ML17251A123
| ML17251A123 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/31/1987 |
| From: | Snow B ROCHESTER GAS & ELECTRIC CORP. |
| To: | Russell W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 8806150181 | |
| Download: ML17251A123 (68) | |
Text
A~CCRLERATED DIRIBUTION DEMONSTTION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8806150181 DOC.DATE: 87/12/31 NOTARIZED: NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G
AUTH.NAME AUTHOR AFFILIATION SNOW,B.A.
Rochester Gas 6 Electric Corp.
RECIP.NAME
, RECIPIENT AFFILIATION DOCKET g
05000244 R
D 05000244 8
NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
SUBJECT:
"Semiannual Radioactive Effluent Release Rept,Jul-Dec 1987."
W/880229 ltr.
DISTRIBUTION CODE:
IE48D COPIES RECEIVED:LTR J ENCL J SIZE:
TITLE: 50.36a(a)(2)
Semiannual Effluent Release Report RECIPIENT ID CODE/NAME PDl-3 LA STAHLE,C COPIES LTTR ENCL 1
0 1
1 I
RECIPIENT COPIES ID CODE/NAME LTTR ENCL PD1-3 PD 5
5 INTERNAL: AEOD/DOA ARM TECH ADV PB 10 01 RGN2/DRSS/EPRPB EXTERNAL: BNL TICHLER,J03 NRC PDR 1
1 1
1 4
4 1
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1 1
1 1
1 AEOD/DSP/TPAB NRR/DEST/PSB 8D NUDOCS-ABSTRACT RGN1 FILE 02 LPDR 1
1 1
1 1
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1 TOTAL NUMBER, OF COPIES REQUIRED:
LTTR 22 ENCL 21
0
SEMIANNUAL RADIOAC IVE EFFLUENT RELEASE REPORT R. E.
GINNA NUCLEAR PLANT ROCHESTER GAS AND ELECTRIC DOCKET NO. 50-244 JULY - DECEMBER 1987 8306150181 871231 PDR ADOCK 05000244 R
nrn..
TE cP Y(
0 0
1.0 2 '
2.1 INTRODUCTION
SUPPLEMENTAL INFORMATION REGULATORY LIMITS 2 '
MAXIMUMPERMISSIBLE CONCENTRATIONS 2 '
2 '
2 '
2.6 3 ~ 0 4 ~ 0 5 '
6 '
7.0 8 '
9 '
10.0
- 11. 0 12 '
13 '
14 '
RELEASE RATE LIMITS MEASUREMENTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY BATCH RELEASES ABNORMAL RELEASES
SUMMARY
OF GASEOUS RADIOACTIVE EFFLUENTS
SUMMARY
OF LIQUID RADIOACTIVE EFFIVENTS SOLID WASTE LOWER LIMIT OF DETECTION RADIOLOGICAL 1MPACT METEOROLOGICAL DATA LAND USE CENSUS CHANGES ANNUAL TABULATION OF PERSONNEL EXPOSURE LEAK TEST OF SEALED SOURCES CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL CHANGES TO THE PROCESS CONTROL PROGRAM MAJOR CHANGES TO RADWASTE TREATMENT SYSTEMS
LZST OF TABLES Table 1A Gaseous Effluents - Gaseous Summation of all Releases Table 1B Gaseous Efiluents - Continuous and Batch Releases Table 2A Liquid Effluents -
Summation of all Releases Table 2B Liquid Efiluents - Continuous and Batch Releases Table 3
Solid Waste and Zrradiated Fuel Shipments Table Table Table Table Table 4A Radiation Dose to Nearest Zndividual Receptor from Gaseous Releases 4B Radiation Dose to Nearest Zndividual from Liquid Releases 5A Liquid Effluents - Summation of all Releases (corrected table for JanMune 1987) 5B Liquid Effluents -
Continuous and Batch Releases (corrected table for Jan-June 1987) 6A Number of Personnel and Man-Rem by Work and Job Function Table 6B Standard Report of Personnel Whole Body Exposure
I ZNTRODUCTZON This Semiannual Radioactive Effluent Release Report is for Rochester Gas and Electric Company's R.E. Ginna plant and is submitted in accordance with the requirements of Tech.
Spec.
Section 6.9.1.4.
The report covers the period from July 1, 1987 through December 31, 1987.
This report includes a
summary of the quantities of radioactive gaseous and liquid effluents and solid waste released from the plant presented iri the format outlined in appendix B of Regulatory Guide 1.21, Revision 1,
June 1974.
Zt includes the annual summary of personnel whole body exposure and man-rem received by work groups.
All gaseous and liquid effluents discharged during this reporting period were in compliance with the limits of the R.E. Ginna Technical Specifications.
2 ~ 0 SUPPLEMENTAL INFORMATION 2 '
The Technical Specification limits applicable to release of radioactive material in liquid and gaseous effluents are:
2 ~ lol 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 vhich would yield
< 500 mrem/yr to the total body and
< 3000 mrem/yr to the skin if allowed to continue ior 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'.2 2'.3 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 vould yield < 1500 mrem/yr to any organ if allowed to continue for a full year.
The dose to an individual, as calculated in the ODCM, from radioiodine, radioactive materials in particulate form and radionuclides other than noble gases vith hali-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.
The release of'adioactive liquid effluents shall be such that the concentration in the circulating water discharge does not exceed the limits specified in accordance vith Appendix B, Table II, Column 2
and notes thereto oi 10CFR20.
For dissolved or entiained noble gases the total activity due to dissolved or entrained noble gases shall not, exceed 2 E-4 uCi/ml.
2 '
2 ~ 2 ~ 1 2 ~ 2 ~ 2 The dose or dose commitment to an individual as calculated in the pDCM from radioactive materials in liquid effluents released to unrestricted areas shall be limited:
(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.
e s
b e o
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 concentration values specified in 10CFR20, Appendix B, Table ZZ, 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'ffluents.
2 '
The release rate limits for fission and activation gases from the R.G.fE Ginna plant are not based on the average energy of, the radionuclide mixture in gaseous effluents therefor, this value is not applicable.
- Hovever, the average energy of'he radionuclide mixture was, 0.285 Mev.
2 '
2 '
Gamma spectroscopy was the primary analysis method used to determine the radionuclide composition and concentration of gaseous,and liquid effluents.
Composite samples vere analyzed for Sr-89, Sr-90 and Fe-55 by a contract lab-oratory.
Tritium and alpha analysis vere done using liquid scintillation and gas flov proportional counting respect-ively.
The total radioactivity in effluent release vas determined from, the measured concentration of each radionuclide present and the total volume oi effluents released.
1.
2 ~
3 ~
4 ~
5 ~
6 ~
Number of batch release:
Total time period ior 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 during periods of release effluent into a flowing stream:
2.89 E+02 F 88 E+04 2'2 E+02 6.5 E+01 1 ~ 10 8+01 li28 8+06 2.5. 2 gg~ua 2 '
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:
7E+00 2.9E+03 min 7.06E+02 min 4.20E+02min 2.95E+02 min 3 '
There were no abnormal releases of liquid or gaseous affluents during the reporting period.
SUMMARY
OP GASEOUS RADZOACTZVE EPFZRJENTS The quantities of radioactive material released in gaseous effluents are summarized in tables 1A and 1B. All releases were considered to be elevated releases.
4 '
SUMMARY
OP LZQUZD RADZOACTZVE EPPHJENTS The quantities of radioactive material released in liquid effluents are summarized in tables 2A and 2B.
5.0 SOLZD 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 de'termined by gamma spectroscopy and non-gamma emitters were calculated from scaling factors de-termined by an independent laboratory from representative samples of that waste type.
6.0 LOWER LIMIT OF DETECTION NOT MET There were no releases for which any gamma emitting radionuclide did not meet the required lover limit for detection.
For the determination of Sr>>89, the contract laboratory results for the third quarter, aa has occured in several quarters during the years of 1984, 1985 and 1986, did not meet the Technical Specification requirement for the lower limit of detection of 5E-8 uCi/ml.
The laboratory LLD value at the time of analysis vas 1E-8 uCi/ml, but the quarterly composite sample must be decay corrected to the midpoint of the.sampling period, causing the Tech Spec LLD requirement to be exceeded.
The decay correction factor exceeded a value of 5 and gave a calculated LLD greater than SE-8 uCi/ml.
A tabulation of the missed LLD's by the contract laboratory for the years oi 1986 and 1987 are!
Date Analysis Sample Value 1986 First Qt.
Sr-89 1986 Fourth {}t. Sr-89 N
Sr-89 ll Sr-89 1987 Third Qt.
Sr-89 Sr-89 Retention Tank Boiler Return House Heating Boiler High Cond.
Waste Tank High Cond.
Waste Tank Retention Tank 6ol E-8 6.4 E-8 8 '
E-8 6.9 E-8 6 '
E-8 8 '
E-8 7o0 RADIOLOGICAL IMPACT 8 '
An assessment of doses to the maximally exposed individual from gaseous and liquid effl'uenta vas performed for locations representing the maximum dose.
Zn all cases, doses vere well below Technical Specification limits.
Doses vere assessed based upon actual meterological conditions considering the noble gas exposure, inhalation, ground plane and ingestion pathways.
The ingestion pathvays considered were the produce, vegetable, goat's
- milk, cow's milk and meat pathvay.
The results of this assessment are presented in tables 4A and B.
METEOROLOGICAL DATA 9 '
The annual summary of hourly meteorological data collected during 1987 is not included vith this report, but can be made available at the R.G.&E Ginna Plant as alloved by our Technical, Specification.
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 )ob function required by Technical Specification 6.9.2.2 and 10CFR20.407 is included as tables 6A and 6B.
ll~ 0 LEAK TEST OF SEALED SOURCES 12.0 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.
During an internal audit of Technical Specif ication reporting requirements, there was no record found that a
copy of Revision 2 of the ODCM was sent to the NRC with the January-June 1986 Semi-Annual Radioactive Effluent Release Report.
A Copy of Revision 2 is included with this report.
The ODCM was revised to show:
a.
A change in the flow of the circulating discharge water used for dilution of releases before leaving the restricted boundary.
All example calculations affected by this change were corrected.
b.
The deletion of the measurement of direct radiation by film badge.
The plant discontinued the use of personnel dosimetry by film badge and it was deleted from the environmental program.
The use of TLD mon'itoring continues.
c.
The addition of the word instantaneous to section VI.
13 ~ 0 14 '
d.
The correction of several typographical errors overlooked previously.
CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)
There were no changes to the PCP during the reporting period.
MAJOR CHANGES TO RADWASTE TREATMENT SYSTEMS There were no ma)or changes to the Radwaste Treatment Systems during the reporting period.
TABLE 1 A GASEOUS EFZXUENTS - SUMMATION OF ALL RELEASES JULY 1 - DECEM'1, 1987 Unit [Quarter )Quarter JEst. Total(
A.
Fission
& activation gases 1.
Tota re e se
+
7.
+00 3 ~
B.
Zodines o t 7
0
<<0
+01 t o t C.
Particulates 3 ~
7 7
<<7 7 <<
+0 D.
Tritium 2o v
e eas 8
+0
<<0
+00
TABLE 1 B GASEOUS EFFIIJENTS - CONTINUOUS AND BATCH RELEASES JULY 1 - DECEMBER 31, 1987 1.
