ML20246L709
| ML20246L709 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 06/30/1989 |
| From: | Cargill E, Fantacci C GULF STATES UTILITIES CO. |
| To: | |
| Shared Package | |
| ML20246L706 | List: |
| References | |
| NUDOCS 8909070026 | |
| Download: ML20246L709 (83) | |
Text
{{#Wiki_filter:_ _, i RIVER BEND STATION, UNIT 1 SEMIJ N UAL RADIOeTTIVE EFFLUEtTT RELEASE REPORT REPORT PERIOD: January 1, 1989 Through June 30, 1989 REVIEWED BY -b L M . Fahtacci, Radiological Engineering Supervisor AhROVED BY ~-
/ / f ~
E. M/Cargill, Dir for f Radiological Programs 8909070026 890829 PDR ADOCK 05000458 R PDC
h. l l l TABLE OF 00Nrn es PAGE I. INTRODUCTION 3 II. SUPPLEME24TAL INFORMATIG4 2 A. REGUIATORY LIMITS 2 B. MAXIMUM PERMISSIBLE CONCEt&FATIONS 6 C. AVERAGE ENERGY 6 D. MEASURDDTTS AND APPROXIMATIONS OF TOTAL RADIOACTIVITY 6 E. BATCH RELEASES 8 F. ABNORMAL RELEASES 9 G. ESTIMATE OF TOTAL ERROR 12 III. GASEQUS EFFLUENTS
SUMMARY
IN00WATION 33 IV. LIQUID EFFLUENTS SUWARY INFORMATION 33 V. SOLID WASTE 13 VI. RADIOLOGICAL IMPACT ON MAN 13 VII. METEOROLOGICAL DATA 13 VIII. RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION OPERABILITY 13 IX. RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATIGi OPERABILITY I4 X. LIQUID HOLD UP TANKS 14 XI. RADIOLOGICAL ENVIRONME2EAL !ONITORING 34 XII. IAND USE CENSUS 34 XIII. OFFSITE DOSE CALCUIATIGJ MANUAL (ODCM) 14 XIV. MAJOR CHANGES 'ID RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATME2C SYSTEMS 14 XV. PROCESS 00tRROL PROGRAM (PCP) 34 f
s PAGE XVI.- TABLES TABLE I RADIOACTIVE GASEQUS b"L9TE SAMPLING AND ANALYSIS PROGRAM 15 TABLE 2 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM 16 TABLE 3 GASBOUS EFFLUD7TS - SUMMATION OF ALL RELETLSES 17 TABLE 4 GASEQUS EFFLUaffS - CONDITIONALLY ELEVATED RELEASES 19 TABLE 5 -GASDOUS EFFLUDTTS - GROUND LEVEL RELEASES 22 TABLE 6 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES 25 TABLE 7 SOLID WASTE AND IRRADIATED FUEL SHIPME27PS 30
. Attachment 1: Offsite Dose Calculation Manual Qianges 33 l
LL _ __ _--- --- _ - _. _ - _ - - - - - - - - _ _ - - >
( SEMIMMEL RADIOACTIVE EFFWElff
-' RE2 EASE REPORP r
19CILITY: -River Bend Station,' Unit 1 LIGNSEE: Gulf States Utilities REPORP PERIOD:: ' January 1, 1989 through June 30, 1989 I. INTRODUCTION Enclosed is the Semiannual Radioactive Effluent Release' Report for the period'of January 1, 1989 through June 30, 1989. .This report is sulnitted in accordance with Technical Specification 6.9.3.8 of Appendix A to River Bend Station (RBS) License Number.NPF-47. II. SUbtRRY Gaseous release totals were a maximum of 5.3% of the quarterly whole
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body and critical organ technical specification limits. . Liquid releases were a maximum of 5.7% of their quarterly whole body and critical organ technical specification limits. 1
II. SUPPLEMENTAL INFORMATION A. Regulatory Limits
- 1. 10CFR20 Limits
- a. Fission and Activation Gases in- accordance with Technical Specification 3.11.2.1, the .
dose rate due to noble gases released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY shal] be limited to less than or equal. to 500 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin: DR TB
= Dose rate to the total body in mrems/yr n .
= '3.15 x 10 I Kg (X/Q) Qg 5 500 mrems/yr i=1 and DP skin
= Dose rate to the skin'in mrems/yr n .
= 3.15 x 10 I (Lj + 1.1M g)(X/Q) Qg 5 3000 mrems/yr l
i=1 (above terms defined in RBS ODCM).
- b. Radiciodines and Particulate l In accordance with Technical Specification 3.11.2.1, the dose rate due to iodine-131, fodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY shall be limited to less than or equal to 1500 mrems/yr to any organ:
DR ygggp = Dose rate to the organ t for the age group of interest from iodines, tritium, and 8 day particulate via the inhalation pathway in mrems/yr n .
= 5 1500 mrems/yr I Pg (X/Q)D 0 1 i=1 (above terms defined in RBS ODCM) 2
l 1-p. L c. Liquid Effluents L In accordance with Technical Specification 3.11.1.1, the-concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10CFR20, Appendix B, Table II, Column 2 for radionuclides other than dissolved and entrained noble gases. For dissolved or entrained noble
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l gases, the concentration shall be limited to 2 x 10 microcuries/ml total activity. 2, 10CFR50, Appendix I Limits-
- a. Fission and Activation Gases In accordance with Technical Specification 3.11.2.2, the air dose due to noble gases released in gaseous effluents to areas at or beyond the SITE BOUNDARY shall be limited:
to: D = The gamma air dose from radioactive noble Gamma-Air gases in mrad n
= I Mg (X/Q) Q g 5 5 mrads/qtr j,7 5 10 mrads/yr D = eta air dose from radioactive noble Beta-Air gases in mrad n
= I Ng (X/Q) Qg 5 10 mrads/qtr g,3 5 20 mrads/yr (above terms defined in RBS ODCM)
- b. Radiofodines and Particulate In accordance with Technical Specification 3.11.2.3, the dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, and all radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents releases to areas at and beyond the SITE BOUNDARY shall be limited to:
D * " * * '" "#** * ** #8"" (*) '" I68DPt specified age group from radiciodines, tritium, and 8 day particulate via the pathway of interest 3
I 1 n
~
= 3.17 x 10 I R 3 (X/Q)D 0 1 ini and/or n
= -8 3.17 x 10 I Rg (D/Q) Q g i=1' and D = Dose in mrem to the organ (t) of a specified age group from radiciodines, tritium, and B day particulate from all pathways n
= I D I&8DPt 5 7.5 mrems/qtr z=1 5 15 mrems/yr (above terms defined RBS ODCM)
- c. Liquid Effluents In accordance with Technical Specification 3.11.1.2, the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS shall be limited to:
Dg = AiTAt Q g (DF) D, n TOTALt it i=1 D = tal dose commitment to the TOTALt organ (t) due to all releases during the desired time Daterval in mrem 4
and D TOTAL 5 1.5 mrems/qtr Total Body 5 3 mrems/yr D TOTAL 5 5 mrems/qtr. Any Organ
$ 10 mrems/yr (above terms defined in RBS ODCM)
~3. 40CFR190 Limits In accordance with Technical Specification 3.11.4, the annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and to radiation from uranium fuel cycle sources, shall be limited to:
5- 25 mrems to the total body or any organ (except' the thyroid) 5 75 mrems to the thyroid
- 4. Miscellaneous Limits
- a. Ventilation Exhaust Treatment System In accordance with Technical Specification 3.11.2.5, the VENTILATION EXHAUST TREATMENT SYS^ TEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses, due to gaseous effluent releases to areas at and beyond the SITE BOUNDARY would exceed 0.3 mrem to any organ in a 31 day period.
- b. Liquid Radwaste Treatment System In accordance with Technical Specification 3.11.1.3, the Ifquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses, due to the liquid effluent, to UNRESTRICTED AREAS would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31 day period.
5
r- > l l B. Maximum Permissible Concentrations
- 1. Gascous Releases The RBS Radiological Effluents Technical Specifications (RETS) for gaseous releases are based on the dose rate restrictions of 10CFR20, rather than the Maximum Permissible Concentrations (MPC) listed in 10CTR20 Appendix B. Table II, Column 1.
- 2. Liquid Releases The Maximum Permissible Concentration of radioactive materials in liquid effluents is limited by 10CFR20, Appendix B, Table II, Column 2. The MPC chosen is the most !
conservative value (i.e., the lowest) of either the soluble l or insoluble MPC for each radionuclides. I C. Average Energy Not applicable to RBS RETS l l l D. Measurements and Approximations of Total Radioactivity i i
- 1. Gaseous Effluents l
- a. Fission and Activation Gases !
Periodic grab samples are obtained from the Main Plant Exhaust Duct, Fuel Building Exhaust Vent and Radwaste ! Building Exhaust Vent. These samples are analyzed ! utilizing high resolution germanium detectors coupled to computerized pulse height analyzers. The sampling and analysis frequencies are described in Table 1. Sampling ! l and analysis of these effluent streams provide noble gas radionuclides relative abundances which can then be ; l applied to the noble gas gross activity and gross l activity release rate to obtain nuclide specific ! activities and release rates. The noble gas gross I activity released within a specific time period is ! I determined by integrating the stack monitor release rate i over the considered time period. An average correction j factor of 0.017 has been utilized for this report period j due to the infrequent detection of noble gas radionuclides in the effluent stream. If no activity was ! detected between stack grab samples and significant l increase in hourly averages were recorded, the nuclide relative abundances of the last sample which indicated the presence of activity was utilized to obtain nuclide ; specific activities. J l 1 l I 1 6 1 1 i i_ __ _ _ _ _ _ _ _ . . . _ _ _ _ . . __ ____ - . _ _ . _ _ _j
- b. Particulate and lodines Particulate and iodines are continuously sampled from each of the three release points utilizing a particulate filter and charcoal cartridge in line with a sample pump (stack- monitor pump). These filters and charcoal cartridges are removed and analyzed in accordance with the. frequencies .specified in Table 1. Analysis is performed to identify 'and' quantify radionuclides utilizing high resolution germanium detectors coupled to computerized pulse height analyzers. Given the nuclide specific activity concentrations, process flow rate, and time which the sample cevered; the nuclide specific activity released to the environment can be obtained.
Due to the continuous sampling . process, it is assumed that the radioactive material is released to the environment at a constant rate within the- sampling period. Sr-89 and Sr-90 are quantitatively analyzed by counting the digested filter precipitate with a get flow proportional counter. Gross alpha analysis is periurced using a zinc sulfide scintillation counter.
- c. Tritium Tritium grab samples are obtained from the three release points at the specified frequencies listed in Table 1 utilizing an ice bath condensation collection method.
The collected sample is then analyzed utilizing a Liquid Scintillation Counter. Given the tritium concentration, process flow rate, and time period for which the sample is obtained, the tritium activity released to the environment can be determined. Due to the frequency of sampling, it is assumed that the tritium is released 'to the environment at a constant rate within the time period for which the sample is obtained.
- 2. Liquid Effluents Representative grab samples are obtained from the appropriate sample recovery tank and analyzed prior to release of the tank in accordance with the frequencies listed in Table 2.
Analysis for gamma emitting nuclides (including dissolved and entrained noble gases) is performed utilizing a high resolution germanium detector coupled to a computerized pulse height analyzer. Tritium concentration is determined utilizing a Liquid Scintillation Counter. Sr-89 and Sr-90 are quantitatively analyzed by counting the precipitate with a gas flow proportional counter. Fe-55 is counted with a Liquid Scintillation Counter after digestion of the iron. Gross alpha analysis is performed using a zine sulfide scintillation counter. Given the nuclide specific activity concentration and total volume of the tank that was released, the activity of each nuclide released to the environment can be determined. 7
p .- i -< E. Batch Releases-
- 1. ' Liquid 1st Quarter 1989
- a. Number of batch releases : 99
- b. Total time period for batch releases : 674.05 lu: 1
- c. Maximum time period for batch releases : 9.50 hr-
'd. Average time period for batch releases : 6.91'hr
- e. Minimum time period for a batch release : 1.73 hr
- f. Average stream flow during periods of 3
release of effluent into a flowing stream : 833,000 ft /sec I i 2nd Quarter 1989
- a. Number of batch releases : 176
- b. Total tire period for batch releases :1279.35 hr j
- c. Maximnm time period for batch releases : 9.35 hr
- d. Average time period for batch releases : 7.39 hr
- e. Minimum time period for a batch release : 1,35 hr
- f. Average stream flow during periods of 3
release of effluent into a flowing stream : 726,000 ft /sec
- 2. Gaseous l
All gaseous releases to River Bend Station are considered continuous releases. , I
)
1 1 8 ' i _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ . . _ _ _ _ _ _ . . _ . _ . _ _ _ . _ . . . . _ . _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ ______.______J
- y if'!
l
' T. Abnormal Releases
- 1. Liquid 1st Quarter 1989
- a. Number of releases : 0
- b. Total activity released : O Curies 2nd Quarter 1989
- a. Number of releases : 3
- b. Total activity released : 4.93E-04 Curies
- 2. Gaseous 1st Quar:er 1989
- a. Number of releases : 0
- b. Total activity released : O Curies 2nd Quarter 1989
- a. Number of releases : 0 B. Total activity released : O Curies During the reporting period 1 -Jan-89 thru 30-Jun-89, RBS experienced three (3) incidents which would be classified as an abnormal liquid release. Each incident is described below:
- 1. April 4, 1989 incident:
During Refueling Outage Two (RT-2), the drywell/ containment unit coolers were realigned from the normal service water system to the turbine chilled water system. After realignment, radioactivity analysis of the turbine chilled water system indicated a slight amount of : radioactivity. Initial concentrations detected were as follows: ISDTOPE CONCENTRATION (uCJ/ml) Mn-54 1.36E-06 Co-58 1.68E-06 Co-60 2.81E-06 Zn-65 2.11E-07 1 In preparation for restoring normal service water to these unit coolers., RBS began flushing these components (to Radwaste) to remove the radioactive contamination. After approximately four (4) days of flushing, Mn-54 at a concentration of 3.89 E-08 uC1/mi still remained in the unit coolers. In order to reduce the total volume of water flushed. to Radwaste, RBS decided to restore normal service water and co;;equently flush this residual radioactivity to the circulating water fiume. This activity will 91timately { 9
n_ . - . p
- ,4 be discharged to the Mississippi River via the coolingL l
- aver - blowdown -line 'which is monitored by the liquid radiation monitor IRMS-RE108.
- A conservative estimate of the volume of the normal service water system inside of containment. and drywell is 2,992-gallons. (1.1 E 07ml). Based on this, on 4/5/89-approximately 0.428 uCi (4.28 E-13 'C1) of Mn-54 was released via this pathway.-
II. ' June 13, 1989 Incident: During Refueling Outage Two (RF-2), a visual inspection of' the "D" circulating water box inlet revealed that a tube plug 'had become dislodged from the tube sheet. This 3/4 W , inch opening provided a pathway allowing condensate water to leak into the circulating water system under the-following conditions:
- 1. . Condenser hot well level greater than 12.5 feet and
- 2. Condenser at atmospheric pressure or greater.
Review ~ of operator logs and discussions with Operations personnel revealed that during the ptriod 6/2/89 thru 6/7/89 ~both of the above conditions existed. Conservative estimates would project tLct approximately 60,750 gallons (2.3 E 08 ml) of condensate was pumped to the circulating water flume via normal service water system. Consequently the following- activity is projected to have been released during this time period. ISOTOPE ACTIVITY (uCi} Mn-54 1.54 E 02 Co-58 3.27 E 01 Co-60 2.65 E 02 This activity will ultimately be discharged to the Mississippi River via the cooling tower blowdown line which is monitored by the liquid radiation monitor 1RMS-RE108. 10
III. June 28, 1989 Incident: On June 28, 1989 approximately four (4)l inches of rain fell during .the period between 5:00 and 9:00 a.m. This coupled with already. saturated soils caused brief. but extreme flooding in areas that normally drain well. At 6:00 a.m. the sewage treatment plant operctor noticed abnormally high flows and suspecting storm water intrusion, made adjustments for maximum throughput in the sewage treatment plant. At this time a hydraulic surge was experienced in the sewage treatment plant. This .in turn disrupted the sludge blanket which caused a slight amount of carryover of radioactivity which was detected in the sewage treatment plant outfall. This composite grab sampler indicated the following concentrations / activities which were ' released during the time period 6/21/89 thru 6/28/89: ISOTOPE CONCENTRATION (uC1/ml) ACTIVITY (uCi) Co-60 1.90 E-08 31.2 Mn-54 6.04 E-09 9.9 I i 11 I ( b _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
( ,
,3, - ,
G. Estimate of Total Error , i
- 1. Liquid The maximum error associated with sample collection, '!
laboratory analysis, and discharge volume are collectively estimated to be: ; 1 Fission and Activation Products : + 14.2% Tritium : + 14.2%
- Dissolved and Entrained Noble Gases
- + 14.2%
_ l Gross Alpha Radioactivity : 1 14.2% j
- 2. Gaseous i l
The maximum errors (not- including sample line loss) j associated with sample flow, process flow, sample collection, ! monitor accuracy and laboratory analysis are collectively ! estimated to be: j i Fission and Activation Gases : + 37.0% l Iodines : 1 18.6% j Particulate : 1 18.6% Tritium : 1 18.2%
- 3. Determination of Total Error !
The total errer (i.e., collective error due to sample l collection, laboratory analysis, sample flow, process flow, monitor accuracy, etc.) is calculated using the following
)
i equation: I
-ll 9
E T
= (E3 )2 + (E2 ) + ... (E )2 -
where: ' i E = t tal error T j E,E = individual errors due to sample 3 2 *** n collect d on, laboratory analysis, sample j flow, process flow, monitor accuracy, i etc. l l l i i l 12
- - - - _ _ - - _ __ - --__-_--_--__--______-_-_--___-a
l' III. GASEOUS EFFLUENTS '
SUMMARY
. IhTORMATION
-Refer to Tables 3, 4 and 5 for Summation of All Releases-and'Nuclides Released, . respectively. .It should benoted that' an entry.:of' "0.00E+00" Ci or 'uC1/sec in this- section1 does.not indicate the absence of' a radionuclides;.,but, rather, . indicates that' .the-concentration of the particular radionuclides- was below the Lower-
- Limit of Detection (LLD) as listed in Table 1.
. IV. .- LIQUID EFFLUENTS
SUMMARY
INFORMATION-L Refer to Table 6.for Summation of All Releases and Nuclides Released. L It.'should be noted that an entry of."0.00E+00" C1 or uCi/ml in this' section does not indicate the absence'of a radionuclides; but, rather,. Indicates that the concentration of the particular radionuclides was below the Lower Limit of Detection (LLD) a4 listed in Table 2. V. SOLID WASTE Refer to Table 7 V1. RADIOLOGICAL IMPACT ON MAN This information will be provided in the end-of-year report as described in Technical Specification 6.9.1.8. VII. METEOROLOGICAL DATA This information will be provided in the end-of-year report as described in Technical Specification 6.9.1.8. VIII. RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION OPERABILITY The Liquid Radwaste Effluent Monitor (IRMS-RE107) was not restored to operable status within 14-days as required by Technical Specification 3.3.7.10. On.5/29/89 at 0730, LCO 89-279 was initiated declaring 1RMS-RE107 inoperable. IRMS-RE107 was declared operable on 6/12/89 at 1459. The reason for exceeding the 14-day restoration period was improper priortization of work activities. During this time period River Bend Station was actively closing out work items in preparation for starting up from its scheduled Refueling Outage 2. The Maintenance Work Order for repair of 1RMS-RE107 was not added to the priority list for start-up. Upon discovery of the expiration of the 14-day window, a new Maintenance Work Order was initiated and the instrument was returned to service at 1459, 7 hours and 29 minutes in excess of the requirement. I 13
IX. RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION OPERABILITY-The minimum number of channels required to be OPERAB14 as described in Table 3.3.7.11-1 of Technical Specification'3.3.7.11 were, if inoperable at any time in the period 1/1/89 through 6/30/89, restored to operable status within the required time. Reporting of these inoperable channelt in this report is, therefore, not required. ; X. LIQUID HOLD UP TANKS The maximum quantity of radioactive material, excluding tritium and dissolved or entrained noble gases, contained in any unprotected outdoor tank during the period of 1/1/89 through 6/30/89 was less than or equal to tle 10 curie limit as required by Technical Specification 3.11.1.4. i XI. RADIOLOGICAL ENVIRONMENTAL MONITORING There were no changes in sampling locations for the Radiological Environmental Monitoring Program (REMP) during the reporting period i 1/1/89 through 6/30/89. l l XII. IAND USE CENSUS l This information will be presented in the end-of-year report. The Land Use Census results will be available in September 1989. XIII. OFFSITE DOSE CALCULATION MANUAL (ODCM) j Attachment 1 identifies the changes made to the ODCM during the 1st j and 2nd Quarter of 1989 Each change and the justification for the i change is identified on the change form. XIV. MAJOR CHANGES TO RADI0 ACTIVE LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS i There were no major change > u the radioactive liquid, gaseous, and I solid waste treatment systems for the period of 1/1/89 through 6/30/89. l XV. PROCESS CONTROL PROGRAM (PCP) L 1 l No changes were made to the RPS Process Control Program (PCP) for the I period 1/1/89 through 6/30/89. l 14 __-___________D
f) oD L) tLl i( m z 0 m /
- 8 '- s 4 8 8
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TABLE 2 _ RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PFOGPAM Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD) Type Frequency Frequency Analysis (uCi/ml)
~
A. Batch Waste P P Principal Ganma 5x10 Release Each Batch Each Batch Emitters; except -6 (Liquid for Ce-144 5x10 Radwaste Recovery -6 Sample I-131 1x10 Tanks P M Dissolved and 1x10
-5 One Batch /M Entrained Gases (Ga:Ima Emitters)
P M H-3 1x10
-5 ;
Each Batch Ccxnposite
~
Gross Alpha 1x10 I Sr-89, S;-90 5x10
-8 P Q Each Batch Ccruposite
-6 Fe-55 1x10 P = Prior to each radioactive release M = At least once per 31 days Q = At least once per 92 dap 4
16
- _ - - - - - - - - - - - - - - - - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -- ------------_----J
N(4'-n t - , s- TAB m 3 Effluent and Waste Disposal Semi-Annual Report .. -1989 Year-m g raaamm Effluents - S-tion of All Relaa- -1/2 Quarters
.j. i I. -I i Est. .!
! ! Unit. ! Quarter l Quarter ! 'Ibtal' l l- ,
!' ! 1 l 2 l' Error, % l A. Noble Gases
~!
i I i -l l
.j. 1. Total release- ! Ci ! 8.31E-01 ! 0.00E+00 l 3.70E+01 !
l- l- l l ! .!
! I i i- 1 l 2.- Average release rate l uCi/sec j 1.05E-01 l 0.00E+00 l l for period' ! l '! !
! I i i !
.l .3. Percent of technical l % ! 3.36E-03-l 0.00E+00 l l specification limit (1) ! ! l l
- B. Iodines l l I-131 i i i !
! 1. Total I-131 and I-133 l Ci l 9.42E-05 ! 0.00E+00 ! 1.86E+01 !
! ! I-133 i i _ .I !
! ! Ci ! 1.35E-03 ! 3.93E-06 l 1.86E+01 l j .. I I-131 i ! i
! 2. Average release rate. l uCi ! 1.19E-05 l 0.00E+00 l j for period ! sec i j j l~ ! I-133 1.71E-04 ! 4.99E-07 l l 1 i !
! 3. 'I-131 +'I-133 m tribu- l % ! ! !
l tion percent of techni- l l 5.37E+00 l 2.77E-03 l l cal specification limit ! ! ! !
- C. Particulate
! I i i i !
-l 1. Particulate with half- l Ci l l l l l- lives of > 8 days ! ! 8.28E-03 l (2) ! 1.86E+01 l l ; I I i l 2. Average release rate l uCi/sec l l l
! -for period ! ! 1.05E-03 ! (2) !
'I .
; ; i !
! 3. Percent of technical l % l l l l specification limit l ! 3.29E-01 ! (2) l l I I I i j 4. Gross alpha l Ci l l !
! radioactivity ! l 0.00E+00 ! 0.00E+00 l 17
. I
k. I i i i ! Est. l l l Unit l- Quarter ! ouarter l Total l l l l 1 l 2 l Error, % l D. Tritium l 1 i i ! ! l- 1. Total release ! Ci l 7.48E-01 l 2.47E-01 l 1.82E+01 l l ! ! l ! ! l 1 i i l 2. Average release rate j uCi/sec ! 9.50E-02 ! 3.13E-02 l l for period ! ! ! ! I i i ! !
.! 3. Percent of technical l % ! 3.09E-03 l 1.04E-03 !
! specification limit l l l l (1) Garana airdose limit of 5 mrads/qtr (T.S.3.11.2.2.a) .
(2) Not available for submission at this time, supplanental report to follow. I8
TABLE 4 Effluent and Waste Disposal Semi-Annual ItTort 1989 Year cer: Effluents - Conditionally Elevated Ibleases 1/2 Ouarters Continuous Mode Batch Mode ! I I I i I l l Nuclides Released ! Unit ! Quarter 1 ! Ouarter 2 ! Quarter 1 l Quarter 2_l
- 1. Fission Gases l I I l l I l l Argon-41 l Ci ! 0.00E+00 l 0.00E+00 !
N/A ! N/A l l l I l i I I l Krypton-85m ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A l N/A l l 1 i i i l ! l Krypton-85 ! Ci ! 0.00E+00 ! 0.00E+00 l N/A l N/A l l l l I I I l l Krypton-87 l Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l l l l I i ! l l Krypton-88 l Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l l l l l l i l l Xenon-133m ! Ci ! O.00E+00 ! 0.00E+00 ! N/A ! N/A l l I i i i I l l Xenon-133 ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l l I i i i l l l ' Xenon-135m l Ci ! 0.00E+00 ! 0.00E+00 l N/A ! N/A ! l l l I I i ! ! Xenon-135 ! Ci ! 8.31E-01 ! 0.00E+00 ! N/A ! N/A l l l l i i l l Xenon-137 ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A l N/A l ! ! I I I I ! ! Xenon-138 ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l l I i i i I l l unidentified ! Ci ! N/A ! N/A ! N/A l N/A l l i i i i I l l Total for period ! Ci ! 8.31E-01 1 0.00E+00 ! N/A ! N/A l
- 2. Gaseous Iodines
! I i i i I l l Iodine-131 l Ci ! 9.39E-05 l 0.00E+00 ! N/A ! N/A l l i ; I ! ! ! l Iodine-132 l Ci ! 0.00E+00 ! 0.00E+00 l N/A ! - N/A l i . . . . . i e e e e e e e l Iodine-133 ! Ci ! 1.35E-03 l 3.93E-06 ! N/A l N/A l l I I I I I a ! ! Iodine-134 ! Ci ! 0.00E+00 l 0.00E+00 ! N/A l N/A l ! j I I I l . ! l Iodine-135 l Ci ! 0.00E+00 ! 0.00E+00 l N/A ! N/A l l j ! ! ! I ! ! Total l Ci l 1.44E-03 ! 3.93E-06 l N/A ! N/A ! 19
Continuous Mode Batch Mode l' I I I I I l l_ Nuclides Released ! Unit ! Quarter 1 l Quarter 2 l Quarter 1 ! Quarter 2 l
- 3. Particulate l l l l I i !
j Strontitrn-89 l Ci ! 6.09E-06 !. (1) N/A l i N/A l i i-* i . , e e : a e e
!- Strontitzn-90 l Ci l 0.00E+00 ! (1) l N/A ! N/A l l 1 1 i ! !
l Cesium-134 ! Ci l 0.00E+00 ! 0.00E+00 l N/A ! N/A l
! I I I I I l l Cesium-137 ! Ci ! 0.00E+00 l 0.00E+00 l N/A l N/A l l l 1 I l l l l Barium-Lanthanum-140! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l
! l l i ! l !
