Text

Offsite Dose Calculation Manual, Revision 31
	ML17130A971
Person / Time
Site:	Prairie Island
Issue date:	10/13/2016
From:	; Northern States Power Company, Minnesota, Xcel Energy
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML17130A980	List: ML17130A969; ML17130A970; ML17130A971;
References
	L-PI-17-018
	Download: ML17130A971 (163)
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ENCLOSURE 4 PRAIRIE ISLAND NUCLEAR GENERATING PLANT OFFSITE DOSE CALCULATION MANUAL (ODCM)

REVISION 31 DATED:- 10/13/2016 162 pages to follow

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 1 of 162 PRAIRIE ISLAND NUCLEAR GENERATING PLANT OFFSITE DOSE CALCULATION MANUAL (ObCM)

DOCKET NO. 50-282 AND 50-306

Procedure may be pelformed from memory.

User remains responsible for procedure adherence.

Procedure should be available, but not necessarily at the work location.

PORC REVIEW DATE: APPROVAL:

PCR #: 01526914 10/13/2016

PRAIRIE ISLAND NUCLEAR GENERATING PLANT 1-i PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 2of162 TABLE OF CONTENTS Section Title Page RECORD OF REVISIONS .......................*................................. :........................ *............. 7 OFFSITE DOSE CALCULATIONS MANUAL INTRODUCTION .................................. 15 DEFINITIONS'. ............................................................................. ~ ................. .-............... 17 1.0 RADIOLOGICAL EFFLUENT SPECIFICATIONS AND SURVEILLANCE REQUIREMENTS ................................................................................................ 23 1 .1 Specifications ............................................................................................ 23 1.2 Surveillance Requirements ....................................... ,.............................. 23 Liquid Radwaste Treatment Systems .............................................................*27 Radioactive Liquid Effluent Monitoring Instrumentation ...................................... 28 Radioactive Liquid Effluent Monitoring Instrumentation ...................................... 28 Liquid Storage Tanks ........................................................................................ 29 Landlocked Area ............................................................................................... 29 3.0 GASEOUS EFFLUENTS ................. :....................................................... :.......... 31 Dose *Rate ........................................................................................................... 31 Dose - lodine-131, lodine-133, Tritium and Particulates ............................... 33 Dose - lodine-131, lodine-133, Tritium and Particulates ............................... 34 Gaseous Radwaste Treatment Systems .......................................................... 35 Radioactive Gaseous Effluent Monitoring Instrumentation ........................... 37 Radioactive Gaseous Effluent Monitoring Instrumentation .......................... 38 Atmospheric Steam Dump Monitoring ........................................................... 38 Atmospheric Steam Dump Monitoring ........................................................... 39 4.0 LIQUID EFFLUENT CALCULATIONS ................................................................ 41

. 4.1 Monitor Alarm Setpoint Determination ..................................................... 41 4.2 Compliance With 10 CFR 20 ........ , .......................................................... 47 4.3 Liquid Effluent Dose - Compliance with* 10 CFR 50 ................................. 49

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 3of162 4.4 References .........................................................................................*..... 53 5.0 GASEOUS EFFLUENT CALCULATIONS ........................................................ :. 55 5.1 Monitor Alarm Setpoint Determination ..................................................... 55 5.2 Gaseous Effluent Dose Rate - Compliance with 10 CFR 20 ............ ;....... 61 5.3 Gaseous Effluents - Compliance with 10 CFR 50 .................................... 64 5.4 References .................................................................................. ;........... 70 6.0 TOTAL DOSE FROM RADIOACTIVE RELEASES AND URANIUM FUEL

soURCES ............. :............................................................... ,............................ 71 7.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ..................... 75 8.0 REPORTING REQUIREMENTS~ ................. ,.. *..................... :................... :......... 79 8.1 . Annual Radioactive Effluent Report ......................................................... 79 8.2 Annual Radiological Environmenta.1 Monitoring *Report ............................ 81 8.3 Annual Summary of Meteorological Data ......... ,....................................... 83 8.4 1.ndustry Initiative on Groundwater Protection .......................................... 83 8.5 Record Retention ..................................................................................... 86 8.7 Reporting Errata in Effluent Release Reports .......................................... 87

BASIS89 2.0 LIQUID EFFLUENTS .........................................................................*...... 89 3.0 GASEOUS EFFLUENTS ..... :................................................................... 91 6.0 TOTAL DOSE .....................................................................:.................... 94 7.0 RADIOLOGICAL ENVIRONMENTAL MONITORING .............................. 95

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LIST OF TABLES Table 1.1 DELETED Table 2.2 Radioactive Liquid Effluent Monitoring lnstrumentation ............................. 101 Table 2.3 Radioactive Liquid Effluent Monitoring Instrumentation Sur\teillance Requirements ............................................................................................. 103 Table 3.1 Radioactive Gaseous Waste Sampling and Analysis Program ................. 105 Table 3.2 Radioactive Gaseous Effluent Monitoring Instrumentation ........................ 111 Table 3.3 - Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements .......................................*.: .............................. 113 Table 4.1 Liquid Source Terms ................................................................. :............... 115 Table 4.2 - Adult Ingestion Dose Values (Ait) for the Prairie Island Nuclear Generating Plant (mrem/Hr Per µCi/ml) ...................................................... 117 Table 5.1 - Monitor Alarm Setpoint Determination for PINGP .................... :................ 119 Table 5.2 Gaseous Source Terms ............................................................................. 121 Table 5.3 Critical Organ Dose Values (Pi) for Child .................................................. 123 Table 5.4 Dose Factors for Noble Gases * ................................................................ 125 Table 7.1 Radiological Environmental Monitoring Program Sample Collection .

and Analysis ............... :........................_. ................................................ :.:.... 127 Table 7.2 - Reporting Levels for Radioactivity Concentration in Environmental Samples ................................................... .-...... :........................................... 131 Table 7.3 - Detection Capabilities for Environ.mental Sample Analysis Lower Limit of Detection (LLD)(a) ............................... :.: .................................................. 133

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LIST OF FIGURES Figure 3.1 Prairie Island Nuclear Generating Plant Site Boundary For Liquid Effluents ...................... ~ .............................. ,................................................ 135 Figure 3.2 Prairie Island Nuclear Generating Plant Site Boundary For Gaseous Effluents .................................................................... :.................. 137 LIST OF APPENDICES Appendix A Meteorological Analyses ......................................................................... 139 Table A-1 Prairie Island Release Conditions ................................................. :.... 143 Table A-2 Distances (Miles) to Controlling Site Boundary Locations .................. 145 Appendix B Dose Parameters for Radioiodines, Particulates and Tritium .................. 147 Table B-1 Parameters for Cow and Goat Milk Pathways .................................... 157 Table B-2 Parameters for the Meat Pathway ..................... *......... , ....................... 159 Table B-3 Parameters for the Vegetable Pathway ...........................................*.. *: 161

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) . REV: 31 Page 7of162 RECORD OF REVISIONS Revision No. Date Reason for Revision Original June.7, 1979 1 April 15, 1980 Incorporation of NRC Staff comments and corrections of miscellaneous errors.

2 August 6, 1982 Incorporation of NRC Staff comments.

3 February 21, Change* in milk sampling location.

1983 4 November 14, Change in milk sampling location and change in cooling 1983. tower blowdown.

5 March 27, 1984 Change Table 3.2-1 6 February 14, Change in location to collect cultivated crops (leafy green 1986 veg.) and removal of meat animals from land use census.

7 July 31, 1986 Retypeand format ODCM. No change in.content.

8 January 8, 1987 Addition of discharge Canal monitor setpoint calculation.

9 June 29, 1987

Change inhalation dose factor to child and address change in land use survey.

10 April 27, 1989 Change in method for calculating liquid effluent monitor setpoints. Fix of various typ*ing errors. Change in location of two REMP sampling locations. Deletion of one REMP

sampling location.

11 Octobe.r 5, 1989 Change in Tables 3.3-6 thru 3.3-16. Appendix C equations corrected. Section 5 figures replaced. Sample point definitions corrected.

12 June 17, 1991 Change in REMP sampling locations Tables 5.1-1. Added text to address the increased volume of the new discharge pipe.

13 September 27, Incorporation of RETS as defined in PINGP Technical

. 1995 Specifications in accordance with GL 89-01 as directed by *

NUREG-1301. Change grab sampling frequency from 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when required on line monitoring equipment is out of service. Define liquid and gaseous monitor calibration. Define radiological effluent and environmental reporting and records retention.

14 May 15, 1996 Correct typing errors and Tech. Spec. references. Update dose factor tables.

15 August30, 1999 Revised Tech Spec references. Added reference to TBS Landlock. Changed environmental LLDs and reporting level values to reflect "Drinking Water Pathway." Consistent usage of Site Boundary and Unrestricted Area.

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H4 31 Page 8of162 RECORD OF REVISIONS [CONTINUED]

Revision No. Date Reason for Revision 16 August 1, 2001 Reformatted to M.S. Word. References to Northern States Power Company removed.

17 October 12, 2002 Revised to comply with Improved Technical Specifications.* Changed T.S. references, redefined monthly as at least every 31 days, removed all references to the OLD 10 CFR-20 and the MPG liquid release rate limits, increased the size of the airborne release. dose factor tables to include all nuclides listed in Reg Guide 1.109, changed REMP milk sampling description to comply NUREG 1301, and a few typographical errors were corrected.

18 June 26, 2003 Adopted airborne radio iodine and particulate sampler locations from NUREG 1301.

19 . July 8, 2005 For out-of-service effluent monitoring instrumentation, removed operational time constraints, and added reporting requirements, IAW NUREG 1301. Applicability requirement, for condensate storage tank level instrumentation, was clarified. Updated Site Boundary Map for Liquid Effluents to reflect extension of discharge piping. Various editorial changes.

20 November 6, 2006 Clarification was added to the Basis section, providing

" guidance for review and approval of monitor set point changes. Direction is that the Operations Committee

. (OC) will review and approve changes to the ODCM, which includes the methodologies for set point determination. Specific set point changes made in accordance.with theses OC reviewed and approved methodologies need not be reviewed by the OC.

21 April 20, 2007 Added the NEI Industry Initiative on Groundwater Protection recommended reporting protocol to Section 8.0, Reporting Requirements. This addition lowers the threshold for reporting of groundwater contamination and clarifies the reporting protocol.

22 June 11, 2008 Revised record retention length for various documents from 5 to Life of the Insurance Policy plus 10 years.

NRG Branch Technical Position, Rev 1, November 1979 added to the Critical Receptor Identification, as a compliant alterative approach, when this approach proves to be conservative with regards to dose. Various typographical errors with no change to intent.

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Revision No. Date Reason for Revision 23 May 29, 2009 Revised Section 8.4 based on guidance in NEI 07-07, "Industry Ground Water Protection Initiative - Final Guidance Document," August, 2007. This revision included the addition of four definitions to the "Definitions" section, an additional condition of Plant Manager discretion for voluntary communication to State and Local official, and the addition of NEI to the list of entities notified in the event of a spill or leak.

24 9/17/09 µ Symbol shows up as an empty box (D) 25 10/21/2010 Revised sections 2.11 and 4.2.1 to remove references to release of Turbine Building Sump water via the land

.locked discharge pathway. *Release to the land locked area was no longer allowed as of 1/8/10.

Added Section 8.5 and 8.6 to direct the processing of correspondence with the NRG and other government agen~9ies to be !AW.corporate directives.

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Revision No. Date Reason for Revision 26 4/07/2011 Adopted the language of Technical Specification SR 3.0.2, for section ODCM 1.2, "Surveillance

. Requirements".

The phase operability requirements, "for a Control for operation" was deleted, as undefined and unsupported.

In section 2.11, "LANDLOCK AREA", reference to NSP was changed to Prairie Island Nuclear Generating Plant.

Methodology for quantification of Carbon-14 curies generated and dose attributed, was added as section 3.5.1.

Removed "at least once per" from "The Land Use Census" frequency to read, "between the dates of May 1 and October 31" *

Entered new calculations for C-14 dose based on Regulatory Guide 1.109 and NU REG -0133 methodologies. - Calculation 8.2-9 Moved Ri tables, Historical Meteorological Joint Frequency Tables and dispersion tables to reference document H4.2, "OFFSITE DOSE CALCULATION MANUAL (ODCM) SUPPORTING DATA"

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Revision No. Date Reason for Revision 27 6/27/2012 The definition of CHANNEL RESPONSE TEST has been deleted. There is no requirement to perform a CHANNEL RESPONSE TEST; therefore the inclusion of the definition is extraneous.

The statement that C-14 will not be included in the totals when assessing compliance with specification 3.7.1.A (section 3.5.1.F) has been deleted as inappropriate.

The term "GALE Code" is defined and referenced. Beyond the definition, all subsequent use of the term "GALE Code",

GALE Code Mix or PWR GALE Code, has been changed to "source term".

Corrected DEi definition. Conversion factor basis has always been T1 D-14844.

Non-gamma emitters, previously treated as a subtraction in the liquid radiation process monitor setpoint calculations will now be treated as a factor, in a similar fashion to gamma emitters. This will reduce calculated setpoints from tt)ose previously generated by past methodology generated values.

Tritium will be accounted for by bounding calculation.

Section 4 and section 5 equations have been restructured to reflect source documentation, with NO change in methodology, other than those identified. This was done to enhance auditability.

Discharge Canal Monitor definition was revised to reflect USAR, and to .direct the maintaining of the al arm setpoint low, reflecting its function as an. atypical release monitor.

Table 5.1 specific dispersion factor values have been removed. Dispersion factors are identified as long or short

term. The ODCM defines methodology. Specific values generated by the prescribed methodology, will be maintained in supporting documentation (H4.2)

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Revision No. Date Reason for Revision 28 11/27/2013 Corrected DEi definition. DEi is a Reactor Coolant System Radiochemistry Technical Specification parameter. With the implementation of Alternate Source Term (License Amendment 206/193) the source document for the determination of conversion factors will change from TID-14844 to EPA Federal Guidance Report No. 11. Added the extra tritium samples taken for the NEI Groundwater Protection Initiative to the list of HEMP waterborne samples in Table 7~ 1.

29 8/24/14 Included various reporting criteria, not previously captured in the ODCM:

Reporting criteria when the concentration of Hquid effluents exceeds 20 times the limits, as averaged over 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Reporting criteria when the dose rates of airborne effluents exceeds 20 times the limits, as averaged over 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

. ISFSI Annual Environmental Report.

ISFSI Annual Radiological Environmental Monitoring Report.

30 2/27/2016

Added paragraph to link to ISFSI License Renewal aging management program.

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Revision No. Date Reason for Revision 31 10/14/2016 Various typographical and editorial errors Changed the terms "Operational" ~nd "Operability", to "Functional" and "Functionality" with regards to grading ODCM required equipment availability.

Revised various definitions to the standard language of Reg Guide 1.21 and NUREG 0133, with no change to requirements or intent.

Revised Surveillance Requirements of section 1.2.2 ts>

standard language of NUREG 1301, with no change to requirements or intent.

Changed dose rate limits refer~nce to Technical Specificaitons, rather than 10 CFR 20

Clarified C-14 approach, and reporting, with no change to methodology or reporting Added requirement and methodology to assess Total

Dose annually, for demonstrating compliance with 40 CFR 190. .

Changed the completion due date for the Annual Land Use Census from October 31 to September 30 Added requirement to Annual Radioactive Effluent.

Report, to include reporting of description of all leaks or spills that are communicated per the Industry Initiative on Groundwater Protection Aligned the ODCM with NUREG 1301, with regards to Ce-144 required LLD Removal of R-15 from the Radioactive Gaseous Effluent Monitoring Instrumentation tables Removed the reporting requirement for Waste Gas Instrumentation in the Annual Radioactive Effluent Report.

Align the ODCM

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Revision No. Reason for Revision 31 (CONT.) Align the ODCM with NUREG 1301 and Technical Specifications, with regards to.CHANNEL CHECK definition.

Remove the requirement in Appendix A, to determine dispersion factors in 0.1 mile increments.

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OFFSITE DOSE CALCULATIONS MANUAL INTRODUCTION The Offsite Dose Calculation Manual (ODCM) describes the methodologies and parameters used in: 1) the calculation of offsite doses resulting from radioactive gaseous and liquid effluents; 2) the calculation of gaseous and liquid effluent monitoring instrumentation Alarm/Trip Setpoints. The methodology stated in this manual is acceptable for use in demonstrating compliance with 10 CFR 20.1301 (a)(1 ), 10 CFR 50.36A, 10 CFR 50, Appendix A (GDC 60 & 64) and Appendix I, and 40 CFR 190.

The ODCM is based on "Radiological Effluent Technical Specification of PWR's (NUREG-0472, October 1978)", "Preparation of Radiological Effluent TechnicaJ Specifications for Nuclear Power Plants (NUREG-0133, October 1978)", and "Offsite Dose Calculation Manual Guidance (NUREG.:1301, April 1991 ). Specific plant procedures have been developed to implement the ODCM.

This m~nual also includes information related to the RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP). Tables 7.1, 7.2 and 7.3 designate specific sample types, reporting 'levels and lower limits of detection currently used to satisfy the sampling requirements for the REMP. *

This procedure is credited by the ISFSI License Renewal Aging Management Programs for implementing monitoring and for the methodologies and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, the calculation of gaseous and liquid effluent monitoring instrumentation alarm/trip setpoints, and information related to the REMP all of which aid in determining that the ISFSI radiation levels remain within acceptable limits. Therefore, consider ISFSI License Renewal requirements when revising this procedure (Ref. H76, ISFSI License Renewal Implementation and Aging Management Programs)

Licensee initiated changes to the ODCM:

1. SHALL be documented and records of reviews performed SHALL contain:

a. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s).

b. A determination that the change(s) maintain the level of radioactive effluent control required by 10 CFR20.1301 (a)(1 ), 10 CFR 50.36A, 40 CFR 190, 10 CFR 50,.Appendix I, and not adversely impact the ~ccuracy or reliability of effluent, .

dose or setpoint calculations.

2. SHALL become effective upon review and acceptance by the Operations Committee.

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3. SHALL be submitted to the NRG in the form of a complete legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Report for the period of the report in which the change in the ODCM was made. Each change SHALL be identified by markings in the margin of the affected pages clearly indicating the area of the page that was changed. The date (i.e., month and year) of the change SHALL be clearly indicated on the "Record of Revision" page.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

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ABNORMAL RELEASE An unplann.ed or uncontrolled release of radioactive material from the plant.

A release which results from procedural or equipment inadequacies, or personnel errors, that cou.ld indicate a deficiency.

ACTION ACTION SHALL be that part of a specification which prescribes remedial measures required under designated conditions. *

. BATCH RELEASE A BATCH RELEASE is a discharge of liquid or gaseous radioactive effluents of a discrete volume. Prior to. release, each bafoh SHALL be isolated and thoroughly mixed* for sampling and analysis.

CHANNEL CALIBRATION A CHANNEL CALIBRATION SHALL be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION SHALL encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by aliy series of ~equential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST consists of injecting a* simulated signal into the channel as close to the primary sensor as practicable to verify that it is FUNCTIONAL, including alarm and/or trip initiating action.

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CONTINUOUS RELEASE A CONTINUOUS RELEASE is the discharge of liq uid or gaseous radioactive effluents of a nondiscrete volume of a system that usually has makeup flow during the release .. CONTINUOUS RELEASES are normally sampled and analyzed either during or following tne release.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 is that concentration of 1-131 (µCi/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, .1-133, 1-134 and 1-135 actua lly present.

DEi is a Reactor Coolant System Radiochemistry T echnical Specification parameter. With the implementation of Alternate S ource Term (License Amendment 206/193) the source document for the determination of conversion factors changes from TID-14844 to EPA Federal Guidance Report No. 11.

EXCLUSION AREA BOUNDARY The EXCLUSION AREA is the area encompassed by the EXCLUSION AREA BOUNDARY at a minimum distance of 715 meters from the center of either reactor.

GALE CODE GALE (Gaseous and Liquid Effluents) Code refers to the computer modeling of plant effluents, using a combination of input data and hard wired parameters to calculate source terms. Jhe gaseou sand liquid source terms presented in the ODCM are calculated using the GALE Code and referenced to USAR tables 9~3-1 and 9.2-3.* Throughout the O DCM, the use of the terms "Liquid Source Term" or "Gaseous Source Term" w ill mean-source terms generated using the Gale code.

GASEOUS RADWASTE TREATMENT SYSTEM The GASEOUS RADWASTE TREATMENT SYSTE M SHALL be any system designated and installed to reduce atmospherically released radioactive I effluents by collecting primary coolant system offga ses from the primary system and providing for delay or holdup for the pu rpose of reducing the total radioactivity prior to release to the environment.

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Page 19of162 GROUNDWATER All water in the surface soil, the subsurface soil, or any other subsurface water.

Ground water is simply water in the groun*d regardless of its quality, including saline, brackish, or fresh water. Ground water can be moisture ih the ground that is above the regional water table in the unsaturated (or vadose) zone, or ground water can be at and below the water table in the saturated zone ..

LIQUID RADWASTE TREATMENT SYSTEM The LIQUID RADWASTE TREATMENT SYSTEM SHALL be any system designated and installed to reduce radioactive effluents by holdup or collecting radioactive materials by means offiltering, evaporation, ion exchange or chemical reaction for the purpose of reducing the total radioactivity prior to release to the environment..

LONG Tf:RM RELEASE LONG TERM RELEASES refer to releases that are generally continuous and stable* in release rate with some anticipated variation in release rate, such as is experienced in normal ventilation system effluents at a nuclear power plant.

Determination of doses due to long-term releases should use the historical annual average relative concentration (X/Q), based on meteorological data summarized.

. MEMBER OF THE pµsuc MEMBER OF THE PUBLIC means any individual except when that individual is receiving an occupational dose.

FUNCTIONAL - FUNCTIONALITY A system, subsystem, train, component, or device shall be FUNCTIONAL or have FUNCTIONALITY when it is capable of performing its specified safety

function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and *seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

POTENTIAL TO REACH GROUNDWATER SPILLS OR LEAKS with the POTENTIAL TO REACH GROUNDWATER include:

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SPILL OR LEAK directly onto native soil or fill,

SPILL OR LEAK onto an artificial surface (i.e. concrete or asphalt) if the surface is cracked orthe material is porous or unsealed., or

A SPILL OR LEAK that is directed into unlined on non-impervious ponds I or retention basins (i.e., water hydrologically connected to GROUNDWATER)

A SPILL OR LEAK inside a building or containment unit is generally unlikely to reach GROUNDWATER, particularly if the building or containment unit has a drain and sump system.

PURGE - PURGING PURGE - PURGING SHALL be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (REMP)

The RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM is established for monitoring the radiation and radionuclides in the environs of the plant. The program SHALL provide representative measurements of radioactivity in the highest potential exposure pathways and verification of the ac.curacy of potential exposure pathways and verification of the accuracy of the effluent monitoring program and modeling of the environmental exposure pathways. The current methodology used in the conduct of the specifications of the REMP described in the ODCM are defined in the RPIP 4700 series of Radiation Protection Implementing Procedures.

SHORT TERM'RELEASE SHORT TERM RELEASES refers to releases that are intermittent in radionuclide concentration or flow and defined as less than 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months per calendar year, but not more than 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months in any quarter, and is subject to more restrictive dispersion factors than long term releases.

SITE BOUNDARY That line beyond which the land or property is not owned, leased or .otherwise controlled by the licensee. The SITE BOUNDARIES for liquid and gaseous releases are defined in Figures 3.1 and 3.2.

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SPILL OR LEAK An inadvertent event or perturbation in a system or component performance that releases liquid outside the system or compone.nt.

SOURCE CHECK .

A SOURCE CHECK SHALL be the quantitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity ..