Fission gases 4
2.
- 3. Particulates 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, 1988.
ZOOID HTXHBGS SUhSKXCN OF ALL RELEASES JULY 1 - IXKXMBER31, 1987
( Unit (Quarter (Quarter (Est.Total(
A.
Fissice and activatim peaducts (1.
Tatal release (nct including tritium, (2.
Anemia diluted ccxxxmtraticn 6
W
.78 W3 I
I B.
Tritium (2.
Avexac~ diluted ccnccmtxatica I
I' 7
C.
Dissolved and entwined gases (2.
Average diluted ocmentratica D.
Gmss alpha radicactivity
TABLE 2B EPFLtHBIT AND WASTE DISPOSAL SEMIANNUAL REPORT LIQUIED EFFLUENTS - CONTINUOUS AND BATCH RELEASES JULY 1 - DECEMBER 31, 1987 8
7 0
B~
7 7-7 7
w 7
- 7. 67 NOTE:
Isotopes for which no value is given were not identified in
'pplicable releases.
- Sample sent out ior analysis but results not yet received.
Data for identified isotopes will be included with next semi-annual report for January - June, 1988
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) 6-month
) Est. Total Unit Period Error 4 m
)1 '7E+01 (5.0E+00 Ci 2.01E+02 S.OE+00 m
le 1 E+01
)5 AL OE+00 Ci 1.44E+00 1.0E+Ol mCi ci
- 2. Estimate of major nuclide composition (by type of waste) a 0 b.
C ~
d0 iron-55 coba t-60 ickel-63 cesium-134 cesium-137 cobalt-58 h dro en-3 cobalt-60 iron-55 cesium-137 9.03E+00 4.89E+Ol 1.19E+Ol 7.05E+00 1.29E+Ol 4.
7E+00 4.07E+00 4.17E+Ol 4.17E+01 1.66E+Ol
- 3. Solid Waste Disposition Number of Shi ments P
Mode of Trans ortation Destination Enclosed Van Barnwell, S.C.
Lowboy Barnwell, S.C.
B.
IRRADIATED FUEL SHIPMENTS (Disposition)
Number of Shi ments Mode of Trans ortation Destination None
TABLE 4A RADIATION DOSES TO NEAREST INDIVIHJALRECEPTOR HNN GASEOUS RELEASES IN REM 1987 QUARTER 1 N
NNE NE BNE 8
SW MSW W
WNW NW
)1 1E-712 'E-7(1.5E-7
)8.38-8)1.6E-7)1.28-7 11.2E-712.0E-711.68-7
)7.08-8)1.18-7)9.28-8
)8 4E-7)1.3E-6)1.6E-6
- 11. OE-611.58-611.9E-6
)1.98-6)3.08-6(4.0E-6
)1.1E-6)lo88-6)2.1E-6 17.8E-711.3E-611.5E-6
)2.78-713.78-7)5.1E-7
)4.5E-S)6.78-8(6.58-8 12 ~2E-811. 9E-S I 3.7E-S
)9.8E-8)2.2E-7)1.5E-7 (7'8 811.6E 7(l 1E 7 19'E 812 F 08 711 4E 7
)5.78-8(1 18-7)7.7E-S
)8'E-7(1.8E-6)1.28-6 12.1E-614.88-612.9E-6 17.08-7)1.38-6(1.4E-6
)8.48-7(1.5E-6)1.6E-6
)1.6E-6(3.0E-613.5E-6
)9.1E-7(1.8E-611.98-6 16 ~ 6E 7(1.38 611.3E 6
) 1 ~ 68 811 9E 812 9E 8
)1.28-7)2 '8-7)1.5E-7
)8.68-8)1.6E-7)l.lE-7
)1.3E-7)2.0E-7(1.6E-7
)7'8 8)1 1E 7)9oOB 8
)1.0E-611.88-6)l 38-6 12'E 6)4 ~ SEWI3 ~ 18 6
)8.88-7)1.38-6)1.4E-6 (1 18-6)1.5BW)1.78-6 (2'E-6)3 'E-6(3 'E-6
)1.18-6)1.88-6)1.9E-6 (Sol8-7)1.38M)1.38-6 (2'8-7)3 '8-7)4 'E-7 (1.68-7)2 '8-7(2 'E-7
)4.68-8)6.78-8)6.18-8
)2.48-8)1.9E-S)3.58-8 11.28 712 2E 711 5E 7 1'8-811.6E-711.28-7 (1.38-7)2.0E-7)1.7E-7
)7.3E-S(1.1E-7(9.48-8 Il.OE-611.8E-611.38-6
)2.5E-614
~ 88-6)3 '8-6 IS. 88-71 1. 3E-61 1. 6E-6 ll 18 6)l 5E 611.98"6 12 F 08 613 ~ OEW(3'8 6
)lolE-6)1.88-6)2 '8-6
)8.18-7)1.38-6)1 5E-6
)2.88-7(3 '8-7)5.0E-7 11.68-712.3E-7(2 'E-7
)4.6E-S)6.78-8)6.58-8
)2.4E-S(1.98-8)3 'E-8
)Total
)
)
)Total
)
)
)Total
)
)
)Total S
TABLE 4A RADIATION DOSES TO NEAREST INDIVIIXJALRECQPXOR HKN GASEOUS RELEASES IN RM 1987 QUiQGKR,2
)Total
)
(
)Total
)
)
)Total
(
)
(Total
(
N NNE NE EHE E
SW MSW W
MNW NM
) 6.6E-S) 5.4E-S) 6.6E-S
)6.4E-S)5.3E-S)6.4E-S
)9.6E-S)7.1EW(9.6E-S
)6.7E-S)5olE 8)6.7E-S
)7.1E-7)6.2E-7)7o2E-7
)3 'E-7)3.2E-7)3.2E-7 (3.6E-7)3.3E-7)3.7E-7 15.5E-7)5.5E-7)5.6E-7
)7.4E-7(6.8E-7!7.5E-7
)1 5E-6)1.2EW)1.5E-6 (9.0E-7(7.6E-7)9.1E-7
)2.5E-7(2.1E-7)2.5E-7
)1.6E-7)1.9E-7)1.7E-7
)3.4E-S)2.2E-S(3.4E-S
)6.6E-S)5.4E-S)6.6E-S
)6.4E-S)5.3E-S)6.4E-S
)9 6E-S)7.1E-S)9.6E-B
)6 7E-S)5.1E-S)6.8E-S
)1.3E-6)9.7E-7)1.4E-6
)7.5E-7)6.2E-7)7.7E-7 (3 2E 7)3.2E 7)3 3E 7 (3 9E-7)3 3E-7)3.9E-7 15 7E-7)5.5E-7)5.8E-7
)1.5E-6(1.2EW)1.6E-6 (9 5E 7(7.6E 7(9 6E 7 (2.6E-7(2.1E-7)2.6E-7 (1.7E-7(1.9E-7(1.7E-7
)3.4E-S)2.2E-S)3.4E-S
)6.1E-B)5.4E-S(6.6E-S 15.9E-B(5.3E-B)5.9E-S IS.SE-S(7'lE-BIB.SE-B
)6.2E-B)5.1E-S)6.2E-B
)1.6E-6(9.7E-7)1.6E-6
)S.SE-7(6.2E-7(9 'E-7
)4.7E-7)3 'E-7)USE 7
)6.4E-7)5.5E-7)6.5E-7
)8.6E-7)6.8E-,7)8 7E-7
)1+BE-6)1.2E-6)l.SE-6
)3 'E-7)2.1E-7)3.1E-7
)1.9E-7)1.9E-7)1 9E-7
)3 'E-S(2.2E-S)3.1E-S
) 4.4E-S) 5.4E-S (6.6E-S
)4.3E-S)5.3E-B)4.3E-S
)6.2E-B(7.1E-B)6 'E-S (4.4E-S)5.1E"8)4.4E-S
)8.7E-7)9.7E-7)9.4E-7
)6.7E 7(6 2E 7(7'E 7
)2.7E-7(3.2E-7(2.8E-7
)4'E-7(3 'E-7)4 'E-7 I
15 ~4E-7)5.5E-7)5.6E-7 I
)6'E-7)6.8E-7)6.9E-7
)1.5E-6)1.2EW)1.6E-6 IS ~28-7)7.6E-7)8 'E-7
)2.4E-7(2.1E-7)2.4E-7
)1.1E-7(1.9E-7)1.1E-7
) 2.1E-B) 2 ~ 2E-S) 2.2E=S
TABLE 4A RADIATION DOSES TO NEAREST INDIVIHJALRECEPMR FBM GASEOUS RiKZASBS IN REM 1987 QUARH9t 3
)Total
)
)
)Total
)
I
)Total
)
)Total
)
N
)1.1E-7)3.2EW)l.lB-7 NNE
)1.5E-7)4.3E-S)1.5B-7 NE
)1 3B-7)5.1B-8)1.3E-7 BNE
)6.9E-S)2.5E-S)6.9B-S E
)1.4E-6)3.9E-7)1.4E-6 EBB
)6.7E-7)2.3E-7)6.9B-6 SE
)5.8E-7)2.5E-7)5.9E-7 SSE
)6.7E-7)2.0E-7)6o9E-7 S
)2.7E-6)1.6E-6)2 'E-6 SSW
)S.OE-7)3.2E-7)8.3B-7 SW
)1~9E 616.5E 7)1.9E 6 MSW
)1.2E-6)3.1E-7)1.3E-6 W
)6 4E-7)1.6E-7)6.4B-7 MNW
) 2 ~ SE-7) 1.2B-7) 2.8E-7 NW
)3.7E-S)1.2E-S)3 7E-S
)1.1E-7)3 'E-S)l.lB-7 ll.SE 7)4'E 8)1.5E 7
)1.3E-7)5.1E-S)1.3E-7
)7.0E-S)2.5E-S)7.0E-S
)1.5E-6)3 9E-7)1.5E-6
)7.2E-7)2.3E-7)7.4E-7
)6.1E-7)2.5B-7)6.3E-7
)7.2E-7).2 OE-7)7.4E-7
)2.9E-6)1.6EW)3.0E-6
)8.6B-'7)3.2E-7)8.9E-7
)2.0E-6)6.5E-7)2.1E-6
)1.4E-6)3 1E-7)1.4B-6
)7.0E-7)1.6E-7)7.1E-7
)3.0E-7)1.2E-7)3.0E-7
)3o7E-S)1.2E-S)3.7E-S
)1.0E-7)3 'B-S)1.0E-7
)1.4E-7)4 ~ 3E-8 I 1. 4E-7 I 1. 2E-7) 5 ~ 1B-8 I 1. 2E-7
)6.3E-S)2.5E-S)6.3E-S
)8.9B-7)2.3E-7)9.3E-7
)7 3E-7)2.5E-7)7.7E-7
)8.7B-7)2.0E-7)9.0E-7
)3 'E-6)1.6E-6)3.8E-6
)1.1E-6)3 'E-7)1.1E-6
)2 'E-6)6.5E-7)2.6B-6
)1.8E-6)3 'E-7)1.8E-6
)9.2B-7)1.6B-7)9.3E-7
)USE-7)1.2E-7)3.9E-7
)3 'E-S)1.2E-S)3 'E-S
)6.5B-S)3.2E-S)6.5E-S
)8+7B-S)4.3B-S)8.7E-S
)7.7E-8)5.1E-.S)7 'E-S
)4.0E-S)2.5E-S)4.0E-S 17 ~ 2E 7)3'E 7)7 9E 7 13.7E-7)2.3B-714.2E-7
)3.8B-7)2.5E-7)4.2E-7
)USE 7)2'E 7)4'B 7 I
)1.8B-6)1.6E-6)2.0E-6 15.2E-7)3.2E-7)5-6E-7
)1.5E-6)6.5E-7)1.6E-6
)1.0E 6)3 lE 7)1 1B 6
)4.3E-7)1.6B-7)4.5B-7
)1.0E-7)1.2E-7)1.0E-7
