ML20206D589

Offsite Dose Calculation Manual

ML20206D589
Person / Time
Site:	Seabrook
Issue date:	06/30/1986
From:	PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:
Shared Package
ML20206D586	List: ML20206D581 ML20206D589
References
PROC-860630, NUDOCS 8606200092
Download: ML20206D589 (144)

Text

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DISCLAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company

(" Yankee"). The use of information contained in this document by anyone other

( than Yankee, or the Organization for which the document was prepared under contract, is not authorized and, with respect to any unauthorized use, neither Yankee nor its of ficers, directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or

{ representation as to the accuracy or completeness of the material contained in this document.

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L I ABSTRACT The Seabrook Station ODCM (Off-Site Dose Calculation Manual) is divided

' into two parts: (1) the in-plant radiological effluent monitoring program

~ requirements for liquid and gas sampling and analysis, along with the environmental radiological monitoring program requirements (Part A); and (2) approved methods to determine effluent monitor setpoint values and estimates L of doses and radionuclide concentrations occurring beyond the boundaries of the station resulting from normal station operation (Part B).

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The sampling and analysis programs in Part A provide the inputs for the p models of Part B in order to calculate off-site doses and radionuclide

' concentrations necessary to determine compliance with the dose and concentration requirements of the Station Technical Specification 3/4.11. The radiological environmental monitoring program required by Technical Specification 3/4.12 and outlined within this manual provides the means to L determine that measurable concentrations of radioactive materials released as a result of the operation of Seabrook Station are not significantly higher than expected.

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I TABLE OF CONTENTS I Page REVISION REC 0RD.................................................. 11 LIST OF EFFECTIVE PAGES.......................................... iii I DISCLAIMER OF RESPONSIBILITY..................................... iv ABSTRACT......................................................... v LIST OF FIGURES.................................................. viii LIST OF TABLES................................................... ix PART A: RADIOLOGICAL EFFLUENT MONITORING PROGRAMS Section 1.0 Introduction................................................ A .1 -1 2.0 Responsibilities for Part A................................. A . 2-1 3,0 Liquid Effluent Sampling and Analysis Program............... A . 3-1 4.0 Gaseous Effluent Sampling and Analysis Program.............. A . 4-1 5.0 Radiological Envi ronmental Monitoring. . . . . . . . . . . . . . . . . . . . . . . A . 5 -1 5.1 Sampling and Analysis Program.......................... A.5-1 5.2 Land Use Census........................................ A.5-2 PART B: RADIOLOGICAL CALCULATIONAL METHODS AND PARAMETERS........... B .1 -1 Section l

1.0 INTRODUCTION

................................................ B .1 -1 1.1 Responsibilities for Part B............................ B .1 -1 1.2 Summary of Methods, Dose Factors, Limits, i Constants, Variables and Definitions................... B.1-2 l

2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS.......... B . 2-1 ENG NG 2.1 Method to Determine F and C B.2-1 I 2.2 MethodtoDetermineRddionuclide.......................

Concentration for Each Liquid Effluent Pathway....................... B.2-2 2.2.1 Waste Test Tanks Pathway........................ B.2-2 i 2.2.2 Turbine Building Sump Pathway................... B.2-3 2.2.3 Steam Generator Blowdown Flash Tank Pathway..... B.2-3 l

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TABLE OF CONTENTS (Continued)

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3.0 0FF-SITE DOSE CALCULATION METH0DS........................... B . 3-1 I 3.1 Introductory Concepts.................................. B.3-2 3.2 Method to Calculate Total Body Dose from Liquid Releases............................................... B.3-4 3.3 Method to Calculate Maximum Organ Dose from Liquid w Releases............................................... B.3-6 3.4 Method to Calculate the Total Body Dose Rate from c Noble Gases............................................ B.3-8 L 3.5 Method to Calculate the Skin Dose Rate from Noble Gases.................................................. B . 3-10 3.6 Method to Calculate the Critical Organ Dose Rate

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from Iodines, Tritium and Particulates with Tl/2 5 Greater Than 8 Days.................................... B.3-12 3.7 Method to Calculate the Gamma Air Dose from Noble r Gases.................................................. B.3-14 3.8 Method to Calculate the Beta Air Dose from Noble Gases.................................................. B . 3-16 3.9 Method to Calculate the Critical Organ Dose from 1 Tritium, Iodines and Particulates . . . . . . . . . . . . . . . . . . . . . . B . 3-18 3.10 Method to Calculate Direct Dose f rom Plant 0peration.............................................. B.3-20 F 3 .11 Do s e P roj e cti o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B.3-21 1

4.0 ENVIRONMENTAL MONITORING PR0 GRAM............................ B . 4-1 5.0 SETPOINT DETERMINATIONS..................................... B . 5 -1 5.1 Liquid Effluent Instrumentation Setpoints.............. B.5-2 5.2 Gaseous Effluent Instrumentation Setpoints............. B.5-8 6.0 LIQUID AND GASE0US EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS.............................. B . 6-1 7.0 BASES FOR DOSE CALCULATION METH0DS.......................... B . 7 -1

.1 B.O BASES FOR LIQUID AND GASEOUS MONITOR SETPOINTS.............. B . 8-1 REFERENCES.................................................. R-1 I

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5 LIST OF FIGURES 7

Number Title Page 8.4-1 Radiological Environmental Monitoring Locations Within 4 km of Seabrook Station B.4-4 I

B.4-2 Radiological Environmental Monitoring Locations Between 4 km and 12 km from Seabrook Station B.4-5 L B.4-3 Radiological Environmental Monitoring Locations Outside 12 km of Seabrook Station B.4-6 I B.4-4 Direct Radiation Monitoring Locations Within 4 km of Seabrook Station B.4-7 Direct Radiation Monitoring Locations Between 4 km m B.4-5 and 12 km from Seabrook Station B.4-8 r B.4-6 Direct Radiation Monitoring Locations Outside 12 km L of Seabrook Station B.4-9 B.6-1 Liquid Effluent Streams, Radiation Monitors and Radwaste Treatment System at Seabrook Station B.6-9

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B.6-2 Gaseous Effluent Streams, Radiation Monitors and Radwaste g Treatment System at Seabrook Station B.6-10 B

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E LIST OF TABLES PART A Title Page Number A . 3-1 Radioactive Liquid Waste Sampling and Analysis Program A.3-2 A . 4-1 Radioactive Gaseous Waste Sampling and Analysis Program A.4-2 A.5-1 Radiological Environmental Monitoring Program A.5-3 A.5-2 Detection Capabilities for Environmental Sample Analysis A.5-7 A.5-3 Reporting Levels for Radioactivity Concentration in Environmental Samples A . 5-10 PART B Number Title Page B .1 -1 Summary of Radiological Effluent Technical Specifications and Implementing Equations B.1-3 i

B.1-2 Summary of Method I to Calculate Unrestricted Area Liquid Concentrations B .1 -6 B .1 -3 Summary of Method I to Calculate Off-Site Doses from Liquid Releases B.1-7 B .1 -4 Summary of Method I to Calculate Dose Rates B .1 -B B .1 -5 Summary of Method I to Calculate Doses to Air from I Noble Gases B.1-9 Summary of Method I to Calculate Dose to an Individual

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B.1-6 from Tritium, Iodine and Particulates B .1 -10 B .1 -7 Sunrnary of Methods for Setpoint Determinations B .1 -11 l

B .1 -B Summary of Variables B .1 -12 B.1-9 Definition of Terms B .1 -16 B .1 -10 Dose Factors Specific for Seabrook Station for Noble Gas B.1-17 Releases

B .1 -11 Dose Factors Specific for Seabrook Station for Liquid

! Releases B.1-18 I -ix-I

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- LIST OF TABLES (Continued)

Number Title Pace B.1-12 Dose and Dose Rate Factors Specific for Seabrook Station F for Tritium, Iodine and Particulate Releases B.1-19 L

B.1-13 Combined Skin Dose Factors Specific for Seabrook Station B.1-20 Special Receptors for Noble Gases

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B.1-14 Dose and Dose Rate Factors Specific for Seabrook Station Special Receptors for Iodines, Tritium and Particulates B.1-21 g

B.4-1 Radiological Environmental Monitoring Stations B.4-2 r B.7.1-1 Usage Factors for Various Liquid Pathways at L Seabrook Station B.7-4 B.7.2-1 Environmental Parameters for Gaseous Effluents at Seabrook Station B.7-22

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B.7.2-2 Usage Factors for Various Gaseous Pathways at Seabrook Station B.7-23

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B.7.3-1 Seabrook Station Dilution Factors B.7-27 l B.7.3-2 Seabrook Station Dilution Factors for Special Receptors B.7-28 l

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I SEABROOK STATION ODCM r

PART A e,oix0e1C < <,,<ut, ,0,1,1,e ,,0ee S g

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s PART A RADIOLOGICAL EFFLUENT MONITORING PROGRAMS d

1.0 INTRODUCTION

The purpose of Part A of the ODCM (Off-Site Dose Calculation Manual) is W to describe the sampling and analysis programs conducted by the station which j provides input to the models in Part B for calculating liquid and gaseous i effluent concentrations, monitor setpoints, and off-site doses. The results of Part B calculations are used to determine compliance with the concentration

{ and dose requirements of Technical Specification 3/4.11.

p The Radiological Environmental Monitoring Progrart required as a minimum to be conducted (per Technical Specification 3/4.12) is described in Part A, E with the identification of current locations of sampling stations being L

utilized to meet the program requirements listed in Part B. The information obtained f rom the conduct of the Radiological Environmental Monitoring Program l provides data on measurable levels of radiation and radioactive materials in the environment necessary to evaluate the relationship between quantities of

} radioactive materials released in effluents and resultant radiation doses to individuals from principal pathways of exposure. The data developed in the surveillance and monitoring programs described in Part A to the ODCM provide a means to confirm that measurable concentrations of radioactive materials released as a result of Seabrook Station operations are not significantly I higher than expected based on the dose models in Part B.

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! 2.0 RESPONSIBILITIES FOR PART A I All changes to Part A of the 00CM shall be reviewed and approved by the Station Operations Review Committee (SORC) and the Nuclear Regulatory I Commission prior to implementation.

It shall be the responsibility of the Station Manager to ensure that l

the ODCM is used in the performance of the surveillance requirements and administrative controls of the appropriate portions of the Technical l

Specifications.

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3.0 LIOUID EFFLUENT SAMPLING AND ANALYSIS PROGRAM N

F Radioactive liquid wastes shall be sampled and analyzed in accordance L

with the program specified in Table A.3-1 for Seabrook Unit 1. The results of f the radioactive ana. lysis shall be used as appropriate with the methodology of Part B of the 00CM to assure that the concentrations of liquid effluents at the point of release from the multiport diffuser of the circulating water i system are maintained within the limits of Technical Specification 3.11.1.1 for Unit 1.

L Radioactive effluent information for liquids obtained from this .

sampling and analysis program shall also be used in conjunction with the

{ methodologies in Part B to demonstrate compliance with the dose objectives and surveillance requirements of Technical Specifications 3/4.11.1.2, 3/4.11.1.3, and 3/4.11.4.

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- TABLE A.3-1 Notations

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(I)The LLO is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net

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count, above system b.:kground, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 s b LLD =

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E x V x 2.22 x 10 x Y x exp (-Ant) 6 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 l minute),

E = the counting efficiency (counts per disintegration),

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

l 2.22 x 10 the number of disintegrations per minute per I microcurie, Y = the fractional radiochemical yield, when applicable.

l g 'A = the radioactive decay constant for the particular radionuclide (s-l), and l

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! at = the elapsed time between the midpoint of sample collection and the time of counting (s).

Typical values of E, V, Y, and At should be used in the calculation.

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

(2)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.

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TABLE A.3-1 Notations (Continued)

(3)The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report in accordance with Technical Specification 6.8.1.4. Isotopes which are not detected should be reported as "not detected." Values determined to be below detectable level are not used in dose calculations.

(4)A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

(5)A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.

(6) Sampling and analysis is only required when Steam Generator Blowdown is directed to the discharge transition structure.

(7) Principal gamma emitters shall be analyzed weekly in Service Water.

Sample and analysis requirements for dissolved and entrained gases, tritium, gross alpha, strontium 89 and 90, and Iron 55 shall only be required when analysis for principal gamma emitters exceeds the LLD.

The following are additional sampling and analysis requirements:

a. PCCW sampled and analyzed weekly for principal gamma emitters.

b. Sample Service Water System (SWS) daily for principal gamma emitters whenever primary component cooling water (PCCW) activity exceeds lx10-3 uC/cc.

c. With the PCCW System radiation monitor inoperable, sample PCCW and SWS daily for principal gama emitters,

d. With a confirmed PCCW/SWS leak and PCCW activity in excess of lx10-4 uC/cc, sample SWS every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for principal gamma emitters. .

e. The setpoint on the PCCW head tank liquid rate-of-change alarm will be set to ensure that its sensitivity to detect a PCCW/SWS leak is equal to or greater than that of an SWS radiation monitor located in the unit's combined SWS discharge, with an LLD of 1x10-E uC/cc. If this sensitivity cannot be achieved, the SWS will be sampled once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(8)lf the Turbine Building Sump (Steam Generator Blowdown Flash Tank) isolate due to high concentration of radioactivity, that liquid stream will be sampled and analyzed for Iodine-131 and principal gamma emitters prior to release.

A.3-6

4.0 GASE0US EFFLUENT SAMPLING AND ANALYSIS PROGRAM F

Radioactive gaseous wastes shall be sampled and analyzed in accordance with the program specified in Table A.4-1 for Seabrook Unit 1. The results of W the radioactive analyses shall be used as appropriate with the methodologies of Part B of the ODCM to assure that the dose rates due to radioactive materials released in gaseous effluents from the site to areas at and beyond the site boundary are within the limits of Technical Specification 3.11.2.1 f or Unit 1.

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sampling and analysis program shall also be used in conjunction with the methodologies in Part B to demonstrate compliance with the dose objectives and I surveillance requirements of Technical Specifications 3/4.11.2.2, 3/4.11.2.3, l 3/4.11.2.4, and 3/4.11.4.

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TABLE A.4-1 Notations (I)The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net

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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 r- the counting rate of a blank sample as appropriate (counts per L minute),

E = the counting efficiency (counts per disintegration),

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

p 2.22 x 10-6 = the number of disintegrations per minute per L microcurie, Y = the f ractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (s-l),and at = the elapsed time between the midpoint of sample collection and the time of counting (s).

Typical values of E, V, Y, and at should be used in the calculation.

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

A.4-4

L TABLE A.4-1 r-Notations (Continued)

L (2)The principal gansna emitters f or which the LLO specifications applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Hn-54, Fe-59, Co-58, Co-60 2n-65, Mo-99,1-131, Cs-134, Cs-137 Ce-141 and Ce-144 in iodine and

( particulate releases. This list does not mean that only these nuclides are to be considered. Other gansna peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the

{ Semiannual Radioactive Effluent Release Report in accordance with Technical Specification 6.8.1.4. Isotopes which are not detected may be reported as p "not detected.' Values determined to be below detectable level are L not used in dose calculation _s.

p (3) Sampling and analysis shall also be performed following shutdown, i L, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERHAL POWER within a one hour period unless; 1) analysis shows that the DOSE EQUIVALLNT l-131 concentrations in the primary coolant has not increased more than a factor of 3; 2) the noble gas activity monitor for the plant

{

5 vent has not increased by more than a factor of 3. For containment purge, requirements apply only when purge is in operation.

(4) Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling canal is flooded.

(5)1he ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation I .

made in accordance with Technical Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3.

(6) Samples shall be changed at least once per seven (7) days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> af ter changing, or af ter removal f rom sampler. Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for I- at least seven (7) days following each shutdown, sta-change exceeding 15 percent of RATED THERMAL POWER wii 'r a one-hour period and analyses shall be complcted within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples

., or THERMAL POWER collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding ! t0s may be I increased by a factor of 10. This requirement does not apply if 1) analysis shows that the DOSE EQUIVALENT l-131 concentration in the reactor coolant has not increased more than a factor of 3; and (2) the noble gas I monitor shows that effluent activity has not increased more than a factor of 3.

'(7) Samples shall be taken prior to start-up of condenser air removal system I ' when there have been indications of a primary to secondary leak. .