SOURCE CONTAINING LICENSED MATERIAL A liquid, including steam, for which a statistically valid positive result is obtained when the sample is analyzed to the lower limits of detection that are required for radioactive liquid effluents for the isotopes listed in Table 2.1.

UNRESTRICTED AREA An UNRESTRICTED AREA SHALL be any area, access to which is neither limited nor controlled by the licensee.

URANIUM FUEL CYCLE The URANIUM FUEL CYCLE is defined in 40 CFR 190.02(b) as: "The I.

operation of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power for public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these operations, and. the use of recovered non-uranium special nuclear and by-product materials from the cycle."

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VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM SHALL be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the pu"rpose of removing iodines or particulates from the gaseous exhaust stream prior .to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered safety feature atmospheric cleanup systems are.not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING VENTING SHALL be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating _condition, in such a manner that replacement air or gas is

_NOT provided or required during VENTING. Vent, used in system names, does not imply a venting process. The release of air or gases via sampling equipment or instrumentation is not considered a controlled process.*

PRAIRIE ISLAND NUCLEAR.GENERATING PLANT H PROCEDURE NUMBER:,

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 23 of 162 1.0 RADIOLOGICAL EFFLUENT SPECIFICATIONS AND .SURVEILLANCE REQUIREMENTS APPLICABILITY AND SURVEILLANCE REQUIREMENTS 1.1 Specifications 1.1.1 Compliance with the Controls contained within the succeeding text is required during the conditions specified. Upon failure to meet the specifications, the associated ACTION requirements SHALL be met.

1.1.2 Noncompliance with a specification SHALL exist when the requiremen_ts of the Control and associated ACTION requirements are not met within the specified time interval. If the Control is restored prior to expiration of the specified time interval, completion of the ACTION requirements is not required.

1.2 Surveillance Requirements 1.2.1 Surveillance Requirement SHALL be met during the conditions specified for individual specifications unless otherwise stated in an individual Surveillance Requirement.

1.2.2 . Each Surveillance Requirement SHALL be performed within the specified time interval with the following exceptions:

A. A maximum allowable extension not to exceed 25% of the

surveillance interval, but B. The combined time interval, for any three consecutive surveillance intervals, SHALL NOT exceed 3.25 times the specified surveillance interval.

1.2.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 1.2.2, SHALL constitute noncompliance with the functionality requirements for a specification. The time limits of the ACTION requirements are applicable at the time it is identified that a Su.rveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the allowable outage time* limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance Requirements do not have to be performed on nonfunctional equipment.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4

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PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 25 of 162 2.0 LIQUID EFFLUENTS 2.1 SPECIFICATION - CONCENTRATION 2.1.1 In accordance with T.S. 5.5.4.b the concentration of radioactive material released in *liquid effluents to UNRESTRICTED AREAS SHALL be limited to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20:1001-20.2402 other than dissolved or entrained noble gases.

For dissolved or entrained noble gases, the concentration SHALL be limited 'to 2 x 104 µCi/ml total activity.

2.1.2 APPLICABILITY

At all times.

2.1.3 ACTION

A. When the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeds the above limits, immediately restore the. concentration to within the above limits.

B. Report all deviations in the Annual Radioactive Effluent Release

I Report.

C. With liquid effluent release, averaged over 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , exceeding 20 times the limits of concentration in unrestricted areas (10 CFR 20, App. B, Table 2, Column 2) to UNRESTRICTED AREAS, a licensee Event Report (LER) SHALL be submitted within 60 days 2.2 SURVEILLANCE REQUIREMENTS-CONCENTRATION 2.2.1 Radioactive liquid wastes SHALL be sampled and analyzed according to the sampling and analysis program of Table 2.1.

2.2.2 The results of radioactive analysis SHALL be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 2.1.1.

PRAIRIE ISLAND NUCLEAR GENERATING. PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 lei'.

~~.*. MANUAL (ODCM) REV: 31

,~~~~~:~:~r~i~;:.w :'. .,. : '. . Page 26 of 162 2.3 SPECIFICATION_; DOSE 2.3.1 In accordance with T.S. 5.5.4.d the dose or dose commitment to a MEMBE.R OF THE PUBLIC from radioactive materials in liquid effl_uents released to UNRESTRICTED AREAS SHALL be limited to:

A. During any calendar quarter to .:::_3 mrem to the total body and to .:::_10 mrem to any organ, and B. During any calendar year to .:::_6 mrem to the total body and to .:::_20 mrem to any organ.

2.3.2 APPLICABILITY

At all times.

2.3.3 ACTION

A. With the calculated dose from the release of radioactive materials in liquid efflu*ents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days a Special Report that includes the following information:

1. Identifies the cause(s) for exceeding the limit(s).

2.

Defines the corrective actions taken to reduce the release.

3. Defines the corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

2.4 SPECIFICATION REQUIREMENT - DOSE 2.4.1 Cumulative dose contributions for the current calendar quarter and current calendar year SHALL be determined at least every 31 days in accordance

. with the methodology and parameters in Section 4.3of the ODCM.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV:. 31 Page 27 of 162 2.5 SPECIFICATIONS -LIQUID RADWASTE TREATMENT SYSTEMS 2.5.1 In accordance with T.S. 5.5.4.f the LIQUID 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 0.12 mrem to the whole body or 0.4 mrem to any organ in a 31 day period.

2;5.2 APPLICABILITY: At all times.

2.5.3 ACTION

A. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days a Special Report that includes the following information:

1. Explanation of why liquid radioactive waste was being discharged without treatment, identification of any non.:.

functional equipment or subsystems, and the reason for the non-functionality.

2. Action(s) taken to restore the non-functional equipment to

\.,

FUNCTIONAL status, and

3. Summary description of action(s) taken to prevent recurrence.

2.6 SURVEILLANCE REQUIREMENTS - LIQUID RADWASTE TREATMENT SYSTEMS 2.6.1 Doses due to liquid releases SHALL be projected at least every 31 days in accordance with the methodology and parameters in Section 4.3 of the ODCM.

2.6;2. The installed LIQUIDRADWASTE TREATMENT SYSTEM SHALL be considered FUNCTIONAL by meeting the Controls specified in 2.1.1 and 2.3.1.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE

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', ,,, .. . --~.),l;11*:,;,;;i',;:~:t NUMBER:

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 28 of 162

. 2.7 SPECIFICATIONS -RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 2.7.1 In accordance with T.S. 5.5.4.a the radioactive liquid effluent monitoring instrumentation channels shown in Table 2.2 SHALL be FUNCTIONAL with their alarm/trip setpoints set to ensure that the limits_ of Specification 2.1.1 are not exceeded. The alarm/trip setpoints of these channels SHALL be determined in accordance with the methodology in Section 4.1 of the ODCM.

2.7.2 APPLICABILITY

During release via the monitored pathway.

2.7.3 ACTION

A. _With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, immediately suspend the release of radioadive

.effluents monitored by the effected channel, or declare the channel non-functional, or change the setpoint so it is acceptably conservative.

B. With less than the minimum required radioactive liquid effluent monitqring instrumentation channels FUNCTIONAL, take the Action shown in Table 2.2 C, Report all deviations in the Annual Radioactive Effluent Release Report.

2.8 SURVEILLANCE REQUIREMENTS - RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION .

2.8.1 Each radioactive liquid.effluent monitoring instrumentation channel SHALL be demonstrated FUNCTIONAL by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 2.3.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 29 of 162 2.9 SPECIFICATION -LIQUID STORAGE TANKS

. 2.9.1 In accordance with T.S. 5.5.10.c the quantity of radioactive material contained in each of the following tanks SHALL be limited to 10 Curies, exduding tritium and dissolved or entrained gases:

A. Condensate Storage Tanks

8. Outside Temporary Storage Tanks

2.9.2 APPLICABILITY

At all times*.

2.9.3 ACTION

A. With the quantity of radioactive material contained in any of the above listed tanks exceeding the limit in 2.9.1 above, immediately suspend all additions of radioactive materials to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the contents to within the limit.

2.10 SURVEILLANCE REQUIREMENTS -LIQUID STORAGE TANKS 2.10.1 The quan"tity of radioactive material contained in each of the tanks listed in Specification 2.9.1 SHALL be determined to be within the limit by analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.

2.11 SPECIFICATION - LANDLOCKED AREA 2.11.1 In accordance with 10CFR20.2001 and NRC interpretations, soil removed from the landlocked area for free release to the UNRESTRICTED AREA SHALL NOT contain licensed radioactivity, i.e., radionuclides are detected when the soil sample is analyzed to the LLDs listed in Table 7.3 for sediment.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 30 of 162 2.11.2 APPLICABILITY:

A. When the soil in the landlocked area is disturbed (construction occurs in the area or the soil is moved to a new location) and during plant decommissioning.

B. The landlocked area is located near the southwest corner of the Prairie Island reacfor building proper. The landlocked area is fully contained within an area controlled by Prairie Island Nuclear Generating Plant.

2.11.3 ACTION:

A. With the quantity of radioactive material contained in the soil exceeding the limit in 2.11.1 above, describe the landlocked location in the 10CFR50.75.g file, conduct a dose assessment, and remediate, as required by applicable regulation.

2.12 SURVEILLANCE REQUIREMENTS - LANDLOCKED AREA 2.12.1 . The presence of licensed radioactive material described in specification 2.11.1 SHALL be determined by analyzing soil samples of the affected landlocked area when the area is distu.rbed and during plant decommissioning, as required by applicable regulations.

PRAIRIE ISLAND NUCLEAR GENERATiNG PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 31 of 162 3.0 GASEOUS EFFLUENTS 3.1 SPECIFICATION -DOSE RATE 3.1.1 In accordance with T.S.5.5.4.g the dose rate due to radioactive materials released in gaseous effluents from the site to areas at or bey~md the gaseous SITE BOUNDARY (Figure 3.2) SHALL be limited to the following:

A.

For Noble Gases: ~500 mrem/yr to the whole body and ~3000 mrem/yr to the skin, and B. For lodine-131, lodine-133, Tritium, and ,Particulates with half-lives greater than 8 days: ~1500 mrem/yr to any organ.

3.1.2 APPLICABILITY

At all times.

3.1.3 ACTION

A. With the dose rate(s) exceeding the above limits, immediately

. restore the release rate to within the above limits(s). .

B. Report all deviations in the Annual Radioactive Effluent Report.*

C. With airborne radioactive release, averc;iged over 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,*exceeding 20 times the limits of concentration in unrestricted areas (10 CFR 20, App. B, Table 2, Column1) a Licensee Event Report (LER) SHALL be submitted within 60 days.

3.2 SURVEILLANCE REQUIREMENTS - DOSE 'RATE 3.2.1 The dose rate due to noble gases in effluents SHALL be determined to be within the above limits of 3.1.1 in* accordance with the methodology and parameters in Section 5.2 of the. ODCM. *

3.2.2 The dose rate due to lodine-131, *lodine-133, Tritium, and Particulates with half-lives greater than 8 days in gaseous effluents SHALL be determined to be within the above limits of 3.1.1 in accordance with. the methodology

. and parameters in the ODCM by obtaining representative samples and

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE

-~r;f~'.'~(~~:~~:~f:. . NUMBER:

OFFSITE DOSE CALCULATION H4 li,JT,. *4 .

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MANUAL (ODCM) REV: 31 Page 32 of 162 performing analyses in accordance with the* sampling and analysis program specified in Table 3.1.

3.3 SPECIFICATION - NOBLE GAS AIR DOSE 3.3.1 In accordance with T.S,5.5.4.h the air dose due to noble gases released in gaseous effluents to areas at or beyond the gaseous SITE BOUNDARY (Figure 3.2) SHALL be limited to the following:

A. During any calendar quarter: ~10 mrad for gamma radiation and ~20 mrad for beta radiation, and B. During any calendar year: ~20 mrad for gamma radiation and -~40 mrad for beta radiation.

3.3.2 APPLICABILITY

At all times.

3.3.3 ACTION

A. With the calculated dose from the release of radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days a Special Report that includes the following:

1. Identifies the cause(s) for exceeding the limit(s).

2. Defines the corrective .actions taken to reduce the release.

3. Defines the proposed corrective actions to be taken to assure

.that subsequent releases will be in compliance with the above limits.

3.4 SURVEILLANCE REQUIREMENTS - NOBLE GAS AIR DOSE 3.4.1 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases SHALL be determined at least every 31 days in accordance with the methodology and parameters in Section 5.3 of the ODCM.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4

. MANUAL (ODCM) .REV: 31 Page 33 of 162 3.5 SPECIFICATION -IODINE-131, IODINE-133, TRITIUM AND PARTICULATE DOSE 3.5.1 In accordance with T.S.5.5.4.i the dose to any organ of a Ml;:MBER OF THE PUBLIC froni lodine-131, Iodine-133, Tritium, and all radioactive particulates with a half-life greater than 8 days in gaseous effluents released to areas at or beyond the gaseous SITE BOUNDARY (Figure 3.2)

SHALL be limited to the following:

A. During any calendar quarter: ~15 mrem to any organ, and.

B.

During any calendar year: ~30 mrem to any organ.

3.5.2 Carbon 14 A. Carbon-14 contribution to dose SHALL be included in the total dose from lodine-131, lodine"."133, Tritium and Particulates, as specified and defined in section 3.5.1: It is recognized that Carbon-14 is NOT a 10 CFR 50 appendix I dose.

B. Carbon 14 dose contribution SHALL be included in the total dose, in determination of total uranium fuel cycle dose, as required by specification 6.2 C. Carbon-14 total curies generated, for a given time period, SHALL be

determined by calculation, in accordance with the methodologies of "EPRI Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents".

D. Carbon'."14 total curies released, for a given time period, SHALL be equal to the Carbon-14 determined to have been generated. No credit for holdup in the Waste Gas Decay Tanks SHALL be taken.

E. Only the portion of Carbon-14 in the Carbon Dioxide (C02) form is available to enter a viable dose path.way. This is via photosynthesis and incorporation into vegetation, during the daylight growing season. Credit SHALL be taken for the portion of Carbon-14 that is in the C02 form, when performing dose calculations.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 34 of 162

3.5.3 APPLICABILITY

- At all times.

3.5.4 ACTION

A. With the calculated dose from the release of lodine-131, lodine-133, Tritium, and Particulates with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days a Special Report that includes the following:

1. Identifies the cause(s) for exceeding the limit(s).

2. Defines the corrective actions taken to reduce the release.

3. Defines the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

3.6 SURVEILLANCE REQ_UIREMENTS - IODINE-131, IODINE-133, TRITIUM AND PARTICULATE DOSE 3.6.1 Cumulative dose contributions for the current calendar quarter and current calendar year for lodine-131, lodine-133, Tritium, and Particulates with half-lives greater than 8 days SHALL be determined at least every 31. days in accordance with the methodology and parameters in Section 5.0 of the ODCM.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 35 of 162 3.7 SPECIFICATION -GASEOUS RADWASTE TREATMENT SYSTEMS 3.7.1 In accordance with T.S.5.5.4.f the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM SHALL be used to reduce releases of radioactivity when the projected doses due to the gaseous effluents to areas at or beyond the gaseous SITE BOUNDARY (Figure 3.2) would exceed any of the following controls over a 31 day period:

A. 0.4 mrad to air from gamm*a radiation, or B. . 0.8 mrad to air from beta radiation, or C. 0.6*mrem to any organ of a MEMBER OF THE PUBLIC.

3.7.2 In accordance with T.S.5.5.10.b the quantity of radioactivity contained in each gas storage tank SHALL be limited to .:::, 78,800 Curies of noble gases (considered as dose equivalent Xe-133).

3.7.3 The radioactive gas contained in the Waste Gas Treatment System SHALL NOT be deliberately discharged.to the environment during unfavorable wind conditions when the cooling fowers are in operation. For purposes of this specification, unfavorable wind conditions are defined as.

wind from 5° West of North to 45° East of North at 1O miles per hour or less.

3.7.4 APPLICABILITY

At all times.

3.7.5 ACTION

A. *With radioactive gaseous waste being discharged without treatment and in excess of the above limits of 3. 7 .1, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days a Special Report that includes the following information:

1. Identification of any non-functional equipment or subsystems, and the reason for the non-functionality. *

2. Action(s) taken to restore the non-functional equipment to

. FUNCTIONAL status, and

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 36 of 162

3. Summary description of action(s) taken to prevent recurrence.

B. With the quantity of radioactive material in any gas storage tank exceeding the limits of 3.7.2, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months r~duce the tank contents to within the limit.

3.8 SURVEILLANCE REQUIREMENTS - GASEOUS RADWASTE TREATMENT SYSTEMS 3.8.1 . Doses due to gaseous releases at and beyond the SITE BOUNDARY SHALL be projected at least every 31 days in accord~nce with the methodology and parameters in the ODCM .. A projected dose in excess of the limits of 3.7.1 indicates that appropriate components or subsystems of I the GASEOUS RADWASTE TREATMENT SYSTEM must be placed in service to reduce radioactive materials in the gaseous effluents.

3.8.2 The installed Waste Gas Treatment System and the VENTILATION EXHAUST TREATMENT SYSTEM SHALL be considered FUNCTIONAL by meeting the Controls specified in 3.1.1, 3.3.1 AND 3.5.1. I 3.8.3 The quantity of radioactive material contained in each gas storage tank in use SHALL be determined to be within the limit specified in 3.7.2 at least I every 31 days. If the inventory of any tanl< exceeds 10,000 curies, daily sampling when m_aking additions SHALL be performed.

3.9 SPECIFICATIONS - EXPLOSIVE GAS MONITORING INSTRUMENTATION 3.9.1 In accordance with T.S.5.5.10.a the explosive gas monitoring instrumentation channels shown in Table 3:2 SHALL be FUNCTIONAL with their Alarm/Trip Setpoints set to ensure the limits of 3.9.2 are not exceeded.

3.9.2 The concentration of oxygen at the outlet of each operating recombiner SHALL be maintained to ~2% by volume.

3.9.3 APPLICABILITY

As shown in Table 3.2.

3.9.4 . ACTION:

A. With an explosive gas monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the specification of 3.9.2,

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 37 of 162 declare the channel non-functional and take the ACTION shown in Table 3.2.

B. With less than the minimum required explosive gas monitoring instrumentation channels FUNCTIONAL, restore the non-functional instrumentation to FUNCTIONAL status within 30 days. If unsuccessful, in lieu of a License Event Report, prepare and submit a Special Report to the Commission explaining why this non-functionality was not corrected in*a timely manner. The Special Report SHALL be submitted within 60 days.

C. With the concentration of oxygen measured at the outlet of operating recombiner(s) >2% by volume but <4% by volume, restore the-concentration of oxygen to .::5_2% by volume within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

D. With the concentration of oxygen measured at the outlet of operating.

recombiner(s) >4% by volume, immediately suspend all additions of waste gases to the system*~nd reduce the concentration of oxygen to .::5_2% within one hour.

3.10 SURVEILLANCE REQUIREMENTS EXPLOSIVE GAS MONITORING INSTRUMENTATION 3.10.1 Each explosive gas monitoring instrumentation channel SHALL be

demonstrated FUNCTIONAL by performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION at the frequencies shown in Table 3.3.

3.11 SPECIFICATION -RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 3.11.1 In accordance with T.S.5.5.4.a the radioactive. gaseous effluent monitoring instrumentation channels shown in Table 3.2 SHALL be FUNCTIONAL

with their alarm/trip setpoints set to ensure that the limits of Specification 3.1.1 are not exceeded. The alarm/trip setpoints of these channels SHALL be determined in accordance with the methodology in Section 5.0 of the ODCM.

3.11.2 APPLICABILITY: As shown in Table 3.2.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

H4 REV:* 31 Page 38 of 162 3.11.3 ACTION:

A. With a radioactive gaseous effluent monitoring instrumentation .

channel alarm/trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive effluents monitored by the effected channel, or declare the channel non-functional, or change the setpoint so it is a*cceptably conservative.

B. With less than the minimum required radioactive gaseous* effluent monitoring instrumentationchannels FUNCTIONAL, take the Action shown in Table 3.2.

C. Report all deviations in the Annual Radioactive Effluent Release Report.

3.12 SURVEILLANCE REQUIREMENTS - RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 3.12.1 Each radioactive gaseous effluent monitoring instrumentation channel SHALL be demonstrated FUNCTIONAi by performance of the CHANNEL .

CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 3.3.

3.13 SPECIFICATION -ATMOSPHERIC STEAM DUMP MONITORING 3.13.1 The dose to a MEMBER OF THE PUBLIC from lodine-131 released, via one steam dump operation, in gaseous effluents from the site at or beyond the gaseous SITE BOUNDARY (Figure 3.2) SHALL NOT be greater.than twice the limit specified in 3.5.1.

3.13.2 APPLICABILITY: During atmospheric steam dump operations with detectable lodine-131 activity in the Steam Generator bulk water.

3.13.3 ACTION:

A. When the calculated dose from the release of lodine-131 in gaseous effluents via steam *dump operations exceeds the above limit:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 39 of 162

1. The milk from dairy cows grazing in the downwind area SHALL be sampled and analyzed for a period of 5 days following the release. The downwind area shall include the 22 1/2 degree sector of a circle having it's center at the plant and a 2 mile radius.

2. The lodine-131 concentration in the milk SHALL be determined daily utilizing instrumentation with a minimum lodine-131 detection limit of 1.0 pCi/ml.

3.14 SURVEILLANCE REQUIREMENTS -ATMOSPHERIC STEAM DUMP MONITORING 3.14.1 The lodine-131 activity released via atmospheric steam dumps SHALL be sampled and analyzed according to the sample and analysis program of Table 3.1:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H _PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (O_DCM) REV: 31

~

~ .~ ~*.;

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PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PRO.CEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 41 of 162 4.0 LIQUID EFFLUENT CALCULATIONS 4.1 Monitor Alarm Setpoint Determination This procedure determines the monitor alarm setpoint that indicates if the concentration of radionuclides in the liquid effluent released to UNRESTRICTED AREAS exceeds the specification defined in-Section 2.1.1.

Since Fe-55, Sr-89, Sr-90, and alpha concentrations are determined from composite samples, the liquid monitor setpoint determinations should be completed using the most recent available composite sample results.

Monitor high alarm or isolation setpoints.will be established or verified each time a release permit is generated, by the methodology described in section 4.1.1 and 4.1.2. Nuclide mix input to the high alarm or isolation setpoint will be:

A. The Liquid Source Terms (Table 4.1 ).

1. Used in the case that n.o gamma emitters are identified in the batch tank pre-release samples or for continuous releases which are not anticipated to have gamma emitters and are not evaluated pre-release.

B. Based on analysis prior to discharge.

1. Used in the case of that gamma emitters are identified in the batch tank pre-release samples ..

In the *event that no release is made for a given liquid release pathway and therefore no evaluation of the associated liquid process radiation monitor is made, a setpoint calculation will be performed based on Liquid Source Terms (Table 4.1) at least once every 31 days.

If the calculated setpoint is less than the existing monitor setpoint, the setpoint SHALL be reduced to the new value. If the calculated setpoint is greater than the existing setpoint, the setpoint may remain at the lower value or increased to the calculated value.

Setpoint calculations are *performed each time a release permit is generated.

