ML20056G075
| ML20056G075 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 06/30/1993 |
| From: | Tremblay L VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BVY-93-91, NUDOCS 9309010287 | |
| Download: ML20056G075 (62) | |
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VERMONT YANKEE
! Ni1 CLEAR POWER CORPORATION *
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- -(- Ferry Road, Brattleboro, VT 05301-7002 , , , .
/, X ENGINEERING OFFICE
'~ ' 580 MMN ST AE ET BOL TON. M A 01740 tbDS) 7;"d 6711 August 30,1993 -
BVY 93 -91 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 l
References:
- a. License No. DPR-28 (Docket No. 50-271)
Subject:
Vermont Yankee Effluent and Waste Disposal Semiannual Repon for the First and Second Quaners,1993
Dear Sir:
Enclosed herewith please find one copy of the Vennont Yankee Nuclear Power Corporation subject repon. This repon covers the period beginning January 1,1993 and ending June 30,1993 and is submitted in accordance with our Technical Specification 6.7.C.1.a and
l l We trust that the enclosed information is satisfactory; however, should you have any questions, please contact this office.
Very truly yours, ,
VERMONT YANKEE NUCLEAR POWER CORPORATION Q .
t%1 r. m.
Ieonard A.Tremblay,Jr.
Senior Licensing Engineer
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i cc: USNRC Region I Administrator USNRC Resident Inspector- VYNPS USNRC Project Manager- VYNPS l
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EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT FOR FIRST AND SECOND QUARTERS, 1993
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TABLE 1A Vermont Yankee Effluent and Waste Disposal Semiannual Report First and Second Ouarters. 1993 Gaseous Effluents - Summation of All Releases Est.
Quarter Quarter Total Unit 1 2 Error. %
A. Fission and Activation Gases
- 1. Total release Ci <1.29E+03 <8.16E+02 1.00E+02
- 2. Average release rate for period uCi/sec <1.64E+02 <1.04E+02
- 3. Percent of Tech. Spec. limit (1) %
B. Iodines
- 1. Total Iodine-131 Ci 2.82E-03 2.63E-03 5.00E+01
- 2. Average release rate for period uti/sec 3.59E-04 3.34E-04
- 3. Percent of Tech. Spec. limit (1) %
C. Particulates
- 1. Particulates with T-1/2 > 8 days Ci 3.38E-03 2.31E-03 5.00E+01
- 2. Aver 69e release rate for period uCi/sec 4.30E-04 2.94E-04
- 3. Percent of Tech. Spec. limit (1) %
4 Gross alpha radioactivity Ci 3.61E-06 1.44E-06 D. Tritium
- 1. Total release Ci 7.05E+00 6.96E+00 5.00E+01
- 2. Average release rate for period uCi/sec 8.96E-01 8.85E-01
- 3. Percent of Tech. Spec. limit (1) %
(1) Percent of Technical Specification limit will be provided in the Supplemental Effluent and Waste Disposal Report to be submitted per Technical Specification 6.7.C.1.
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P TABLE IB Verm7nt Yankee Effluent and Waste Disposal Semiannual Report First and Second Ouarters. 1993 Gaseous Effluents - Elevated Release Continuous Mode Batch Mode (l)
Quarter Quarter Quarter Quarter Nuclides Released Unit 1 2 1 2
- 1. Fission Gases Ci Arypton-55 Ci 1.06E-03(2) ND Arypton-65m Ci <3.72E+00 <3.64E*00 Arypton-87 Ci <2.74E+01 <2.31E+01 e.rypton-ee Ci <l.3eE+01 <1.26E+01 Aenon-133 Ci (1.70E+00 <6.80E+01 i menon-135 Ci (2.32E+01 <2.30E+01 xenon-135m Ci (1.69E+02 <1.35E+02 xencn-135 Ci (6.99E*02 <5.41E+02 Unicentifiec Ci Total for perico Ci (9.3SE+02 (8.06E+02
- 2. Iodines loaine-131 Ci 2.32E-03 2.25E-03 locine-133 Ci 1.2EE-02 1.10E-02 Ioaine-135 Ci ND ND Total for perica Ci 1.51E-02 1.33E-02
- 3. Particulates Strontium-69 Ci 9.69E-04 7.20E-04 5trontium-90 Ci 1.0SE-05 3.22E-05 Cesium-134 Ci ND ND Cesium-137 Ci 2.20E-05 6.94E-06 barium-Lantnanum-140 Ci 1.65E-03 1.07E-03 Manganese-54 Ci ND ND Cnromium-51 Ci 1.90E-06 ND Cocalt-58 Ci ND 2.39E-06 CoDalt-60 Ci 4.29E-05 3.90E-05 Cerium-141 Ci ND ND Zinc-65 Ci ND 4.49E-06 Unicentifiec Ci Total for period Ci 2.70E-03 1.55E-03 (1) There were no batch mode gaseous releases for this reporting period.
(2) F.elease of monitor calibration gas.
ND Not detected at the plant stack.
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TABLE IC Vermont Y3nkee Effluent and Waste Disposal Semiannual Report First and Second Quarters, 1993 Gaseous Effluents - Ground Level Releases Continuous Mode Batch Modell)
Quarter Quarter Quarter Quarter Nuclides Released Unit 1 2 1 2
- 1. Fission Gases Ci krypton-55 Ci ND ND trypton-65m Ci 1.39E+00 3.82E-02 Krypton-87 Ci 1.02E+01 2.61E-01 Arypton-55 Ci 5.16E+00 1.40E-01 xenon-133 Ci 6.25E-0] 1.70E-02 xenon-135 Ci 8.65E+00 2.36E-01 xencn-135m Ci 6.24E+01 1.76E+00 Aenon-138 Ci 2.61E+02 7.21E400 Unicentifiec Ci Total for perica Ci 3.49E+02 9.70E+00 ,_
- 2. Iodines locine-131 Ci 5.04E-04 3.75E-04 locine-133 Ci 3.22E-03 2.33E-03 locine-135 Ci ND ND Totai for perica Ci 3.73E-03 2.71E-03
- 3. P a rti cul a t e s Strontium-59 Ci 1.45E-04 1.61E-04 Strontium-90 Ci 2.02E-06 1.33E-06 Cesium-134 Ci ND ND Cesium-137 Ci ND .N D
! Earium-Lantnanum-140 Ci 5.31E-04 2.65E-04 Manganese-54 Ci ND ND Cnrcmium-51 Ci ND ND EoDait-56 Ci ND ND Cobalt-60 Ci ND ND Cerium-141 Ci ND ND Zinc-65 Ci ND ND unidentifiea Ci Total for period Ci 6.76F 04 4.28E-04 (1) There were no batch mode gaseous releases for this reporting period.
ND Not detected at the plant stack.
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TABLE ID Vermont Yankee Effluent and Waste Disposal Semiannual Report First and Second Guarters.1993 Gaseous Effluents - Nonroutine ReleasesII}
Continuous Mode Batch Mode Quarter Quarter Quarter Quarter Nuclides Released Unit 1 2 1 2
- 1. Fission Gases Ci Arypton-65 Ci Krypton-85m Ci Arypton-57 Ji Arypton-SS Ci xenon-133 Ci xenon-135 Ci xenon-135m Ci xenon-135 Ci Unicentifiec Ci Total for period Ci
- 2. lodines locine-131 Ci locine-133 Ci locine-135 Ci Total for period Ci
- 3. Pa rti cul ates Strontium-59 Ci Strontium-90 Ci Cesium-134 Ci Cesium-137 Ci Barium-Lantnanum-140 Ci Manganese-54 Ci Enromium-51 Ci CoDait 58 Ci CODait-60 Ci Cerium-141 Ci Zinc-65 Ci Unicentifiec Ci Total for perica ci (1) There were no nonroutine gaseous releases for this reporting period.
ND hot detected at the plant stack.
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TABLE 2A Vermont Yankee Effluent and k'aste Disposal Semiannual Report First and Second Ouarters, 1993 Liovid Effluents - Summation of All Releases There were no liquid releases during the first or second quarters of 1993.
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TABLE 2B Vermont Yankee Effluent and Waste Disposal Semiannual Report First and Second Ouarters. 1993 Licuid Effluents - Nonroutine Releases There were no nonroutine or accidental releases during the first or second quarters of 1993.