)2 'E-S)1.2E-S)2'lB-S
TABLE 4A RADIATION DOSES TO NZ3REST INDIVIHJALRECEPTOR FRY GASEOUS RE'XZASES IN REM 1987 QUARTER 4 I fat
)Total
)
)
)Total
)
(
)Total
(
I
)Total
)
Bod S
N NNE NE ENE E
MNW NW 15.6E-SI4.9B-SI5,6E-S
)4.7E-B(4.0B-S)4 'E-S
)5.5E-S(4.4E-7)5.6E-S (4'E-S)3 'B-S)4.6E-S 16 3E 715.9E 716 5E 7 (USE-7)4.3E-7)4.9E
.7 (6.8E-7)5.6B-7)6.8E-7 15.4E-7)EBB-715.6E-7
)7.9E-7(7.1E-7)S.OE-7
)1.2E-7)1.4E-7)1.2E-7
)1.1E-7)1.5E-7)l.lE-7 (6.2E-S)6.9E-S(6 'E-B (1.5E-7)1.5E-711.5E-7 (9.1E-B)1.3E-7)9.2B-B
)1.4E-S)1.2E-S)1.4E-S
) 5.6E-S) 4.9E-S) 5.7E-S
)4.7E-S)4.0E-S)4 'E-S
)5.6E-S)4.4E-S)5.6E-S 14.6E-S)3.4E-S)4.6E-S 16.4E-715.9E-716.5E-7 (4.8E-7(4.3E-7(4.9E-7
)6.8E-7)5.6E-7(6.8E-7
)5.4E-7)4.8E-7(5.7E-7
)8 ~ OE-7)7 'B-7)8 ~ 1E-7 11.2E 711 4E 711 2E 7 (1.1E-S(1.5E-S(1.1E-7
)6.3E-S)6.9E-S)6 3E-B
)1.5E-7)1.5E-7)1.5E-7
)9o2E 8)lo3E 7)9 2B 8
)1.4E-S)1.2E-S)1.4E-S
)5.2E-S)4.9E-S)5.2E-B
)4.4E-S)4.0E-S)4.4E-S
)5.1E-S)4 'E-S)5.2E-B
)4 'E-S)3.4E-S)4.2E-B 16.1E 7)5 9E 7)6'E 7 14.6E-7)4.3E-7(4.7E-7
)6.3E-7(5.6E-7(6.4E-7
)5.3E-7)4.8E-7)5.7E-.7
)7.5E-7)7.1E-7)7.7E-7
)lo2E-7(1.4E-7)1.3E-7 ll lE 7(l 5E 711 lE 7 (6+4E-S)6.9E-B(6.5E-S 11.5B-711. 5E-711. 5E-7 (9+OE-S) 1.3E-7) 9.2E-B
)1.3E-S)1.2E-S)1.3E-S
)3+BE-8)4 'E-S)3.8E-S 13 'E-814 'E-SI3.2E-S
)3.7E-B)4 'E-S)3 'E-S
)3.0E-B)3.4E-S)3 'E-B 14.5E-715.9E-714.7E-7 1'E-714 'E-713 'E-7 14.6E-7)5.6E-7)4.7E-7 13 'E-714 USE-714.2E-7 15.7E-717.1E-715.8E-7
)9.3E-B(1.4E-719.9E-B
)1.0E-7(1.5E-711.0E-7 15.0E-SI6 9E-815.2E-S 11.2E-711. SE-711. 2E-7 18 ~ OE-811. 3E-7(9.2E-S 19 ~ 3E-9) 1. 2E-B ) 9.4E-9
TABLE 4B RADIATION DOSE TO NEAREST INDIVIDUAL FROM LI{}UIDRELEASES IN MREM haut
~ee Child Infant First Quarter Total Body Bone Thyroid Second {}uarter Total Body Bone Thyroid Third {}uarter Total Body Bone Thyroid Fourth {}uarter Total Body Bone Thyroid 1.5E-l 1.1E-1 3 'E-3 8 'E-3 6.6E-3 9.0E-4 3 'E-2 2.3E-2 5.4E-4 9.1E-3 6.5E-3 1.0E-3 8 'E-2 1.2E-l 2 'E-2 5 'E-3 7 'E-3 1.0E-2 1 ~ 7E-2 2.4E-2 4.6E-3 5.3E-3 6.9E-3 4 'E-3 3 'E-2 1 5E-1 3.5E-3 2.7E-3 8.7E-3 1.0E-3 7.1E-3 3.0E-2 6.0E-4 2 'E-3 8 'E-3 1.1E-3 1.4E-3 2 'E-4 1.9E-3 7 'E-4 2.0E-5 7 'E-4 4 'E-4 5.7E-5 4 'E-4 8 'E-4 1.6E-5 8 'E-4
TMKZ 5 A EPFIHEHZ N %LPGA MSEMALSHGXMRL HERR%
LIQUID IHTYIJZÃIS SUhSAXXCN OF ALL HEXHSES CGFKMIKD NGE KR JMUAKf 1 - ZINE 30, 1987 HEKRC I
A.
Fission and activation products I 1.
Total release (not including tritium, I 2.
Average diluted ccacentratica
/ Unit, /Quater (Quarter JEst.Verb&/
B.
( 2.
Average diluted ocacentratica I
I C.
Dissolved anR entxainxi gases I 2 ~
Average diluted accxx~tice I
I D.
Gross alpha radioactivity
TABLE 5 B
EFFLUENT AND WASTE DISPOSAL SEIKAKWALREPORT LIQUID EFFLUENTS - CONTINUOUS AND BATCH RELEASES CORRECTED PAGE FOR JANUARY 1 - JUNE 30'987 REPORT 7
7 Note:
isotopes for which no value is given were not identii'ied in applicable releases.
RQfiE GDQQ, STATI(N NQQZR OP PERSCSHEXs AND lQN RM Kf NKKMD JOB 1%HCXXOg (1987)
I
~RK PER~T SUFFIX I CONTRACT I STATION I UTXLITY I CONTRACT I STATION I UTILITY I
I REACTOR OPXSATZONS IIMatntenance Personnel I Operating Personnel IHealth Physics Personnel I
I Supervisory Personnel I
i IMaintenance Personnel IOperating Personnel IHealth Physics Personnel I
I Supervisory Personnel I
I IMaintenance Personnel I Operating Personnel IHealth Physics Personnel I
I Supervisory Personnel I
105 0
28.
24 144 0
28 26 42 0
15 6
54 23 12 16 51 8
12 16 18 2
4 8
70 0
0 13 141 0
3 15 49 0
0 6
1.995 Oo000 3 ~ 168
- 1. 552 29 '498 0 F 000 4.429 4'68 3.686 0+000 0.595 0.105 6'35 9'74 4 '08 2.570 16.148 0.322 2'58 2 '33 Oe285 0'02 0.156 0.166 lo 652 0.000 0.000 0.487 0.08 19.048 0.000 0.330 3.042 4.089 0.000 0.000 I
0'79 5
IMaintenance Persannel I Operating Personnel I Health Physics Personnel I
I Supervisory Personnel I
132 0
26 24 49 12 10 13 142
- 46. 635 I
- 8. 401 0
- 0. 000 I
- 0. 122 2
I 1.832 I
1.504 12 I
4'57 I
0.855 I
63.306 0.000 0.000 0.617 0
0 IMaintenance Personnel IOperating Personnel IHealth Physics Personnel I
I Supervisory Personnel I
IMainterMLnce Personnel IOperating Baraannel I Health Hxysics Personnel I
I Supervisory Persannel I
24 0
6 1
61 0
19 15 22 5
6 5
29 3
8 2
13 0
3 2
130 0
0 7
4.560 0 F 000 2.554 0.092
- 7. 031
- 0. 000 2.328 1.280 0.460 0'47 0.095 0.097 3 '45
- 0. 001 1.957 0.365 0.183 I
- 0. 000 I
0.575
- 0. 175 24.293 0.000
- 0. 000.
- 0. 075 IHealth Physics Personn ISupervisory Personnel I
28 IMaintenance Persannel 159 IOperating Personnel 0
el I 28 54 23 12 16 153 0
3 15 93.405 0+000 14.906 11.954 35'74 10'68 10.978 6+286 111. 571 0.000 0.905 4.775
~ 7
%7TE: This report is based era SRPD (~ ReaRing Pacjaat Dosimeter)
~mgasmes tahar' the NXK PIICXTS.
TABLE 6B ROCHESTER GAS 4 ELECTRIC CORPORATION ROBERT E GINNA STATION STANDARD REPORT OF PERSONNEL WHOLE BODY KICPOSURE (REM)
BY K&OSURE GEKRJPS FOR THE YEAR OF 1987 RBQUZRED AS PER 10CPR20 PARAGRAPH 20 ~ 407(h)
DOSE (REM)
NUMBER OF PEOPLE 0 F 000 0'00 Oo250 0 '00 0'50 1 F 000 2.000 F 000 F 000 5.000 F 000 F 000 F 000 F 000 10.000 11 F 000 12 F 000 Oe 000 Oi 100 0'50 0'00 0'50 1 F 000 F 000 F 000 F 000 5.000 F 000 7.000 8o000 F 000
- 10.000 11.000 12 F 000 496 258 153 132 90 39 88 13 0
0 0
0 0
0 0
0 0
0 TOTAL NUMBER OF PERSONNEL MONITORED 1269 PERSONNEL WITH THE FIVE HIGHEST EXPOSURES FOR-YEAR
- Bement, John S.
Polfleit, Peter S.
- Galletto, James E.
Bodley, Warren A.
- Slade, Timothy R.
117-58 1594 065 54-9637 058-48-0822 058-46-2638 060-42-7098 2.905 rem 2.421 rem 2.403 rem 2 388 rex 2 '32 rem
Offsite Dose Calculatin Manual Ginna Station Rochester Gas and Electric Corporation Revision 2
l
Ginna Station Offsite Dose Calculatin Manual TABLE OF CONTENTS Radiological Effluent Technical Specification Section Dose Calculation Manual
~Pa e g 3.5.4 4.12.1 I
I.
Liquid Effluent Monitor Setpoints 3.5.4 4.12.2 II.
Gaseous Effluent Monitor Setpoints 3.9.1.1 III.
Liquid Effluent Release Concen-trations 3.9.1.2 IV.
