9E-8 6.4E-8 4.0E-8 4.1E-8 6 4E-8 4.3E-8 3 4E-8 6 4E-8 3-4E-8 NNW 4.5E-8 3.7E-8 4.5E-8 4.5E-8 3.7E-8 4.5E-8 4.1E-8 3.7E-8 4.2E-8 2.9E-8 3.7E-8 3.0E-8

5A RADIATION DOSES TO NEAREST INDIVIDUALRECEPTOR FROM GASEOUS RELEASES IN REM 1990 QUARTER 4 Adult Child Infant

/Total f

/

/Total

/

f

/Total

/

f (Total Direction Bod Skin Th oid Bod Skin Th roid Bod Skin Th roid Bod Skin Th roid N

4.1E-8 4.4E-8 4.1E-8 4.3E-8 4.4E-8 4.4E-8 4.4E-8 4.4E-8 4.5E-8 3.4E-8 4.4E-8 3.5E-8 NNE 3.4E-8 3.6E-8 3.5E-8 3.5E-8 3.6E-8 3.6E-8 3.4E-8 3.6E-8 3.5E-8 2.6E-8 3.6E-8 2.6E-8 NE 5.2E-8 5.6E-8 5.3E-8 5.4E-8 5.6E-8 5.5E-8 5.3E-8 5.6E-8 5.5E-8 4.1E-8 5.6E-8 4.2E-8 ENE 3-5E-8 3.5E-8 3.6E-8 3.5E-8 3.5E-8 3.6E-8 3.3E-8 3.5E-8 3.4E-8 2.4E-8 3.5E-8 2-5E-8 E

2-3E-7 1.9E-7 2 5E-7 2.4E-7 1.9E-7 2.6E-7 2.8E-7 1 9E-7 3 2E-7 1.9E-7 1.9E-7 2 3E-7 ESE 1.6E-6 1.5E-6 1.7E-6 2.3E-6 2.2E-6 2.4E-6 5.0E-6 4.9E-6 5.2E-6 2.7E-6 2.7E-6 2-9E-6 SE 4.0E-7 4.2E-7 4.3E-7 5.5E-7 5.5E-7 5.8E-7 1.1E-6 1.0E-6 1.1E-6 2.5E-7 2.9E-7 3-4E-7 SSE 2.0E-7 2.7E-7 2.1E-7 2.2E-7 2.7E-7 2.3E-7 2.5E-7 2.7E-7 2-7E-7 2.2E-7 2.7E-7 2.7E-7 J.

S 1.7E-7 1.8E-7 2.0E-7 1.8E-7 1.8E-7 2.2E-7 2.0E-7 1.8E-7 2.7E-7 1.3E-7 1.8E-7 2.1E-7 SSW 1.9E-7 2.1E-7 2.2E-7 2.0E-7 2.1E-7 2.3E-7 2.3E-7 2.1E-7 2-8E-7 1.7E-7 2.1E-7 2-2E-7 SW 5.3E-7 5.5E-7 6.2E-7 7.3E-7 7.4E-7 8.2E-7 1-4E-7 1.4E-6 1-6E-6 2-6E-7 3.3E-7 4-lE-7 WSW 4.7E-7 4-1E-7 5.8E-7 5.5E-7 4.7E-7 6.8E-7 8.6E-7 7.1E-7 1.1E-6 3.3E-7 3.7E-7 6.3E-7 2.7E-7 2.1E-7 3.1E-7 3.0E-7 2.2E-7 3.4E-7 3.8E-7 2.5E-7 4.6E-7 2.1E-7 2.2E-7 3-5E-7 WNW 3.3E-8 4.3E-8 3.6E-8 3 4E-8 4.3E-8 3.7E-8 3.8E-8 4.4E-8 4.2E-8 2-4E-8 4-3E-8 2-4E-8 NW 1.4E-8 1.4E-8 1.5E-8 1.5E-8 1.4E-8 1.5E-8'.3E-8 1.4E-8 1.4E-8 1.0E-8 1.4E-8 1.1E-8 NNW 1.8E-8 1.7E-8 1.8E-8 1.8E-8 1.7E-8 1.8E-8 1.7E-8 1.7E-8 1.7E-8 1.2E-8 1.7E-8 1.3E-8