Tritium may constitute a significant portion of the nuclide mix, however tritium will not generate a monitor response (cpm). A bounding calculation may be performed, and a correction factor generated, as a basis for bounding tritium in the set point calculations.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE

.NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 42 of 162 4.1.1 Setpoint Safety Factor Determination:

A. Determine the maximum postulated tritium, based on the maximum postulated RCS tritium.

B. Determine the maximum postulated diluted tritium in the discharged effluent, based on 110% of rated pump discharge flow rate and minimum allowable dilution flow rate:

H301L =H3MAX

FPMP I Fo1L (4.1-1)

Where:

H301L- Diluted tritium concentration (µCi/ml)

H3MAX_. Maximum postulated tritium (µCi/ml)

FPMP - 110% of rated pump flow (gpm)

Fo1L - Minimum dilution flow (gpm)

C. Determine maximum postulated ECL fraction for tritium:

=H3rnL / 1.00E-02 µCi/ml Tritium ECL Fraction (4.1-2)

I D. Determine the Setpoint Safety Factor (SPSF):

SPSF =1 - Tritium ECL Fraction (4.1-3)

The Set Point Safety Factor is used as the tritium adjustment factor. When the maximum tritium contribution to ECL Fraction is accounted for by bounding calculation, it is represented in the set point calculation as a*reduced value of the Setpoint Safety Factor.

The bounding calculation will be maintained in supporting documentation and validated as appropriate.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 43 of 162 4.1.2 Liquid Effluent Monitor Setpoints The following methodology applies when determining the isolation setpoints for:

Waste Effluent Liquid Monitor (R-18)

Steam Generator Slowdown Liquid Monitor - Unit 1 (1 R-19) *

Steam Generator Slowdown Liquid Monitor - Unit 2 (2R-19)

The following calculations assume the radioactive waste liquid discharge flow rate will be maintained constant, at the maximum design flow rate and that dilution flow will be maintained constant at a minimum flow rate.

A. Nuclide "mix" Determination (representative liquid source terms of the liquid effluent)

1. For short term batch releases, the gamma and tritium source

terms will be determined, by analysis, prior to release.

2. In the absence of quantification *Of gamma source terms, the Liquid Source Terms (Table 4.1) may be used, generating a default set point.

3. Fe-55, Sr-89 and Sr-90 are determined by quarterly composite samples. Input to the source terms will be the most recent values.

B. Required Dilution Factor (RDF) Determination RDF =L{AC/ECLj) * (4.1-4)

DSF Where:

RDF Require Dilution Factor (unitless)

A Ci - Activity Concentration of nuclide "i" (µCi/ml)

.ECLi Effluent Concentration Limit of nuclide "i" (µCi/ml)

DSF Dilution Safety Factor; 0.8 C. Specific Activity (SP) Determination

1. Specific Activity equates all nuclide activities determined to be in the mix to the equivalent Cs-137 activity, based on monitor.

response.

2. If the gamma activity concentration is 0, then the Specific Activity need not be determined. The setpoint will be the default setpoint, based on the Liquid Source Terms (Table 4.1 ).

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 44 of 162

3. The allocation factor is the fraction of the radioactivity from the site that may be released via each release point to ensure that the unrestricted area limit is not exceeded due to simultaneous releases from multiple release points. The summation of all the allocation factors for active release points SHALL NOT be greater than unity.

SP= L.(AC; *Cs&];)* SPSF *AF* (FnIL + F;vsT)

(4.1-5)

RDF* FwsT Where:

SP - Specific Activity, adjusted for monitor response (µCi/ml)

ACi - Activity Concentration of nuclide "i" (µCi/ml)

CsEqi - Cs-137 Equivalence, monitor response of nuclide "i" SPSF - Setpoint Safety Factor AF - Allocation Factor FrnL - Dilution Flow (gpm)

FwsT - Waste Flow (gpm)

RDF - Required Dilution Factor.

D. Liquid Set Point (LSP) Determination

1. * .If no gamma emitter activity is quantified, then the Liquid Set Point Calculation is the default Liquid Set Point value, based on Liquid Source Terms (Table 4.1 ).

2. If the Waste Flow exceeds the Maximum Permissible Waste Flow then the Liquid Set Point is zero and no release is permitted.

LSP = eCOA+(COB*In(SP)) (4.1-6)

Where:

LSP Liquid Set Point (cpm)

COA Monitor Calibration Coefficient A COB Monitor Calibration Coefficient B

3. SP Specific Activity adjusted for monitor response

(µCi/ml) The monitor high alarm setpoint above background (ncpm), SHALL be set at or below the LSP value.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31

. Page 45of162

.4.1.3 Maximum Waste Flow Determination As a further verification, and to ensure the release concentration limits are not exceeded, and to identify flow throttling requirements, a maximum waste flow rate is determined, as follows:

If the Required Dilution Factor is < 1, the release rate limit is unlimited ..

If the Required Dilution Factor is.:::. 1, then:

F. - FDIL * (1- L.(DAC; I EGL;))* AF* SPSF (4.1-7)

MAX- RDF Where:

FMAX. Maximum Permissible Waste Flow {gpm)

Fo1L Dilution Flow (gpm)

DA Ci Dilution Activity concentration of nuclide "i;' (µCi/ml)

EC Li Effluent Concentration of nuclide "i" (µCi/ml)

AF Allocation Factor SPSF Setpoint Safety Factor RDF Required Dilution Factor In the absence of activiW in the dilution water, the equation becomes:

F. - FDIL

AF

SPSF (4.1-8)

MAX - RDF 4.1.4 Discharge Canal Monitor (R-21)

The Discharge Canal Monitor*(R-21) provides direct measurement of the diluted plant effluent concentration, monitoring the various streams feeding the discharge canal, with the exception of the Waste Liquid Discharge Header.

The Waste Liquid Discharge Header is extended to the end of the discharge canal to a point just upstream of the river release sluice gates.

This line effectively bypasses R-21.

The Waste Liquid Discharge Header effluents are Steam Generator Slowdown releases and Radioactive Liquid Waste Tank Batch releases from the Auxiliary Building.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31.

Page 46 of 162 The activity of effluents released from the site, other than the Waste Liquid Discharge Header are typically negligible. R-21 is an atypical release monitor, for the purpose of identifying unanticipated releases.

The Discharge Canal Monitor alarm set point is set at a minimal value to detect minimal activity, without generating* spurious alarms.

4.1.5 Rad Effluent Monitor Calibration Liquid effluent monitors are calibrated periodically using a Cs-137 standard. Since the actual isotopic mixes of the liquids released may contain nuclides with different gamma energies and yields than the calibration standard, the response of the monitor varies with respect to the actual energies and abundances of the nuclides in the mix being monitored when compared to Cs-137.

Setpoint determinations or expected monitor readings during or prior to a release are compensated for the difference in gamma energies and yields.

The monitor setpoint calculations and predicted monitor readings are adjusted according to reflect the actual nuclide mix.

The assumption is made that the monitor's response is directly proportional to the gamma energies and yields.

The cumulative errors associated with the monitor calibration methodology are not accounted for in the determination of individual monitor setpoints.

Sufficient conservatism is built into the monitor setpoint determination, such as the required dilution safety factor. Additionally, the use of allocation factors would require that all release paths exceed their respective monitor set points before the limits of ten times the water effluent concentrations of 10 CFR 20, Appendix B, Table 2, Column 2 (ECLs) were challenged.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 47 of 162 4.'2 Compliance With 10 CFR 20 In order to comply with 10 CFR 20, in accordance with T.S.5.5.4.b, the I.

. concentrations of radionuclides in liquid effluents will not exceed 10 times the water effluent concentrations as defined in Appendix B, Table 2, Column 2 of 10 CFR 20.

For CONTINUOUS RELEASES, the alarm trip setpoints discussed in Section 4.1 will assure that these concentrations are not exceeded. For BATCH RELEASES, concentrations of diluted effluents will be compared to effluent concentrations limits pre-release, providing protection in addition to the alarm trip setpoints discussed in Section 4.1 .

. 4.2.1 Continuous Releases Continuous liquid releases can occur from PINGP through Steam Generator Slowdown. The alarm trip setpoints discussed in Section 4.1 will assure that releases from this pathway will not exceed the limits of ten times the water effluent concentrations of 10 CFR 20, Appendix B, Table 2, Column 2.

Other continuous releases* occur at PINGP, through the turbine building sump system. These releases are minor. A continuous composite sample will be .maintained at the discharge from the turbii:ie building sump with samples being taken and analyzed weekly. If these samples indicate significant levels of radionuclides, the methodologies given in section 4.2.2 will be applied to the turbine .sump weekly releases and the limit in Equation 4.1-6, as input to Steam Generator Slowdown and BATCH RELEASES, will be lowered to account for this source term. This will be done by the adjustment of allocation factors to .these releases.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL {ODCM) REV: 31 Page 48 of 162

. 4.2.2 Batch Releases To demonstrate compliance with 10.CFR 20, Append ix B, Table 2, Column 2, the radioactivity content of each BATCH RELEAS E will be determined I

prior to releas.e. The concentration of the various rad ionuclides in the BATCH RELEASE prior to dilution, is divided by the minimum dilution flow to o,btain the concentration at the UNRESTRICTED A REA.

Cone. = C; Fwsr (4.2-1)

I FOIL Where: .

Conci Concentration of radionuclide i at the site boundary (µCi/ml) I Ci Concentration of radionuclide i in the potential batch release Fwsr

Release rate of the batch Fo1L minimum dilution flow (65,900 gpm)

In accordance with T.S.5.5.4.b, the projected concen tration at the UNRESTRICTED AREA is compared to the ten time s the water effluent concentrations of Appendix B, Table 2, Column 2 of 10 CFR 20. Before a I release may occur, Equation 4.2-2 must be met for a II isotopes.

L. =Cone; (4.2-2)

' ECL.l ECLi - Ten times the water effluent concentration of radionuclide i, from 10 CFR 20, Appendix B, Table 2, Colu mn 2 (µCi/ml)

I The summation of all source terms, as input to the to tal contribution to ECL SHALL NOT be greater than 1.0.

The volume of the discharge pipe could contain the v olume of 2 to 3 waste batch tanks. To ensure compliance with 10CFR20 w hen the maximum acceptable discharge flow rate, as calculated in section 4.1.3, is less than the maximum possible release rate from all release s ources, the discharge pipe SHALL be flushed with a volume of at least the volume of the discharge pipe. The flush rate SHALL NOT exceed the maximum discharge flow rate and may be accomplished with w ater from other release paths. If more than one waste batch tank re quiring flushing are to be.released, the discharge pipe may be flushed follow ing the final tank release.

Volume of discharge pipe =15,500 gal.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 49 of 162 4.3 Liquid Effluent Dose - Compliance with 10 CFR 50 Doses resulting from liquid effluents will be calculated at least every 31 days to show compliance with 10 CFR 50. A cumulative summation of total body and organ doses for each calendar quarter and calendar year will be maintained as well as projected doses for the next month.

Since Fe-55, Sr-89, Sr-90, and alpha conc*entrations are determined from composite samples, the monthly liquid effluent dose calculations and comparisons

to quarterly and annual limits should be completed using the most recent available composite sample results. The quarterly and annual dose calculations SHALL be completed using.the actual composite sample results.

The limits of 10 CFR 50 are on a per reactor unit basis. The liquid radwaste system I.

at PINGP is shared by both reactor units making it impossible to separate the releases of the two units. The releases that can be separated by uriit, for steam generator blowdown and turbine building sump releases, contribute a very small portion of the total liquid releases from PINGP. Therefore, for compliance with 10 CFR 50, the releases from both units will be summed and the limits of Appendix I will be doubled.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 50 of 162 4.3.1 Dose Calculations The dose contribution from the release of liquid effluents will be cal.culated for each release permit and will be assessed at least every 31 days. The dose contributio.n will be calculated using the following:

A. Waste Flow Fraction Determination To determine doses from liquid effluents, the waste flow fraction for the period of release must be calculated. This dilution factor must be calculated for each BATCH RELEASE and each CONTINUOUS RELEASE mode. The waste flow fraction is determined by:

WFF= FwsT (4.3-1)

(FotL + FwsT) *MF where:

WFF Waste Flow Fraction (unitless)

FwsT Waste Flow (gpm)

FoiL Dilution Flow (gpm)

MF Mixing Factor*.

The value of MF is the site specific factor for the mixing effect of the PINGP discharge structure. This value is 10 for PINGP while operating in the closed cycle cooling mode. In once through, or helper mode, the value of MF is 1.0.

A waste flow value of the rated pump flow and a dilution value of 65,900 gpm (147 CFS) is used on dose projections when generating a release permit. Actual values are used for final reported dose.

B. Effluent Activity Determination EC.I =WC./* *WFF (4.3-2)

Where:

ECi Effluent Concentration for nuclide i (µCi/ml) wci *Waste Concentration for nuclide I (µCi/ml)

. WFF Waste Flow Fraction (unitless) .

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 51 of 162 C. Effluent Concentration Duration. Determination (4.3-3)

Where:

ECDi *Effluent Concentration *Duration (µci-hr/ml) I T Release Duration (hours) I i

I D. Dose Determination . '

D _ I.(Ecq

Uap** Bip

DFa!i)*109 (4.3-4)

T - OIL *8760 Where:

Dr the dose commitment to the total body or any organ T, from the liquid effluents for the period of release (mrem)

ECDi - Effluent Concentration Duration for nuclide i (µCi-hr/ml)

.. Uap Consumption rate for age group a, pathway p (Kg/year)

Bip Bioaccumulation Factor for nuclide i for pathway p DFari - Dose Factor for age group a, receptor r, nuclide i (mrem/hour I pCi/L)

10 9 Conversion factor (µCi/ml to pCi/L)

PIL Dilution Factor between discharge to collection point 8760 - hours per year *

The factor Air assesses and accounts for the site specific inputs to dose. The factor Air also includes the correction factors of equation 4.3-4. Air must be reassessed and updated if assessment of specific site dose inputs should change," as identified in the Annual Land Use Census. For instance, .a change to Air could be required iri the event that the Mississippi River were to be used as a potable water source.

By employing the Air factor the_ dose equation can be reduced to:

(4.3-5)

Air is the site related ingestion dose commitment factor to the total body or any organ r for each identified principal gamma and beta emitter (mrem/hr per µCi/ml)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 52 of 162 The dose factor Air was calculated for an adult for each isotope using the following equation:

(4.3-6)

Where:

1.14 x 105 - 1OE+06 pCi/uCi

1E+03 mill* 1 year/8,760 hours0.0088 days 0.211 hours 0.00126 weeks 2.8918e-4 months 21 adult fish consumption (Kg/yr)

Bit Bioaccumulation factor for radionuclide i, pathway fish, from Table A-1 of Regulatory Guide 1.109 Rev. 1 5

( ) (pCi/Kg per pCi/I)

Dose conversion factor for radionuclide i for adults for a particular organ r from Table E-11 of Regulatory Guide 1.109 Rev. 1, (5 ) (mrem/pCi)

Mississippi River water is not used as a potable water supply within 300 miles downstream of the PINGP. Wells are used for irrigation downstream of the plant.

Applicable pathway(s) and age group(s) are determined. by the Annual Land Use Census. If changes to the Air is required, calculations are performed using the methodologies of Regulatory Guide 1.109 Rev. 1. The current values are captured in Table 4.2.

A table of Air values, for an adult age group and a fish pathway, are presented in Table 4.2.

4.3.2 Accumulation *of Doses Doses calculated at least every 31 days will be summed for comparison with quarterly and arinual limitS. The monthly results should be added to the doses cumulated from the other months in the quarter of interest and in.

the year of interest for the combined releases of both reactor units and compared to the limits given in Section 2.3.

The quarterly limits represent one half of the annual design objective. If these quarterly or annual limits are exceeded, a special report should be submitted to the USNRC identifying the cause and corrective action to be taken. If twice the quarterly or annual limits are exceeded, a special report SHALL be submitted showing compliance with 40 CFR 190.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 53 of 162 4.3.3 Projection of Doses Anticipated doses resulting from the release of liquid effluents will be projected monthly. If the projected doses would exceed 2 percent of the J limit specified in Section 2.3.1, the liquid radwaste treatment system will be used to process waste (T.S. 5.5.4.f).

Projected dose will be the dose for the preceding 31 days, as calculated by .

Equation 4.3-4.

The total source term utilized for the most recent dose calculation should be used for the projections unless information exists indicating that actual future releases could differ significantly. In this case, the source term would be adjusted to reflect this. information and the justification for the adjustment noted. This adjustment. should account for any radwaste equipment which was operated during the previous month that could be out of service in the coming month:

4.4 References 4.4.1 "Prairie Island Final Environmental Statement," USAEC, May, 1973, p. V-26.

4.4.2 "Prairie Island Nuclear Generating Plant, Appendix I Analysis -

Supplement No.1 - Docket No. 50-282 and 50-306," Table 2.1-1.

4.4.3 "10 CFR 20," Appendix 8, Table II, Column 2.

4.4.4 "Prairie Island Nuclear Generating Plant, Appendix I Analysis "."'"

Supplement No. 1 - docket 50-282 and 50.;306," July 21, 1976, Table 2.1-2.

4.4.5 U.S. Nuclear Regulatory Commission, "Regulatory Guide 1.109 -

Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Compliance with 10 CFR 50, Appendix I,"

Rev. 1, 1977.

H PROCEDURE

PRAIRIE ISLAND NUCLEAR GENERATING PLANT NUMBER:

H4 1; .":3;;,!:,~_:

OFFSITE DOSE CALCULATION MANUAL (ODCM) REV: 31 Page 54 of 162

{ .)~--f't'.~'.)~ ',, '

THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) *REV: 31 Page 55 of 162 5.0 GASEOUS EFFLUENT CALCULATIONS 5.1 Monitor Alarm Setpoint Determination This procedure determi.nes the monitor alarm setpoint that indicates if the dose rate beyond the SITE BOUNDARY due to noble gas radionuclides in the gaseous effluent released .from the site exceeds 500.mrem/year to the whole body or exceeds 3000 mrem/year to the skin.

Monitor high alarms or isolation setpoints will be established each time a release permit is generated, calculated by the methodology described in section 5.1.1.

Nuclide mix input to the high alarm or isolation setpoint will be:

a. The Gaseous Source Terms (Table 5.2).

1. Used in the case that no gamma emitters are identified in the pre-release samples odor continuous releases which are not anticipated to have gamma emitters and are not evaluated pre-release.

b. Pre-release analysis results

1. Used in the case that gamma emitters are identified in the batch release pre-release samples.

If the calculated setpoint is less than the existing monitor setpoint, the setpoint SHALL be reduced to the calculated setpoint. If the calculated setpoint is greater than the existing setpoint, the setpoint may remain at the lower value or increased, not to exceed the calculated set point.

5.1.1

Effluent Monitors

The following method applies when determining the isolation or high alarm setpoint for the monitors listed in Table 5.1.

A Determine the "mix" (noble gas radionuclides and composition) of the gaseous effluent. This is the gaseous source terms that are representative of the gaseous effluent. Gaseous source terms are the total curies of each noble gas. If measured gas source terms are

~elow the lower limits of detection (LLD), Table 5.2 source terms are the mix.

B. Determine the maximum effluent release rate in µCi/sec, for Whole .,

Body Dose Limits.

PRAIRIE ISLAND NUCLEAR GEN.ERATING PLANT H PROCEDURE

_ *.*' NUMBER

.' .. ' ~

OFFSITE DOSE CALCULATION H4

--:-Ill "

. {~r~~A MANUAL (ODCM) REV: 31

'. I;;~:,.'. '"'"* ' *:

Page 56 of 162 MRRVIB = 500

SPSF *AF* L:(NGi)

(5.1-1)

L:(NGi *Ki)* X/Q Where:

MRRwB Max Effluent Release Rate based on Whole Body Dose Rate Limit.(µCi/sec) 500 . Whole Body Dose Rate Limit (mrem/year)

SPSF Setpoint Safety Factor; ~ 1)

AF - Allocation Factor NGi Noble Gas "i" Concentration (µci/cc)

Ki The total whole body dose factor due to gamma emissions from noble gas radionuclide "i" from Table 5.4 (mrem/year/µci/m 3)

X/Q Highest calculated annual average relative concentration of effluents released via the plant vents for any area at or beyond the site boundary, for all sectors. (Table 5.1)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 57 of 162

.C. Determine the maximum effluent release rate in µCi/sec, for Skin Dose Limits.

MRR (uCi I sec)= 3000

SPSF *AF* L,(NG;) (5.1-2)

SKIN L,(NG; *(~ +1.1M;))* x IQ Where:

MRRsKIN

Max Effluent Release Rate based on Skin Dose Rate Limit (µCi/sec) 3000 Skin Dose Rate Limit (mrem'/year)

SPSF Setpoint Safety Faetor AF "Allocation Factor NGi Noble Gas "i" Concentration (µCi/cc)

Li+ 1.1 Mi - The total skin dose factor due to gamma and beta emissions from noble gas radionuclide "i" from

Table 5.4 (mrem/ye.ar/µCi/m 3) I XIQ Highest calculated annual average relative.

concentration of effluents released via the plant vents for any area at or beyond the site boundary, for all sectors, from Table 5.1 D. Define the limiting maximum effluent release rate (MRR), as the lesser value generated (µCi/sec); per equations 5.1-1 and 5.1-2, as input to subsequent calculations.

E. Determine the.maximum effluent concentration'(µCi/cc).

MOC= MAA . (513)

. (FWST + FDIL}* 472 . -

MEG Maxim.um effluentconcentration (µCi/cc)

MRR Maximum Effluent Release Rate (µCi/sec)

FWST Waste Flow (cfm)

FOIL Dilution Flow (cfm) 472 Conversion factor In the absence of dilution flow (typical), the equation becomes:

MEC= MAA. (5.1-4)

FWST*472

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4

. MANUAL (ODCM) REV: 31 Page 58 of 162 F. Determine Xe-133 fraction of the total radioactivity in the gaseous effluent. *

s = Axe-133 (5.1-5) xe A total Where:

Axe-133 - The radioactivity of Xe-133 in the gaseous effluent (µCi/cc)

Atota1 - The total radioactivity of noble gas radionuclides in the gaseous effluent (µCi/cc)

G. Determine Non-Xe-133 fraction of the total radioactivity in the gaseous effluent comprised by all noble gases, excluding Xe-133.

s.1

= Ai (5.1-6)

Atotal Where:

Ai The total radioactivity in the gaseous effluent comprised by all noble gases, excluding Xe-133.

Atota1 -

The total radioactivity of noble gas radionuclides in the gaseous effluent.

H. Determine the setpoint contribution from Xe-133 (cpm).

SP _ e (xcoA+cxcos*1ncMEC*s...-133))) (5.1-7)

Xe-133 -

WHERE:

SPxe-133 - Gas Set Point Contribution from Xe-133 (cpm)

XCOA Monitor Xe-133 Calibration Coefficient A XCOB Monitor Xe-133 Calibration Coefficient B MEC Maximum Effluent Concentration

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 59 of 162 I. Determine the Monitor High Set Point.

HSP = [SPXe-133 + e COA+(COB'ln(MEC'SQ)] * (29 - VAC/29)* (5.1-8)

Where:

HSP High Set Point (cpm)

GOA Monitor Non-Xe-133 Calibration Coefficient A COB Monitor Non-Xe-133 Calibration Coefficient B MEG Maximum Effluent Concentration 29-VAC/29 -

Correction for vacuum of .the monitor The isolation or high alarm setpoint above background (ncpm) for the monitors should be set at *or below the HSP value.

5.1.2 Air Ejector Monitors Radiation monitors 1R-15 and 2R-15 provide an indication of gross noble

gas activity at the main condenser air ejector of Unit 1 and Unit 2, respectively. These monitors are provided to give rapid indication of steam generator tube leakage. They are not effluent monitors since the air ejectors are vented to the auxiliary building vents during normal plant operation and releases are monitored by the auxiliaiy building vent monitoring system.

PRAIRIE ISLAND N~CLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 60 of 162 5.1.3 Monitor Calibration Gaseous effluent monitors are calibrated periodically. Available gas mixes existing in plant systems may be used. Since effluent gas mixes vary in isotopic ratios and the energies of those isotopes span a range of energies, more than one gas mix is used during the calibration.