Liquid Effluent Dose 10 3.9.1.3 3.9.2.3 5.5 V.
Liquid and Gaseous Radwaste Treatment 13 and Operability 3.9.2.1 VI.
Gaseous Effluent Dose Rate 17 3.9.2.2 VII.
Gaseous Effluent Doses 18 4.10.1 VIII.
Environmental Monitor Sample Locations 20 3.9.2.4
~ IX.
Preparation of Special Report to Demonstrate Compliance with Environmental Radiation Protection Standards 32 X.
References 33
LIST OF TABLES AND.FIGURES
~Pa e
Table 1
Dose Parameters for Radioiodines and Radioactive Particulate, Gaseous Effluents Table 2
Dose Factors for Noble Gases and Daughters Table 3
Dispersion Parameter (X~Q) for Long Term Releases, Plant Vent 25 Table 4
1 Dispersion Parameter (D~Q) for Long Term Releases, Plant Vent 26 Table 5
Dispersion Parameter (X/Q) for Iong Term Releases, Containment Purge 27 Table 6
Dispersion Parameter (D~Q) for Long Term Releases, Containment Purge 28 Table 7
Dispersion Parameter
(~X Q) for Long Term Releases, Ground Vent 29 Table 8
Dispersion Parameter (D~Q) for Long Term Releases, Ground Vent 30 Table 9
Pathway Dose Factors Due to Radionuclides Other than Noble Gases 31 Figure 1
Ginna Station Liquid Waste Treatment System 15 Figure 2
Ginna Station Gaseous Waste Treatment and Ventilation Exhaust Systems 16 Figure 3
Location of Onsite Air Monitors and Post Accident TLD's 21 Figure 4
Location of Farms for Milk Samples and Ontario Water District Intake 22 Figure 5
Location of Offsite TLD's 23 Figure 6
Location of Offsite Air Monitors 24
I.
Li uid Effluent Monitor Set pints The Gin'na Technical Specifications, Section 3.5.4, require alarm and/or trip setpoints for radiation monitors on each liquid effluent line (reference 1).
Precautions, limitations and setpoints applicable to the operation of Ginna Station liquid effluent monitors are provided in plant procedures P-9 and RD-13.1.
Setpoint values are calculated =to assure that alarm and trip actions occur prior to exceeding the limits of 10 CFR 20 at the release point to the unrestricted area.
For added conservatism, liquid effluent release rates are administratively set so that only small fractions of the applicable 10 CFR 20 maximum permissible concentrations can be reached in the discharge canal.
The calculated alarm and trip action setpoints for each radioactive liquid effluent line monitor and flow determination must satisfy the following equation:
Where:
Equation (1):
cf
<C F+f C =
the effluent concentration which implements the 10 CFR 20 limit for unrestricted
- areas, in pCi/ml.
c =
the setpoint, in pCi/ml, of the radioactivity monitor measuring the radioactivity concentration in. the discharge line prior to dilution and subsequent release.
f =
the flow as measured at the radiation monitor location, in volume per unit time, in the same, units as F below.
F =
the dilution water flow as determined prior to the release point, in volume per unit time.
Liquid effluent batch releases from Ginna Station are discharged through a
liquid waste disposal monitor.
The liquid waste stream (f) is diluted (by F) in the plant discharge canal before it enters Lake Ontario.
The limiting batch release concentration (c) corresponding to the liquid waste monitor setpoint is calculated from the above expression.
Since the value of (f) is very small in comparison to (F), the expression becomes:
Where:
Equation (2):
c C Ff the maximum permissible concentration of gross beta, gamma activity above background in the circulating water discharge at the unrestricted area boundary (1 x 10 pCi/ml).
the dilution flow assuming operation of only 1 circulating water pump (170,000 gpm).
the maximum waste effluent discharge rate through the designated pathway.
The limiting release concentration (c) is then converted to a setpoint count rate by use of the monitor calibration factor determined per procedure RD-13.1.
The expression becomes:
Equation (2a):
Setpoint (cpm) =
c( Ci/ml)
Cal. Factor (pCi/ml per cpm E~xam le (Liquid Radwaste Monitor R-18):
If one assumes, for example, that the maximum pump effluent discharge rate (f) is 30 gpm, then the limiting batch release concentration (c) would be determined as follows:
( Ci/ l) <
1 x 10
( Ci/ml) '70 000 (
m) 30 (gpm c
< 5.7 x 10 (pCi/ml)
The monitor R-18 alarm and trip setpoint (in cpm) is then determined utilizing the monitor calibration factor calculated in plant procedure RD-13.1.
Assuming a calibration factor of 9.5 x 10
~(Ci/ml) cpm and a limiting batch release concentration determined above the alarm and trip setpoint for monitor R-18 would be:
5.7 x 10
( Ci/ml) 6 104 9.5 x 10
~(Ci/ml) cpm The setpoint values for the Containment Fan Cooler Monitor (R-16), Spent Fuel Pit Heat Exchanger Service Water monitor (R-20),
Steam Generator Blowdown monitor (R-19), the Retention Tank monitor (R-21),
and the All Volatile Treatment waste discharge monitor (R-22) are calculated in a similar manner using Equation (2),
substituting appropriate values of (f) and the corresponding calibration factor.
II.
Gaseous Effluent Monitor Set pints The Ginna Technical Specifications (reference
- 1) require alarm and/or trip setpoints for specified radiation monitors on each noble gas effluent line.
Precautions, limitations and setpoints applicable to the operation of Ginna Station gaseous effluent monitors are provided in plant procedures P-9 and RD-13.1.
Setpoints are conservatively established for each ventilation noble gas monitor so that dose rates in unrestricted areas corresponding to 10 CFR Part 20 limits will not be exceeded.
Setpoints shall be determined so that dose rates from releases of noble gases will comply with the Technical Specifi-cation requirements of 3.9.2.l.a(i),
which stipulate that the dose rate for noble gases shall be
< 500 mrem/yr to the total body and
< 3000 mrem/yr to the skin.
The calculated alarm and trip action setpoints for each radioactive gaseous effluent monitor must satisfy the following equation:
Equation (3):
c
< ~iv f'k'K where:
setpoint in cpm Qiv =
release rate limit by specific nuclide in pCi/sec discharge flow rate in cfm units conversion factor (cc/sec/cfm) calibration factor (pCi/cc/cpm)
The general methodology for establishing plant ventilation monitor setpoints is based upon a vent concentration limit (in pCi/cc) derived from site specific meteorology and vent release characteristics.
Additional radiation monitor alarm and/or, trip setpoints are calculated for radiation monitors measuring radioiodines, radioactive materials in particulate form and radionuclides other than noble gases.
Setpoints are determined to assure that dose rates from the release of these effluents shall comply with Technical Specification 3.9.2.1.a (ii), which requires that the dose rate for all radioiodines, radioactive materials in particulate form,'nd radionuclides other than noble gases with half-lives greater than 8 days shall be
< 1500 mrem/yr to an organ;
The release rate limit for noble gases shall be calculated by the following equation for total body dose:
Equation (4):
Q. (pCi/sec) iv 500 (mrem/ r) 3 3
K. (mrem/yr per pCi/m )
X/Q (sec/m
)
i and by the following equation for skin doses:
Equation (5):
Q. (pCi/sec) iv 3000 (mrem/ r) 3 3
(L. + 1.1M.) (mrem/yr per pCi/m )
X/Q (sec/m
)
Where:
K. = The total body dose factor due to gamma emissions for earth identified noble gas radionuclide, (in mrem/yr per pCi/m )
from Table 2.
L. = The skin dose factor due to beta emissions fog each identified 1
noble gas radionuclide,
.in mrem/yr per pCi/m from Table 2-.
M. = The air dose factor due to gamma emissions f~r each identified 1
noble gas radionuclide, in mrad/yr per pCi/m from Table 2.
(unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose).
X/Q
= The highest calculated annual average dispersion parameter for estimating the dose to the critical3offsite receptor from vent release point (v).
(in sec/m ).
The X/Q is calculated by the method described in Reg.
Guide 1.511 (reference 6).
Q.
= The release rate of radionuclide (i) from vent (v) which iv results in a dose rate of 500 mrem/yr to the whole body or 3000 mrem/yr to the skin of the critical receptor, (in pCi/sec).
Historically, xenon-133 is the principal noble gas released from all vents and is appropriate for use as the reference isotope for establishing monitor setpoints
~
The whole body dose will be limiting, and the Xe-133 release rate limit is calculated by substituting the appropriate values in equation (4).
After the release rate limit for Xe-133 is determined for each vent, the corresponding vent concentration limits are calculated based on applicable vent flow rates.
Annually-derived monitor calibration factors (pCi/cc per cpm) convert limiting vent concentrations to count rate.
Containment Plant Vent'nd Air E'ector Noble Gas Monitors (Monitors R"12 R-14 and R-15)
Monitor R-12 measures noble gas activity in containment when it is isolated, or in the containment vent during purge releases.
Noble gases being released via the plant vent are detected by R-14.
Monitor R-15, on the air ejector, normally indicates only background noble gas activity; however it serves as one of the first indicators of primary-to-secondary leakage.
Additional noble gas monitoring capability for the containment, plant and air ejector vents is provided by high-range effluent monitors R-12A, R-14A and R-15A, respectively.
Noble gas monitor setpoints are conservatively set in Procedure P-9 to correspond to fractions of the applicable 10 CFR 20 maximum permissible concentrations (MPCs) for unrestricted areas.
Fractions are small enough so as to assure the timely detection of any simultaneous discharges from multiple release points before the combined downwind site boundary concentration could exceed MPC.
Additional conservatism is provided by basing these setpoints upon instantaneous downwind concentrations.
Release rates during the remainder of a given year, combined with any infrequent releases at setpoint levels, are likely to result in only a very small fraction of the 10 CFR 20 annual limits.
E~xam le:
(Plant vent monitor R-14)
Using Xe-133 as the controlling isotope for the setpoint and assuming a measured activity at 1.63 E-6 pCi/cc, a ratemeter reading of 16 cpm above background and a vent flow of 6.45 E4 cfm.
Xe-133 efficiency = Activit Net Ratemeter Reading Ee-133 efficiency = 1.63E-6
1.02E-7 ECi/cc 16 cpm Q'"
I (Ki)(X/Qv)
Qiv =
500 (2.94E2)(2.7E-6)
= 6.3E5 pCi/sec
setpoint
c
iv Ic) K) c =
6.3E5
( Ci/sec) 6.45E4 (cfm) '72 1.0E-7 cfm
'pm c = 2.1E5 cpm (R-14 is set at 1/10 of this value, per Procedure P-9)
Continuous radioiodine. and particulate monitoring on the air ejector is not required.
Calculations and plant measurements have indicated that the iodine and particulate source terms via the air ejector compared to other airborne release
- pathways, are negligible.
(See references 5, 7, and 8).
TABLE 1
DOSE PARAHETERS FOR RADIOIODINES AND RADIOACTIVE PARTICULATE GASFOUS EFFLUFNTS" Radio-nuclide H-3 Cr-51 Hn-54 Fe-59 Co-58 I
Co-60 g Zn-65 Rb-86 Sr-89 Sr-90 Y-91 Zr-95 Nb-95 Ho-99 Ru-103 R>>-106 Ag-110m P.