! Cobalt-60 l Ci ! 1.30E-05 ! 0.00E+00 ! 11/A ! N/A l
! I I l I I !
l Chrcxnium-51 l Ci ! 0.00E+00 0.00E+00 ! N/A ! N/A l-
! I i ! l !
! Zirconium-Niobium-951 Ci l 0.00E+00 ! 0.00E+00 l N/A ! N/A l l ! I I I I l l Zine-65 ! Ci ! 0.00E+00 l 0.00E+00 ! N/A l N/A l
! I I I I l !
'l Iron-59 l Ci ! 0.00E+00 l 0.00E+00 ! N/A ! N/A l l l l l l I !
! Manganese-54 l Ci 8.26E-05 ! 0.00E+00 ! N/A l N/A l l- ! u l l l l l Iodine-131 ! Ci ! 0.00E+00 1 0.00E+00 ! N/A ! N/A l I ! I i ! l l' l N/A l j Cerium-141 l Ci l 0.00E+00 l 0.00E+00 ! N/A ! !
! ! I I ! l !
l Cerium-144 ! Ci ! 0.00E+00 ! 0.00E+00 l N/A ! N/A l l l ! I I i ! l Cobalt-58 ! Ci ! 4.99E-06 l 0.00E+00 ! N/A ! N/A ! l t i l ! l
! Silver-110m ! Ci l 0.00E+00 ! 0.00E+00 l N/A ! N/A l l l l I I I !
l j Molybdenum-99 Ci l 0.00E+00 ! 0.00E+00 ! N/A l N/A l I I I l l l l unidentified l Ci l N/A ! N/A ! N/A l N/A l l I I I I ! l l Total for period l Ci l 1.07E-04 ! (1) l N/A l N/A l l 20
Continuous Mode Batch Mode
! I I I I l !
l Nuclides Released ! Unit ! Quarter 1 ! Quarter 2 ! Quarter 1 ! Quarter 2 i 4.0 Tritium I- I I I i I l l Hydrocen-3 l Ci ! 7.31E-01 ! 2.08E-01 1 N/A ! N/A l
.(1) Not available for submission at this time, suppl mental report to follow.
Main Plant Exhaust Duct is considered a conditionally elevated release point. 1 21
. - - - - - - _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____i
i i i TABLE 5 l Effluent and Waste Disposal Smi-Annual Report 1989 Year , Gaseous Effluents - Ground Ievel Releases 1/2 Quarters Continuous Mode Batch Mode
! I I I I i !
l Nuclides Released ! Unit ! Quarter 1 ! Cuarter 2 ! Quarter 1 ! Quarter 2 !
- 1. Fission Gases
! I l l l l l l Argon-41 l Ci ! 0.00E+00 ! 0.00E+00 ! N/A I N/A l l l i i ! i !
! Krypton-85m ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A !
! l 1 I l l Krypton-85 ! Ci l 0.00E+00 l 0.00E+00 l N/A ! N/A l l l I i i l l l Krypton-87 ! Ci ! 0.00E+00 ! 0.00E+00 l N/A ! N/A l l l l l 1 i !
! Krypton-88 ! Ci l 0.00EM0 l 0.00E+00 ! N/A ! N/A l l l l l l 1 !
l Xenon-133m ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l l i i i i ! !
! Xenon-133 ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l l l I I I I !
! Xenon-135m ! Ci ! 0.00E+00 ! 0.00E+00 l N/A ! N/A l l l I I I I ~!
! Xenon-135 ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l l l l l l l l
! Xenon-137 ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l l l 1 i i !
! Xenon-138 l Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A !
l I l i i ! l unidentified ! Ci ! N/A ! N/A ! N/A ! N/A l l 1 l l ! ! ! l Total for period ! Ci l 0.00E+00 ! 0.00E+00 ! N/A l N/A l
- 2. Gaseous Iodines
! l
! I i ! l l Iodine-131 l Ci l 3.93E-07 l 0.00E+00 ! N/A ! N/A l l I l l l l l l Iodine-:_32 ! Ci ! 0.00E@0 ! 0.00E+00 ! N/A ! N/A l l I I l l 1 l
! ! Iodine-133 l Ci ! 0.00E+00 ! 0.00E+00 ! N/A l N/A I l l l I l l l l Iodine-134 ! Ci ! 0.00E+00 ! 0.00E+00 l N/A ! N/A l l } I i I l ! l Iodine-135 ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l
! I i i I I l l Total l Ci ! 3.93E-07 ! 0.00E+00 ! N/A l N/A l l
22
P 1 t. Continuous Mode Batch Mode l . I i i i I l l Nuclides Released l Unit ! Ouarter 1 ! Ouarter 2 l Ouarter 1 l Ouarter 2 !
- 3. Particulate
! I I i i i !
l Strontium-89 ! Ci ! 0.00E+00 ! (1) l N/A ! N/A. I
! I i i i i !
Strontium-90 ! Ci ! 0.00E+00 l -(1) ! N/A ! N/A j l-l l l I i i I Cesium-134 ! Ci 0.00E+00 ' O.00EMO ! N/A' l N/A l: l e ' s e
! Cesium-137 l C1 , 0.00E+00 l 0.00E+00 l N/A l N/A l ! I i i i i ! ! Barium-Lanthanum-140! Ci ! 0.00E+00 ! 0.00E+00 ! N/A l N/A l ! I i i- i i I l Cobalt-60 l Ci ! 7.84E-07 ! 2.19E-06 l N/A l N/A l l l l I I i !
l Chranium-51 l Ci ! 0.00E+00 1 0.00E+00 l N/A' l N/A l l l 1 i i i ! l Zirconium-Niobium-95! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l j i i i I i !
!- Zine-65 ! Ci ! 0.00E+00 ! 0.00E+00 l N/A ! N/A l ! ! ! ! I i ! ! Iron-59 ! Ci ! 0.00E+00 ! 0.00E+00 ! N/A ! N/A l ! I i i i i !
l Manganese-54 l Ci ! 0.00E+00 l 8.45E-07 ! N/A ! N/A l l i i i i i I l Iodine-131 ! Ci ! 0.00E+00 l 0.00E+00 l N/A ! N/A l l I i i i I l
! Iodine-132 l Ci ! 0.00E+00 l 0.00E+00 l N/A ! N/A l I i I i I i i Iodine-133 ! Ci ! 0.00E+00 ! 0.00E+00 l N/A l N/A !
I i I i i i l Cerium-141 ! Ci ' O.00E+00 1 0.00E+00 ! N/A ! N/A l l l 1 i i l l Cerium-144 l Ci ! 0.00E+00 l 0.00E+00 l N/A l N/A l I i i i i ! ! l Cobalt-58 ! Ci ! 0.00E+00 ! 3.26E-07 ! N/A ! N/A l l l l l l l ! l Silver-210m l Ci ! 0.00E+00 ! 0.00E-00 l N/A l N/A l l l l l i i !
! Molybdenum-99 ! Ci ! 0.00E+00 l 0.00E+00 ! N/A ! N/A l l l l I i i I
! unidentified l Ci ! N/A ! N/A l N/A ! N/A l l l i i i l l l Total for period l Ci ! 7.84E-07 ! (1) ! N/A l N/A l 23
. _ - _ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ . __ ___________ _ _ _ _ _ _ - _ _ a
i i Continuous Mode Batch Mode l i i i i. i !
.! 'Nuclides Released ! Unit '! Quarter 1 ! Quarter 2 I Quarter 1 1 Ouarter 2'l 4.0 Tritium
) i i i i i !
l Hydrcnen-3 i Ci ! 1.74E-02 ! 3.89E-02 l N/A ! N/A l (1) Not available for subnission at this time, supplemental report to follow. Fuel Building Exhaust Vent and Radwaste Building Exhaust Vent are considered ground level release points. 24
-_--__-___ _ _______ _ _ - A
TAB 2 6 Effluent and Waste Disposal Semi-Annual Report 1989 Year TArmid Effluents - Sunanation of All Beleases j i l l l Est. l l l Unit l Quarter l Quarter l Total l l ! ! 1 ! 2 l Error, % l A. Fission and activation products
! I I i i !
! 3. Total release (not in- l Ci ! 5.83E-01 l (4) l 1.42E+01 l l cluding tritium, gases, l l l l l l alpha ! ! ! ! !
l I I I I l 2. Average diluted concen- I uCi/11 l 4.61E-07 l (4) l l tration during period ! ! ! l l l l l l l 3. Percent of applicable j % l 4.64E+01 l (4) l l limit (1) ! ! ! l l B. Tritium l 1 I i i l l 1. Total release ! Ci l 1.48E+00 l 2.23E+00 l 1.42E+01 l l ! ! ! ! !
! l ! l !
l l 2. Average diluted concen- ! uCi/ml l 1.17E-06 l 1.89E-06 l l tration during period ! ! ! l l l l l l l 3. Percent of applicable ! % l 3.93E-02 ! 6.31E-02 l l limit (2) ! ! ! ! C. Dissolved and entrai3ed gases ;
! I I I ! I I i 1. 'Ibtal release ! Ci l 3.66E-03 l 6.56E-05 ! 1.42E+01 l l
! ! ! ! ! l l l ! l l l 2. Average diluted concen- l uCi/ml l 2.90E-09 l 5.56E-31 l ,
j l tration during period ! ! ! l
! I I I l
, l 3. Percent of applicable ! % l 1.45E-03 l 2.78E-05 l l limit (3) ! ! ! l l 25 i l E____________ .____ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _
.__ __ _ __ __ __ _ j
r
! i i j l Est. l l l Unit l Quarter l Quarter l Total l l l l 1 ! 2 ! Ener, % !
D. Gross alpha radioactivity
. I .I I I !
! 1. Total release' l Ci ! 0.00E+00 l 0.00E+00 l 1.42E+01 !
! I i i l E. Volume of waste released- l liters l 5.84E+06 l 1.05E+07 l 8.73E-01 l (prior to dilution) ! ! ! ! l
! I I I !
F. Volume of dilution water l liters l 1.26E+09 l 1.18E+09 l 5.70E-01 '! used during period l l ! ! ! (1) One quarter of 5 Ci annual limit (1.25 Ci) ' for liquid releases except for tritium and dissolved or entrained noble gases frcrn 10CPR50 Appendix I. (2) 100FR20, Appendix B, Table II, Column 2 MPC limit of 3.00E-03 uCi/ml. (3) Technical Specification 3.11.1.1 liInit of 2.0E-04 uCi/ml for dissolved or entrained noble gases in liquid effluents. (4) Not available at this time, supplemental report to follow. 26 1__________________ _ 1
Effluent and Waste Dipl Smi-Annual Report 1989 Year G. Liquid Effluents 1/2 Quarters Continuous Mode Batch Mode I l l l l l ! l Nuclides Released ! Unit ! Ouarter I l Ouarter 2 ! Quarter I ! Ocarter 2 !
! i l 1 i l l l Hydrogen-3 l Ci ! N/A ! N/A ! 1.48E+00 l 2.23E+00 l I- i i i i I l
! Arsenic-76 ! Ci ! N/A l N/A l 1.31E-03 ! 0.00E+00 l
! l l I i l l
! Strontium-89 l Ci l N/A ! N/A l 1.17E-04 l (1) l
! l I I l I l l Strontium-90 l Ci l N/A ! N/A ! 0.00E+00 ! (1) l l l i i ! ! !
! Cesium-134 l Ci ! N/A ! N/A l 0.00E+00 ! 0.00E+00 l l l 1 l 1 i
! Cesium-137 ! Ci ! N/A ! N/A ! 0.00E+00 1 0.00E+00 l l l l l 1 i ~!
l Iodine-131 ! Ci l N/A ! N/A l 0.00E+00 ! 0.00E+00 l l l 1 : l l l l Iodine-132 l Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l l i i ! ! l I i Iodine-133 ! Ci l N/A ! N/A ! 0.00E+00 ! 0.00E+00 l
! l l l i I I
! Iodine-134 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l i- i i i i i 9 I I e t i 8
! Iodine-135 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l l l l l l 1 l l Sodium-24 ! ~
Ci ! N/A ! N/A l 3.11E-05 ! 0.00E+00 l
! I I I I I !
l Cobalt-58 ! Ci ! N/A l U/A ! 1.45E-02 ! 7.83E-03 l l l i l l l l l C& 21t-60 ! Ci ! N/A ! N/A ! 6.12E-02 ! 6.37E-02 i
! i l 1 i l l Iron-55 ! Ci ! N/A ! N/A ! 9.93E-03 ! (1) !
j' I i l l I ! l Iron-59 ! Ci ! N/A l N/A ! 8.55E-03 ! 3.42E-03 l
! l l l I l !
! Zinc-65 ! C1 ! N/A ! N/A ! 4.07E-03 ! 3.27E-03 l l I l l l I I l Mancanese-54 ! Ci ! N/A l N/A ! 2.55E-02 ! 2.53E-02 l l l i ! i l l
! Manganese-56 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l l l l l l l l l Chrmium-51 ! Ci ! N/A ! N/A ! 4.48E-01 ! 3.18E-02 l
- i i i i i -- s i i f i i t f l Zirconium-Niobium-95l Ci ! N/A ! N/A ! 3.70E-04 l 1.26E-04 !
! ! ! l l l l l Molybdenum-99 ! Ci ! N/A l N/A ! 4.14E-04 ! 0.00E+00 l l l I i i l l
! Technicium-99m ! Ci l N/A ! N/A ! 4.46E-04 ! 0.00E+00 l l l l l ! ! !
l Copper - 64 ! Ci l N/A ! N/A ! 0.00E+00 ! 0.00E+00 l l l l l ! l l l Tin - 113 ! Ci ! N/A ! N/A ! 1.45E-04 ! 2.91E-05 l 27
I G. (cont ) I Continuous Modo Batch Mode i
! I J l I i !
l Nuclides Released ! Unit ! Quarter 1 ! Quarter 2 ! Quarter 1 ! Quarter 2 l l l 1 i i i I l l Barium-lanthanum-140! Ci ! N/A l N/A ! 2.33E-04 1 0.00E+00 l l l l l l- l ! l Cerium-141 ! Ci ! N/A ! N/A ! 0.00E+00 0.00E+00 l i i i I ., j 1 l Cerium-144 ! Ci l N/A l N/A ! 6.97E-04 l 0.00E+00 l l 1 I I I i ! l Antimony-122 ! Ci l N/A l N/A ! 4.78E-05 ! 0.00E+00 l' I i i i 1 ! . ! l Antimony-124 ! Ci ! N/A ! N/A ! 2.10E-03 ! 1.14E-03 l l 1 i i i i i
! Rhodium - 105 ! Ci ! N/A ! N/A~
! 0.00E+00 ! 0.00E+00 l I i ! i I i !
j Brcznine - 82 l Ci l N/A l N/A l 0.00E+00 ! 0.00E+00 l l 1 i i i i l l Neptunium - 239 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l l l 1 i i i I l Niobium - 97 l Ci l N/A ! N/A ! 4.63E-04 ! 7.41E-04 !
! 1 i i i i !
l Silver - 110m ! Ci ! N/A ! N/A ! 8.88E l 2.53E-04 l l ; i i i I l l Strontium - 92 l Ci ! N/A ! N/A l 3.20E-04 ! 9.15E-05 l l l 1 i ! i ! l Tungsten - 187 ! Ci ! N/A ! N/A ! 1.26E-03 l 2.65E-05 l l i i i I i ! Total for period ! Ci ! N/A l N/A ! 2.06E+00- ! (1) j l , (1) Not available at this time. Supplemental report to follow. t 28 c . .. . . . . ____-_-_-_--___-_--_____--__---_----_--__-_-______-_-__-_-___-____________-______-____________-_a
7 j 1. i l B. Dissolved and Entrained Gases Continuous Mode Batch Mode
! I l I ! ! ! 1 l Nuclides Released ! Unit ! Ocarter 1 ! Ouarter 2 ! Ocarter 1 ! Ouarter 2 l ! ! ! l ! l l l Arcon 41 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l I I ! I l l l l Krypton-85m ! Ci ! N/A .! N/A ! 0.00E+00 ! 0.00E+00 l ! l l l l l l l Krypton-85 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l ! I I i ! l l l Krypton-87 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l ! I I I ! l l
l Krypton-88 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l
! l ! I i l ! ! enen-133m ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l ! I l l I I l l Xenon-133 ! Ci ! N/A ! N/A ! 1.99E-03 ! 9.00E-06 l l l l l l l l l Xenon-135m ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l ! i l l l l l l Xenon-135 ! Ci ! N/A l N/A ! 1.67E-03 ! 5.66E-05 l l I i l l I l l Xenon-137 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l ! ! ! ! ! I !
l Xenon-138 ! Ci ! N/A ! N/A ! 0.00E+00 ! 0.00E+00 l I l l I ! l l l unidentified ! Ci ! N/A ! N/A ! N/A ! N/A l I I I I I i l l Total for period ! Ci ! N/A ! N/A ! 3.66E-03 ! 6.56E-05 l s 29
TABLE 7 Effluent and Waste Diane =al munimranal Report 1989 Year Solid Maste and Irradiated Fuel 44 % Reporting Period 01/01/89 to 06/30/89 Otr 1/2 A.. Solid Waste Shipped for Burial or Disposal _ (Not. irradiated fuel)
! j l- 6-month l Waste i Est. 'Jbtal l l 1. Type of waste ! Unit ! Period l Class- l Error, % !
! . l - ! i i 1 l a. Spent resins, filter sludges l- m# l 1.74E+02 ! A-U, l l evaporator bottans, etc. . l ! -l l- l
.l ! Ci ! 2.95E+02 l A-S ! 3.50E+01 l
'l l I I I- !
l b. Dry compressible waste, l' m3 ! 1.48E+02 l A-U l 3.50E+01 l , l contaminated eauip, etc. ! .Ci ! 1.83E+00 l l l l l I I I I-m3 l c. Irradiated camponents, l l 0.00E+00 l N/A l 'O.00E+00-. l-l control rods, etc. ! Ci ! 0.00E+00 ! ! !
-l . i -
i i i l
-l d. Other (None) l m* l 0.00E+00 i N/A l . l~
l ! Ci ! 0.00E+00 l l 0.00E+00 l 1 30 l-l L________. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ i
l L 1 J
~2.. l '} ' TYPE OF l Spent Resins, tilter l Dry expressible j Irradiated Camponentsi Other ! j l
i j WASTE l sludges, evaporator l waste, contaminated l control rods, etc. l (None) ! - l ! bottams, etc. ! equip, etc. l l l
!. I I I i !
jPrinciple ! 11 - 3 3.71E-02 .! C-34 1.26E+00 l N/A ! N/A i l radionuclides l C-14 3.59E-01 l Cr-51 2.68E+00 l' ! ! j 1 l(Identity and l Na-24 4.4BE-27 l Mn-54 2.13E+03 ! I ! j % Abundance) j Cr-51 1.38E+03 l Fe-55 1.61E+03 l ! ! l ! j- Mn-54 1.89E+01 ! Co-58 2.97E+00 l l l
-l j l' ! Co-57 1.85E-02 l Fe-59 1.23E+00 i j j Co-58 5.48E+00 l Co-60 4.59E+03 l' ! ! .! j' Fe-59 3.49E-01 l Ni-63 9.54E-01 l l l u '! ! Co-60 3.03E+01 l Zn-65 6.97E+00 l l l l l Ni-63 8.02E-01 ! Am-241 2.90E-03-l l !
l l Cu-64' O.00E+00 l Pu-243 3.70E-03 ! ! l l_ l Zn-65 2.46E+00 ! Pu-239' 4.68E-03 l ! ! l l As-76 0.00E+00 l Cm-242 4.43E-03 i ! !
~! l 'Nb-95 4.00E-02 l Cm-244 3.51E-03 ! l l-l l Zr-95 2.46E-02 l Nb-95 4.48E-01 l l !
4.09E-05 j j j l l Mo-99 !
! ! Ag-310m 4.88E-02 j ! ! i' ! ! Sn-133 7.36E-04 I l l l l l Sb-122 7.04E-07 l l ! ! ! ! Sb-324 9.20E-02 l l l !
j j I-133 9.42E-04 ! ! ! l-l l I-333 0.00E+00 l l l l l l Te-131m 0.00E+00 l l l l l l Te-132 0.00E+00 l l l l l l.Ba-340 0.00E+00 ! ! ! !
! ! La-340 3.68E-07 l l ! !
l Ce-141 0.00E+00 l l l l l l l Ce-144 0.00E400 ! l l l l W-187 0.00E+00 l l l l l l Np-239 0.00E+00 l l l l j l Cm-242 3.34E-04 l l ! l l l Sc-46 0.00E+00 l l l ! j' ! Cm-244 2.86E-04 l l l l l- l Am-243 2.46E-04 l l l l l l Fe-55 2.70E+03 l l l l
! l Pu-241 3.42E-02 l l l l j j Pu-238 5.42E-04 l ! l l
- j. j Pu-239 3.97E-04 l l ! l l l Sr-89 5.02E-01 ! ! ! !
- _ __ _ _ -_ _ --_-_ - _-___-_____-___-__--__________-_-_____-__-_-____-___________ a
.___;4 j TYPE OF i Spent Resins, filter ! Dry compressible j Irradiated Camponentsi Other l
' j '. WASTE ! sludges, evaporator j waste, contaminated l control rods, etc. ! - (None) l
'l l bottoms, etc. ! equip, etc. l l l I- I. I i l l jAbove Detennined! l l l l u -!by: .
! l ,
l lA. measurement -l ! ! ! -! i . lB. estimation l C' l- C l N/A ! N/A I l lC. measurement l' l l l l l and correlation ! ! ! ! !
-! l 1 i I !
- l. l TYPE OF l Strong, Tight Liners l Strong, Tight Drums ! N/A l N/A l l CONTAINER !' ! ! ! l I i i i i !
l SOLIDIFICATIm 'l j N/A l N/A l N/A !
- l AGENT OR j Cement j j .
! ABSoRawr ! ! I ! !
. 3. SOLID WASTE DISPOSITION Number of Shipnents Mode of Transportation Destination 37 Truck, Exclusive Use CNSI, Barnwell, South Carolina B. IRRADIATED FUEL SHIPVENTS (Disposition)
Number of Shipnents Mode of Transportation Destination 0 N/A N/A 32
AmOM2fr #1 OPP SI7E DOGE % g Mumt. aga;s:s . 1 ) 33 l
n-_. . , - 1 I I I
-1 I
l ATTACILET - I l ODCM/ PROCEDURE REVISION SEEET NO.89-01 l 1 1 I I I l DESCRIBE THE IhTORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR AND A COMPLETE l l DESCRIPTION OF THE CHANGE (S) MADE TO THE ODCM: l 1 I l Chance all FSAR to USAR l l naces 3.6.14.30.68 l 1 l l Channe re ferences to USAR to reflect new revision of Safety Analysis Report j l i I I I I I I I I I I I i i i i I I i 1. I l l I I I I I - 1 I I I I i 1 I I l COMMENTS: l l l l l l l 1 1 I I I I I WILL THIS CHANGE REDUCE THE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS OR l l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO X l l l l WILL THIS CHANGE REQUIRED REVISION TO LOWER TIER IMPLEMENTING PROCEDURES OR l l COMPUTER PROGRAPS. YES NO X l 1 . l l THESE CHANGES HAVE BEE!! REVIEWED AND FOUND ACCEPT BLE, PURSUANT TO TECHNICAL l l SPECIFICATION 6.S.2. l l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: ' 92 / ~ ~/ 8 3/F9 l REVIEVED, DIRECTOR - RADIOLOGICAL PROGRAMS:
'/~
VY /9 MI DATE I I I I I l l ll l l l l ATTACHMEhT l PAGE I ll l 1 1 1 I I 0F I ll RSP-0008 I REV - 2 l PAGE 110 0F 110 l 1 1 II I I I M c - - _ _-- _ __ _ _. _ .
1.0 INTRODUCTION
l'.1 PURPOSE l This manual provides a concise description of the environmental dose models and techniques.used to calculate offsite doses resulting from
. measured or projected releases of radioactive materials from Gulf States Utilities'. River Bend Nuclear Station. It also provides the methodology for calculating effluent mor.itoring setpoints and allowable release rates to ensure compliance with the Radiological Effluent Technical Specifications (RETS) of Gulf States Utilities, River Bend Station. This manual also contains a description of the Radiological Environmental Monitoring Program which includes san.ple point descriptions for both onsite and offsite locations and sampling and analysis frequencies.
The ODCM follows the methodology and models suggested by the " Guidance. Manual for Preparation of' Radiological Effluent Technical Specifications for Nuclear Power Plants" (NUREG-0133, dated October 1978) and " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I" (Regulatory Guide 1.109, Rev.'1, dated October 1977). Alternate calculational methods may be used from those presented as long as the overall methodology does not change or as long as the alternative methods. provide results that are more limiting. Also, as available, the most up-to-date revision of Regulatory Guide 1.109 dose conversion factors and site-specific environmental transfer factors may be substituted for those currently included and used .in this document. 1.2 RERERENCES 1.2.1 NUREG 0133; Guidance Manual for Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants; October, 1978. 1.2.2 REG. GUIDE 1.109, Rev. 1, October, 1977; Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Compliance with 10 CFR Part 50, Appendix I. 1.2.3 U.S. Code of Tederal Regulations; 10CTR20. 1.2.4 River Bend Environmental Report, OLS. 1.2.5 REG. GUIDE 1.111; Methods for Estimating Atmospheric 1 Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water - Cooled Reactors. 1.2.6 River Bend Station USAR RSP-0008 REV. 3 PAGE 3 0F 110 35 _ - _ - _ - - ____- ___ __ _ __ ___ ___ _ - _ _ - _ _ _ _ _ _ - _ __ _ _ _ _ - _ - - _ _ _ _ _ _ _ _ _ ___-___a
I 2.0 LIQUID ETTLUENT METHODOLOGY 2.1 River Bend Site Description The River Bend Station Updated Safety Analysis Report (USAR) contains l the official description of the site characteristics. The description that follows is a brief summary for dose. calculation purposes: The River Bend Station (RBS) is on a site in West Feliciana Parish, Louisiana, located approximately 24 miles north-northwest of Baton Rouge, Louisiana. This site is just east of the Mississippi River which as used as the source of the RBS major water requirements and which receives the FBS liquid effluents. The reactor is a General Electric boiling water reactor of the BWR-6 or 1972 product line. Containment is of the Mark 3 design, a free-standing cylindrical steel structure surrounded by a reinforced concrete shield building. 2.2 Compliance with 10CFR20 (Liquids) 2.2.1 Requirements In accordance with Technical Specification 3.11.1.1, the concentration of radioactive material released in liquid effluents to Unrestricted Areas (Figure 1) shall be limited to the concentrations specified in 10CFR20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 uCi/ml total activity. The concentration of radionuclides in liquid waste is determined by sampling and analysis in accordance with Technical Specification Table 4.11.1.1-1. 2.2.2 Methodology This section describes the calculational method to be used to determine T , the fraction of 10CFR20 limits of release concentrations of liquid radioactive effluents. 2.2.2.1 General Approach Liquid effluent releases from River Bend Station are discharged through the cooling tower water blowdown which is directed to the Mississippi River. Principal sources of radwaste are from floor drains, phase separators / backwash tank subsystem, sample recovery tanks, and reactor water cleanup (as shown in Figure 4). The liquid radwaste system is operated as a batch system. Only one tank of liquid radwaste is released at a time and is considered a batch. l RSP-0008 REV. 3 PAGE 6 0F 110 36 l
n DR = (3.15 x 10 ) I (Lg + 1.1 Mg ) ( ) () 3.3.1.2.1-2 i = 'I where: DR TB
* "" ##** * * * * ** I I" *#""II'"#*
K g
= The total body dose factor due to gamma emissions for each ider4ff ed noble gas radionuclides (i) in
'3 mrem /sec per ur.i. m . Appendix C.