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E TABLE 3 Vermont Yankee Effluent and Waste Disposal Semiannual Report First and Second Ouarters. 1993 Solid Waste and Irradiated Fuel Shipments A. Solid Waste Shipped Off-Site for Burial or Disposal (Not Irradiated fuel) 6-Month Est. Total Unit Period Error, i
- 1. Type of Waste
- a. Spent resins filter sludges, evaporator m3 5.66E+01 17.50E+01 bottcms, etc. Ci 6.24E+02
- b. Dry compressible waste, contaminated m3 ecuipment, etc. 6.02E+00 17.50E+01 Ci 7.01E-02
- c. Irradiated components, control rods, m3 etc. Ci
- 2. Estimate of Major Nuclide Composition (By Type of Waste):
- a. Zinc-65 t 5.54E+01 Cesium-137 % 1.27E+01 Cobalt-60 1 1.19E+01 Barium-140 1 3.47E+00 Cesium-134 1 3.89E+00 lodine-131 1 3.49E+0D Manganese-54 1 2.07E+00 Nickel-63 1 2.44E+00
- b. Iron-55 i E.52E+01 Zint-65 1 1.09E+01 Cobalt-60 1 1.02E+01 Manganese-54 % 5.67E+00
- 3. Solid Weste Disposition:
Number of Shipments Mode of Transportation Destination 29 Truck Barnwell SC B. Irradiated Fuel Shipments (Disposition): None C. Supplemental Information
- 1) Class of solid waste container shipped: 17A (Unstable), 11A, 1B
- 2) Types of containers used: 12 Type A 17 Strong-Tight Container
- 3) Solidification agent or absorbent: None U2\70
APPENDIX A EFFLUENT AND WASTE D!SPOSAL SEMIANNUAL REPORT Suoplemental Information First and Second Ouarters. 1993 Facility: Vermont Yankee Nuclear Power Station Licensee: Vermont Yankee Nuclear Power Corporation IA. Technical Specification Limits - Dose and Dose Rate Technical Specification and Catecory Limit
- a. Noble Gises 3.8.E.1 Total body dose rate 500 mrem /yr 3.F.E.1 Skin dose rate 3000 mrem /yr 3.8.F.1 Gamma air dose 5 mrad in a quarter 3.8.F.1 Gamma air dose 10 mrad in a year 3.8.F.1 Beta air dose 10 mrad in a quarter 3.8.F.1 Beta air dose 20 mrad in a year
- b. Iodine-131. lodine-133. Tritium and Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days 3.8.E.1 Organ dose rate 1500 mrem /yr 3.8.G.1 Organ dose 7.5 mrem in a quarter 3.8.G.1 Organ dose 15 mrem in a year
- c. Licuids 3.8.B.1 Total body dose 1.5 mrem in a quarter 3.8.B.1 Total body dose 3 mrem in a year 3.8.B.1 Organ dose 5 mrem in a quarter 3.8.B.1 Organ dose 10 mrem in a year 2A. Technical Specification Limits - Concentration Techr.ical Specification and Cateoory Limit
- a. Noble Gases No MPC limits
- b. Iodine-131. Iodine-133. Tritium and Radionuclides in Particulate Form With Half-Lives Greater Than 8 Days No MPC limits croc A-1
Table II, Column 2): 11.0 3.8.A.1 Total noble gas concentration: 12.00E-04 uti/cc
- 3. Average Energy Provided below are the average energy (E) of the radionuclide mixture in releases of' fission and activation gases, if applicable.
- a. Average gamma energy: 1st Quarter 8.90E-01 MeV/ dis 2nd Quarter 8.40E-01 MeV/ dis
- b. Average beta energy: Not Applicable 4 Measurements and Approximations of Total Radioactivity Provided below are the methods used to measure or approximate the total radioactivity in effluents and the methods used to determine radionuclide composition.
- a. Fission and Activation Gases Continuous stack monitors monitor gross Noble Gas radioactivity -
released from the plant stack. Total Noble Gas release rates are calculated using this monitor. To determine the isotopic breakdown of the release, samples are taken of the Steam Jet Air Ejector, which is the source ges for the releases. These samples are analyzed by gamma spectroscopy to determine the isotopic composition. The isotcpic composition is then proportioned to the gross releases determined from the stack monitor to quantify the individual isotopic releases. These are indicated in Table IB and the totals of TaDie 1A.
Beginning in the fourth quarter of 1991, grab samples were obtained from the Turbine Building roof vents. Fission and activation gases, and their daughters, were not detected in grab samples during April and June. The activity of Xe-138 released from the Turbine Building roof vents was , assumed to De the same as the LLD value of Xe-138. The remainder of the gases indicated were calculated by ratioing the indicated Xe-138 to the other gases using the Steam Jet Air Ejector samples as mentioned above.
For the remainder of this reporting period, only Cs-138 was detected in these samples. The remainder of the gases indicated were calculated by assuming Cs-138 solely from the decay of Xe-138 crun A-2
and then raticing Xe-138 to the other gases using the Steam Jet Air Ejector samples as mentioned above. These results are indicated in Table IC and the totals of Table 1A.
The error involved in these steps may be approximately 100 percent.
Iodines b.
Continuous isokinetic samples are drawn from the plant stack through a particulate filter and charcoal cartridge. Beginning in the fourth quarter of 1991, continuous particulate and charcoal samples were also taken at the Turbine Building roof vents. The filters and cartridges are normally removed weekly and are analyzed for lodine-131, 132, 133, 134, and 135. The error involvec in these steps may be approximately 50 percent.
- c. Particulates The particulate filters described in b. above are also counted for particulate radioactivity. The error involved in this sample is also approximately 150 percent.
- d. Liquid Effluents Radioactive liquid effluents released from the facility are continuously monitored. Measurements are also made on a reoresentative sample of each batch of radioactive liquid effluents released. For each batch, station records are retained of tne total activity (mci) released, concentration (uCi/ml) of gross racioactivity, volume (liters), and approximate total quantity of water (liters) used to dilute the liquid effluent prior to release to the Connecticut River.
Each batch of radioactive liquid effluent released is analyzed for gross gamma and gamma isotopic radioactivity. A monthly proportional composite sample, comprising an aliquot of each batcn released during a month, is Lnalyzed for tritium and gross alpha radioactivity. A quarterly proportional composite sample, comprising an aliquot of each batch released during a quarter, is analyzed for Sr-89. Sr-90, and Fe-55.
There were no licuid releases during the reporting period.
- 5. Batch Releases
- a. Liquid There were no routine liquid batch releases during the reporting period.
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- b. Gaseous There were no routine gaseous batch releases during the reporting period.
- 6. Abnormal Releases
- a. Liquid .
There were no nonroutine liquid releases during the reporting period.
- b. Gaseous There were no nonroutine gaseous releases during the reporting period.
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APPENDIX B L10VID HOLDUP TANKS Recuirement: Technical Specification 3.8.D.1 limits the quantity of radioactive material contained in any outside tank. With the quantity of radioactive material in any outside tank exceeding the limits of Technical Specification 3.8.D.1 a description of the events leading to this condition is required in the next Semiannual Effluent Release Report per Technical Specification 6.7.C.1.
Response: The limits of Technical Specification 3.8.D.1 were not exceeded during this reporting period.
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l l APPENDIX C l
RADIDACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION Reavirement: Radioactive liquid effluent monitoring instrumentation channels are required to be operable in accordance with Technical Specification Table 3.9.1. If an inoperable radioactive liquid effluent monitoring instrument is not returned to operable status prior to a release pursuant to Note 4 of Table 3.9.1. an explanation in the next Semiannual Effluent Release Report of the reason (s) for delay in correcting the inoperability are required per Technical Specification 6.7.C.I.
Response: Since the requirements of Technical Specification Table 3.9.1 governing the operability of radioactive liquid effluent monitoring instrumentation were met for this reporting period, no response is required.
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APPENDIX D RADIDACTIVE GASE0US EFFLUENT HONITORING INSTRUMENTATION Recuirement: Radioactive gaseous effluent monitoring instrumentation channels are required to be operable in accordance with Technical Specification Table 3.9.2. If inoperable gaseous effluent monitoring instrumentation is not returned to operable status within 30 days pursuant to Note 5 of Table 3.9.2 an explanation in the next Semiannual Effluent Release Report of the reason (s) for the delay in correcting the inoperability is required per Technical Specification 6.7.C.I.
Response: Since the requirements of Technical Specification Table 3.9.2 governing the operability of radioactive gaseous effluent monitoring instrumentation were met for this reporting period, no response is required.