Inhalation Pathway i~i 6.5E+02 3.6E+02 2.5F+04 2.4E+04 1.1E+04 3.2E+04 6.3K+04 1.9E+05 4.0E+05 4.1E+07 7.0E+04 2.2E+04 1.3E+04 2.6E402 1.6E+04 1.6E+05 3.3K+04 P.l.
food 8 Ground Pathways (m
mrem/ r er Ci/sec) 2.4F+03 1.1E+07 1.1E+09 7.0E+08 s.7r+og 4.6E+09 1.7E+10 1.6E+10 1.0E+10 9.5E+10 1.9E+09 3.5E+08 3.6F+08 3.3E+08 3.4E+10 4.4E+ll 1.5F+10 Radio-nuclide Cd-115m Sn-126 Sb-125 Te-127m Te-129m Te-132 Cs-134 Cs-136 Cs-137 Ba-140 Ce-141 Ce-144 Np-239 I-131 I-133 Unidentified P.
Inhalation Pathway (mrem/ r er Ci/m )
7 AL OE+04 1.2E+06 1.5E+04 3.8F+04 3.2F+04 1.0E+03 7.0E+05 1.3E+05 6.1r+oS 5.6E+04 2.2F+04 l.sr+os 2.5E+04 1.5E+07 3.6E+06 4.1E+07 P..l.
F~od d Ground Pathway%
(m
. mrem/ r er Ci/sec) 4.8E+07 1.1K+09 1.1E+09 7.4E+10 1.3E+09 7.2E+07 5.3E+10 5.4F+09 4.7E+10 2.4K+08 8.7E+07 6.5K+08 2.5E+06 1.1K+12 9.6E+09 9.5E+10
""The listed dose parameters are for radionuclides that may be detected in gaseous effluents.
These and additional dose parameters for isotopes not included in Table 1 may be calculated
>>sing the methodology described in NUREG-0133, Section 5.2.1 (reference 2).
TABLE 2 DOSE FACTORS FOR NOBLE GASES AND DAUGIITERS>
Radionuclide Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Kr-90 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Ar-41 Total Body Dose Factor K.r (mrem/
r.
er Ci/m 7.56E-02~
1.17E+03 1.61E+01 5.92E+03 1.47E+04 1.66E+04 1.56E+04 9.15E+01 2 '1E+02 2.94E402 3.12E+03 1.81E+03 1.42E+03 8.83E+03 8.84E+03 Skin Dose Factor L.i (mrem/ r er Ci/m )
1.46E+03 1.34E+03 9.73E+03 2.37K+03 1.01E+04 7.29E+03 4.76E+02 9.94K+02 3.06E+02 7.11E+02 1.86E+03 1.22K+04 4.13E+03 2.69E+03 Gamma Air Dose Factor H.i (mrad/ r er Ci/m )
1.93E+01 1.23E+03 1.72E+01 6.17E+03 1.52E+04 1.73E+04 1.63E+04 1.56E+02 3.27E+02 3.53E+02 3.36E+03 1.92E+03 1.51E+03 9.21E+03 9 '0E+03 Beta Air Dose Factor N.1 (mrad/ r er Ci/m )
2.88E+02 1.97E+03 1.95E+03 1.03E+04 2.93E+03 1.06E+04 7.83E+03 1.11E403 1.48E+03 1.05E+03 7.39E+02 2.46E+03 1.27E+04 4.75E+03 3.28E+03
- The listed dose factors are for radionuclides that may be detected in gaseous effluents.
These dose factors for noble gases and daughter nuclides are taken from Table B-1 of Regulatory Guide 1.109 (reference 3).
A semi-infinite cloud is assumed.
++ 7.56E-02
= 7.56.x 10 -2
Li uid Effluent Release Concentrations Liquid batch releases are controlled individually and each batch release is authorized and based upon sample analysis and the existing dilution flow in the discharge canal.
Plant procedures RD-7 and RD-8 establish the methods for sampling and analysis of each batch prior to release.
A release rate limit is calculated for each batch based upon analysis, dilution flow and all procedural conditions being met before it is authorized for release.
The waste effluent stream entering the discharge canal is continuously monitored and the release will be automatically terminated if the pre-selected monitor setpoint is exceeded.
(See Section I.)
If gross beta analysis is performed for each batch release in lieu of gamma isotopic analysis, then a weekly. composite for principal gamma emitters and I-131 is performed.
Additional monthly and quarterly composite analyses are to be performed as specified in Table 4.12-1 of the Ginna Technical Specifications.
The equations used to calculate activity are:
Gamma S ectromet eak area counts-bk d counts pCijcc Act.
(Count,Time)(Eff.)(Vol.)(T1/2 correction)(3.7E4)(Branching fraction)
Gross Beta Gamma:
C./
A t total counts - bk d counts (Count Time (Eff.)(Vol.)(T1/2 correction)(3.7E4) where:
count time is in seconds; eff. = counting efficiency, in counts er sec disintegrations per sec.
vol. = volume, in milliliters; T /
correction = decay correction factor, dimensionless; 3.7E4
= conversion constant, in disinte rations er sec; pCi Branching fraction is the fraction disintegrating by a particular decay mode, dimensionless.
IV.
Li uid Effluent Dose The dose contribution received by the maximally exposed individual from the ingestion of Lake Ontario fish and drinking water is determined using the following methodology.
These calculations will assume a near field dilution factor of 1.0 in evaluating the fish pathway dose, and a dilution factor of 20 between the plant discharge and the Ontario Water District drinking water intake located 1.1 miles away (Figure 4).
The dilution factor of 20 was derived from drift and dispersion studies documented in reference 4.
Dose contributions from shoreline recreation, boating and swimming have been shown to be negligible in the Appendix I dose analysis (reference
- 5) and do not need to be routinely evaluated.
Also, there is no known human consumption of shellfish from Lake Ontario.
The dose contribution to an individual will be determined to ensure that it complies with the Technical Specification requirements of 3.9.1.2.a (i) and 3.9.1.2.a (ii).
The dose or dose commitment to an individual from radioactive materials in liquid effluents released to unrestricted areas shall be limited:
(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.
Offsite receptor doses will be determined for the limiting age group and organ, unless census data show that actual offsite individuals are of a less limiting age group.
The following expression is used to calculate ingestion pathway dose contributions m
for the total release period g $ l 6tg from all radionuclides identified in liquid effluents released to unrestricted areas:
Equation (6):
D
= X [A.
Z At< C.
F<]
l.
R-1 Where:
D the cumulative dose commitment to the total body or any
- organ, X, from the liquid effluents for the total time period (in mrem).
m Z
4t~,
L = l bt
= the length of the Rth time period over which C.< and F< are averaged for all liquid releases, (in hours).
iR C.~ =
the average concentration of radionuclide, i, in undiluted liquid effluent during time period At< from any liquid
- release, (in pCi/ml).
A.iX the site-related ingestion dose commitment factor to the total body or any organ x for each identified principal gamma and beta emitter (in mrem/hr per pCi/ml).
See equation (7).
10-
F< =
the discharge canal dilution factor for C.< during any liquid effluent release.
Defined as the ratio of the maximum undiluted liquid waste flow during release to the average flow from the site discharge structure to unrestricted receiving waters.
The dilution factor will depend on the number of circulation pumps operating and, during icing conditions, the percentage opening of the recirculating gate.
Reference curves are presented in plant procedure RD-7.
Equation (7):
A.
= k (U /D
+ U BF.) DF.
Where:
A.i' k0 the site-related ingestion dose commitment factor to the total body or to any organ t for each identified principal gamma and beta emitter, (in mrem/hr per pCi/ml).
units conversion factor, 1
~ 14 x 10
= 10 pCi/pCi x 10 ml/kg 5 6...
3 8760 hr/yr.
U
=
a receptor person's water consumption by age group from w
table E-5 of Regulatory Guide 1.109 (reference 3).
D
= Dilution factor from the near field area of the release w
point to potable water intake.
The site specific dilution factor is 20.
This factor is assumed to be 1.0 for the fish ingestion pathway.
UF =
a receptor person's fish consumption by age group from table E-5 of Regulatory Guide 1.109.
BF. =
Bioaccumulation factor for nuclide, i, (in fish pCi/kg per PCi/2),
from Table A-1 of Regulatory Guide 1.109.
DF. =
Dose conversion factor for the ingestion of nuclide, i, for a receptor person in pre-selected organ, t, (in mrem/pCi),
from Tables E-ll, E-12, E-13, E-14 of Regulatory Guide 1.109:
The monthly dose contribution from releases for which radionuclide concentrations are determined by periodic composite sample analysis may be approximated by assuming an average monthly concentration based on the previous monthly or quarterly composite analyses.
However, in the radioactive effluent release report (submitted within 60 days of January 1 per Technical Specification 6.9.1.4) the calculated dose contributions from these radionuclides shall be based on the actual composite analyses.
Exam le which illustrates how to com ute the dose to the whole bod via the fish and drinkin water athwa s
assumin an initial Cs-137 dischar e
concentration of 3.0 E-4 Ci/ml:
Given the following discharge factors, where:
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Dw 3.0 E-4 pCi/ml 170,000 gpm 20 And, taking the following values from Regulatory Guide 1.109 which concern the receptor of interest, which we assume is the child in this case:
U
=
510 R/year w
U
=
6.9 Kg/year BF.
=
2000 pCi/kg per pCi/2 DF.
=
4,62 E-5 mrem/pCi 1
- Then, the site-related ingestion dose commitment factor, A.
, is calculated as follows:
= 1.14 E5
(
+ 6.9
~ 2000) 4.62 E-5 510 20 A.
=
7.28 E4 mrem/hr per pCi/mR i'C And', the whole body dose to the child is then:
D mrem = (A. )(btR)(C.R)(FR)
= (7.28 E4)(1)(3.0 E-4)(1.2 E-4)
D
= 2.6E-3 mrem to the whole body from Cs-137 (The dose contribution from any other isotopes would then need to be calculated and summed).
12-
V.
Li uid and Gaseous Radwaste Treatment and 0 erabilit An objective of the Ginna Technical Specifications which implement the overall requirements of 10 CFR Part 50, Appendix I, is to ensure that the plant radwaste treatment equipment is used and maintained.
This equipment is to be utilized to reduce radioactive discharges from nuclear plants to levels "as low as reasonably achievable" or ALARA.
ALARA levels warranting equipment operability have been defined by the NRC in the form of monthly dose "trigger" values.
The trigger values, which are provided below, correspond to approximately 1/48 of the annual design objective doses given by 10 CFR Part 50, Appendix I. If continued at this rate, these monthly doses would correspond to just under 1/4 of the Appendix I annual design objectives.
Li uid Radwaste S stem Gaseous Radwaste S stem Ventilation Exhaust 31-day 0.06 mrem (w. body)
Trigger Values 0.2 mrem (any organ) 0.2 mrad (gamma air) 0.4 mrad (beta air) 0.3 mrem (any organ)
Figures 1 and 2 show the Ginna liquid and gaseous waste/ventilation exhaust systems.
These systems are normally in routine use at the plant.
Because discharges are being treated, the trigger values in the Technical Specifications may be exceeded but compliance with the stated quarterly and annual dose limits is required.