One calibration is performed with a mix that is predominately Xe-133 with lower level beta and gamma energies. A second calibration is performed with a mix containing longer lived plant gases that more accurately represent the higher beta energy range.

The result of this method of calibration is two separate calibration curves

. for each monitor. The Xe-133 curve is applied to setpoint calculations for

the Xe-133 activity. The second curve is applied to setpoint calculations for balance of noble gases activities.

Setpoint determination and projected monitor reading during release utilize a combination of the two calibration curves, according to the actual nuclide mix.

The cumulative errors associated with the monitor calibration methodology are not accounted for in the determination of the individual monitor setpoints. There is sufficient conservatism built into the selection of the actual monitor setpoint. Additionally, the. monitor fractions used in the setpoint determination equation make it necessary for all the effluent monitors to be in alarm before the. limits of 10CFR Part 20 would be exceeded.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT . H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 61 of 162 5.2 Gaseous Effluent Dose Rate - Compliance with 10 CFR 20 Dose rates resulting from the release of noble gases, and radioiodines and particulates must be calculated to show compliance with 10 CFR 20. The limits of 10 CFR 20 must be met on an instantaneous basis at the hypothetical worst case

. location, and apply on a per site basis.

Releases made via the shield building vents as a result of routine surveillance tests or scheduled short term maintenance/work activities of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or less do not require the sampling and analysis of shield building vent stack samples described in Table 3.1 for the following reasons:

a. Shield building effluent particulates and iodines are filtered through a PAC (Particulate Absolute Charcoal) system and the auxiliary building vent normal ventilation has no filtration.

b. The lower limit of detection limits specified in Table 3.1 can not be obtained on all the specified nuclides with normal sample flow and sample duration of less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

c. Shield building vent releases are monitored via a noble gas monitor.

d. .Auxiliary building normal ventilation flow is high~r than the special ventilation fans that vent via the shield building vent stack.

Therefore, it is conservative to assume that the auxiliary building normal ventilation system would continue to run during the testing/maintenance period. The surveillance test or maintenance/work being performed should be evaluated to ensure the*airborne activity in the affected areas will not increase during the evolution. If this evaluation indicates a possible increase in airborne effluents, or radiation monitors or continuous ai,r monitors in the affected buildings indicate higher tha.n normal background airborne activity before the evolution begins, the shield building vent stack sample SHALL be sampled and analyzed as described in Table 3.1.

Since Sr-a9 and Sr-90 concentrations are determined from composite samples, the pre-release, weekly and monthly airborne dose calculations and comparisons to quarterly and annual limits should be completed using the most recent available composite sample results. The quarterly dose values and critical receptors reported to the USNRC SHALL be calculated using the actual composite results.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 62 of 162 5.2.1 Noble Gases 5.2.2 To comply with the 10 CFR 20 dose limit of 100 mrem TEDE to MEMBERS OF THE PUBLIC, the dose rate at the SITE BOUNDARY resulting from noble gas effluents is limited, per Techical Specification 5.5.4.g.1, to 500 mrem/yr to the total body and 3000 mrem/yr to the skin. The setpoint determinations discussed in the previous section are based on the dose calculation method presented in NUREG-0133. They represent a backward solution to the limiting dose equations in NUREG-0133. Setting monitor alarm trip points in this manner will assure that the limits of Technical Specification 5.5.4 are met for noble gas releases. Therefore, no routine dose calculations for noble gases will be needed to show compliance with this part. Routine calculation~ will be made for doses from noble gas releases to show compliance with 10 CFR 50, Appendix I as discussed in Section 5.3.1.

5.2.3 Radioiodine, Radioactive Particulates, and Other Radionuclides For compliance with Techical Specification 5.5.4.g.2, the dose rate at the SITE BOUNDARY resulting from the release of radioiodine and particulates with half lives greater than 8 days is limited to 1500 mrem/yr to any organ. Calculations showing compliance with this dose rate.limit will be performed for BATCH RELEASES prior to the release. To show compliance, Equations 5.2-1 will be evaluated using 1-1.31, 1-133, tritium, and radioactive particulates with half-lives greater than eight days.

(5.2-1) 1.

Where:

Pi Child critical organ dose parameter for radionuclide i for the inhalation pathway, from Table 5.3 (mrem/yr per µCi/m 3 )

(X/Qv) . -

Annual average relative concentration for LONG TERM release at the critical location, from H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data" (sec/m 3 ) I Oiv Total release rate of radionuclide i from all vents from both units for the batch or week of interest (µCi/sec)

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 63 of 162

. Radioiodines, tritium, and radioactive particulates will be released from up to six individual vents all within 300 feet of each other. For showing .

compliance with Techical Specification 5.5.4.g.2, calculations based on Equation 5.2-1' will be made each release. The source terms (Qiv) will be determined from the results of analysis of vent particulate filters and charcoal canisters and vent flow rate. These source terms include all

. gaseous releases from PINGP.

5.2.4 Critical Receptor Identification Compliance with 10 CFR 20 radiation dose limits for individual MEMBERS OF THE PUBLIC will be demonstrated by identifying critical receptor locations based on 10 CFR 50 App I ALARA design objectives. Since the doses associated with* 10 CFR 50 are more restrictive than the 10 CFR 20 limits,. this method satisfies the 10 CFR 20 requirements.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT . H PROCEDURE NUMBER:*

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 64 of 162 5.3 Gaseous Effluents - Compliance with 10 Ci=R 50 Doses resulting from the release of noble gases, radioiodines and particulates must be calculated to show compliance with Appendix I of 10 CFR 50. The calculations will be performed at least every 31 days for all gaseous effluents.

The limits of 10 CFR 50 are on a per reactor unit basis. The GASEOUS RADWASTE TREATMENT SYSTEM and the auxiliary building at PINGP is shared by both reactor units making it impossible to separate the releases of the two units.

The releases that can be separated by unit contribute a very small portion of the total gaseous releases from PINGP. Therefore, for compliance with 10 CFR 50 the releases from both units will be summed and the limits of Appendix I will be doubled.

Releases made via the shield building vents as a result of routine surveillance tests or scheduled short term maintenance/work activities of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or le.ss do not

. require the sampling and analysis of shield building vent.stack samples described in Table 3.1 for the following reasons:

a. Shield building effluent particulates and iodines are filtered through a PAC (Particulate Absolute Charcoal) system and the auxiliary building vent normal ventilation has no filtration.

b.

The lower limit of detection limits specified in Table 3.1 can not be obtained on all the specified nuclides with normal sample flow and a sample duration of less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .* *

. c. Shield building vent releases are monitored via noble gas*monitor.

d. Auxiliary building normal ventilation flow is higher than the special ventilation fans that vent via* the shield building vent stack.

Therefore, it is conservative fo assume that. the auxiliary building normal ventilation system would continue to run during the testing/maintenance period. The surveillance test or maintenance/work being performed should be evaluated to ensure the airborne activity in the affected areas will not increase during the evolution. If this evaluation indicates a possible increase in airborne effluents, or radiation monitors or continuous air monitors in the affected buildings indicate higher than normal background airborne activity before the evolution begins, the shield building vent stack sampled SHALL be sampled and analyzed as described in Table 3.1.

Since Sr-89 and Sr-90 concentrations are determined from composite samples, the pre-release, weekly and monthly airborne dose calculations and comparisons to quarterly and annual limits should be completed using the most recent available composite sample results. The quarterly dose values and critical receptors reported to the USNRC SHALL be calculated using the actual composite results.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 65of162 5.3.1 Noble Gas A. Dose Equations The air dose at the critical receptor due to noble gases released in

gaseous effluents is determined by Equations 5.3-1 and 5.3-2. The critical receptor will be identified as described in Section 5.3.4. For gamma radiation:

GammaAirDcse = 3.17X1 o-a :L[M; *(XI Ov

O;v) + (x I qv

q;v))]

(5.3-1)

Gamma Air Dose Limits:

< 10 mrad for any calendar quarter

< 20 mrad for any calendar year BetaAirDoSB = 3.17X10-a :L[N; *((XI Ov

O;J + (x I qv

q;v))]

(5.3-2)

Beta Air Dose Limits:

< 20 mrad for any calendar quarter

< 40 mrad for any calendar year Where:

3.17 x 10-8 The inverse of the number of seconds in a year.

The air dose factor due to gamma emission for each identified noble gas radionuclide i, from Table.5.4 (mrad/yr per µCi/m 3)

The air dose factor due to beta emission for each identified noble gas radionuclide i, from Table 5.4 (mrad/yr per µCi/m 3)

X!Ov The annual average relative concentration for areas at or beyond the restricted area boundary for LONG TERM vent releases from H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data" Table 6.0,(sec/m 3 ).

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31

., ..*._,.,..--~*. '-~~L*-~*~2:\1-:~r:. '.: ~_.;.... ;-,._"':'f.

t. rt~jt}.~~.::* : : _ : , , Page 66 of 162 The total release of noble gas radionuclide i in gaseous effluents for LONG-TERM vent releases from both units (µCi)

(x/q)v The relative concentration for areas at or beyond the restricted area boundary for SHORT-TERM vent releases, from H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data" Table 6.0,{sec/m 3).

The total release of noble gas radionuclide in

. gaseous effluents for SHORT-TERM vent releases

. from both units (µCi)

I Noble gases will be released from PINGP from up to six vents.

LONG-TERM (X/Q) and SHORT-TERM (x/q) dispersion factors were calculated using the USN RC computer code "XOQDOQ" assuming*

100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months per year SHORT TERM RELEASES (H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data"). Values of Mand N are taken directly from Reg Guide 1.109 and are given in Table 5.4.

B. Accumulation of Doses Doses calculated monthly will be summed for comparison with quarterly and annual limits. The monthly results will be added to the doses calculated from the other months in the quarter of interest and the year of interest and compared to the limits given in Section 3.3. If these limits are exceeded, a special report will be submitted to the

USN RC. If twice the limits are exceeded, a special report showing

.compliance with 40 CFR 190 will be submitted.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 67 of 162 5.3.2 Radioiodiile, Particulates, and Other Radionuclides A. Dose Equations The worst case dose to an individual from 1-131, 1-133, tritium, and radioactive particulates with half-lives greater than eight days in gaseous effluents released beyond the SITE BOUNDARY is determined by the following expressions:

Dose due to 1-131, 1-: 133, Tritium and Radioactive Particulates with half-lives greater than eight days =

(5.3-3)

< 15 mrem (per quarter)

< 30 mrem (per calendar year)

Where:

The Wand w values are in terms of X/Q (sec/m 3) for the inhalation .

pathways, tritium and Carbon-14. For all other pathways and/or

.w nuclides the Wand value~ are in terms of D/Q (1/m 2). Current dispersion factors are maintained in H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data" 3.17 x 10-8 * - the inverse of the. number qf seconds in a year (sec-1)

Ruak the dose factor for each identified radionuclide i, pathway j, age group a, and organ k, m2 mremfYr per

µCi/sec or mrem/yr per µCi/m 3

Wv Dispersion (deposition) parameter for estimating the dose to an individual at the controlling location for LONG-TERM vent releases release of radionuclide i for LONG-TERM vent releases from both units (µCi)

Wv Dispersion (deposition) parameter for estimating the dose to an individual at the controlling location for SHORT-TERM vent releases release of radionuclide i for SHORT-TERM purge releases from both units (µCi)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

~-jr~~t

'~~;: *1~: :~}::1>-t~i>~~ '.

-.* OFFSITE DOSE CALCULATION MANUAL (ODCM) REV:

H4 31 Page 68 of 162 Equation 5.3-3 will be applied to each combination of age group anci organ. Values of Riiak have been calculated using the methodology given in NUREG-0133 and are maintained in H4.2, "Offsite Dose Calcuiation Manual (ODCM) Supporting Data". Dose factors for isotopes not listed Will be determined in accordance with the methodology in Appendix B.

B. Accumulation of Doses Doses calculated monthly will be summed for comparison with quarterly and annual limits. The monthly results should be added to the doses cumulated from the other months in the quarter of interest and in the year of interest and compared with the limits in Section 3.5.1. If these limits are exceeded, a special report will be submitted to the USN RC. If twice the limit_s are exceeded, a special report showing compliance with 40 CFR 190 will be submitted.

5.3.3 Projection of Doses Doses resulting from the release of gaseous effluents will be projected at least every 31 days. The doses calculated for the present month will be used as the projected doses unless information exists indicating that actual releases could differ significantly in the next month. In this case the source terms wm be adjusted to reflect this information and the justification for the adjustment noted. If the projected release of noble gases for the month exceeds 2 percent of the calendar year limits of section 3.3.1, additional waste gas treatment will be provided. If the projected releas_e of 1-131, 1-133, tritium, and radioactive particulates with half-lives greater than 8 days exceeds 2 percent of the calendar year limit of, equation 5.3-3, operation of the ventilation exhaust treatment equipment is required if not currently in use.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) . REV: 31 Page 69 of 162 5.3.4 Critical Receptor Identification

.To demonstrate compliance with 10 CFR 50 App I ALARA design objectives, two locations of interest are identified.

For noble gases effluents, the location identified for demonstrating compliance will be based on the beta and gamma air doses only. This location will be the offsite location with the highest long term vent X/Q values maintained in H4.2, "Offsite Dose Calculation Manual (ODCM)

Supporting Data". This location will remain the same unless meteorological data is reevaluated or the SITE BOUNDARY changes.

For the 1-131, 1-133, tritium, long-lived particulate and C-14 effluents, the location indentfied for demonstrating compliance will be based on

maxiumum organ dose. It will be known as the Critical Receptor Location.

The selection will follow the annual land use census, performed witt)in 5 miles of the PINGP. Each of the following locations will be evaluated as*

potential critical receptors.

1. Residence in each sector

2. Vegetable garden producing leafy green vegetables

3. All identified milk animal locations Following the annual survey, doses will *be calculated using Equation 5.3-3 for all new identified receptors and those receptors whose characteristics have changed significantly. The calculation will include appropriate information about each new location. The dispersion parameters given in this manual should be employed. The total releases reported for the previous calendar year should be used as the source terms.

In certain cases, the Critical Receptor identified may not produce conservative doses in comparison to a past Critical Rec.eptor. A past Critical Receptor may no longer qualify, based on such criteria as discontinuing the maintenance of a qualifying garden. In this case the option to consider a qualifying garden to still exist may be chosen, when doses may be proven to be conservative, with regards to the newly identified Critical Receptor, based on radioactive effluent releases. This position complies with the U.S. Nuclear Regulatory Commission Branch Technical Position, Revision 1, dated November, *1979.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 70 of 162 5.4 References 5.4.1 "Prairie *Island Nuclear Generating Plant, Appendix I Analysis - Supplement No. 1 -Docket No. 50-282 and 50-306", Table 2.1-4.

5.4.2 "10 CFR 20" 5.4.3 "10 CFR 50" Appendix I 5.4.4 U.S. Nuclear Regulatory Commission, "Regulatory Guide 1.. 109 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Compliance with 10 CFR 50, Appendix I", Rev. 1, 1977.

5.4.5 U.S. Nuclear Regulatory Commission, NUREG 0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants",

dated October, 1978.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 71 of 162 6.0 TOTAL DOSE FROM RADIOACTiVE RELEASES AND URANIUM FUEL SOURCES 6.1 SPECIFICATION - TOTAL DOSE 6.1.1 In accordance with T.S.5.5.4.j and 40 CFR 190 the annual dose or dose commitment to any MEMBER OF THE PUBLIC, beyond the SITE BOUNDARY, due to releases of radioactivity and to radiation from URANIUM FUEL CYCLE sourc~s SHALL be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which SHALL be limited to less than or equal to *75 mrems. *

6.1.2 APPLICABILITY

At all times

6.1.3 ACTION

A. With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 2.3.A, 2.3.B, 3.3.A, 3.3.B, 3.5.A, or 3.5.B, calculations SHALL be made including direct radiation contributions from the reactor units (including outside storage tanks and the ISFSI) to determine whether the above limits have been exceeded and to determine compliance with 40 CFR 190. If such is the case, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, a Special Report that includes the following: *

1. Defines the corrective action(s) to be taken to redu.ce subsequent releases to prevent reoccurrence of exceeding the above limits.

2. Includes the schedule for achieving conformance with the above

limits.

I

3. . This special report as defined in 10 CFR 20.2203(a), SHALL include an analysis that estimates.the radiation exposure (dose) to a MEMBEROF THE PUBLIC from uranium fuel sources, including all effluent pathways and direct radiation, for the calendar year that includes the. release(s) covereo by this report.

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OFFSITE DOSE CALCULATION . H4 MANUAL (ODCM) REV: 31 Page 72 of 162

4. Describe levels of radiation and concentrations of radioactive material involved, and cause of the exposure levels and concentrations.

5. If the estimated dose(s) exceed the above limits, and if the release condition resulting in violation of 40 CFR 190 has not already been corrected, the special report SHALL include a request for a variance in accordance with the provisions of 40 CFR 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

6.2 SURVEILLANCE REQUIREMENTS - TOT AL DOSE 6.2.1 Cumulative dose contributions from liquid and gaseous effluents SHALL be determined in accordance with Surveillance Requirements 2.4, 3.4, and 3.6, and in accordance with the methodology and parameters in the*

ODCM. u 6.2.2 Cumulative direct radiation dose contributions, from the reactor units and the ISFSI SHALL be determined:

A. When the conditions set forth in ACTION (a) of Specification 6.1 are met, AND B. In demonstration of compliance with the limits of 40 CFR 190, for

preparation of the Annual Radioactive Effluent Report.

6.3 METHODOLOGY - TOT AL DOSE 6.3.1 Direct dose component SHALL be assessed A. Pirect dose component, due to gamma radiation, SHALL be assessed per the methodology of ANSl/HPS N13.37-2014, Environmental Dosimetry - Criteria for System Design and Implementation.

B. Neutron shine dose, due t0 the ISFSI, SHALL be assessed.

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. OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31

Page 73of162 6.3.2 Organ dose contribution, due to release of lodine-131, lodine~133, Tritium, Particulates with half-lives greater than 8 days and Carbon-14 in gaseous effluents SHALL be assessed.

6.3.3 An evaluation of the dose contributions from the sources noted in 6.3.1, 6.3.2 and 6.3.3 wiil be performed to determine the maxium individual dose .

location.

6.3.4 Noble gas dose (mrem) to the maximum.individual dose location will be

determined as follows:

A. Gamma Dose (mrem)

=3.1 ?E-08 *SF * ((X/Q

2:(K

Q)) + (x/q *2:(K

q)))

Where:

3.17E-08 years/sec SF occupancy factor (typically 1)

X/Q Long Term Release Dispersion Factor (sec/m3)

Ki Nuclide i Total Body Dose factor due to gamma emissions (mrem/year per µCi/m3)

Qi Total Release of noble gas i, released as a continuous release (µCi) x/q Short Term Release* Dispersion Factor (sec/m3) qi Total Release of noble gas i, released as a batch rel.ease (µCi)

B. Beta Dose (mrem) =

3.1 ?E-08 *SF* ((X/Q

l((L+1.1 M)

Q)) + (x/q *2:((L +1.1 M)

q)))

Where.:

3.1 ?E-08 years/sec SF occupancy factor (typically 1)

XJQ Long Term Release Dispersion Factor (sec/m3)

Li Skin Dose factor due to beta emissions for nuclide i (mrem/year per µCi7m3)

Mi* Air Dose Factor due to gamma emissions for nuclide i (mrad/year per µCi/m3) 1 .1 mrem/mrad

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE ii~tili:!ii7l OFFSIT:~~~~~~~g~~TIO~ NUMBERP:::~4 !!

m

'-'--'---'--'-'---'---'-~~---'-~~~~~~~~~~~~~~~~~~~-'-~~

Qi Total Release of noble gas i, released as a continuous release (µCi) x/q Short Term Release Dispersion Factor (sec/m3) qi Total Release of noble gas i, released as a batch release (µCi) 6.3.5 Doses SHALL be totaled from the direct radiation dose and effluent dose contributors, for the maximum individual dose location.

6.3.6 Liquid effluent dose is not required to be assessed for the determination of compliance with 40 CFR 190 limits.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 75 of 162 7.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM MONITORING PROGRAM SPECIFICATIONS 7.1 In accordance with T.S.5.5.1 the Radiological Environmental Monitoring Program (REMP) SHALL be conducted as specified in Table 7.1.

APPLICABILITY At all times.

ACTION I

a. Whenever the Radiological Environmental Monitoring Program is not being conducted as described in Tab.le 7.1 the Annual Radiological Environmental Monitoring Report SHALL include a description of the reasons for not conducting the program as required and the plans for the prevention of a recurrence.

b. Deviations are permitted from the required sampling schedule if samples are unobtainable due to hazardous conditions, seasonable unavailability, or to malfunctions of automatic sampling equipment.. If the latter occurs, every effort SHALL be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule SHALL be reported in the Annual Radiological Environmental Monitoring Report.

c. With the level of radioactivity as the result of plant effluents in an environmental, sampling medium at a specified location exceeding the reporting levels of Table 7.2 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare an~ submit to the Commission within.30 days from the end of the affected calendar quarter a Special Report that includes the following:

1. Identifies the cause(s) for exceeding the limit(s).

2. Defines the corrective actions that have been taken to reduce radioactive effluents so that the potential annual dose 1 to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 2.3.1, 3.3.1, or 3.5.1.

1 The Methodology and parameters used to estimate the potential annual dose to a member of the public SHALL be indicated in the report.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 76 of 162 When more than one of the nuclides in Table 7.2 are detected in the sampling medium, this report SHALL be submitted if:

concentration (1) concentration (2)

+. -------------------:---------------+ ... -> 1 .0 reporting level (1) reporting level (2)

When nuclides other than those in Table 7.2 are detected and are the result of plant effluents, this report SHALL be submitted if the potential annual dose to a MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of Specifications 2.3.1, 3.3.1, or 3.5.1 This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such a*n event, the condition SHALL be reported and described in the Annual Radiological Environmental Monitoring Report.

d. Although deviations from the sampling schedule are permitted under Paragraph b.

above, whenever milk or leafy vegetation samples can no longer be obtained from the

designated sample locations required by Table 7.1, the Annual Radiological Environmental Monitoring Report SHALL explain why the samples can no longer be obtained and identify the new locations added to and deleted from the monitoring program.

SURVEILLANCE REQUIREMENTS 7.2 The radiological environmental monitoring s*amples SHALL be collected pursuant to Table 7.1 from the specific locations of the radiological environmental monitoring sampling program described in the Radiation Protection Implementing Procedure (RPIP) 4700, and SHALL be analyzed pursuant to the requirements of Table 7.1 and the detection capabilities required by Table 7.3.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 77 of 162 LAND USE CENSUS SPECIFICATIONS 7.3 A Land Use Census SHALL be conducted and SHALL identify:

a. The location of the nearest milk animal, the nearest residence, and the nearest garden of greater than 500 ft2 producing fresh leafy vegetation in each of the 16 meteorological sectors within a distance of 5 miles.

b. Fields or gardens of greater than 500 ft2 producing corn that are irrigated with water taken from the Mississippi River between the plant and a point 5 miles downstream.

APPLICABILITY At all times.

ACTION

a. With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 3.6, in lieu of a Licensee Event Report, identify the new location(s) in the next Annual Radiological Environmental Monitoring Report.

b. With the Land Use Census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Specification 7.1, add the new location(s) to the Radiological Environmental Monitoring Program within

30 days. The sampling location(s) excluding the control station location, having a lower calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program. Identify the new location(s) in the next Annual Radiological Environmental Monitoring Report.

c. If fields or gardens larger than 500 ft 2 producing corn are being irrigated with Mississippi River water, appropriate samples SHALL be collected and analyzed per Table 7.1.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 78 of 162 SURVEILLANCE REQUIREMENTS 7.4 . The Land Use Census SHALL be conducted during the growing season, between the dates of May 1 and September 30 by door to door survey, aerial survey, or by consulting local agricultural authorities or associations. A summary of the results of the land use census SHALL be included in the Annual Radiological Environmental Monitoring Report.