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APPENDIX E RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Reavirement: The Radiological Environmental Monitoring Program is conducted in accordance with Technical Specification 3.9.C. With milk samples no longer available from one or more of the sample locations required by Technical Specification Table 3.9.3 Technical Specification 6.7 C.1 requires the following to be included in the next Semiannual Effluent Release Report:
(1) identify the cause(s) of the sample (s) no longer being available, (2) identify the new location (s) for obtaining available replacement samples, and (3) include revised ODCM figure (s) and table (s) reflecting the new location (s).
Response
No changes were needed in the milk sampling locations specified in Technical Specification Table 3.9.3 due to sample unavailability, cruo E-1
4 APPENDIX F LAND USE CENSUS Recuirement: A land use census is conducted in accordance with Technical Specification 3.9.D. With a land use census identifying a location (s) which yields at least a 20 percent greater dose or dose commitment than the values currently being calculated in Technical Specification 4.8.G.I. Technical Specification 6.7.C.1 requires the identification of the new location (s) in the next Semiannual Effluent Release Report.
Response
The 1993 land use census was not performed during this reporting period. It will be completed during the next reporting period (second half of 1993).
m:uo F-1
APPENDIX G PROCESS CONTROL PROGRAM Recuirement: Technical Specification 6.12. A.1 requires that licensee initiated changes to the Process Control Program (PCP) be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made.
Response: There were no licensee-initiated changes to the Process Control Program during this reporting period.
troo G-1
APPENDIX H OFF-SITE DOSE CALCULATION MANUAL Recuirement: Technical Specification 6.13. A.1 requires that licensee initiated changes to the Off-Site Dose Calculation Manual (0DCM) be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective.
Response: The ODCM was expanded to add on-site incineration of waste oil that has low levels of contamination. This change was prompted by a Federal Register Notice (Volume 57. No. 235: Monday.
December 7. 1992. Page 57649.
- Disposal of Waste Oil by Incineration *) that allows for on-site incineration of waste oil under existing operating effluent Technical Specifications.
Prior to the inclusion of waste oil burner, the ODCM considered ground level releases from the Turbine Building roof vents.
This amendment identified that releases from the waste oil burner located in a structure next to the Turbine Building could be treated for the purpose of dose assessments by the same models as applied to the Turbine Building.
The amendment also added a new Appendix D to explain the basis for the surveillance criteria to be applied to waste oil sampling and analysis. Since waste oil prior to burning is in a liquid form. the Technical Specification effluent lower limits of detection (LLDs) for liquids are to be used to determine the radioactivity content of the material prior to its incineration.
Appendix D indicates that the use of the liquid effluent LLDs will provide sufficient sensitivity to ensure that overall site gaseous dose / dose rate limits will not be violated by the practice of oil burning due to any limitation on sample analyses. All activity determined to be present in the waste oil prior to burning is assumed to be released in the effluent gas during incineration. No retention or partitioning in the combustion chamber is assumed with respect to calculating off-site doses.
Since the methodology for determining off-site doses from oil burning is the same as that currently applied to existing ground level release points covered by the ODCH. these changes will not reduce the accuracy or reliability of any dose calculations or setpoint determinations presently controlled by the Off-Site Dose Calculation Manual.
The revised ODCM pages for the above revision are attached.
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1 VERMONT YANKEE NUCLEAR POWER STATION OFF-SITE DOSE CALCULATION MASTAL ,
- REVISION #14 Reviewed /S/27[f3 Plard) Operations Review Committee Date Approved , h- / S/2 7/93 Plant - 'e V8 I) ate' Approved 3 e / C4 dd/ / 4V-98 Vice President [ Operations Date
LIST OF AFFECTED PAGES Pace Revision Date ii 13 05/01/92 lii 12 09/19/91 iv - v 0 03/01/84 vi 12 09/19/91 vii 14 06/01/93 viii - x 12 09/19/91 1.1 10 04/04/91 1.2 - 1.5 0 03/01/84 1.6 - 1.7 3 12/22/86 1.8 12 09/19/91 I 1. 9 14 06/01/93 1.10 12 09/19/91 1.11 9 03/02/90 1.12 - 1.16 12 09/19/91 1.17 14 06/01/93 1.18 - 1.21 12 09/19/91 1.22 14 06/01/93 2.1 - 2.4 9 03/02/90 3.1 - 3.3 3 12/22/86 3.4 14 06/01/93 3.5 - 3.6 3 12/22/86 3.7 9 03/02/90 3.8 - 3.13 3 12/22/86 <
3.14 14 06/01/93 3.15 12 09/19/91 3.16 - 3.18 14 06/01/93 3.19 12 09/19/91 3.20 - 3.36 14 06/01/93 3.37 12 09/19/91 3.38 - 3.39 14 06/01/93 3.40 - 3.44 12 09/19/91 3.45 14 06/01/93 3.46 - 3.56 12 09/19/91 4.1 10 04/04/91 4.1A 9 03/02/90 4.1B 4 12/30/87 4.2 - 4.5 13 05/01/92 4.6 - 4.9 10 04/04/91 5.1 - 5.8 0 03/01/84 5.9 - 5.10 4 12/30/87 5.11 3 12/22/86 5.12 - 5.17 12 09/19/91 5.18 4 12/30/87 5.19 - 5.20 3 12/22/86 5.21 13 05/01/92 5.22 9 03/02/90 6.1 - 6.8 0 03/01/84 6.9 9 03/02/90 6.10 14 06/01/93 A1 - A3 3 g 12/22/86 A4 - A5 12 i 09/19/91 AG 3 12/22/86 A7 - AB 12 09/19/91 Revision 14 Date 06/01/93 iiia
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LIST OF AFFECTED PAGES Pace Revision Date A9 3 12/22/86 A10 - All 12 09/19/91 A12 3 12/22/86 A13 - A24 12 09/19/91 A25 - A29 4 12/30/87 B1 - C39 9 03/02/90
. D1 - D5 14 06/01/93 Revision .14 Date 06/01/93 iiib
TABLE OF CONTENTS (Continued)
Pace 4.0 ENVIRONMENTAL MONITORING PROGRAM . . . . . . . . . . . . . . . . . . 4-1 5.0 SETPOINT DETERMINATIONS . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 Liquid Effluent Instrumentation Setpoints . . . . . . . . . . 5-2 5.2 Gaseous Effluent Instrumentation Setpoints . . . . . . . . . . 5-9 6.0 LIQUID AND GASEOUS EFFLUENT STREAMS. RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS . . . . . . . . . . . . . . . . . . . 6-1 6.1 In-Plant Liquid Effluent Pathways . . . . . . . . . . . . . . 6-1 6.2 In-Plant Gaseous Effluent Pathways . . . . . . . . . . . . . . 6-3 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . R-1 APPENDIX A: Method I Example Calculations . . . . . . . . . . . . A-1 APPENDIX B: Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont Yankee . . . . . . . . . . . . . . . . . . . . . . . . B-1 APPENDIX C: Response to NRC/EG&G Evaluation of ODCM Update Through Revision 4 . . . . . . . . . . . . . . . . . . C-1 APPENDIX D: Assessment of Surveillance Criteria for Gas Releases from Waste Oil Incineration . . . . . . . . . . . . . D-1 Revision Date crut -vii-
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TABLE 1.1-5 f .Summarv of Methods to Calculate Doses to Air from Noble Gases Equation Reference Number Category Equation Section 3-21 Gamma Dose to Air from 3.7.1 Noble Gases Released D airs (mrad) - 0.022i E Of DFf from Stack 3-41 Gamma Dose to Air from 3.7.1 Noble Gases Released D airg (mrad) - 0.13 E Of' DFf from Ground Level i t 3-2 Beta Dose to Air from 3 .8.1 Noble Gases Released 0,0 irs (mrad) - 0.019 E oft DFf from Stack i 3-43 Beta Dose to Air from 3.8.1 Noble Gases Rele) sed D airg (mrad) - 0.55 E Of from Gcound Leve5 i
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l s
5 Re.ision Date enn 1-9
i TABLE 1.1-8 ;
(Continued)
Summarv of va ri ables l variable ,
Definition Units 6 '. - Release rate for radionuclide "i" uCi at the point of i'iterest sec i
- The noble ga 'adionuclide "i" release uCi rate at the plant stack sec
.GL
- The ncble gas radionuclide "i" release uCi 0 '. rate from ground level releases sec
.SJAE = The noble gas radionuclide "i" release bsec G
i rate at the steam jet air ejector
.A0G 0 '.
- The noble gas radionuclide "i" release rate uti at the exhaust of the augmented Off-Gas System sec
.STP 0 - The iodine. tritium, and particulate uti I
ra?ionuclide "i" release rate from the sec ph nt stack
.GLP
- The iodine tritium, and particulate uCi .