TABLE 5B RADIATION DOSE TO NEAREST INDIVIDUAL FROM LIQUID RELEASES IN MREM First Quarter Total Body Bone Thyroid Second Quarter Total Body Bone Thyroid Third Quarter Total Body Bone-Thyroid Fourth Quarter Total Body Bone Thyroid Adult 4.7E-3 3.8E-3 5.1E-3 6.6E-3 6.0E-3 1.4E-3 2.2E-3 1.7E-3 2.4E-3 1.2E-2 9.3E-3 3.9E-3 Teen 2 ~ 8E-3 4.0E-3 4.5E-3 3 'E-3 6.3E-3 1.2E-3 1.3E-3 1 ~ 8E-3 2.1E-3 7.2E-3 9.8E-3 3.5E-3 Child 2.0E-3 5.1E-3 7.5E-3 2.0E-3 7.9E-3 2.0E-3

- 7.9E-4 2.3E-3 3.6E-3 3.2E-3 1.2E-2 5.7E-3 Infant 1.1E-3 9.6E-5 7.5E-3 6.4E-4 1.9E-4 2.0E-3 3.2E-4 4.5E-5 3.6E-3 4.1E-4 1.8E-4 5.8E-3

Table 6A ACTUAL WHOLE BODY DOSE HO.

OF PERSONNEL (> or =

100)

ROCHESTER GAS 8 ELECTRIC CORPORATION GINNA STATION NUMBER OF PERSONHEL AND MAH-REM BY WORK AHD JOB FUNCTION FOR 90/01/01 - 90/12/31 TOTAL MAH-REM WORK PERMIT CATERGORY r

REACTOR OPERATIONS 8 SURV WORK GROUP MAINTENANCE PERSOHHEL OPERATING PERSONNEL HEALTH PHY.

PERSONNEL SUPERVISORY PERSOHHEL EHGIHEERIHG PERSONNEL CONTRACT WORKERS 205 2

53 35 9

STATION EMPLOYEES 36 30 14 16 0

UTILITY EMPLOYEES 84 0

4 9

2 CONTRACT WORKERS 1.261 0.822 16.935 1.901 0.223 STATION EMPLOYEES 1.641 6.990 2.669 2.462 0.000 UTILITY EMPLOYEES 0.401 0.000 0.265 1.000 0.000 ROUTINE MAIHTENAHCE MAINTENANCE PERSONHEL OPERATING PERSONNEL HEALTH PHD PERSONNEL SUPERViSORY PERSONNEL ENGINEERING PERSOHNEL 264 2

50 30 8

37 21 14 16 0

158 0

4 7

2 48.354 0.000 10.180 4.769 1.170 4.340 1.074 5.046 0.549 0.000 12.416 0.000 0 ~ 186 0.654 0.197 IHSERVICE INSPECTION MAINTENANCE PERSONNEL OPERATING PERSONHEL HEALTH PHY.

PERSONNEL SUPERVISORY PERSONNEL EHGINEERIHG PERSONNEL 48 0

9 14 4

20 1

2 8

0 28 0

0 6

0 1 ~ 007 0 ~ 000 0.064 0.710 0.202 0.176 0.063 0.181 0.318 0.000 0.803 0.000 0.000 0.127 0.000 MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHD PERSONNEL SUPERVISORY PERSONNEL EHGIHEERIHG PERSONNEL 266 2

43 36 8

37 22 10 16 0

157 0

3 9

2 80.717 0.017 3.127 7.717 1 ~ 120 6.974

1. 519 1.288 2.263 0.000 69.181 0.000 0.011 3.547 0.108 WASTE PROCESSING MAINTENANCE PERSONNEL OPERATING PERSOHHEL HEALTH PHY. PERSONNEL SUPERVISORY PERSOHHEL EHGINEERIHG PERSONHEL 28 2

15 2

0 1.131 0.020 1.806 0.006 0.000 0.035 0.035 1.244 0.117 0.000 0.011 0.000 0.098 0.003 0.000 REFUELING MODIFICATIONS MAINTEHANCE PERSOHHEL OPERATING PERSONNEL HEALTH PHY.