Lg = Skin dose factor due to beta emissions for each identified nor.e gas radionuclides (i) in mrem /sec per uCi/m . Appendix C. Mg = The air dose factor due to gamma emissions for each identified nob k gas radionuclides (i) in mrad /sec per uCi/m . Appendix C. (X/Q) = The highest calculated annual average relative dispersion factor for any area at or beyond the unrestricted area boundary for all Sectors (sec/m ). Appendix F. hg = The release rate of radionuclides (i) in gaseous effluents from all releases in uCi/sec. 1.1 = Conversion factor for M from y mrad to mrem. 3.15 x 10 = Number of sec/ year. In order to comply with the limits of 10CTR20, DR TB 1500 mrem / year and DR, g 13.000 mrem / year must be met at the most limiting location, at or beyond the site boundary. The radionuclides mix was based upon source terms tabulated in the River Bend Station USAR, Table 11.3-1 and are summarized in Appendix D. The X/Q values utilized in equations 3.3.1.2.1-1 and 3.3.1.2.1-2 are based upcn maxim w long-term annual average (X/Q) in the unrestricted area. Appendix T lists the maximum X/Q values for the RBS release points at the appropriate receptor locations. RSP-0008 REV. 3 PAGE 14 0F 110 l 37 L___--------.
n __ -- _ . l- - L=g The skin dose factor due to beta emissions from noble gas radionuclides (i) (mrem /sec per uCi/m ) from Appendix C, Table C-1. 1.1 = Average ratio of tissue to air energy absorption coefficients. S = 0.7, attenuation factor accounting for shielding provided by F residential structures for maximally exposed individual. 3.4.1.3 Simplified Approach A single effective gamma air dose factor (M,ff) and beta air dose factor (N,ff) have been derived, which are representative of the radionuclides abundances and corresponding dose contributions that are projected in the RBS USAR. (See Appendix C for a detailed e:tplanation and evaluation of M gf and N,ff). The values of M,ff and N,ff which l have been derived from the projected radioactive noble gas effluents are: Radwaste Building Exhaust Duct:
~
= mrad-m /uCi-sec M,ff 8.07 x 10
~
3 N,ff = 7.40 x 10 mrad-m /uCi-sec Main Plant Exhaust Duct and Fuel Building Exhaust Duct:
~ 3 M,ff = 5.96 x 10 mrad-m /uCi-sec
~
N,ff = 8.99 x 10 mrad-m /uC1-sec NOTE The Mg end N,ff f actors should only be ussi if the actual eff hant is similar to that described in Appendix D or similar to the noble gas isotopic wivture dcacribed in the previas Femi.tnnus) Effirco, hpar.t . (Reference 1.2.19) b e 1 RFP-0008 REV. 3 PAGE 30 0F 110 38 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ J
=
11 -. EXPECTED RELEASE OF RADI0 ACTIVE NOBLE GASES IN GASEOUS EFFLUENTS FROM RIVER BEND STATION USAR*
' Containment Building ** Radwaste Building Nuclide -C1/yr Fraction C1/yr Fraction Kr-83m 4.7E (-2) 1.07E (-5) <1 ---
Kr-85m 218 0.050 <1 --- Kr-85 210 0.048 <1 --- Kr-87 14.2 0.003 <1 --- Kr-88 47.2 0.011 <1 --- Kr-89' 118 0.027 29 .03
^f.
Xe-131m 21 -0.005 <1 --- Xe-133m 6.6E (-2) 1.504E (-5) <1 --- Xe-133 2,340 0.533 220 .19 Xe-135 693 0.158 280 24 Xe-135m 140 0.032 530 46 Xe-137 380 0.087 83 .07 Xe-138 208 0.047 2' 1.75E (-3) 4,389. 1.0000 1,144 .99
- RBS USAR Table 11.3-1
** Containment Building contains releases from Fuel Building 3,
RSP-0008 REV. 3 PAGE 68 0F 110 39 __________-_-_w
I I l 1 I l l ATTACHMENT - 1 l CDCM/ PROCEDURE REVISION SHEET l NO. 89-02 l l l l l l DESCRIBE ~~HE INFORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR AND 1 DESCRIPTION OF THE CHANGE (S) MADE TO THE ODCM: l I Page 5 of 110, Section 1.3,5 i l g I I l Addr and/ to first sentence to read... installed to reduce gaseous radiciodings l and/or radioactive material in particulate form... l I l l Grammatical correction to make definition more accurate. g i I I I I I I I I _ l l l 1 l l l I I I I I I l l l - l l l I I l l l l l COMMENTS: l l l l l 1 I i l I I I i l WILL THIS CHANGE REDUCE THE ACCURACY OR RELIA 21LITY OF DOEE CALCULATIONS OR l l SETPolNT DETEPEINATIONS (TECHNICAL SPECIFICATION 6.14)? YES FO X l . I 1 i l WIIJ TMDi CHWGE REQUIED REVISION 70 INER TIER IMPLEdraTING PROCEDr?ES OR l l UJMFUTER PROGRAtt. yES ho X l 3 ! i TFESE CHAFGES MAVE BErN REVIEVED AND FOUND ACCEPTABLE,' PURSUANT TO TECHNICAL l { SPECIFICATION 6.S.2. s l
) j 1
i REVIEVED, MIO~d)GICAL IEGIfEERING SUTE2 VISOR: Q L'" _ s- T / Adf/f 9
'DATE __ l V) / l l REVIEWED, f;IRECTCR - RADIOLOGICAL PROGRAMS: ,/ [/y -
/4'/87 l I /
/ f DATE l I
1 1 I I I il l I i { l ATTACHMENT l PAGE 1 ll l l l l 1 1 0F 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l l l ll I I I ' 40
.-____________d
l 1 l } I l ATTACHMEhT - 1 l l ODCM/ PROCEDURE REVISION SHEET NO. 84-03 l 1 l I l I l DESCRIBE TE IhTORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR AND l DESCRIPTION OF THE CHANGE (S) MADE TO THE ODCM: l 1 l p t, n c s nr 110. section 1.3.5 I 1 l l nrenvn. /or in first sentence to read... gases through charcoal l l HEPA filters... absorbers ana l l I l RBS filter trains are such that airflow must pass through bot.h charcoal I g l absorbers and HEPA filters. There is no mechanism for an either or situation.g I 1 l I I l I I I I I ! l l l I I I I I I l l l l l l l l 1 1 I I l COMMENTS: l l I I I I I I I I I I I I l VILL THIS CHANGE REDUCE 'IE ACCURACY OR RELIABILIIT OF DOSE CALCULATIONS OR l l SETPOINT DETERMINATIONS (TECENIC,iL SPECIFICATION 6.14)? YF.S _ NO X l I 1
! VILL THIS CHAME RIQt/IFID REVISION 70 LC(ER TIER IMPLEENf1NG FROCECRES OP l l COMPUTER PROGR!AS- TE$ NO X l 1
. I
! TEESE C?.A!ES HA7E ITIR EVIEWED RM FOUND ACCEPTA37E, P'JRSUAhT TO TECEhICAL l 1 SPECIn CATIGN 6.5.2. ti l REVIEWED, RAC10 LOGICAL ENGINEERING SUPERVISOR.: j 2. #C l
~/2//J/P"/__ l G DATE j l
l PJNIEVED, DIICCTIUR I RADICIt#1 CAL PROGR!.MS: [D/' 7tC A / f
/
/ 9/<Elll
'DLTE I I
I I I l l ll l l l l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l l l 11 I I I i 41
i I l i l l ) l i l ATTACHMENT - 1 l ODCM/ PROCEDURE REVISION SHEET NO.89-04 l l l- l l l l DESCRIBE THE IhTCIMATION TO BE CHANGED INCLUDE THE RATIONALE FOR AND A COMPLETE l l DESCRIPTION OF THE CHANGE (S) MADE TO THE ODCM: l l l l Page 5 of 110 Section 1.3.5 l l l l Remove for purpose of removing iodines or particulate trom the gaseous g l exhaust stream l l l l Redundant information g I l I l l l 1 I I I I i 1 l l l l 1 I i l l 1 l 1 I I I I I I - l l l l COMMENTS: l 1 1 I I I I I I I _. - . _ _ _ . _ _ .
. . _ _ .._._ l I
l s i Vill THIS CHANG 7 T'dDECE TEL ACCURACY OR RELIA 3!LITY OF DOSE CALCULATIONS GR l l SLTPOINT DETERhiKtTIONS (""ECHNICAL 2PECIFICATICh 6.14)? YES _ NO X l I l 1 WIII THIS CHANGE REQUIL*D REVISION 79 'LDLTR TIER IMPLEMEhT1'NG FAOCEDURES OR l l LOMPUTER PROGRAMS. YES NQ,_X, l 1 . I l TEI.FE CHANCES HAVE BEEN REVIEWED AND FCUND ACCEPTA LE, PURSUANT TO IICENICAL i l SPECIFICATION 0.5.2. . l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: F Q,gC7~72 2[3//1 1 DATE
#3 f- l REVIEWED, DIRECTOR - RADIOLOGICAL PROGRAMS: M [~
/ //[8j
/ l l / ' DATE l 1 l l l l l ll l l l l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 i REV - 2 l PAGE 110 0F 110 l l l II I I I 42
1.3.3 SITE BOUNDARY - The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee. 1.3.4 UNRESTRICTED AREA - An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY
. access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
1.3.5 VENTILATION EXHAUST TREATMEhT SYSTEM - A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiciodine and/or radioactive material in ! particulate form in effluents by passing ventilation or vent exhaust gases through charcoal abst -bers and HEPA filters prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. 1.4 REQUIRED EQUIPMENT 1.4.1 None 1.5 PRECAUTIONS AND LIMITATIONS 1 5.1 As per Reference 1.2.16, Licensee-initiated changes to the ODCM/ Procedure shall be submitted to the Commission in tbs Semiannual Radioactive Effluent Release Report for the pa.riod in which the change (s) was made effective. ) 1,5.2 No c%tngesO.) shall be mada to the ODCM/ Procedure that will ' reduce the acr.uracy or reliability of dase calculations or ) setpoint determinations. }' l.S.3 Any changefs) shall be recorded on the'CDCM Revision Sheet and made in acc.ordance with Re;erence 1.2.16. 1.6 PREREQUISITES 1.6.1 None RSP-0008 REV. 3 PAGE 5 0F 110 43
i I I 1 l 1 l ATTACHMENT - 1 I l CDCM/ PROCEDURE REVISION SHEET NO.89-06 1 l l I i I I DESCRIBE E E INFORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR A l DESCRIPTION OF THE CHANGE (S) MADE TO THE OL2M: l l I Pape 6 of 110, Section 2.1 second paracraph i I l I l Retnove : "used as" in second sentence I l Gray 1matical correction I l l l 1 I I I ~I 1 l i I I l l l _ l 1 I I I 1 1 I i l i I I I __ l l 1 I I I I I I I COMMEhTS: 1 I I._. _l _ l I ___ _ ,. ..____ ! l _ l { I _ _ __ _ _. . _i i I VILL TNIS Clul4GU RE7)CE TEE ACCWACY OR RELIABILITY OF DDSY CALCCATIONF OR I l SETPGINT DE1TKdINATIO:4 (TIQiNICAI, SPECIFICATION 4.%M YES._,_, XO X l 1 I . l WILL THIE CHAN75 EQUTPED REVISION TO ICJER TIER IlfPIAENTING ?ROCFDUEES OR I i 1 COMPLTG PROGRAMS. YES N0_X ,, l 1 l IEEE CFJNGES HAVE BEEN REVIEWED AND FOUND ACCEPTABTE. PURSW2T TO SCHNICLL i i SFECITICATION 6.5.2. \ l l REVIEVT.D RADIOLOGICAL ENGDGERING EUPERVISOR: /! _ # --- / 2//gry l , 7 DATE I I REVIEVED, DIRECTOR - RADIOLOGICAL PROGRAMS: 4
/
/2 f/d*/ I
#DATE l 1
1 I I I I II I I I I ATTACHMENT I PAGE 1 1I I l l l 1 1 0F 1 II RSP-0008 i REV - 2 I PAGE 110 0F 110 1 I I il i I I 44
l ! I I I I l ATTACHMENT - 1 I ODCM/ PROCEDURE REVISION SHEET NO.89-05 l l l l l l
'l DESCRIBE TE INFORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR AND A COMPLETE I l DESCRIPTION OF THE CHANGE (S) MADE TO THE ODCM: l I
l l Pane 6 of 110,Section 2.1
]
I I l Change: " Final" to updated { l i I I l Change references to updated safety analysis report to reflect new revision l l l l l l l l l l l l l 1 1 I I I I I I I I I I I 1 I l 1 I I I I I l COMMENTS: l l 1 i __ ___ __ _.__ i I _. __ _ . . I I _ _ l I ... - - 1 I I l VIZ THIS CHANGE REDUCI TE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS CR l l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NC X l l l l WILL THIS CHANGE REQUIRED REFIL!ON TO LOWER TIER IMPLE'ENTING PROCEDGRES OR l l COMPUTER PRCfRAMS. YE3_ _ NO X _ l l 1 . I ! I I TESE TIJLNGES HAVE BEEN FIVIEVED Afr0 FZ O ACCEPTABLE, PURSUANT 't0 TECHNICAL l l SPECIFICATION 6.5.2. i l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: /]S f,2/)y/f } l l
] DATE l REVIEWED, DIRECTOR - RADIOLOGICAL PROGRAMS: @/- ^[
?
/ '/
l l / 'DATE I I I I I I I ll l l 1 l ATTACHMENT I PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 i REV - 2 l PAGE 110 0F 110 l l l l1 I I I 45 _ -_______-_________D
2.0 L1' QUID 7 EFFLUENT METHODOLOGY 2,1-' River Bend Site Description g' -The River Bend Station Updated Safety Analysis Report (USAR) contains the' official' description of the site characteristics. 'The description .l
'that follows is'a brief summary for dose calculation purposes:
L . The. River Bend Station (RBS) is on a site in West Feliciana Parish, Louisiana, located approximately 24 miles north-northwest of Baton Rouge, Locisiana.. This site is just east of the Mississippi River which is used as the source of the RBS major water requirements and which receives the RBS liquid effluents.
.The reactor is a General-Electric boiling water reactor of the BWR-6
'or 1972 product line. Containment is of the Mark 3 ' design, a free-standing cylindrical steel structure surrounded by a reinforced concrete shield building.
2.2 Compliance with 10CFR20 (Liquids) 2.2.1 Requirements In accordance with Technical Specification 3.11.1.1 the concentration of radioactive material released in liquid effluents to Unrestricted Areas (Figure' 1) shall be limited to the concentrations specified in 10CFR20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble
~
gases, the concentration shall'be limited to 2 x 10 ' uCf/mi total activity. The concentration of radionuclides in liquid wn'ae is determinei by sampling and ant,1ysis in accordance with Technicc1 Specification Table 4.11.1.bl. 2J.2 Methodology This sectico describes tho calculational method to be used to determine F g, the fractice of 10CTR20 limits cf release concentrations of :,1gu.id radioactive effluents. 2.2.2.1 General Approach
-Liquid effluent releases from River Bend Station are discharged through the cooling tower water blowdown which is directed to the Mississippi River. Principal sources of radwaste are from floor drains, phase separators / backwash tank subsystem, sample recovery tanks, and reactor water cleanup (as shown in Figure 4). The liquid radwaste system is operated as a batch system. Only one tank of liquid radwaste is released at a time and is considered a batch.
RSP-0008 REV. 3 PAGE 6 0F 110 46
. . _ _ . , - _ _ :_ _ . _ . - _ . __. -. -- - _ -.._.-_._.m.____._._____-_..________ . _ _ _ _ _ _ _ - - - . '
l l 1 I I I I ' i l ATTACILMENT - 1 1 ODCM/ PROCEDURE REVISION SHEET NO. 89-07 l l l l l l l DESCRIBE THE INFORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR Ado A COMPLETE l l DESCP:P/ ION OF THE CHANGE (S) MADE TO THE ODCM: l 1 1 I Pace 10 of 110, Section 2.4.2 third paracraph l l l l Removes (Dra) l l l l Updated calcula san method no lonner uses this l 1 l l Removet for a specific radioactive liquid batch release l l l 1 I l Updated calculation method is nc longer based on specific batch releases l l l l l 1 I i l
.I I
I I I I I l - l l l l l l l
} l l COMMENTS: l 1 1 I I l !
I i i l i VILL THIS CHANGE REDUCE THE A^ CURACY OR RELIABILITY OF DCSE CALCULATION 5 OR l 1 SETPOIh7 DEIIRMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO X l 1 l l VILL THIS (2 FAN 3E REQUIRED REVISION TO LOWER TIER IMPLEMENTING PROCEDL7JS OM i l C0!ffUTER FROGRAMS. YLS X NO - l l , I LE, PURSUAhT TO '1ECHNICAL I Il SPECIFICATION TESE CHANGES 6.5.2. HAVE BEEN REVIEWED AND FOUND ACCEPTAi[9 l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: 4~ b ' 7 2//y/D l l DATE I l REVIEVED, DIRE 7iOR - RADIOLOGICAL PROGRAMS: 4& ' W /2 / l / ' DATE l l l l ~ I 1 l ll 1 1 I l ATTACHMEhT l PAGE 1 ll l l l l 1 1 0F 1 ll RSP-0008 i REV - 2 i PAGE 110 0F 110 l l l 11 I i 1 47 _ _ _ _ _ _ _ _ . - - - _ - _ i
l L
-l l {
l -1 l L -ATIACHMEhT
'l l I l ODCM/ PROCEDURE REVISION SHEET NO.89-08 l 1 l l 1' I l DESCRIBE THE 'ISTORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR AND A COMPLETE l l DESCRIPTION OF THE CHANGE (S) MADE TO THE ODCM: l l
l
- j. Pace 10 of 110. Equation 2.4.2-2 l
l l l Remove Ecuation 2.4.2-2 l l l l Updated calculation method no longer uses this equation l 1 - I-1 I l l l 1
-l l l- 1 I
-l
'l l I
I I I I I I I I I I - 1 I I I I I I I i l COMMENTS: , __l 1 - _ . _ _ l I _ ~._ __ l l_ __ _l _ _ . _ _. _____ l 1 - _ -.___ l l I l WILL THIS CHANGE REDUCE TE ACCURACY OR RELIABILITY OF DOSE PECULATIONS OR ( t SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)1 YEE N0_X ,, l
- 1. l l WIII THIS CHANGE REQUIRED REVISION TO LOWER TIER IMPLEMENTING PROCEDURES OR l j COMPUIT.R PROGRAFS. YES X NO l 1 . I I THESE CHANGES HAVE BEEN REVIEWED AND FOUNT ACCEPTABLE, PURSUANT TO TECHNICAL l l SPECIFICATION 6.5.2. \ l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: AE -
/ M/3/P') l
] DATI l REVIEWED, DIRECTOR - RADIOLOGICAL PROGRAMS: // /2 MAE l / 'DATE l l l l 1 I I il l I I l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l l l 11 I I I 48
I 2.4 Determining the Dose for Radioactive Liquid Effluents 2.4.1 Requirements Technical Specification 3.11.1.2 requires the dose or dose commitment to a person offsite due to radioactive material released in liquid effluents be calculated on a cumulative basis at least every 31 days. Dose or dose commitment shall be limited to: a) Less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, during any quarter; and b) Less than or equal to 3 mrems to the total body and less than or equal to 10 mrems to any organ during any calendar year. 2.4.2 Methodology This section provides the methodology to calculate dose to all age groups and organs from all radionuclides identified in the liquid effluents. The method is based on the methodology suggested by Sections 4.3 and 4.3.1 of NUREG-0133, Rev. 1, hovember 1978. The dose factors A for all viable pathways are listed in Appendix B. f The following equation provides a dose calculation to the total body or any organ for a given age group based on actual release conditions. Dg =A g
- At
- Q f 2.4.2-1 DF = Dw n 2.4.2-2 D = I D 1
, TOTAL T i=1 i l where: D TOTAL t = The total dose commitment to the organ (t) due to all releases during the desired time period in mrem. 7 RSP-0008 REV. 3 PAGE 10 OF 110 49 t __ _ _ _ _ _ _ _________________________-_____-_a
F
'I i I l i
r l ATTACHMENT - 1. 1 l l ODCM/ PROCEDURE REVISION SEET NO.89-09 l 1 l l
--l l DESCRIBE TE INFORMATION TO BE CHANGED INCLUDE THE RATIONAII
-l DESCRIPTION OF TE CHANGE (S) MADE TO THE ODCM:
l l l Pane 11 of 110 l g I l 1.) Change definition of DJT to read: I Dose commitment from radionuclides.(1/ g l received by organ (T) of the adult age group during the time period (mrem). l- l 2.) l l Chance definit un ofat to reads- The total time for all baten releases l l that occured in the time period (hrs). l l l 3.) Remove terms: " batch releases" from detinition oi yi. g l I 4.) Change definition of DF to read: I l The total volume oi dilution triat g l occured during the time period (ml). 1 l l 5.) In first paragraph remove: l
" release" replace witti: ' isotope g
I I l 6.) in first narscraoh remover "for releases" l I
.l Move countion no. 2.4. 2-3 to page 10, change i 7.) eque: ion no. to 2.4.2-2 l l chnnee- "D-ni" to "DiT". Chances i=1 to i=1 1 l l _,8.) I Mnve definition of D total to pace 10.
! l 1 l I
l COMMENTS: chnneem ,gde to reflect updated licuid release _ dose cniculation l -chna. l l
, l 1 i
, __. __ 1 i
i _i I l VILL THIS CRANGE REDUCE TE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS l OR
} SETPOINT DETERMINATIONS f. TECHNICAL SPECIFICATION 6.I4)? YESNO X l l
t 1 1 l WILL THIS CHANGE REQUIRED REVISION TO LUliER TIER IMPLEMENTING PROCEDURES l OR f I COMPUTER PROGRAMS.
.; YES_X_ NO l l TESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE,' PURSUANT TO TECHNICAL l l SPECIFICATION 6.S.2.
l l REVIEWED, RADICIDGICAL ENGINEERING SUPERVISOR: %i- -/ M/Ms'9 l DATE l l REVIEWED, DIRECTOR - RADIGIDGICAL PROGRAMS: l 46 _/m
/
/
~DATE hhl l I
I I I I i 11 l l l l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l 1 1 II I I I 5o
l-D it = Dose commitment from radionuclides (i) received by organ. (t) of the adult age group during the time period (mrem). A it = Site related dose commitment factor to the total body or any organ (t) for each identified radionuclides (1). The A values listed in Appendix B are site-related g to RBS (mrem /hr per uCi/ml).
~"
-At = The total time for all batch releases that occurred in the period (hrs).
Q i = The total quantity of nuclide (1) released during the interval at (uCi). D = Thn near field dilution factor. Site specific value is 77.4. DF = The total volume of dilution that occurred during the time l period (ml). The doses associated with each isotope may then be summed to provide' l the cumulative dose over a desired time period (e.g., sum all doses during a 31 day period, calendar quarter, or a year). RSP-0008 REV. 3 PAGE 11 0F 110 51 __________--.----_---._________---___--______----__--_______________---__.----_--_-__--._-----_--------_--_--------------_._-----_----_--Y
i I j l I j l ATTACFM.NT - 1 l ODCM/ PROCEDURE REVISION SHEET NO.89-10 ; I I I ! l
; l
' IiESCRIBE TE INFORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR AND A l DESCRIPTION OF THE CHANGE (S) MADE TO THE ODCM:
l l l Pane 12 of 110, De f init. ion or bru l g I I l Remove: "of age group ta)" l l l Move open parenthesis to next line 1 l l l Clarify definition l l l l 1 I I I I. l I i l l l 1 , 1 I I I I I I I I I I - 1 I l l I I i 1 l l COMMENTS: l _l I ___ _
.__ _l 1
_ _ . _ _ _ _ . _ . _. _ . - _ . _ - ___.- .___ l __. _ __ l l
- . _ _ _ . _ - . _ _ _ . I I __
l i I I VILL THIS CHANGE REDUCE DU. ACCURACY OR RELIABILITY OF DOSE CALCULATIONS OR l l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO X l l l l VILL THIS CHANGE REQUIRED REVISION TO LOWER TIER IMPLEMENTING PROCEDURES OR l l COMPUTER PROGRAMS. YES NO X l l
. l l THESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL l l SPECIFICATION 6.5.2. l l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: %E ==-/2//J/f? l l DATE 7 l l REVIEWED, DIRECTOR - RADIOLOGICAL PROGRAMS: 4 4' /
l #~ 'DATE l 1 I I I t i I II I I I I l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l 1 1 I!_ I I I 52
~
l l l I i ATTACEMENT - 1 l l l ODCM/ PROCEDURE REVISION SHEET NO.89-11 l l l l l l l DESCRIBE THE IhTORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR AND l DESCRIPTION OF THE CHANGE (S) MADE TO THE ODCM: l l l Page 12 of 110, Section 3.1 l l l l Changer Figure 5 to Figure 2 l l 1 l Typographical Corrections I l I 1 i I i i i l I i i l i I I I I I I I I I I I I I i - 1 I I i 1 I I I I l COMMENTS: l l l l l l l I _ l i I I I l VILL TiiIS CHANGE REDUCE THE AL* CURACY OR RELIABILITY OF DOSE CALCULATIONS lOR l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO X l 1 l l WILL THIS CHANGE REQUIRED REVISION TO LOWER TIER IMPLEMENTING PROCEDURES l OR l COMPITTER PROGRAMS. YES NO X l 1 I l THESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL l SPECIFICATION 6.5.2. l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: I 11/ ' ~/ W/7/f7 l
, DAW I !
i l REVIEWED, DIRECTOR - RADI0I4GICAL PROGRAMS: .M w / //!//7 I [" # DATE l I I I I I II I I I l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l j l l II I I I 53 l _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ . _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _l
4 E 2.5- Projecting Dose for Radioactive Liquid Effluents 2.5.1 Requirements 1 1 Technical Specification 3.11.1.3 requires the liquid radwaste treatment system be used to reduce the radioactive materials in liquid wastes prior to their discharge when projected doses due to liquid effluents, to unrestricted areas ( Figure 1 ) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31 day period. 2.5.2 Methodology The following calculational methodology shall be performed at least once per 31 day period: TOTAL t L * * * *
~
PD D L PD = Projected dose commitment (mrem) to organ (t) during the 31 day period from liquid effluents. X = Number of days to date in the current quarter D l 3.0 GASEOUS EFFLUENT METHODOLOGY 3.1 Introduction The River Bend Station charges gaseous effluents through the Main Plant Exhaust Duct, Fuel 8uilding Exhaust Duct, and Radwaste Building Exhaust Duct. The location of these release points in relation to the River Bend site is found in Figure 3. The gaseous effluent streams, radioactivity monitoring points, and effluent discharge points are shown schematically in Figure 2. For purposes of simplicity, Fuel l Building exhaust effluents are included in the Main Plant exhaust duct releases. All gaseous effluent releases from the Radwaste Building Exhaust Duct are assumed to be ground level releases. The Main Plant Exhaust Duct routine releases are treated as a wake split (conditionally elevated) release. 3.2 Data Requirements for Gaseous Effluents For the purpose of estimating offsite radionuclides concentrations and radiation doses, measured radionuclides concentrations in gaseous effluents and in ventilation air exhausted from the station are relied upon. Table 4.11.2.1.2-1 in the Technical Specifications identifies the radionuclides in gaseous discharges for which sampling and analysis is done. RSP-0008 REV. 3 PAGE 12 0F 110 54 , l
l I l i I l ATTACHENT - 1 l l ODCM/ PROCEDURE REVISION SEET NO.89-12 g i i
! l I
ll DESCRIPTION DESCRIBEOFTE INFORMATION TO BE CHANGED INCLUDE TE RATIONAII FOR THE CHANGE (S) MADE TO TE ODCM: l l l Pare 20 of 110 l 1 l l 1.) Equation no. 3.3.2.2-2 l l- Changer to n l l E I l l 1 1-1 -- l 2.) ; l Definition of Keff l l l kemover E K,f. l l 1 l l l l l Keep all formulas and definitions concistant throughout procedure
! l I I I I I I I I 1 1 i - I i _. )
I I I I I i 1 l CO. W S: I 1 l 1 I
! I I I I
_ l I I I I WILL THIS CHANGE REDUCE TE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS l OR l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YESNO X l l 1 l VILL THIS CHANGE REQUIRED REVISION TO IDWE TIER IMPLEMENTING PROCEDURES l OR l COMPUTER PROGRAMS. YES NO X l l
. I l THESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL l
l SPECIFICATION 6.5.2. 7 l l l l DEVIEVED, RADI0I4GICAL ENGINEERING SUPERVISOR: ( h ~/A T b '/MS/Ml l
~ 7 DATE I l REVIEVED, DIRECTOR - RADIOLOGICAL PROGRAMS:
l 4/ 6-t / / l
/ DATE l l
l 1 I I I 11 1 I I l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l 1 1 11 I I I 1 i 55
] Step 1 Determine hTB utilizing one of the following methods: Qg = (3.17 x 10-8) (500) (0.8) 3.3.2.2-1 (5 76) (K,ff) or NOTE The Kg methodology for determining Qg should be used only.if isotopic analyses is available and the relative abundances of noble gas nuclides in the effluent stream are not similar to those listed in Appendix D or not similar to-the noble-gas' isotopic mixture described in the previous Semiannual Effluent Report. (Reference l.2.19)
~0 TB
= 3.17 x 10 (500) 3.3 2.2-2
_ n (X/Q).I (K,) (fg) i=1 where: hTB
- maximum 3CCeptable total release rate of all noble gas radionuclides in the gaseous effluent (uCi/sec).