0 '.
radionuclide "i" release rate from sec l ground level releases l
ST - The release of noble gas radionuclide curies i O
i "i" from the plant stack g GL
- The release of noble gas radionuclide "i* from curies :
i ground level releases l STP - The release of iodine, tritium, and particulate curies O
I radionuclide "i* from the plant stack !
GLP - The release of iodine tritium, and particulate curies O
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radionuclide "i" from ground level releases l L = Liquid monitor res onse for the limiting cps R
S P ,'
concentration at t e point of discharge j 1
Revision Date unn 1-17 f
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l TABLE 1.1-12 Dose and Dose Rate Factors Specific for Vermont Yankee j for lodines, Tritium, and Particulate Releases i Stack Release Ground level Release -
Critical Organ Critical Organ Critical Organ Critical Organ ;
Dose Factor Dose Rate Factor Dose Factor Dose Rate factor i rem' mrem-s e c ' ' mrem-s e c ' )
DFGsito cj DFG sico y r-pCi j 0FG gico Ci DFG gi co y r-pCi Radio- ( , ( , , ,
nuclide l -
H-3 1.81E-04 5.70E-03 5.24E 1.65E-01 C-14 1.10E-01 3.47E+00 3.18E+00 - 1.00E+02 ;
Cr-51 3.80E-03 1.32E-01 4.33E 1.49E+00 Mn-54 4.36E-01 1.72E+01 4.86E+00 1.92E+02 Fe-55 1.97E-01 6.21E+00 2.20 E+00 - 6.94E+01 Fe-59 4.35E-01 1.44E+01 4.89E+00 1.62E+02 Co-58 2.26E-01 8.07E+00 2.53E+00- 9.05E+01 Co-60 4.76E+00 2.12E+02 5.29E+01- 2.35E+03 Zn-65 2.32E+00 7.51E+01 2.57E+01 8.33E+02 Sr-89 7.08E+00 2.23E+02 7.88E+01 2.49E+03 Sr-90 2.69E+02 8.48E+03 3.01E+03 9.49E+04 Zr-95 4.31E-01 1.42E+01 4.83E+00 1.59E+02 l Sb-124 7.86E-01 2.63E+01 8.86E+00 2.96E+02 Sb-125 7.78E-01 3.04E+01 8.66E+00 3.37E+02 1 1-131 4.80E+01 1.51E+03 5.38E+02 1.70E+04 f I-133 5.12E-01 1.61E+01 6.81E+00 2.15E+02 Cs-134 9.88E+00 s 3.28E+02 1.10E+02 3.66E+03 Cs-137 1.01E+01 3.44E+02 1.13 E+02 - 3.85E+03 Ba-140 7.021-02 3.27E+00 1.10E+00 - 3.53E+01 Ce-141 1.06E-01 3.37E+00 1.22E+00 3.88E+01 Ce-144 2.40E+00 7.60E+01 2.69E+01 8.52E+02 I l l
l 1
's'
, Revision 3 _ Date < /, /c o N \h1 i
. l time duration, this approach of limiting dose rates equivalent to the annual dose limits then assures that 10CFR20.106 limits on an annual average air .
t concentration in unrestricted areas will be met.
Each of the methods to calculate dose or dose rate are presented in ;
l separate sections of Chapter 3. and are summarized in Tables 1.1-1 to 1.1-7.
i Each method has two levels of complexity and conservative margin and are called Method I and Method II. Method I has the greatest margin and is the i I
simplest: generally a linear equation. Method II is a more detailed analysis which allows for use of site-specific factors and variable parameters to be selected to best fit the actual release. Guidance is provided but the ,
appropriate margin and depth of analysis are determined in each instance at l the time of analysis under Method II.
l The plant has both elevated and ground level gaseous release points:
the main vent stack (elevated release), the Turbine Building roof vents ;
l (ground level release), and the waste oil burner (ground level release). j l Therefore, total dose calculations for skin, whole body, and the critical i
! organ from gaseous releases will be the sum of the elevated and ground level I doses. Appendix D provides an assessment of the surveillance needs for waste l oil to ensure that off-site doses frcm its incineration is maintained within ;
the ALARA limits of the Technical Specifications.
I i
i l
Revision Date enn 3-4
3.4 Method to Calculate the Total Body Dose Rate From Noble Gases Technical Specification 3.S.E.1.a limits the dose rate at any time to the total body from all release sources of noble gases at any location at or beyond the site boundary equal to or less than 500 mrem / year. By limiting the maximum R tb to a rate equivalent to no more than 500 mrem / year, assurance is provided that the total body dose accrued in any one year by any member of the general public will be less than 500 mrem in accordance with the annual dose limits of 10CFR Part 20 to unrestricted areas.
Use Method I first to calculate the Total Body Dose Rate from the peak release rate via both elevated and ground level release points. The dose rate limit of Technical Specification 3.8.E.1.a is the total contribution from both ground and elevated releases occurring during the period of interest.
Use Method II if Method I predicts a dose rate greater than the l Technical Specification limit (i.e., use of actual meteorology over the period 1 of interest) to determine if, in fact, Technical Specification 3.8.E.1 had ,
j actually been exceeded during a short time interval.
Compliance with the dose rate limits for noble cases are continuously demonstrated when effluent release rates are below the plant stack noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site dose rate limit of Technical Specification 3.8.E.1, or a value below it, taking into account the potential contribution of releases from all ground level sources.
Determinations of dose rates for compliance with Technical Specifications (3.8.E.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Specification 3.8.E.2 is unsuccessful, or as required by the notations to Technical Specification Table 3.9.2 when the stack noble gas monitor is J
] inoperable.
Revision Date muut 3-14
4 S
g
- Appropriate or conservative plant stack monitor detector counting efficiency for the given nuclide mix (cpm /(pCi/cc)). l F - Stack flow rate (cc/sec).
6JAE - The last measured release rate at the steam jet air ejector of noble gas i ( Ci/sec).
DFB j - Total body gamma dose f actor (see Table 1.1-10).
For ground level noble gas releases, the total body dose rate is calculated as follows:
Atbg - 4.0 Ei 0 DFBi
'pci -s e c 'pci ' mrem-m 3'
Ci-m 3 (sec, pCi-yr ,
where: 0 - Ground level release rate ( Ci/sec) of noble gas.
The total body dose rate for the site is equal to Atbs + ktbg -
During periods (beyond the first five days) when the plant is shutdown and no radioactivi.ty release rates can be measured at the SJAE. Xe-133 may be used in place of the last SJAE measured mix as the referenced radionuclide to determine off-site dose rate and monitor setpoints. In this case, the ratio ofeachb tothesumofallbf^inEquation3-28aboveisassumedto '
reduce to a value of 1 and the total body gamma dose factor DFBi for Xe-133 (2.94 E-04 mrem-m 3 /pCi-yr) is used in Equations 3-5 and 3-39. Alternately, a relative radionuclide "i" mix fraction (f ) may $
be taken from Table 5.2-1 as a functionoftimeaftershutdown,andsubstitutedinplaceoftheratioofb tothesumofallbfA in Equation (3-28) above to determine the relative fraction of each noble gas potentially available for release to the total (example calculations can be found in Appendix A). Just prior to plant ,
startup before a SJAE sample can be taken and analyzed, the monitor alarm ,
i Revision Date unst 3-16 i
setpoints should be based on Xe-138 as representing the most prevalent high dose f actor noble gas expected to be present shortly after the plant returns to power. Monitor alarm setpoints which have been determined to be conservative under any plant conditions may be utilized at any time in lieu of the above assumptions.
Equations 3-5 and 3-39 can be applied under the following conditions:
- 1. Normal operations (not emergency event), and
- 2. Noble gas releases via elevated and ground level vents to the atmosphere.
3.4.2 Basis for Method i Method I may be used to show that the Technical Specification which limits total body dose rate from noble gases released to the atmosphere (Technical Specification 3.8.E.1) has been met for the peak noble gas release f rate. . (?'
Method I for stack releases was derived from Regulatory Guide 1.109 as
{
follows:
Atbs
- IE+06 SF [X/0]y E01ST g731 l (3-6) 3
'crem' 'pci ' () sec' 'pCi ' mrem-m
- f. ( yr , ( Ci , g3 sec, p C1 -y r ,
{
where:
57 - Shielding factor - 1.0 for dose rate determination.