I I A.4-5 I

5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 5.1 Samplina and Analysis Program The Radiological Environmental Monitoring Program (REMP) provides representative measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of members of the public resulting from station operation. This monitoring program is required by Technical Specification 3.12.1. The monitoring program implementsSection IV.B.2 of Appendix ! to 10CFR, Part 50, and thereby supplements the radiological effluent moaitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of effluent measurements and the modeling of the environmental exposure pathways which have been incorporated into Part B of the ODCH.

The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, ,

program changes may be initiated based on operational experience.

In accordance with Technical Specification surveillance requirements, 4.12.1, sampling and analyses shall be conducted as specified in Table A.5-1 for locations shown in Section 4 of Part B to the ODCH. Detection capability requirements, and reporting levels for radioactivity concentrations in environmental samples are shown on Tables A.5-2 and A.5-3, respective,1y.

It should be noted that Technical Specification 3.12.1.C requires that if milk or f resh leafy vegetable samples are unavailable f rom one or more sample locations required by the REMP, new specific locations for obtaining replacement samples (if available) shall be added to the REMP within 30 days, and the specific locations, from which the samples are unavailable may then be deleted from the monitoring program. In this context, the term unavailable means that samples are no longer available to be collected now or in the future for reasons such as the permission from the owner to collect the samples has been withdrawn or he has gone out of business, thus causing the permanent lose of the sample location.

A . 5-1

L

( 5.2 Land Use Census F As part of the Radiological Environmental Monitoring Program, Technical L

Specification 3/4.12.2 requires that a land use census be conducted annually c during growing season to identify within a distance of 8 km the location in L each of the 16 meteorological sectors of the nearest milk animal, the nearest 2

residence, and the nearest garden of greater than 50 m producing broad leaf

[ vegetation.

The land use census ensures that changes in the use of area beyond the

(.

site boundary are identified, and appropriate modifications to the monitoring r program and dose assessment models are made,1f necessary. This census satisfies the requirements of Section IV.3.3 of Appendix I to 10CFR Part 50.

For the purpose of conducting the land use census as required by Technical Specification 4.12.2, station personnel should determine what survey

( methods will provide the necessary results considering the type of information to be collected and the use to which it will be put, such as the location of potential milk animal pathway for use in routine dose calculations. Land use

( census results shall be obtained by using a survey method, or combination of methods, which may include, but are not limited to, door-to-door surveys

{ (i.e., roadside identification of locations), aerial surveys, or by consulting local agricultural authorities.

Technical Specification 3.12.2.b requires that new locations identified h from the census that yield a calculated dose of dose commitment 20 percent greater than at a location from which samples are currently being obtained be added within 30 days to the REMP. These new locations required to be added to

{ the sampling program shall only be those from which permission from the owner to collect samples can be obtained and sufficient sample volume is available.

{

[-

[

A.5-2

{

E

m mE m m m u r1 n rm _n El ra o n o r TABLE A.5-1

' Radiological Environmental Monitoring Program Number of Representative i

Samples and Sampling and Type and Frequency l

Exposure Pathway a of Analysis i and/or Sample Sample Locations Collection Frequency I

40 routine monitoring stations Quarterly. Gamma dose quarterly.

1. DIRECT RADIATIONb with two or more dosimeters placed as follows:

An inner ring of stations, one in each meteorological sector in the generai area of the SITE BOUNDARY; An outer ring of stations, one in each meteorological sector, generally in the 6 to 8-km range

- from the site; i

The balance of the stations to be placed in special interest areas such as population centers, nearby residences, schools, and control locations.

l

2. AIRBORNE i Samples from five locations : d Continuous sampler Radioiodine Cannister: l Radioiodine and operation with sample l Particulates I-131 analysis weekly.

Three samples from close to the collection weekly, or three SITE BOUNDARY locations, more frequently if i I

in different sectors, of high required by dust Particulate Sampler:

calculated long-term average loading.

Gross beta radioactivity ground-level D/Q.

analysis following filter One sample from the vicinity of changeC-a community having the highest Gamma isotopic analysis e l calculated long-term average of composite (by location) quarterly.

ground-level D/Q.

I

- ~ l

W M'E'M M M M M M7 R M R TV R R P TABLE A.5-1 (C:ntinu;d)

Number of Representative l Exposure Pathway Samples and Sampling and Type and Frequency l a of Analysis and/or Sample Sample locations Collection Frequency.

One sample from a control location, as for example 15-30 km distant and in the least prevalent wind direction.

3. WATERBORNE One sample.in the discharge area. Monthly grab sample. Gamma isotopic analysis e

a. Surface monthly. Composite for i One sample from a control location.

tritium analysis quarterly.

One sample from area with existing Semiannually. Gamma isotopic analysis e

> b. Sediment from semiannually.

from or potential recreational value.

L shoreline

4. INGESTION Samples from milking animals in Semimonthly when Gamma isotopic e and I-131 }

a. Milk milking animals are on analysis on each sample.

three locations within 5 km distance having the highest dose pasture, monthly at potential. If their are none, other times.

then, one sample from milking animals in each of three areas between 5 to 8 km distant where I;

doses are calculated to be greater than 1 mrem per yr.f One sample from milking animals at a control location, as for example,15-30 km distant and in the least prevalent wind )

direction.

E E TABLE A.5-1 I (Continued) l Number of Representative Exposure Pathway Samples and a Sampling and Type and Frequency f Collection Frequency of Analysis l and/or Sample Sample Locations

b. Fish and One sample of three conenercially Sample in season, or Gamma isotopic analysise Invertebrates and recreationally important semiannually if they on edible portions.

species in vicinity of plant are not seasonal.

discharge area.

One sample of similar species in areas not influenced by plant discharge.

c. Food Samples of three (if practical) Monthly, when Ganna isotopice and I-131 Products different kinds of broad leaf available. analysis.

> vegetation 9 grown nearest each of two different off-site locations

& of highest predicted long-term average ground-level D/0 if milk sampling is not performed.

One sample of each of the similar Monthly, when Gansna isotopic e and I-131 broad leaf vegetation 9 grown at available. analysis.

a control location, as for example 15-30 km distant in the least prevalent wind direction, if milk sampling is not performed.

V n _n F1 El n__fl Fl Fl El EU" TABLE A.5-1 (Crntinurd)

Table Notation l

a) Specific parameters of distance and direction sector f rom the centerline of the Unit 1 reactor, and additional description where pertinent, shall be provided for each and every sample location in Table B.4-1 in the ODCM, Part B. Deviations are permitted from the required sampling schedule if I specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability and malfunction of automatic sampling equipment. If specimens are unobtainable due to sampling ,

equipment malfunction, ef fort shall be made to complete corrective action prior to the end of the next sampling period. All deviations frem the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time.

In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. Identify the cause of the unavailability of samples for that pathway and identify the new location (s), if available, for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

Y b) A thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.

c) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

d) Optimal air sampling locations are based not only on D/Q but on factors such as population in the area, year-round access to the site, and availability of power.

e) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

f) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM, Part B.

g) If broad leaf vegetation is unavailable, other vegetation will be sampled.

~~q w rm v m_ rm_ v rw a_ rm. _fum rn_ cm r-TABLE A 5-2 i

Detection Capabilities for Environmental Sample Analysis a ,f,g

~

Lower Limit of Detection (LLD)b Fish and Water Invertebrates Milk Food Products Sediment AirbornePart1cglate Analysis (DCi/kg) or Gas (DCi/m ) (DCi/kg. wet) (DCi/kg) (DCi/ka. wet) (DCi/kg. dry)

Gross Beta 4 0.01 H-3 3,000 Mn-54 15 130 Fe-59 30 260 Co-58, 60' 15 130 260

( Zn-65 30 Zr-Nb-95 15c 0.07 1 60e I-131 15 l

15 0.05 130 15 60 150 Cs-134 l

0.06 150 18 80 180 Cs-137 18 15c,d 15c,d l Ba-La-140 l

l l

L

[ TABLE A.5-2 (Continued)

Table Notation 7 a) This list does not mean that only these nuclides are to be considered.

Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report.

b) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioact M material in a sample that will yield a net count, above s; stem bacxground, that will be detected with 95%

probability with oniy 5% probability of falsely concluding that a blank observation represents a "real" signal.

u For a particular measurement si stem, which may include radiochemical separation:

4.66 s b LLD = E

• V

• 2.22

• Y

• exp(-hat)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume; I 4.66 is a constant derived from the Kalpha and Kb eta values for the 95% confidence level; sb is the standard deviation of the background counting rate or of I the counting rate of a blank sample as appropriate, as counts per minute; E is the counting efficiency, as counts per disintegration; V is the sample size in units of mass or volume; 2.22 is the number of disintegrations per minute per picocurie; Y is the fractional radiochemical yield, when applicable; h is the radioactive decay constant for the particular radionuclide as per second; and At for environmental samples is the elapsed time between sample collection and time of counting, as seconds.

Typical values of E. V, Y, and at should be used in the calculation.

In calculating the LLD for a radionuclide determined by gamma ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g., Potassium-40 in milk samples).

A.5 B B

L TABLE A.5-2 (Continued)

' It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as c an a posteriori (after the fact) limit for a particular measurement. This does not preclude the calculation of an a_ oosteriori LLD for a particular measurement based upon the actual parameters for the sample in question and appropriate decay correction parameters such as decay while sampling

and during analysis. Analyses shall be performed in such a manner that

^ the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of E interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

l c) Parent only.

d) The Ba-140 LLD and concentration can be determined by the analysis of its

[ short-lived daughter product La-140 subsequent to an eight-day period following collection. The calculation shall be predicated on the normal ingrowth equations for a parent-daughter situation and the assumption that L any unsupported La-140 in the sample would have decayed to an insignificant amount (at least 3.6% of its original value). The ingrowth equations will assume that the supported La-140 activity at the I time of collection is zero, e) Broad leaf vegetation only, f) If the measured concentration minus the three standard deviation uncertainty is found to exceed the specified LLD, the sample does not have to be analyzed to meet the specified LLD.

g) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with reconsnendations of Regulatory Guide 4.13 Revision 1, July 1977.

1 l

1 l

1 A.5-9 o

v v m. .m- m_ v Mm .n m v _r w rm TABLE A 5-3 Reportina Levels for Radioactivity Concentrations in Environmental Samples Fish and Water Invertebrates Milk- Food Products.

Analysis (DCi/ka)

AirborneParticy) or Gas (DCi/m late (DCi/ka. wet) (DCi/ka) (oci/ka. wet)

H-3 30,000 Mn-54 1.000 30,000 l

Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 2n-65 300 20,000 y

[ Zr-Nb-95 400*

O 100 0.9 3 100**

I-1 31 10 1,000 60 1,000 Cs-134 30 20 2,000 70 .2,000 Cs-137 50 200* 300*

Ba-La-140 I

I

• Parent only.

• • Broad leaf vegetation only.

l

W I

F L

I SEABROOK STATION ODLM L

PART B r

L RADIOLOGICAL CALCULATIONAL METHODS AND PARAMETERS g

L

[

[

I

1.0 INTRODUCTION

Part B of the ODCM (Off-Site Dose Calculation Manual) provides formal and approved methods for the calculation of of f-site concentration, of f-site doses and effluent monitor setpoicts, and indicates the locations of environmental monitoring stations in order to comply with the Seabrook Station Radiological Effluent Technical Specifications (RETS) Sections 3/4.3.3.9, 3/4.3.3.10, and 3/4.11, as well as the REMP detailed in Part A of the manual.

The ODCM forms the basis for station procedures which document the off-site doses due to station operation which are used to show compliance with the I numerical guides for design objectives of Section II of Appendix I to 10CFR Part 50.

The methods contained herein follow accepted NRC guidance, unless otherwise noted in the text. The basis for each method is sufficiently documented to allow regeneration of the methods by an experienced Health Physicist.

1.1 Responsibilities for Part B All changes to Part B of the ODCM shall be reviewed and approved by the Station Operations Review Committee (SORC) in accordance with Technical Specification 6.13 prior to implementation. Changes made to Part B shall be submitted to the Commission for their information in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective.

It shall be the responsibility of the Station Manager to ensure that I the ODCM is used in the performance of the surveillance requirements and administrate controls of the appropriate portions of the Technical Specifications.

I i

I  :

B.1-1 I

I

L 1.2 Summary of Methods. D@se Factors. Limits. Constants. Variables and

[

Definitions L This section summarizes the Method I dose equations which are used as the primary means of demonstrating compliance with RETS. The concentration I

and setpoint methods are identified in Table B.1-2 through Table B.1-7. Where more refined dose calculations are needed, the use of Method II dose g

determinations are described in Sections 3.2 through 3.9 and 3.11. The dose factors used in the equations are in Tables B.1-10 through B.1-14 and the f Regulatory Limits are summarized in Table 8.1-1.

L The variables and special definitions used in this ODCM, Part 8, are in f

k Tables B.1-8 and B.1-9.

E F

I I

I 1

1 1

P l

B.1-2

.. _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .__ _ _ _ _ . l

y v m_ rm_ rm_ m _ rm m rx n_ rn. .I t_ cu- u-TABLE B.1-1 Summary of Radiological Effluent Technical Specifications and Implementine Equations l

Technical (1)

Specification Catcaory Method I Limit 3.11.1.1 Liquid Effluent Total Fraction of Eq. 2-1 51.0 I Concentration MPC Excluding Noble Gases Total Noble Gas Eq. 2-2 5 2 x 10-4 pCi/ml Concentration 3.11.1.2 Liquid Effluent Total Body Dose Eq. 3-1 5 1.5 mrem in a qtr.

Dose 5 3.0 mrem in a yr.

1 Organ Dose Eq. 3-2 5 5 mrem in a qtr.

[

b 5 10 mrem in a yr. l l

3.11.1.3 Liquid Radwaste Total Body Dose Eq. 3-1 5 0.06 mrem in a mo.

Treatment Operability Organ Dose Eq. 3-2 5 0.2 mrem in a mo. j l

Gaseous Effluents Total Body Dose Rate Eq. 3-3 $ 500 mrem /yr. 1 3.11.2.1 '

l Dose Rate from Noble Gases 1

Skin Dose Rate Eq. 3-4 5 3000 mrem /yr. l from Noble Gases Organ Dose Rate Eq. 3-5 5 1500 mrem /yr.

f rom I-131, I-133, Tritium and Particulates with T1/2 > 8 Days

TABLE B.1-1 (continued)

Sunnary of Radiological Ef fluent Technical Specifications and Implementing Ecuations Technical (1)

Category Method I Limit Specification 3.11.2.2 Gaseous Effluents Gamma Air Dose from Eq. 3-6 5 5 mrad in a qtr.

Dose from Noble Noble Gases Gases 510 mrad in a yr.

Beta Air Dose from Eq. 3-7 510 mrad in a qtr.

Noble Gases 5 20 mrad in a yr.

3.11.2.3 Gaseous Effluents Organ Dose from Eq. 3-8 5 7.5 mrem in a qtr.

to L Dose f rom I-131, Iodines, Tritium and Particulates with 515 mrem in a yr.

E I-133, Tritium, and Particulates T1/2 > 8 Days 3.11.2.4 Ventilation Organ Dose Eq. 3-8 5 0.3 mrem in a mo.

Exhaust Treatment Total Body Dose Footnote (2). 5 25 mrem in a yr.

3.11.4 Total Dose (from All Sources)

Organ Dose 5 25 mrem in a yr.

Thyroid Dose 5 75 mrem in a yr.

l

-'W Nm m m m m m m uma ums - T o r7 T - U - u L _._ J T r TABLE B.1-1 (continued) l

[

i Summary of Radiological Effluent Technical Specifications and Implementing Ecuations Technical (1)

Specification Category Method I Limit 3.3.3.9 Liquid Effluent Monitor Setpoint Liquid Waste Test Alarm Setpoint Eq. 5-1 T.S. 3.11.1.1 l l Tank Monitor l l 3.3.3.10 Gaseous Effluent l Monitor Setpoint to W

b Plant Vent Alarm / Trip Setpoint Eq. 5-9 T.S. 3.11.2.1 Wide Range Gas for Total Body Dose (Total Body) l Monitors Rate Alarm / Trip Setpoint Eq. 5-10 T.S. 3.11. 2.1 I for Skin Dose Rate (Skin) l l

(1)

More accurate methods may be available (see subsequent chapters). l (2)

Technical Specification 3.ll.4.a requires this evaluation only if twice the limit of equations 3-1, 3-2, 3-12, 3-15 or 3-18 is reached. If this occurs a Method II calculation, using actual release point parameters with annual average or concurrent meteorology and identified pathways for a real individual, shall be made.