'(X76) = The highest calculated annual average relative dispersion factor for any area at or beyond the unrestricted area boundary for all Sectors (sec/m ). Appendix F.
K g
= The total whole body dose factor due to gamma emissions from noble gas radionuclides (1) (mrem /sec per uCi/m ) from Appendix C, Table C-1.
f = Fraction of noble gas radionuclides (1) to total f noble gas concentration. K,ff = Effective dose factor 3 (mrem /sec per uCi/m ) from Appendix C, Table C-3.
-8 3.17x10 = Inverse of number of seconds per year in year /sec.
RSP-0008 REV. 3 PAGE 20 0F 110 l.- . 56
1 I l l I l ATTACHMENT - 1 I l ODCM/ PROCEDURE REVISION SHEET NO.89-13 1 l I I I I ll DESCRIPTION DESCRIBEOFTHE INFORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR THE CHANGE (S) MADE TO THE ODCM: I l l Pace 21 of 110, Equation 3.3.2.2-4 l l l Chancer T to b l 1 i=1 l 1 l Keep fomula consistant throughout procedure l l l I i I I i 1 l l 1 I I i 1 1 I I I I I I I I I i - 1 I I I
, i i
I I I l COMMENTS: l l l 1 I I I l l 1 I I l WILL THIS CHANGE REDUCE THE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS ORl l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO X l
- 1
' I l WILL THIS CHANGE REQUIRED REVISION TO LOWER TIER IMPLEMENTING PROCEDURES OR I l COMPUTER PROGRAMS. YES NO X l 1
. I i l THESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL I l l SPECIFICATION 6.5.2.
) l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: AP~ -
N
/ 2[3M7 ]
REVIEVED, DIRECTOR - RADIOLOGICAL PROGRAMS: //
/
/ DATE [
I I I I I I I ll 1 1 I l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 i REV - 2 l PAGE 110 0F 110 l 1 I il I I I 57 {
1 0.8 = Conservative factor to account for changing isotopic inventory. 500 =- Whole body exposure limits of 500 mrem / year. 3.17x10' = Inverse of number of seconds per year in year /sec. 1 Step 2 . Determine Q, utilizing one of the following methods. Q, = (3.17 x 10~0) (3,000) (0.8) 3.3.2.2-3 (576)(L+1.1M),ff or NOTE The -(L + 1.1M)g methodology for determining Q3 should be used only if isotopic analyses is available and the relative abundances of noble-gas nuclides in the effluent stream are not similar to those listed in Appendix D or not similar to the noble gas isotopic mixture described in the previous Semiannual Effluent Report. (Reference 1.2.19)
~0
= (3.17 x 10 L (3,000) 3.3.2.2-4
~(5[d)I [(L g t 1.1Mg )ff }
. i=1 Q,
= the maximum acceptable release rate of all gas radionuclides in the gascous effluent [uCi/sec))
= Total skin dose factor due to emission from noble gas Lg + 1.1Mg radionuclides (i) mrem /sec/uCi/m from Appendix C.
(5 76 ) = The highest calculated annual average relative l dispersion factor for any area at or beyond the L unrestricted area boundary for all Sectors 3 (sec/m ). Appendix F. f l RSP-0008 REY. 3 PAGE 21 0F 110 58
l 1 i 1 l l ATTACEMEST - 1 l l CDCM/ PROCEDURE REVISION SEET l NO. n o _ u l I
~~
i l I ll DESCRIPTION DESCRIBEOF THE INFORMATION TO BE CHANGED INCLUDE TE RATIONALE FOR Ah THE CHANGE (S) MADE TO TE ODCM: l 1 1 l Pnce ?? nf 110. Definition nf (141.1M)eff l l l l Remnver T (T 1+41.1M-)1 (f ) L l l i l I l I Keen definfrinns consistant throuchout nrocedure l I I I I I I I I I I I I I I I i I I l i l I I I - I 1 I -- __ I I I l i l I l CO.TfENTS:
<l l
I
.I I
I I I - 1 I I I i l WILL THIS CHANGE REDUCE "nIE ACCURACY OR RELIABILITY OF DOSE CALCULATIONSl OR l l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO y l l l l l l WILL THIS CHANGE REQUIRED REVISION TO LOWER TIER IMPILMENTING PROCEDURES OR l l l COMPUTER PROGRAMS. YES NO y l l 1
. I I TESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL l
l SPECIFICATION 6.5.2. 1 REVIEWED, RADIGIDGICAL ENGINEERING SUPERVISOR: /h Y ~ ^ / 2 [ y /f> S l l
~
DATE l REVIEWED, DIRECTOR - RADIOLOGICAL PROGRAMS: ild 47 l ,/ P
/2 9[#7 l
'DATt l 1
I I I I I il i I I l ATTACHMENT l PAGE 1 l1 l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l I I 11 I I I 59 L__---. __ __ _ - . - - J
_ - mm-= .
= - - - - - - - - --- - --, - -
- - = --
j-{q -;i.< = ' ' , 4-M v': ; . . y. I :lI , c
;(L+1.1M);ff = Effective total skin dose factor 3
0
'(mrem /sec/pci/m ) from Appendix C-1 Table C-4 53000. ' = ' ' Skin' exposure limit of 3000 mrem / year
=
3.17x10 Inverse of number'of-seconds per year in year /sec. Step 3 e i
' Select.the'lowerofthehvaluas(hTB # s) obtained in Step 1 and.-
Step 2.
' NOTE i .=
Actual alarm setpoint in'the data-base may be:
. modified to account for loop accuracy.
Step 4
- Multiply the Q value. selected in Step 3 by 0.33. =By multiplying the Q-value by a factor.of 0.33, the allowable operating.setpoints will;be administrative 1y controlled to allocate one-third (1/3) of the total
' allowable release rate to ecch of the release points.- The resu'J tant product will;be the actual ODCM release rate HIGH ALARM setpoint. fo:-
the' appropriate WRGM Monitor. ii. Particulate and Gas Monitor-(P&G) (gas channel only). Step 1 Perform Steps 1 through 3 of Section 3.3.2.2a.1 above Step 2 Det- ae C,-(the maximum acceptable total radioactivity concentration of all noble gases radionuclides for all release points in the gaseous effluent [uCi/ce]):
~3 3.3.2.2-5 C,= (2.12 x 10 I
where: 2.12 x 10' = Unit conversion factor to convert uCi/sec/cfm u to uCi/cc. h = lower of the two h values, hTB # s' F = The maximum acceptable effluent flow rate at the point of release based on design flow rates (cfm) RSP-0008 REV. 3 PAGE 22 0F 110 f 60 _w .-
_ 1 l i 1 , I I I ! l ATTACHEhT - 1 l ODCM/ PROCEDURE REVISION SEET NO.89-15 j ! I I I I I l DESCRIBE . IhTORMATION TO BE CHANGED INCLUDE TE RATIONALE FOR AND A COMPLETE l DESCRIPTION F TE CHANGE (S) MADE TO THE ODCM: 1 1_ l l Pace 23 of 110. Equation 3.3.2.2 7 l I n l Chanter T to r i I 1 i=1 l l l 1 1 I I I i 1 l Keep equation consistant throughout procedure l 1 1 I I I l l 1 I I I I I I l l __ l I
!_ _ l I _ . ___ .
._ . _ I I .
_ _ . . . _ __ l l I I I I I l COMENTS: l l l 1 I _i i I I l l l 1 l VILL THIS C(ANGE REDUCE THE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS OR l l SETPOINT DE ' TERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO X l 1 l l VILL THIS CHANur. REQUIRED REVISION TO LOWER TIER IMPLEMENTING PROCEDURES OR l l COMPUTER PROGRAMS. YES NO X l l
. I l TESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL l l SPECIFICATION 6.5.2. g i
I ' l l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: b_ b"- -y 2//Md9 l REVIEWED, DIRECTOR - RADIOLOGICAL PROGRAMS: ,[ / / l / ' DATE l l l I i i I ll l 1 l l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l I I 11 I I l 61
b NOTE Actual alarm setpoint in the data-base may be modified to account for loop accuracy. Step 3 Multiply the C , value determined in Step 2 by 0.33. By multiplying the C, value by a factor of 0.33, the allowable operating setpoints will be administrative 1y controlled to allocate one-third (1/3) of the total allowable release to each of the release points. The resultant product will be the actual ODCM activity concentration HIGH ALARM setpoint for the appropriate P&G monitor gas channel,
- b. ALERT Setpoint Determination (Reference 1.2.12)
- 1. Wide Range Gas Monitor (WRGM)
Step 1 Determine GQ -A utilizing ne f the following methods: t. G-A
= . x 0.8) 3.3.2.2-6
) ("eff)
OR NOTE The M g methodologyfordeterminingh s ud G-A be used only if isotopic analyses is available and the relative abundances of noble gas nuclides in the effluent stream are not similar to those listed in Appendix D or not similar to the noble gas isotopic mixture described in the previous Semiannual Effluent Report. (Reference 1.2.19)
~
= (1.26 x 10 ) (5) 3.3.2.2-7 G-A n j (5[Q)I Mg ff j i=1 i
Where: J l Q = maximum acceptable total release rate of all noble G-A gas radionuclides in the gaseous effluent [uCi/sec] I RSP-0008 REY. 3 PAGE 23 0F 110 62 j
\
l l 1- I l ) i : l ATTACHMENT - 1 l ODCM/ PROCEDURE REVISION SEET NO.89-16 '
'I I l
I i 1 l DESCRIBE TE INFORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR AND A COM i
'I DESCRIPTION OF TE CHANGE (S) MADE TO THE ODCM: l 1
- j. Pnce 24 of 110 Equation 3.3.2.2-9 l l
l -n l l Chancer E to I l l 1 i=1 l l 1 I Keep eaustion consistant throughout procedure l l l l l
.I I I
I I I I I I I 1 I I
. _. I I
I I l
'I .. ___ l 1 -
1 I _____ . .
. . _ . _ _ I I _ , . -_
_. 1 l __ l 1 l l COMMENTS: l _{ l 1 i
-- I 1
I __ l l I l l' I l VILL THIS CHANGE REDUCE THE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS OR l 1 SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO X l 1 I I WILL THIS CHANGE REQUIRED REVISION TO LOWER TIER IMPIL*fENTING PROCEDURES OR l l COMPUTER PROGRAMS. YES NO X l I
, I l TESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL I I SPECIFICATION 6.5.2.
]i1% 8 I
I ' I l REVIEWED, RADIGI4GICAL ENGINEERING SUPERVISOR: ~
~/ 245/g7 l l DATE I
! REVIEWED, DIRECTOR-RADIOLOGICALPROGRAMS:
l 4R Y
/ ' DATE l l
l I m, I I I i i l I l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l l I II I I I 63
F (X3) = The highest calculated annual average relative dispersion factor for any area at or beyond the unrestricted area boundary for all Sectors (sec/m ). Appendix F.
.M = ffective gamma air dose factor (mrad-m /uCi-sec).
df Appendix C, Table C-5. 5 = 5 mrads/ quarter (92 days) gamma air dose limit at the unrestricted area boundary. Mg = The gamma air dose factor for radioactive noble gas nuclide (i) in mrad-m /uCi-sec (Appendix C). fg = The fractional abundance of noble gas radionuclides i 1.26 x 10 = Inverse of number of seconds per quarter in quarters /second. 0.8 = Conservative factor to account for changing isotopic inventory. Sjen 2 Determine36,3 utilinng me of the follwing methods: B-A
= 0. 6 x 10 b 00) (0.8) 3.3,2.2-8 (576)(N,gf) or NOTE The Ng methodology for determiningB Q -A sad be used only if isotopic analyses is available the relative abundances of noble gas nuclides in the effluent stream are not similar to those listed in Appendix D or the previous Semiannual Effluent Report. (Reference 1.2.19)
B-A
= (. x O b (10) 3.3.2.2-9 (5/6) (Nj ) (f )
i=1 RSP-0008 REV. 3 PAGE 24 0F 110 64 _ n____ _ _ _ _ _ _ _ _ _ _
'l i 1, 1 l j' ATIACHENT - 1 i i CDCM/ PROCEDURE REVISION SEET NO. 89-17 l 1 1 I
I I l DESCRIBE TE IhTORMATION TO BE CHANGED INCLUDE TE RATIONAII TOR AND A COM l DESCRIPTION OF THE CHANGE (S) MADE TO TE ODCM: l l
-l h 31 of 110. First Paragraph 1 l
l l 1 Remove: "(IOi)" l l l Unnecessary term in paragraph 1 l I I I I I I i I i I L I I I I I
, I I
I I i
.- 1 i __ _
i i ___ _ l 1 -. .. _ .._ _ I I .._ _ _._._ 1 1-- ._ _- =_- -- I
} - I I
I co m TS: _ .__ l I i __
-_i 1
3 _ _ . . ____ l
.I ___ _ l 1
._ _l l
l .YILL THIS CHANGE REDUCE TE ACCURAC7 OR RELIABILITY OF DOSE CALCULATIONS OR l l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? TES NO X l i I I WILL THIS CHANGE REQUIRED REVISION 'It) LOWER TIER IMPIIMEhTING PROCEDURES OR l l COMPETTER PROGRAMS. YES NO X l l i
. I i TESE CHANGES HAVE BEEN REVIEWED AND TODND ACCEPTABII, PURSUAhT TO TECHNICAL l
l SPECIFICATION 6.S.2. \ l
! REVIEWED,RADIOLOGICALENGINEERINGSUPERVISOR: A, [ --
/ 24,4 7 I i DAtt I
! REVIEWED, DIRECMR - RADIOLOGICAL PROGRAMS: 74 / [ /
i e 'DrIt i I , I I I I I 11 1 I l i ATTACHMENT l PAGE 1 1I l l l l 1 1 OF 1 ll RSP-0008 i REV - 2 l PAGE 110 OT 110 l i I il .! I I 65 m________________--_-_ -- -
The effective gamma air dose factor may be used'in conjunction with the total noble gas release to simplify the dose evaluation and to verify that the' cumulative gamma and beta air dose is within the: l equivalence of the limits of Technical Specification 3.11.2.2. To compensate for any unexpected variability.in the radionuclides j distribution, a conservatism fattor of 0.8 is introduced into the calculation. The simplified equation is: (H,ff) (X/Q) n D " Ni * 'I' 'I Gamma-Air 0.8 i=1 (N,ff) (X/Q) n D
- 01 ' ' '
~
Beta-Air == 0.8 i=1 l 3.4.2 _ Determining the Radioiodine and 8 Day Particulate Dose to Any Organ from Cumulative Releases 3.4.2.1 Requirements Technical Specification 3.11.2.3 states that the dose to a Member of the Public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with' half-lives greater than 8 days in gaseous . j effluents released, from each reactor unit, to areas at and beyond the site boundery shall be limited to the following
- a. During any calendar quarter: less than or equal to 7.5 mrem to any organ; and i
- b. During.any calendar year: less than or equal to 15 mrem to any I organ.
1 l l l l ) l l l l RSP-0008 REV. 3 PAGE 31 0F 110 66 ! I i L --- - - - - - _ _ _ _ _ _ _ _.------_____ ____ __________9
1 I i l I ATTACHMENT - 1 I l l ODCM/ PROCEDURE REVISION SEET NO. 89-18 l g I l l i 1 l DESCRIBE TE INFORMATION TO BE CHANGED INCLUDE TE RATIONALE FOR AND AI l DESCRIPTION OF TE CHANGE (S) MADE TO THE ODCM: l l j Page 32 of 110 Equation 3.4.2.2-4 ._ _ l . g l Add subscript total to D to read: D total l l g I
.__ l I
Typographical correction l l l 1 I I l l l l i I I I l 1
. I I
I I I l l l I I I I - 1 I l i I I I I I l C0!ffENTS: l l l l l l l l t I I I l WIII THIS CHANGE REDUCE TE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS OR l l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO X j i I l WILL THIS CHANGE REQUIRED REVISION TO LOVER TIER IMPLEMENTING PROCEDURES OR l l COMPUTER PROGRAMS, YES NO X l 1 , I l TESE CHAWEE HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL 1 l SPECIFICATION 6.5.2. l l (' I l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR:
~
[- A - i-2//p/97 l l DATE l l l REVIEWED, DIRECTOR - RADIOLOGICAL PROGRAMS: 4(2 / / // /l
~/
' l 'DA*It l I I I I I I il l I i l ATTACHME!ir l PAGE 1 ll l l ! I 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l l 1 11 I I I 67
k y .. 3.',.2.2 Methodology 1 The following calculational method is provided for determining the critical organ dose doe to releases of radio' iodines.(1131 I133), trituim and particulate. Ir. is. based'on Section 5.3.1 of NUREG-0133', Rev. :1, November .1978. The equation can be used.for any age group. provided.that the appropriate dose factors are used and the. total dose reflacts only those pathways that are applicable to the age group. The symbol (X/Q)D; represents a depleted (X/Q) which is different from the noble gas (X/Q)-in that (X/Q)g takes into account the lose of radioiodines (1-131, I-133), 8' day particulate, and tritium from the plume as the semi-infinite cloud travels.over a given distance. The dispersion factor (D/Q) represents the rate of fallout from the cloud that affects a square meter of ground at various distances from the site. 'The total dose to an organ can then be determined by summing the pathways that apply to the receptor in the sector. The equations are: Inhalation Pathway: n D I&8DPt = (3.1 x 10 )I (Rit) !O)D IEi ) 3 * ~ i=1 Ground Plane Pathway: n D =(. x 0 )I (Rgt) (D/Q) (Qg ) 3.4.2.2-2 I68DPt i=1 Contaminated Forage / Cow / Milk Pathway: n D I&BDPt
= (3.17 x 10~0) I (Rh) (D/Q) (Qg ) 3.4.2.2-3 i=1 Total Dose:
-D Total = I D 3'4"2'2~4 I&BDPt z=1 IMPORTANT When calculating organ. doses due to the release of C-14 and/or tritium (H-3), X/Q values (not D/Q values) must be used for cow milk, goat milk, meat and vegetation pathway calculations.
RSP-0008 REV. 3 PAGE 32 0F 110 68 L _m __ u___. __ _ . . . _ _ _ _ . _ _ _ _ _ _ ._
-.W
l l 1 l I l ATTACHENT - 1 I l ODCM/ PROCEDURE REVISION SEET l NO.80-1o l l 1 l I l DESCRIBE THE INFORMATION TO BE CHANGED INCLUDE TE RATIONALE FOR l DESCRIPTION OF THE CHANGE (S) MADE TO TE ODCM: 1 l l Pare 33 of 110 I l l I Changer D" t o D To t a l i 1 l l Typographical correction l 1 l I 1 I I I I i I 1 i I l I I I 1 I I I I I I I I I I - 1 I i i 1 I l i I l C0&fENTS: l I I I I I I I I I I i i I WILL THIS CHANGE REDUCE THE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS lOR l SEIPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.I4)? YES NO X l l l l l VILL THIS CHANGE REQUIRED REVISION TO IEw'ER TIER IMPIIMENTING PROCEDURES ORl 1 l COMPUTER PROGRAMS. YES NO X l I
, I l TESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUAhT TO TECHNICAL l l SPECIFICATION 6.5.2. ( l I '-
l l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: - F4 M -T-24f87 ll t DATE l
! REVIEWED,DIRECIVR-RADIOLOGICALPROGRAMS: #// A MM l l / / DATE I I
I i 1 I I il l I l l ATTACHMENT l PAGE 1 ll l l l
.l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l l l 1I i 1 1 69 l l
p where: D 168DPt = Dose in mrem to the organ (t) of a specified age group from radioiodines (I-131, I-133), tritium and 8 day particulate due to a particular pathway. z = All the applical,le pathways for the age grcup of interest. D Total = Total dose in mrem to the organ (t) of a specified age group from gaseous radioiodine (I-131, I-133), tritium and particulate effluents, summed over all applicable pathways (z).
~0 =
3.17 x 10 The inverse of the number of seconds per year (years /sec). R it = The dose factor for nuclide (i) for pathway (z) to organ (t) of the specified age group. The units are either:
~
3 mrem-m for pathways using (X/Q)D yr-uCi or
~
2 mrem-m -sec for pathways using (D/Q) yr-uCi (See Appendix I.) (X/Q) D = The depleted (X/Q) value for a specific location where the receptor is located (sec/m ). (See Appendix F.) Note: No credit is taken for depletion and decay. (X/Q)p = (X/Q) (5) = The deposition value for a specific location where the
~
receptor is located (m ). (See Appendix F.) RSP-0008 REV. 3 PAGE 33 0F 110 70
t I j 1 l l l ATTACFN.NT - 1 I [ CDCM/ PROCEDURE REVISION SEET NO.89- 20 g i l I I l l DESCRIBE THE INFORMATION TO BE CHANGED INCLUDE TE RATIONALE FOR l DESCRIPTION OF THE CHANGE (S) MADE TO THE ODCM: l l I l Pane 36 of 110 De fini tion s l l l 1.) Deleter "R" from DRT+8DPT 1 j i 2.) Adds I l " Annual" to .efinition remove: " rate" l 1 I 1 i j Clarifv definitions, and correct typographical error I I I I I I I I l l I I i I I l I I I I I i - 1 I I I I I I i I l COMMENTS: l l l : I I I I I I I I I I l VILL THIS CHANGE REDUCE IIIE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS OR I l SETPOINT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO x ! I l l WILL THIS CHANGE REQUIRED REVISION TO IDWER TIER IMPLEMENTING PROCEDURES l OR l COMPUTER PROGRAMS. YES NO x l l
. I l THESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL l 1 SPECIFICATION 6.5.2.
( l l l REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: I 'l N7 l DATE l 1 REVIEWED, DIRECTOR - RADIOLOGICAL PROGRAMS: d4 / O d/M4 l 1 DATE fl I I I I I ll l l 1 l ATTACHMENT l PAGE 1 ll l l l l 1 1 0F 1 ll RSP-0008 i REV - 2 l PAGE 110 0F 110 l l l Il i I I 71
[ l Fresh Fruits and Vegetables: n D ygggp = (3.17 x 10-8) I (Rp)' ( ) (Q ) 3.4.2.5-6 i=1 Total Dose: n
'U t
- U I&SDPt 3.4.2.5-7 z=1 -
where: D I&8DPt = Annual dose to the organ (t ) l for the age group of interest from radioiodines (I-131 I-133), tritium and 8-day particulate via the pathway of interest in mrem /yr. For radiciodines (I-131, I-133), the entire cource term was used to calculate these values. z = All the applicable pathways for the age group of interest. Qg = The number of uCi of nuclide (i) released during the year of interest. R = e dose factor for nuclide (i) for organ it (t) for the pathway specified [ units vary with pathway). For tritium, a site-specific absolute humidity (H) value 3 of 12.9 gir/m was used for calculation. (See Appendix I.) (D3) = A long-term relative deposition value for elevated and ground level releases. A factor with units of m' which describes the deposition of particulate matter from a plume at a point downrange from the source. Actual meteorological data and sector wind frequency distribution will be used to determine annual average D/Q for the year of interest. RSP-0008 REV. 3 PAGE 36 0F 110 72 _ = _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ - -
l I l l 1 ATTACHMENT - 1 I
- l. l ODCM/ PROCEDURE REVISION SEET NO. 89-21 l 1 I l l 1
l DESCRIBE TE IhTORMATION TO BE CHANGED INCLUDE THE RATIONALE FOR ANT l DESCRIPTION OF TE CHANGE (S) MADE TO THE ODCM: l
\
l Pape 62 and 63 of 110 l l 1 1.) Remove terms Equation C-1 thru C-4 l l l l
'] 2.) Replace with equation no. consistant with rest 01 procecure g
l l ie C.2-1 C.2-2 l l C.2-3 l l C.2-4 l l l 1 D l l 3.) Type: E for all equations l l i=1 l l l 1 I I I I I I - _ l I i 1 - 1 I I I I I I i l l COMMENTS: l 1 l 1 1 I i I l I I I 1 l WILL THIS CHANGE REDUCE TE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS OR l
] SETPOINT DETERMINATIONS (TECENICAL SPECIFICATION 6.I4)? YES NO X _ l 1
1 I WILL THIS CHANGE REQUIRED REVISION IV LOVER TIER IMPLEMENTING PROCEDURES ORI l COMPUTER PROGRAMS. YES NO X l l
. 1 I THESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTA , PURSUANT TO TECHNICAL l l SPECIFICATION 6.5.2. l l
- ! REVIEWED,RADIOLOGICALENGINEERINGSUPERVISOR: b/4:# M / 2 // g /> 9 l
! DATE I l REVIEWED, DIRECTOR-RADIOLOGICALPROGRAMS: 24 / eA I l e
/ VMEl
' DATE I l
l l 1 : 1 I ll l l I j l ATTACHMENT l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l l l 11. I I I 73
L TABLE C-2 TEC5NICAL BASES FOR EFFECTIVE DOSE FACTORS-The evaluation of-doses due to releases'of. radioactive material to the atmosphere can be simplified by the use of effective dose transfer
- factors'instead'of using: dose factors which are radionuclides specific.
l- These effective factors, which are based on the typical radionuclides distribution in the releases,'can be applied tc. the total radioactivity released to approximate the dose in'the environment, 1 e.. instead of having to. sum the isotopic distribution multiplied by the isotope specific dose factor only a single multiplication (K,ff, (L + 3.1M),ff, M,ff, or N,ff times the total quantity of radioactive l ' material released) would be needed. This approach provides a reasonable estimate of'the actual dose while eliminating the need for a detailed calculational: technique. Use of effective dose factors should only be used if isotopic analyses are not available (i.e., prior to initial criticality), if the relative abundances of the noble gas isotopic mixture are similar to
- those listed in Appendix D or if the relative abundances of the noble gas isotopic mixture are similar to those listed in the previous Seminannual Radioactive Effluent Release Report. (Reference 1.2.19)
Determination of Effective Dose Factors The effective dose transfer factors should be based on past operating data. The radioactive effluent distribution for the-past years.can be used to derive single effective factors by the following equations: n K,ff =I K . f g C.2-1 f i=1 where: K,ff = The effective total body dose factor due to gamma emissions from all noble gases released. K = The total body dose factor due to gamma emissions from f each noble gas radionuclides "i" released. f = The fractional abundance of noble gas radionuclides "1" of the total noble gas radionuclides. RSP-0008 REV. 3 PAGE 62 0F 110 74
a 4 TABLE C-2 TECHNICAL BASES FOR EFFECTIVE DOSE FACTORS (Continued) n (L + 1.1 M) ,ff = I (Lg + 1.1 Mf ) . fg C.2-2 i=1 where:
= . (L + 1.1. M),ff The effective skin dose factor due to beta and gamma emissions from all noble gases released.