INTERLABORATORY COMPARISON PROGRAM SPECIFICATIONS 7.5 An analysis SHALLbe performed on radioactive materials, supplied by an NRC approved crosscheck program. This program involves the analyses of samples provided by a control laboratory as well as with other laboratories which receive.

portions of the same samples. Media used in this program (air, milk, water, etc.)

SHALL be limited to those found in the radiation environmental monitoring program.

APPLICABILITY At all times.

ACTION

a. When required analyses are not performed, corrective action SHALL be reported in the Annual Radiological Environmental Monitoring Report.

SURVEILLANCE REQUIREMENTS 7.6 The summary results of analyses performed as part of the above required lnterlaboratory Comparison Program SHALL be included in the Annual Radiological Environmental Monitoring Report.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM). REV: 31 Page 79 of 162 8.0 REPORTING REQUIREMENTS 8.1 Annual Radioactive Effluent Report Annual Radioactive Effluent Reports include, T.S.5.6.3 "The Annual Radioactive Effluent Report" and ISFSI T.S 5.3, "The ISFSI Annual Environmental Report".

These reports have different submittal dates and are generated separately.

8.1.1 The Annual Radioactive Effluent Report In accordance with T.S.5.6.3 the Annual Radioactive Effluent Report covering the operation of the units SHALL be submitted in accordance with 10 CFR 50.36A and SHALL include:

a. The Annual Radioactive Effluent Report covering the operation of the plant during the previous calendar year SHALL be submitted by May 15 of each

. calendar year to the Administrator of the appropriate Regional NRC office or designee.

b. The Annual Radioactive Effluent Report SHALL include a summary of the quantities of radioactive liquid and gaseous effluents released from the plant as outlined in Appendix B of Regulatory Guide 1.21, Revision 1, June, 1974, with data summarized on a quarterly basis. In the event that some results are not available for inclusion with the report, the report SHALL be submitted noting and explaining the reasons for the missing* results. The missing data

SHALL be submitted as soon as possible in a supplementary report.

c. The Annual Radioactive Effluent Report SHALL include an assessment of the radiation doses from radioactive effluents released from the plant during the previous calendar year. The report SHALL also include an assessment of the radiation doses from radioactive liquids and gaseous effluents to individuals due to their activities inside the SITE BOUNDARY (Figures 3.1 and 3.2) during the report period. All assumptions used in making these assessments (i.e., specific activity, exposure time and location) SHALL be included in the report.

d. The Annual Radioactive Effluent Report SHALL include the following information for solid waste shipped offsite during the report periqd ..

1. Container volume,

2. Total curie quantity (spedfy whether determined by measurement or estimate), *

3. Principal radionuclides (specify whether determined by measurement or estimate),

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM). REV: 31 Page 80of162*

4. Type of waste (e.g., spent resin, compacted dry waste, evaporated bottoms),

5. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and

6. Solidification agent (e.g., cement, urea formaldehyde).

e. The Annual Radioactive Effluent Report SHALL include ABNORMAL RELEASES from the site of radioactive materials in gaseous and liquid effluents on a quarterly basis.

f. The Annual Radioactive Effluent Report SHALL also include an assessment of radiation doses to the most likely exposed MEMBER OF THE GENERAL PUBLIC from reactor releases and other nearby uranium fuel cycle sou.rces (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show compliance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation

. g. The Annual Radioactive Effluent Report SHALL include a description (including cause, response and prevention of reoccurrence) of occurrences when the sampling frequency, minimum analysis frequency, or lower limit of detection requirements specified in Tables 2.1 and 3.1 were not-met.

h. The Annual Radioactive Effluent Report SHALL include a description of occurrences when less than the minimum required radioactive liquid and/or gaseous effluent monitoring instrumentation channels were functional as requir~d in Tables 2.2 and 3.2.

i. The Annual Radioactive Effluent Report SHALL include a description of the circumstances which caused the failure to complete the minimum sample and/or analysis frequency required by Tables 2.1 and 3.1. The report SHALL include the actions taken to *restore the ~ampler, actions taken to prevent recurrence, and a summary of the occurrences effect on the analysis validity.

j. The Annual Radioactive Effluent Report SHALL include a description of the circumstances which result in LLD's higher than those listed in Tables 2.1 and 3.1.

k. The Annual Radioactive Effluent Report SHALL include an assessment of the radiation doses from radioactive effluents released from the ISFSI during the previous calendar year.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 81 of 162 I. Licensee initiated changes to the ODCM SHALL be submitted to the NRC in the form of a complete legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Report for the period of the report in which the change in the ODCM was made. Each change SHALL be identified by markings in the margin of the affected pages clearly indicating the area of the page that was changed. The date (i.e., month and year) of the*

change SHALL be clearly indicated on the Record of Revisions page.

m. The Annual Radioactive Effluent Report SHALL include description of changes to the Process Control Program.

n.

The Annual Radioactive Effluent Report SHALL include a description of all leaks or spills that are communicated per section 8.4 below, including a dose

. assessment.

8.1.2 . ISFSI Annual EnvironmentalReport In accordance with ISFSI T.S 5.3, the ISFSI Annual Environmental Report SHALL be submitted to the NRC Region Ill, Office, with a copy to the Director, Office of Nuclear Material Safety and Safeguards in accordance with 10 CFR 72.44(d)(3) and SHALL include:

a. The ISFSI Annual Environmental Report SHALL be submitted within 60 days after January 1 of each year.

[m. This report should specify the quantity of each of the principal radionuclides released to the environment in liquid and in gaseous effluents during the previous year of operation and such other information as may be required by the Commission to estimate maximum potential radiation dose commitment to the public resulting from effluent release.

8.2 Annual Radiological Environmental Monitoring Report The Annual Radiological Environmental Monitoring Report, covering the operation of the offsite monitoring program is submitted in accordance with T.S.5.6.2 .. The ISFSI Annual Radiological Environmental Monitoring Report, covering the operation of the ISFSI monitoring program is submitted in accordance with ISFSI T.S.5.2.

These reports are typically submitted as a single report. The two reports are inclusively referred to as the Annual 'Radiological Environmental Monitoring Report.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 82 of 162

a. The Annual Radiological Environmental Monitoring Report covering the operation of the plant during the previous calendar year SHALL be submitted by May 15 of each year to the Administrator of the appropriate Regional NRC office or his designee. *

b. The Annual Radiological Environmental Monitoring .Report SHALL include summarized and tabulated resu_lts in the. format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period. In the event that some results are _not available for inclusion with the report, the report SHALL be submitted noting and explaining the reasons for the missing results. The missing data SHALL be submitted as soon as possible in a supplementary report.

c. The Annual Radiological Environmental Monitoring Report SHALL include summaries, interpretations, and an analysis of trends of the radiological environmental surveillance (3ctivities for the report period, including a comparison with preoperational studies; operational *controls (as appropriate),

and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The report SHALL also include a summary of the results of the land use census. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report SHALL provide an analysis of the problem and a planned course of action to alleviate the problem.

d. The Annual Radiological Environmental Monitoring Report SHALL also include the following: a summary description of the radiological environmental monitoring program; a map of sampling locations within a distance of five miles keyed to a table giving distances and directions from .the reactor; and .

the results of licensee participation in the lnterlaboratory Comparison Program.

e. The Annual Radiological Environmental Monitoring Report SHALL include reasons for all deviations from the REMP sampling program as specified in

Table 7.1 and plans for the prevention of a recurrence, if applicable.

f. The Annual Radiological Environmental Monitoring Report SHALL contain a description of when and why- milk or leafy vegetable samples specified in Table 7.1 cannot be obtained from the designated sample locations, and*

identify the new locations added to and deleted from the monitoring program.

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g. If the level of radioactivity in an environmental sampling medium at a specified location exceeds the reporting levels of Table 7.2 for the sample type specified in Table 7.1 and is NOT the results of plant effluents, the condition SHALL be reported in the Annual_ Radiological Environmental Monitoring Report.

h. A summary of the lnterlaboratory Comparison Program SHALL be included in the Annual Radiological Environmental Monitoring Report. If the required lnterlaboratory Comparison Program analyses are NOT performed, corrective action SHALL be reported in the Annual Radiological Environmental Monitoring Report_

i. The Annual Radiological Environmental Monitoring Report SHALL NOT include the Complete Analysis Data Tables. These contain the results of each sample analysis and SHALL be maintained by the licensee.

j. The Annual Radiological Environmental Monitoring Report SHALL include all on-site and off-site groundwater sample results taken in support of the Industry Initiative. *

k. The Annual Radiological Environmental Monitoring Report SHALL include a description of all leaks or spills that are communicated per section 8.4 below.

8.3 Annual Summary of Meteorological Data An annual .summary of meteorological data SHALL be submitted, at the request of the Commission, for the previous calendar year in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.

8.4 Industry Initiative on Groundwater Protection NOTE: For purposes of this section, groundwater is defined as any subsurface moisture or water, regardless ofwhere it is locked beneath the earth's surface; any water located in wells, regardless of depth, type, or whether it is potable; water in storm drains, unless it has been demonstrated that the storm drains do not leak to ground; and water in sumps that communicate with subsurface water.

a. 30-day Report to the NRC

1. Submit to the NRC within 30 days, a special report for any on-site or off-site GROUNDWATER sample that:.

Exceeds the ODCM criteria for 30-day reporting for off-site samples(see Section 7.0); and

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Has a POTENTIAL TO REACH GROUNDWATER that is or could be used in the future as a source of drinking water. Any GROUNDWATER that is potable should be considered as a potential source of drinking water.

The initial discovery of GROUNDWATER contamination greater than the REMP reporting criterion is the event documented in a written 30-day report. It is not expected that a written 30-day report will be generated each time a subsequent sample(s) suspected to be from the same "plume" identifies concentrations greater than any of the REMP criteria as described in the ODCM. Evaluate the need for additional reports or communications based on unexpected changes in conditions .

2. The 30-day special report should include:

A statement that the report is being submitted in support of the Groundwater Protection Initiative,

A list of the contaminant(s) and verified concentration(s),

Description of the action(s) taken.

. An estimate of the potential or bounding annual dose to a member of the public, and

Corrective action(s), if necessary, that will be taken to reduce.the projected annual dose to a member of the public to less than the limits in 10 CFR 50 Appendix I.

3. Concurrently, provide copies of the 30-day written report to the designated State and Local Officials.

b. Voluntary Communications to State and Local Officials

1. Make informal communications by end of next business day to the designated State and Local officials if a SPILL OR LEAK has the POTENTIAL TO REACH GROUNDWATER an.d exceeds.any of the following criteria:

If a SPILL OR LEAK exceeding 100 gallons from a source containing licensed material,

. * . If the volume of a SPILL OR LEAK cannot be quantified but is likely to exceed 100 gallons from a sou~ce containing licensed material, or

Any SPILL OR LEAK, regardless of volume or activity, is deemed by the Plant Manager or designee to warrant voluntary communication.

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2. Communication with the designated State and Local officials SHALL be made before the end of the next business day for a water sample result

~ . .

Off-site GROUNDWATER or surface water that exceeds any of the REMP reporting criteria for water as described in the ODCM (see Section 7.0), or

On-site surface water that is hydrologically.connected to GROUNDWATER, or GROUNDWATER that is or could be used as a source of drinking water, that exceeds any of the REMP reporting criteria for water as described in the ODCM.

Document the basis for concluding that on-site GROUNDWATER is not or would not be considered a source of drinking water. Examples of a defensible basis are documents from the regulatory agency with jurisdiction over GROUNDWATER use.

3. When communicating with State and Local officials, be clear and precise when quantifying the actual release information as it applies to the. .

appropriate regulatory criteria (i.e. put it in perspeC?tive). The following information should be provided as part of the information communication:

A statement that the communication is being made as part of the NEI Groundwater Protective Initiative,

The date and time of the SPILL OR, LEAK, or sample result(s),

Whether or not the spill has been contained or the leak has been stopped,

If known, the location of the SPILL OR LEAK or water sample(s),

The source of the SPILL OR LEAK, if known,

A list of the contaminant(s) and the verified concentration(s),

Description of the action(s) already taken and a general .description of future actions,

An estimate of the potential or bounding annual dose to a member of the public if available at this time, and

An estimated time/date to provide additional information or follow:-up.

4. Following communication with State/Local offidals, complete a 4-hour 10CRF50.72 NRG notification.

5. Contact NEI. by email address GW_Notice@nei.org with the information provided to t~e State Local Officials.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 86 of 162 8.5 Record Retention 8.5.1 Records will be retained for the "Life of Insurance Policy, plus ten (10) years".

8.5.2 . Records to be retained include, but not limited to, the following:

A. Periodic checks, inspections, tests and calibrations of components and systems as related to the specifications and treatment systems defined in the ODCM.

B. Records of wind speed and direction.

C. Liquid and airborne radioactive releases to the environment.

D. Off-site environmental monitoring surveys.

E. Records of reviews performed for changes made to the Offsite Dose Calculation Manual.

8.6 Official correspondences with the NRG and other government agencies SHALL be processed IAW:

A. CP 0061 B.

CP 0067 C. FP-R-LIC-13

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM)

REV: 31 Page 87 of 162 8.7 Reporting Errata in Effluent Release Reports 8~7.1 Small errors should be corrected within one year of discovery and the correction may be submitted with the next normally scheduled submittal of the ARERR(Annual Radiological Effluent Release Report). Small errors criteria are:

Inaccurate reporting of dose that equates to < 10% of the applicable 10 CFR 50 Appendix I design objectives of< 10% of the EPA public dose criterion. *

Inaccurate reporting of curies, release rates, volumes, etc., that equate to< 10% of the affected.curie total, release rate, volume, etc.,

after correction.

Omissions that do not impede theNRC's ability to adequately assess the information supplied.

Typographical errors or other errors that do not alter the intent of the report.

8. 7 .2 Large errors should be corrected within 90 days of discovery and the correction should be submitted within 90 days of the discovery. The correction may be submitted with the next ARERR, if the next ARERR is to be submitted within 90 days of the discovery. Large error criteria, are those which do not meet the criteria of a small error.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 89 of 162 BASIS 2.0 LIQUID EFFLUENTS 2.1/2.2 CONCENTRATION This control is provided to ensure that the concentration of radioactive materials released .

in liquid waste effluents from the site to UNRESTRICTED AREAS will be less than ten times the concentration levels.specified in 10 CFR 20, Appendix B, Table 2, Column 2.

This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures exceeding (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, and (2) ten times the limits of 10 CFR

20. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

This control applies to the releases of radioactive materials in liquiq effluents from all units at the site.

Secondary condenser drains were not included in the routine sampling requirements of Table 2.1. Operating experience has shown that the condenser activity during plant transients normally consists of very low levels of tritium. Condensers are normally only released directly to the environment during plant startups and shutdowns and these volumes combined with the low levels of activity are insignificant when compared to the waste tank activities. Condenser releases should be sampled and analyzed during a significant plant event (i.e. steam generator tube rupture, or steam dump to the condenser with a primary to secondary leak >725 gpd).

2.3/2.4 DOSE Provided to implement the requirements of Sections II.A, Ill.A and IV.A of Appendix I, 10 CFR 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Considering that the nearest drinking water supply using the river for drinking water is more than 300 miles downstream, there is reasonable assurance that the operation of the facility will not result in radioactive concentrations in the drinking water that.

are in excess of the 40 CFR 141 limits.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 90 of 162 2.5/2.6 LIQUID RADWASTE TREATMENT SYSTEMS Provides assurance that the liquid radwaste treatment system will be available for use 1 whenever liquid effluents require treatment prior to release to the environment. The requirements that appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents be kept "as low as reasonably achievable". This control implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50 and the design objective given in Section 11.D of Appendix I to 10 CFR 50. The limhs governing the use of appropriate portions of the liquid radwaste system were specified as a suitable fraction of the guide set forth in Section II.A of Appendix I, 10 CFR 50, for liquid effluents.

The liquid radwaste treatment system is shared by both units. It is not practical to determine the contribution from*each unit to liquid radwaste releases. For this reason, liquid radwaste releases will be allocated equally to each unit.

. 2.7/2.8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or poteritial releases of liquid effluents. The Alarm/Trip Setpoint for these instruments SHALL be calculated and adjusted in accordance with the methodologies and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding ten times the water effluent concentration limits of 10 CFR 20. The FUNCTIONALITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50.

Ra.diation monitor set points are calculated to provide alarm and trip functions to ensure concentration of radioactive materials in liquid waste effluents released from the site to UNRESTRICTED AREAS, does not exceed the noted specific limits. The methodology prescribed in the ODCM for these calculations is acceptable for use in demonstrating compliance with 10 CFR 20.1301 (a)(1 ), 10 CFR 50.36a, 10 CFR 50, Appendix A (GDC 60

& 64) and Appendix I, and 40 CFR 190.

Revision to the ODCM requires Operations Committee review and approval to ensure the revision continues to demonstrate compJiance.

Specific monitor set point changes, when performed in accordance the methodology as reviewed and approved by the Operations Committe~ need not be reviewed by the Operations Committee. Specific monitor set point changes will be reviewed and approved by the Department Manager administering the ODCM program and the Radiation Monitor

Engineer. The calculation sheet supporting the set point change is submitted to engineering for documentation.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 91 of 162 2.9/2.10 LIQUID STORAGE TANKS Restricting the quantities of radioactive material contained in the specified tanks provides

. assurance that in the event of an uncontrolled release of the contents of the tank, the resulting concentrations would be less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, in an UNRESTRICTED AREA.

3.0 GASEOUS EFFLUENTS 3.1/3.2 DOSE RATE This specification provides reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure to a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA either at or beyond the SITE BOUNDARY in excess of the design objectives of appendix I to 10 CFR 50. This specification is provided to ensure that gaseous effluents from all units on the site will be appropriately controlled. It provides operational *flexibility for releases gaseous effluents to satisfy the Section ILA and 11.C design objectives of appendix I to 10 CFR 50. *For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy.of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to less than or equal to 1500 mrem/year at or beyond the SITE BOUNDARY.

This control applies to the release of radioactive materials in gaseous effluent from all units

,at the site.

3.3/3.4 . DOSE FROM NOBLE GAS This control is provided to implement the requirements of Sections 11.B, Ill.A and IV.A of Appendix I, 10 CFR 50. The Limiting Conditions for Operation implement the guides set forth in Section 11.B of Appendix I. The ACTION statement provides*the required operating

flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I to assure that the release of radioactive material in gaseous effluents will be kept "as low as reasonably achievable".

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OFFSITE DOSE CALCULATION H4

MANUAL (ODCM) REV: 31 Page 92 of 162 3.5/3.6 DOSE FROM IODINE 131, IODINE 133, TRITIUM & PARTICULATES Implements the requirements of Section 11.C, Ill.A and IV.A of Appendix I, 10 CFR 50. The Limiting Conditions for Operation are the guides .set forth in Section 11.C of Appendix I.

The ACTIONS statement provides the required operating flexibility and at the same time implements t~e guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive m:aterials in gaseous effluents will be kept "as low as reasonable achievabl.e".

The release rate specifications for 1-131, 1-133, tritium and radioactive particulates with half-lives greater than eight days are dependent on the existing radionuclide pathways to

MEMBERS OF THE PUBLIC in the UNRESTRICTED AREA, using child dose conversion factors. The pathways which are examined in the development of these calculations are:

1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto

green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

3.7/3.8 GASEOUS RADWASTE TREATMENT SYSTEMS This control provides assurance that the Waste Gas Treatm.ent System and the VENTILATION EXHAUST TREATMENT SYSTEMS will be available for use whenever gaseous wastes are released to the environment. The requirement that the appropriate portions of the Waste Gas Treatment System be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as reasonably achievable". This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50, and the design objective given in Section 11.D of Appendix I to 10 CFR 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections 11.B and 11.C of Appendix I, 10 CFR 50, for gaseous effluents.

The Waste Gas Treatment System, containment purge release vent, and spent fuel pool

are shared by both units. Experience has also shown that contributiqns from both units are released from each auxiliary building vent. For these reasons, it is not practical to allocate releases to a specific unit. All releases will be allocated equally in determining conformance to the design objectives of 10 CFR 50, Appendix I.

Restricting the*quantities of radioactivity which can be stored in one decay tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest EXCLUSION AREA BOUNDARY will not exceed 0.5 rem.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 93 of 162 The cooling towers at Prairie Island are located to the south of the plant and are within 500 to 2000 feet from the point of release. At low wind velocities (below 10 mph) the gaseous activity released from the gaseous radwaste system could be at or near ground level near the cooling towers and remain long enough to be drawn into the circulating water in the tower. This control minimizes the possibility of releases of gaseous effluents from entering the *river from cooling tower scrubbing.

3.9/3.10 EXPLOSIVE GAS MONITORING INSTRUMENTATION To ensure the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen and oxygen.

Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. Maintaining the concentrations below the flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR 50. .

The system is operated with flow through the recombiners and with excess hydrogen in the system: By verifying that oxygen is less than or equal to 2% at the recombiner outlet, there will be no explosive mixtures in the system. Waste gas system oxygen is monitored by the two recombiner oxygen analyzers and the 121 gas analyzer. The 121 gas analyzer only monitors the low level loop of the waste gas system. If the required gas analyzers are noffunctional, the oxygen to the recombiner will- be isolated to prevent oxygen from entering the system from this source. Tanks that may undergo maintenance are normally purged with nitrogen before placing them in service to eliminate this as a source of oxygen.

3.11/3.12 RADIOACTIVE GASEOUS EFFLUENT.MONITORING INSTRUMENTATION The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.* The Alarm/Trip Setpoint for these instruments SHALL be calculated and adjusted in accordance with the methodologies and parameters in the ODCM to ensure that the alarm/trip will occur.prior to exceeding the limits of 10 CFR 20. The FUNCTIONALITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50.

Radiation monitor set points are calculated to provide alarm and trip functions to ensure concentration of radioactive materials in airborne effluents releas~d from the site do not exceed the noted specific limit$. The methodology prescribed in the ODCM for these

. calculations is acceptable for use in demonstrating compliance with 10 CFR 20.1301 (a)(1 ),

10 CFR 50.36A, 10 CFR 50, Appendix A (GDC 60 & 64) and Appendix I, and 40 CFR 190. , .

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 94 of 162 Revision to the ODCM requires Operations Committee review and approval to ensure the revision continues to demonstrate compliance.

Specific monitor set point changes, when performed in acco~dance the methodology as reviewed and approved by the Operations Committee need not be reviewed by the Operations Committee. Specific monitor set point changes will be reviewed and approved by the Department Manager administering the ODCM program* and the Radiation Monitor Engineer. The calculation sheet supporting the set point change is submitted to engineering for documentation.

6.0 *TOTAL DOSE This control is provided to meet the dose limitations of 40 CFR 190 that have been incorporated info 10 CFR 20 by FR 18525. The control requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or to any organ, except the thyroid, which SHALL be limited to less than or equal to 75 mrems.

For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units (including outside storage tanks, etc.) are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within 40 CFR 190 limits.

With the advent of the ISFSI, direct dose SHALL be assessed in demonstrating compliance with 40 CFR 190, for Special Reporting and for inclusion i*n the Annual Radioactive Effluent Report.

If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in

accordance with the provisions of 40 CFR 190.11 & 10 CFR 20.2203(a)(4), is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR 20, as addressed in Specification 2.1 and 3.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle. Demonstration of compliance with the limits of 40 CFR 190 will be considered to demonstrate compliance with the 0.1 rem limit of 10 CFR 20.1301.