H TABLE B.1-2 Summary of Method I Ecuations to Calculate Unrestricted Area Liouid Concentrations Equation Number Category Ecuation 2-1 Total Fraction of MPC in ENG i Fj

={MPC I

( Liquids, Except Noble Gases j j

( 2-2 Total Activity of Dissolved and Entrained Noble Gases C

NG 1

uCi ml)={3 i C

NG from all Station Sources < 2E-04

[

1 I

I I

I I

I l

I B.1-6 I

I

N TABLE B.1-3 Sumary of Method I Eauations to Calculate Off-Site Doses from Liauid Releases Equation Number Category Eauation J 3-1 T Do e U

Dtb(mrem) = k Q$ DFL itb 3-2 Maximum 0 0i DFL imo Organ Dose mo(mrem) = k $

r t

u

[

E E

B .1 -7

L F TABLE B.1-4 Summary of Method I Ecuations to Calculate Dose Rates u

Equation P Number Category Ecuation 3-3 Total Body Dose Rate gmrem) = 0.62 tb hgDFB i from Noble Gases yr 3-4 Skin Dose Rate hi DF'i from Noble Gases skin (mrem) yr = g r

L p Ra rom o n beg ("yr ) " i DFG ico L Tritium i and Particulates with T 1/2 Greater Than

{ Eight Days E

F I

1 l

1 l

B .1 -B

L TABLE B,1-5

{.

Summary of Method I Ecuations to Calculate p Doses to Air from Noble Gases L,

Equation Number Category Eauation

(

3-6 Gamma Dose to Air T T Qi DF i F from Noble Gases Dair (mrad) = 2.0E-08 i L

3-7 Beta Dose to Air O 0

{ from Noble Gases Dair (mrad) = 4.4E-08 3 Qi DF i

[

[

[

[

1

[

[

[

[

[

(

B.1-9 n,..., , . . ..

- TABLE B.1-6 Summarv of Method I Ecuations to Calculate Dose to an Individual from Tritium. Iodine and Particulates Equation

[ Number Category Ecuation 3-8 Dose to Critical Qi DFG ico r Organ from Iodines, Dco (mrem) = 3 L Tritium and Particulates I

L Y

L r

L.,

E L

[

1 I

I I B.1-10 l

I

u e

" TABLE B.1-7

~ Sumary of Methods for Setpoint Determinations k

L Equation Eauation Number._ Cateaorv 5-1 Liauid Effluents:

Liquid Waste Test OF F Tank Monitor setpoint ( " ') )*fl DF min i C ,$

L (RM-6509) 5-23 gggggaj$a-

~

RCset(gph) = lx108 , 397 , PCC

[ Gaseous Effluents:

L Plant Vent Wide Range Gas Monitors (RM-6528-1, 2, 3) p L Total Body 11 5-5 R tb ( uti3;, = 806 0FB c

cm y

1 1 5-6 Skin R F DF'c

~ skin ( cm" 3 )= 3000 L

[

1 I

I I

I B .1 -i l

L TABLE B.1-8 7

Summary of Variables r-Variable Definition Units C = Concentration at point of discharge of pCi/mi

" II dissolved and entrained noble gas "i" in liquid pathways from all station sources I

C G = Total activity of all dissolved and entrained uti noble gases in liquid pathways from all ml

[ station sources Cg = Concentration of radionuclide "i" at the point uti p of liquid discharge ml L

C = Concentration of radionuclide "i" pCi/mi I

L C = Concentration, exclusive of noble gases, of uti pg radionuclide "i" from tank "p" at point of mi r discharge L

= Concentration of radionuclide "i" in mixture pCi/ml

-C,$

at the monitor 0

D ar

- Beta dose to air mrad E

l D ir E

= Beta dose for air at Education Center mrad 6 mrad D

air R = Beta dose to air at " Rocks"

= Gamma dose to air mrad Da {r b D = Gamma dose to air at Education Center mrad air E

= Gamma dose to air at " Rocks" mrad

{ D[irR mrem D,c

= Dose to the critical organ D = Direct dose mrem d

= Gamma dose to air, corrected for finite cloud mrad Dfinite h B .1 -12

[

H TABLE B.1-8 L (continued) r- Summary of Variables Variable Definition Units D, = Dose to the maximum organ mrem S

" D = Dose to skin from beta and gamma mrem O = Dose to the total body mrem tb DF = Dilution factor ratio r

L DF min

= Minimum allowable dilution factor ratio 3

{ DF' = Composite skin dose factor mrem-m pCi-yr DFB = Total body gamma dose factor for nuclide "i" pCi-yr I

(Table B.1-10)

I 3 DFB = Composite total body dose factor C c

l I DFL = Site-specific, total body dose factor for a mrem itb liquid release of nuclide "i" (Table B.1-11) pCi DFL9 ,3 = Site-specific, maximum organ dose factor for a mrem l liquid release of nuclide "i" (Table B.1-11) pCi DFG 4g

= Site-specific, critical organ dose factor for a mrem gaseous release of nuclide "i" (Table B.1-12) pCi

= Site-specific, critical organ dose rate factor mrem-sec DFG'C for a gaseous release of nuclide "i" pCi-yr l

(Table B.1-12)

• "'*-*3 DFS = Beta skin dose factor for nuclide "i" pCi-yr (Table B.1-10) mrem-sec DF' = Combined skin dose factor for nuclide "i" pCi-yr (Table B.1-10) 3 mrad-m I DF T = Gamma air dose factor for nuclide "i" (Table B.1-10) pCi-yr B .1 -13 I

L T ABLE 8.1 -8 (continued)

' Summary of Variables Variable Definition Units 3

mrad-m _

( OF = Beta air dose factor for nuclide "i. pCi-yr D = Critical organ dose rate due to iodines yr en and particulates mrem b = Skin dose rate due to noble gases yr

{ skin h = Total body dose rate due to noble gases yr tb

= Deposition f actor for dry deposition of I D/0 elemental radioiodines and other particulates' ,2

(

= Flow rate out of discharge tunnel gpm or F

[ d ft3 j3,c F, = Flow rate past liquid waste test tank monitor gpm

= Flow rate past plant vent monitor cc F

sec f ;f if = Fraction of total MPC associated with path 1,2,3. Dimensionless j 2 3 Dimensionless

( F "O = Total fraction of MPC in liquid pathways (excluding noble gases)

= Maximum permissible concentration for uti

{ MPCg .

radionuclide "i" (10CFR20, Appendix B, cc Table 2, Column 2)

Og = Release to the environment for curies or radionuclide "i" microcuries

= Release rate to the environment for pcuries/sec

(_ hy radionuclide "i" R = qu mn r response for Oe Hm W ng

{ setpoint concentration at the point of discharge R = Response of the noble gas monitor at the cpm

[ skin limiting skin dose rate 8.1-14 m... . a, m.sn e .. ..

L T ABLE B.1-8 F

L (continued)

Summary of Variables e

L Variable Definition Units I

L

= Response of the noble gas monitor to cpm R

tb limiting total body dose rate L

S = Shielding factor Dimensionless p

F L = Detector counting ef ficier.cy f rom the com mR/hr S r 9 gas monitor calibration pCi/cc Ci/cc r mR/hr L S

= Detector counting ef ficiency for noble com r Ci/cc 9I gas "i" pCi/cc F = Detector counting ef ficiency from the ces S j L liquid monitor calibration pCi/ml

= Detector counting efficiency for ces S jj

{ radionuclide "i" pCi/ml

= Average undepleted atmospheric

{ X/0 dispersion factor m

= Effective average gansna atmospheric

( [X/Q)T dispersion factor m{'

= Service Water System flow rate gph SWF

{ = Primary component cooling water measured uCi/ml PCC (decay corrected) gross radioactivity

( concentration

[

[

[ .

1 B .1 -15

[

L TABLE B.1-9 g

Definition of Terms L

Critical Receptor - A hypothetical or real individual whose location and behavior cause him or her to receive a dose greater than any other possible real individual.

r L

Dose - As used in Regulatory Guide 1.109, the term " dose," when applied to individuals, is used instead of the more precise term " dose equivalent," as defined by the International Comission on Radiological Units and Measurements c

(ICRU). When applied to the evaluation of internal deposition or L radioactivity, the term " dose," as used here, includes the prospective dose component arising f rom retention in the body beyond the period of

( environmental exposure, i.e., the dose commitment. The dose commitment is evaluated over a period of 50 years. The dose is measured in mrem to tissue or mrad to air.

{

Dose Rate - The rate for a specific averaging time (i.e., exposure period) of dose accumulation.

Liquid Radwaste Treatment System - The components or subsystems which comprise the available treatment system as shown in Figure 8.6-1.

[

[

[

[

[

E B .1 -16

p TABLE B.1-10 L

Dose Factors Specific for Seabrook Station for h Noble GaTiteleases Gam 7,a i

[

L Total Body 8 eta Skin Combined Skin Beta Air Gamma Air Dose Factor Dose Factor Dose Factor Dose Factor Dose Factor 3 3 3 3 I

L R dio.wclide DFB $ (mrem-m DCi-vr ) DFS $(mrem-m)DFj(mrem-sec) oci-vr uCi-vr DF8 (mrad-m ) DF}

$ DCi-vr (mrad-m )

DCi-vr

[ Ar-41 8.84E-03* 2.69E-03 1.01E-02 3.28E-03 9.30E-03 Kr-83m. 7.56E-08 1.33E-05 2.88E-04 1.93E-05 F _

L Kr-85m 1.17E-03 1.46E-03 2.86E 03 1.97E-03 1.23CiO3 Kr-85 1.61E-05 1. 34 E-O'3 1.86E 1.95E-03 1.72E-05

~ .

Kr-87 5.92E-03 9.73E-03 ,'~ 1 /17 E-02 ;1.03E-02 6.17E-03 Kr-88 1.47E-02 2.37E-03.~ t.38E-02 2.33E-03 :1.52E-02

{

Kr-89 1.66E-02 1.01E-02 2.59E-02 1.06E-Oi , 1.73E-02 Kr-TJ l.56E 7.29E-03 2.13E-02 '7.83E-03 1.63E-02 Xe-131m 9.15E-05 4.76E-04 7.65E-04 1.11E-03 1.56E-04

. Xe-133m 2.51E-04 9.94E-04 1.60E 1.48E-03 3.27E-04 1.05E-03 3.53E-04

{ Xe-133 2.94E-04 3.06E 6.66E-04 Xe-135m 3.12E-03 7.11E-04 3.30E-03 7.39E-04 3.36E-03 Xe-135 1.81E-03 1.86E-03 3.89E-03 2.46E-03 1.92E-03:

Xe-137 1.42E-03 1.22E 1.79E-02 1.27E-02 1.51E-03 I

Xe-138 8.83E-03 4.13E-03 1.21E-02 4.75E-03 9.21E-03 l

0 8.84E-03 = 8.84 x 10-3 l 4 1

B .1 -17 1

l

I TABLE B.1-ll Dose Factors Specific for Seabrook Station for Liouid Releases Total Body Maximum Organ I Dose Factor DFL mrem itb I ci )

Dose Factor DFL$ ,, (mrem) uci Radionuclide u H-3 3.02E-13 3.02E-13 Cr-51 1.83E-11 1.48E-09 Mn-54 5.14E-09 2.68E-08 Fe-55 1.26E-08 7.67E-08 Fe-59 8.74E-08 6.66E-07 -

Co-58 2.45E-09 1.40E-08 Co-60 6.14E-09 9.21E-08 Zn-65 2.49E-07 5.49E-07 Br-83 1.31 E-14 1.89E-14 Rb-86 4.18E-10 6.96E-10 Sr-89 2.17 E-10 7.59E-03 Sr-90 3.22E-08 1.31E-07

. Mo-99 3.10E-l l 2.62E-10 Tc-99m 4.95E-11 7.16E-11 Te-127m 7.07E-08 1.81E-06 Te-127 3.50E-10 9.46E-08 Te-129m 1.54E-07 3.46E-06 Te-129 6.97 E-14 8.22E-14 Te-131m 3.16E-08 2.94E-06 Te-132 9.05E-08 3.80E-06 I-130 2.77E-11 3.20E-09 I-131 2.21 E-10 1.00E-07 I I-132 I-133 I-134 3.30E-12 2.55E-11 1.18E-12 4.03E-12 1.15E-08 1.40E-12 1-135 8.84E-12 4.39E-10 Cs-134 3.24E-08 3.56E-08 Cs-136 2.46E-09 3.27E-09 Cs-137 3.58E-08 4.03E-08 Ba-140 1.64 E-10 3.48E-09 La-140 5.13E-11 4.13E-08 Ce-141 3.67E-11 9.31E-09 6.46E-08 I Ce-144 Np-239 1.95E-10

4. 55E-12 5.71E-10 I B .1 -18

TABLE 8.1-12 Dose'and Dose Rate Factors Specific for Seabrook Station for (n -

Iodines.' Tritium and Particulate Releases Critical Organ

{ Critical Organ Dose Factor Dose Rate Factor DFG 4c,(mrem-sec) ico Imrem)

DFG P- Radionuclide uCi vr uCi H-3 4.47E-10 1.41E-02 Mn-54 1.60E-07 5.05E+00

'Fe-59 1.99E-07 6.28E+00 Co-58 8.41E-08 2.65E+00 Co-60 2.48E-06 7.82E+01 In-65 1.19E-06 3.75E+01 h Sr-89 4.33E-06 1.36E+02 Sr-90 1.57E-04 4.95E+03 Mo-99 1.61E-08 5.08E-01 1-130 1.08E-07 3.41E+00 1-131 5.24E-05 1.65E+03 1-132 7.67E-09 2.42E-01 1-133 6.25E-07 1.97E+01 1-134 2.01E-09 6.34E-02 1-135 3.24E-08 1.02E+00

[ Cs-134 Cs-137 1.31E-05 1.24E-05 4.13E+02 3.91E+02 Ce-141 5.83E-08 1.84E+00 Ce-144 1.28E-06 4.04E+01

{

[

[

[

[

[-

8.1-19

[

[

'I' ' ' _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . . - . . . _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _

r TABLE B.1-13 l

Combined Skin Dose Factors Specific for Seabrook Station

~

SDecial Receptorstl) for i

Noble Gas Release Education Center The " Rocks" E Combined Skin Combined Skin Dose Factor Dose Factor I #'*-

k Radionuclide DF'E("uCi-vr i ) DF'R(mrem-sec) i uCi-vr Ar-41 3.85E-02 1.21E-01 E Kr-83m 4.25E-05 1.2SE-04

'~ Kr-85m 1.25E-02 4.08E-02 Kr-85 9.03E-03 3.03E-02 Kr-87 7.89E-02 2.59E-01

[ Kr-88 Kr-89 4.93E-02 1.06E-01 1.52E-01 3.40E-01 Kr .10 8.48E-02 2.70E-01 I Xe-131m 3.54E-03 1.17E-02 L Xe-133m 7.39E-03 2.45E-02 Xe-133 2.83E-03 9.18E-03 Xe-135m 1.22E-02 3.78E-02 Xe-135 1.67E-02 5.43E-02 Xe-137 8.52E-02 2.84E-01 Xe-138 4.80E-02 1.53E-01 I

(I) See Seabrook Station Unit 1 Technical Specification Figure 5.1-1.

l l

l 1

I B.1-20 I

F-

' TA8LE 8.1-14

/ Dose and Dose Rate Factors SDecific for Seabrook Station SDecial ReceDtorstl) for Iodine.