PERSONNEL SUPERVISORY PERSONNEL EHGINEERIHG PERSOHHEL MAIHTENANCE PERSONNEL OPERATIHG PERSONNEL HEALTH PHY.

PERSONNEL SUPERVISORY PERSOHNEL ENGINEERING PERSONNEL 1

11 4

2 19 0

0 6

1 13 13 4

3 0

23 0

1 3

0

/

1 0

0 0

0

13. 181 0.003 1.703 0.187 0.216 0.102 0.000 0.000 0.070 0.000 1.477 2.267 1.034 0.443 0.000 0.052 0.000 0.000 0.000 0.000

1. 123 0.000 0.012 0.153 0.000 0.047 0.000 0.000 0.000 0.000 TOTAL MAINTEHANCE PERSONNEL OPERATING PERSOHHEL HEALTH PHY.

PERSONNEL SUPERVISORY PERSONNEL EHGIHEERING PERSONNEL 277 2

55 37 10 37 30 14 16 0

161 0

4 9

2 145.753 0.862 33.815 15.360 2.932 14.694 11.948 11.463 6.153 0.000 83.982 0.000 0.572 5.483 0.305 GRAND TOTAL 381 96 176 198.722 44.257 90.342

<f I'4

Table 6B STANDARD REPORT OF PERSONNEL WHOLE BODY EXPOSURE 1990 DOSE REM NUMBER OF PEOPLE 00.000 00.001 00.101 00.251 00.501 00.751 01.001 02.001 03.001 04 F 001 00.000 00.100 00.250 00.500 00.750 01.000

-, 02.000 03.000 04.000 05.000 903 349 183 201 125 62 67 4

0 0

Total number of personnel monitored 1894 The total collective dose.for 1990 is 346.7 person-rem based on the sum of all personnel TLD badge readings.

FIVE HIGHEST EXPOSURES FOR THE YEAR B

C D

E 2 '10 2 '69 2.250 2 '50 1.946 This report contains all personnel monitored during 1990.

I T

Process Control Pro ram for Ginna Station Rochester Gas and Electric Corporation Revision 4

December 21 1990

~'I C.

~

I

) I. Introduction

~ 'he Radiological Effluent Technical Specifications require the establishment of a Process Control Program (PCP).

The PCP herein is a manual outlining the method for processing wet solid wastes and for solidification of liquid wastes.

It includes applicable process parameters and evaluation methods used at Ginna Station to assure compliance with the require-ments of 10 CFR Part 71 prior to shipment of containers of radioactive waste from the site.

The Ginna PCP encompasses five types of solid wastes:

a ~b.

C ~

d I e.

Cemented Evaporator Bottoms Solidified Sludge Oily Waste Dewatered Bead Resin Filters A radwaste sampling and analysis program has been instituted to assure compliance with 10CFR Part 61.

Scaling factors have been developed to calculate concentrations of hard to measure isotopes from more easily determined isotopes.

The scaling factors will enable concentrations of all required isotopes to be determined for each radwaste shipment.

A'll radioactive waste is shipped to a licensed burial site in accordance with applicable Nuclear Regulatory Commission, Department of Transportation, and'State Regulations, including burial site regulation requirements.

To assure personnel exposure is minimized, ALARA consider-ations are addressed in all phases of the solidification process.

k k

'l. Cemented Eva orator Bottoms A.

General Description The waste holdup tank, located in the auxiliary building, accepts liquid waste from all floor drains, certain system

drains, resin sluice water, laundry and shower waste, spent fuel pit leak off and the chemical drain tank.

The liquid from the waste holdup tank is processed through cuno type filters to the waste evaporator.

The waste evaporator processes water from the waste holdup tank in a batch mode.

The distillate is polished by mixed bed (HOH) demineralizer resin, collected in waste condensate and monitor

tanks, sampled, analyzed and released to the discharge canal.

As an alternative, the boric acid evaporator can be used to process excess waste water which has been transferred to a

CVCS holdup tank.

Mixed bed (HOH) resins can also be used to polish the distillate.

Sampling and release is done as for waste evaporator distillate.

The concentrates from either evaporator are processed through the waste evaporator feed tank for disposal through the drumming system as described in this section.