[X/0]S - Maximum annual average gamma atmospheric dispersion factor
[ for stack (elevated) releases 6.9BE-07 (sec/m3 )
Revision Date nnn -
3-17 r
1 __ _ _ -__
l l
( - Release rate from the plant stock of noble gas *i" ( Ci/sec).
6fT 3
DFB j ~- Gamma total body dose f actor, (*[**-h*r ). See Table 1.1-10.
Equation 3-6 reduces to:
Atbs - 0.70 ][i ojT gggi (3-5) f 3 3
f mrem' pCi-sec f pCi ' mrem-m yr 3 sec , ( pCi-yr ,
( Cisn s k For ground level releases, the ground level maximum annual average gamma 3
atmospheric dispersion factor - 3.95E-06 sec/m , thus leading to:
i Atbg - 1E+06
- 3.95E-06 ]p Of DFBj 1
(3-39)
Rt eg - 4.0 }p Of' DFBj
?
1 The selection of critical receptor, outlined in Section 3.10, is inherent in Method I, as are the maximum expected off-site annual or long-term P.verage atmospheric dispersion factors. Due to the holdup and decay of gases allowed i l in the A0G. off-gas concentrations at the plant stack during routine plant :
operations are usually too low for determ'ination of the radionuclide mix at the plant stack. It is then conservatively assumed that most of the noble gas i i
activity at the plant stack is the result of in-plant steam leaks which are 4 removed to the plant stack by building ventilation air flow, and that this air
! flow has an isotopic distribution consistent with that routinely measured at the SJAE.
! Regarding the calculation of ground level release doses from the Turbine l Building roof vents or waste oil burner vent, the ground level atmospheric {
dispersion factors are based on the same methodologies as used for the stack dispersion factors -
l l
Revision Date ;
3-18 l unn i
3.5 Method to Calculate the Skin Dose Rate from Noble Gases Technical Specification 3.8.E.1.a limits the dose rate at any time to the skin from all release sources of noble gases at any location at or beyond the site boundary to 3.000 mrem / year. By limiting the maximum Rskin to a rate equivalent to no more than 3.000 mrem / year, assurance is provided that the skin dose accrued in any one year by any member of the general public Is much less than 3.000 mrem.
Use Method I first to calculate the Skin Dose Rate from both elevated and ground level release points to the atmosphere. The dose rate limit of Technical Specification 3.8.E.1.a is the total contribution from both ground and elevated releases occurring during the period of interest.
Use Method II if Method I predicts a dose rate greater than the Technical Specification limits (i.e., use of actual meteorology over the period of interest) to determine if, in f act. Technical Specification 3.8.E.1 had actually been exceeded during a short time interval.
Compliance with the dose rate limits for noble gases are continuously demonstrated when effluent release rates are below the plant stack noble gas activity monitor alarm setpoint by virtue of the f act that the alarm setpoint is based cn a value which corresponds to the off-site Technical Specification dose rate limit or a value below it, taking into account the potential contribution of releases from all ground level sources.
Determinations of dose rate for compliance with Technical Specifications (3.8.E.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Specification 3.8.E.2 is unsuccessful, or as required by the notations to Technical Specification Table 3.9.2 when the stack noble gas monitor is inoperable.
Revision Date enn 3-20
3.5.1 Method 1 The skin dose rate due to noble gases is determined by multiplyio; the individual radionuclide release rates by their respective dose factors and summirg all the products together as seen in the following Equation 3-7 (an example calculation is provided in Appendix A):
kskins " hb i 1
0F j
' mrem' Ci ' fmrem-s e c '
yr , sec, pCi-y r ,
where:
U - In the case of noble gases the noble gas release rate from the plant stack ( Ci/st c) for each radionuclide. "i". identified.
The release rate at the plant stack is based on the measured radionuclide distribution in the off-gas at the Steam Jet Air Ejector (SJAE) during plant operation when the activity at the stack is below detectable levels, and the recorded total gas effluent count rate from the Stack Gas Monitor I or II. The release rate at the stack can also be stated as follows:
g SJAE I
M 1Sg F
3- SJAE O'j -
1 D
i (3-28)
(cpm) ( CiIcc) (cc) cpm sec M - Plant stack gas monitor I or II count rate (cpm).
S
- Appropriate or conservative plant stack monitor g
detector counting efficiency for the given nuclide mix (cpm /( Ci/cc)).
F - Stack flow rate (cc/sec).
The last measured release rate at the steam jet air 6f# -
ejector of noble gas i ( Ci/sec).
DF j3 - combined skin dose f actor (see Table 1.1-10) for stack release.
For ground level releases the skin dose rate from noble gases is calculated by Equation 3-38:
Revision Date nnn 3-21
Asking - { 0 DF jg (3-38) 1 where: hf - The noble gas release rate from ground level vents to atmosphere ( Ct/sec) for each radionuclide "i" identified.
DF $g - Combined skin dose factor for a ground level release (see Table 1.1-10A).
The skin dose rate for the site is equal to kskins + Asking -
During periods (beyond the first five days) when the plant is shutdown and no radioactivity release rates can be measured at the SJAE Xe-133 may be used in place of the last SJAE measured mix as the referenced radionuclide to determine off-site dose rate and monitor setpoints. In this case, the ratio of each j tothesumofallE in Equation 3-28 above is assumed to f reduce to a value of 1, and the combined skin dost: factor DF j for Xc. 133 (4.57 E-04 mrem-sec/ Ci-year) is used in Equations 3-7 and 3-38. Al te rr.a tely ,
a relttive radionuclide *i" mix fraction (f) may be taken from Table 5.2-1 as a function of time after shutdown, and substituted in place of the ratior, of each5 E tothesumofallbj in Equation 3-38 above to determine the relative fraction of each noble gas potentially available for release to the total (example calculations can be found in Appendix A). Just prior to plant startup before a SJAE sample can be taken and analyzed, the monitor alarm setpoints should be based on Xe-138 as representing the most prevalent high dose factor noble gas expected to be present shortly after the plant returns to power. Monitor alarm setpoints which have been determined to be conservative under any plant conditions may be utilized at any time in lieu of the above assumptions.
Equations 3-7 and 3-38 can be applied under the following conditions:
)
- 1. Normal operations (not emergency event), and
- 2. Noble gas releases via both elevated and ground level vents to the atmosphere.
Revision Date 3-22 enn
. 3.5.2 Basis for Method I The methods to calculate skin dose rate parallel the total body dose rate methods in Section 3.4.3. Cniv the differences are presented here.
Method I may be used to show that the Technical Specification which limits skin dose rate from noble gases released to the atmosphere (Technical Specification 3.8.E.1) has been met for the peak noble gas release rate.
Method I was derived from Regulatory Guide 1.109 as follows:
T D 5 - 1.11 S F D air
+ 3.17 E+04 E Oi [X/0]S DFSj 1
3 (3-8) 3
' mrem' ,' mrem' 'm r a d ' 'pCi -y r ' Ci sec' mrem-m
()
yr , mrac , yr , C1 -s e c , W M 3 ( pCi-yr ,
where:
1.11 - Average ratio of tissue to air absorption coefficients will convert trad in air to mrem in tissue.
D i r - 3.17 E+04 EOi [X/0]s DF{
i 3
(3-9) 3
'p C i -y r ' 'C i ' sec mrad-m (mra d '
( yr j C 1 -s e c , yr, n3 pCi -y r now Dfinite ~D air [X/0]}/[X/0)s e
mrads y
mrad 3
r sec 3
m 3' (3-10)
(
yr , (
yr s
3 sec,
<m >
and Oi 31.54 0fT '
'Ci ' PCi-sec ' ' Ci '
W,t C i -y r , sec, t
Revision Date
5T so Askins " l 11 SF 1E+06[X/0]}E0 $ DF{
1
' mrem' ' mrem'(
f pCi '
) #sec' fpci ' mr a d-<n 3
yr j ,mr a c , Ci m 3 j( s e c , ( pC1 -y r ,
( s 5
+ 1.E+06 X/O s [ 01 DFSj 1
f 3 3
P pCi' set pCi mrem-m Ci s m 3 sec p C 1 -y r ,
substituting
[X/0]y -
6.98E-07 sec/m3 X/0 -
5.99E-07 sec/m 3 S
F
- Shielding Factor - 1.0 for dose rate determinations gives 5T kskins = 0.77 { Of DFJ + 0.60 S0$ DFSi (3-13) 1 1
'cr em ' 'pCi -s e c-mrem ' ' Ci ' trad-m 3'rpCi-sec' f pci' mrem-m 3'
(
yr s
3 C i -m -mr a d , (
sec, ( pCi-y r s pCi-m 3 (sec, ( pCi -y r
- E Of [0.77 DFJ + 0.60 DFS j]
i define DF 0.77 DF[ + 0.60 DFS$
s then Of kskins " h 0 1 1 is
' mrem' ' Ci ' 'crem-s e c '
yr , sec, Ci -y r ,
For determining combined skin doses for ground level releases, a 7X/0g - 3.95E-06 sec/m 3and an undepleted X/0 - 1.74E-05 sec/m have 3 been substituted into Equation 3-12 to give:
Revision Date ensi 3-24
}; Of (4.38 DF{ + 17.4 DFSj) i then DF g
L and R sking -
}[0 i
DF jg (3-38) where:
U - The noble gas release rate from ground level release points
( Ci/sec) for each radionuclide "1" identified.