(Lg + 1.1 Mf ) = The skin dose factor due to beta and gamma emissions from each noble gas radionuclides "i" released, n M,ff = I Mg . f g C.2-3 i=1 where:
=
M,ff The effective air dose factor due to gamma emissions from all noble gases released. Mg = The air dose factor due to gamma emissions from each noble gas radionuclides "i" released. n N,ff =I Ng . f C.2-4 f i=1 where N,ff = The effective air dose factor due to beta emissions from all noble gases released. Ng = The air dose factor due to beta emissions from each noble gas radionuclides "i". To provide an additional degree of conservatism, a factor of 0.8 is introduced into the dose calculation process when the effective dose transfer factor is used. This added conservatism provides additional assurance that the evaluation of dose by the use of a single effective l factor will not significantly under-estimate any actual dose in the l environment. l l Each year the dose factors should be determined and the average annual values be used. l RSP-0008 REV. 3 PAGE 63 0F 110 75
i l I 1 I l ATTACIDT.NT - 1 1 l CDCM/ PROCEDURE REVISION SHEET NO.89-22 1 l I l l I l DESCRIBE THE INFORMATION TO BE CHANGED INCLUDE TE RATIONALE FOR A l DESCRIPTION OF THE CHANGE (S) MADE TO TE ODCM: I 1 i l Pancs 85 throuch 88 of 110 I l I
! Tables I-2 throuch I-4 l
l _ l l ChanPer "uCf /m" t o uCf /m" l l I 1 Typographical Error I l l l l 1 I I I I I I I I I I I I I l _. I i I I i l i I I I I I I I COMMENTS: l l l l l 1 l l 1 1 I I ! I l WILL THIS CHANGE ELDUCE TE ACCURACY (G RELIABILITY OF DOSE CALCULATIONS OR l l SETPOINT DETERMINATIONS (TECHNICAL SPECIH CATION 6.14)? YES NO y l l 1 l WILL THIS CHANGE REQUIRED REVISION TO LOWER TIER IMPLEMENTING PROCEDURES OR l l COMPUTER PROGRAMS. YES NO y l 1
, I l TTESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCE , PURSUANT TO TECHNICAL l l SPECIFICATION 6.5.2. I l
REVIEWED, FADI0 LOGICAL ENGINEERING SUPERVISOR: l% "8 '- = / 2/1/87
$ DATE l l REVIEWED, DIRECTOR - RADI0IDGICAL PROGRAMS: 7 8
/
l r ~~ 'DATE l 1 I I I l l 11 I I I l ATTACHMEKr l PAGE 1 ll l l l l 1 l OF 1 ll RSP-0008 l REV - 2 l PAGE 110 0F 110 l l l 11 ! I I 76
m O
.C. +
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'W=
% C N O + O m m N+ONw +0 O C OwoomOOwO +
3 +cC+@ CON @ w O
- .e wc2WO@NNw @ C. e Xm4JCNAND O .=
wecemmNwmOO N. -- e e . o e e a w O w a Q O c h $
= =
U V . W
@ += N m @ Q 0 O@O > <
O '6 +C+ - 4 A w+w .a m C no w @ e O >= = 0 000000000 m. m. N. W
'w Ns* 6 M
b $ e b
- - > , s O
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= EC O
- NV OC O + 6
- == 0 CON @CW D E Q ac O @ nan 3ec@ w 3 2 m NN *@NNN W G D CN@NN e E 6 S O @ o O N N *= @ e e" - O. m O G ~
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o w
*= a= h Z e N O
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>= >
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2Nmm@@ eN@N o C- O m e - C osO.1 # a c N e . c U Q +8 a 3 u w e a 6 e m m e O
@ e b O 8 W e
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+ e 6 O
e e o w w e A O OwCO - e e g c' so 880223282 a:NNeomeN o s R " N 30 C NomNNemON e .a 6 4 O O O m m m* m # mea ne= O. C 8 L U - e w G D 6 6 m E O w m E ** m e e E U O eo E 0 e O Ca CO e- a g 6 OOOONO COOOW 6 V
*= Den @ COO 20@@@ C
-> e @ e E = m 3 2mmaNNDOO#eO N 80 s= O NDNNN N. m W Q N N A m > *= C N o- *= * = ee #
m .= Z = a O @ U
== 6
>= m 6 4 % CO O U O w e= D CO "O C O O + G %
( .J c @C@C+OOce@Ow e CO O @CANwee0@NSE@ N e EO N G @ N *= N N e t O O N == se
-4 . @*2C@N#Naam. O -
N . v= 3 C ll il 3 A S D C &
>O e o -
4E V X P. O 3Q += 3Cr>Omm2**m D C - G ..
== W Q b@@cm=eennm=m e e e e a a a e e= == a w (Q 3 EOKEEMM4weee a w 44 E E U N M W M O C60 - - 2 O O E
77
,, f . !'
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+
s- 0 0 0 0 0 0 0 0 0 0,0.w % A. N. s:
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@ + O-C nCO++0C+O
++wwOOwO +
I .
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. (C@t t.s 3 = 2 @ :3mm N243NOe N O e
=
- w. e. .g . N. . - - N. @ c o s- .
O w a 0- -O- @ E & ED . ( c Z' V U V w. M = ' h@m. W. o O OCC -> ( O 6 + + + . - M 4. A W e.J W C man e
-m.
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'W e
U w ah N m @ #2m X
- EC O0C
= ND CC +++ 6 w == 0. O '0C wwW C E O aC N. O NCeOme% t 3 % 2 D3 c@wN M N '@ P- O N e . * *
- g. b
- C c Q O m m fu c - N w O. m O G e e s 6 >
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=2 5 N O .
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> .> d
- p*=e ce m O O' *= 3 e@ e e c.e. N.
E Q U. M G C w eD
. maw 9
ce 3=U 0 w e 'a L eD ~ O m 6 = 9 O w CL C
-Q O O e e b L
a 9 O @ 22 9 e
-O C000 00 = G W
> @ 0C+CC00++ = h 5
-4 CC @XwasNoww o %
3 eO A#eNONamm 3 5 3 40 C e
* . b- C O ED M amw.C e@ O Nefw # e C.#. O - 6 4
m
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5 e
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.= b e Z O w .
m E d ##m e e t U 000 ** Z O e O CC 00+++ e g=- 6 00000 Cowww b g
. e= ONENSOceXC@N C
> 8 Q @ @ w a @ e G @ m* O.N.
D e m E = @@@N@%2NN W- Q @NamNmeNewN* e X G = E a
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==e U 6
sam e- >= - b ( h C00 000 0 0 J V COO +CO +++ e b (E D C E O c w e N O eD w w w a
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.d
= e= # 7 N N @ M = ut* N.N. N.
eDe@-@cmmN U = a . m e ,. 3 C L il Il 3 e b 9 C E
>O (g
O @ sho O > D 3G = X O e @ C m m K **
- m 5
** e @ @ e > = == e a m m D ..
.'=w p e e o e e a e e .= .- e- m a w e -
40 D ECEEEMM4eet e b( 2 K U N W m u u CD Q = = 2 c0 0 2 78
f_ ,. rj_,.
. ,e
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+ -
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*=.
2 W e m N . wo-
-CN C+ -C
+c +w + m m W+CwcCC+JO C C' NwC-=mNmc +-
g 3' @Nm>@mmNm W .
.,d ' *= @ e e % C m a n N C e .
com"2 NOME e e
.o.
w.-e. m N. =. = = =. e c c-e .. a w
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. =g E a
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Q '6 -+++ - A h www- d M & .ecN e C. p. - N w - O w
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. e 6
m
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oc-w=h
+++'
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<- 6 g
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~>. 4 e e.
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.O G 6= N . ggm L o oh h C 00. -
. e++
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A e O C f. NO O o -
<E D BNCe gQ ~ XCemOmmsXai9 V
- e@@em----mm e ..
e= W D e e o e e e e e s *=
- m m e .=
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? ;').:. Y (jf
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=
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==
a e 0'- OON- WWW aNaoceN w O
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..= - 3O-- e 5 .*
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'e
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M u = y '. - g e
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> ^ QCwQ@ Q++ == h- g
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== commOma esme N N e.@. O C
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=. >= @N@@C8R@@N-> C
>- t e g a
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*= N X#N $=
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# *= m s= N. P= g N ,= M s= @
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*= W e e e 4 e eee e ** *= m a w
<g 8- * -
= Ev~**m***a*"'='z'. o .u . g l
80
RIVTER BEND STATION t APPROVAL SHEET I STATION OPERATING PROCEDURES NO. RSP-0008 _ TITLE OFFSITE DOSE CALCULATION M UAL (ODCM) PROCEDURE SAFETY RELATED YES13 N0((l QA APPLICABIA YES((Xl NOl[l
-l REV. l PAGES INDEP. l l l l NO. l ISSUED REVIEW l APPROVED BY I EFFECT I l l SIGNATURE /DATE I SIGNATURE /DATE I DATE I I I I I I I i -
1 I i 0 1 Thru 112 07 07/11/86 I jjg DL /bj 08/13/
- l l l
"" k i ~ 07/29/88l l l l%I ,f%[ ! ,, l 3 1 thru 110 02/14/89.
l l 43 0 l l l 1 I I I I I I I I I I I I I l l OFHCI AL WORK COPY; l l l l ISSUED I . I 1 I I I I I I I AUG 2 81989 I I I l l RECEIVED l l suuos oacumua ccNm )l FEB 141989 l l l l l SDC l l l l l l 1 1 I i l i I I I I I I I I I l l l l l 1 i i 1 % i l i
'\1 I I I bdk'#
l l
~ . , . . . -
N l I l I l I s
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1
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l il @ w . , h l c ~n p -
' Table of' Contents y
p y Section - Page . I c 1
'1.0 ' Introduction..............................................' 1 3
~
!. -1.1L LPurpose'............ 4.................................... -3 1.2 . References ......................................... ..... -3
, 1.31 , Definitions.............................................. 4 J- 1.4' Required Equipment;....................................... 5 1.5 Precautions and Limitations,.............................. 5 11 . 6 - Prerequisites ........................-...................... 5 2.0; Liquid Effluent Methodology............................... ~6' 2.1 River. Bend Station Site: Description....................... 6 l2.2' '. Compliance with 10CFR20.(Liquids)...... .................. 6 2.3 - Determination of Setpoints'for Radioactive' Liquid Effluent ~ Monitors....................... ................. -7 2.4- Determining the Dose for Radioactive Liquid Effluents..... 10 2.5 Projecting Dose for Radioactive Liquid Effluents.......... 11 13.0 Gaseous Effluent Methodology............................... 12 3.1 Introduction...............................-............... 112 3.2 ' Data Requirements for Gaseous Effluents................. . 12.
3.3 Instantaneous' Release Rate and Setpoint Determination..... 13'
' 3.4 ~ Cumulative Dose-Determination for Radioactive Gaseous E f f l u e nt s .~ . . . . . . . . . . . .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 c 3.5 - Dose Projection - Determination of Need to Operate Ventilation Exhaust Treatment System.................... 38
' 4~. 0 - Radiological Environmental Monitoring' Program.... ......... 39
.5 . 0 - - 40CFR190 Considerations................................... 49 5.1 Compliance with 40CFR190.......................... ....... 49 15 . 2 . Calculations Evaluating Conformance with 40CFR190......... 49 5.3 Calculations for Total Body Dose.......................... 50 5.4 ~ Thyro i d Dos e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 0 5.5 <0r g an Do s e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 5.6 Skin Dose.................................................. 52 6.0 Interlaboratory Comparison Studies........................ 52 6.1 Requirement................................................ 52 6.- 2 Program..................................-................ 52 RSP-0008 REV. 3 PAGE 1 0F 110
_o______________________ _ _ __ _ ___ . _ _ a
a W; , u .. L Appendices
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PAGE-A Liquid MPC Values - 54
'B; ELiquid Environmental Dose Transfer Factors A g 56
'C- .K g LgAir Dose Transfer Factors 60 D ExEected_ Gaseous Radionuclides Mixture 67.
E. X/Q and Dlg Values _for Restricted Area Boundary' 69 F. Maximum X/Q and D/Q'for. Individual Locations 78 G : Instantaneous Dose Transfer Factor Tables 80 [ ~H' Gaseous MPC Values 82 I Environmental Dose Transfer Factors for' Gaseous Effluents 84 Figures-
'l Restricted Area and Near-Field Environmental Monitoring Locations 104 2 Schematic of Gaseous Radwaste System 105 3 Effluent Release Points 106 4 Schematic of Liquid Radwaste System 107 5 Far-Field Radiological Environmental Monitoring Locations 108 6 Schematic of the Solid Waste Treatment System 109 Attachments
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1 ODCM/ Procedure Revision Sheet 110 RSP-0008 REV. 3 PAGE 2 0F 110 [.
1 '. 0' INTRODUCTION 1.1 PURPOSE This manual'provides a concise description of the environmental dose-
- modelstand' techniques used to calculate offsite doses resulting from
. measured.or projected releases of. radioactive' materials from. Gulf States Utilities' River Bend Nuclear Station. It also provides the methodology for calculating effluent monitoring setpoints and allowable release rates to ensure compliance with the Radiological Effluent Technical Specifications (RETS) of Gulf States Utilities,.
River Bend Station. This manual also contains a description of the Radiological' Environmental Monitoring Program which includes sample point descriptions ~for.both onsite and offsite locations and sampling and analysis frequencies. The ODCM follows the methodology and models suggested by the " Guidance
. Manual for Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants" (NUREG-0133, dated October 1978) and " Calculation of Annual Doses to Man from Routine Releases of-Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50 Appendix I" (Regulatory Guide 1.1'09, Rev. 1, dated October
-1977). -Alternate calculational methods may be used from those presented as long as the overall methodology does not change or as
.long as the alternative methods provide results that are more
' limiting. Also, as available, the most up-to-date revision of Regulatory Guide 1.109 dose conversion factors and site-specific environmental transfer factors may be substituted for those currently included and used in this document.
1.2 RERERENCES 1.2.1 NUREG 0133; Guidance Manual for Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants; October, 4978. 1.2.2 REG. GUIDE 1.109, Rev. 1, October, 1977; Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Compliance with 10 CFR Part 50, Appendix I. 1.2.3 U.S. Code of Federal Regulations; 10CFR20. 1.2.4 River Bend Environmental Report, OLS. 1.2.5 REG. GUIDE 1.111; Methods for Estimating Atmospheric Transport and Dispersion of Geseous Effluents in Routine Releases from Light-Water - Cooled Reactors. 1.2.6 River Bend Station USAR RSP-0008 REV. 3 PAGE 3 0F 110 __-_:__-__-___-_-_--_-___-___---- __ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ - _ _-____-____--___a
1.2.7 River Bend Technical Specifications; Section 3/4.11. 1.2.8 River Bend Environmental Report, CPS. d 1.2.9 U.S. Code Of Federal Regulations, 10CFR50. 1.2.10 U.S. Code of Federal Regulations, 40CFR190. 1.2.11 NUREG 0543, Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel Cycle Standard (40CFR Part 190) 1.2.12 QAFR # P-86-03-004 1.2.13 QAFR # P-86-03-005 1.2.14 QAFR # P-86-03-002 1.2.15 CONDITION REPORT # 86-0495 1.2.16 River Bend Technical Specification; Section 6.14. 1.2.17 River Bend Technical Specification 3.3.7.10 1.2.18 River Bend Station Radiological Environmental Operating Report for 1985 1.2.19 QAFR #P-86-03-003 1.3 DEFINITIONS 1.3.1 MEMBER (s) 0F THE PUBLIC - MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant. 1.3.3 0FFSITE DOSE CALCULATION MANUAL - l The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints. It shall also contain a table and figure defining current radiological environmental monitoring sample locations. RSP-0008 REV. 3 PAGE 4 0F 110
- x c-l'. 3. 3 " -SITE' BOUNDARY -
The SITE BOUNDARY shall be that line beyond which the land is not-owned,:1 cased,'or otherwise controlled by the licensee.
'1.3.4 ' UNRESTRICTED AREA -
(An' UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY.
' access to which is not controlled by the licensee for purposes of protection of individuals,from exposure to radiation and radioactive materials, or'any' area within the site boundary us.ed'for residential quarters or for industrial . commercial, institutional, and/or recreational purposes.
-1.3.5 VENTILATION EXHAUST TREATMENT SYSTEM -
' A VENTILATION EXHAUST TREATMENT SYSTIM is any system designed and Installed to reduce gaseous radioiodine and/or radioactive material in l.
particulate form in effluents by passing ventilation or' vent exhaust gases through charcoal absorbers and HEPA filters prior to the release
- to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
1.4 REQUIRED EQUIPMENT 1.4.1 None 1.5 PRECAUTIONS AND LIMITATIONS
' 1.5.1 As per Reference 1.2.16, Licensee-initiated changes to the-ODCM/ Procedure shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report-for the period in which the change (s) was made effective.
, 1.5.2- No chtsnges(s) shall be made to the ODCM/ Procedure that will L reduce the accuracy or reliability of dose calculations or setpoint determinations. l 1.5.3 Any change (s) shall be recorded on the ODCM Revision Sheet and made in accordance with Reference 1.2.16. 1.6 PREREQUISITES 1.6.1 None RSP-0008 REV. 3 PAGE 5 0F 110 1
. _ _ _ _ . _ _ _ . _ .m______ _ _ _ . . _ _ _ . . - - _. . ~ . - . . . , - -
'I. i 2.0 L1 QUID EFFLUENT METHODOLOGY 1 2.1 River Bend Site Description The River Bend Station Updated Safety Analysis Report (USAR) contains l the official description of the site characteristics. The description that follows is a brief summary for dose calculation purposes: The River Bend Station (RBS) is on a site in West Feliciana Parish, Louisiana, located approximately 24 miles north-northwest of Baton Rouge, Louisiana. This site is just east of the Mississippi River which is used as the source of the RBS major water requirements and which receives the RBS liquid effluents. The reactor is a General Electric boiling water reactor of the BVR-6 or 1972 -'oduct line. Containment is of the Mark 3 design, a free-st 'g cylindrical steel structure surrounded by a reinforced concret .1d building. 2.2 Compliance with 10CFR20 (Liquids) 2.2.1 Requirements In accordance with Technical Specification 3.11.1.1, the concentration of radioactive material released in liquid effluents to Unrestricted Areas (Figure 1) shall be limited to the concentrations specified in 10CFR20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10' uCi/ml total activity. The concentration of radionuclides in liquid waste is determined by sampling and analysis in accordance with Technical Specification Table 4.11.1.1-1. 2.2.2 Methodology This section describes the calculational method to be used to determine F g, the fraction of 10CFR20 limits of release concentrations of liquid radioactive effluents. 2.2.2.1 General Approach Liquid effluent releases from River Bend Station are discharged through the cooling tower water blowdown which is directed to the Mississippi River. Principal sources of radwaste are from floor drains, phase separators / backwash tank subsystem, sample recovery tanks, and reactor water cleanup (as shown in Figure 4). The liquid radwaste system is operated as a batch system. Only one tank of liquid radwaste is released at a time and is considered a batch. RSP-0008 REV. 3 PAGE 6 0F 110
The radioactive content of each batch release will be determined prior k to release in accordance with Table 4.11.1.1-1 of the RBS Technical Specifications. Compliance with 10CFR20 limits will be determined
- with the following equation
f g n Cg 2.2.2.1-1 F = I 3 i=1 f1+#2 (MPC)g where: Tg = The f; action of 10CFR20 MFC limits resulting from the release source being discharged f y
= The undiluted release rate of the release source at the monitor location, in gpm f = The cooling tower blowdown release rate, in gpm 2
C = The undiluted concentration of nuclide (1), in f uCi/ml from sample assay. (MPC)g
= Maximum Permissible Concentration of nuclide (i) from Appendix A, in uCi/ml as long as F is less than 1.0, the concentration of the tank is 3
within compliance with 10CFR20 limits. 2.2.2.2 Simplified Approar.h
-8 For purposes of simplifying the calculations, the value of 3 x 10 uCi/ml (unidenti.fied 10CFR20 MPC value) could be substituted for (MPC)g and the cumulative concentration ( C-Total = sum of all identified radionuclides concentrations) or the gross beta-gamma concentration should be substituted for C . As long as the diluted f
concentration (C-Total xg f /(f1 + # 2 )) is less than 3 x 10' uCi/ml, the nuclide by nuclide calculation is not required to demonstrate compliance with 10CFR20 MPC limits. 2.3 Determination of Setpoints for Radioactive Liquid Effluent Monitors ! I 2.3.1 Requirements l Technical Specification 3.3.7.10 requires the radioactive liquid effluent monitor be operable with their high alarm / trip setpoints set to ensure that limits of Technical Specification 3.11.1.1 are not exceeded. The high alarm / trip setpoints shall be determined and adjusted by the methodology which follows. RSP-0008 REV. 3 PAGE 7 0F 110
ll. I The high alarm setpoint for the liquid effluent radiation monitor is derived from the concentration limit provided in 10CFR20, Appendix B, Table II, Column 2 applied at the restricted area ~ooundary where the discharge flows into the Mississippi River. 2.3.1.2 Liquid Effluent Monitors Two General Atomics RD-53 monitors are provided to ensure compliance with Technical Specification limits for liquid releases. The RD-53 is an offline gamma scintillation (NaI) monitor derigned for detecting radioactivity in liquids. The monitors consists of a removable sample canister surrounded by Pb shielding. A well inside the canister holds the detector within the sample fluid. The two monitors are as follows:
- 1. Cooling Tower Blowdown Line Monitor (1RMS-RE108)
I
- a. Range: 10 to 10 cpm
- 2. Liquid Radwaste Effluent Monitor (1RMS-RE107) 1
- a. Range: 10 to 10 cpm 2.3.2 Methodology The high alarm setpoint does not consider di htion, dispersion, or decay of radioactive material beyond the s!La boundary. That is, the alarm setpoint is based on a concentration '.imit at the end of the blowdown line discharge.
2.3.2.1 Liquid Radwaste Effluent Monitor (1RMS-RE107) A sample of each batch of liquid radwaste is analyzed for I-131 and other principal gamma emitters as specified in Table 4.11.1.1-1 of Technical Specification 3.11.1.1, for total activity concentration prior to release. The fraction F of the 10CFR20 MPC limits for 3 unrestricted areas is determined in accordance with the preceding section for the activity concentration released. The liquid radwaste effluent monitor will terminate a liquid radwaste discharge if activity levels exceed the Technical Specifications limits. The automatic actions associated with a trip of the monitor are:
- 1. ILWS-TV197 closes
- 2. ILWS-A0V258 opens RSP-0008 REV. 3 PAGE 8 0F 110 l
_ _ _ _ _ _ - _ = -
- o. p+
3
'ri< t p ---
% An alarm will'also be annunciated-in the main control room. c:. The:11guid radwast_e effluent.line radiation monitor. alarm setpoint is'
' determined,with the. equation:
if Sl =! A x 'g 2.3.2-1 ! I .F 3 where:
. S- = the radiation monitor setpoint (cpm or uCi/ml)
A = -the counting rate'(cpm /ml) or activity 1 concentration (uci/ml) of-the sample ~as determined in the laboratory. j g = the ratio of effluent radiation monitor counting rate to laboratory counting rate or activity-concentration in-a given batch of liquid (cpm-per epm /ml, cpm per uCi/ml, or uCi/ml per l uC1/ml) l I Note: A/F represents g the counting rate of a solution having the same-radionuclides distribution as the sample and having the maximum j
, permissible concentration (MPC) of that mixture.
2.'3.2.2 Cooling Tower Blowdown Line Monitor (1RMS-RE108) . i 1 j The cooling tower monitor alarms at high levels of radioactivity in
.the ' normal' plant service water'/ circulating water effluent to the j L Mississippi River. An alarm will be annunciated in the main control ]
room if' predetermined setpoints are exceeded. j The cooling tower monitor alarm setpoint is determined by the
~l equation: j S = 2 x BKG 2.3.2.2-1 .!
I where: .l S- .= the radiation monitor setpoint (cpm or uCi/ml)
]
BKG = monitor background value (cpm or uCi/ml) The cooling tower blowdown line is not expected to be a contaminated stream and normally would serve as a dilution source for the final radwaste system effluent discharge. Any significant upward fluctuation in the background level is indicative of a release which
.could approach 10CTR50 Appendix I limits or 10CFR20 limits when combined with'the liquid radwaste effluent.
RSP-0008 REV. 3 ? AGE 9 0F 110 ._x______--- _ _ . _
f-2.4. Determining the Dose for Radioactive Liquid Effluents
'2.4.1 Requirements-Technical Specification 3.11.1.2 requires the dose or. dose commitment
'to.a person offsite due to radioactive material released in liquid effluents be calculated on a cumulative basis ati.least every 31 days.