A third alternative available is vendor supplied deminerali-zation systems which utilize mixed media filtration, anion (OH), cation (H) and mixed bed (HOH) resin to process water from the waste holdup system.

This process can be utilized on a once through and/or recirc mode.

The product water is

sampled, analyzed and released to the discharge

canal, the same as for the evaporator systems.

Spent resins are sluiced from the vessels and shipped as described in Section V, Spent Bead Resin.

The operation of the evaporators is controlled by several operating procedures, S-3.4C, D, E, and F for the boric acid evaporator and S-4.1A, B,

and C for the waste evaporator operation.

The vendor 'demineralization systems are also controlled by plant procedures.

The currently used system is controlled via S-4.1.27.

The parameters used to control the batch operations are boric acid concentration and gross

,degassed activity.

These concentrations are limited by procedure although activity may be further limited by burial ground dose rate limits.

F w I V

w ll' J

~E 14 kC

-1 lg S

The drumming process is currently, controlled by procedure RD-16.2.

The only chemical parameter which is controlled for

~

~ -"solidification is the solution pH.

A pH, between 6.0 and 8.5 is required to insure that the cement will properly solidify in the least amount of time. If waste is to be drummed with the pH of the waste outside of these parameters a solidifi-cation test is performed to insure proper solidification may be achieved.

The drum filling is controlled by weight and/or level indication to control the amounts of liquid and cement to a predetermined ratio.

This ratio is determined by performing a solidification test.

For normal waste evaporator bottoms within the above pH requirements approximately 1 gal. of evaporator bottoms is solidified with 20 pounds of masonry cement and one pound of meta silicate as an accelerator.

The minimal amount of accelerator limits the rate of the solidification process.

The drums are not. sealed until the solidification is verified complete and the drums are at ambient temperature.

This is typically two weeks after drum filling.

The quality control section is notified prior to solidifi-cation and also prior to shipping so they may perform periodic surveillance on these processes.

A minimum of 10% of all drums are visually checked for proper solidification by the QC section.

If a

drum is found which is not. properly solidified the remaining drums in that batch are also checked.

The drums are then set aside to give additional time for curing.

If the drum(s) still do not solidify, the material can be removed from the drum and mixed with more of the solidifying agent and allowed to cure.

An alternative to resolidification is to place the drum in an acceptable (to the burial site) overpack and ship to the burial site.

As a precaution, a lab test would also be performed on the next evaporator waste to insure there were no unknown matrix or chemical changes in the system which would cause the failure.

Also on one drum from approximately every tenth evaporator bottoms batch, a,drum will be mechanically checked to insure that the total drum contents have properly solidified.

E

+

QC also monitors the drum loading and shipping to insure compliance with all shipping and burial regulations.

After the drumming process has been

. completed, the drums are

weighed, surveyed, serialized and stored in one of the drum storage areas.

Prior to shipment the drums are

cleaned, resurveyed, and labeled, in accordance with the RD-10 series procedures.

l

~@~

<t ~'

RII. Solidified Slud e

A.

General Description Suspended solids and other sludges occasionally require processing.

This material is processed using

. a vendor supplied system.

A Topical Report demonstrating satisfactory processing by a vendor is required.

The vendors procedures would then be PORC

approved, and if necessary, a

50.59 review.

Lab samples are then created and tested.

Following quality control review, full scale solidification would be performed.

r l ~

\\

I 1

II

~

r I

~

(

A.

General Description Oily waste is solidified by methods acceptable to licensed burial sites.

An approved method is to add an emulsifier to the oily waste, then water at a neutral pH.

The mixture is then solidified by adding "Envirostone" gypsum cement.

The method is described in the RD-16 series of procedures.

An alternative method that may be employed would utilize filtration.

As a vendor supplied system, this would require PORC review and approval.

4 q

I gkh Pl A

fp t

lJ Qs

~

p

S ent Bead Resin A.

General Description Bead resin is used to remove chemical and radioactive contamination from the reactor

coolant, the chemical and volume control system, the spent fuel pool, and the liquid waste processing

system, and may be used to process the effluent stream described in Section IIA of the PCP.

B.

Primary Processing When the resin is exhausted or reaches a radiation limit, the spent resin is sluiced to one of two 150 cubic foot spent resin storage tanks.