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 95 of 162 7 .o* RADIOLQGICAL ENVIRONMENTAL MONITORING 7.1/7.2 MONITORING PROGRAM Provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the plant operation. This program therepy supplements the radiological effluent monitoring by verifying that the measurable concentrations of radioactive materials and levels are not higher than expected in the bases of the effluent

. measurements and modeling of the environmental exposure pathways.

y The detection capabilities required by Table 7 .1 are state-of-the art for routine environmental measurements in industrial la_bor_atories and the LLDs for drinking water meet the requirement of 40 CFR 141.

7.3/7.4 LAND USE CENSUS This control is provided to ensure that changes in the use of off site areas are identified and that modifications to the monitoring program are made if required by the results of the census. The best survey information from door-to-door, aerial or consulting with local agricultural authorities SHALL be used. This census satisfies the requirements* of Section IV.B.3 of Appendix I to 10 CFR 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored sinc*e a garden of this size is the minimum

required to produce the quantity (26 kg/year) of leafy vegetables assumed .in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were used: 1) that 20% of the garde.n was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of

_2 kg/square meter.

7.5/7.6 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in*an interlaboratory comparison program is provided to ensu.re that independent checks on the precision and accuracy of the measurements of radioactive material in. environmental sample matrices are performed as part. of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

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'.* ' ' *~:f;_:-,:~~;J0~*~:{3--~~** NUMBER:

' '1 \ ;;'! I '.'i *:~

H4

-~"t~:.ltl:"*~::~: . OFFSITE DOSE CALCULATION MANUAL (ODCM) REV: 31

-:*-.....::,;.....,..:*-t**'.,,_ .. _;,.,,.-

~,~ ri.t~'.~ F*. ~. .- Page 96 of 162 THIS PAGE IS LEFI INTENTIONALLY BLANK.

H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 97 of 162 Table 2.1 Radioactive Liquid Waste Sampling and Analysis Program Liguid Release Tll~e Sampling Minimum Type of Activity Lower Lim it of Frequency Analysis Analysis Detection (LLD)

Frequency (µCi/ml)a, d Batch Releases9 : Each Batch Each Batch Principal Gamma 5 x 10-7 Waste Tanks (Prior to (Prior to Emitters0 Release) Release) 1-131 1 x 10-6 One Batch One Batch Dissolved and 1 x 10-5 Each Month Each Month Entrained Gases Each Batch Morithly H-3 1 x 10-5 Compositeb Gross alpha 1 x 10-7 Each Batch Quarterly Sr-89, Sr-90 5 x 10-8 Compositeb Fe-55 1 x 10-6 Continuous Release 0 : Continuousi,h,k' Weekly Principal Gamma 5 x 10-7 Turbine Building Compositet Emitters 0 Sumps . 1-131 1 x 10-6 Weekly Grab Each Sample

Dissolved and 1 x 10-5 Sample Entrained Gases Continuousi.k Monthly H-3 1 x 10-5 Compositet Gross Alpha 1 x 10-7 Continuousi.k Quarterly Sr-89, Sr-90 5 x 10-8 Compositet Fe-55 1 x 10-6 Continuous Release 0 :

Weekly Grab Each Sample Principal Gamma 5 x 10-7 Steam Generator Sample During Compositeb Emitters 0 Slowdown Releasesi 1-131 1 x 10-6 Grab Sa_mple Each Sample .Dissolved and 1 x 10-5 Each Month Entrained Gases During Releases Weekly Grab Monthly H-3 1 x 10-5 Sample Dur!ng Conipositeb Releases' Gross Alpha 1 x 10-7 Weekly Grab Quarterly Sr-89, Sr-90 5 x 10-8 Sample Dur!ng Compositeb Releases' Fe-55 1x10..£

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 98 of 162 Table 2.1 Radioactive Liquid Waste Sampling and Analysis Program Table Notations

a. The LLD is defined, for purposes of these controls, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will d detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

4.66Sb

LLD =

E

V

2.22X106

Y *exp (-A.Ai-)

where:

LLD = the "a priori" lower limit of detection (microCurie per unit mass or volume),

Sb = the standard deviation of the background counting rate or of the .

counting rate of a blank sample as appropriate (counts per minute),

E = the counting efficiency (counts per disintegration),

v = the sample size (units of mass or volume),

2.22 x 106 = the number of disintegrations per minute per microCurie, y = the fractional radiochemical yield, when applicable,

')., = the radioactive decay constant for the particular radionuclide (sec-1),

and

= the elapsed time between the midpoint of sample collection and the time of counting (sec).

Typical values of E, V, Y, and A"L should be used in the calculation ..

It should be recognized that the LLD is defined as an _g_ priori (before the fact) limit representing the capability of a measurement system and not as an _g_ posteriori (after the fact) limit for a particular measurement. .

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE

NUMBER:

OFFSITE oos*e CALCULATION H4 1

MANUAL (ODCM) REV: 31 Page 99 of 162 Table 2.1 Radioactive Liquid Waste Sampling and Analysis Program Table Notations [Cont'd]

b. A composite ~ample is one in which th~ quantity of liquid sampled is proportional to the quantity of liq:uid waste discharge and in which* the method of sampling employed results in a specimen which is representative of the liquids released. *

c. The principal gamma emitters for which the LLD control applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo..:.99, Cs-134, Cs-137 and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 x 10-6 . This list does not mean that only these nuclides are to be detected and reported. Other gamma peaks which are measurable and identifiable, together with the above nuclides, SHALL also be identified an_d reported. * *

d. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result.in LLDs higher than required, *the reasons SHALL be documented in the Annual Radioactive Effluent Report.

e.* *A CONTINUOUS RELEASE is the discharge of liquid wastes of a non-discrete volume; e.g., from a volume of system that has an input flow during the continuous release.

f.

To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples SHALL be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite SHALL be thoroughly mixed in order for the composite sample to be representative of the effluent release.

g. A BATCH RELEASE is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch SHALL be isolated, and then thoroughly mixed to assure representative sampling.

h. Daily grab samples from the turbine building sumps SHALL be collected and analyzed for principal gamma emitters, including 1-131, whenever primary to secondary leakage exceeds 1SO gpd in any steam generator. This sampling is provided in lieu of continuous monitoring with automatic isolation.

i. Grab samples SHALL be collected at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when steam generator blowdown releases are being made and the specific activity of the. secondary coolant is ~0.01 µCi/gram DOSE EQUIVALENT 1-131 or primary to secondary leakage exceeds 150 gpd.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 100of162 Table 2.1 Radioactive Liquid Waste Sampling and Analysis Program Table Notations [Cont'd]

j. A continuous sample is one in which the sampling media is in place at all times during the release period, with the exception of periods necessary to change sampling media and scheduled short term equipment maintenance. If the sample media is not in place during the entire release period, an explanation of the occurrence, actions taken to restore the sampler and to prevent recurrence, and a summary description to explain the occurrence's effect on the analysis validity SHALL be included in the Annual Radioactive Effluent Report.

k.

Continuous samples of the Turbine Building Sumps are collected via on-line composite samplers. These samplers function on timers and collect a predetermined volume of effluent whenever the TBS pumps are in operation. Samples from these compositors are collected daily and saved for the preparation of a weekly composite prepared utilizing volumes proportional to the sample volumes collected daily by the compositor. If the use of a submersible pump is necessary to maintain sump level, that pump should be positioned above the normal TBS pump controlling level and include a timer to allow the calculation of the additional 'release volume.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 101 of 162 Table 2.2 Radioactive Liquid Effluent Monitoring Instrumentation With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels FUNCTIONAL, take the actions directed in Table 2.2. Restore the non-functional instrumentation to FUNCTIONAL status within 30 days. If instrumentation is not restored within 30 days, explain in the next Annual Radioactive Effluent Release Report, why this non-functionality was not corrected in a timely manner.

MINIMUM CHANNELS INSTRUMENT FUNCTIONAL APPLICABILITY . ACTION . I

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Liquid Radwaste Effluent Urie 1 During releases 1 I

b: Steam Generator Slowdown 1/Unit During releases 2 Effluent Line

2. *Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line 1 During releases 4 requiring throttling of flow

b. Steam Generator Slowdown Flow 1/Gen During. releases 4

3. *Continuous Composite Samplers

a. Each Turbine Building Sunip 1/Unit During releases 3 Effluent Line

4. Discharge Canal Monitor 1 At all times 6

5. Tank Level Monitor

a. Condensate Storage Tanks 1/Unit . When containing 5 radioactive material

b. Temporary Outdoor Tanks Holding 1/Tank When tanks are 5 Radioactive Liquid in use

6. Discharge Canal Flow System (Daily NA At all times determination and following changes in flow)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 102of162 Table 2.2 Radioactive Liquid Effluent Monitoring Instrumentation Table Notations ACTION 1 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue, provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Specification 2.2.1, and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge .

line valving. ..

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 2 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue provided grab samples are analyze.d for radioactivity at a lower limit of detection of not more than that specified in Table 2.1 for Principal Gamma Emitters.

1. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the secondary coolant is

~0.01 µCi/gram DOSE EQUIVALENT-1-131, or

2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is

<0.01 µCi/gram DOSE EQUIVALENT 1-131.

ACTION 3 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may I

continue provided that at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and saved for weekly composition and analysis in accordance with Table 2.1.

ACTION 4 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requireinent, effluent releases via this pathway may continue provided that the flow rate is estimated at least once per four (4) hours during actual releases. Pump curves may be used to estimate flow.

ACTION 5 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue provided that tank liquid level is estimated during all liquid additions.

ACTION 6 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gamma emitters.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 103of162 Table 2.3 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL SOURCE FUNCTIONAL

cHECK. CHECK TEST CALIBRATION Instrument Frequency (4) Frequency . Frequency Frequency Liquid Radwaste Effluent Daily during Prior to each Quarter1y< 1> At least once every Line Gross Radioactivity releases release 18 months<3>

Monitor Liquid Radwaste Effluent Daily during ---- ---- At least once every Line Flow Instrument releases 18 months Steam Generator Slowdown Daily during Monthly Quarter1y< 1>* At least once every Gross Radioactivity Monitors releases 18 months<3>

Steam Generator Slowdown Daily during ---- ---- At least once every Flow releases 18 months Turbine SuUding Sump Daily during ---- ----

Continuous Composite releases

. Samplers (Includes sample volume check)

Discharge Canal Monitor Daily during Monthly Quarter1y<2 > At least once every releases 18 months<3>

Discharge Canal Daily during ---- ---- At least once every Flow Instruments releases 18 months Condensate Storage Tank Daily ---- Quarterly At least once every Level Monitors 18 months Level Monitors for Daily when in use ---- Quarterly when in At least once every Temporary Outdoor Tanks use 18 months when in Holding Radioactive Liquid use

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION. H4 MANUAL (ODCM) REV: 31 Page 104of162 Table 2.3 Radioactive Liquid Effluent Monitoring Instrumentation Surve.illance Requirements Table Notations

1. The CHANNEL FUNCTIONAL TEST SHALL also demonstrate that automatic isolation of this pathway and control room annunciation occurs if any of the following Gonditions exists: .

a. Instrument indicates measured levels above the alarm/trip setpoint.

b. Circuit failure (if provided).

c. Instrument indicates a downscale failure (if provided).

d. Instrument controls not set in operate mode (if provided).

2. The CHANNEL FUNCTIONAL TEST SHALL also demonstrate that alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the alarm/trip setpoint.

b. Circuit failure (if provided).

c. Instrument indicates a downscale failure (if provided).

d. Instrument controls not set in operate mode (if provided).

3. The initial CHANNEL CALIBRATION SHALL be performed using one or more of the reference standards certified by the National Institute of Standards and Tech.nology or using sources traceable to NIST standards. These ~tandards SHALL permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATIONS, sources that have been related to the initial calibration SHALL be used.

4. The CHANNEL CHECK SHALL consist of verifying indication of flow during periods of release. A CHANNEL CHECK SHALL be made at least once daily on any day on which continuous, periodic, or batch releases are made.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 105of162

Table 3.1 Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Release Type Sampling Minimum Type of Activity Lower Limita,t Frequency Analysis* Analysis of Detection Frequency (LLD)(µCi/ml)

CONTINUOUS RELEASE Weekly b, i Weekly Principal Gamma 1 x 10-4 Points: Gas Grab Sample Emitters e g, i, h 1 x 10-12 Weekly c 1-131, 1-133 Plant Vents: Continuous Charcoal Sample g, i, h Unit 1 Aux Bldg. Weekly c Principal Gamma 1 x 10-11 Unit 2 Aux Bldg. Continuous

Particulate Emitters e Radwaste Bldg. Sample*

g, i, h Spent Fuel Pool Monthly H-3 1x10-5 Unit 1 Shield Bldg. Continuous Silica Gel Unit 2 Shield Bldg. *sample g, i, h Each Gross Alpha 1 x 10-11 Continuous Particulate Sample g, i, h 1 x 10-11 Quarterly d Sr-89, Sr'."90 Continuous Particulate Composite g

Noble Gas Noble Gases, 1x10-4 Continuous Monitor Gross beta and gamma Atmospheric Steam Daily i Releases k Grab Sample Each I Principal Gamma During Release Sample Emitters e 5 x 10-7 1-131, 1-133 1 x 10-5 Daily i Grab Sample Monthly 1 During Release Composite H-3 1x10-5 Gross Alpha 1 x 10-7 Daily i Quarterly 1 Sr-89, Sr-90 5 x 10-B Grab Sample Composite :

During Release

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 106of162 Table 3.1 Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Rel*ease Type Sampling Minimum Type of Activity Lower Limita, t Frequency *Analysis Analysis of Detection Frequency (LLD){DµCi/ml)

Containment Purgem Gas Grab Sample Each Principal Gamma 1 x 10-4 Prior to each Purge Sample Emitters e (Prior to Release)

Grab9

h, m Each H-3 1 x .10-6 Prior to Release and Sample Continuous Grabg,h,m Charcoal 1-131, 1-133 1 x 10-12 Prior to Release and Sample Continuous Grab 9

h, m Particulate Principal Gamma 1 x 10-11 Prior to Release and Sample Emitters e Continuous Grab9

h, m Each Gross Alpha 1 x 10-11 Prior to Release and Particulate Continuous Sample Grab9 *h* m Quarterly d Sr-89, Sr-90 1 x 10-11 Prior to Release and . Particulate Continuous Composite Waste Gas Gas Grab Sample Each Principal Gamma 1 x 10-4 Storage Tan ks Prior to each Sample Emitters e Release (Prior to Release)

Grab Sample .Each H-3 1 x 10-6 Prior to each Sample Release (Prior to Release)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE.DOSE CALCULATION H4

.. MANUAL (ODCM) REV: 31 Page 107of162 Table 3.1 ~ Radioactive Gaseous Waste Sampling and Analysis Program .

Table Notations

a. The LLD is defined, for purposes of these. controls, as the smallest concentration of radioactive material in a sample that will yield a .net count, above system backgrou.nd, that will be detected with 95% probability with only 5% probability of falsely concluding

.that a bla~k observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

LLD =

E

V

2.22X10 6

Y *exp (- A-61) where:

LLD = the "a priori" lower limit of. detection (microCurie per unit mass or volume.

Sb = of the standard deviation the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute).

E = the counting efficiency (counts per disintegration),

v = the sample size (units of mass or volume),

2.22 x 106 = the number of disintegrations per minute per microCurie, y = the fractional radiochemical yield, when applicable,

')., = the radioactive decay constant for the particular radionuclide (sec-1), and the elapsed time between the midpoint of sample collection and the time of c9unting (sec).

Typical values of E, V, Y, and .1.,; should be used in the calculation:

It should be recognized that the LLD is defined as an g_ priori (before the fact) limit representing the capabi.lity of a measurement system and not as an g_ posteriori (after the fact) limit for a particular mea~urement.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 108of162 Table 3.1 - Radioactive Gaseous Waste Sampling and Analysis Program Table Notations [Cont'd]

b. Grab samples taken at the ventilation exhausts are generally below minimum detectable levels for most nuclides with existing analytical equipment. If this is the case, Gaseous So1:1rce Terms (Table 5.2) noble gas isotopic ratios may be assumed.

c. With >1 µCi/gm DOSE EQUIVALENT 1-131 in either Unit 1 or Unit 2 reactor coolant system, the iodine and particulate collection devices for all° release points SHAL.L be removed and analyzed daily until it is shown that a pattern exists which can be used to predict the release rate. Sampling may then revert to weekly. When samples collected for one day are analyzed, the corresponding LLD's may be increased by a factor of 10. Samples SHALL be analyzed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after removal.

d. To be representative of the average quantities and concentrations of radioactive materials in particulate form in gaseous effluents, samples should be collected in proportion to the rate of flow of the effluent streams.

e. The principal gamma emitters for which the LLD control applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for noble gas analysis and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 in iodine and particulate analysis. This list does not mean tha.t only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, SHALL also be detected and reported.

f. Nuclides which are below the LLD for analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than reported, the reasons. SHALL be documented in the Annual Radioactive Effluent Report.

g. For continuous samples, the ratio of the sample flow rate to the samples stream flow rate SHALL be known for the time period sampled (Conservative assumptions may be used). Design flow rates may be. used for building exhaust vent flow rates.

h. A continuous sample is one in which the sampling media is in plac;;e at all times during the release period, with the exception of periods necessary to change sampling media and scheduled short term equipment maintenance of two hours or less. If the sample media is not in place during the entire release period (except as described above), an explanation of the occurrence, actions taken to restore the sampler and to prevent reoccurrence, and a summary description to explain the occurrence's effect on the analysis validity SHALL be included in the Annual Radioactive Effluent Report.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 109of162 Table 3.1 Radioactive Gaseous Waste Sampling and Analysis Program Table Notations [Cont'd]

i. Releases are made via the shield building vents only during PURGING, or operation of special ventilation systems. When ventilation fans in any vent path are not'in service for the entire sample period, in lieu of weekly removal and analysis of iodine and particulate coll~ction devices, these devices may be removed and analyzed following each release provided that the release lasts less than one week. Releases made via the plant ventilation paths as a result of routine surveillance tests, operational testing or scheduled short term maintenance activities of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or less do not require special sampling and analysis provided that plant conditions do not indicate the completion of these activities would cause an increase in the release of activity. Removal and analysis of collection .devices is not required if releases are not being made.

j. Grab samples for atmospheric steam releases are representative liquid grab samples from the respective steam generator.

k. Atmospheric steam releases are the timed releases of steam from the steam generators to the atmosphere via either the power operated reliefs, steam dump valves.or flash tank vents. It does not include steam. dumped via the condenser.-

I. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of steam released and in which the method of sampling employed results in a specimen which is representative of the total steam released form the respective steam generator.

m. Containment Purges includes PURGE releases with either the lnservice Purge or Containment Purge Fans and also VENTING of containment utilizing the Post Loca Vent System. When the release is completed via the Post Loca Vent, the pre-release tritium, particulate and charcoal samples should be used for all analyses, and continuous samples collected during the release are not required. During Cold*

Shutdown periods, the availability of ventilation systems and the position of containment air-lock doors may require that portions of the required samples be collected with installed continuous monitors or portable sampling equipment.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: *31 Page 110of162 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) .REV: 31 Page 111 of 162 Table 3.2 Radioactive Gaseous Effluent Monitoring Instrumentation MINIMUM CHANNELS INSTRUMENT FUNCTIONAL APPLICABILITY ACTION

1. Waste Gas Holdup System Explosive Gas 2 During system 2 (Oxygen) Monitors operation.

2. Effluent Release Points Unit 1 Aux Bldg.

Unit 2 Aux Bldg.

Rad Waste Bldg.

Spent Fuel Pool Unit 1 Shield Bldg.

Unit 2 .Shield Bldg.

a. Noble.Gas Activity Monitor* 1 During releases 3,4,6

b. Iodine Sampler Cartridge 1 During releases 2

c. Particulate Sampler Filter 1 During releases 2

d. Sampler Flow Integrator 1 During releases 1

Noble gas activity monitors providing automatic termination of releases (except the Radwaste Building which has no automatic isolation function).

With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels FUNCTIONAL, take the actions directed in Table 3.2. Restore the non-functional instrumentation to FUNCTIONAL status within 30 days.

If Radioactive Gaseous Effluent Monitoring Instrumentation is not restored within 30 days, explain in the next Annual Radioactive Effluent Report, why this non-functionality was not corrected in a timely manner.. The Waste Gas Holdup System Explosive Gas (Oxygen) Monitors are NOT subject to this requirement. They are subject to the Special Reporting requirements, as defined in.

section 3.9.4.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 112of162 Table 3.2 Radioactive Gaseous Effluent Monitoring Instrumentation Table Notations ACTION 1 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue provided that the flow rate is estimated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, operating of this system may continue for up to 14 days. With two channels not FUNTIONAL, manually isolate the oxygen addition line.

ACTION 3 With the numbers of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may continue provided that samples are collected with auxiliary sample equipment as required in Table 3.1.

ACTION 4 With the num.ber of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, effluent releases via this pathway may

continue provided that samples are taken and analyzed to LLD per Table 3.1, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 5 With the number of channels FUNCTIONAL less than required by the Minimum Channels FUNCTIONAL requirement, immediately suspend Purging of radioactive

effluents via this pathway during periods when containmentintegrity is required or the primary system is initially opened to the atmosphere. (applicable to Reactor

Building Vents)

ACTION 6 With the number of channels FUNCTIONAL less than required 'by the Minimum Channels FUNCTIONAL requirement, the contents of the waste gas decay tanks may be released to the environment provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge valve lineup;

Othervitise, suspend release of radioactive effluents via this pathway (applicable to Unit 2 Auxiliary Building Vent).

PRAIRIE ISLAND NUCLEAR GENERATING PLAl)IT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION 1-14 MANUAL (ODCM) REV: 31 Page 113of162 Table 3.3 - Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements CHANNEL SOURCE FUNCTIONAL CHECK. CHECK TEST CALIBRATION Instrument Frequency Frequency Frequency Frequency .

Waste Gas Holdup System Daily during ---- Monthfy<2) . Quarter1y<5l Explosive Gas (O)(ygen) system operation Monitors Effluent Release Points Unit 1 Aux Bldg.

Unit 2 Aux Bldg.

Rad Waste Bldg.

Spent Fuel Pool Unit 1 Shield Bldg.

Unit 2 Shield Bldg.

Noble Gas Activity Daily during Monthly* Quarter1y'1l At least once every

Monitor (4) releases 18 months<3l (Except Radwaste -

Building)

Noble Gas Activity Monitor Daily during Monthly. Quarter1y<2l At least once every Radwaste Building (4) releases 18 months(3)

Iodine and Particulate Weekly ---- ----

Samplers Sampler Flow Rate Monitor Weekly ---- ---- At least once every 18 months

A SOURCE CHECK of the applicable nobles gas monitor SHALL be conducted prior to each waste gas decay tank or containment purge release.

J

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 114of162 Table 3.3 - Radioactive Gas.eous Effluent Monitoring Instrumentation Surveillance Requirements Table Notations

1. The CHANNEL FUNCTIONAL TEST SHALL also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following exists. *

a. Instrument indicates measured levels above the alarm/trip setpoint.

b. Circuit failure (if provided).

c. Instrument indicates .a downscale failure (if provided).

d. Instrument controls not set in operate mode (if provided).

2. The CHANNEL FUNCTIONAL TEST SHALL also demonstrate that alarm .

annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the alarm/trip setpoint.

b. Circuit failure (if provided).

c. Instrument indicates a downscale failure (if provided).

d. Instrument controls not set in operate mode (if provided).