Dose or dose commitment shall be limited to: a) Less than'or. equal to-1.5 mrems to;the total body and to less-than or equal to 5 mrems to any organ, during any quarter; and b) Lessithan or equal to 3~mrems to the total body and less than or equal to 10 mrems to any organ during any calendar year. 2.4.2 Methodology This section provides the methodology to calculate dose to all age groups and organs from a11' radionuclides identified in the liquid effluents. The method is based on the methodology suggested by Sections 4.3 and
- 4. 3.1 - of NUREG-0133, Rev. 1 , November 1978. The dose factors A
g for all viable pathways are listed in Appendix B. The following equation provides a dose calculation to the total body or any organ for a given age group based on actual release conditions. Dg =A g
- At
- Q g 2.4.2-1 DF
- Dy n 2.4.2-2 D = I Dg TOTAL t i=1 where:
D TOTAL t = The total dose commitment to the organ (t) due to all releases during the desired time period in mrem. RSP-0008 REV. 3 PAGE 10 0F 110
1 1:; - V
*i D1 .
- it = ~ Dose commitment from radionuclides (1) received by organ (t) of:the adult age group during the time period'(mrem).
A-it ~=- Site related dose ' commitment factor to the total' body. or any organ-(t) for each identified radionuclides (1). The-Ag values-listed in Appendix'B are site-related
.to RBE (mrem /hr per uCi/ml).
A't = .The total time ~for all batch releases that occurred in the
. period (hrs)..
.i = The total quantity of nuclide (1): released during-the interval
.At-(uCi).
=
D, The near field dilution factor. Site specific value is 77-4. . DF = The total volume of dilution that occurred during the time period-(ml).
.The doses associated with each isotope may.then be summed to provide. .l
- the cumulative dose over a desired time' period (e.g., sum all doses during a 31. day period, calendar quarter, or a year).
RSP-0008 REV. 3 PAGE 11 0F 110
l l 2.5 Projecting Dose for Radioactive Liquid Effluents 2.5.1 Requirements Technical Specification 3.11.1.3 requires the liquid radwaste treatment system be used to reduce the radioactive materials in liquid wastes prior to their discharge when projected doses due to liquid effluents, to unrestricted areas ( Figure 1 ) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31 day period. 2.5.2 Methodology The following calculational methodology shall be performed at least once per 31 day period: TOTAL t bD
=
- 31 2 1 2-3 D
L PD = Projected dose commitment (mrem) to organ (t) during the 31 day period from liquid effluents. X = Number of days to date in the current quarter D 3.0 GASEOUS EFFLUENT METHODOLOGY 3.1 Introduction The River Bend Station discharges gaseous effluents through the Main Plant Exhaust Duct, Fuel Building Exhaust Duct, and Radwaste Building Exhaust Duct. The location of these release points in relation to the River Bend site is found in Figure 3. The gaseous effluent streams, radioactivity monitoring points, and effluent discharge points are shown schematically in Figure 2. For purposes of simplicity, Fuel l Building exhaust effluents are included in the Main Plant exhaust duct releases. All gaseous eff1 cent releases from the Radwaste Building Exhaust Duct are assumed to be ground level releases. The Main Plant Exhaust Duct routine releases are treated as a wake split (conditionally elevated) release. 3.2 Data Requirements for Gaseous Effluents For the purpose of estimating offsite radionuclides concentrations and radiation doses, measured radionuclides concentrations in gaseous effluents and in ventilation air exhausted from the station are relied upon. Table 4.11.2.1.2-1 in the Technical Specifications identifies the radionuclides in gaseous discharges for which sampling and analysis is done. RSP-0008 REV. 3 PAGE 12 0F 110
m_ _
- i. ._
l l ll ' I When a nuclide concentration;is.below the LLD for the analysis, it is not reported as being present inithe sample. )
.In the absence of'real-time meteorological data,' historical information will be used to calculcte off-site dose. Modelling will i be performed in accordance with the methodologies described in Reg. ;
Guide 1.111 . Rev. 1. l
- 0. <
3.3 Instantaneous Releast Rate and Setpoint Determination
; 3. 3.1 - . Instantaneous Release Rate Determination ;
i Tha instantaneous release rate determination is performed to show ! compliance with the limits' set forth-in 10CFR20. j 3.3.1.1 Requirements.
-1
~
Technical Specification 3.11.2.1 states that the dose rate due to l l radioactive materials released in gaseous effluents from the site to l areas at and_beyond the site boundary (see Figure 1) shall be limited j to the following: j
- a. For noble gases: Less than or equal to 500 mrem / year to the l total body and less than or equal to 3,000 mrem / year to the- !
skin; and
- b. 'For I-131 -I-133, tritium,Jand for all radionuclides in particulate form with half-lives greater than 8 days: less'than or equal to 1,500 mrem / year to any organ. ,
I
?
3.3.1.2 Methodology ' 3.3.1.2.1 General Approach - Total Body and Skin Instantaneous j i Release Rate Calculations i To durvrmine the dose rate from noble gases in unrestricted areas, the l folicsing formulae are used: n LR TB
= (3.15 x 10 ) I (Kg ) (X/Q) (Qg) 3.3.1.2.1-1 l
4 i=1 l l 4 l i RSP-0008 REV. 3 PAGE 13 0F 110 i i 1
p n l DR skin = (3.15 x 10,) I (Lg + 1.1 Hg) (X/Q) (Qg ) 3.3.1.2.1-2 1=1 ' where: DR = Dose rate to the total body in mrem / year. TB Kg = n e total body dose factor due to gamma emissions
.for each identified noble gas radionuclides (i) in mrem /see per uCi/m . Appendix C.
L = Skin dose factor due to beta emissions for each identified noble gas radionuclides (i) in mrem /sec per uCi/m3 . Appendix C. Mg = The air dose factor due to gamma emissions for each identified noble gas radionuclides (i) in mrad /sec per uCi/m . Appendix C. (X/Q) = The highest calculated annual average relative dispersion factor for any area at or beyond the unrestricted area beundary for all Sectors (sec/m ). Appendix F. Qg = The release rate of radionv lide (i) in gaseous effluents from all releases in uCi/sec. 1.1 = Conversion factor for M gfrom mrad to mrem. 3.15 x 10 = Number of sec/ year. In order to comply with the limits of 10CFR20, DRTB 5500 mrem / year and DR 53,000 mrem / year must be met at the most limiting location, at or beyond the site boundary. The radionuclides mix was based upon source terms tabulated in the River Bend Station USAR, Table 11.3-1 and are summarized in Appendix D. The X/Q values utilized in equt.tions 3.3.1.2.1-1 and 3.3.1.2.1-2 tre based upon maximum long-term annual average (X/Q) in the unrestricted area. Appendix F lists the maximum X/Q values for the RBS release points at the appropriate receptor locations. RSP-0008 REV. 3 PAGE 14 0F 110
% e: w l
*to select the most limiting location, the highest X/Q for each release f point is used (from Appendix F): '
l
-6 3 (X/Q)s = 3.31 x 10 sec/m 1 p .l ,
3 ,
.(X/Q), = 4.21 x 10 sec/m*~ .]
' l, where:
(X/Q), = .. Chi /Q for Main Plant exhaust duct and Fuel-Building' exhaust duct
.X/Q),
( =~ Chi /Q for Radwaste Building exhaust duct , i. l- Appendix F contains the maximum X/Q and D/Q values used in calculating l individual doses. j
, Release rates for all release points must be considered at the same. !
time. LIf releases are occurring at the same time, the total l instantaneous, dose for all releases most be less than the-limits of l Technical Specification 3.11.2.1. An a"'inistrative. centrol. limits I the release rates for each of the three talease points to 1/3 the ! total Technical Specification doses. ., l 1 l i 3.3.1.2.2 Limited Analysis Approach - Instantaneous Noble Gas Release Rate j NOTE This approach for K,ff and (L + 1.1M),ff should only be used if the relative abundances of the noble gas radionuclides in the effluent stream are similar to those listed in Appendix D or cf the previous Semiannual Effluent Report, as appropriate. (Reference 1.2.19) f 1 i l i L RSP+0008 REV. 3 1 AGE 15 0F 110 ; 1 i
h i [ ' lL . l The above methodology can be simplified to' provide for a rapid-determination of cumulative noble gas release limits based on the requirements specified,in~Section 3.3.1.1. . Beginning with equation 3.3.1.2.1-1 the implication proceeds as follows: I From an evaluation of projected releases, an effective total body dose factor (K,ff) can be derived. This dose factor is, in effect, a weighted average total body dose factor. See Appendix C for a.- detailed explanation and evaluation of K,ff. The value of K,ff.has been derived from the radioactive noble gas effluents listed in RBS-USAR and included in Appendix D. The values are:
'Radwaste Building' Exhaust Duct:
K,ff = (8.05 x 10-5) (y ,,,,3/ Ci-sec) Nain Plant Exhaust Duct and Fuel Building Exhaust Duct: 3
'K,ff = 5.56 x 10" (mrem-m /uCi-sec)
Either of these values, as appropriate, may be u' sed in conjunction with the total noble gas release rate (Q g ) to verify that the instantaneous dose rate is within the allowable limits. To compensate for any unexpected variability in the radionuclides distribution,~a conservatism factor of 028 is introduced into the calculation. The simplified equation is: n . DRg =
'(3.15 x 10 ) (K,ff) ( ) I Qg 3.3.1.2.2-1 0.8 - i= 1 where:
DR TB
= Total body dose rate from noble gases in airborne releases in mrem / year.
(X/Q) = The highest calculated annual average relative dispersion factor for any area at or beyond the unrestricted area boundary for all Sectors (sec/m ). Appendix F. Qg = The total release rate of all noble gas nuclides from the release source of interest in uCi/sec.
.3.15 x 10 = Number of seconds / year RSP-0008 REV. 3 PAGE 16 0F 110
'l
This limited' analysis approach methodology is also available for stermining skin dose. rates;from. noble gas release rates: Beginning with equation 3.3.1.2.1-2 . the simplification proceeds as follows: From an evaluation of projected releases, an effective skin dose
' factor, (L + 1.1M)eff, can be derived. This' dose factor is,.in effect, a weighted average skin dose factor. See Appendix C for a detailed explanation and evaluation of -(L + 1.1M),ff. The.value.of (L L+ 1.1M),ff has been derived from the radioactive noble gas effluents listed in RBS USAR and included in Appendix D. The values are:
Radwaste Building' Exhaust Duct: 3 (L + 1.1M),ff = 1.59 x 10 (mrem -m /uci-sec) Main Plant Exhaust Duct and Fuel Building Exhaust Duct: 3 (L t 1.1M)gff = 1.36 x 10 (mrem -m /uCi-sec) Either of these values, as appropriate, may be used in conjunction with the total noble gas release rate g( h)toverifythatthe instantaneous dose rate is within the allowable-limits. To compensate for an unexpected variability in the radionuclides distribution, a conservatism factor of 0.8 is introduced into the calculation. The ! simplified equation is: n . DR Skh = (3.15 x 10 ) (L + 1.1M),ff (X/Q) I Qg . 3.3.1.2.2-2 1. 0.8 i=1 where: DR gg = Skin dose rate from noble gases in airborne releases in mrem / year. (X[Q) = The highest calculated arawal average relative dispersion factor for any area at or beyond the vntestricted area boundary for all Sectors (sec/m ). Appendix F. hg = The total release rate of all noble gas nuclides from the release source of interest in uCi/sec. 3.15 x 10 = Number of seconds / year RSP-0008 REV. 3 PAGE 17 0F 110
,s
'I
. ..v
.,j,,
~-3.3.1.2.3~~ Determining the Radioiodine' and 8-day Particulate R Release Rates The following calculation 1Imethod is provided for determining the
, idose rate from radiciodine (I-131, 1-133),; Tritium and particulate with half-lives greater than 8 days and tc' determine if they are within the limits listed in Section 3.3.1.1-b.
m :In the calculation:to show compliance witti 10CFR20, only the
': inhalation pathway;is considered,-since it is the most limiting pathway.
Inh'alation Pathway: DR I&8DPt ( i) ' ( ) i) 3.3.1.2.3-1 i=1 where:
- DR I&BDPt = Dose' rats % the organ.t for the age group of interest from radiciodines (I/131, I-133), tritium and 8
; day particulate via the inhalation pathway (mrem /yr) hg = Release rate of nuclide-(1), (uCi/sec).
(X/Q) = The highest calculated' annual average relative dispersion factor for any area at or beyond the unrestricted area boundary for all Sectors (sec/m ).3 Appendix F. P i = The dose factor for applicable environmental pathway (mrem /yr per uCi/m ). Appendix G. Values for P were calculated for a child using the inhalation pathway 1 methodology of NUREG-0133. The Pg values are presented in Appendix G. 7 RSP-0008 REV. 3 PAGE 18 0F 110 [. ;.;
k 3.3.2 Setpoint Determination 3.3.2.1 Requirements Instrumentation is provided to monitor beta-gan.ma radiation from radioactive materials released from the River Bend Station in gaseous effluents. Each release point process monitor listed in Tech. Spec. Table 4.11.2.1.2-1 includes an alarm (HIGH AL'.RM) that is set to report when the radioactive noble gas in gaseous effluents (Main Plant exhaust duct, Fuel Building exhaust duct and/or Radwaste Building exhaust duct) is expected to cause a noble gas concentration at ground level offsite resulting in a dose rate equal to or greater than 500 mrem /yr to the total body and/or 3000 mrem /yr to the skin. The ALERT alarm is set to report when the radioactive noble gas in gaseous effluents (Main Plant exhaust duct, Fuel Building exhaust duct and/or Radwaste Building' exhaust duct) is expected to cause a noblo gas concentration at ground level offsite that would result in meeting or exceeding either the 5 mrad per quarter gamma air dose or 10 mrad per quarter beta air dose limit (Technical Specification 3.11.2.2). It is permissible to set the ALERT alarm at twice (2.0) normal (approximately 100 % unit power) detector background if nuisance alarms would result from setpoints based on gamma and beta air dose. (Reference 1 2.12) The distribution of radioactive noble gases in a gaseous effluent stream is determined by gamma spectrum analysis of identifiable radionuclides in effluent gas sample (s). Results of one or more previous analyses may be averaged to obtain a representative spectrum. In the event the distribution is unobtainable from measured data, the , distribution of radioactive noble gases based on past data or calculated by the BWR-GALE code appearing in Appendix D may be assumed. To allow for multiple sources of releases from the three different release points, the allowable operating setpoints will be administrative 1y controlled to allocate one-third (1/3) of the total allowable release to each of the release sources. 3.3.2.2 Methodology
- a. HIGH ALARM Setpoint Determination This section describes the methodology for determining HIGH ALARM / trip setpoints for the three release points:
- 1. Wide Range Jas Monitor (VRGM)
RSP-0008 REV. 3 PAGE 19 0F 110
f( ;. r, [- 1 Step 1,- Determine hTB utilizing one of the following methods: Qg = (3.17 x 10~0) (500) (0.8) 3.3.2.2-1 (5Id) (K,ff) or NOTE The K g methodology for determining h D should be used only if isotopic analyses is available
-and the relative abundances of noble gas nuclides in the effluent stream are not similar to those listed in Appendix D or not similar to the noble gas isotopic mixture described in the previous Semiannual Effluent Report. (Reference 1.2.19)
= . x 10 (500) 3.3.2.2-2 TB (E[Q)I (Kg ) (f f) i=1 where:
h3 = maximum acceptable total release rate of all noble gas radionuclides in the gaseous effluant (uCi/sec). (X/Q) = The highest calculated annual average relative dispersion factor for any area at or beyond the unrestricted area boundary for all Sectors (sec/m ). Appendix F. Kg = The total whole body dose factor due to gamma emissions from noble gas radionuclides (1) (mrem /sec per uCi/m ) from Appendix C, Table C-1. f = Fraction of noble gas radionuclides (i) to total f noble gas concentration. K,ff = Effective dose facter (mrem /sec per uCi/m ) from Appendix C, Table C-3. 3.17x10'8 = Inverse of number of seconds per year in year /sec. RSP-0008 REV. 3 PAGE 20 0F 110
m >
.. ~l 1- ~ ,
0.8 ,= Conservative factor to account for changing l'sotopic ! inventory. l 1 500- = Whole body exposure limits of 500 mrem / year. l
~
3.17x10 8- = : Inverse of number of seconds per year in year /sec. '! 1 i
' Step 2 .
Determine Q, utilizing one of the following methods.
.= (3.17 x 10-8) (3,000) (0.8) 3.3.2.2-3 ,
.l l
1 (ETQ)-(L+1.1M),ff ! or
. NOTE The (L + 1.1M)3 methodology for determining Q3 should be used only if isotopic analyses is )
available and the relative abundances of noble j gas nuclides in the effluent stream are not j similar to those listed in Appendix D or not j similar to the noble gas isotopic mixture described in the previous Semiannual Effluent Report. (Reference 1.2.19)
= (3.17 x 10-8) (3,000) 3.3.2.2-4 n
(27Q)I [(Lgf 1.1M1 )fg ]
. i=1 Q,
= the maximum acceptable releass' rate of all gas radionuclides in the gaseous effluent [uCi/sec]) !
= Total skin dose factor due to emission from noble gas Lg + 1.1Mg radionuclides (i) mrem /sec/uCi/m from Appendix C.
= The highest calculated annual average relative j!
(5[Q) dispersion factor for any area at or beyond the j unrestricted area boundary for all Sectors j (sec/m ). Appendix F. ] 1 i RSP-0008 REV. 3 PAGE 21 0F 110
r _ _ _ _ N , p - . %. ,
- r. -
g v;
'(L+1.1M)gf. = . Effective total skin dose factor (mrem /sec/pC1/m )-from Appendix C f
-Table'C-4 3000 = Skin' exposure limit _of 3000 mrem / year
-8 Inverse.of number of seconds per year'in year /sec.
-3.17x10 =
ll Step 3 Select the.Iower_of the h values-l(hTB # s) btained in Step l'and Step 2. NOTE ~
' Actual alarm setpoint in the data-base may be.
. modified to account for loop accuracy.
Step 4-Multiply the Q value. selected in Step 3 by.0.33. By multiplying the Q
- value by 'a: factor of 0.33, the allowable operating setpoints will- be administrative 1y' controlled to ' allocate one-third (1/3) of the total allowable release rate to each of: the release points. The resultant product will be the actual ODCM release rate HIGH ALARM:setpoint for the appropriate WRGM Monitor.
- 11. Particulate and Gas Monitor (P&G) (gas channel only).
Step 1. Perform Steps.1 through 3 of Section 3.3.2'.2a.1 above Step 2
' Determine C ,_(the maximum acceptable total radioactivity concentration
.of all noble gases radionuclides for all release points in the gaseous effluent [uCi/cc]):
~
C,= (2.12 x 10 3.3.2.2-5 F where: 2.12 x 10~3 = Unit conversion factor to convert uCi/sec/cfm to uCi/cc. h = lower of the two h values, hTB # s' l F = The maximum acceptable effluent flow rate at the point of release based on design flow rates (cfm) i RSP-0008 REV. 3 PAGE 22 0F 110
I NOTE Actual alarm setpoint in the data-base may be modified to account for loop accuracy. Step 3 Multiply the C,value determined in Step 2 by 0.33. By multiplying the C ,value by a factor of 0.33, the allowable operating setpoints will be administrative 1y controlled to allocate one-third (1/3) of the total allowable release to each of the release points. The resultant product will be the actual ODCM activity concentration HIGH ALARM setpoint for the appropriate P&G monitor gas channel.
- b. ALERT Setpoint Detern.ination (Reference 1.2.12)
- 1. Wide Range Gas Monitor (WRGM)
Step 1 Determineh G-A utilizing ne the following methods: Qg .g = (1.26 x 10' ) (S)(0.8) 3.3.2.2-6 (7Q) (M,ff) OR NOTE The M g methodologyfordeterminingh should G-A be used only if isotopic analyses is available and the relative abundances of noble gas nuclides in the effluent stream are not similar to those listed in Appendix D or not similar to the noble gcs isotopic mixture described in the previous Semiannual Effluent Report. (Reference 1.2.19)
= . x 0 ) (S) 3.3.2.2-7 G-A (57Q)I Mg fg i=1 Where:
h-A G
" maximum acceptable total release rate of all noble gas radionuclides in the gaseous effluent [uC1/sec)
RSP-0008 REV. 3 PAGE 23 0F 110
7 . . l! (5f[6) = The highest calculated annual average relative dispersion factor for any area at or beyond the unrestricted ares boundary for all Sectors (sec/m ). Appendix F. ti,f f = Effective gamma air dose factor (mrad-m /uCi-sec). Appendix C, Table C-5. 5 = 5 mrads/ quarter (92 days) gamma air dose limit at the unrestricted area boundary. M = The gamma air dose factor for radioactive noble f 3 gas nuclide (i) in mrad-m /uci-sec (Appendix C). f g
= The fractional abundance of noble gas radionuclides i 1.26 x 10" = Inverse of number of seconds per quarter in quarters /second.
0.8 = Conservatism factor to account for changing isotopic inventory. Step 2 Determineh,gg utilizing one of the follwing methods. o3 ,3 = (1.26 x 10-7) (to) (o.sj 3.3.2.2-8 (57d) (N,ff) or NOTE The Ng methodology for determining QB-A should be used only if isotopic analyses is available the relative abundances of noble gas nuclides in the effluent stream are not similar to those listed in Appendix D or the previous Semisnnual Effluent Report. (Reference 1.2.19) B-A (5[d) I (Ng ) (f ) g i=1 RSP-0008 REV. 3 PAGE 24 0F 110
;. 'i l
~
L Where: QB-A, s. maximum acceptable total release rate of all !. noble gas radionuc1' ides in the gaseous effluents (uCi/sec).: ). . (X[6) = The highest calculated annual average relative dispersion factor for an area at or beyond the unrestricted area boundary for all sectors-(sec/m3 ) (Appendix F). 10 = 10 mrad / quarter (92 days)_ beta air dose limit' at the unrestricted area BOUNDARY. Nggf
=' Effective beta air dose factor (mrad --
m /uCi-sec). Appendix C, Table C-5. Ng = The air dose factor due'to beta emissions from each' noble gas radionuclides i. fg = The fractional abundance of noble gas radionuclides 1. 1.26 x 10 = Inverse of number of seconds per quarter in quarters /second. 0.8 = Conservatism factor to account for changing isotopic inventory. Step 3 SelectthelowerofthehvaluesobtainedinSteps1and2, eitherh-A G B-A" Step 4
'MultiplythehvalueselectedinStep3by0.33. Bymultiplyingtheh v4.lue by this- f actor, the allowable operating setpoints will be administrative 1y controlled to allocate one-third (1/3) of the total allowable release rate to'each of the release points. The resultant product will be the actual ODCM ALERT setpoint to be entered into the applicable WRGM's RM-80.
RSP-0008 REV. 3 PAGE 25 0F 110
I 1 l' Step 5 If the actual ODCM ALERT setpoint determined in Step 4 is less than two (2.0) times the detector background, it is permissible to enter an ALERT setpoint equal to two (2.0) times the normal (approximately 100% unit power) detector background to reduce the possibility of nuisance alarms. T% twice background setpoint should provide sufficient indication that an offsite dose limit could possibly be exceeded, ii. Particulate and Gas Monitor (P&G) (gas channel only) Step 1 Perform Steps 1 through 3 of Section 3.3.2.2.b.i above. Step 2 Determine C, (the maximum acceptable total radioactivity concentration of all noble gas radionuclides for all release points in gaseous effluent [uCi/cc)):
~3 C,= (2.12 x 10 3.3.2.2-10 F
Where: 2.12 x 10' = Unit conversion factor to convert uC1/sec/ cfm to uCi/cc. h = Lower of the two h values, O # G-A B-A F = The maximum acceptable effluent flow rate at the point of release based on design flow rates (cfm). Step 3 Multiply the C, value determined in Step 2 by 0.33. By multiplying the C,value by this factor, the allowable operating setpoints will be administrative 1y controlled to allocate (1/3) of the total allowable release to each of the release points. The resultant product will be the actual ODCM activity concentration ALERT setpoint. This value is the setpoint to be entered into the appifcable P&G monitor's RM-80. Step 4 If the actual ODCM ALERT setpoint determined in Step 3 is less than two (2.0) times the gas detector background, it is permissible to enter an ALERT setpoint equal to two (2.0) times the normal (approximately 100'.' unit power) gas detector background to reduce the possibility of nuisance alarms. The twice background setpoint should provide sufficient indication that an offsite dose limit could possibly be exceeded. RSP-0008 REV. 3 PAGE 26 CF 110
h. 3.4 Cumulative Dose Determination for Radioactive Gaseous Effluents 3.4.1 Noble Gases 3.4.1.1 Requirements
- a. Air Dose Technical Specification 3.11.2.2 states that the air dose due to noble gases released in gaseous effluents from each reactor unit to areas at and beyond the site boundary (see Figure 1) shall be limited to the following:
- 1. During any calendar quarter: less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation; and
- 11. During any calendar year: less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
- b. . Total Body and Skin Dose (Reference 1.2.13.)
- 1. Technical Specification 3.11.4 states that the annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and to radiation from uranium fuel cycle sources, shall be limited to less than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.
ii. Technical Specification 6.9.1.8 (Semi-Annual Effluent Release Report) requires that an assessment of radiation doses to the likely most-exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) be performed for the previous calendar year to show conformance with 40 CFR190, Environmental Radiation Protection Standards for Nuclear Power Operation. Cummulative doses from liquid effluents and gaseous pathways (radioiodines (I-131, I-133), Tritium and particulate with T 1/2 > 8 days) are determined in accordance with Sections 2.4.2 and 3.4.2.5. Cummulative total body and skin doses from noble gas releases are determined in 'accordance with Section 3.4.1.2b. 3.4.1.2 Methodology
- a. Air Dose This section provides the methodology to calculate the gamma and beta air doses to a maximum receptor location at the site boundary from all noble gas radionuclides identified in the gaseous effluents.
RSP-0008 REV. 3 PAGE 27 0F 110
f' The method is based on the methodology suggested by sections 5.3 and 5.3.1 of NUREG-0133, Rev. 1 November, 1978. The dose factors for beta and gamma air dose are listed in Appendix C and are obtained from Table B-1 of RG 1.109, Revision'1, October 1977. The following equations provide for air dose calculations based on actual noble gas releases during a specific time interval for
. radioactive gaseous release sources at the site boundary:
n D Gamma-Air ("i) ) (01) 3 . . a-1 i=1 n
. D Beta-Air
( i) ( (01 ) . . a-2 1=1 where: D Gamma-Air
= The gamma air dose from radioactive noble gases in mrad.
Hg = The gamma air dose factor for radioactive noble gas nuclide (i)'in mrad-m /uCi-sec_(Appendix C).
= The highest calculated annual average relative (3I[Q) dispersion factor for an area at or beyond the unrestricted area boundary for all sectors (sec/m ) (Appendix F).
Qg = The number of uCi of nuclide (i) released during the period of interest. D = eta ah dose from radioactive nome gases in mrad. Beta-Air Ng = The beta air dose factor for radioactive noble gas nuclide (1) in mrad-m /uCi-sec (Appendix C), Table C-1.
- b. Total Body and Skin Dose (Reference 1.2.13)
This section provides the methodology to calculate the total body and skin doses to the likely most-exposed MEMBER OF THE PUBLIC from all
-noble gas radionuclides identified in the gaseous effluents.
l The method is based on the methodology suggested in Section C.2 and i Appendix B of RG 1.109, Revision 1, October, 1977. The dose transfer factors required for the calculations are listed in Appendix C of this document and are obtained from Table B-1 of RG 1.109, Revision 1, October, 1977. RSP-0008 REV. 3 PAGE 28 0F 110
, e!jp r .,
i o
;The following equations provide for total. body and skin dose'.
l calculations based'on actual noble l gas' releases during a specific time interval for' radioactive gaseous release sources'at the site boundary: Totai' Body." 8F: I IE II Q)(0 1) 3.4.la2b. , i=1 DSk h " '8F' Z ILi +'I*1Hi )I )IOi ) 3.4.1.2b.-2. i=1:
'Where:
?D .