Tritium and Particulate Releases L The " Rocks" Education Center Critical Organ critical Organ critical Organ critical Organ I Dose Factor Dose Rate Factor Dose Factor Dose Rate Factor k mrem icoEI uCi ) DFGjcoEImrem-sec) uCi-vr G$ (mrem) DN qc,p(mrem-sec) copvCi vCi-vr Radionuclide H-3 2.71 E-10 8.55E-03 9.07 E-10 2.86E-02 Mn-54 5.78E-07 1.82E+01 2.58E-06 8.14E+01 9.93E-07 3.13E+01 r Fe-59 2.98E-07 9.40E+00 8.26E+00 8.73E-07 2.75E+01 L. Co-58 2.62E-07 Co-60 8.99E-06 2.84E+02 4.01E-05 1.26E+03

-Zn-65 3.06E-07 9.65E+00 1.36E-06 4.29E+01 Sr-89 4.72E-07 1.49E+01 1.57E-06 4.95E+01 h.. Sr-90 2.11 E-05 6.65E+02 7.02E-05 2.21E+03 Mo-99 5.25E-08 1.66E+00 1.75E-07 5.52E+00 F

1-130 3. 61 E-07 1.14E+01 1.20E-06 3.78E+01 I-1 31 3.17E-06 1.00E+02 1.06E-05 3.34E+02 h 3.97E+00

/ I-132 3.78E-08 1.19E+00 1.26E-07 1-133 7.51E-07 2.37E+01 2.50E-06 7.89E+01 1-134 9.90E-09 3.12E-01 3.29E-08 1.04E+00 1-135 1.55E-07 4.89E+00 5.15E-07 1.62E+01 Cs-134 2.83E-06 8.93E+01 1.26E-05 3.97E+02 Cs-137 4.27E-06 1.35E+02 1.90E-05 5.99E+02 Ce-141 1.20E-07 3.78E+00 3.99E-07 1.26E+01 Ce-144 2.61E-06 8.23E+01 8.68E-06 2.74E+02 E

E c

E (I) See Seabrook Station Unit 1 Technical Specification Figure 5.1-1.

B .1 -21

[

u 2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS

[

Chapter 2 contains the basis for station procedures that the station operator requires to meet Technical Specification 3.11.1.1 which limits the r

total fraction of MPC in liquid pathways, other than noble gases, denoted here NG L as F , at the point of discharge from the station to the environment (see Figure B.6-1). F "O is limited to less than or equal to one, i.e.,

L r

F"1

< 1.

L The total concentration of all dissolved and entrained noble gases at k the point of discharge from the multiport diffuser from all station sources combined,denotedCf,islimitedto2E-04pCi/ml,i.e.,

Cf<_2E-04pCi/ml. '

r O

2.1 Method to Determine F " and C"1 1 t

L g

First, determine the total fraction of MPC (excluding noble gases), at the point of discharge from the station from all significant liquid sources denoted F "U; and then separately determine the total concentration at F the point of discharge of all dissolved and entrained noble gases from all stationsources,denotedCf,asfollows:

1 F NG,{MPCp.i < ), {p_))

1 IuCi/ml) pCi/ml y

L and:

B.2-1

7

< 2E-04 (2-2)

{ Cf = C (pCi/ml) (pCi/ml) (pCi/ml)

L where:

F F = Total fraction of MPC in liquids, excluding noble g

gases, at the point of discharge from the multiport diffuser i Concentration at point of discharge from the multiport Cp ,j =

dif fuser of radionuclide "i", except for dissolved and F

entrained noble gases, from all tanks and other significant L sources, p, f rom which a discharge may be made (including the waste test tanks and any other significant source from which r a discharge can be made) (pC1/ml) k = Maximum permissible concentration of radionuclide "i" except MPCj for dissa:ved and entrained noble gases f rom 10CFR20, Appendix B, Table II, Column 2 (pCi/ml)

= Total concentration at point of discharge of all dissolved Cf and entrained noble gases in liquids from all station f sources (pci/ml)

L C = Concentration at point of discharge of dissolved and entrained I noble gas "i" in liquids from all station sources (pCi/ml)

>' 2.2 Method to Determine Radionuclide Concentration for Each Liauid Effluent Source l

2.2.1 Waste Test Tanks l

C p,j is determined for each radionuclide detected from the activity in a representative grab sample of any of the waste test tanks and the predicted flow at the point of discharge.

The batch releases are normally made from two 25,000-gallon capacity waste test tanks. These tanks normally hold liquid waste evaporator distillate. The waste test tanks can also contain other waste such as liquid taken directly from the floor drain tanks when that liquid does not require processing in the evaporator, distillate f rom the boron recovery evaporator B B.2-2 I

r I

l l

)

when the BRS evaporator is substituting for the waste evaporator, and distillate f rom the Steam Generator Blowdown System evaporators and flash steam condensers when that system must discharge liquid off-site.

If testing indicates that purification of the waste test tank contents is required prior to release, the liquid can be circulated through the waste demineralizer and filter.

The contents of the waste test tank may be reused in the Nuclear System if the sample test meets the purity requirements.

Prior to discharge, each waste test tank is analyzed for principal gamma emitters in accordance with the liquid sample and analysis program outlined in Part A to the ODCM.

2.2.2 Turbine Buildino Sump The Turbine Building sump collects leakage from the Turbine Building floor drains and discharges the liquid unprocessed to the circulating water system.

Sampling of this potential source is normally done once per week for determining the radioactivity released to the environment (see Table A.3-1).

2.2.3 Steam Generator Blowdown Flash Tank The steam generator blowdown evaporators normally process the liquid f rom the steam generator blowdown flash tank when there is primary to secondary leakage. Distillate from the evaporators can be sent to the waste test tanks or recycled to the condensate system. When there is no primary to secondary leakage, flash tank liquid is processed through the steam generator blowdown demineralizers and returned to the secondary side.

Steam generator blowdown is only subject to sampling and analysis when all or part of the blowdown liquid is being discharged to the environment instead of the normal recycling process (see Table A.3-1).

B.2-3

L I 3.0 0FF-SITE DOSE CALCULATION METHODS

- Chapter 3 provides the basis for station procedures required to meet the Radiological Effluent Technical Specifications (RETS) dose or dose rate requirements contained in Section 3/4.11 of the station operating Technical A Specifications. A simple, conservative method (called Method I) is listed in Tables B.1-2 to B.1-7 for each of the requirements of the RETS. Each of the

[ Method I equations is presented in Sections 3.2 through 3.9. In addition, those sections include more sophisticated methods (called Method II) for use when more refined results are needed. This chapter provides the methods,

[

data, and reference material with which the operator can calculate the needed

~ doses, dose rates and setpoints. The bases for the dose and dose rate k equations are given in Chapter 7.0.

f The Semiannual Radioactive Effluent Release Report, to be filed after y

January 1 each year per Technical Spacification 6.8.1.4, requires that

[ metecrological conditions concurrer,with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and F measurement, be used for determir,ing the gaseous pathway doses. For L

contiruous release sources (i.e., plant vent, condenser air removal exhaust, and gland steam packing exhauster), concurrent quarterly average meteorology will be used in the dose calculations along with the quarterly total radioactivity released. For batch releases or identifiable operational I activities (i.e., containment purge or venting to atmosphere of the Waste Gas L

System), concurrent meteorology during the period of release will be used to determine dose if the total noble gas or iodine and particulates released in

[

the batch exceeds five percent of the total quarterly radioactivity released f rom each unit; otherwise quarterly average meteorology will be applied.

{

Quarterly average meteorology will also be applied to batch releases if the I

1 hourly met data for the period of batch release is unavailable.

i B.3-1

... _ . _ _ _ J

u F

3.1 Introductory Concepts F

In part, the Radiological Effluent Technica~1 Specifications (RETS) limit dose or dose rate. The term " dose" for ingested or inhaled

( radioactivity means the dose commitment, measured;in mrem, which results from the exposure to radioactive materials that, because of uptake and deposition

[ in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is ' stopped. The time frame over which F- the. dose comitment is evaluated is 50 years. The: phrases " annual dose" or dose in one year" then refers to the 50-year dosehomitment resulting from y exposure to one year's worth of releases. " Dose in a quarter" similarly means N the 50-year dose comitment resulting from exposure to one quarter's

, releases. The term " dose," with respect to external exposures, such as to noble gas clouds, refers only to the doses received during the actual time period of exposure to the radioactivity released from the plant. Once the source of the radioactivity is removed, there is no longer any additional

{ accumulation to the dose comitment.  ;

I

" Dose rate" is the total dose or dose comitment divided by exposure period. For example, an individual who is exposed via the ingestion of milk l.

for one year to radioactivity from plant gaseous effluents and receives a 50-year dose comitment of 10 mrem is said to have been exposed to a dose rate l of 10 mrem / year, even though the actual dose received in the year of exposure may be less than 10 mrem.

l In addition to limits on dose.comitment, gaseous ef fluents f rom the station are also controlled so that the maximum or peak dose rates at the site l

boundary at any time are limited to the equivalent annual dose limits of I

l 10CFR, Part 20 to unrestricted areas (if it were assumed that the peak dose rates continued for one year). These dose rate limits provide reasonable I

assurance that memuers of the public, either inside or outside the site boundary, will not be exposed to annual averaged concentrations exceeding the

--limits specified in Appendix B, Table II of 10CFR; Part 20 (10CFR20.106(a)).

t i

i B.3-2 i

I L

I L ThequantitiesADandbareintroducedtoprovidecalculable quantities, related to of f-site doses or dose rates that demonstrate F

L compliance with the RETS.

Delta D, denoted AD, is the quantity calculated by the Chapter 3, Method I dose equations. It represents the conservative increment in dose.

g The AD calculated by Method I equations is not necessarily the actual dose received by a real individual, but usually provides an upper bound for a given g release because of the conservative margin built into the dose factors and the selection and definition of critical receptors. The radionuclide specific h

dose factors in each Method I dose equation represent the greatest dose to any

( organ of any age group. (Organ dose is a function of age because organ mass and intake are functions of age.) The critical receptor assumed by ' Method I" I equations is then generally a hypothetical individual whose behavior - in h terms of location and intake - results in a dose which is higher than any real individual is likely to receive. Method II allows for a more exact dose calculation for each individual if necessary.

Ddot,denotedb,isthequantitycalculatedintheChapter3doserate equations. It is calculated using the station's effluent monitoring system reading and an annual or long-term average atmospheric dispersion factor. b predicts the maximum off-site annual dose if the peak observed radioactivity release rate f rom the plant stack continued for one entire year. Since peak

[

release rates, or resulting dose rates, are usually of short time duration on the order of an hour or less, this approach then provides assurance that 10CFR20.106 limits will be met.

Each of the methods to calculate dose or dose rate are presented in separate subsections of Chapter 3, and are summarized in Tables B.1-1 to B.1-7. Each method has two levels of complexity and conservative margin called Method I and Method II. Method I has the greatest margin and is the simplest; generally a linear equation. Method II is a more detailed analysis which allows use of site-specific factors and variable parameters to be selected to best fit the actual release. Guidance is provided, but the appropriate margin and depth of analysis are determined in each instance at the time of analysis under Method II.

s B.3-3 I

L k 3.2 Method to Calculate the Total B7dv Dose from Liouid Releases L

~ Technical Specification 3.11.1.2 limits the total body dose commitment to a member of the public f rom radioactive material in liquid ef fluents to 1.5 y mrem per quarter and 3 mrem per year per unit. Technical Specification 3.11.1.3 requires liquid radwaste treatment when the total body dose estimate exceeds 0.06 mrem in any 31-day period. Technical Specification 3.11.4 limits s the total body dose commitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year.

/

b use Method I first to calculate the maximum total body dose from a liquid release from the station as it is simpler to execute and more conservative than Method II.

Use Method II if a more refined calculation of total body dose is

(

needed, i.e., Method I indicates the dose might be greater than the Technical Specification limits.

{

r To evaluate the total body dose, use Equation 3.1 to estimate the dose from the planned release and add this to the total body dose accumulated from prior releases during the month. See Section 7.1.1 for basis.

l 3.2.1 Method I l

The increment in total body dose from a liquid release is:

D tb

=k Q$ DFl itb (3-1) l "

(mrem) ( ) (pCi) ("C )

l where:

DFlitb = Site-specific total body dose factor (mrem /pCi) for a liquid release. It is the highest of the four age groups.

See Table B.1-11.

Qi = Total activity (pCi) released for radionuclide "i". (For strontiums, use the most recent measurement available.)

B.3-4 i

.. J

K o 918/Fd ; where Fd is the average (typically monthly average) dilution flow of the Circulating Water System at I the point of discharge from the multiport diffuser (in ft 3/sec). For normal operations with a cooling water flow of 918 ft 3 /sec, K is equal to 1.

I Equation 3-1 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1 1. Liquid releases via the multiport diffuser to unrestricted areas (at the edge of the initial mixing or prompt dilution zone that corresponds to a factor of 10 dilution), and

2. Any continuous or batch release over any time period.

3.2.2 Method II If Method I cannot be applied, or if the Method I dose calculations I appear to exceed a Technical Specification limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-speci'ic models, data or assumptions are more applicable. The general equations and parameters taken f rom Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, can also be applied to a Method II assessment, which, in addition, incorporates site receptor-specific information which coincides conditions at the time of release, such as identified pathways of I exposure and dilution values associated with the receptor.

,I

!I iI I

lg

! 8.3-5 I

L 1

3.3 Method to Calculate Maximu, Organ Dose f rom Liauid Releascs c Technical Specification 3.11.1.2 limits the maximum organ dose commitment to a Member of the Public f rom radioactive material in liquid

~ effluents to 5 mrem per quarter and 10 mrem per year per unit. Technical L. Specification 3.11.1.3 requires liquid radwaste treatment when the maximum organ dose projected exceeds 0.2 mrem in any 31 days (see Subsection 3.11 for r

( dose projections). Technical Specification 3.11.4 limits the maximum organ dose consnitment to any real member of the public from all station sources I (including liquids) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year.

Use Method I first to calculate the maximum organ dose from a liquid release to unrestricted areas (see Figure B.6-1) as it is simpler to execute L and more conservative than Method II.

I Use Method II if a more refined calculation of organ dose is needed, i.e., Method I indicates the dose may be greater than the limit.

l Use Equation 3-2 to estimate the maximum organ dose from individual or combined liquid releases. See Section 7.1.2 for basis.

3.3.1 Method I l

The increment in maximum organ dose from a liquid release is:

l D,g =k Qg DFL ho (3-2) l (mrem) ( ) (pCi) ("C I where:

DFLimo

= Site-specific maximum organ dose factor (mrem /pCi) for a liquid release. It is the highest of the four age groups.

See Table B.1-11.

Qj = Total activity (pCi) released for radionuclide "i". (For strontiums, use the most recent measurement available.)

B.3-6 I

I K = 918/Fd ; where Fd is the averago (typically monthly average) dilution flow of the Circulating Water System at I the point of discharge frem the multiport diffuser (in ft 3/sec). 3 For normal operations with a cooling water flow of 918 ft /sec, K is equal to 1.

I Equation 3-2 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Liquid releases via the multiport diffuser to unrestricted areas 1 (at the edge of the initial mixing or prompt dilution zone that corresponds to a factor of 10 dilution), and

2. Any continuous or batch release over any time period.

E 3.3.2 Method II I If Method I cannot be applied, or if the Method I dose calculations appear to exceed a Technical Specification limit, or if a more exact I calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. The general equations and parameters taken f rom Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, can also be applied to a Method II assessment, which, in addition, incorporates site receptor-specific information which coincides conditions at the time of release, such as identified pathways of exposure and dilution values associated with the receptor.

I I

I I

B.3-7 8

7 L

C 3.4 Method to Calculate the Total Body Dose Rate From Noble Gases Technical Specification 3.11.2.1 limits the dose rate at any time to y the total body from noble gases at any location at or beyond the site boundary L to 500 mrem / year. The Technical Specification indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate. Bylimitingh to a rate equivalent to no more than 500

( tb mrem / year, we assure that the total body dose accrued in any one year by any

( member of the general public is less than 500 mrem.

Use Method I first to calculate the Total Body Dose Rate from the peak H release rate via the station vents . Method I applies at all release rates.

Use Method II if a more refined calculation of Dtb is desired by the station (i.e., use of actual release point parameters with annual or actual meteorology to obtain release-specific X/Qs) or if Method I predicts a dose rate greater than the Technical Specification limit to determine if it had actually been exceeded during a short time interval. See Section 7.2.1 for basis.

{

Compliance with the dose rate limits for noble gases are continuously demonstrated when effluent release rates are below the plant vent noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site dose rate limit, or a value below it. Determinations of dose rate for compliance with Technical Specifications are performed when the effluent monitor alarm setpoint is exceeded.

(I) The Turbine Building vent ground level release X/Qs are used in the

( ODCM Method I equations. This is to conservatively account for the station vent stack, and, any potential ground level releases.

B.3-8 l

, . , l

L I 3.4.1 Method I The Total Body Dose Rate due to noble gases can be determined as L follows:

b = 0.62 hg DFB g (3-3) tb 3

(mrem) yr DCi-sec) 3 uC1) sec mrem-mpCi-yr )

Cim where:

h4 = The release rate at the station vents (uCi/sec), for each L noble gas radionuclide, "i", shown in Table B.1-10.