If the liquid or gaseous radwaste/ventilation exhaust systems were to be inoperable in excess of 31 days, then effluents are considered "untreated" waste.
- Should, over a 31-day period, the plant discharges exceed the dose
- system, then Technical Specifications 3.9.1.3.b and 3.9.2.3.c apply.
In this
- case, a 30-day report must be submitted to the Commission which identifies the inoperable equipment and describes appropriate corrective actions (see Technical Specifications 3.9.1.3.b and 3.9.2.3.c).
E~xam le:
Assume a case where plant modifications were underway on the waste evaporator and demineralizer piping, and a reduction of the liquid waste volume contained in the waste holdup tank (WHUT) was needed.
Assuming no other means of treatment were readily available, WHUT liquid (untreated) might be transferred to the waste condensate
- tanks, sampled and then discharged on a controlled basis.
In this case, assume also that a decision was made to proceed this way at the beginning of a given month, knowing that the waste evaporator and other portions of the liquid radwaste treatment system would be unavailable for the next 45 days.
The followin method would be used to determine the need for a 30-da re ort:
1.
Using existing plant procedures, sample the concentration contained in the WHUT (C.<).
Decide a sample frequency (e.g.
1-day) since the tank concentration could change.
13-
2.
Determine the permissible release rate to maintain the concentration in the discharge canal well within the applicable maximum permissible concentration (e.g.
1/100-1/10 MPC) for the mixture.
(The discharge canal concentration is equal to C.
~ F ).
xi 3.
Calculate the incremental dose from all identified isotopes via the drinking water.and fish ingestion pathways for each receptor group.
The critical receptor will probably be the child.
Assume the release will be continuous and that doses will be evaluated"each day, corresponding to the WHUT sampling frequency.
(We thus compute D
using Equations 6 and 7, taking ht as the duration of each release; in this case, 24 hr/day).
4, The offsite receptor dose due to a controlled discharge of the WHUT contents is thus determined and cumulated over each daily release time interval. If the WHUT isotopic mixture and the discharge canal dilution factor, F>, are relatively constant, then each day's dose increment should be approximately the same.
One can then estimate the number of release days it will take to reach the applicable dose trigger value.
5.
The 30-day reporting requirement applies if a radwaste treatment system is inoperable and dose trigger values are exceeded.
In the example, one reporting criterion is already met, since the treatment systems will be out of service for more than 31 days.
If we determine that the liquid pathway dose does not exceed the trigger values in 31 days or less, then a 30-day report is not required.
However, if a liquid pathway dose attains a trigger level within 31 days, then a report submittal would be required.
6.
In the last case, it would be prudent to avoid a situation requiring the 30-day report.
- First, a trigger level dose, when added to the calculated doses resulting from all other liquid release sources (e.g.
high conductivity waste tank, blowdown, retention tank),
may significantly impact upon the plant's "dose budget" for the calendar quarter or the calendar year.
Also, more realistically, other treatment options would likely be available at the plant and could be utilized.
GINNA STATION LIQUID WASTE TJ<EATHEN'I'YSTEM FIGIJRE 1
SPENT RESIN STORAGE TANKS REACTOR COOLANT DRAIN TANK CONTAINMENT SUMP "A" CJIEHICAL DRAIN TANK LAUNDRY & JIOT SIIOWER TANKS S/G BLOWDOWN TANK DRAIN
- 0-AUX.
& INTERMEDIATE BLDG. DRAINS BLOWDOWN SAMPLE LINE
'ONITOR R-19 RECYCLE S/G BJ,OWDOWN LINJ.,
TO IIOTWELL MIXED BED DI WASTE JIOLDUP TANK WASTE L'VAPOIUVCOI IIIGII CONDUCTIVITY WASTE TANK MONITOR R-22 gr CIRCGLATING WATER, DISCIIARGE
/
'I l
I I
I I
I MIXED BED DI WASTE CONDENSATE TANK TURBINE BLDG DRAINS MONITOR R-21 Sl P IIX MONITOR R-20 MONITOR R-18 WASTE CONDENSATE TANK
~
SERVICL'ATJ R
HONITOR R-16 CV l'N
. COnr,l;RL
- OR 'I'0 CIRC.
Wh'I'I'..R I)JSCIIARGE
GINNA STATION GASEOUS HASTE TREATMENT AND VENTINGATION EXUAUS'I' YSTEHS rIGURE 2
AUXILIARY DUIL'DING VENTILATION SYSTEH "G"PILTERS "C"FILTERS MONITORS R-108,13,14,lych PLANT VEN'i "A"FILTERS GASEOUS WASTE TREATMENT. SYSTEM CVCS HASTE GAS COMPRESSORS GAS DECAY TANl(S I
II'O PLANT VENT HONITORS CONTAINMENT PURGE'ONTAINMENT MONITORS R-10A, 11, l2, 12A CONTAINMENT V CONDENSER AIR EJECTOR OPFCAS VE~
NOTE: A=IIEPA F'ILTERS C=ClthRCOAL PI LTERS
~
~
r:=rhNS
VI.
Gaseous Effluent Dose Rate Gaseous effluent monitor setpoints as described in Section II of this manual are established at concentrations which permit some margin for corrective action to be taken before exceeding offsite dose rates corresponding to 10 CFR Part 20 limitations.
Plant procedures RD-1.1, RD-1.2, RD-1.3, RD-2, RD-3, RD-5 and RD-12 establish the methods for sampling and analysis for continuous ventilation releases and for containment purge releases.
Plant procedure RD-6 establishes the methods for sampling and analysis prior to gas decay tank releases.
The instantaneous dose rate in unrestricted areas due to unplanned releases of airborne radioactive materials may be averaged over a 24-hour period according to Technical Specification 3.9.2.1.b.
Dose rate shall be determined using the following expressions:
For noble gases:
Equation (8):
D = Z [K.
(X/Q)
Q. ]
< 500 mrem/yz (to total body) 3.
Equation (9):
D = Z [(L. + 1.1 M.) (X/Q)
Q. ]
< 3000 mrem/yr (total gamma 8
l.
beta dose to the skin)
For radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases.
Equation (10):
D = Z [P.
W Q. ]
< 1500 mrem/yr (critical organ)
K.
The total body dose factor due to gamma emissjons for each identified noble gas radionuclide, (in mrem/yr per pCi/m from Table 2).
L.
The skin dose factor due to beta emissions fog each identified noble gas radionuclide, (in mrem/yr per pCi/m from Table 2).
M.
The air dose factor due to gamma emissions fog each identified noble gas radionuclide, (in mrad/yr per pCi/m from Table 2),
(unit conversion constant of 1,1 mrem/mrad converts air dose to skin dose).
P.
The dose parameter for radionuclides other t)an noble gases for the inhalation/pathway, in modem/yr per pCi/m and for food and ground plane pathways, (in m mrem/yr per,pCi/sec) from Table l.
The dose factors are based on the critical individual organ and most restrictive age group.
(X/Q)
The highest calculated annual average relative concentration fog any area at or beyond the unrestricted area boundary, (in sec/m ).
W Q v The highest annual average dispersion parame)er for estimating the dose to the critical receptor; (in sec/m for the inhalation pathway, and in m for the food and ground pathways).
the release rate of radionuclide i from vent (v), (in pCi/sec).
VII. Gaseous Effluent Doses The air dose in unrestricted areas due to noble gases.released in gaseous effluents from the site shall be determined using the following expressions:
During any calendar year, for gamma radiation:
Equation (ll):
D
= 3.17 x 10 X[M. ~X/Q Q. ] < 10 mrad, and
-8 3.
During any calendar year'for beta radiation:
1 Equation (12):
D
= 3.17 x 10 Z [N. ~X/Q Q. ]
< 20 mrad
-8 i
Where:
M. = The air dose factor due to gamma emissions fog each identified i
noble gas radionuclide, (in mrad/yr per pCi/m from Table 2).
N. = The air dose factor due to beta emissions for each identified 3.
noble gas radionuclide, (in mrad/yr per pCi/m from Table 2).
3
~X/Q
= For vent releases.
The highest calculated annual
- average, V
relative concentration for any area at or beyond the unrestricted area boundary, including uninhabited areas, (in sec/m ).
Y Dp
= The total gamma air dose from gaseous effluents, (in mrad).
= The total beta air dose from gaseous effluents, (in mrad).
Qv 3.17 x
= The release of noble gas radionuclides, i, in gaseous effluents from all vents, in pCi.
Releases shall be cumulative over the time period.
10
= The inverse of the number of seconds in a year.
-8 18-
The dose to an individual from radioiodines and radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents released from the site to unrestricted areas shall be determined using the following expression:
During any calendar year:
Equation (13):
D
= 3.17 x 10 Z R.
[W Q. ],
15 mrad
-8 i
Where:
The release of radioiodines, and radioactive materials in particulate form in gaseous effluents, i, with half-lives greater than 8 days, (in pCi).
Releases shall be cumulative over the desired time period as appropriate.
D> = The total dose from radioiodines and radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents, (in mrem).
W The annual average dispersion parameter for estimating the dgse to an individual at the critical logation; (in sec/m for the inhalation pathway, end in m for the food and ground pathways).
R. = The d~se factor for each identified radionuclide,3i, i
(in m
~ mrem/yr per pCi/sec or mrem/yr per pCi/m from Table 9).
VIII.
Environmental Monitor Sam le Locations Figure 3 shows the onsite* indicator sample locations for airborne particulates, radioiodine and direct radiation.
Respective sample locations are specified below.
Also indicated on Figure 3 is the onsite vegetable
- garden, as well as the placement of post-accident TLD's (locations 13-24).
The onsite garden is located in the sector having the highest D/Q value.
Figure 4 gives. the location of the only milk herds within 5 miles of the plant.
On this map is also included the Ontario Water District intake pumping station where lake water is sampled prior to treatment.
Figure 5 shows the offsite control sample locations for airborne particulates, radioiodine and direct radiation.
Sample stations 9 and 11 are situated near population centers (Webster and Williamson) located approximately 7 miles from the Ginna site.
Ke to Fi ures 3 to 5:
Location Radioiodine:
2 onsite 2 offsite
//4 and 7
/j9 and ll Particulate:
7 onsite 5 offsite
/f2, 3, 4, 5, 6, 7 and 13
//8, 9, 10, 11 and 12 Direct Radiation:
TLD 18 onsite 10 4-5 miles 11
>5 miles
//2) 3, 4, 5, 6, 7) 13, 14 15) 16, 17, 18,
- 19) 20, 21, 22, 23 and 24 jf31 32 33 ) 34) 35 ) 36) 37 38, 39 and 40
/N) 9) 10) 11)
- 12) 25) 26) 27, 28, 29 and 30 Surface Water 1 control (Russell Station) 1 indicator (Ginna Condenser Water Discharge)
Drinking Water 1 indicator (Ontario Water District Intake)
Milk 1 control (/j4) 3 indicator (Ill 2, 3)
Fish 4 control 4 indicator (offshore at Ginna)
Food Products 1 control 2 indicator (onsite)
Note:
"Onsite" refers to the area surrounding the Ginna plant bounded by RG&E property lines.
"Offsite" refers to the area beyond the immediate RG&E property.