3. The initial CHANNEL CALIBRATION SHALL be performed using one or more of the reference standards certified by the National Institute of Standards and Technology or using sources traceable to NIST standards. These standards SHALL permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATIONS, sources that have been related to the initial calibration SHALL be used.

4. Noble gas monitor in the Radwaste Building vent not provided with automatic isolation trip.

5. The CHANNEL CALIBRATION SHALL include the use of a nitrogen zero gas and an oxygen span gas with a nominal concentration suitable for the range of the instrument.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 115of162 Table 4.1 Liquid Source Terms EFFLUENT WASTE EFFLUENT SGBD RADIONUCLIDE CONCENTRATION LIMIT Ai. A-l

{!:!Ci/ml}* {CiNr} (CiNr}

Mo-99 2E-4 6.42E-3 1.415E-2 1-131 1E-5 3.061E-2 4.11 E-2 Te-132 9E-5 2.12E-3 3.61E-3 1-132. 1E-3 2.83E-3 1.BBE-2 1-133 1E-6 2.365E-2 4.856E-2 Cs-134 9E-6 1.464E-1 4.047E-2 1-135 ' 3E-4 4.84E-3 1.792E-2 Cs-136 6E-5 5.743E-2 1.862E-2 Cs-137 1E-5 8.214E-2 2.69E-2 All Others 1E-7 0 2E-5 H-3 1E-2 1.89E2 1.41E2 Noble gases 2E-4 --- ---

TOTAL 1.894E2 1.412E2

Effluent Concentration Limit =

Ten times the values listed in 10 CFR-20.1001-20.2402, App. B, Table 2, Column 2.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 116of162 THIS PAGE IS LEFT INTENTIONALLY BLANK

(

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 117of162 Table 4.2 - Adult Ingestion Dose Values (Ait) for the Prairie Island Nuclear Generating Plant (mrem/Hr Per µCi/ml)

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG Gl-LLI H-3 O.OOE-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 2.26E-01 C-14 3.13E 04 6.26E 03 6.26E 03 6.26E 03 6.26E 03 6.26E 03* 6.26E 03 NA-24 4.07E 02 4.07E 02 4.07E 02 4.07E 02 4~07E 02 4.07E 02 4.07E 02 CR-51 O.OOE-01 O.OOE-01 1.27E 00 7.61E-01 2.81 E~01 1.69E 00 3.20E 02 MN-54 O.OOE-01 4.38E 03 8.35E 02 O.OOE-01 1.30E 03 O.OOE-01 1.34E 04 MN-56 O.OOE-01 1.10E 02 1.!;l5E 01 O.OOE-01 1.40E 02 O.OOE-01 3.51E 03 FE-55 6.58E 02 4~55E 02 1.06E 02 O.OOE-01 O.OOE-01 2.54E 02 2.61E 02

FE-59 1.04E 03 2.44E 03 9.36E 02 O.OOE-01 O.OOE-01 6.82E 02 8.14E 03 C0-57 O.OOE-01 2.10E 01 3.48E 01 O.OOE-01 O.OOE-01 O.OOE-01 5.32E 02 C0-58 O.OOE-01 8.92E 01 2.00E 02 .O.OOE-01 O.OOE-01 O.OOE-01 1.81E 03 C0-60 O.OOE-01 2.56E 02 *5.65E 02 O.OOE-01 O.OOE-01 O.OOE-01 4.81E 03 Nl-63 3.11E 04 2.16E 03 1.04E 03 O.OOE-01 O.OOE-01 O.OOE-01 4.50E 02

Nl-65 1.26E 02 1.64E 01 7.49E 00 O.OOE-01

O.OOE-01 O.OOE-01 4.17E 02 CU-64 O.OOE-01 9.97E 00 4.68E 00 O.OOE-01 2.51E 01 O.OOE-01 8.50E 02 ZN-65 2.32E 04 7.37E 04 3.33E 04 O.OOE-01 4.93E 04 O.OOE-01 4.64E 04 ZN-69 4.93E 01 9.43E 01 6.56E 00 O.OOE-01 6.13E 01 O.OOE-01 1.42E 01 BR-83 O.OOE-01 O.OOE-01 4.04E 01

O.OOE-01 O.OOE-01 O.OOE-01 5.82E 01 BR-84 O.OOE-01 O.OOE-01 5.24E 01 O.OOE-01 O.OOE-01 O.OOE-01 4.11E-04 BR-85 O.OOE-01 O.OOE-01 2.15E 00 O.OOE-01 O.OOE-01 O.OOE-01 1.01E-15 RB-86 O.OOE-01 1.01E 05 4.71E 04 O.OOE-01 O.OOE-01 O.OOE-01 1.99E 04 RB-88 O.OOE-01 2.90E 02 1.54E 02 O.OOE-01 O.OOE-01 O.OOE-01 4.00E-09 RB-89 O.OOE-01 1.92E 02 1.35E 02 O.OOE-01 O.OOE-01 O.OOE-01 1.12E-11 SR-89 2.21E 04 O.OOE~01 6.35E 02 O.OOE-01 O.OOE-01 O.OOE-01 3.55E 03 SR-90 5.44E 05 O.OOE-01 1.34E 05 O.OOE-01 O.OOE-01 O.OOE-01 1.57E 04 SR-91 4.07E 02 O.OOE-01 1.64E 01 O.OOE-01 O.OOE-01 O.OOE-01 1.94E 03 SR-92 1.54E 02 O.OOE-01 6.68E 00 O.OOE-01 O.OOE-01 O.OOE-01 3.06E 03 Y-90 5.76E-01 O.OOE-01 1.54E-02 O.OOE-01 O.OOE-01 O.OOE-01 6.10E 03 Y-91M 5.44E-03 O.OOE-01 2.11E-04 O.OOE-01 O.OOE-01 O.OOE-01 1.60E-02 Y-91 8.44E 00 O.OOE-01 2.26E-01 O.OOE-01 O.OOE-01 O.OOE-01 4.64E 03 Y-92 5.06E-02 O.OOE-01 1.48E-03 O.OOE-01 O.OOE-01 O.OOE-01 8.86E 02 Y-93 1.60E-01 O.OOE-01 4.43E-03 O.OOE-01 O.OOE-01 O.OOE:-01 5.09E 03 ZR-95 2.40E-01 7.70E-02 5.21 E-02 O.OOE-01 1.21E-01 O.OOE-01 2.44E 02 ZR-97 1.33E-02 2.68E-03 1.22E-03 O.OOE-01 4.04E-03 O.OOE-01 8.30E 02 NB-95 4.47E 02 2.48E 02 1.34E 02 O.OOE-01 2.46E 02 O.OOE-01 1.51E 04 NB-97.

3.76E 00 9.48E-01 3.46E-01 O.OOE-01 1.11E 00 O.OOE-01 3.50E 03 M0-99 O.OOE-01 1.03E 02 1.96E 01 O.OOE-01 2.34E 02 O.OOE-01 2.39E 02 TC-99M 8.87E-03 2:51 E-02 3.19E-01 O.OOE-01 3.81E-01 1.23E-02 1.48E 01 TC-101 9.12E-03 1.31 E-02 1.29E-01

O.OOE-01 2.37E-01 6.72E-03 .3.95E-14 RU-103 4.43E 00 O.OOE-01 1.91E 00 O.OOE-01 1.69E 01 O.OOE-01 5.17E 02 RU-105 3.69E-01 O.OOE-01 1.46E-01 O.OOE-01 4.76E 00 O.OOE-01 2.26E 02 RU-106 6.58E 01 O.OOE-01 8.33E 00 O.OOE-01 1.27E 02 O.OOE-01 4.26E 03

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 118of162 Table 4.2 - Adult Ingestion Dose Values (Ait) for the Prairie Island Nuclear Generating Plant (mrem/Hr Per µCi/ml)

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG Gl-LLI RH-105 2.92E 00 2.12E 00 1.40E 00 O.OOE-01 9.00E 00 O.OOE-01 3.38E 02 AG-110M 8.81E-01 8.15E-01 4.84E-01 O.OOE-01 1.60E 00 O.OOE-01 2.9E 02 SB-124 6.74E 00 1.27E-01 2.66E-01 1.63E-02 O.OOE-01 5.23E 00 1.91E02 SB-125 5.34E 00 5.75E-02 1.07E 00 4.74E-03 O.OOE-01 5.58E 02 4.72E 01 SB-126 2.75E 00 5.60E-02 9.94E-01 1.69E-02 O.OOE-01 1.69E 00 2.25E 02 TE-125M 2.57E 03 9.301; 02 3.44E 02 7.72E 02 1.04E 04 O.OOE-01 1.02E 04 TE-127M 6.48E 03 2.32E 03 7.90E 02 1.66E 03 2.63E 04 O.OOE-01 2.17E 04 TE-127

1.05E-02 3.78E 01 2.28E 01 7.80E 01 4.29E.02 O.OOE-01 8.31E 03 TE-129M 1.10E04 4.11E 03 1.74E 03 3.78E 03 4.60E 04 O.OOE-01 5.54E 04 TE-129 3.01E 01 1.13E01 7.33E 00 2.31E 01 1.26E 02 O.OOE-01 2.27E 01 TE-131M 1.66E 03 8.10E 02 6.75E 02 1.28E 03 8.21E 03 O.OOE-01 8.04E 04 TE-131 1.89E 01 7.88E 00 5.96E 00 1.55E 01 8.26E 01 O.OOE-01 2.67E 00 5TE-132 2.41 E 03 1.56E 03 1.47E 03 1.72E 03

1.50E04 O.OOE-01 7.38E 04 1-130 2.71 E 01 8.01E 01 3 ..16E 01 6.79E 03 . 1.25E 02 O.OOE-01 6.89E 01 1-131 1.49E 02 2.14E 02 1.22E 02 7.00E 04 3.66E 02 O.OOE-01 5.64E 01 1-132 7.29E 00 1.95E 01 6.82E 00 6.82E 02 3.11E01 O.OOE-01 3.66E 00.

1-133 5.10E 01 8.87E 01 2.70E 01 1.30E 04 1.55E 02 O.OOE-01 7.97E 01 1-134 3.81E 00 1.03E 01

3.70E 00 1.79E 02 1.64E 01 O.OOE-01 9.01E-03 1-135 1.59E 01 4.17E 01 1.54E 01 2.75E 03 6.68E 01 O.OOE-01 4.70E 01 CS-134 2.98E 05 7.09E 05 5.79E 05 O.OOE-01 2.29E 05 7.61E 04 1.24E 04 CS-136 3.12E 04 1.23E 05 8.'86E 04 O.OOE-01 6.85E 04 9.38E 03 1.40E 04 CS-137 3.82E 05 5.22E 05 3.42E 05 O.OOE-01 1.77E 05 5.89E 04 1.01E 04 CS-138 2.64E 02 5.22E 02 2.59E 02 O.OOE-01 3.84E 02 3.79E 01 2.23E-03 BA-139 9.29E-01 6.62E-04

2.72E-02 O.OOE-01 6.19E-04 3.75E-04 1.65E 00 BA-140 1.94E 02 2.44E-01 1.27E 01 O.OOE-01 8.30E-02 1.40E-01 4.00E 02 BA-141 4.51E-01 3.41E-0.4 1.52E-02 O.OOE-01 3.17E-04 1.93E-04 2.13E-10 BA-142 2.04E-01 2.10E-04 1.28E-02 O.OOE-01 1.77E-04 1.19E-04 2.37E-19 LA-140 1.50E-01 7.54E-02 1.99E-02 . O.OOE-01 O.OOE-01 O.OOE-01 5.54E 03 LA-142 7.66E-03 3.48E-03 8.68E-04. O.OOE-01 O.OOE-01 O.OOE-01 2.54E 01 CE-141 2.24E-02 1.52E-02 1.72E-03 O.OOE-01 7.04E-03 O.OOE~01 5.79E 01 CE-143 3.95E-03 2.92E 00 3.23E-04 O.OOE-01 1.29E-03 O.OOE-01 1.09E 02 CE-144 1.17E 00 4.88E-01 6.27E-02 O.OOE-01 2.90E-01 O.OOE-01 3.95E 02 PR-143 5.51E-01 2.21E-01 2.73E-02 O.OOE-01 1.27E-01 O.OOE-01 2.41E 03 PR-144 1.80E-03 7.48E-04 9.16E-05 O.OOE-01 4.22E-'04 0.00E-01. 2.59E-10 ND-147 3.76E-01 4.35E-01 2.60E~02 O.OOE-01 2.54E-01 O.OOE-01 2.09E 03 W-187 2.96E 02 2.47E 02 8.65E 01 O.OOE-01 O.OOE-01 O.OOE-01 8.10E 04 NP-239 2.85E-02 2.80E-03 1.54E-03 O.OOE-01 8.74E-03 O.OOE-01 5.75E 02 The values in the above table are calculated utilizing an adult fish consumption of 21 Kg/yr.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE

.NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 119of162 Table 5.1 - Monitor Alarm Setpoint Determination for PINGP SOURCE Dis~ersion EFFLUENT FLOW RELEASE RELEASE SOURCE OF TERMS {A;} factor selection* RATE FRACTION 3

MONITOR POINT RELEASE* {TABLE 5.2} XIQ {sec/m } {F} {cfm} {Tm}

1R-30 Aux. Bldg. Aux. Bldg. Long Term and Vent - Unit 1 Unit 1 Exhaust Aux. BldQ. Release 2.9E+4 0.2 1R-37 Air Ejector

Air Ejector NA 2.9E+4 Unit 1 2R-30 Aux. Bldg. Aux. Bldg. - Long Term and Vent - Unit 2 Unit 2 Exhaust Aux. Bldg. Release 4.1E+4 0.3 2R-37 Gas Decay Xe-133 (100%) Short Term 4.1E+4 Tanks Release Air Ejector Air Ejector NA 4.1E+4 Unit2 1R-12 and Shield Bldg. Cont. - Units 1&2 1R-22 Vent- Unit 1 Purge, Unit 1 Shield Bldg. Short Term 3.2E+4 0.3 lnservice PurQe Release (Note 2) 2R-12 and Shield Bldg. Cont. - Unit 2 Shield Bldg. Short Term 4.6E+3 0.3 2R-22 Vent - Unit 2 lnservice PurQe Release R-35 Radwaste Bldg. Radwaste Bldg. Aux. Sldg. Long Term 6.1E+3 0.1 Vent Exhaust Release R-25 and Spent Fuel Pool Air Spent Fuel Pool Air Aux. Bldg. Long Term 1.8E+4 0.1 R-31 Vent Exhaust Release

Current dispersion.factors are maintained in H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data".

Values listed for Tm are nominal values only. They may be adjusted as necessary to allow a reasonable margin to the monitor setpoint. Duplicate values of Tm are assigned to both Shield Building vents since only one containment will be purged at' any one time. The assigned Tm values of all active release points SHALL NOT be greater than unity.

When purging the Unit 1 containment via the inservice purge system, the monitor setpoints may be based on 4.6E+3 cfm for the duration of the release.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION

H4 MANUAL (ODCM) REV: 31 Page 120of162 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 121 of 162 Table 5.2 Gas~ous Source Terms **

AUX.BLDG SHIELD BLDG. AIR EJECTOR

RADIONUCLIDE Ai (CiNr) Ai (CiNr) Ai (CiNr)

Kr-85m 3EO - 2EO Kr-85 2EO 2.2E1 -

Kr-87 1EO - -

Kr-88 5EO 1EO 3EO Xe-131m 2EO 2.1 E1 1EO Xe-133m

5EO' 2E1 3EO Xe-133 3.7E2 2.7E3 2.3E2 Xe-135 BED 6EO 5EO Xe-'138 1EO - -

TOTAL 3.97E2 2.77E3

2.44E2 '

"-" indicates that the release is less than 1 Ci/yr.

\;

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

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PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4

. MANUAL (ODCM) REV: 31 Page 123of162 Table 5.3 Critical Organ Dose Values (Pi) for Child mrem/~r ISOTOPE (Pt) µCi/m H-3 1.12E3 Cr-51 1.70 E 4 Mn-54 1.58 E 6 Fe-59 1:27 E 6 Co-58 1.11E6 Co-60 7.07 E 6 Zn-65 9.95 E 5 Rb-86 1.98 E 5 Sr-89 2.16 E 6 Sr-90 1.01 E 8 Y-91. 2.63 E 6 Zr-95 2.23 E 6 Nb-95 6.14 E 5 Ru-103 6.62 E 5 Ru-106 *

1.43 E 7 Ag-110m 5.48 E 6 Te-127m 1.48 E 6 Te-129m 1.76 E 6 Cs-134 1.01 E 6 Cs-136 1.71E5 Cs-137 9.07 E 5 Ba-140 1.74 E 6 Ce-141 5.44 E 5 Ce-144 1.20 E 7

1-131 1.62 E 7

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PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE *oosE CALCULATION H4

. MANUAL (ODCM) REV: 31 Page 125of162 Table 5.4 Dose Factors for Noble Gases

Total Body Dose Beta Air Dose Factor Skin Dose Factor Gamma Air Dose Factor Ki Li Factor Mi " Ni (mrem/yr per (mrem/yr per (mrad/yr per (mrad/yr per Radionuclide µCi/m 3 ) µCi/m 3 ) µCi/m 3 ) µCi/m 3 )

Kr-83m 7.56E-02 ---- 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03. 1.23E+03 1.97E+03

-Kr-85 1.61 E+01 1;34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 .

1.01E+04* 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 163E+04 7.83E+03 Xe-131m 9.15E+01 . 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 *3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 . 1.92E+03 2.46E+03 Xe-137 1.42E+03

1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03

9.21 E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03

The listed dose factors are for radionuclides that may be detected in gaseous effluents.

All others are 0.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 127of162 Table 7.1 Radiological Environmental Monitoring Program Sample Collection and Analysis Exposure Pathway Number of Samples Sampling and Type and Frequency and/or Sam le and Sam le Locations** Collection Fre uenc of Anal sis

1. AIRBORNE Samples from 5 locations: Continu.ous Sampler operation Radioiodine analysis weekly for Radioiodine and a. Three samples from close to the three

with sample collection weekly 1-131 Particulates SITE BOUNDARY locations (in different sectors) of the highest Particulate:

calculated annual average ground level *. Gross beta activity on each filter D/Q; weekly*. Analysis SHALL be *

b. One sample from the vicinity of a performed more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months community having the highest following filter change. Perform calculated annual average ground level gamma isotopic analysis on D/Q. composite (by location) sample c.. One sample from a control location quarterly.

specified in the REMP.

2. DIRECT RADIATION 32 TLD stations established with duplicate* Quarterly Gamma dose dosimeters placed at the following quarterly locations:

1. Using the 16 meteorological wind sectors as guidelines, an inner ring of stations in.the general area of the site boundary is established and an outer ring of stations in the 4 to 5 mile distance from the plant site is established. Because of inaccessibility, seven sectors in the inner and outer rings are not covered

.

If Gross beta activity in any indictor sample exceeds 10 times the yearly average of the control sample, a gamma isotopic analysis is required.

- Sample locations are further described by the REMP.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE

- ~*-

NUMBER:

,H OFFSITE DOSE CALCULATION MANUAL (ODCM) REV:

H4 31 j

Page 128of162 Table 7 .1 Radiological Environmental Monitoring Program Sample Collection and Analysis.

Exposure Pathway Number of Samples Sampling and Type and Frequency and/or Sam le and Sam le .Locations** Collection Fre -of Anal sis

2. DIRECT RADIATION

[Cont'd]

2. Seven dosimeters are established at special interest areas and a control station.

3. WATERBORNE

a. Surface Upstream & <;lownstream locations Monthly Composite of weekly Gamma isotopic analysis of each samples (water & ice conditions monthly composite permitting)

Tritium analysis of quarterly composites of monthly composites

b. Ground 3 samples from wells within 5 miles *of the Quarterly Gamma isotopic and tritium plant site and 1 sample from a well greater analyses of each sample than 10 miles from the plant site

c. Drinking 1 sample from the City of Red Wing water Monthly Composite of weekly 1-131 Analysis and Gross beta supply samples and gamma isotopic analyses of each monthly composite Tritium analysis of quarterly composites of monthly

composites

d. All Surface, ground, and drinking water Monthly, Quarterly, or Annually Tritium analysis, Gamma samples in support of the NEI analysis, and hard-to-detect Groundwater Protection Initiative analysis of selected samples

- Sample locations are further described by the REMP.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 129of162 Table 7.1 Radiological Environmental Monitoring Program Sample Collection and Analysis Exposure Pathway Number of Samples Type and Frequency and/or Sam le and Sam le Locations** of Anal sis

3. WATERBORNE

[Cont'd]

d. Sediment from One sample upstream of plant, one Semiannually Gamma isotopic analysis of each shoreline sample downstream of plant, and one sample from shoreline of recreational area.

4. INGESTION

a. Milk* One sample from dairy farm having Semimonthly when animals are Gamma isotopic and 1-131 highest D/Q, one sample from each of on pasture; monthly at other analysis of each sample three dairy farms calculated to have doses times.

from 1-131 >

1 mRem/yr, and one sample from 10..:20

-* miles

b. Fish and One sample of one game specie of fish Semiannually Gamma isotopic analyses on Invertebrates

located upstream and downstream of the each sample (edible portion only

plant site on fish)

One sample of Invertebrates upstream and downstream of the plant site

- Sample locations are further described by the REMP.

- J

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

.... ;., . ;';::1'.:~~1\<o )&~L!'.c))'. ::;;;_J OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 130of162 Table 7.1 Radiological Environmental Monitoring Program Sample Collection and Analysis Exposure Pathway Number of Samples Sampling and Type and Frequency and/or Sam le and Sam le Locations** Collection Fre uenc of Anal sis

4. INGESTION

[Cont'd]

c. Food Products One sample of corn from any field that is At time of harvest Gamma isotopic analysis of irrigated by water into which liquid plant edible portion of each sample wastes have been discharged***

One sample of broad leaf vegetation from *At time of harvest 1-131 analyses of edible portion highest D/Q garden and one sample from of each. sample 10-20 miles

- Sample locations are further described by the REMP.

- - As determined by methods outlined in the ODCM.

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) . REV: 31 Page 131 of 162 Table 7.2 - Reporting Levels for Radioactivity Concentration in Environmental Samples WATER AIRBORNE FISH MILK *FOOD ANALYSIS (pCi/I) PARTICULATE OR. (pCi/kg, wet) (pCi/I.) PRODUCTS GASES (pCi/m 3) (pCi/kg, wet)

H-3 20 ooo<a)

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000. 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95 400(b) 1-131 2<a) 0.9 3 100 Cs-134 . 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 2oo 300(b)

(a) Drinking water pathway level.

(b) Total for parent and daughter.

. H PROCEDURE PRAIRIE ISLAND NUCLEAR GENERATING PLANT NUMBER:

H4 OFFSITE DOSE CALCULATION MANUAL (ODCM) REV: 31 Page 132of162 THIS PAGE IS LEFT INTENTIONALLY BLANK

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:.

OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 133of162 Table 7.3 - Detection Capabilities for Environmental Sample Analysis Lower Limit of Detection *(LLD )(a)

ANALYSIS WATER AIRBORNE FISH MILK FOOD SEDIMENT (pCi/I) PARTICULATE (pCi/kg, wet) PRODUCTS (pCi/kg, dry)

(pCi/I) .