DFB g = Total body gamma dose factor (see Table B.1-10).

Equation 3-3 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

I 1. Normal operations (nonemergency event), and

2. Noble gas releases via any station vent to the atmosphere.

l 3.4.2 Method II l

If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method 11 may be l

applied. Method 11 consists of the models, input data and assumptions in 5 Regulatory Guide 1.109, Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. The general equations and parameters taken f rom Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, can also be applied to a Method 11 assessment, which, in addition, incorporates site receptor-specific information which coincides conditions at the time of release, such as identified pathways of exposure.

B.3-9 l

L F

3.5 Method to Calculate the Skin Oose Rate from Noble Gases U Technical Specification 3.11.2.1 limits the dose rate at any time to y the skin from noble gases at any location at or beyond the site boundary to I 3,000 mrem / year. The Technical Specification indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate. By limiting D to a rate equivalent to no more than skin 3,000 mrem / year, we assure that the skin dose accrued in any one year by any member of the general public is less than 3,000 mrem. Since it can be expected that the peak release rate on which D skin is derived would not be exceeded without corrective action being taken to lower it, the resultant g

L average release rate over the year is expected to be considerably less than the peak release rate.

Use Method I first to calculate the Skin Dose Rate from the peak release rate via the station vents . Method I applies at all release rates.

l Use Method II if a more refined calculation of D skin is desired by the I station (i.e., use of actual release point parameters with annual or actual meteorology to obtain release-specific X/Qs) or if Method I predicts a dose g rate greater than the Technical Specification limit to determine if it had l actually been exceeded during a short time interval. See Section 7.2.2 for basis.

l Compliance with the dose rate limits for noble gases are continuously demonstrated when effluent release rates are below the plant vent noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site dose rate limit, or a value below it. Determinations of dose rate for compliance with Technical Specifications are performed when the effluent monitor alarm setpoint is exceeded.

(I) The Turbine Building vent ground level release X/Qs are used in the 00CM Method I equations. This is to conservatively account for the station vent stack, and, any potential ground level releases.

B . 3-10 1

E 3.5.1 Method I The Skin Dose Rate due to noble gases is:

I DFj (3-4) skin " i I mrem yr y

uti) sec mrem-sec) pCi-yr I where:

h3 = The release rate at the station vents (pCi/sec) for each radionuclide, "i", shown in Table B.1-10.

DFj

= combined skin dose f actor (see Table B.1-10).

Equation 3-4 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

I 1. Normal operations (nonemergency event), and

2. Noble gas releases via any station vent to the atmosphere.

3.5.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method 11 may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. The general equations and parameters taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, can also be applied to a Method II assessment, which, in addition, incorporates site I receptor-specific information which coincides conditions at the time of release, such as identified pathways of exposure.

I I B . 3-11 8

r>

3.6 Method to Calculate the Critical Organ Dose Rate from Iodines. Tritium and Particulates with T1/2 Greater Than 8 Days b

j Technical Specification 3.11.2.1 limits the dose rate at any time to 31 any organ from 7,1337, 3H and radionuclides in particulate form with half lives greater than 8 days to 1500 mrem / year to any organ. The Technical L Specification indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate. By limiting 0 g to a rate equivalent to no more than 1500 mrem / year, we assure that the critical organ dose accrued in any one year by any member of the general public is less than 1500 mrem.

Use Method I first to calculate the Critical Organ Dose Rate from the l peak release rate via the station vents ( }. Method I applies at all release rates.

l .

Use Method II if a more refined calculation of O co is desired by the station (i.e., use of actual release point parameters with annual or actual l

meteorology to obtain release-specific X/Qs) or if Method I predicts a dose rate greater than the Technical Specification limit to determine if it had 1

actually been exceeded during a short time interval. See Section 7.2.3 for basis.

l 3.6.1 Method I l

The Critical Organ Dose Rate can be determined as follows:

D, = Qg 0FGjeg (3-5) c j mrem) yr uti) sec mrem-sec) pCi-yr (I) The Turbine Building vent ground level release X/Qs are used in the ODCM Method I equations. This is to conservatively account for the l station vent stack, and, any potential ground level releases.

B. 3 -12 l

7 L

/

where:

DFGy,= Site-specificcriticalorgandoseratefactor mrem-sec) pCi-yr h for a gaseous release. See Table B.1-12.

f Qi = The activity release rate at the station vents of L radionuclide "i" in pCi/sec (i.e., total activity measured of radionuclide "i" averaged over the time period for which r

k the filter / charcoal sample collector was in the effluent j stream). For i = Sr89 or Sr90, use the best estimates

( (such as most recent measurements).

Equation 3-5 can be applied under the following conditions (otherwise, f

' justify Method I or consider Method II):

1. Normal operations (not emergency event), and t 2. Tritium,1-131 and particulate releases via monitored station vents

( to the atmosphere.

3.6.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. The general equations and parameters taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, can also be applied to a Method 11 assessment, which, in addition, incorporates site

( receptor-specific information which coincides conditions at the time of release, such as identified pathways of exposure.

C E

B .3-13

3.7 Method to Calculate the Gamma Air Oose from Noble Gases k Technical Specification 3.11.2.2 limits the gamma dose to air from j noble gases at any location at or beyond the site boundary to 5 mrad in any L quarter and 10 mrad in any year per unit. Dose evaluation is required at least once per 31 days.

L Use Method I first to calculate the gamma air dose for the station

/ vent III releases during the period.

Use Method II if a more refined calculation is needed (i.e., use of b actual release point parameter with annual or actual meteorology to obtain release-specific X/Qs), or if Method I predicts a dose greater than the r' Technical Specification limit to determine if it had actually been exceeded.

See Section 7.2.4 for basis, b

3.7.1 Method I The gamma air dose from station vent releases is:

0,}r=2.0E-08 09 (3-6) 0F}

3 l (mrad) (DCi-V") (pCi) (mrad-m )

3 B pCi-m pCi-yr Qg = total activity (pCi) released to the atmosphere via station vents of each radionuclide "i" during the period of interest.

DF} = gamma dose f actor to air for radionuclide "1". See Table B.1-10 1

(I) The Turbine Building vent ground level release X/Qs are used in the ODCM Method I equations. This is to conservatively account for the I station vent stack, and, any potential ground level releases.

B.3-14 1

r....._...

L F

Equation 3-6 can be applied under the following conditions (otherwise justify Method I or consider Method II):

1. Normal operations (nonemergency event), and

( 2. Noble gas releases via station vents to the atmosphere.

[ 3.7.2 Method II H If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. The general equations and parameters taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, can also be applied to a Method II assessment, which, in addition, incorporates site

{ receptor-specific information which coincides conditions at the time of release, such as identified pathways of exposure.

{

I I

1 1

1 1

l B. 3-15 i

L 3.8 Method to Calculate the Beta Air Oose from Noble Gases Technical Specification 3.11.2.2 limits the beta dose to air from noble j gases at any location at or beyond the site boundary to 10 mrad in any quarter L and 20 mrad in any year per unit. Dose evaluation is required at least once per 31 days.

F L

Use Method I first to calculate the beta air dose for the station vent (II stack releases during the period. Method I applies at all dose levels.

s Use Method II if a more refined calculation is needed (i.e., use of actual release point parameters with annual or actual meteorology to obtain release-specific X/Qs) or if Method I predicts a dose greater than the i Technical Specification limit to determine if it had actually been exceeded.

( See Section 7.2.5 for basis.

3.8.1 Method I The beta air dose from station vent releases is:

0 0 = 4.4E-08 Qj OF (3-7) ir i

(mrad) (DCi-vr) 3 ( C1) (mrad-pC _ r) pCi-m where:

0 DF = beta dose factor to air for radionuclide "i". See Table B.1-10

( Qg = total activity (pC1) released to the atmosphere via station vents of each radionuclide "i" during the period of interest.

(I) The Turbine Building vent ground level release X/Qs are used in the ODCM Method I equations. This is to conservatively account for the station vent stack, and, any potential ground level releases.

B.3-16

Equation 3-7 can be applied under the following conditions (otherwise justify Method I or consider Method II):

1. Normal operations (nonemergency event), and

( 2. Noble gas releases via station vents to the atmosphere.

3.8.2 Method II

{

If Method I cannot be applied, or if the Method I dose exceeds the f

limit or if a more exact calculation is required, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The general equations and parameters taken from Regulatory Guide 1.109,and used in the derivation of the simplified Method I approach as described in the Bases section, can also be applied to a Method II assessment, which, in addition, incorporates site receptor-specific information which coincides conditions at the time of release, such as identified pathways of exposure.

I I

l l

l l

l l

l '

B.3-17

3.9 Method to Calculate the Critical Organ Dose from Iodines. Tritium and Particulates h

y Technical Specification 3.11.2.3 limits the critical organ dose to a L member of the public from radioactive iodines, tritium, and particulates with half-lives greater than 8 days in gaseous effluents to 7.5 mrem per quarter and 15 mrem per year per unit. Technical Specification 3.11.4 limits the

( total body and organ dose to any real member of the public from all station sources (including gaseous effluents) to 25 mrem in a year except for the h thyroid, which is limited to 75 mrem in a year.

P k Use Method I first to calculate the critical organ dose from a vent release as it is simpler to execute and more conservative than Method II.

Use Method II if a more refined calculation of critical organ dose is needed (i.e., Method I indicates the dose is greater than the limit). See Section 7.2.6 for basis.

I 3.9.1 Method I O, = Q j DFG gg (3-8) c l (mrem) (pCi) (*C I 0 = Total activity (pC1) released to the atmosphere of radionuclide 9

"i" during the period of interest. For strontiums, use the most I

l recent measurement.

DFG g ,= Site-specific critical organ dose factor (mrem /pC1). For each I radionuclide it is the age group and organ with the largest dose factor. See Table B.1-12.

Equation 3-8 can be applied under the following conditions (otherwise, justify Method I or consider Method II):

1. Normal operations (nonemergency event),

B.3-18

- -_-- _______________________________________________.__u

h Iodine, tritium, and particulate releases via station vents to the f 2.

atmosphere, and

3. Any continuous or batch release over any time period.

3.9.2 METHOD II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in v Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The general equations and parameters taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases section, can also be applied to a Method II assessment, which, in addition, incorporates site receptor-specific information which coincides conditions at the time of release, such as identified pathways of exposure to a real individual.

b B

l l

1 l

• k l

B.3-19

I i 3.10 Method to Calculate Direct Dose f rom Plant ODeratien L

Technical Specification 3.11.4 restricts the dose to the whole body or any organ to any member of the public from all uranium fuel cycle sources (including direct radiation from station facilities) to 25 mrem in a calendar year (except the thyroid, which is limited to 75 mrem). It should be nott-d that since there are no uranium fuel cycle facilities within 5 miles of the station, only station sources need be considered for determining compliance

(

with Technical Specification 3.11.4.

3.10.1 Method The direct dose from the station will be determined by obtaining the dose from TLD locations situated on-site near potential sources of direct radiation, as well as those TLDs near the site boundary which are part of the environmental monitoring program, and subtracting out the dose contribution from background. Additional methods to calculate the direct dose may also be used to supplement the TLD information, such as high pressure ion chamber measurements, or analytical design calculations of direct dose from identified sources (such as solid waste storage facilities).

The dose determined from direct measurements or calculations will be related to the nearest real person off-site, as well as those individuals on-site involved in activities at either the Education Center or the Rocks boat landing, to assess the contribution of direct radiation to the total dose limits of Technical Specification 3.11.4 in conjunction with liquid and gaseous effluents.

i 5

B.3-20

_ - - - - _ - - - - - - J

l '

Dose Projections I 3.11 l

l I Technical Specifications 3.11.1.3 and 3.11.2.4 require that , appropriate i portions of licuid and gaseous radwaste treatment systems, respectively, be I used to reduce radioactive effluents when it is projected that the resulting dose (s) would exceed limits which represent small fractions of the "as low as reasonably achievable criteria of Appendix I to 10CFR Part 50. The surveillance requirements of these Technical Specifications state that dose projections be perfortred at least once per 31 days when the liquid radwaste treatment systems or gaseous radwaste treatment systems are not being fully utilized. ,

I Since dose assessments are routinely performed at least once per 31 days to account for actual releases, the projected doses shall be determined by comparing the calculated dose from the last (typical of expected operations) completed 31-day period to the appropriate dose limi,t for,use of radwaste equipment, adjusted if appropriate for known or expect.ed dif ferences between past operational parameters and those anticipated for the,next 31 days.

3.11.1 Liquid Dose Projections I 1 4 The 31 -day liquid dose projections are calculated by the following:

(a) Determine the total body' D tb and organ , dose D, (Equations 3-1 and 3-2, respectively) for the last typical completed 31-day period. The last typical 31-day period should be one without significant identified operational dif ferences f rom the period being projected to, such as full power operation vs. periods when the plant is shut down.

I (b) Calculate the ratio (R)) of the total estimated volume of batch releases expected to be released for the projected period to that actually released in the reference period.  ;

I B.3-21 I

(c) Calculate the ratio (R )p cf the estimated grass primary ccolant activity for the projected period to the average value in the p reference period. Use the most recent value of primary coolant h activity as the projected value if no trend in decreasing or y increasing levels can be determined.

4 (d) Determine the projected dose from:

I Total Body: 0 =D

. R) .R Max. Organ: D mo pr =

D,, . R) .R 2 3.11.2 Gaseous Dose Proiections

% For the gaseous radwaste treatment system, the 31-day dose projections are calculated by the following:

b (a) Determine the gamma air doseaD }r (Equation 3-6), and the beta air l dose Dair (Equation 3-7) f rom the last typical 31-day operating period.

l (b) Calculate the ratio (R3) f anticipated number of curies of noble gas to be released from the hydrogen surge tank to the atmosphere I over the next 31 days to the number of curies released in the reference period on which the gamma and beta air doses are based.

l If no differences between the reference period and the next 3i days can be identified, set R

• 3 l

(c) Determine the projected dose from:

g l

Gansna Air: 0,{rpr=D{r.R a 3 Beta Air: D r pr

=D r.R3 B.3-22

Fcr the ventilaticn exhaust treatment system, the critical crgan dose f rom iodines, tritium, and particulates are projected for the next 31 days by the following:

(a) Determine the critical organ dose D , (Equation 3-8) from the y last typical 31-day operating period.

[ (b) Calculate the ratio (R 4) of anticipated primary coolant dose equivalent I-131 for the next 31 days to the average dose equivalent I-131 level during the reference period. Use the most current determination of DE I-131 as the projected value if no  !

trend can be determined.

(c) Calculate the ratio (R 5 an c pa e p ma n system leakage rate to the average leakage rate during the reference period. Use the current value of the system leakage as an estimate of the anticipated rate for the next 31 days if no trend can be determined.

(

(d) Determine the projected dose from:

Critical Organ: D co pr =D,.R4.R5 c

[

E d

c

[

E B.3-23

~

4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring stations are listed in Table B.4-1. The locations of the stations with respect to the Seabrook Station are shown on the maps in Figures B.4-1 to B.4-6.

L Direct radiation measurements are analyzed at the station. All other radiological analyses for environmental samples are performed at the Yankee Environmental Laboratory. The Laboratory participates in the U.S.

Environmental Protection Agency's Environmental Radioactivity Laboratory H

Intercomparison Studies Program for all the species and matrices routinely p analyzed.

L g

Pursuant to Specification 4.12.2, the land use census will be conducted k "during the growing season" at least once per 12 months. The growing season is defined, for the purposes of the land use census, as the period from June 1 to October 1. The method to be used for conducting the census will consist of one or more of the following, as appropriate: door-to-door survey, visual inspection from roadside, aerial survey, or consulting with local agricultural authorities.

r Technical Specification 6.8.1.3 requires that the results of the e Radiological Environmental Monitoring Program be summarized in the Annual Radiological Environmental Operating Report "in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, 1979." The l general table format will be used with one exception and one clarification, as follows. The mean and range values will be based not upon detectable measurements only, as specified in the NRC Branch Technical Position, but upon

[

all measurements. This will prevent the positive bias associated with the i

calculation of the mean and range based upon detectable measurements only.

l Seondly, the Lower Limit of Detection column will specify the LLD required by I ODCM Table A.5-2 for that radionuclide and sample medium.