DF - Combined skin dose factor for ground level releases (see
$g l Table 1.1-10A).
The selection of critical receptor, as outlined in Section 3.10 is inherent in Method I, as it determined the maximum expected off-site atmospheric dispersion factors based on past long-term site-specific meteorology.
Regarding the calculation of ground level release doses from the Turbine .
Building roof vents or the waste oil burner vent, the ground level atmospheric dispersion f actors are based on the same methodologies as used for the stack dispersion factors (same noble gas mix, meteorological history [1981-1985),
and meteorological models), and are for the site boundary location that will have the highest dose.
3.5.3 Method 11 If Method I cannot be applied. or if the Method I dose exceeds the limit, then Method 11 may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Revision 1 (Reference A),
except where site-specific models, data or assumptions are more ap311 cable.
The base case analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.
Analyses requiring Method 11 calculations should be referred to it4SD to be performed and documented.
Revision Date uns2 3-25
3.6 Method to Calculate the Critical Orcan Dose Rate from Todines, Tritium and Particulates with T 1/2 Greater Than 8 Days Technical Specification 3.8.E.1.b limits the dose rate to any organ, denot'ed E gg , from all release sources of I-131,1-133, 3H, and radionuclides in particulate form with half lives greater than 8 days to 1500 mrem / year to any organ. The peak release rate averaging time in the case of iodines and particulates is commensurate with the time the iodine and particulate samplers are in service between changeouts (typically a week). By limiting the maximum k
co to a rate equivalent to no more than 1500 mrem / year, assurance is provided that the critical organ dose accrued in any one year by any member of the general public will be less than 1500 mrem.
Use Method I first to calculate the critical orga, dose rate from the peak release rate via the plant vent stack. Turbine Building roof vents, and waste oil burner if in operation. The dose rate limit of Technical Specification 3.8.E.1.b is the total contribution from both ground and elevated releases occurring during the period of interest.
Use Method II if Method I predicts a dose rate greater than the Technical Specification limits (i.e.. use of actual meteorology over the period of interest) to determine if, in fact Technical Specification 3.8.E.1.b had actually been exceeded during the sampling period.
3.6.1 Method 1 The cr1+.ical organ dose rate from stack releases can be determined by multiplying the individual radionuclide stack release rates by their respective dose factors and summing all their products together, as seen in the following Equation 3-16 (an example calculation is provided in Appendix A):
STP DFG ACOS * )) 01 sico 1
(3-16)
' mrem' f Ci ' ' mrem-s ec '
yr j sec, pCi-y r , _
Revision Date eru 3-26
where:
b - Stack activity release rate determination of radionuclide "i" (Iodine-131. Iodine-133, particulates with half-lives greater than 8 days, and tritium), in Ci/sec. For i - Sr89, Sr90 or tritium use the best estimates (such as most recent measurements).
DFGsico - Site specific critical organ dose rate f actor (*[8*-s 3 .y r ) for a stack gaseous release. See Table 1.1-12.
For ground releases (Turbine Bui', ding and waste oil burner) the critical organ dose rate from Iodine, Tritium, and Particulates with T1/2 greater than 8 days is calculated as follows:
Acog = $ 0 0FG gico (3-40) 1 where:
5f - Ground activity release rate determination of radionuclide
'i' (Iodine-131. Iodine-133, particulates with half-lives greater than 8 days, and tritium), in pCi/sec. For j strontiums. Fe-55, or tritium, use the best estimates (such I as most recent measurements). For waste oil, the release rate is the total activity by radionuclide divided by the estimated burn time. (See Appendix D for surveillance criteria on waste oil burning.)
0FG g$cg - Site specific critical organ dose rate factor I mrem-sec) u yr for a ground level gaseous release. See Table 1.1-12.
The critical organ dose rate for the site is equal to Acos + b cog -
Equations 3-16 and 3-40 can be applied under the following conditions:
- 1. Normal operation (not emergency event), and
Turbine Building vents, and waste oil burner vent to the atmosphere.
Revision Date unn 3-27
3.6.2 Basis for Method I The methods to calculate critical organ dose rate parallel the total body dose rate methods in Section 3.4. Oniv the differences are presented here.
Method I may be used to show that the Technical Specification which limits organ dose rate from iodines, tritium and radionuclides in particulate form with half lives greater than 8 days (hereafter called Iodines and Particulates or "I+P") released to the atmosphere (Technical Specification 3.8.E.1.b) has been met for the peak I + P release rates.
The equation for k gg3 and k cog is derived by modifying Equation 3-25 from Section 3.9 as follows:
Ocos - E03 DFGsico I (3-17)
(mrem) (Ci)
C1 ,
t applying the conversion factor, 31.54 (ci-sec/uCi-yr) and converting 0 to in Ci/sec as it applies to the plant stack yields:
k cos = 31.54' {0 3 DFGsico (3-18)
'm r em ' 'C i -s e c ' ' C i ' ' mrem'
, yr , C 1 -y r , sec, Ci ,
Equation 3.8 is rewritten in the form:
$cos " 'EDi DFG sico (3-19)
'mr em ' ' Ci ' (mrem-sec '
g yr , sec j
( C1-yr ,
DFG sico and DFG ggeg (for Turbine Building vents and waste oil burner vent releases) incorporates the conversion constant of 31.54 and has assumed that the shielding f actor Revision Date
(SF ) applied to the direct exposure pathway from radionuclides deposited on the ground plane is equal to 1.0 in place of the SF value of 0.7 assumed in the determination of DFG sico and CFG gico for the integrated doses over time.
The selection of a critical receptor (based on the combination of exposure pathways which include direct dose from the ground plane, inhalation, and ingestion of vegetables, meat, and milk) which is outlined in Section 3.10 is inherent in Method I, as are the maximum expected off-sit? atmospheric dispersion factors based on past long-term site-specific me:eorology.
Regarding the calculation of ground level release doses from the Turbine Building roof vents or waste oil burner vent, the selection of a critical receptor is based on the site boundary with the highest ground level atmospheric dispersion factor (based on same meteorological history and models
- as used in the stack) and the assumption that all of the exposure pathways used for the stack release occur at this site boundary.
t Should Method II be needed the analysis for critical receptor critical pathway (s) and atmospheric dispersion factors may be performed with actual meteorologic and latest land use census data to identify the location of those pathways which are most impacted by these type of releases.
3.6.3 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109. Revision 1 (Reference A),
except where site-specific models, data or assumptions are more applicable.
The base care analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method 11 analysis.
Analyses requiring Method 11 calculations should be referred to YNSD to be performed and documented.
Revision Date nnu 3-29
3.7 Method to Calculate the Gamma Air Dose from Noble Gases Technical Specification 3.8.F.1 limits the gamma dose to air from all release sources of noble gases at any location at or beyond the site boundary to 5 mrad in any quarter an 10 mrad in any year. Dose evaluation is required at least once per month.
Use Method I first to calculate the gamma air dose for elevated and ground level vent releases during the period. The total gamma air dose limit of Technical Specification 3.8.F.1 is the total contribution from both ground and elevated releases occurring during the period of interest.
Use Method II if a more accurate calculation is needed.
3.7.1 Method I The gamma air dose from plant stack release is:
D irs = 0.022 $ 1 O f DF}
e 3 (3-21) pCi-yr mrad-m (mrad) (Ci) 3 k p C 1 -y r ,
u Ci-m >
where:
- total noble gas activity (Curies) released to the atmosphere Of via the plant stack of each radionuclide "i" during the period of interest.
DF[
- gamma dose f actor to air for radionuclide "i". See Table 1.1-10.