ORGASES . (pCi/kg, wet)

(pCi/m 3 )

Gross Beta 4 0.01 H-3. 2 ooo<b)

Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-Nb-95 15 (c) 1-131 (d) 1(b) 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 *0.06 150 18 80 180 Ba-La-140 15(c) 15(c)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:

. . El . ---~'
~ ... ., ~ ...._4.-..- .* :**~~ :*,_..;:,
r;*....,,,,~_...,::...* ...."~~--... .:.
OFFSITE DOSE CALCULATION MANUAL (ODCM) REV:
H4 31
-1 *I Page 134of162 Table 7 .3 - Table Notation a - The LLD is the smallest concentration of radioactive material in a sample that will be detected wHh 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system*(which may include radiochemical separation):*
4.66sb LLD = ---* ------"----
E*v*2.22*Y*exp(-Mt)
Where:
LLD is the appropriate lower limit of detection as defined above (as Pico curie per unit mass or volume), Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute). In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background SHALL include the typical contributing of other radionuclides normally present in.the samples (e.g., potassium-40 in milk samples). Typical values of E, V, Y and L\.'t SHALL be used in the calculations.
E is the counting efficiency (as counts per transformation),
2.22 is the number of transformation per minute per Pico curie, r
Y is the fractional radiochemical yield (when applicable),
'A is the radioactive decay constant for the particular radionuclide, and L\'t is the elapsed time between sample collection (or end of the sample collection period) and time of counting.
b - Drinking water pathway limit.
c - Total for parent and daughter d - These LLDs*apply only where "131 1analysis" is specified.
e - Where "Gamma Isotopic Analysis" is specified, the LLD specification applies to th~ following radionuclides:
95 54 Mn, 59 Fe, 58 Co, 6°Co, 65Zn, Zr-Nb, 137Cs, 134Cs, and 140Ba-La. Other peaks which are measurable and identifiable, together with the above nuclides, SHALL also be identified and reported.
PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:
OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 135of162 Figure 3.1 - Prairie Island Nuclear Generating Plant Site Boundary For Liquid Effluents
.. ()
Plant Scre¢~H6u~e *
--~-
_ Cooling Tower .
_ Corifr~l !1ol!s~5 \
LEGEl'.JD
~ - - --S:iTE - *06U;;iLJAl='<Y- FOil
" -- - ' t,lOf..;liO- Ef'FLLJENIS
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 137of162 Figure 3.2 *- Prairie Island Nuclear Generating Plant Site Boundary For Gaseous Effluents
-- 0 z
c .....
... c
~j c
z w "'.., ::;
I!)
w 0
I;-
0 * ..
_J - .. up,,,
A.I Cf) 0 ID
.C 0
[&.,,P.,.
0
~
~
I') r-0:;;
,; al Ill wzfu g11.1J ,,

iii /~;/
l J f- w :JI- /. :r~
~~ 'i,_ "'-z /:',-_,,.
~~ ~~ ~

~-"--L _______ -" , ,

--<~ I
II
/,_J II
PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER:
.t I;~M
'* *'" -.... ..,.:.,-"* ~ 1. ..... , *'.
. OFFSITE DOSE CALCULATION MANUAL (ODCM) REV:
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MANUAL (ODCM) REV: 31 Page 139of162 Appendix A Meteorological Analyses Table A-1 Release Conditions Table A-2 Distance to Site Boundary
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 141 of 162 Appendix A Summary of Dispersion Calculation Procedures Undepleted, undecayed dispersion parameters were computed using the computer program XOQDOQ (Sagendorf and Goll, 1977). Specifically, sector average x!Q and D/Q values were obtained for a sector width .of22.5 degrees. Building wake corrections were used to adjust calculations for ground-level releases. Standard open terrain recirculation correction factors were also applied as available as default values in XOQDOQ.
Dispersion calculations were based on ground level releases for the shield buildings, turbine buildings, and auxiliary building (hereafter referred to as the plant complex). A summary of release conditions used as input-to XOQDOQ is presented in Table A-1 and controlling site boundary distances are defined in Table A-2. Computed X/Q and D/Q values for site boundary locations (relative to release points) and for standard distances to five miles are maintained in H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data".
Onsite meteorological data is collected over a representative time period. A 5 year period is suggested to ensure year to year variances do not bias the data set. This data reduced to joint frequency tables and used as input to the XOQDOQ determinations. Data is collected and delta-T stability classes are defined in conformance with NRC Regulatory Guide 1.23. Dispersion calculations for the plant complex is based on delta-T for 60 meter and 10 meter Go int data recovery of 90 percent. Joint frequency tables and resultant XOQDOQ determinations are maintaine<:I H4.2, "OFFSITE DOSE CALCULATION MANUAL (ODCM) SUPPORTING DATA". Meteorological data may be reassessed periodically to assure proper representation of local meteorological profiling.
REFERENCES
1. Sagendorf, J.F. and Goll, J.T., XOQDOQ Program for the Evaluation of Routine Effluent Releases at Nuclear Power Stations, NUREG-0324,_ U.S. Nuclear Regulatory Commission, September 1977.
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 143of162 Table A-1 Prairie Island Release Conditions Shield Buildings Auxiliarv Building Turbine Building Type Release Ground Level Ground Level Ground Level
{Long Term and Short Term} (Long Term) (Long Term)
Release Point Height (m) 56.4 24.4 33.6, 12.2 Adjacent Building Height 62.2 62.2* 62._2*
Relative Location to Adjacent Structures Adjacent to Adjacent to Adjacent to Auxiliary Building AuxiliaQL Building Auxiliary Building Exit Velocity (m/sec). . N.A. N.A. N.A.
Internal Stack Diameter (m) N.A. N.A. N.A.
Building Cross-Sectional Area (m 2) 2.170 2.170** 2.170**
Purge Freguency *** (times/yr) 20 N.A. N.A.
Purge Duration*** (hours/release) 5 N.A. N.A.
Height of Shield Buildings
Shield Building cross-sectional area
Applied to short-term calculations only
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Table A-2 Distances (Miles) to Controlling Site Boundary Locations
As Measured from Edge of Plant Complex Sector Distance N 0.28 NNE 0.26 NE 0.84*
ENE 0.62*
E 0.59*
ESE 0.61*
SE 0.67 SSE 0.43 s 0.43 SSW 0.40 SW 0.40 WSW 0.37 w 0.36 WNW 0.36 NW 0.43 NNW 0.48
Over-water distances
PRAIRIE ISLAND NUCLEAR GENERATING PLANT H PROCEDURE NUMBER: .
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MANUAL (ODCM) REV: 31 Page 147 of 162 Appendix 8 ~ose Parameters for Radioiodines, Particulates and Tritium This appendix contains the methodology which was used to calculate the dose parameters for radioiodines, particulates, and tritium to show compliance with 10 CFR 20 and Appendix I of 10 CFR 50 for gaseous effluents. These dose parameters, Pi and Ri, were calculated using the methodology outlines in NUREG-0133 along with Regulatory Guide 1.109 Revision 1. The following sections provide the specific methodology which was utilized in calculating the Pi and Ri values for the various exposure pathways.
8.1 Calculation of Pi The parameter, Pi, contained in the radioiodine and particulates portion of Section 5.2, includes pathway transport parameters of the ith radionuclide, the receptor's usage of the pathway media and the dosimetry of the .exposure .. Pathway usage rates a.nd the internal dosimetry are functions of the receptor's age: however, the child age *group, will always receive the maximum dose under the exposure conditions assumed.
8.1.1 Inhalation Pathway Pi = K' (BR) DFAi (B.1-1)
I..
Where:
Pi = dose parameter for radionuclide i. for.the inhalation pathway, mrem/yr per µci/_m3; K' = a constant of unit conversion:
6
= 10 pCi/µCi; BR = the breathing rate of the child age group, 3
. m /yr; DFAi = the maximum organ inhalation dose factor for the child age group for radionuclide i, mrem/pCi.
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 148of162 Appendix 8 Dose Parameters for Radioiodiries, Particulates and Tritium The age group considered is the child group. The child's breathing rate is taken as 3700 m3/yr from Table E-5 of Regulatory Guide 1.109 Revision 1. The inhalation dose factors for the child OFAi, are presented in Table E-9 of Regulatory Guide 1.109 in units of mrem/pCi. The to~al body is considered as an organ in the selection of DFAj. The incorporation of breathing rate of the child and the unit conversion factor results in the following: *
= 3.7 x 109 DFAi (B.1-2) 8.2 Calculation of Ri The radioiodine and particulate specification is applicable to the location in the unrestricted area where the combination of existing. pathways and receptor age groups indicates the maximum potential exposure occurs. The inhalation an*d ground plane exposure pathways SHALL be considered to exist at all locations. The grass-goat-milk, the grass-cow-milk, grass-cow-meat, and vegetation pathways are considered based on their existence at the various locations. Ri values have been calculated for the adult, teen, child, and infant age groups for the ground plane, cow milk, goat milk, vegetable and beef ingestion pathways. The methodology which was utilized to calculate these values is presented below.
8.2.1 Inhalation Pathway Rf =K' (BR)a (DFAi)a (B.2-1)
I*
where:
R~t = dose factor for each identified radionuclide 3
I of the organ of interest, mrem/yr per µCi/m K' = a constant of unit conversion:
= 106 pCi/µCi; I
(BR)a = breathing rate of the receptor of age.group a,
. m3/yr;
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 149of162 Appendix 8 Dose Parameters for Radioiodines, Particulates and Tritium (DFAi)a= organ inhalation dose factor for radionuclide i for the receptor of age group a, mrem/pCi.
The breathing rates (BR)a for the various age groups are tabulated below, as given in Table E-5 of the Regulatory Guide 1.109 Revision 1.
Age Group (a) Breathing Rate (m 3/yr)
Infant 1400 Child 3700 Teen 8000 Adult 8000 Inhalation dose factors (DFAi)a for the various age groups are given in Tables E-7 through E-10 of Regulatory Guide 1.109 Revision 1.
8.2.2 Ground Plane Pathway (8.2-2) where:
Rf = dose factor for the ground plane pathway for each. identified radionuclide i for the organ of interest, m2 -mrem/yr per µCi/sec per; K' = a constant of unit conversion; 6
= 10 pCi/µCi;
K" = a constant of unit conversion;
= 8760 hr/year; A.i = the radiological decay constant for 1
radionuclide i, sec- ;
t = the exposure time, sec; 8 .
= 4.73X10* sec (15 years)'
DFGi = the ground plant dose conversion factor for 2
radionuclide i; mrem/hr per pCi/m ;
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' .*: .. _..~ ./.., ; .*...\' :~,:. ~2~.,,-0-*~-.,. ..,. i
-:~r~--<" cif~::Q"~:.:~~f Page 150of162 Appendix B Dose Parameters for Radioiodines, Particulates and Tritium SF = the shielding factor (dimensionless)
Ii = factor to account for fractional deposition of radionuclide i.
For radionuclides other than iodine, the factor Ii is equal to one. For radioiodines, the value of h may vary. However, a value of 1.0 was used in calculating the R values in H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data".
A shielding factor of 0. 7 from Table E-15 of Regulatory Guide 1.109 Revision 1 is used. A tabulation of DFGi values is presented in Table E-6 of Regulatory Guide 1 .109 Revision 1.
8.2.3 Grass-Cow or Goat-Milk Pathway where:
Rf'1M = dose factor for the cow milk or goat milk pathway, for each identified radionuclide i for the* organ of interest, m 2 - mrem/yr per µCi/sec; K' = a constant of unit conversion;
= 106 pCi/µCi; OF = the cow's or goat's feed consumption rate, kg/day (wet weight);
Uap = the receptor's milk consumption rate for age group a, liters/yr;
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) . REV: 31 Page 151 of 162 Appendix B Dose Parameters for Radioiodines, Particulates and Tritium Yp = the agricultural productivity 2
by unit area of pasture feed grass, kg/m ;
2 Ys = the agricultural productivity by unit areas of stored feed, kg/m ;
Fm = the stable element transfer coefficients, pCi/liter per pCi/day;
r = fraction of deposited activity retained on cow's feed grass; (DFLi)a = the organ ingestion dose factor for radionuclide I for the receptor in age group a, mrem/pCI; A.i = the radiological decay constant for re)dionuclide I,
-1 sec ;
A.w = the decay constant for removal of activity on leaf and
-1 plant surfaces by weathering, sec
-7 -1 .
= 5.73 X 10 sec (corresponding to a 14 day half-lift);
tf = the transport time from feed to cow or goat to milk to receptor, sec; th = the transport time from harvest, to cow or goat, to consumption, sec; tb = period of time that activity builds up in soil, sec; Biv = concentration factor for uptake of radionuclide i from the soil by the edible parts of crops, pCi/kg (wet weight) per PCi/kg (dry soil);
P = effective surface density for soil, (dry weight) kg/m 2 ;
fp = .fraction of the year that the cow or goat is on pasture; fs = fraction of the cow feed that is pasture grass while the cow is on pasture;
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 152of162 Appendix B Dose Parameters for Radioiodines, Particulates and Tritium tep = period of pasture grass exposure during the growing season, sec; tes = period .of crop exposure during the growing season, sec; h = factor to account for fractional deposition of radionuclide i.
For radionuclides other than iodine, the factor h is equal to one. For radioiodines, the value of Ii may vary. However, a value of 1.0 was used in calculating the R values in H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data".
Milk cattle and goats are considered to be fed from two potential sources, pasture grass and stored feeds. Following the development in Regulatory Guide 1.109 Revision 1, the value offs.was considered unity in lieu of site-specific information. The value of fp was 0.5 based upon a 6-month grazing period.
Table 8-1 contains the appropriate parameter values and their source in Regulatory Guide 1 .109 Revision 1 .
The concentration of tritium in milk is based on the airborne concentration rather than the deposition. Therefore, the Riis based on X/Q:
R't/_ 3 = K'K" FmOFUap(DFLi)a 0.75 (0.5/H) (B.2-4) where:
' RH-3 M. = dose factor for the cow or goat milk pathway for tritium for the organ 3
of interest, mrem/yr per µCitm
K" = a constant of unit conversion; 3
= 10 gm/kg; H = absolute humidity of the atmosphere, gm/m 3 ;
0. 75 = the fraction of total feed that is water; 0.5 = the ratio of the specific activity of the feed grass to the atmospheric.water.
and other parameters and values are given below. A value. of Hof 8 grams/meter3 , was used in lieu of site-specific informatic;m.
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 153of162 Appendix B Dose Parameters for Radioiodines, Particulates and Tritium B.2.4 Grass-Cow-Meat Pathway The integrate*d concentration in meat follows *in a similar manner to the development for the milk pathway, therefore:*
(B.2-5) where:
= dose factor for the meat ingestion *pathway for radionuclide i for any organ of interest, m2 - mrem/yr per µCi/sec; Ft = the stable element transfer coefficients, pCi/Kg per pCi/day; Uap = the receptor's meat consumption rate for age group a, kg/yr; ts = the transport time from slaughter to consumption, sec; th = the transport time from harvest to animal consumption, sec; tep = period of pasture grass exposure during the growing season, sec; tes = period of crop exposure during the growing season, sec; lj = factor to account for fractional disposition of radionuclide i.
For radionuclides other than iodine, the factor h is equal to one. For radioiodines, the value of h may vary. However, a value of 1.0 was used in calculating the R values in H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data".
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 154of162 Appendix 8 Dose Parameters for Radioiodines, Particulates and Tritium All other terms remain the same as defined in Equation B.2-3. Table B-2 contains the values which were used in calculating Ri for the meat pathway.
The concentration of tritium in meat is based on its airborne concentration rather than ttie deposition. Therefore, the Riis based on X/Q.
(B.2-6) where:
= dose factor for the meat ingestion pathway for tritium for any organ of interest, mrem/yr per µCi/m 3 .
All other terms are defined in Equation B.2-4 and B.2-5, above.
8.2.5 Vegetation Pathway The integrated concentration in vegetation consumed by man follows the.
expression developed in the derivation of the milk factor. Man is considered to consume two types of vegetation (fresh and stored) that differ only in the time period between harvest and consumption, therefore:
= lj K'(DFLi)a [u~fLe-A.itL*[r(l-e-A.Eite) + Biv(l-e-A.itb)] +
.YvA.Ei P,i.i
(B.2-7) where:
dose factor for vegetable pathway for radionuclide i for organ of interest, m 2 - mrem/yr per µCi/sec; K' = a constant of unit conversion;
= 106 pCi/µCi;
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 155 of162 Appendix B Dose Parameters for Radioiodines, Particulates and Tritium L
Ua = _the consumption rate of fresh leafy vegetation by the receptor in age group a, kg/yr; usa = the consumption the or stored vegetation by the receptor in age group* a, kg/yr; fL = the fraction of the annual intake of fresh leafy vegetation grown
-locally; fg = the fraction of the annual intake of stored vegetation grown locally; tL = the average time between harvest of leafy vegetation and its consumption, sec; th = the average time between harvest of stored vegetation and its consumption, sec; Yv = the vegetation aerial density, kg/m 2 ;
I
= period of leafy vegetable exposure during growing season, sec;
= factor to account for fractional deposition of radionuclid~ i.
For radionuclides other than iodine, the factor h is equal to one. For radioiodines, the value of Ii may vary. However, a value of 1.0 was used in calculating the R values in H4.2, "Offsite Dose Calculation Manual (ODCM) Supporting Data".
All other factors were defined above.
Table B-3 presents the appropriate parameter values and their $OUrce in Regulatory Guide 1.109 Revision 1.
In lieu of site-specific data default values for fL and f9 , 1.0 and 0.76, respectively were used in the calculation of Rj. These values were obtained from Table E-15 of Regulatory Guide 1.109 Revision 1.
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 156of162 Appendix 8 Dose Parameters for Radioiodines, Particulates and Tritium The concentration of tritium in vegetation is based on the airborne concentration rather than the deposition. Therefore, the Ri is based on X/Q:
R~-3 = K'K'; [U~ fL + U~ f 9 ] (DFLi)a 0.75 (0.5/H) (B.2-8) where:
V RH-3-- dose factor for the vegetable pathway for tritium for any organ of interest, m 2 - mrem/yr per Ci/m 3 .
All other terms remain the same as those in Equations B.2-4 and B.2-7.
The concentration of Carbon-14 in milk, meat, or vegetation, is based on the airborne concentration rather than the deposition. Therefore, the Ri is based on X/Q:
9 (R C-14)aj = 10
Uc-14
0.11 * (DFLc-14)aj
1/0.19 . (B.2-9) where:
(R C-14)aj = Site specific Carbon-14 Dose Factor, for age group a, organ j , 1 mrem/yr per µCi/m3 10 9
= a constant of unit conversion (pCi/uCi, gm/Kg)
Uc-14 = Annual Carbon Ingestion via specific Pathway in Kg-Carbon per year for age group a 0.11 = Carbon Fraction (regulatory guide 1.109, Revision 1)
(DFLc-14)aj = C-14 Ingestion Dose Factor in mrem/pCi for age group a and organ j 0.19 = Atmospheric Concentration of Natural Carbon in gm/m 3
based on 383 ppm
stated value in Regulatory Guide 1.109, Revision 1, is 0.16. Due to atmospheric changes, latest EPA data is 0.19.
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Page 157of162 Table B-1 Parameters for Cow and Goat Milk Pathways
.Parameter Value Reference (Reg. Guide 1.109 Rev. 1)
OF (kg/day) 50 (cow) Table E-3 6 (goat)
Table E-3 Yp (kg/m 2) 0.7 Table E-15 5 Table E-15 if (seconds) 1.73x10 (2 days) r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 (DFLi)a (mrem/pCi)
Each radionuclide Tables E-11 to E-14 Fm (pCi/day per pCi/liter) Each stable element Table E-1 (cow)
Table E-2 (goat) 8 Table E-15 tb (seconds)
4.73x10 (15yr)
Ys (kg/m 2 ) 2.0. Table E-15
,yp (kg/m2 ) 0.7 Table E-15
th (seconds) 7.78 x 106 (90 days) Table E-15 Uap (liters/yr)
330 infant Table E-5 330 child Table E-5 400teen Table E-5 310 adult Table.E-5 6 Table E~15 tep (seconds) 2.59 x 10 (30 days)
. 6 .
tes (seconds) 5.18x10 (60 days) Table E-.15 Biv (pCi/Kg (wet weight) Each stable element Table E-1
per pCi/Kg (dry soil))
P (Kgtrr? (dry weight)) 240 Table E~15
H PROCEDURE PRAIRIE ISLAND NUCLEAR GENERATING PLANT NUMBER:
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 159of162 Table 8-2 Parameters for the Meat Pathway Parameter Value Reference (Reg. Guide 1.109 Rev. 1) r
1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 Ft (pCi/Kg per pCi/day) Each stable element Table E-1 Uap (Kg/yr) 0 infant Table E-5 41 child Table E-5 65 teen Table E-5 110 adult Table E-5 (DFLi)a (mrem/pCi) Each radionuclide Tables E-11 to E-14 2 0.7 Table E-15 Yp (kg/m )
2 2.0 Table E-15 Ys (kg/m )
8 Table E-15 tt) (seconds) 4.73x10 (15 yr) 6 Table E-15 ts (seconds) 1.73 x 10 (20 days) 6 . Table E-15 th (seconds) 7.78 x 1O (90 days) 6
Table E-15
tep (seconds) 2.59 x 1O (30 days) 6 Table E-15 tes (seconds) 5.18x10 (60 days)
Ot (kg/day) 50 Table E-3 Biv (pCi/Kg (wet weight) Each stable element Table E-1 per pCi/Kg (dry soil))
2 240 Table E-15 P (Kg/m (dry weight))
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OFFSITE DOSE CALCULATION H4 MANUAL (ODCM) REV: 31 Page 161 of 162 Table 8-3 Parameters for the Vegetable Pathway Parameter Value Reference (Reg. Guide 1.109 Rev. 1) r (dimensionless) 1.0 (radioiodines) Table E-1 0.2 (particulates) Table E-1 (DFLi)a (mrem/pCi) Each radionuclide Tables E-11 to E-14 L 0 Table E-5 Ua (kg/yr) - Infant
-Child 26 Table E-5
-Teen 42 Table E-5
. -Adult 64 Table E-5 s
Ua (kg/yr) - Infant 0 Table E-5
- Child 520 Table E-5
-Teen 630 Table E-5
-Adult 520 Table E-5 4 Table E-15 I
tL (seconds) 8.6 x 10 (1 day)
. 6 th (seconds) 5.18x10 (60 days) Table E-15 Yv (kg/m 2) 2.0 Table E-15 6 . Table E-15 te (seconds) 5.18x10 (60 days) 8 Table E-15 tb (seconds) 4.73 x 10 (15 yr)
. 2 .
P(Kg/m (dry weight)) 240 Table E-15 Biv (pCi/Kg (wet weight) Each stable element Table E-1 per pCi/kg (dry soil))*
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ML17130A971

Contents

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2.1.2 APPLICABILITY

2.1.3 ACTION

2.3.2 APPLICABILITY

2.3.3 ACTION

2.5.3 ACTION

2.7.2 APPLICABILITY

2.7.3 ACTION

2.9.2 APPLICABILITY

2.9.3 ACTION

3.1.2 APPLICABILITY

3.1.3 ACTION

3.3.2 APPLICABILITY

3.3.3 ACTION

3.5.3 APPLICABILITY

3.5.4 ACTION

3.7.4 APPLICABILITY

3.7.5 ACTION

3.9.3 APPLICABILITY

6.1.2 APPLICABILITY

6.1.3 ACTION

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2.1.2 APPLICABILITY

2.1.3 ACTION

2.3.2 APPLICABILITY

2.3.3 ACTION

2.5.3 ACTION

2.7.2 APPLICABILITY

2.7.3 ACTION

2.9.2 APPLICABILITY

2.9.3 ACTION

3.1.2 APPLICABILITY

3.1.3 ACTION

3.3.2 APPLICABILITY

3.3.3 ACTION

3.5.3 APPLICABILITY

3.5.4 ACTION

3.7.4 APPLICABILITY

3.7.5 ACTION

3.9.3 APPLICABILITY

6.1.2 APPLICABILITY

6.1.3 ACTION

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