B . 4-1

___________________________j

TABLE B.4-1 Radioloaical Environmental Monitorina Stationf a)

Distance From Exposure Pathway Sample Location Unit 1 Direction From and/or Sample and Desianated Code Containment (km) the Plant e

L

1. AIRBORNE (Particulate and Radioiodine)

" AP/CF-01 PSNH Barge 2.7 ESC L Landing Area AP/CF-02 Hampton Marina 2.7 E p AP/CF-03 SW Boundary 0.8 SW y AP/CF-04 W. Boundary 1.0 W AP/CF-05 Winnacunnet H.S.(b) 4.0 NNE AP/CF-06 Georgetown 24.0 SSW

( Substation (Control)

2. WATERBORNE

a. Surface WS-01 Hampton-Discharge Area 5.3 E WS-51 Ipswich Bay (Control) 16.9 SSE

( l

b. Sediment SE-02 SE-07 Hampton-Discharge Area (b)

Hampton Beach 5.3 3.1 E

E SE-08 Seabrook Beach (b) 3.2 ESE SE-52 Ipswich Bay (Control)(b) 16.9 SSE SE-57 Plum Island Beach 15.9 SSE (Control)(b)

3. INGESTION

a. Milk TM-04 Salisbury, MA 5.2 SW TM-08 Hampton Falls, NH 4.3 NNW TM-10 Hampton Falls, NH 4.8 WNW TM-20 Rowley, MA (Control) 16.3 S

b. Fish and Invertebrates (c)

FH-03 Hampton - Discharge 4.5 ESE Area FH-53 Ipswich Bay (Control) 16.4 SSE HA-04 Hampton - Discharge 5.5 E

( HA-54 Area Ipswich Bay (Control) 17.2 SSE MU-06 Hampton - Discharge 5.2 E f Area L MU-56 Ipswich Bay (Control) 17.4 SSE

[

B.4-2

'~

N s TABLE B 4-1

( (centinued)

RadioloaicalEnvironmentalMonitorinaStation[a)

Distance From Exposure Pathway Sample Location Unit 1 Direction From the Plant and/or Sample and Designated Code Containment (km)

4. DIRECT RADIATION TL-1 Brimmer's Lane, 1.1 N Hampton Falls TL-2 Landing Rd., Hampton 3.2 NNE g

y TL-3 Glade Path, Hampton' 3.1 NE Beach TL-4 Island Path, Hampton 2.4 ENE Beach

{ TL-5 Harbor Rd., Hampton 2.7 E beach

( TL-6 PSNH Barge Landing 2.7 ESE L Area TL-7 Cross Rd., Seabrook 2.6 SE Beach TL-8 Farm Lane, Seabrook 1.1 SSE

( TL-9 Farm Lane, Seabrook 1.1 S TL-10 Site Boundary Fence 1.0 SSW TL-11 Site Boundary Fence 1.0 SW TL-12 Site Boundary Fence 1.0 WSW TL-13 Inside Site Boundary 0.8 W TL-14 Trailer Park, Seabrook 1.1 WNW TL-15 Brimmer's Lane, 1.4 NW Hampton Falls TL-16 Brimmer's Lane, 1.1 NNW Hampton Falls

( TL-17 South Rd., N, Hampton 7.9 N TL-18 Mill Rd., N. Hampton 7.6 NNE TL-19 Appledore Ave., 7.9 NE

[, N. Hampton TL-20 Ashworth Ave., 3.4 ENE Hampton Beach TL-21 Route 1A, Seabrook 2.7 SE Beach TL-22 Cable Ave., 7.6 SSE Salisbury Beach

( TL-23 Ferry Rd., Salisbury 8.1 S TL-24 Ferry Lots Lane, 7.2 SSW r Salisbury

( TL-25 Elm St., Amesbury Route 107A, Amesbury 7.6 8.1 SW TL-26 WSW

( B.4-3

l TABLE B 4-1 L (continued)

I Radioloaical Environmental Monitorina Stationf a)

L Distance From F Exposure Pathway Sample Location Unit 1 Direction From L and/or Sample and Designated Code Containment (km) the Plant

~ TL-27 Highland St., 7.6 W j S. Hampton TL-28 Route 150, Kensington 7.9 WNW TL-29 Frying Pan Lane, 7.4 NW

[ Hampton Falls L TL-30 Route 101C, Hampton 7.9 NNW TL-31 Alumni Drive, Hampton 4.0 NNE TL-32 Seabrook Elementary 1.9 S r

L School TL-33 Dock Area, Newburyport 9.7 S TL-34 Bow St., Exeter 12.1 NW I TL-35 Lincoln Ackermaa 2.4 NNW

" School TL-36 Route 91, Georgetown 22 SSW F' (Control)

L TL-37 Plaistow, NH (Control) 26 WSW TL-38 Hampstead, NH (Control) 29 W TL-39 Epping, NH (Control) 27 NW I

TL-40 Newmarket, NH (Control) 24 NNW E TL-41 Portsmouth NH 21 NNE E (Control)(b)

TL-42 Ipswich, MA (Control)(b) 27 SSE l

I (a) Sample locations are shown on Figures B.4-1 to B.4-6.

(b) This sample location is not required by monitoring program defined in Part A of CDCM; program requirements specified in Part A do not apply to samples taken at l this location.

(c) Samples will be collected pursuant to 0DCM Table A.5-1. Samples are not required from all stations listed during any sampling interval (FH = Fish; l HA = Lobsters; MU = Mussels). Table A.5-1 specifies that "one sample of three I

l commercially and recreationally important species" be collected in the vicinity of the plant discharge area, with similar species being collected at a control location. (This wording is consistent with the NRC Final Environmental Statement for Seabrook Station.) Since the discharge area is off-shore, there is a great number of fish species that could be considered commercially or recreationally important. Some are migratory (such as striped bass), making them less desirable as an indicator of plant-related radioactivity. Some pelagic species (such as herring and mackerel) tend to school and wander throughout a large area, sometimes making catches of significant size difficult to obtain. Since the collection of all species would be difficult or impossible, and would provide unnecessary redundancy in terms of monitoring important pathways to man, three fish and invertebrate species have been specified as a minimum requirement. Samples may include marine fauna such as lobsters, clams, mussels, and bottom-dwelling fish, such as flounder or hake.

Several similar species may be grouped together into one sample if sufficient sample mass for a single species is not available af ter a reasonable effort has been made (e.g., yellowtail flounder and winter flounder).

B.4-4

h N

P

% \

?

l u p l l

( d 0

s BROWS RIVER I

L L

jEABROOK1 AP/ct-02@

susRooK 2 / , gggp7py gggggg AP/cr-o4 @

AP/CF-0 .\

F a KUMTS LAND CREEK l AP/CF-01 l

SE-08 e l

l

~ E .

\ 5 5 l {

0 500 1000 METERS h

~

e

/ n Figure B.4-1 Radiological Environmental Monitoring Locations Within 4 Kilometers of Seabrook Station B.4-5

-_-__-_?__________

I y 5 r.- o g l

.._.i w_. >

I E lif*tt Tl R S b

I i e

N I \

RYE BEACH I ..

I SEE ENTARCDENT IN TICURE 2.1 TH-03 Ap/cr-GS

__ S______'

I l

"~** @  ; l umnon etAcu

', leSE-07 g '. .

8 SEABR00K, STATION E i

g - DISCHARE SITE

• . WS-01 MU-06

.I . --.,- e FH-03 HA-04 SE-02 s'

- r, sEAsaoor stAcu I . -

l-- s

~

g_' f [- _ - .-'e

< a

@ TM-04

-SAllS8URY BtACH D

.s I

I ~

RRIurc ATLANTIC OCEAM

)

il Figure B.4-2 Radiological Environmental Monitoring Locations Between 4 Kilometers and 12 Kilometers from Seabrook Station B.4-6 I

tr I

I O 5 10 E I If>ti.TI.It S 15 i

4 YORK

'oY e lI s DURHAM e y I ,

'A PORTSMOUTH e M NEWHARKET e I ND e s\'g EPPING a

?

______ .____. ..________ g N

,EXETER e

, . , i SLT EN!ARCDGMT IN FIGURE 2. 2 e s HAMPTON e e s SEABROOK STATION l l KINGSTON e N(#

e SEABROCK e a --

,-'s, i

, ,e' \_-- DI)CHARGESITE

/ AM.ESBURY

• I PLAISTOW e [ l

l auric oceAu

. -.J.ii.  ! ,

i l

/ 4qg'y .

J e NEW3URYPORT e I HAVERHille i

k l e SE-57 s . 'I ' TM-20 PLtD8 ist.ANO FH-53 nEtuuEn .

"'" ' @ I-s$

e N EMCE IPsWICN 8Av IPSWICHe d

@f E ')

UCESTER 1

I Figure B.4-3 Radiological Environmental Monitoring Locations Outside 12 Kilometers of Seabrook Station I B. 4-7 I l

1

1. t I

F pq /

• NNE / i

\ N NNW

/ i s

e \

f . TL-2 ,!

/

? ,I u ,

/

~'

~ .. q @TL-3 "w

L 4 I

d 7

E E @TL-4 ENE i TL-15 TL-16 E

I @

@ TL-1 47 e WNW

• TL-14 @

W TL-13@ @

smR 2/ 's . HAMFroM HARS0R TL-12@ 'g

- i k C TL-11 @

ws @ , ,

ESE TL-7 @

l TL-32@

o l ~ g y

/ o 500 1000 x=

9

U s

TL-21g SSW METERS Ge /

p SSE SE :g S

Figure B.4-4 Direct Radiation Monitoring Locations Within 4 Kilometers I of Seabrook Station I B. 4-8 l

l

~ '

N g 4 o s yi NNW < - - < >---i 2 ---J P 6 : u>ii Ti n 3 NNE 9

F h NW NE y @ TL-34 L

5 Miles

\N J g rye BEACH

. TL-17 TL-30 TL 18 [

e,TL-19 F ,

L @ TL-29 j WNW sit ENTARc a su rzcune 2.4 ENE

~ ~~

• TL-28 '

l' ' ' I

\l T HAMPToM BEACH -

I W E DISCHE SITE

@ TL-27  ;

,. ~~' ' ' ' ~ b

, g

~

/ '

s

/ '

e TL-26 '

l i , ,.

,./

_./ g g %

s'.

s. . - -l l WSW

@ TL-25 SALIS8uty cu ESE l TL-24 g

TL-23 TL-2

• D NESturc l SE

) SW 7t.33g Ar m ic ocEAu

)

SSW SSE I -

S Figure B.4-5 '

Direct Radiation Monitoring Locations Between 4 Kilometers and 12 Kilometers from Seabrook Station B. 4-9 I

L r

[- - N O $ 10 15 ggg E ll> >u li.ns .

NNW $ , , ,

NW

/ k

[

.,e ts P/#

TL-40. Pc NE

[ @

1 PORT UTHe 9 U.

NEWMARKET e ,

@TL-39 TL-41 mMono

• trrises e 'o "'8*$ 's

[. /

- *#'\,k

.. .-. . . . . . .... ...... \ ENE

[- i i KETER , ,

N e

I 8 d

EAMCENENT IN FICURE 2.5 e ,

SEARROOK TI KINGSTliN e ' f"# -- E W i Sun i

{ . x.3, ~

-) ai

/s AMESB e' '

PLAIST0W e [  ! i l ' ATLANTIC QCEAN N.y l TL-31 % ,

,' 44,,5 -J- p -- --.

at vaTe0=T eSe I PLUM LAND s..

TL-36 @

e LAWREE E IP3WICN SE IPSWICH 8 e TL-42 SW

[- "

33W SSE Figure B.4 6 Direct Radiation Monitoring Locations outside 12 Kilometers of Seabrook Station

{-

B.4-10

L F 5.0 SETPOINT DETERMINATIONS L.

~

Chapter 5 contains the plant procedures that the plant operator L requires to meet the setpoint requirements of the Radioactive Effluent Monitoring Systems Technical Specifications. They are Specification 3.3.3.9

[. for liquids and Specification 3.3.3.10 for gases. Each outlines the instrumentation channels and the basis for each setpoint.

r L

r L

u r

L'

[

E E

[

[;

[

[

B.5-1

L 5.1 Liouid Effluent Instrumentation Setpoints L

Technical Specification 3.3.3.9 requires that the radioactive liquid effluent instrumentation in Table 3.3-12 of the Technical Specifications have

[

alarm setpoints in order to ensure that Specification 3.11.1.1 is not E exceeded. Specification 3.11.1.1 limits the activity concentration in liquid

^

effluents to the appropriate MPCs in 10CFR20 and a total noble gas MPC.

E 5.1.1 Liauid Waste Test Tank Monitor (RM-6509)

! The liquid waste test tank effluent monitor provides alarm and automatic termination of release prior to exceeding the concentration limits

! specified in 10CFR20, Appendix B, Table II, Column 2 to the environment. It is also used to monitor discharges from various waste sumps to the environment.

l Method to Determine the Setpoint of the Liquid Waste Test Tank I 5.1.1.1 Monitor (RM-6509)

The instrument response (pCi/ml) for the limiting concentration at the point of discharge is the setpoint, denoted Rsetpoint, and is determined as follows:

l R

setpoint " fl D mi (5-1) min i (pCi/ml) ()() ( )

where:

l d

DF = p = Dilution factor (dimensionless) (5-2) m F, = Flow rate past monitor (gpm)

F = Flow rate out of discharge tunnel (gpm) d DF g = Minimum allowable dilution factor (dimensionless)

I B.5-2 I

{ f)

= 1 - (f 2 + I3); where f) is the fraction of the total contribution of MPC at the discharge point to be associated j with the test tank effluent pathway and,2f and f3 are the

' similar fractions for Turbine Building sump and steam generator blowdown pathways, respectively: (f) +f2+f3 I II

• L DF Ib- )

r min " M i

( 1 MPC = MPC for radionuclide "i" from 10CFR20, Appendix B, Table II, r $

L Column 2 (uci/ml). In the event that no activity is expected to be discharged, or can be measured in the system, the b liquid monitor setpoint should be based on the most restrictive MPC for an " unidentified" mixture given in 10CFR20, Appendix B, notes.

{

C ,g

= Activity concentration of radionuclide "i" in mixture at the

{ monitor (vCi/ml) 5.1.1. 2 Liouid Waste Test Tank Monitor Setpoint Example b The activity concentration of each radionuclide, C ,g, in the waste test tank is determined by analysis of a proportional grab sample obtained at the radwaste sample sink. This setpoint example is based on the following

( data:

i C ,g (pCi/ml) MPC4 (yCi/ml)

Cs-134 2.15E-05 9E-06 Cs-137 7.48E-05 2E-05 Co-60 2.56E-05 3E-05 C,$ = 2.15E-05 + 7.48E-05 + 2.56E-05 i

IuCi) ml Iuti) m1 Iuti) ml Iuti ml I

= 1.22E-04 uti Iml )

{

B.5-3

u C

7 ni

- DF " (5-3) min MPC i i r

- uCi-m1 Iml pCi)

T L

" 2.15E-05 + 7.48E-05

• 2.56E-05 9E-06 2E-05 3E-05 I

L uCi-m1 uCi-m1 Imi pC1) IuCi-m1)ml pC1 Imi pCi)

F L

DF min "

The minimum dilution factor, DFmin, needed to discharge the mixture of radionuclides in this example is 7. The release rate of the waste test tank L is between 10 and 150 gpm. The circulating water discharge flow can vary f rom 10,500 to 412,000 gpm of dilution water. With the dilution flow taken as

[ 412,000 gpm and the release rate from the waste test tank taken as 150 gpm, the DF is:

d DF = -

[ F, I (qpm) (5-4) l (gpm)

I l 412.000 apm 150 gpm I

= 2750 I

l l

B.5-4

L Under these conditions, and with the fraction f) of total MPC to be associated with the test tank selected as 0.6, the setpoint of the liquid L radwaste discharge monitor is:

r Rsetpoint " I l D C ,$ (5-1) min i H uCi Iuti)

L ml I II ml 2 0 I = 0.6 1.22E-04 u ()() ("n I

= 2.87E-02 pCi/ml or pCi/cc

{

In this example, the alarm of the liquid radwaste discharge monitor should be set at 2.87E-02 pCi/cc above background.

5.1. 2 Turbine Building Drains Liouid Ef fluent Monitor (RM-6521)

The Turbine Building drains liquid effluent monitor centinuously monitors the Turbine Building sump effluent line. The only sources to the Sump Effluent System are from the secondary steam system. Activity is expected in the Turbine Building Sump Effluent System only if a significant primary-to-secondary leak is present. If a primary-to-secondary leak is present, the activity in the sump effluent system would be comprised of only I those radionuclides found in the secondary system, with reduced activity from decay and dilution.