,= The total body dose from radioactive noble' gases in mrem.
Kg = The, total whole body dose factor due to gamma emmissions from inoble gas radionuclides.(i) (mrem /sec per uCi/m3) from Appendix C, Table C-1.
' (T/6) - = Th$ highest calculated annual average relative
, dispersion factor for an area at or beyond the unrestricted area boundary for all sectors
- (sec/m ) (Appendix F).
NOTE For purposes of calculating D Total Body and D Skin r the Semiannual Radioactive Effluent-Release Report, X/Q values based on meteorological data for the actual considered time period should beusedratherthanhistorical-(I/6) values. 'If at all possible, these real time'X/Q values should also be used when determining 40CFR190 compliance when Technical Specification limits have been exceeded by a factor of two (2.0). Q g = The number of uCi of noble gas nuclide (i) released during the period of interest. D Skin
= The skin dose from radioactive noble gases in mrem.
.M g = The gamma air dose factor due to gamma emissions from each noble gas radionuclides (i) released.
RSP-0008 REV. 3 PAGE 29 0F 110
L. l: ~c l
'L g = ' The skin dose factor due to beta, emissions from noble gas radionuclides (i) (mrem /sec per uCi/m )-from Appendix C, Table C-1. '
-1.1 =. Average ratio of' tissue to air energy absorption coefficients.
~S F
= 0.7, attenuation factor' accounting for shielding pro *;ided by residential structures'for maximally exposed individual.
3.4.1.3 Simplified Approach
~
A single effective gamma air dose factor (M,f.f) and beta air dose . factor (N,ff) have been' derived, which are representative of the . radionuclides abundances and' corresponding dose contributions that are projected in the RBS USAR.. (See Appendix C for a detailed explanation
-l and evaluation of M,ff:and N,ff). .The values of M,ff and N,ff which
.have been derived from the~ projected radioactive noble gas effluents are:
Radwaste Building Exhaust Duct:
~$
M,ff = 8.07 x 10 mrad-m /uCi-sec
='
~
3 N,ff 7.40 x 10 mrad-m /uci-sec Main Plant Exhaust Duct and Fuel Building Exhaust Duct:
~
M,ff = 5.96 x 10 mrad-m /uCi-sec N,ff = 8.99 x 10 mrad-m /uCi-sec NOTE The M,ff and N,ff factors should only be used if the actual effluent is similar to that describ'ed in Appendix D or similar to the noble gas isotopic mixture described in the previous Semiannual Effluent Report. (Reference 1.2.19) RSP-0008 REV. 3 PAGE 30 0F 110
3
; i$$
The effective gamma air dose factor'may be used in conjunction with
-the total noble gas release to simplify the dose evaluation and to verify that the cumulative gamma and beta air dose is within the equivalence of the limits of Technical Specification 3.11.2.2. To compensate for any unexpected variability in the radionuclides distribution, a conservatism factor of 0.8 is introduced into the calculation. The simplified equation is:
.(M,ff) M) n D
- O . .1.3-1 Gamma-Air f 0.8 i=1 (N,ff)
@) n D " E 3 1* ~
Beta-Air i 0.8 -i=1 3.4.2 Determining the Radioiodine and 8 Day Particulate Dose to Any
' Organ from Cumulative Releases 3.4.2.1 Requirements Technical Specification 3.11.2.3 states that the dose to a Member of the Public from iodine-131, iodiae-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the site boundary.shall be limited to the following:
- a. During any calendar quarter: less than or equal to 7.5 mrem to any organ; and-
- b. During any calendar year: less than or ecual to 15 mrem to any organ.
RSP-0008 REV. 3 PAGE 31 0F 110
3.4.2.2 Methodology The following calculational method is provided for determining the ! critical organ dose due to releases of radiciodines (1131, 1133), trituim and particulate. It is based on Section 5.3.1 of NUREG-0133, Rev. 1, November 1978. The equation can be used for any age group provided that the appropriate dose factors are used and the total dose reflects o'ily those pathways that are applicable to the age group. The symbol (X/Q)D represents a depleted (X/Q) which is different from the noble gas (X/Q) in that (X/Q)D takes into account the loss of radioiodines (I-131, I-133), 8 day particulate, and tritium from the plume es the semi-infinite cloud travels over a given distance. The dispersion factor (D/Q) represents the rate of fallout from the cloud that affects a square meter of ground at various distances from the site. The total dose to an organ can then be determined by summing the pathways that apply to the receptor in the sector. The equations are: Inhalation Pathway: n
~
D I&BDPt = (3.1 x 10 )I (Rit) ( IO)D (01 ) ' ' ' ~1 i=1 Ground Plane Pathway:
~
D IS8DPt =(. x 0 ) (Rgt) (D/Q) (Qg ) 3.4.2.2-2 i=1 Contaminated Forage / Cow / Milk Pathway: n D I&8DPt =( . x 10~0) I (Rit) (D/Q) (Qg ) 3.4.2.2-3 i=1 Total Dose: n D Total = 1 D I&BDPt 3.4.2.2-4 z=1 IMPORTANT When calculating organ doses due to the release of C-14 and/or tritium (H-3), X/Q values (not D/Q values) must be used for cow milk, goat milk, meat and vegetation pathway calculations. RSP-0008 REV. 3 PAGE 32 0F 110
f ..
,f where:
D. ._ l
.I&8DPt = Dose in mrem to the organ (t)- 1 of a specified aga' group from radiciodines i
.(I-131, I-133), tritium'and 8 day-particulate due to a particular pathway.
l; ' z -- = All the applicable pathways for the age group of l
' interest. ;
D ~
- Total- = Total dose in erem to,the organ (t)'of a- i specified age group from gaseous radioiodine (I-131, _j I-133), tritium and particulate effluents,' summed over !
all applicable pathways (z).
~0=
'3.17 x 10 'The--inverse of the number of, seconds per year' (years /sec). !
1 i i
-R _ _
it = The dose factor for nuclide (i) for pathway (z) to organ ,(t) of-the specified age group. The units are either:
-l mrem-m for pathways using (X/Q)D I yr-uCi l
or l l 2 mrem-m -sec for pathways using (D/Q). yr-uCi (See Appendix I.) i
'l
'l (X/Q).
D. = The depleted (X/Q) value for a specific location where 3 the receptor is located (sec/m ). (See Appendix F.) Note: No credit is te. ken for depletion and decay. (X/Q)g =-(X/Q) .j (D/Q) '= The deposition value for a specific location where the ! receptor is located (m-2). (See Appendix F.) l 4 i RSP-0008 REV. 3 PAGE 33 0F 110 l
l NOTE For purpose of calculating D r the R8DPt Semiannual Radioactive Effluent Release Report, X/QD and D/Q values based on meteorological data for the actual considered time period shouldbeusedratherthanhistorical(X/Q)D and (D/Q) values. If at all possible, real time X/QD and D/Q values should also be used when determining 40CFR190 compliance when Technical Specification limits have been exceeded by a factor of two (2.0). Q i = The number of microcuries of nuclide (1) released (or projected) during the dose co?culation exposure period. 3.4.2.3 Limited Analysis Approach The contaminated forage / cow / milk pathway has been identified in Section 5.4 of the RBS ER-OLS as the most limiting, with the infant thyroid being the most critical age group and organ. It is possible to demonstrate compliance with the dose limit of Technical specification 3.11.2.3 for radiciodines (I-131, I-133), tritium and particulate by only evaluating the inf ant's thyroid dose due to the release of radiciodines via the contaminated forage / cow / milk pachway. The calculational method to be used includes a conservatism factor of 0.8 which assures that the calculated dose is always greater than or l equal to the actual dose despite possible atypical distributions of I radionuclides in the gaseous effluent. The simplified dose equation reduces to: D= (3.17 x 10-8) (D/Q) E (Rg ) (Qg ) 3.4.2.3-1 0.8 radioiodines , 3.4.2.4 Approach Selection Criteria l The limited analysis may be used in all cases to demonstrate compliance with the dose limit of Technical Specification 3.11.2.3 (7.5 mrem /qtr) for radioiodines (I-131, I-133), tritium and particulate. However, for the dose assessment included in the Semi-annual Radioactive Effluent Release Report, doses will be evaluated for all designated age groups and organs via all designated pathways from radiciodines (I-131, I-133), tritium and particulate measured in the gaseous effluents according to sampling and analyses required by the Technical Specifications. RSP-0008 REV. 3 PAGE 34 0F 110 k _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _
!. L:
?, ,
, s 5;
L3i4;2.5l : Annual Dose Due to'Radioiodines (I-131, I-133), tritium, andl 8-Day' Particulate Technical Specification 3.11.2.3. required the annual dose be calculated at least once per.31 days for all pathways. The following-formulae are used'to calculate the annual dose for radioiodines,
- (I-131, 1-133),l tritium and 8- day particulate:
Inhalation Pathways: D _ I&8DPt
= .1 x 0 ): L (Rgt) (X/Q)D (01) 3.4.2.5-1 i=1-Ground Plane Pathway:
'n.
.g (01 ) .4.
DI&BDPt'L= (3.17 x 10-8) ;(Rit ) ' ( IO) . -2
.i=1 Contaminated Forage / Cow / Milk Pathway:
~
D ygggp = (3.17 x 10 ) I (Rgt) ( ) -(Qg ) 3.4.2.5-3 i=1 Contaminated Forage / Goat / Milk Pathway: n D I&BFPt = (3.17 x 10-8) I (Rit) - I ) IN1 ) * * * ~ i=1 Conta:ainated Forage / Meats: n
~
D I&BDPt = (3.17 x 10 ) I (Rit) ) (01) "
~
i=1 s RSP-0008 REV. 3 PAGE 35 0F 110 L . _ _ __ - ._ --. _ _ - - _ . - _ _ _ - . _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - -
l' Fresh Fruits and Vegetables: n D I&BDPt ( .1 x 0 ) I (Rit) -( ) (01 ) *2'0~0 i=1 Total Dose: n t I&BDPt 3.4.2.5-7 z=1 where: I&8DPt = Annual dose to the organ (t ) l for the age group of interest from radiciodines (I-131 I-133), tritium and 8-day particulate via the pathway of interest in mrem /yr. For radioiodines (I-131,'I-133), the entire source term was used to calculate these values. z = All the applicable pathways for the age group of interest. Qf = The number of uCi of nuclide (i) released during the year of intarest. Rg = The dose factor for nuclide (i) for organ (t) for the pathway specified [ units vary with pathway]. For tritium, a site-specific absolute humidity (H) value of 12.9 gm/m was used for calculation. (See Appendix I.) (D7Q) = A long-term relative depesition value for elevated and
~
ground level releases. A fa-tor with units of m which describes the deposition of particulate matter from a plume at a point downrange from the source. Actual meteorological data and sector wind frequency I distribution will be used to determine annual average D/Q for the year of interest. l RSP-0008 REY. 3 PAGE 36 0F 110 _ _ - _ . _ _ - ____-___-___-__-_______Q
I I (X7Q)D
= A long-term depleted and 8-day decayed relative concentration value for elevated and ground level release (sec/m ). It describes the physical dispersion characteristics of a semi-infinite cloud traveling downwind. Since radioicdines (I-131, I-133), and particulate sy tle out (fallout of the cloud) on the ground, the (X/Q)D represents what physically remains of the cloud at a given location downwind from the release point. Actual meteorological data and sector wind frequency d M ributions will be used to determine annual average (X/Q)D f r the year of interest. Total body and organ doses will be calculated for pathway and age group on an annual basis using the above-described methodology.
IMPORTANT When calculating organ doses due to the release of C-14 and/or tritium (H-3), (X/Q) values (not D/Q values) must be used for cow milk, goat milk, meat and vegetation pathway calculations. NOTE For purposes of calculating DR I&BDPt '#
- Semiannual Radioactive Effluent Release Report, X/Q and D/Q values based on D
meteorological data for the actual consid-ered time period should be used rather than historical (X/Q)D and (D/Q) values. If at all possible, real time X/Q and D/Q D values should also be used when determining 40CTR190 compliance when Technical Specifi-cation limits have been exceeded by a factor of two (2.0).
-8 =
3.17 x 10 The inverse of the number of seconds per year (in year /sec). Meteorological data (X/Q, X/Q D, D/Q) will be determined from actual meteorological data and sector wind frequency distributions for the year of interest. Release rates (uCi/ year) will be based on total activity released through elevated and ground level (total of all vent pathways) as reported in the Semi-annual Radioactive Effluent Release Report. RSP-0008 REV. 3 PAGE 37 0F 110
l l I 3.5 Dose Projection - Determination of Need to Ope rate i Ventilation Exhaust Treatment System 3.5.1 Requirement Technical Specification 3.11.2.5 requires that the ventilation exhaust treatment system be used to reduce radioactive material in waste prior to discharge when the projected dose due to gaseous effluents (radiciodines (I-131, 1-133), particulate T 1/2 > 8 days and H-3) would exceed 0.3 mrem to any organ in a 31 day period. NOTE The ventilation exhaust treatment system does not reduce the noble gas concentration in plant effluents (See Definition 1.3.5). 3.5.2 Methodology The following calculation method is provided for determining the projected doses: ID t G =
- 31 3.5.2-1 PD D
where: G = Projected dose due to radiciodines (I-131, I-133). PD particulate with T > 8 days and H-3 during the 1/2 current 31 day period (mrem). Xp = The number of days to date in the current quarter D = Cumulative total dose due to radioiodines (I-131, I-133), particulate with T 1/2 > 8 days and H-3 during the current quarter (mrem). A dose projection would be based on the latest results of the monthly calculations of the dose due to radiciodines (I-131, I-133), particulate with T 1/2 > 8 days, and H-3 (Section 3.4.2.5). The value may need to be adjusted to account for any changes in operating conditions that could significantly alter the actual releases, such as failed fuel, or changes in ventilation flow rate. RSP-0008 REV. 3 PAGE 38 0F 110
i .l I 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Table 4.1 centains the sample point description, sampling and collection frequency, analysis, and analysis frequency for various exposure pathways in the vicinity of RBS for the radiological monitoring program. Figures 1 and 5 indicate the locations of the various onsite and offsite sampling points and TLD locations. This section describes only those elements of the radiological environmental monitoring program required by the RBS Technical Specifications. Additional exposure pathways, sample points, analyses, and/or frequencies are performed as described in ER-OLS Section 6.2. Samples of groundwater are taken from onsite wells located to intercept any potential contamination of the Upland Terrace Aquifer so thr; any such contamination would be detected before migrating beyond RBS site boundaries. l l l l l I RSP-0008 REV. 3 PAGE 39 0F 110 l l l
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*0- - - e-o6 M b Q **> C C a = C e .d 4 3- .. .E. .. e C3 e - -e3 - ea GD p-= a == e ** h e@ -EC e da 0-0 )e e : C .* Em VeO C* -@ e o a em WE 60e i i b es 6e am@ m **v .e o e e m6 O E6 CL ea 4 = 4.* D 3 :W e C -EE C G3 07 CL C O %m= .nn *= D -%C 4.J ** r W 6.a . . == e= N O E 2 --___ J _ _ _ _ _ _ _ _ - . _ _ - _ _ _ - - . . _ _ _ - - _ _ . _ _ . _ _ _ l. i 5.0 40CFR190 CONSIDERATIONS 5.1 Compliance with 40CFR190 Compliance with 40CFR190 as-prescribed by Technical Specification 3.11.4 is to be demonstrated only nhen one or more of Technical Specification (s) 3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a. 3.11.2.2.b, 3.11.2.3a, and 3.11.2.3 b, including direct radiation are exceeded- by , a factor of 2. : Once this occurs, GSU has 30 days to submit a report N in accordance with Specification 3.11.4. 5.2 Calculations Evaluating Conformance with 40CFR190 To perform the calculations to evaluate'conformance with 40CFR190, an effort is made to develop doses that are realistic by removing assumptions that lead to overestimates of dose to a Member of the Public (i.e., calculations for compliance with 20CFR50 Appendix I.). To accomplish this, the following calculational rules are used: 5.2.1 Doses to Members of the Public via the liquid release pathway are considered to be'< 1 mrem /yr (Ref NUREG-0543). 5.2.2 Doses to a member of the Public due to a milk pathway will be ; evaluated only as can be shown to exist. Otherwise, doses via this pathway will be estimated as < 1 mrem /yr. , 5.2.3 Environmental sampling data which demonstrate that no pathway ! exists may be used to delete a pathway to man from a , calculation.' ! 5.2.4 To sum numbers represented as "less than" (<) , use the.value i of the largest number in the group. I e.g., <5 + <1 + <1 + <3 = 5 ! 5.2.5 k' hen doses via direct radiation are added to doses via ! inhalation pathway, they will be calculated for the same distance in the same sector. 5.2.6 The calculational locations for-a Member of the Public will j only be at residences or places of employment. ; 5.2.7 If at all p ;ble; X/Q, X/Q ' and D/Q values based on l D meterological 'ata for the actual considered time period should be used in determining compliance with 40CFR190. Note: Additional assumptions may be used to provide situation specific parameters, provided they are documented along with ' their concomitant bases. 1 I l l 4 RSP-0008 REV. 3 PAGE 49 0F 110 .) 1 J: D:- O, p b2 ) s F '5.3 ' Calculations'of Total Body Dose < k i Estimates will be made for each of the following exposure ~ pathways to the same location'by-age class. Only those age classes known to exist at a location are considered. ~ 5.3.1. Direct Radiation (from storage tanks, N-16 sources, etc.) ~ The, component'of dose'to a Member of the Publir due to direct radiation will be determined by thermoluminescent dosimeters (TLDs). 5.3.2- Inhalation Dose The inhalation dose will be determined at the calculational locations for each age. group according to the methods outlined in Sections 2.0 and-3.0 of this manual. 5.3.3. . Ingestion Pathway (cow milk,' goat milk, meat, vegetation) ~ The dose via the ingestion pathway will be calculated at the consumer-locations'for the consumers at risk. If no milk pathway exists in a < sector, the dose via this pathway will be treated as < 1 mrem /yr. 5.3.4- Total Body Noble Gas Immersion Dose This dose will be calculated in accordance with Section 3.4.1.2b. for the maximally exposed MEMBER OF THE PUBLIC in the limiting sector. 5.3.5 Ground Plane Deposition 5.3.6 Other Uranium Fuel Cycle Sources The dose from other fuel sources will be treated as < 1 mrem /yr. 5.4 Thyroid Dose The dose to the thyroid will be calculated for the limiting sector as the sum of: 5.4.1 Direct Radiation (from storage tanks, N-16 sources, etc.) The component of dose to the thyroid due to direct radiation will be determined by thermoluminescent dosimeters (TLDs). 5.4.2 Inhalation Dose The inhalation dose to the thyroid will be determined at the !- calculational locations for each age group according to the methods outlined in Sections 2.0 and 3.0 of this manual. RSP-0008 REV. 3 PAGE 50 0F 110 _ _ _ _ - . _ _ _ _ --._---__-__________.________________m__ 5.4.3 Ingestion Pathway (cow milk, goat milk, meat, vegetation) The dose to the thyroid via the ingestion pathway will be calculated at the consumer locations for the consumers at risk. If no milk pathway exists in a sector, the dose via this pathway will be treated as < 1 mrem /yr. 5.4.4 Noble Gas Icmersion Dose It is assumed that an external total body dose from noble gases irradiates internal body organs at the same numerical rate (Reference 1.2.11). This dose for the thyroid will therefore be equal to the dose calculated in Step 5.3.4 above. 5.4.5 Ground Plane Deposition 5.4.6 Other Uranium Fuel Cycle Sources The dose from other fuel cycle sources will be treated as < 1 mrem /yr. 5.5 Organ Dose (other than thyroid and skin) The dose to any organ will be calculated for the limiting sector as the sum of: 5.5.1 Direct Radiation (from storage tanks, N-16 sources, etc.) The component of dose to an organ due to direct radiation will be determined by thermoluminescent dosimeters (TLDs). 5.5.2 Inhalation Dose The inhalation dose to an organ will be detern ined at the calculational locations for each age group according to the methods atlined in Sections 2.0 and 3.0 of this manual. 5.5.3 Ingestion Pathway (cow milk, goat milk, meat, vegetation) The dose to an organ via the ingestion pathway will be calculated at the consumer locations for the consumers at risk. If no milk pathway exists in a sector, the dose via this pathway will be treated as < 1 mrem /yr. 5.5.4 Noble Gas Immersion Dcse It is assumed that an external total body dose from noble gases irradiates internal body organs at the same numerical rate (Reference 1.2.11). This dose for an organ will therefore be equal to the dose l calculated in Step 5.3.4 above. l 5.5.5 Ground Plane Deposition l l RSP-0008 REV. 3 PAGE 51 0F 110 i . 5.5.6 Other Uranium Fuel Cycle Sources The dose from other-fuel cycle sources will be treated as < 1 mrem /yr. 5.6 Skin Dose The dose to the skin will be calculated for the limiting sector as the sum of: 5.6.1 Direct Radiation (from storage tanks, N-16 sources, etc.) The component of dose to the skin due to direct radiation will be determined by thermoluminescent dosimeters (TLDs). 5.6.2 Inhalation Dose The inhalation dose to the skin (only tritium is considered) will be determined at the calculational locations for each age group according to the methods outlined in Sections 2.0 and 3.0 of this manual. 5.6.3 Ingestion Pathway (cow milk, goat milk, meat, vegetation) The dose to the skin via the ingestion pathway (only tritium and C-14 considered) will be calculated at the consumer locations for the consumers at risk. If no milk pathway exists in a sector, the dose via this pathway will be treated as < 1 mrem /yr. 5.6.4 Skin Neble Gas Immersion Dose This dose will be calculated in accordance with Section 3.4.1.2b for the maximally exposed MEMBER OF THE PUBLIC in the limiting sector (s). 5.6.5 Ground Plane Deposition 5.6.6 Other Uranium Fuel Cycle Sources This dose from other fuel cycle sources will be treated as < 1 mrem /yr. 6.0 INTERLABORATORY COMPARISON STUDIES 6.1 Requirement Technical Specification 3.12.3 states " Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission." RSP-0008 REV. 3 PAGE 52 0F 110 l l1 6.2 . Program 6.2.1 Environmental Sample Analyses Comparison Program-1 , Environmental samples from the River Bend Station are to be analyzed by the Rivor, Bend Station. Environmental Services Group or by a qualified contractingLlaboratory. These laboratories will participate in the U.S. Environmental Protection Agency's-Environmental Radioactivity Laboratory Intercomparison Studies (Crosscheck) Program or an' equivalent program. -This participation will include all of the. determinations (sample-radionuclides combinations) that-are offered by EPA'and~that are'also included'in the licensee's environmental' monitoring program. Results of the Interlaboratory Program will be included in the Annual Radiological Environmental Operating Report. 6.2.2 Effluent Release Analyses Program RBS Chemistry Group will perform sample analyses for gamma-emitting
- radionuclides in effluent releases. The radiochemistry. laboratory will participate annually in a corporate interlaboratory comparison study or an equivalent study. The results of these studies will be provided to.the NRC upon request.
6.2.3 Abnormal Results If the GSU laboratory or vendor laboratory results lie at' greater than three (3) standard deviations from the " recognized value " an . evaluation will be performed to identify any recommended remedial actions to reduce anomalous errors. . Complete documentation on the evaluation will be available to RBS Environmental Services Group and will be provided to the NRC upon request. RSP-0008 REV. 3 PAGE 53 OP 110 ..,n,--_ , ...y-- _ - - --- _- .~ _--- _ APPENDIX A LIQUID MPC VALLTS 'l RSP-0008 REV. 3 PAGE 54 0F 110 L l l l MAXIMUM PERMISSIBLE CONCENTRATIONS IN WATER IN UNRESTRICTED AREAS l MPC MPC MPC l Nuclide* (uC1/ml) Nuclide* (uCi/ml) Nuclide* (uCi/ml) H-3 3 E-3 Y-90 2 E-5 Te-129 8 E-4 i Na-24 3 E-5 Y-91 3 E-3 Te-131m 4 E-5 ! P-32 2 E-5 Y-91 3 E-5 Te-131 None l Cr-51 2 E-3 Y-s2 6 E-5 Te-132 2 E-5 l Mn-54 1 E-4 Y-93 3 E-5 I-130 3 E-6 Mn-56 1 E-4 Zr-95 6 E-5 I-131 3 E-7 Fe-55 8 E-4 Zr-97 2 E-5 I-132 8 E-6 Fe-59 5 E-5 Nb-95 1 E-4 I-133 1 E-6 Co-57 4 E-4 Nb-97 9 E-4 I-134 2 E-5 Co-58 9 E-5 Mo-99 4 E-5 I-135 4 E-6 Co-60 3 E-5 Tc-99m 3 E-3 Cs-134 9 E-6 Ni-65 1 E-4 Tc-101 None Cs-136 6 E-5 Cu-64 2 E-4 Ru-103 8 E-5 Cs-137 2 E-5 Zn-65 1 E-4 Ru-105 1 E-4 Cs-138 None Zn-69 2 E-3 Ru-106 1 E-5 Ba-139 None Br-82 4 E-5 Ag-110m 3 E-5 Ba-140 2 E-5 Br-83 3 E-6 Sn-113 8 E-5 Ba-141 None Br-84 None** In-113m 1 E-3 Ba-142 None Br-85 None Sb-122 3 E-5 La-140 2 E-5 Rb-86 2 E-5 Sb-124 2 E-5 La-142 None Rb-88 None Sb-125 1 E-4 Ce-141 9 E-5 Rb-89 Ncne Te-125m 1 E-4 Ce-143 4 E-5 Sr-89 3 E-6 Te-127m 5 E-5 Ce-144 1 E-5 Sr-90 3 E-7 Te-127 2 E-4 Pr-144 None. Sr-91 5 E-5 Te-129m 2 E-5 W-187 6 E-5 Sr-92 6 E-5 Np-239 1 E-4
- If a nuclide is not listed, refer to 10CFR20 Appendix B and use the most conservative insoluble / soluble MPC where they are given in Tsble II, Column 2.