For ground level noble gas releases, the gamma air dose is calculated as:
(3-41)
D airg
= 0.13 { Of DFf 1
Revision Date enn 3-30 I
where: GL Oi - Total noble gas activity (curies) released to the atmosphere via ground level vents of each radionuclide, "i", during the l period of interest.
The gamma air dose for the site is equal to Dairs + Dairg
- l Equations 3-21 and 3-41 can be applied under the following conditions:
- 1. Normal operations (not emergency event), and
- 2. Noble gas releases via the plant stack Turbine Building vent, and waste oil burner to the atmosphere. !
3.7.2 Basis for Method 1 Method I may be used to show that the Technical Specification which limits off-site gamma air dose from gaseous effluents (3.8.F.1) has been met for releases over appropriate periods. This Technical Specification is based on the Objective in 10CFR50 Appendix I, Subsection B.1, which limits the estimated annual gamma air dose at unrestricted area locations.
Exceeding the Objective does not immediately limit plant operation but requires a report to the NRC.
For any noble gas release. in any period, the dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109 with the added assumption thatD}inite -DY [X/0]7 /[X/0):
D a rs - 3.17E+04 [X/0]T3 I 1
( DF[ (3-22) 3 mrad (mrad) ([C{;;r) g (sec/m3 ) (Ci) (y )
where:
[X/0]{ - maximum annual average gamma atmospheric dispersion factor for a stack release
- 6.98E-07 (sec/m3 )
Revision Date unu 3-31
Of - number of curies of noble gas "i" released from the plant stack which leads to:
D,}rs- 0.022 ETj DFj (3-21) 3 (trad) (DCi-vr) (Ci) (mrad-m EU '# "
)
Ci-m#
For the ground level release:
D a rg - 3.17E+04 [X/0379 hlg DF[ (3-42) where:
(X/0)79 - Maximum annual average gamma atmospheric dispersion factor for a ground level release
- 3.95E-06 sec/m 3 leading to:
Dirg-0.13 a
Ol g
DFj (3-41)
Regarding the calculation of ground level release doses from the Turbine Building roof vent and waste oil burner, the ground level atmospheric dispersion l f actors are based on the same methodologies as used for the stack dispersion f actors (same noble gas mix, meteorological history [1981-1985), and meteorological model s ) , and are for the site boundary location that will have the highest dose.
The main difference between Method I and Method 11 is that Method II would allow the use of actual meteorology to determine [X/037 rather than use the maximum icng-term average value obtained for the years 1981 to 1985.
Revision Date casi 3-32
3.7.3 Method II t
If the Method I dose determination indicates that the Technical L Specification limit may be exceeded, or if a more exact calculation is required, then Method 11 may be applied. Method II consists of the models, input data and i assumptions in Regulatory Guide 1.109. Rev.1 (Reference A). except where site-specific models, data or assumptions are more applicable.
Analyses requiring Method 11 calculations should be referred to YNSD to be performed and documented.
Revision Date ems: 3-33
3.8 Methed to Calculate the Beta Air Dose from N0ble Gases Technical Specification 3.8.F.1 limits the beta dose to air from al' release sources of noble gases at any location at or beyond the site boundary to 10 mrad in any quarter and 20 mrad in any year. Dose evaluation is required at least once per month.
Use Method I first to calculate the beta air dose for elevated and ground level vent releases during the period. The total beta air dose limit of Technical Specification 3.8.F.1 is the total contribution from both ground and elevated releases occurring during the period of interest.
Use Method II if a more accurate calculation is needed or if Method I cannot be applied.
3.8.1 Method I The beta air dose from plant vent stack releases is:
0 D
airs = 0.019 E Of 1 DFf
, s e 3 (3-23)
(trad) pi 1r (Ci) trad-m r
Ci-m' s
< 9 C ' ~1 ' i where:
DFf - beta cose factor to air for radionuclide *i". See Table 1.1-10.
Of - total noble gas activity (Curies) released to the atmosphere via the plant stack of each radionuclide *i' during the period of interest.
For ground level noble gas releases. the beta air dose is calculated as:
D airg = 0.55 0 DFf (3-43)
Revision Date
where: 09' I
- Total noble gas activity (curies) released to the atmosphere l.
via ground level vents of each radionuclide "i" during the period I of interest.
The beta air dose for the site is equal to Dairs *Oairg
- Equations 3-23 and 3-43 can be applied under the following conditions:
- 1. Normal operations (not emergency event), and
- 2. Noble gas releases via the plant stack Turbine Building vent, and waste oil burner to the atmosphere.
3.S.2 Basis for Method I This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. The methods to calculate beta air dose parallel the gamma air dose methods in Secticn 3.7.3. Only the differences are presented here.
Method I may be used to show that the Technical Specification which limits off-site beta air dose from gaseous effluents (3.8.A.1) has been met for releases over appropriate periods. This Technical Specification is based on the Objective in 10CFR50. Appendix I. Subsection B.1, which limits the estimated annual beta air dose at unrestricted area locations.
Exceeding the Objective does not immediately limit plant operation but requires a report to the NRC within 30 days.
For any noble gas release, in any period, the dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109:
0 Dairs - 3.17E+04 X/O s k oft DFf (3-24)
Revision Date czm 3-35
(mrad) ([f.]) ( ) (Ci) ( [(f )
substituting:
X/0 - Maximum annual average undepleted atmospheric s
dispersion factor for a stack release.
- 5.99E-07 sec/m 3 We have
- 0.019 OgT (3-23)
Dfrs a DFf (trad)
(D I[)
(C1)
({['! )
For the ground level release:
Dfrg-3.17E+04 (X/0)g Ofl DFf (3-44) a where:
- Maximum annual average undepleted atmospheric dispersion (X/0)9 factor for a ground level release.
- 1.74E-05 sec/m 3 leading to:
Dfrg-0.55 a Ofl DFf (3-43)
Regarding the calculation of ground level release doses from the Turbine Building roof vent and waste oil burner, the ground level atmospheric dispersion factors are based on the same methodologies as used for the stack dispersion f actors (same noble gas mix meteorological history [1981-1985]. and meteorological models), and are for the site boundary location that will have the highest dose.
Revision Date emn 3-36
4 3.9 Method to Calculate the Critical Orcan Dose from Todines. Tritium and Particulates Technical Specification 3.8.G.1 limits the critical organ dose to a Member of the Public off-site from all release sources of radioactive lodines, Tritium, l and particulates with half-lives greater than 8 days (hereafter called "I+P*) in gaseous effluents to 7.5 mrem per quarter and 15 mrem per year. Technical Specification 3.8.M.1 limits the total body and organ dose to any real member of the public off-site from all station sources (including gaseous effluents) to l 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year.
Use Method I first to calculate the critical organ dose from both elevated and ground level vent releases. The total critical organ dose limit of Technical Specification 3.8.G.1 is the total contribution from both ground level and elevated releases occurring during the period of interest.
Use Method II if a more accurate calculation of critical organ dose is needed (i.e., Method I indicates the dose is greater than the limit).
3.9.1 Method I STP 0
Ocos=J' 4 DFGsico (3-25)
(mrem) (Ci)
C1
( ,
STP 0 5 Total iodine, tritium, and particulate activity (Ci) released from the stack to the atmosphere of radionuclide "i" during the period of interest. For strontiums and tritium, use the most recent measurement.
DFGsico - Site-specific critical organ dose factor for a stack gaseous release of radionuclide 'i' (mrem /Ci). For each radionuclide it is the age group and organ with the largest dose factor. See Table 1.1-12.
Revision Date urus 3-38
The critical organ dose is calculated for ground level releases as:
GLP Drog-J: O j DFGgico 1
(3-44) mrem '
(mrem) (Ci)
C1 s ,
Total iodine, tritium, and particulate activity (Ci) released from 0}' -
ground level vents to the atmosphere of radionuclide "i" during the period of interest. For tritium, strontiums, and Fe-55 use the most recent measurement.
DFGgico- Site-specific critical organ dose factor for a ground level release of nuclide "i" (mrem /Ci). For each radionuclide it is the age group and organ with the largest dose factor. See Table 1.1-12.
The critical organ dose for the site is equal to Dcos + Dcog.
Equations 3-25 and 3-44 can be applied under the following conditions:
- 1. Normal operations (not emergency event). F
- 2. I+P releases via the plant stack. Turbine Building, and waste oil burner (see Appendix D for surveillance criteria on waste oil burning) to the atmosphere, and
- 3. Any continuous or batch release over any time period.
3.9.2 Basis for Method I This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations. (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II.