The Turbine Building drains liquid effluent monitor provides alarm and l automatic termination of release prior to exceeding the concentration limits specified in 10CFR20, Appendix B, Table II, Column 2 to the environment. The alarm setpoint for this monitor will be determined using the same method as B

B B.5-5 I

I that of the liquid waste test tank monitor if the total sump activity is greater than 10 percent of MPC. If the total activity is less than 10 percent of MPC, the setpoints of RM-6521 are calculated as follows:

High Trip Monitor =f2 (DF') (1.0E-07 pCi/ml) (5-21)

I Setpoint (uci/ml) where:

I DF' =

Circulatina water flow rate (apm)

Flow rate post-monitor (gpm) 1.0E-07 pCi/ml = most restrictive MPC value for an unidentified mixture given in 10CFR20, Appendix B, Note 3b.

f2 " l - (f 1 + f 3); where the f values are described above.

Warning Alarm High Trip (5-22)

I Monitor Setpoint (pCi/ml)

=IMonitor Setpoint) (0.25) 5.1. 3 Steam Generator Blowdown Liquid Sample Monitor (RM-6519)

The steam generator blowdown liquid sample monitor is used to detect abnormal activity concentrations in the steam generator blowdown flash tank liquid discharge.

The alarm setpoint for the steam generator blowdown liquid sample monitor, when liquid is to be discharged from the site, will be determined using the same approach as the Turbine Building drains liquid effluent monitor.

For any liquid monitor, in the event that no activity is expected to be g

E discharged, or can be measured in the system, the liquid monitor setpoint should be based on the most restrictive MPC for an " unidentified" mixture given in 10CFR20, Appendix B notes.

'I B.5-6

r

" 5.1. 4 PCCW Head Tank Rate-of-Change Alarm Setpoint

- A rate-of-change alarm on the liquid level in the Primary Component Cooling Water (PCCW) head tank will work in conjunction with the PCCW radiation monitor to alert the operator in the Main Control Room of a leak to the Service Water System from the PCCW System. For the rate-of-change alarm, E a setpoint is selected based on detection of an activity level equivalent to

-8 Ci/ml in the discharge of the Service Water System. The activity in 10 the PCCW is determined in accordance with the liquid sampling and analysis

(

" program described in Part A, Table A.3-1 of the ODCM and is used to determine g the setpoint.

u The rate-of-change alarm setpoint is calculated f rom:

C l

RC = l x10 -8 , 39p , p (5-23) set aal uti I hr I"I ml I Iaal) hr Iml vCi) where:

RCset = the setpoint for the PCCW head tank rate-of-change alarm (in gallons per hour).

l x10-8 = the minimum detectable activity level in the Service I Water System due to a PCCW to SWS leak (vCi/ml).

SWF = Service Water System flow rate (in gallons per hour).

PCC = Primary Component Cooling Water measured (decay corrected) gross radioactivity level (vCi/ml).

I -5 pCi/mi As an example, assume a PCCW activity concentration of lx10 with a service water flow rate of only 80 percent of the normal flow of 21,000 i

gpm. The rate-of-change setpoint is then:

-0 -5 RCset = lx10 1.0x106 g h (1/1x10 )

RC = 1000 gph set B.5-7 l

i e

H As a result, for other PCCW activities, the RC set which would also y relate to a detection of a minimum service water concentration of

-8 pCi/ml can be found from:

L 1x10

-5 uti/ml 1000 aph

" 1x10 (5-24)

RCset = PCC I

u

?

L e

E E

1 I

I I

I l B.5-8 I

5.2 Gaseous Effluent Instrumentation SetDoints Technical Specification 3.3.3.10 requires that the radioactive gaseous effluent instrumentation in Table 3.3-13 of the Technical Specifications have their alarm setpoints set to insure that Technical Specification 3.11.2.1 is not exceeded.

f~

5.2.1 Plant Vent Wide-Ranae Gas Monitors (RM-6528-1.2 and 3)

The plant vent wide-range gas monitors are shown on Figure B.6-2.

5.2.1.1 Method to Determine the Setpoint of the Plant Vent Wide Range Gas Monitors (RM-6528-1.2 and 3)

I The setpoint for the plant vent wide-range gas monitor (readout 3

response in pCi/cm ) is set by limiting the off-site noble gas dose rate to

{ the total body or to the skin, and is denoted Rsetpoint. R setpoint is the lesser of:

R tb

= 806 h DFB c

3 3

(pCi/cm )(mrem yr-pCi-sec uci-m ) Isec) 3 I oCi-vr) 3 cm mrem-m and:

R skin

= 3,000 h DF' c

(5-6) i 3 uti-(uci/cm) (mrem) yr (sec) (mrem sec vr )

e ,3 where:

R = Response of the monitor at the limiting total body dose tb 3 rate (vCi/cm )

[

B.5-9

L 3

500 806 = (mm-uci-m yr-pCi-sec )

(1E+06) (6.2E-07)

- I L 500 = Limiting total body dose rate (mrem /yr)

F L. 1E+06 = Number of pCi per pCi (pCi/pCi) 6.2E-07 = [X/Q)T, maximum annual average gamma atmospheric dispersion f actor (sec/m )

L 3

F = Appropriate plant vent flow rate (cm /sec) g DFB = Composite total body dose factor (mrem-m /pCi-yr) c hj DFB q i

= (5-7) i Qg

= The release rate of noble gas "i" in the mixture, for each noble gas identified in the off-gas (pci/sec)

{

3 DFB = Total body dose f actor (see Table B.1-10) (mrem-m / Ci-yr) 4 R = Response of the monitor at the limiting skin dose rate skin (pCi/cm3 )

[

B . 5-10

I 3,000 = Limiting skin dose rate (mrem /yr)

DF' = Composite skin dose factor (mrem-sec/pCi-yr) bj DFj i

=

(5-8) 63 i

DFj = Combined skin dose fa: tor (see Table B.1-10)

(mrem-sec/pCi-yr)

5. 2.1. 2 Plant Vent Wide Range Gas Monitor Setpoint Example The following setpoint example for the plant vent wide range gas monitors demonstrates the use of equations 5-5 and 5-6 for determining setpoints.

The nominal plant stack flow is 4.3E+07 cc/sec ((153,200 cfm x 28,300 I 3 cc/ft )/60 sec/ min).

This setpoint example is based on the following data (see Table B.1-10 for DFB$ and DF4 ):

h4 DFB 4

0Fj 3

$ gjiq mrem-m ) mrem-sec) sec DCi-vr uCi-vr Xe-138 1.03E+04 8.83E-03 1.21E-02 Kr-87 4.73E+02 5.92E-03 1.77E-02 Kr-88 2.57E+02 1.47E-02 1.38E-02 Kr-85m 1.20E+02 1.17E-03 2.86E-03 Xe-135 3.70E+02 1.81E-03 3.89E-03 Xe-133 1.97E+01 2.94E-04 6.66E-04 B . 5-11

F h4DFB g

=

I (5-7)

DFB c ,

r 01 m i F .

Q4 DFB4 = (1.03E+04)(6.83E-03) + (4.73E+02)(5.92E-03)

[ + (2.57E+02)(1.47E-02) + (1.20E+02)(1.17E-03) y + (3.70E+02)(1.81E-03) + (1.97E+01)(2.94E-04)

L 3

= 9.83E+01 (pCi-mrem-m /sec-pCi-yr)

I L .

Q = 1.03E+04 + 4.73E+02 + 2.57E+02 4

L

+ 1.20E+02 + 3.70E+02 + 1.97E+01

= 1.15E+04 pCi/sec p 0F8 9.83E+01 L c " 1.15E+04 3

= 8.52E-03 (mrem-m /pCi-yr)

{ Rg = 806 h DF8c

= (806) (4.3E+07) (8.52E-03)

= 2.20E-03 pCi/cm bj DFj DF' = (5-8) l 63 i

h40Fj = (1.03E+04)(1.21E-02) + (4.73E+02)(1.77E-02) i B . 5-12

+ (2.57E+02)(1.38E-02) + (1.20E+02)(2.86E-03)

+ (3.70E+02)(3.89E-03) + (1.97E+01)(6.66E-04)

= 1.38E+02 (pCi-mrem-sec/sec pci-yr)

DF'c

= 1. 8N2 1.15E+04

= 1.20E-02 (mrem-sec/pci-yr)

Rskin = 3,000 hDF' '

= (3,000)

(4.3E+07) (1.20E-02)

= 5.80E-03 pCi/cm 3 The setpoint, Rsetpoint, is the lesser of Rtb and R skin. For the noble gas mixture in this example R tb is less than R skin, n cadng hat

the total body dose rate is more restrictive. Therefore, in this example the 3

plant vent wide-range gas monitors should each be set at 2.20E-03 pCi/cm above background, or at some administrative f raction of the above value.

In the event that no activity is expected to be released, or can be measured ir, the system to be vented, the gaseous monitor setpoint should be l

based on Xe-133.

1 I

1 B . 5-13

~ 6.0 LIOUID AND GASE0US EFFLUENT STREAMS. RADIATION MONITORS AND RADWASTE-5 TREATMENT SYSTEMS Figure B.6-1 shows _the liquid ef fluent streams, radiation monitors and the appropriate Liquid Raowaste Treatment System. Figure B.6-2 shows the J gaseous effluent streams, radiation monitors and the appropriate Gaseous Radwaste Treatment System.

For more detailed information concerning the above, refer to the P Seabrook Station Final Safety Analysis Report, Sections 11.2 (Liquid Waste

' System), 11.3 (Gaseous Waste System) and 11.5 (Process and Effluent r Radiological Monitoring and Sampling System).

The turbine gland seal condenser exhaust is an unmonitored release

! path. The iodine and particulate gaseous releases will be determined by continuously sampling the turbine gland seal condenser exhaust. The noble gas releases will be determined by the noble gas released via the main condenser air evacuation exhaust and ratioing them to the turbine gland seal condenser exhaust by use of the flow rates.

l l

l l

1 l

1 l

1 8.6-1

{

L_

F 5 O A

MAKEUP

" MAMEUP STOR AG E STORAGE pag TANK TANK ' g UNIT 1 UNIT 2 UNIT 2 UNIT 1 UNIT 2 UNIT 1 I i t s l .

i

, l @' l 6,"i '

r ,L_ _G

_. e t _ __ _g _i _

' I 3 BORON -

RECOVERY g

-1 i

SYST EM

(

i 9

', " _ _ ._g_ . . . . .

_ . .h i

g STEAM Lwps TURBINE I B L NG R YCLE BLOhvDOWN POR ON UNI 2 UNIT 1 PORT ON SYSTEM - - - e-I I j' 1, 8

l i L t. _ _f  ; __ __

', i l

( I

\- -

d ii 3

~

-RM-6521) ( RM-6519) f ,

y ( RM-6509 ) i CIRCULATING q it W AT E R 2 ---

g SYSTEM PCCW _DllC SWS

{ (2-RM-6521)

SYSTDI SYSTEM RELEASE

" Future" CCLT-2172-1 CCLT-2272-1 h CVCS LETDOWN DIVERSION EQUIPMENT DR AIN AGE EQUIPMENT LEAKAGE LEGEND h PAB FLOOR DRAINS NON. RECYCLABLE AND MISC.

CONTAINMENT SUMPS

- - - - - CONDENSATE LEAK AGE LABORATORY DR AINS D OMAM ATION WATER RECYCLABLE DE AER ATED

---

RECYCLABLE AEREATED h TURBINE BLOG SUMP NON-RECYCLABLE AND MISC. h TRtTIUM CONTROL RELEASE

- - - - ST E AM GE N. BLOWDOWN RM Radiation Monitor h SECONDARY SIDE STM. GEN. BLOWDOWN

@ Service Water System CCLT Level Transmitter Figure B.6*1 Liquid Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Seabrook Station B.6-2

r L -

, Turbine Gland / RM-6528-1,2,3 ,

e Seal Condenser RM-6530-1,2 Exhaust , i L- CONT AINME NT

/

[ d BUILDING VENTIL a TORS }

d' TURBINE d6 d' STEAM Btf t LOING REACTOR GENERATOR , ,

. VACUUM

.TURS'NE . CONDENSER PUMP

'g' m SECONDARY ,

E VACU.JM F ] STEAM i W. H C +

g -

pygg ,

{ g g [

C ,

j -hb EFFLUENT j/ ,,

' i t CONT AINMEtJT PtlRGE AIRc '

@j!

w ,,

10 d

1 I- q.

PLC OWV W AST E J FLASH TANK SUILDING f "

REACTOR ftR d' COOLANT ,

CAJEOUS W ASTE PROritSSING SYS?EM '

- ;y-: r 5,

• k.9: , f ,8.  %

[

- , ' , e ' r %- .s - ti PRIM AR Y

+ . ..d

.cd . TYPICAL OF 3 I di d 6 ii i i s "*

GUARD

'2

. I a I VENT

• e SED <'if i

'd' 5 '

f STACK

~

AFTER

• COOLER m

j 2 H @ )ol Es @ SURGE DRYER CHARCOALBECS PRIMARY '

COMPRESSOR AUxittARY 8UILDING

" \

VOLLME @ (Rft-6504 /

' ODNTROL ,

TANK

{ ,

DEG A,5?FIE R l 1f y AUXf LIARY PgLDING VENT AIR LEGEND H flEPA FILTER C - CHARCOAL FILTER FUEL h_f -  !

( RM - Radiation Monitor BUILDING E 3 l

E F gurc 6-2 Gaseous Effibent Streams, Radiation Monitors, and Radwaste Treatment System at Scabrook Station B.6-3 b

l 7.0 BASES FOR DOSE CALCULATION METHODS

, 7.1 Liauid Release Dose Calculations This section serves: (1) to document the development and conservative nature of Method I equations to provide background infonnation to Method I users, and (2) to identify the general equations, parameters and approaches to Method II-type dose assessments.

y Method I may be used to show that the Technical Specifications which limit of f-site total body dose f rom liquids (3.11.1.2 and 3.11.1.3) have been met for releases over the appropriate periods. The quarterly and annual dose h

limits in Technical Specification 3.11.1.2 are based on the ALARA design objectives in 10CFR50, Appendix I Subsection II A. The minimum dose values noted in Technical Specification 3.11.1.3 are " appropriate fractions," as determined by the NRC, of the design objective to ensure that radwaste equipment is used as required to keep off-site doses ALARA.

Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated" (10CFR50,

( Appendix I). The definition, below, of a single " critical receptor" (a hypothetical or real individuH whose behavior results in a maximum potential dose) provides part of the conservative margin to the calculation of total

{ body dose in Method I. Method II allows that actual individuals, associated with identifiable exposure pathways, be taken into account for any given release. In fact, Method I was based on a Method II analysis for a critical receptor assuming all principal pathways present instead of any real individual. That analysis was called the." base case;" it was then reduced to form Method I. The general equations used in the base case analysis are also

[ used as the starting point in Method II evaluations. The base case, the method of reduction, and the assumptions and data used are presented below.

h The steps performed in the Method I derivation follow. First, the dose impact to the critical receptor [in the form of dose factors DFLitb

[ (mrem /pCi)] for a unit activity release of each radioisotope in liquid effluents was derived. The base case analysis uses the general equations, h methods, data and assumptions in Regulatory Guide 1.109 (Equations A-3 and

{ B . 7-1 i

I A-7, Reference A). The liquid pathways contributing to an individual dose are l due to consumption of fish and invertebrates, shoreline activities, and l swimming and boating near the discharge point. A normal operating plant discharge flow rate of 918 f t3/sec was used with a mixing ratio of 0.10.

I The mixing ratio of 0.10 corresponds to the minimum expected prompt dilution or near-field mixing zone created at the ocean surface directly above the multiport diffusers. (Credit for additional dilution to the outer edge of the prompt mixing zone which corresponds to the 1 F surface isotherm can be applied in the Method II calculation.) The location of the critical receptor is assumed to be the edge of the mixing zone at the ocean surface. The transit time used for the aquatic food pathway was 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and for shoreline activity 0.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Table B.7.1-1 outlines the human consumption and environmental parameters used in the analysis. The resulting, site-specific, I total body dose f actors appear in Table B.1-11.

Note that the liquid dose factors calculated reflect a one unit operation. Liquid waste from both units is processed by a common processing facility. In the case of two-unit operation, the liquid waste releases must be apportioned accordingly to each unit (the method to apportion between each unit will be addressed prior to Unit 2 completion).