After sufficient resin has been collected in one of the storage tanks, a QA order is initiated for use of a transport cask certified by the NRC for trans-porting greater than Type A quantities of radioactive material.

Upon arrival on site, the transport cask is inspected using a

Quality Control Inspection Procedure (QCIP) specific for each type of cask to ensure the cask meets all the requirements of the Certificate of Compliance and 10 CFR 71.

A liner, which contains internal piping to completely dewater the resin, is installed in the cask.

The cask is handled, loaded and unloaded using procedure RD-10 series specific for the model cask used.

Piping is run from the drumming station to the manway in the top of the liner.

Using procedures, spent resin is then slurried from the spent resin storage into the liner with water used for sparging and mixing the resin and nitrogen gas pressure used to move the resin.

A representative sample of the resin is obtained and the concentration of each radioisotope is calculated.

After the resin is dewatered or

cemented, the liner is capped and sealed and the top is put back on the cask.

C.

Effluent Stream Processing Liquid waste from all floor drains, regeneration

wastes, certain system

drains, resin sluice

water, laundry and shower waste and the chemical drain tank may be processed by a vendor supplied temporary demineralizer system in lieu of the evaporator described in Section IIA of the PCP.

The demin system located in the drumming station consists of a

control skid for system isolation and flow control~

mechanical prefilter for roughing filtration and five 15 ft sluicable demin vessels arranged in series.

s

<'a p

l ~

4

Ldt, l

~f gag

When the resin is exhausted or reaches a radiation limit, the spent resin is sluiced to a certified cask, utilizing procedures.

A liner, which contains internal piping to completely dewater the resin, is installed in the cask.

A representative sample of the resin is obtained and the concentration of each radioisotope is calculated.

D.

Shipment Preparation:

The cask is surveyed for radiation and contamination and properly labeled and marked as specified in procedures for packaging shipment of radioactive materials.

The procedures include instructions on any special recgxirements of the burial site to which the shipment is being sent.

A radio-active shipment record is prepared and all necessary shipping papers and instructions are given to the carrier.

The vehicle is placarded, the cask sealed with security seals, and the Quality Control inspection is complete.

The resin is then transported to the burial site.

C ~

I

VI. Filters When filters.become saturated or have a high dose rate, they are dewatered and then replaced.

The spent filters are placed in a

High Integrity Container or solidified in an approved media and shipped in accordance with 10CFR71p 10CFR61 and burial site licenses.

The maximum dose rate allowed on the surface of the container is determined by the shielding of the package in which the container is shipped.

Shipping recpxirements for specific packages are addressed in the RD-10 series of procedures.

CVCS Reactor Letdovn

)axed Bed CVCS lon Gos Tacic GOl CVCS Tanks (3)

)axed Bed ion changers Q)

Narrator Tanks Spent Resns Spent Rcsn Storage Tank

<SRST)

Q)

Hector Tank A

Tenp.

)4cteor Radvaste Dorm S)od Cat)on CVCS lon Exchanger Car trtdge GDl Gas S~r Car tr)dge Pter To Ortreong Statxl

)F Hecessory Bore Ac)d Evaporator Reactor Hokcty Vater Tank GOl Volta'ontrol Tank To Lncr

)tscegoneovs Vastes Vaste Hotdup Todc Car trn)ge Fater Vaste Evaporator feed Tank To Dron Staten

)axed Bed ion Exchangers Q)

Bor):

Ac)d Storage Tanks Encrgency Borate Rehrn Reactor Coo[ant Bran Tonk Contonnent S~s La@oratory Orans E~t Orans F)oor Orans Dcnnerout~

Regeneronts ljPERATIGNAL RADVASTE DIAGRAM R.

E. GINNA FIGURE I RM-15 RE'vr'.

00 11/6/90

Vaste Evaporator Feed Tank

~ent Mixer 55 Gallon Drum Sur yey Inspect Seal Decon~

For ~

Drun ~Storage Label Free Water Cap Scales Spent Resin Storage Tank Sample High Integrity Container Licensed Bur ial Cask~ Ground Was.e Hold-up Tank Dewater Pun AQUI"8 2

RV-16 REV. 00 11/6/90

h 1