Method I may be used to show that the Technical Specifications which limit off-site organ dose from gases (3.8.G.1 and 3.8.L.1) have been met for releases over the appropriate periods. These Technical Specifications are based on Objectives and Standards in 10CFR and 40CFR. Technical Specification 3.8.G.1 is based on the ALARA objectives in 10CFR50. Appendix I.
Revision Date uns: 3-39
- 2. The point of maximum ground level air concentration and deposition of radionuclides.
The point of maximum gamma exposures (5 sector, 400 meters) was determined by finding the maximum five-year average gamma X/0 at any off-site location. The location of the maximum ground level air concentration and deposition of radionuclides (NW sector. 2900 meters) was determined by finding the maximum five-year average depleted X/0 and D/0 at any off-site location. For the purposes of determining the Method I dose f actors for iodines, tritium, and particulates. a milk animal was assumed to exist at the location of highest calculated ground level air concentration and deposition as noted above. This location then conservatively bounds the deposition of radionuclides at all real milk animal locations.
Regarding the calculation of ground level release doses from the Turbine Building roof vents or waste oil burner vent, the selection of a critical receptor is based on the site boundary with the highest ground level atmospheric dispersion f actor (based on same meteorological history and models as used in the stack) and the assumption that all of the exposure pathways used for the stack release occur at this site boundary.
3.10.2 Vermont Yankee Atmospheric Discersion Model The annual average atmospheric dispersion factors are computed for routine (long-term) releases using Yankee Atomic Electric Company's (YAEC) AE0LUS Computer Code (Reference B). AEOLUS is based, in part, on the straight-line airflow model discussed in Regulatory Guide 1.111 (Reference C). The valley in which the plant is located is considered by the model.
AE0LUS produces the following annual average atmosphe c dispersion factors for each location.
- 1. Undepleted X/0 dispersion factors for evaluating ground level concentrations:
Revision Date est 3-45
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liigurc 6-2 Cascous lif fluent Streams, Raillation Honitors, anil Itaiheaste Treatment System at Vermont Yankou 6-10 Revision Date
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APPENDIX D ASSESSMENT OF SURVEILLANCE CRITERIA ,
f FOR GAS RELEASES FROM WASTE OILINCINERATION 9 t
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APPENDIX D ASSESSMENT OF SURVEILLANCE CRITERIA FOR GAS RELEASES FROM WASTE OIL INCINERATION i
Introduction:
The Nuclear Regulatory Commission amended its regulations (10CFR20) in a Federal Register Notice (Vol. 57, No.235; pg 57649 / Monday, December 7,1992) that permitted the onsite incineration of contaminated waste oil generated at licensed nuclear power plants without the need to amend existing operating licenses. This action will help to ensure that the limited capacity oflicensed low level waste disposal facilities is used ;
efficiently while maintaining releases from operating nuclear power plants at levels which !
are "as low as reasonably achievable". Incineration of this class of waste must be in full compliance with the Commission's current regulations that restdct the release of radioac- ;
tive matedals to the environment. Any other applicable Federal, State, or local requirements that relate to the toxic or hazardous characteristics of the waste oil would also have to be satisfied.
Incineration of waste oilis to be carried out under existing effluent limits, recordkeeping and reporting requirements. Specifically, licensees must comply with the effluent release limitations of 10CFR Part 20, and Part 50; Appendix I. This includes the site gaseous pathway dose and dose rate limits contained in the plant's Technical Specifications (
Section 3.8). The dose contdbution to members of the public resulting from the onsite i
incineration of contaminated waste oil must not cause the total dose or dose rate from all l i
effluent sources to exceed the dose or concentration limits imposed by 10CFR20, l l
l Revision Date l D-2 ,
l -
s 10CFR50; Appendix I, and the Radiological Effluent Technical Specifications (RETS).
It is expected that the actual contribution to public exposures caused by waste oil burning will be 2 small fraction of the site's effluent limits, as well as a small portion of the total releases from the site.
4 Source
Description:
l Contaminated waste oil suitable for onsite incineration can be b4 rned in the Waste Oil i Burner located in the Containment Access Building (CAB). The burner has its own exhaust stack situated at the north end of the CAB roof. However, due to the locadon of the CAB adjacent to the Turbine Hall which extends above the height of the oil burner f stack, this release point is considered to be a ground level point source for modeling discharges to the environment similar to the Turbine Building Roof Vents. The building wake effects for both release points are also similar due to their common proximity to the Turbine Hall / Reactor Building complex. Consequently, the remaining variable which can differ in determining the maximum atmospheric dispersion factors for these two release
[ points is the downwind distance to the site boundary.
The site boundary distances used to analyze Turbine Building releases were determined by taking the minimum distance from the nearest point on the Turbine Building to the Site Boundary for Gaseous Effluents within a 45 degree sector centered on the compass direction ofinterest. Similar measurements for the Waste Oil Burner releases show that the Waste Oil Burner distances are all equal to or greater than the Turbine Building distances, except for two sectors (E and ESE). Since the differences in site boundary distances for these two sectors is small (less than 10%), and the resulting atmospheric Revision Date
[
D-3
{
dispersion factors for the these two sectors are approximately half of the maximum site boundary dispersion factors determined for the Turbine Building, it is unlikely that the maximum site boundary Waste oil Burner dispersion factors would exceed the maximum site boundary atmospheric dispersion factors. Therefore, for simplicu; evaluating the releases from oil incineration, the dispersion factors for the Turbine Building will also be used to access releases from this release point.
l l
l The Waste Oil Burner can process oil at a rate of 2 gal./ hour from a 500 gallon day tank.
The offgas flow rate for the burner is rated as 199 cfm. This provides an air to oil dilution during the incineration of 44,800.
1 Waste Oil Sampling / Surveillance Requirements:
1 The oil burner stack is not equipped with continuous air monitoring or sampling capability for the direct determination of radiological effluent releases during the incineration process. As a consequence, sampling and analysis of the waste oil prior to l its incineration is necessary to project the dose and dose rate consequences of burning contaminated oil. Calculations of projected dose from the incineration of total quantity l of oil to be added to the Waste Oil Burn Day Tank for each series of burns will be l performed in accordance with the methods in the ODCM and compared to the ,
I accumulated site total dose for that period before initiation of incineration. Dose rate I
j determinations will be determined by averaging the projected dose for the quantity of radioactivity determined to be present in the oil over the expected duration of the burn necessary to incinerate the total volume to be added to the Day Tank. Inherent in this l determination is the assumption that all radioactivity found to be present in each batch of l oil will be released to the atmosphere during the incineration. No retention of activity Revision Date l
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_ _ _ _ _ ~ ~
. l in the combustion chamber is assumed in calculating the offsite radiological impact.
l l
Normal sampling and analysis methods for gaseous release streams cannot be applied directly to liquids (waste oil). Therefore, the sampling and analysis requirements for liquids as identified on Technical Specification Table 4.8.1 shall be used to determine the level of contamination in waste oil. The stated Lower Limits of Detection (LLD) given on Table 4.8.1 provide assurance that undetectable levels of contaminadon up to the LLD j values will not result in a significant dose impact to the maximum offsite receptor. If i
waste oil was burned continuously for an entire calendar quarter, and the radionuclides listed in the ODCM Dose Conversion Factor Table 1.1-12 were assumed to be present in the oil at the LLD values specified in Tech Spec Table 4.8.1, the resultant maximum organ dose would amount to only 0.2% of the ALARA quanerly limit of 7.5 mrem. .
The principle limitation in the incineration of waste oil is that the site release limits contained in RETS, and implemented by the ODCM methodology, shall not be exceeded.
The use of the liquid LLDs on waste oil sample analyses provide sufficient sensitivity to l t
ensure that site dose limits will not be exceeded as a consequence of burning slightly I contaminated oil. j r
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APPENDIX 1 RADI0 ACTIVE LIQUID GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS I
Recuirement: Technical Specification 6.14. A requires that licensee initiated f major changes to the radioactive waste systems (liquid, gaseous, '
and solid) be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the '
evaluation was reviewed by the Plant Operation Review Committee.
Response: There were no licensee initiated major changes to the l radioactive waste systems (liquid, gaseous, and solid) during j this reporting period, l
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APPENDIX J ON-SITE DISPOSAL OF SEPTIC WASTE Recuirement: Off-Site Dose Calculational Manual. Appendix B requires that the dose impact due to on-site disposal of septic waste during the-reporting year and from previous years be reported to the j Commission in the Semiannual Radioactive Effluent Report filed j after January 1, if disposals occur during the reporting year, !
Response: No response is required for this reporting period. !
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