FIGURE 3
I.AKE ONTARIO if ee l
\\ a 0>>
'8 4IW'4 Vt/ ~ 4 IIt 3
(.=.r Onside Garden
/
-- 2g )(
/6
~ Il
/8 09 20
+e(
r g't ~ ~~ y
~
2/'i<)'/
f-)- r
'i
(
ONSITE AIR llONITOR 0
200 SCALE METERS 400 600 r)-
.~(
~)
~)(
F~g).,f 3
~.C;3;-~
1l
Uater
~~
le Station Mill:Sample Station LAKE 0NTA Rlo js HiL'F=s--....
nd Vew Heighle
nl Be hach 2
i 5MtLE r
.J
~4<
W LLI SON Q~
NQ 0
U
~
(p~
"t 5
P NF LWO 0
ghto
,fe TP (w
nfield e
e
I I
~ ~
g i
)44')4>>rC S Sf A)I I'Al<C I.h KI Oll'I'hit10 FIGURE 6
Ol I S I I'I All( IIOHIIOR I
O1IIIIA SI'I'E lll>il>>4< sess>rs I I II II N I) I: II II0 I I I
r j) l (II Ij
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4 k~
Ii~ )
ssr
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'I~
j C(W
)") ) ygL).
UG)
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~
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I>
I sl.s li) I I
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la.>>a s,
p
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TABLE 4 DXSPERSZON PARA>lETER (D O)
FOR LO!<G TERM RELEASES 5DO llR YR OR
> 125 RR OTR Plan t.Vent Distance to the control location, in miles Sector 0-0.5
- 0. 5-1. 0 l. O.-l. 5 l, 5.-2. 0.
- 2. 0-?.. 5
- 2. 5-3. 0 3.0-3.5 3.5-4.0 4.0-4.5 '4.5-5.0 EPEE
- 8. 3,E-S 4.5 E-S 6.5 E-S 8.3 E-8 1.4 E-7 1.4 E-7 1.3 E-7 5.8 E-S 2.8 E-8 3.1 E-S 4.5 E-8 5.6 E-S 4.2 E-8 2.2 E-8 1.5 E-8 4.0 E-8 1.7 H-S 6.1 E-9 2.5 H-.9 1.2 E-.9
- 1. 0 E-S
- 3. 7 E-9 l. 5 8-9
- 7. 0 E-.10 1.5 E-S 5.4 H-9 2.2 H-9 1.0 E-9 7,3 E=lo 4.4 E-10 6,5:E-10 1.8 E-8 6.4 E-9 2.6 E-9 1.2 E-. 9 7.5 E-10 2.9 E-S 1.0 E-8 4.2 E-9 1.9 H-9 1.2 E-9 3.0 E-8 1.1 E-8 4.3 E-9 1.9 E-9 1.2 E-9 2.7 E-8 9.3 E-9 3.7 H-9 1.7 E-9 1.0 E-9 1.4 E-S 4.7 E-9 1.9 E-9 8.9 E-10 5.6 E-10 8.6 E-9 3.1 E-9 1.3 E-9 5.8 E-10 3;8 E-10 7.8 E-9 3.1 E-9 1.3 E-9 5.9 E-10 3.7 E-10 1.0 E-8 3.6 E-9 1.5 E-9 6.8 E-10 4.4 E-10 1.3 E-8 4.6 E-9 1.8 E-9 8.4 E-10 5.-3 E-10 1.0 E-8 3.9 E-9 1.6 E-9 7.4 E-10 4.7 E-10 5.9 E-9 2.4 E-9 1.0 E-9 4.7 E-10 3.0 E-10 4.1 E-9 1.7 H-9 7.0 E-10 3.3 E-10 2.1 E-10 9.2 E-9 3.5 E-9 1.4 H-9 6.6 E 10 4.2 E-10 5.1 E=10 3,1 E 10 4.5 H 10 5.3 E-10 8.6 E-10 8.7.H-10 7.7 E-10 4.1 E-'10
- 2. 9 E-10 2.7 E-10 3.1 E-10 3.7 E-10 3.3 E-10 2.1 E-10 1.5 E-10 2.9 E-10 2.9 E-10 2.1 E-10 2.6 E-10 1.9 E-10 l. 8 E-1(
1.6 1(
1.7 E-10 1.3 E-10 1.0 E-1(
1.2 E-10 8.8 E-ll 7 4
E-1'.3 E-10 1.7 E-10 1.4 E-1(
4.1 E=10 2.9 E-10 2.5 E-10 2.4 E-. 10 1.8 E-10 1.5 10 3.6 E-10 2.6 E-10 2.2 E-10 4.1 E-10 3.1 E-10 2.6 E-10 6.7 E-10 4.8 E-10 4.1 E-10 6.7 E-10 5.2 E-10 4.5 E-lO 6.1 E-10 4.6 E-10 4.0 E-1('.5 E-10 2.7 E-10 2.3 E-1C 2.4 E-10 1.8 E-10 1.6 E-1(
2.2 H-10 1.8 E-10 1.5 E-1(
2.5 E-10 1.9 E-10 1.6 E-1(
- Direction wind blows into A
A
TABLE 5 C.
DISPERSION PARAMETER (X/Q)
FOR LONG TERM RELEASES
> 500 IJR/YR OR > 125
}JR/QTR c
Containment:
Pur e
Distance to th control location, in miles Sector 0-0. 5 0.5-1.0 1.0-1,5 l. 5.-2..0 2.. 0-2...5 2.5-3.0 3.0-3.5 3.5-4.0
- 4. 0-4. 5
- 4. 5'-5. (
N NNE hlE
- -S=
7 E-6 3.1 E-6 4.1 E-6 3.9 E-6 4.9 E-6 4.3 E-6 4.2 E-6 2.3 E-6 1.3'-6
- l. 2 E-6 1.3 E-6 1.7 E-6 1.7 E-6 1.2 E-6 8.5 E-7 1.8 E-6
- l. 2 E-6
- 7. 2 E-7
- 3. 6 E-7
- 1. 0 E-6
- 6. 6 E-7 1.4 E-6 9.0 E-7 3.5 E-7 4.7 E-7 1.3 E-6 7.7 E-7 3.9 E-7 1.6 E-6 8.8 E-7 4.1 E-7 1.5 E-6 1.2 E-6 9.1 E-7 3.9 E-7
- 6. 1 E ~ 2. 8 9.7 E-7 4.6 E-7 2.2 7.7 E-7 4.1 E-7 1.9 E-7 4.5 E-7 3.3 E-7 1.7 E-7 4.1 E-7 2.7 E-7 1.3 E-7 5.3 E-7 3.2 E-7 1.5 E-7 7.2 E-7 4.4 E-7 2.1 E-7 6.0 E-7 3.9 E-7 2.0 E-7 4.4 E-7 3.0 E-7 1.6 E-7" 7.0 E-7 4.4 E-7 2.2 E-7 2.0 E-7 2.0 E-7 2.7 E-7 2.1 E-7 2.2 E-7 2.0 E-7 1.4 E-7 1.2 E-7 1.0 E-7 9.5 E-S 7.3 E-8 8.6 E-S 1.2 E-7 1.1 E-7 8.9 E-8 1.2 E-7 1.4 E-7 1.1 E-'7 9.6 E-S 1.5 E-7 1.2 E-7 1.0 E-7 2.0 '-7 1.6 E-7 1.3 E-7 1.5 E-7 1.2 E-7 1.0 E-7 1.5 E-7 1.2 E-7 1.0 E-7 1.4 E-7 1.3, E-7 8.6 E-8 9.9 E-8 8.0 E-S 6.5 E-8 8.1 E-8 6.1 E-8 5.0 E-8 7.4 E-S 5.8 E-8 4.7 E-8 6.7 E-8 5.3 E-8 4.5 E-8 5:2 E-8 4.1 E-S 3.4 E-8 6.0 E-8 4.5 E-S 3.8 E-8 8.6 E-8 6.6 E-8 5.5 E-8 8.2 E-S 6.3 E-8 5.3 E-8 6.5 E-S 5.1 E-8 4.3 E-8 9.0 E-8 7.1 E-8 6.-0 E-8
- 8. 1 E-8
- 8. 9 E-8 1.1 E-7 8.5 E-8 8.3 E-S 7.4 E-8 5.4 E-8 4.0 E-8 3.8 E-8 3.7 E-8
2.7 E-8 3.2 E-8 4
6 E-8 4.5 E-8 3.5 E-8 5.0 E-8 7.1 E-8
- 7. 98
- l. 0 E-7 7.5 E-8
- 7. 3 E-8 6.4 E-8 4.6 E-8 3.4 E-8
- 3. 2 E-8
- 3. 2 E-8
- 2. 3 E-8 2.8 E-8
-8
- 3. 9 E-8
- 3. 2 E-8 4.4 E-8
- Direction wind blows into
~
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i
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TABLE 7 0
C0 DISPERSION PARAMETER (X/Q) FOR LONG TERM RELEASES
> 500 HR/YR OR
> 125 HR/QTR Ground Uent Distance to the control location, in miles Sector*
0-0.5 0.5-1.0
- 1. 0.-1. 5 1, 5;2..0 2.. 0-2. 5.
- 2. 5-3..0 3..0-3.S 3.5-4.0 4.0-4.5
- 4. 5-5. 0 NE ENE ESE SSE S
SS'H S!0 4.4 E-5 5.5 E-5 6.5 E-5
- 4. 4,E-5 3.7 E-5 2.6 E-5 1.7 E-5 1.3 E-5 1.2 E-5 1.2 E-5 9.7 E-6 1.4 E-5 2.5 E-5 2.4 E-5 2.1 E-5 2.9 E-5 8.2 E-6 1.0 E-5 1.2 E-5 8.3 E-6 7.1 E-6 4.8 E-6 3.1 E-6 2.4 E-6 2.2 E-6 2'.1 E-6 1.7 E-6 2.4 E-6 4.5 E-6 4.6 E-6 4.0 E-6 5.4 E-6
- 3. 4 E-6 l. 4 E-6
- 6. 9 E~7 4.2 E-6 1.8 E-6 8.7 E-7 5.1 E-6 2.1 E-6 1.0 E-6 3.5 E-6 1.4 E-6 6.9 E-7 2.9 E-6 1.2 E-6 5.7 E-7 2.0 E-6 7.8 E-7 3.8 E-7 1.3 E-6 5.0 E-j 2;4 E-7 9.5 E-7 3.7 E-7 1.8 E-7 9.0 E-7 3.5 E-7 1.7 E-7 8.7 E-7 3.5 E-7 1.7 E-7 6.8 E-7 2.7 E-7 1.3 E-7 9.9 E-7 4.0 E-7 1.9 E-7 1.8 E-6 7.5 E-7 3.6 E-7 1.9 E-6 7.7 E-7 3.7 E-7 1.6 E-6 6.7 E-7 3.3 E-7 2.2 E-6 9.2 E-7 4.5 E~7 4.7 E-.7 5.9 E-7 6
9 E-7 4.8 E-7 3.7 E-7 2.5 E-7 1.6 E-7 1.2 E-7 1.1 E-7 1.1 E-7 8.7 E-8 1.3 E-7 2.4 E-7 2.5 E-7 2.2 E-7 3.0 E-7 3.4 E-7 2.8 E-7 1.8 E-7 2.8 E-7 2.2 E-7 1.5 E-7 2.2 E-7 1.8 E-7 1.1 E-7 1.1 E-7 9.3 E-8 7.6 E-8 8.6 E-8 7
0 E-8 5.7 E-8 8.4 E-8 8.3 E-8 6.7 E-8 6.6 E-8 5.4 E-8 5.4 E-8 6.3 E-8 S.l E-8 4.1 E-8
- 9. 3 E-8 1.8 E-7 7.6 E-8 1&E-7
- 6. 3 E-8 1.1 E-7 1.8 E-7 1.5 E-7 1.2 E-7 1.6 E-7 1.3 E-7 1.1 E-7 2.2 E-7 1.8 E-7 1,.5 E-7 3-4 g-.7 2.7 g-.7 2.2 E-7 4.3 E-7
- 3. 5 E-7
- 2. 9.E-7 5.1 E-7 4.1 E-7 3.4 E-7 1.9 E-7 2.d 2.8 E-7 1.9 E-7 1.5 E-7 9.9 E-8 6.3 E-8 4.6 E-8
- 4. 5 E-8
- 4. 5 E-8 3.4 E-8 5.2 E-8 9.8 9.7 E-8 8.8 E-8 1.2 E-7
- Direction wind blows into
TABLE 8 DISPERSION PARAMETER (D Q)
FOR LONG TERM RELEASES
> 500 HR YR OR > 125 HR/QTR Ground Vent Distance to the control location, in miles Sector 0-0.5
- 0. 5-1. 0 l.. 0.-1, 5 1, 5.-2..0
- 2. 0-2. 5.