** None (as per 10CFR20, Appendix B) "No MPC limit for any single radionuclides not listed above with decay mode other than alpha emission or spontaneous fission and with radioactive half-lives less than 2 hours." RSP-0008 REV. 3 PAGE 55 0F 110 _-____-______D , _ _ , - , . ..-,-.v,. 5 - ' ~ E-APPENDIX B 1 LIQUID ENVIR0h?iEhTAL DOSE TRANSTER FACTORS F- A i1 i f l ) I I -l -i 4 l J i i I i i RSP-0008 REV. 3 PAGE 56 0F 110 l _ _ _ _ _ _ - - _ - _ _ _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ - .h / c s , m: . % : gl - TABLE B-1 , LIQUID EFFLUENT DOSE PARAMETERS (Page'1)' Ag , mrem /hr,per uC1/ml. Radionuclides- Total Body Critical Crgan- 'Na 6.13E02 6.13E02 - P-32 1.87E06 4.84E07-Cr-51 4.31 1.08E03 Mn-54' 4.56E04 7.32E05 Mn-56 1.07E03 1.92E05 l Fe-55: 9.14E02- 5.67E03 l.< Fe-59 8.07E03 7.02E04' L Co-58 4.02E02 3.64E03 Co-60: 1.14E03 9.68E03: N1-63 1.29E03 3.85E04 Ni-65 9.28 5.16E02 Cu 64 1.36E01 2.47E03 Zn-65 7.30E04- 1.61E05. Zn-69 1.40E01. 2.07E02 SI-89 1.20E03 4.19E04. Sr-90 2.50E05 1.03E06 Sr-91 3.01E01 3.68E03 Sr-92 1.30E01 5.80E03 Y-91 2.37 '4.89E04 Y-92' 1.56E-02. .9.32E03' ,Y-93 4.66E-02. 5.35E04 Zr-95 7 74E-02 3.62E02 Zr-97! 1.82E-03 1.23E03 Nb-95 1.34E02 1.51E06= l Mo-99: 6.84 2.96E02-Tc-99m 3.45E '1.60E01 >Tc-101 1.39E-01 -2.55E-01 Ru-103 1.55E01 4.21E03 Ru-105- 1.19 1.84E03 zRu-106 6.78E01 3.47E04 Ag-110M 1.21E01 8.35E03 Te-129M 6.08E03 2.57E05 l Te-131M 3.13E03 3.73E05 Te-132 6.80E03 3.42E05 Ba-139 3.51E-01 2.13E01 L Ba-140 1.64E02 5.17E03 Ba-141 1.96E-01 5.82 Ba-142 1.66E-01 2.63 l La-142 9.13E-03 2.68E02 Ce-141 4.11E-01 1.39E04 Ce-143 7.73E-02 2.61E04 Ce-144 1.50E01 9.44E04 RSP-0008 REV. 3 PAGE 57 0F 110 . .. . ' .1 I-p. TABLE B-1 LIQUID EFFLUENT DOSE PARAMETERS (Page 2) Radionuclides Total Body Critical Organ L Pr-143 2.87E-01 2.54E04 Nd-147 2.74-01 2.20E04 W-187 8.66E01 8.12E04 Np-239 1.70E-02 6.22E03 Br-85 4.80E01 6.91E01 Br-84 6.22E01 6.22E01 I-131 1.57E02 8.98E04 I-132 8.75 8.75E02 I-133 3.49E01 1.68E04 I-134 4.74 2.30E02 I-135 1.97E01 3.52E03 Rb-89 1.51E02 2.15E02 Cs-134 6.49E05 7.94E05 Cs-136 9.93E04 1.38E05 Cs-137 3.83E05 5.85E05 Cs-138 2.90E02 5.85E02 H-3 2.80E-01 2.88E-01 RSP-0008 REV. 3 PAGE 58 0F 110
- l-TABLE B-2 CALCULATIONAL ASSUMPTIONS FOR A g
~ A g- = 1.14 x 10 (Uw/Dw + UpBF g + UyBIg)DFg Uw = 730 kg/yr adult water consumption (Reg. Guide 1.109 Table E-5) Dw = 24,800 dilution factor for potable water intake (RBS Environmental Report page 5.4-5) U F = 21 kg/yr adult fish consumption (Reg. Guide 1.109 Table E-5) BF i = bioaccumulation factor for nuclide i in fish (pCi/kg per pCi/E) RBS Environmental Report Table 5.4-3 (Table B-3) Reg. Guide 1.109 Table A-1 RBS ER-CPS Appendix N, " Stable Element Study" U I = 5 kg/yr adult invertebrate consumption (Reg. Guide 1.109 Table E-5) BI I = bioaccumulation factor for nuclide i in invertebrates (pCi/kg per pCi/t) Reg. Guide 1.109 Table A-1 DF i = dose conversion factor for nuclide i for adults in pre-selected organ t (mrem /pCi). Reg. Guide 1.109, Table E-11. RSP-0008 REV. 3 PAGE 59 0F 110 _.1 g . .., _ ;f' l APPENDIX C KgL AIR g DOSE TRANSFER FACTORS I RSP-0008 REV. 3 PAGE 60 0F 110 e / t TABLE C-1 DOSE TRANSFER FACTORS FOR EXPOSURE TO A SEMI-IhTINITE CLOUD OF RADI0 ACTIVE NOBLE GASES D OFS E' 'T-R A N S F E R 'F.A C T 0 R S Gamma Beta- Beta and Gamma K g Lg (L+1.1M)g mrem- mrem mrem Nuclide uCi sec/m3 uCi sec/m3 uCi sec/m3 - Kr-83m 2.4E-9 --- 6.7E-7 Kr-85m' '3.7E-5 4.6E-5 8.9E-5 Kr-85L 5.1E-7 4'.2E-5 4.3E-5 ' Kr-87 1.9E-4 3~1E-4 . 5.3E-4 - Kr-88. 4.7E-4 7.5E-5 6.0E-4 Kr-89 5.3E-4 3.2E-4 9.3E-4 Kr-90 4.9E-4 2.3E-4 8.0E-4 Xe-131m- 2.9E-6 1.5E-5 2.0E-5 Xe-133m' 8.0E-6 3.1E-5 4.2E-5 Xe-133 9.3E-6 9.7E-6 2.2E-5 ~Xe-135m. 9.9E-5 2.3E-5 1.4E-4 Xe-135 '5.7E-5 5.9E-5 1.3E-4' ' Xe-137 4.5E-5 3.9E-4 4.4E-4 Xe-138 2.8E-4 1.3E-4 4.5E-4 Ar-41 2.8E-4 8.5E-5 4.0E-4 AIR DOSE TRANSFER FACTORS Gamma Beta M.g Ng mrad mrad Nuclide uCi sec/m3 uCi sec/m3 -Kr-83m 6.1E-7 9.1E-6 Kr-85m 3.9E-5 6.2E-5 Kr-85 5.4E-7 6.2E-5 Kr-87 2.0E-4 3.3E-4 Kr-88 4.8E-4 9.3E-5 Kr-89 5.5E-4 3.4E-4 Kr-90 5.2E-4 2.5E-4 Xe-131m 4.9E-6 3.5E-5 Xe-133m 1.0E-5 4.7E-5 Xe 133 1.1E-5 3.3E-5 Xe-135m 1.1E-4 2.3E-5 Xe-135 6.1E-5 7.8E-5 Xe-137 4.8E-5 4.0E-4 Xe-138 2.9E-4 1.5E-4 i- Ar-41 2.9E-4 1.0E-4 l l Ref. Regulatory Guide 1.109, Revision 1 Table B-1. RSP-0008 REV. 3 PAGE 61 0F 110 b, , OL i e , R . .. \ l' . TABLE C-2 . TECHNICAL BASES FOR EFFECTIVE DOSE FACTORS-The evaluation of doses due to-releases of radioactive material to the -atmosphere can be sCmplified by the'use of effective dose transfer factors instead of using dose factors which are radionuclides specific. These effective factors,-which are based on the typical radionuclides distribution in_the relsases, can be applied to the total radioactivity released to approximate the dose in the environment, i.e.,'instead of having to sum the isotopic distribution multiplied by-the isotope specific dose factor only a single multiplication (K,ff, (L.+ 1.1M),ff, M,ff, or N,ff times the total quantity of radioactive material released) would be.needed. This approach provides a reasonable estimate of.the actual dose while eliminating the need for a detailed calculational technique. Use of effective dose factors should only be used if isotopic analyses are not available (i.e., prior to initial criticality), if the relative abundances lof the noble gas isotopic mixture are similar to -those listed in Appendix D or if the relative abundances of the noble gas isotopic mixture are similar to those listed in the previous Seminannual Radioactive Effluent Release Report. (Referenc's 1.2.19) . Determination'of Effective Dose Factors -The effective dose transfer fr.ctors should be based on past operating data. The radioactive effluent distribution for the past years can be used to derive single effective factors by the following equations: n K,ff =I Kg . f g C.2-1 i=1 _ where: K,ff = The effective total body dose factor due to gamma emissions from all noble gases released. Kg = The total body dose factor due to gamma emissions from each noble gas radionuclides "i" released, f g = The fractional abundance of noble gas radionuclides "i" of the total noble gas radionuclides. 'SP-0008 . REV. 3 PAGE 62 0F 110 TABLE C-2 TECHNICAL BASES FOR EFFECTIVE DOSE FACTORS (Continued)* n (L + 1.1 M) ,ff = I (Lg + 1.1 Mg ) . fg C.2-2 i=1 where: (L + 1.1. M),ff = The effective skin dose factor due to beta and gamma emissions from all noble gases released. (Lg + 1.1 Mg ) = The skin dose factor due to beta and gamma emissions from each noble gas radionuclides "i" released. n M,ff = I Mg . f C.2-3 f i=1 where: H,ff = The effective air dose factor due to gamma emissions from all noble gases released. Mg = The air dose factor due to gamma emissions from each noble gis radionuclides "i" released. n N,ff =I Ng . f g C.2-4 i=1 where N,ff = The effective air dose factor due to beta emissions from all noble gases released. Ng = The air dose factor due to beta emissions from each noble gas radionuclides "1". To provide an additional degree of conservatism, a factor of 0.8 is introduced into the dose calculation process when the effective dose transfer factor is used. This added conservatism provides additional assurance that the evaluation of dose by the use of a single effective factor will not significantly under-estimate any actual dose in the environment. Each year the dose factors should be determined and the average annual values be used. RSP-0008 REV. 3 PAGE 63 0F 110 l: t ' f l l TABLE C-3 EFFECTIVE DOSE FACTORS FOR NOBLE GASES TOTAL BODY EFFECTIVE DOSE -'K,ff Main Plant Radwaste Building Exhaust Duct
- Exhaust Duct Year K,ff (mrem-m /uCi-sec) K,ff (mrem-m /uCi-sec) 1 1
Projected ** 5.56E (-5)*** 8.05E (-5)
- Main Plant exhaust duct contains contributions from Fuel Building.
** Projected values from RBS FSAR. When RBS becomes operational, actual release rates reported in semi-annual effluent report should be used to generate effective dose factors. *** 5.56E (-5) = 5.56 x 10' l l RSP-0008 REV. 3 PAGE 64 0F 110 l TABLE C-4 ETTECTIVE DOSE TACTORS FOR NOBLE GASES SKIN ETTECTIVE DOSE (L + 1.1 M),ff Main Plant Radwaste Building Exhaust Duct
- Exhaust Duct Year (L+1.1M),ff (mrem-m /uCi-sec) (L+1.1M),ff (mrem-m /uCi-sec)
Projected ** 1.36E (-4) 1.59E (-4)
- Main Plant exhaust duct contains contributions from Fuel Building.
** Projected values from RBS TSAR. When RBS becomes operational, actual release rates reported in semi-annual effluent report should be used to generate effective dose factors. l RSP-0008 REV. 3 PAGE 65 0F 110 TABLE C-5 l EFFECTIVE DOSE FACTORS FOR NOBLE GASES AIR DOSES M,ff and N,ff Main Plant Redwaste Building Exhaust Duct * - Exhaust Duct 3 3 (mrad-m /uCi-sec) (mrad-m /uCi-sec) Gamma Air Beta Air Gamma Air Beta Air "eff N,ff M,ff N,ff Projected ** 5.96E(-5) 8.99E(-5) 8.07E (-5) 7.40E (-5)
- Main Plant exhaust duct contains contributions from Fuel Building.
** Projected values from RBS TSAR. When RBS becomes operational, actual release rates reported in semi-annual effluent report should be used to generate effective dose factors. RSP-0008 REV. 3 PAGE 66 0F 110 A' ., y APPENDIX D EXPECTED GASEOUS' RADIONUCLIDES MIXTURE RSP-0008 REV. 3 PAGE 67 0F 110 il. ,1 i I EXPECTED RELEASE OF RADI0 ACTIVE NOBLE GASES g IN GASEOUS EFFLUENTS FROM RIVER BEND STATION USAR* I ; i -Containment Building ** , Radwaste Building Nuclide Ci/yr Fraction Ci/yr Fraction j Kr-83m 4.7E.(-2) 1.07E (-5) <1 --- Kr-35m 218 0.050 <1 --- ) Kr-85 210 0.048 <1 --- 1 Kr-87 14.2 0.003 <1 --- Kr-88 47.2 0.011 <1 --- Kr-89 118 0.027 29 .03 l Xe-131m 21 0.005 <1 --- l Xe-133m 6.6E (-2) 1.504E (-5) <1 --- Xe-133 2,340- 0.533 220 ,19 , Xe-135 693 0.158 280 .24 i Xe-135m- -140 0.0L2 530 .46 l Xe-137 '380 0.087 83 .07 ) Xe-138 208 0.047 2 1.75E (-3) i 4,389. 1.0000 1,144 .99 l
- RBS USAR Table 11.3-1
** Containment Building contains releases from Fuel Building 1 l RSP-0008 REY. 3 PAGE 68 0F 110 I . APPENDIX E -X/Q AND D/f} VALUES FOR RESTRICTED AREA BOUNDARY RSP-0008 REV. 3 PAGE 69 OF 110 .1 ,i l Long Term Diffusion Estimates E.1 Objective Annual average CHI /Q and D/Q estimates for continuous and intermittent releases were calculated for.each of the sixteen 22.5-deg sectors at . receptor locations use ' to determine the maximum individual and popu-lation dose receptors. The methodology described in Reg drtory Guide 1.111, Rev. 1 provided guidance for the aforementioned analysis. The resultant CHI /Q and D/Q values for the maximum individual dose receptors are displayed in Appendix F. E.2 Calculation Techniques Nomenclature ~ 2.032 = (2/2) I (2w/16) (dimensionless)' s = 3.14159... (dimensionless) exp. = 2,71828.. (dimensionless) = E T ntrainment coefficient (dimensionless) D = Terrain recirculation factor T (dimensionless) x = Downwind receptor distance (m) = e, Vertical dispersion (plume spread) coefficient (m) u 30 = 30-ft average wind speed corresponding to a given hour of onsite meteoro- ~ logical data (m sec ) 150 = 150-ft average wind speed corresponding to a given hour of onsite meteoro-logical data (m sec~1) = Average concentration (CHI /Q) normalized by source strength (see m'3) RSP-0008 REV. 3 PAGE 70 0F 110 4 h (CHI /Q =. Depleted CHI /Q '(see m~3) D )' Fg = ' Momentum flux :(mksec~3): '=: : Maximum adjacent-b b building height '(m). .h = . Release height: (m) r - h, = Effective release height- .(m)- ;-, h = N nbuoyant plume rise (m) pr h '= . Topographic height of receptor above plant grade (m) d =. Stack or. vent diameter- (m) u,. = Efflux. velocity (m sec~I) N- = Total number of valid hours of onsite wind. data 'in all sectors for appli- . cable averaging period.. (dimensionless) 6/Q = Relative deposition rate normalized by source ~ strength (m ) = Relative deposition per D/Q '~ unit area normalized by source strength' (m-2) l G = Ground release (subscript) (dimensionless) l i- = .Index'for atmospheric stability group (Classes A through G) (dimensionless) .j = Index for number of hours (dimensionless) y k = Index for a particular receptor distance (dimensionless) + t = Index for a particular 22.5-deg sector. (dimensionless) n- = Number of hours onsite wind data in a particular 22.5-deg sector (dimensionless) ~ S- = Stability parameter (sec ) [-. RSP-0008 REV. 3 PAGE 71 0F 110 4 L y E.3 : CHI /Q Modeling. Technique-Annual. average values of relative concentration were calculated for continuous-ganaous releases of activity from the containment building ^ vent.and the radwaste building. vent.according to the straight-line airflow (Gaussian)!model described.in Regulatory Guide 1.111, Rev. 1. An adjustment wasinade to the model to characterize the regional airflow pattern. The equation of this model is as follows: (emy"'.2.onf(.a_)k U "
- 0
% .4.s.#($) G EM-g o 30('a*+eh*/w)4 13e e,g Since the River Bend Station site is located in relatively open. , terrain,.the terrain recirculation factor (D)k (Presented in Figure 2. of Regulatory Guide 1.111) was applied. ~ The entrainment coefficient.(E T ) is a function of the ratio of efflux .vel oc ity (u,)_to elevated wind speed (u150) f r the conditionally elevated release points. For vent releases occuring below the level of a nearby structure, 100-percent.downwash (total entrainment) is conservatively assumed (ET * . 1). For vent releases occuring between 1 and 2 times the height of a nearby structure, a conditionally elevated release is assumed, and the - entrainment coefficient is defined as follows: l -E = 0.0'when u,/E 150 > 5.0 (totally elevated)- T = E T 0.30-0.06 (u,/u150) 1 when 1.5 < u,/u150 5 5.0 (partially entrained) = E T 2.58-1.58 (u,/u150) when 1.0 < u,/u150 5 1.5 (partially entrained) E T = 1.0 when u,/u150 5 1.0 (totally entrained) RSP-0008 REV. 3 'PAGE 72 0F 110 _-_..__-_.____________________________._________m_______._____. Within'5 km in each downwind sector, Equation E.3-1 was evaluated by sector at.the property and restricted area boundaries and nearest a resident, vegetable garden, milk cow, and meat animal. There were no !6 . goats whose milk is consumed in the area'of interest. This evaluation was performed for each continuously emitting release point and the " > intermittent release from the mechanical-vacuum pump with onsite data L collected during the period of March 17, 1977.through March'16, 1979. .e The effective release height was-computed from the following equation: , H h e = hr - (h t)k + hpr E.3-2 .Where the'downwash correction factor (as defined by Equation (5) in Regulatory Guide 1.111,. Rev.1) is included in the equation for h pr (see Equation E.3-4). L Values'of_ topographic heights were conservatively assessed as the maximum height.within a particular annulus-sector (annsect). An annsect is an' area bounded by a 22.5-deg sector and any two radial distances from the release point. For A-D. stability conditions, plume rise for nonbuoyant sources was calculated by the following algorithm: when: > 1.5 "el 150 _ 2/3 1/3 h " 'I'44 (x/d) d E.3-3 r I'el"150) RSP-0008 REV. 3 PAGE 73 0F 110 '} : when: "e/ 150 2/3 1/3 h " I'44 l_, (x/d) (d-3) [1.5 - (u,/u150) d) E.3-4 pr ' ("e "150) and h 5 3 (u,/u150) E.3-5 p The result from Equation E.3-3 or E.3-4 (whichever condition exists) is then compared to Equation E.3-5 and the smaller value of h is used. For E-G stability conditions, Equations E.3-3, E.3-4, and E.3-5 are compared with: h = 4 (F,/s) and, 1/3 -1/6 h = 1.5 (F,/u,,150) 0 pr where: (u,) d F, = 4 and the smallest value was chosen. RSP-0008 REV. 3 PAGE 74 0F 110 j In the ground level portion of Equation E.3-1, the vertical dispersion term: (a + 0.5hb I") 1,k ! was constrained to be less than or equal to 1.732a g ik E.4 (CHI /Q) and D/Q Modeling Techniques Annual average depleted relative concentration values were ccuservatively assumed to be equal to annual average relative concentration values (CHI /Q == (CHI /Q)D). Therefore, no credit was taken for attendant plume depletion of radiciodines and particulate. Annual average relative deposition values were calculated using Regulatory Guide 1.111, Rev. I with the following equation: )[h)x b 13 1 cx n1 a , For the conditionally elevated release points, Figures 6 through 9 of Regulatory Guide 1.111, Rev. I were used to calculate the (6/Q)g and (6/Q)g values, while for the ground level release points, Figure 6 was utilized to calculate the (6/Q)g value. E.5 Methodology Employed for Intermittent Release The methodology employed in the calculation of intermittent release CHI /Qs and D/Qs was as follows:
- 1. Two-hour sector-averaged CHI /Q values were calculated without terrain recirculation factors.
- 2. The 15 percent, I hour value was plotted at 2 hours on log-log coordinates, while the annual average value was plotted at 8,760 hr. A straight line connecting the two points was drawn.
- 3. Log-log interpolation based on total ground intermittent release hours versus annual hours yielded a CHI /Q multiplier.
- 4. The multiplier was applied to annual average CHI /Q and D/Q values to obtain intermittent CHI /Q and D/Q values.
For River Bend Station, a 320 hr/yr intermittent release through the containment building vent from the mechanical vacuum pump was evaluated. RSP-0008 REV. 3 PAGE 75 OF 110 .. I l l F g TABLI E-1 ~ 3 ANNUAL AVERAGE CHI /Q VALUES x 10 (sec/m ) . FOR RESTRICTED AREA BOUNT)ARY ' Main Plant Exhaust Radwaste Building Sector Duct (Continuous) Exhaust Duct-(Continuous) L S 11.4 105. SSW 19.7 186 SW 16.4 215 WSW 19.5 326 W 23.6 654 WNW 33.1 421 NW 15.7 262 NNW 14.8 138 N 18.8' 180 NNE 24.9 211 NE 16.6 150 ENE 12.2 146 E 9.07 168 ESE 10.4 154 SE 8.19 93.1 SSE 7.69 45.6 RSP-0008 REV. 3 PAGE 76 0F 110 ] .. TABLE E-2 ~ ANNUAL AVERAGE D/Q VALUES x 10 ' (m-2) FOR RESTRICTED AREA BOUNDARY Main Plant Exhaust Radwaste Building Sector Duct (Continuous) Exhaust Duet (Continuous) S 7.61 21.4 SSW 11.3 39.6 SW 10.4 36.1 WSW' 9.79 38.5 W 13.8 68.8 WNW 18.0 50.3 NW 8.68 40.8 NNW- 10.5 24.7 N 11.8 28.6 NNE 11.2 27.1 NE 8.26 22.3 ENE 9.73 22.7 E 7.75 23.0 ESE 7.76 24.6 SE 6.60 17.2 SSE 5.34 11.8 l RSP-0008 REV. 3 PAGE 77 0F 110 _ _ - _ _ ___ - __--_____ D L, . APPENDIX F MAXIMUM X/Q AND D/Q VALUES FOR INDIVIDUAL. LOCATIONS RSP-0008 REV. 3 PAGE 76 0F 110 S 0 N 1 9 3 11 1 O t . 5 02 . T I tu c 3 3 3 5 72 39. 1 1 8 31 f A nD 7 O L a - - - - - - U l t n. 9 C Ps 2 2 - 2 2 o 7 L u Q Q Q Q i A na / / 2 /2 /2 t C ih l l Q lQ IQ a E ax H H / li/ H/ c G E ME C C D CD CD o A S l P O D s . i L ) .h A d rt U e otr D i I p co V u ef I D c s c e N o td I 0 0 9 88 0. 3 n sa M 53 1
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>W O e e ee e e e eaeeawem a 40 3 ErwCOEEEMM4weee m A4 E QEwCONOMOUCO-- O o, , m e f::~ ' 11 is ! -l I i APPENDIX H l ^t GASEOUS MPC VALUES ~l ( l i. I' 'i t i ' 6 t f 1 k i i .i 4 i 5 RSP-0008 REV. 3 PAGE 82 0F 110 .--_________--__--_--_-__-._D it: TABLE H-1~ MAXIMUM PERMISSIBLE CONCENTRATIONS IN AIR IN UNRESTRICTED AREAS MPC. MPC Nuclide* (uCi/ce) Nuclide* (uCi/cc) Ar-41 4;E-8 Y-91 1 E-9 .Kr-83m 3 E-6 Zr-95 1 E-9 Kr-85m 1 E-7 Nb-95 3 E-9 Kr-85 3 E-7 Ru-103 3 E-9 Kr-87 2 E-8 Ru-106 2 E-10 Kr-88 2 E-8 Ag-110m 3 E-10 Kr-89 3 E-8 Sn-113 2 E-9 Kr-90 3 E-8 In-113m 2 E-7 Xe-131m 4 E-7 Sn-123 1 E-10 Xe-133m 3 E-7 Sn-126 1 E-10 Xe-133 3 E-7 Sb-124 7 E-10 Xe-135m 3 E-8 Sb-125 9 E-10 Xe-135 1 E-7 Te-125m 4 E-9 Xe-137 3 E-8 Te-127m 1 E-9 Xe-138 3 E-8 Te-129m l'E-9 H-3 2 E-7 I-130 1 E-10 P-32 2 E-9 I-131 1 E-10 Cr-51 8 E-8 I-132 3 E-9 Mn-54 1 E-9 I-133 4 E-10 Fe-59 2 E-9 I-134 6 E-9 Co-57 6 E-9 I-135 1 E-9 Co-58 2 E-9 Cs-134 4 E-10 Co-60 3 E-10 Cs-136 6 E-9 Zn-65 2 E-9 Cs-137 5 E-10 Rb-86 2 E-9 Ba-140 1 E-9 Sr-89 3 E-10 La-140 4 E-9 Sr-90 3 E-11 Ce-141 5 E-9 Rb-88 3 E-8 Ce-144 2 E-10
- If a nuclide is not listed, refer to 10CFR20 Appendix B and use the most conservative insoluble / soluble MPC where they are given in Tabic II, Column I.
** None (as per 10CTR20, Appendix B) "no MPC limit for any single radionuclides not listed above with decay mode other than alpha emission or spontaneous fission and with radioactive half-lives less than 2 hours." RSP-0008 REV. 3 PAGE 83 0F 110 I . c APPENDIX I ENVIRONMENTAL DOSE TRANSFER FACTORS FOR GASEOUS EFFLUENTS )' t. I' RSP-0008 REV 3 PAGE 84 0F 110 _m._ __ -- ._ ._ _ __ + 'm 3 r -C + JC - M ' O.O. O.O. O.O.O.O.O.O. O.@w N 4 .d N e; -0 N 0- + O- .m ..p- N + O P= w +C O C CwCCmOOwo + -+@C+@ CON @ w 3 . O J w @ # w O c N P= = @ C - r>@s@e-cN@ N - .c = .e.c e.m.mr= .m O Csa. a w
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l OFFICIAL WORK COPY L- .I .1- iS5UED l- 1 e 3 g , 4 l ,AITACHMENT:- 11 l l ODCM/IROCEDUltkl%EN51009SHEE7 NO. ., "l- l' l i ' DESCRIBE THE IhTORMATION TO BE CHANGEI, YhCEFMMrWi11 r0R iND.i COMPttrE-i L .l DESCRIPTION OF THE CHANGE (S) MADE TO THE.0DCM: l L 1. I 'l' I I .. I I I I l ; .I I l- I m ~l l. I I , l I I l l l l .I l-I .I I l' u I .I l 1 l I l- 'l l 1 1 I l- I ; l' l j i i ! l: I i l COMMENTS: l , I I i i i i 1 I l- I , I ! I I I I WILL THIS CHANGE REDUCE THE ACCURACY OR RELIABILITY OF DOSE CALCULATIONS OR l l l SETPOIhT DETERMINATIONS (TECHNICAL SPECIFICATION 6.14)? YES NO l j 'I. l I WILL THIS CHANGE REQUIRED REVISION TO LOWER TIER IMPLEMENTING PROCEDURES OR l l COMPUTER PROGRAMS. YES NO l l 1 l THESE CHANGES HAVE BEEN REVIEWED AND FOUND ACCEPTABLE, PURSUANT TO TECHNICAL l l SPECIFICATION 6.5.2. I i l I
- l. REVIEWED, RADIOLOGICAL ENGINEERING SUPERVISOR: / l l
- l. DATE I !
I I I REVIEWED, DIRECTOR - RADIOLOGICAL PROGRAMS: / I , i l DATE I i l I i l l ll l 1 l i AITACHMENT l PAGE 1 1I l l l l 1 l OF l' ll RSP-0008 I REV - 3 l PAGE 110 0F 110 l , 1 I ll l I I 1 _ _ _ _ ._- ______________________-___-___a}}