- 2. 5-3. 0 3.0-3.5 3.5-4.0 4.0-4.5 4.5-5.(
N SSE SSH SH hSH 7.6 E-8 9.9 E-8 1.4 E-S
/l. 8 E-8 4.7 E-9 6.1 E-9 1.1 E-7 2.0 E-8.
6.7 E-9 8.9 H-S 1.6 E-8 5.4 E-9 7.0 E-S 1.3 E-S 4.3 E-9 1.2 H-.7 1.2 E-8 7.1 E-9 2.0 E-7 3.7 E-8 1.2 E-S 1.8 E-7 3.4 E-S 1.1 E-8 2.5 H-7 4.5 E-S 1.5 E-8 2.1 E-7 3.9 E-8 1.3 E-8 2.5 E-7 4.6 E-8 1.5 E-8 2.2 E-7 4.1 E-8 1.3 E-S 1.8 E-7 3.7 E-8 1.1 E-S 9.8 E-8 1.8 E-8 6.0 E-9 6.8 E-8 1.3 E-S 4.2 E-9 6.7 E-8 1.2 E-8 4.1 E-9 5.0 E-9 2.3 E-9 1.4 E-9 4.5 E-9 2.1 E-9 6.1 E-9 2.8 E-9 1.3 H-9 1.7;H-9 1.7 E-9 1.9 E-9 7.9 E-10 4.9 E-10 1.3 H-9'.1 E-10 5.3 E-9 2.4 E-9 1.5 E-9 6.2 E-9 2.-8 E-9 1.7 E-9 5.5 E-9 2.5 E-9 1;6 E-9 4.5 E-9 2.1 E-9 1.3 E-9 2.4 E-9 1.1 E-9 6.8 H-10 1.7 E-9 7.7 E-10 4.8 E-10 1.7 E-9 7.6 E-10 4.7 E-10 1.9 E-9 8.6 E-10 5.5 E-10 1.5 H-9 1.1 H-9 6.9 E-10 2.7 E-9 1.2 H-9 7.5 E-10 2.2 E-9 1.0 E-9 6.3 E-10 9.7 E-10 9.0 H-10 1.1 E-9 1.0 E-9 1.2 E-9.
1.1. E-9 9.0 E-10 4.8 E-10 3.3 E-10 3.3 E-lo 3.8 E-10 4.9 E-10 5.4 E-10 4.3 H-10 3.4 E-10 5.7 E-10 7.6 E-10 5.5 E-10 6.9 E-10 5.0 E-10 9.2 E-10 6.9 E-10 8.0 E-10 6,0 E-10
- 1. 7 E-1<
4.3
- 5. 8 E-,l<
5.0 E-li 3, 7 E-10 2. 8 E-10
- 4. 1 E-10 3. 0 E-10 3.3 E-10 2.5 E-10 2.6 E-10 2.0 E-10 4.4 E-10 3.2 E-10 2;3 E-1
- z. z.
2.1 E-l 1-6 E-1 2.7 E-1 9.4 E-10 7.0 E-10 5.8 E-li 8.4 E-10 6.3 E-10 5.2 E-li 6.9 E-10 5.1 E-10 4.3 E-l~
3.7 E-10 2.7 E-10 2.3 E-1<
2.6 E-10 1.9 E-10. 1.6 E-1 2.5 E-10 1.8 E-10 1.5 E-3.
2.9 E-10 2.1 E-10 1.7 E-1
- Direction wind blaws into
% ~
C TABLE 9 PATHWAY DOSE FACTORS DUE TO RADIONUCLIDES OTHER THAN NOBLE GASES" Radio-nuclide
]1-3 CR-51 MN-54 FE-59 CO-58 CO-60 ZN-65 SR-89 SR-90 ZR-95 I-131 I-133.
CS-134 CS-136 CS-]37 BA-140 CE-141 Inhalation Pathway l.
(mrem/yr3 er Ci/m )
1.12E 03 1.70E 04 1.57E 06
].27E 06 1.]OE 06 7.06E 06 9.94E 05 2.15E 06 1.0]E 08 2.23E 06 1.62E 07 3.84E 06 1.01E 06 1.7]E 05 9.05E 05 1.74E 06 5.43E 05 Meat Pathway i
(m 'mrem/yr er Ci/sec) 2.33E 02 4.98E 05 7.60E 06 6.49E 08 9.49E 07 3.61E 08 1.05E 09 4.89E 08 1.01E 10 6.09E 08 2.60E 09
,6.45E 01 1.42E 09 5.06E 07 1.27E 09 5.00E 07 1.45E 07 Ground Plane Pat)way 3.
(m mrem/yr er ICi/sec) 0.
5.31E 06 1.56E 09 3.09E 08 4.27E 08 2.44E 10 8 '8E 08 2.42E 04 0.
2.73E 08 1.01E 07 1.43E 06 7.70E 09 Z 64E 08 1.15E 10 2.26E 07 1.48E 07 Cow-Milk-Infant Pathway (m
mrem/yr'I'I 2.38E 03 5.75E 06 3.70E 07 4.0]E 08 7.01E 07 2.25E 08 1.99E 10 1.28E 10 1.19E 10 8.76E 05 4.95E ll 4.62E 09 6.37E 10 6.61E 09 5.75E 10
'.75E 08 1.43E 07 Leafy Vegetables Pat)way l.
(m.mrem/yr)
'I'.47E 02 1.63E 06 5.38E 07 1.10E 08 4.55E 07 1.54E 08 2.24E 08 5.39E 09 9.85E 10 1.13E 08 2.08E 10 3.88E 08 1.96E 09 1.60E 08 1.80E 09
'.03E 08 8.99E 07
>Additional dose factors for isotopes not included in Table 9 may be calculated using the methodology described in NUREG-O]33, Section 5.3.1 (reference 2).
IX.
Pre aration of S ecial Re ort to Demonstrate Com liance with Environmental Radiation Protection Standards Ginna Technical Specification 3.9.2.4.a requires the preparation and submittal of a Special Report to the Commission, when calculated 'effluent release doses exceed twice the limits of Specifications 3.9.1.2.a, 3.9.2.2.a or 3.9.2.2.b.
In addition, subsequent releases are to be limited so that the dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to
< 25 mrem to the total body or any organ (except the thyroid, which is limited to
< 75 mrem) for the calendar year that includes the release(s) in the Special Report.
This includes the dose contributions from the calendar quarter in which the limits were exceeded and the subsequent calendar quarters within the current calendar year.
The following general guidelines are presented for preparation of the Special Report:
1)
The maximally exposed real member of the public will generally be the same individual considered in the Technical Specification; 2)
Dose contributions to the maximally exposed individual need only be considered to be those resulting from the Ginna plant itself.
All other uranium fuel cycle facilities or operations are of sufficient distance to contribute a negligible portion of the individual's dose.
3)
For determining the total dose to the maximally exposed individual from the maj or gaseous and liquid effluent pathways and from direct radiation, dose evaluation techniques used in preparing the Special Report may be those described in this manual or other applicable methods where appropriate.
4)
The contribution from direct radiation may be estimated by effluent dispersion modelling or calculated from the results of the environmental monitoring program for direct radiation.
References 1.
R.E.
Ginna Nuclear Power Plant Unit No.
1, A
endix A to Provisional 0 eratin License No.
DPR-18 Technical S ecifications, Rochester Gas and Electric Corporation, Docket No. 50-244.
2.
- USNRC, Pre aration of Radiolo ical Effluent Technical S ecifications for Nuclear Power Plants, NUREG-0133
- October, 1978 3.
USNRC, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Pur ose of Evaluatin Com liance with 10 CFR Part 50 A
endix I, Regulatory Guide 1.109, Revision 1 (October, 1977).
4.
R. E.
Ginna Nuclear Power Plant, Environmental Re ort, Appendix B
(August, 1972).
5.
R.E.
Ginna Nuclear Power Plant, Calculations to Demonstrate Com liance with the Desi n Ob'ectives of 10 CFR Part 50 A
endix I.
Rochester Gas and Electric Corporation,
- June, 1976 6.
USNRC, Methods for Estimatin Atmos heric Trans ort and Dis ersion of Gaseous Effluents in Routine Releases from Li ht-Water-Cooled Reactors, Regulatory Guide 1.111, Revision 1
- July, 1977).
7.
R. E. Ginna Nuclear Power Plant, Incident Evaluation Ginna Steam Generator Tube Failure Incident Janua 25 1982, Rochester Gas and Electric Corporation, April 12, 1982).
8.
Pelletier, C. A. et. al.,
Sources of Radioiodine at Pressurized Water
- Reactors, EPRI NP-939 -(November, 1978).
"33-
0 c
r
~/
g
~
lit'l/I/~y/0</~'~Ji'i7i ROCHESTER GAS AND ELECTRIC CORPORATION
~ 89 EAST AVENUE, ROCHESTER, N.K 14649.0001 TCLKPHONC ARcA cooE 1ld 546.2700 February 29, 1988 Mr. William T. Russell Regional Adminstrator U.S. Nuclear Regulatory'Commission Region 1
631 Park Avenue King of Prussia, Penn 19406
Dear Mr. Russell:
This Semi-annual Radioactive Effluent Release Report is being forwarded to you in accordance with the requirements of Tech Spec.
Section 6.9.1.4.
We have changed the overall format of this report from our earlier submittals;
- however, we have kept the summary report of effluent releases in the format outlined in appendix B of Reg-ulatory Guide 1.21 Revision 1, June 1974.
Very truly yours, Bruce A. Snow
Vl S
~
P