7.1.1 Dose to the Total Body For any liquid release, during any period, the increment in total body dose from radionuclide "i" is:

AD =k Q 0FL (7-1 )

tb 4 itb (mrem) ( ) (9C1) (*C )

where:

DFLitb = Site-specific total body dose factor (mrem /pCi) for a liquid release. It is the highest of the four age groups.

See Table B.1-11.

I Qt

= Total activity (pCi) released for radionuclide "i".

918/Fd ; where Fd is the average dilution flow of the I K =

Circulating Water System at the point of discharge from the multiport diffuser (in ft /sec).

3 I B.7 2 I

l I M2thod I is more conservative than Method II in the region of the Technical Specification limits because the dose factors DFL g used in Method I were chosen for the base case to be the highest of the four age gr ups (adult, teen, child and inf nt) f r that r di nuclide. In effect each 5

5 radionuclide is conservatively represented by its own critical age group.

7.1.2 Dose to the Critical Organ i

The methods to calculate maximum organ dose parallel to the total body dose methods (see Section 7.1.1).

I For each radionuclide, a dose factor (mrem /pci) was determined for each of seven organs and four age groups. The largest of these was chosen to be

.g

.g the maximum organ dose factor (DFLg) for that radionuclide. DFL $

also includes the external dose contribution to the critical organ.

For any liquid release, during any period, the increment in dose from radionuclide "i" to the maximum organ is:

AD,o =k Q DFL gg (7-2)

(mrem) ( ) (pCi) (*C )

I where:

I DFlimo = Site-specific maximum organ dose factor (mrem /vCi) for a liquid release. See Table B.1-ll.

Qi

= Total activity (yCi) released for radionuclide "i".

K = 918/Fd ; where Fd is the average dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in ft 3/sec).

I I

'I B.7 3 I

m mM'm m m m3 U n F3 n F3 _ FD a F- T F3 5 L-- r' TABLE B.7-1 Usage Factors for Various Liquid Pathways at Seabrook Station (From Reference A, Table E-5.* Zero where no pathway exists) l LEAFY MILK MEAT FISH INVERT. POTABLE SHORELINE SWIMMING *** BOATING *** l AGE VEG.

VEG. WATER (LITER /YR) (KG/YH) (KG/YR) (KG/YR) (LITER /YR) (HR/YR) (HR/YR) (HR/YR)

(KG/YR) (KG/YR) 0.00 0.00 21.00 5.00 0.00 334.00** 8.00 29.00 Adult 0.00 0.00 0.00 0.00 16.00 3.80 0.00 67.00 45.00 52.00 l Teen 0.00 0.00 0.00 0.00' 6.90 1.70 0.00 14.00 28.00 52.00 Child 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Infant 0.00 0.00 5

Y i

i

• Regulatory Guide 1.109

• • Regional shoreline use associated with mudflats - Maine Yankee Atomic Power Station Environmental Report

• • • HERMES; "A Digital Computer Code for Estimating Regional Radiological Effects from Nuclear Power Industry,"

HEDL, December 1971 i

[ 7.2 Gaseous Release Dose Calculations F 7.2.1 Total Body Dose Rate From Noble Gases L

e This section serves: (1) to document the development of the Method I b equation, (2) to provide background information to Method I users, and (3) to identify the general equations, parameters and approaches to Method II-type L dose rate assessments.

Method I may be used to show that the Technical Specification which

( limits total body dose rate from noble gases released to the atmosphere (Technical Specification 3.11.2.1) has been met for the peak noble gas release rate.

Method I was derived from general equation B-8 in Regulatory Guide 1.109 as follows:

~

b = lE+06 [X/Q] hg DFB (7-3) tb 4 3

mrem sec mrem-m I yr I IDCi) I)I,3 ) Iuti) I I

( uci sec pCi-yr where:

[X/Q]T = Maximum receptor location long-term average gamma atmospheric dispersion factor.

3

= 6.2E-07 (sec/m ),

h4 = Release rate to the environment of noble gas "i" (pCi/sec).

3 DFB g = Gamma total body dose f actor, ( ). See Table B.1-10.

{ (Regulatory Guide 1.109, Table B-1).

y Equation 7-3 reduces to:

{

b tb

= 0.62 Eh 4 DFB i (3-3)

E B.7-5

[

3 rrem DCi-sec uCi rrem-m I yr I* 3 I I sec II pci-yr I Ci-m r

The selection of critical receptor, outlined in Section 7.3 is inherent in the j L derived Method I, since the maximum expected off-site long-tr.rm average f atmospheric dispersion factors were used. All noble gases in Table B.1-10 I

L should be considered.

A Method II analysis could include the use of actual concurrent meteorology to

[ assess the dose rates as the result of a specific release.

7.2.2 Skin Dose Rate From Noble Gases This section serves: (1) to document the development of the Method I equation, (2) to provide background information to Method I users, and (3) to

[ identify the general equations parameters and approaches to Method II-type dose rate assessments. The methods to calculate skin dose rate parallel the total body dose rate methods in Section 7.2.1. Only the differences are

{

presented here.

Method I may be used to show that the Technical Specification which limits skin dose rate from noble gases released to the atmosphere (Technical Specification 3.11.2.1) has been met for the peak noble gas release rate.

The annual skin dose limit is 3,000 mrem (from NBS Handbook 69 Reference D, pages 5 and 6, is 30 rem /10). The factor of 10 reduction is to account for nonoccupational dose limits.

(

It is the skin dose commitment to the critical, or most limiting,

{ off-site receptor assuming long-term site av.erage meteorology and that the release rate reading remains constant over the entire year.

Esthod I was derived from the general equation B-9 in Regulatory Guide h 1.102 as follows:

[ D S

= 1.11 D[ir + 3.17E+04 i Q$ [X/Q]

DFS (7-4)

B.7-6 Y

i - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

L 3

rrem trad Ci sec rrem-m

' I yr I"III yr I IDCi-vr}

Ci-sec  % I ,3 II pCi-yr I

[ where:

r 1.11 = Average ratio of tissue to air absorption coefficients (will L convert mrad in air to mrem in tissue).

7 DFSj = Beta skin dose factor for a semi-infinite cloud of L radionuclide "i" which includes the attenuation by the outer

" dead" layer of the skin.

F T (7-5)

= 3.17E+04 Qg [X/Q] DF g D}ir r 3 L mrad IDC1-vr)

Ci sec mrad-m I pCi-yr I yr I Ci-sec IF) I,3 DF{ = Gammaradionuclide air dose "i". factor for a uniform semi-infinite cloud of

( Now it is assumed for the definition of (X/Q ) for Reference B that:

Djinite = D air [X/Q] /[ /Q] ( -6) 3 mrad sec m Imrad) I yr II I Isec) l yr ,3 and Q = 31.54 hg (7-7) l $

.C_i) yr (Ci-sec) pCi-yr uti) sec so: D = 1.11 1E+06 [X/Q]Y Qg DF} (7-8) skin i 3

mrem mrad-m I yr I"II uti (sec)

IDCi)

,3-Iuti) sec I pci-yr I

+ 1E+06 X/Q hg DFS g i

3 sec uCi (DCi)

Ci sec Imrem-m pCi-yr )

,3 I B.7-7

u f substituting L

[X/Q]Y = 6.2E-07 sec/m3 r = 1.4E-06 sec/m3 L X/Q (7-9)

L gives b skin = 0.62 hg DF}+ 1.40 hg DFS g i i 3 3

[ gmrem) yr DCi-sec-mremguci){ mrem-m pCi-m3-mrad sec pCi-yr oci-secguci)(mrem-m )

pCi-m3 sec pCi-yr F

= (7-10) i h4 [0.62 DF} + 1.40 DFS g ]

( define DFj = 0.62 DF} + 1.40 DFS g (7-11 )

then: b "

DFj (3-4) skin $

i

~

mrem) yr uCi) sec mrem-sec) pCi-yr The selection of critical receptor, outlined in Section 7.3, is inherent in the derived Method I, as it is based on the determined maximum expected off-site atmospheric dispersion factors at the most limiting location. All noble gases in Table B.1-10 must be considered.

I 7.2.3 Critical Organ Dose Rate From Iodines. Tritium and Particulates With Half-Lives Greater Than Eight Days This section serves: (1) to document the development of the Method I equation, (2) to provide background information to Method I users, and (3) to identify the general equation's parameters and approached to Method 11 type dose rate assessments. The methods to calculate skin dose rate parallel the total body dose rate methods in Section 7.2.1. Only the differences are presented here.

i I B.7-8 I

L Method I may be used to show that the Technical Sp2cificaticn which limits organ dose rate f rom iodines, tritium and radionuclides in particulate r form with half lives greater than 8 days released to the atmosphere (Technical L Specification 3.11.2.1) has been met for the peak above-mentioned release rates.

L The annual organ dose limit is 1500 mrem (from NBS Handbook 69, Reference D, pages 5 and 6). ~It is evaluated by looking at the critical organ

[

dose connitment to the most limiting off-site receptor assuming long-term site I average meteorology.

L The equation for D cg is derived by modifying Equation 3-8 from

{ Section 3.9 as follows:

D g

= Q$ DFG $g (3-8)

(mrem) (pCi) (*C I applying the conversion factor, 3.154E+07 (sec/yr) and converting Q to I .

L Q in pCi/sec yields (7-12)

( bc , = 3.154E+07 h4 DFG

$co sec mrem F uti)

L (mrem) yr yr ) sec pCi y Eq. 3-8 is rewritten in the form:

O, c =Eh j 4 DFGjco (3-5) mrem) yr uti) sec mrem-sec) pCi-yr

[ where DFGjco

- 3.154E+07 DFG ico (7-13) mrem-sec I pCi-yr I" Isee) yr Imrem) pci

[

E B.7-9 m a e sii ei e i

E The selection of critical receptor, outlined in Section 7.3 is inherent in Method I, as are the maximum expected off-site atmospheric dispersion factors.

Should Method 11 be needed, the analysis for critical receptor, critical pathway (s) and annual average atmospheric dispersion factors may be

{ performed with concurrent meteorology and latest land use census data to identify existing pathways.

Because of the choice of atmospheric dispersion factors and pathways, it is expected that Method I results always will exceed Method II calculations. Either method provides adequate margin to ensure that the

( annual average concentrations based on organ dose f rom 10CFR20.106(a) are not exceeded and that the derived peak release rates are conservative.

7.2.4 Gamma Dose to Air From Noble Gases This section serves: (1) to document the development and conservative nature of Method I equations to provide background information to Method I users, and (2) to identify the general equations, parameters and approaches to Method II-type dose assessments.

Method I may be used to show that the Technical Specification which limits of f-site gamma air dose f rom gaseous ef fluents (3.11.2.1) has been met

{ for releases over appropriate periods. This Technical Specification is based on the objective in 10CFR50, Appendix I, Subsection B.1, which limits the estimated gamma air dose at unrestricted area locations.

For any noble gas release, in any period, the increment in dose is taken f rom Equations B-4 and B-5 of Regulatory Guide 1.109 with the added b- assumptionthatDfinite = DDQ]DQD Q DF{ (7-14)

{ AD{r a = 3.17E+04 [X/Q]T $

r

,c)(sec/mb (Ci) }

( (mrad) = ( -pC B . 7 -10

{

u. . .. . . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ ___J

whero:

3.17E+04 = number of pCi per Ci divided by the number of seconds per year.

[X/Q)T = maximum annual average gamma atmospheric dispersion factor 3

= 6.2E-07 (sec/m )

Q = number of curies of noble gas "i" released 4

I 0F} = Gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i".

l which leads to:

D ar

= 2.0E-08 Q 4 DF} (3-6) l i rd (mrad) (DCi-v"3 ) (pCi) ("C _ )

l pCi-m The major difference between Method I and Method II is that Method II would use actual or concurrent meteorology with a specific noble gas release I spectrum to determine [X/Q]T rather than use the most limiting meteorological dispersion value obtained for the years 1979 to 1981.

I l 7.2.5 Beta Dose to Air From Noble Gases I

l This section serves: (1) to document the development and conservative nature of Method I equations to provide background information to Method I users, and (2) to identify the general equations, parameters and approaches to l

Method II-type dose assessments.

Method I may be used to show that the Technical Specification which limits off-site beta air dose f rom gaseous ef fluents (3.11.2.1) has been met l for releases over appropriate periods. This Technical Specification is based on the Objective in 10CFR50, Appendix I, Subsection B.1, which limits the estimated beta air dose at unrestricted area locations.

For any noble gas release, in any period, the increment in dose is taken f rom Equations B-4 and B-5 of Regulatory Guide 1.109:

l B . 7 -l l I  !

1

L AD = 3.17E+04 X/Q Q DF (7-15) ir $

1

~#

(mrad) = (DCi-sec) (sec) ,3

( Ci) (mra 6m )

pCi-yr r

E where: DF = Beta air dose factors for a uniform semi-infinite cloud of radionuclide "i".

[ substituting X/Q = Maximum long-term average undepleted atmospheric dispersion

{ factor

= 1.4E-06 sec/m 3, We have I

D = 4.4E-08 Q DF (3-7) i (mrad) =pCi-m(D@V3) (pCi) (*C - I l

7.2.6 Dose to Critical Oraan From Iodines. Tritium and Particulates With Half-Lives Greater Than Eiaht Days l

This section serves: (1) to document the development and conservative nature of Method I equations to provide background information to Method I users, and (2) to identify the general equations, parameters and approaches to Method II-type dose assessments.

I Method I may be used to show that the Technical Specifications which limit off-site organ dose from gases (3.11.2.3 and 3.11.4) have been met for I releases over the appropriate periods. Technical Specification 3.11.2.3 is based on the ALARA Objectives in 10CFR50, Appendix I, Subsection II C.

Technical Specification 3.11.4 is based on Environmental Standards for Uranium Fuel Cycle in 40CFR190, which applies to direct radiation as well as liquid and gaseous effluents. These methods apply only to iodine, tritium, and particulates in gaseous effluent contribution.

I B . 7 -12

u Method I was develcped such that "the actual cxposure of cn 5

individual ... is unlikely to be substantially underestimated" (10CFR50, Appendix I). The use below of a single " critical receptor" provides part of the conservative margin to the calculation of critical organ dose in

- Method I. Method II allows that actual individuals, associated with L identifiable exposure pathways, be taken into account for any given release.

In fact, Method I was based on a Method II analysis of a critical receptor assuming all pathways present. That analysis was called the " base case"; it was then reduced to form Method I. The base case, the method of reduction,

[ and the assumptions and data used are presented below.

F The steps performed in the Method I derivation follow. First, the dose impact to the critical receptor [in the form of dose factors DFGico

- (mrem /pti)] for a unit activity release of each iodine, tritium, and L particulate radionuclide with half lives greater than eight days to gaseous effluents was derived. Seven exposure pathways (ground plane, inhalation, I stored vegetables, leafy vegetables, cow milk, goat milk, and meat ingestion) were assumed to exist at the site boundary (not over water or marsh areas) which exhibited the highest long-term X/Q. Doses were then calculated to six organs (bone, liver, kidney, lung, GI-LLI, and thyroid), as well as for the f whole body and skin for four age groups (adult, teenager, child, and infant) due to the seven combined exposure pathways. For each radionuclide, the I highest dose per unit activity release for any organ (or whole body) and age group was then selected to become the Method I site-specific dose factors.

The base case, or Method I analysis, uses the general equations methods, data, and assumptions in Regulatory Guide 1.109 (Equation C-2 for doses resulting from direct exposure to contaminated ground plane; Equation C-4 for doses a.isociated with inhalation of all radionuclides to different organs of individuals of different age groups; and C-13 for doses to organs of individuals in different age groups resulting from ingestion of radionuclides in produce, milk, meat, and leafy vegetables in Reference A). Tables B.7-2 and B.7-3 outline human consumption and environmental parameters used in the analysis. It is conservatively assumed that the critical receptor lives at the " maximum off-site atmospheric dispersion factor location" as defined in Section 7.3.

I I B.7-13

I The resulting site-sp;cific d se facters are for the maximum organ which combine the limiting age group with the highest dose factor for any I organ with each nuclide. These critical organ, critical age dose factors are given in Table B.1-12.

For any iodine, tritium, and particulate gas release, during any period, the increment in dose f rom radionuclide "i" is:

ADjc, = Q jDFGjc, (7-16) where DFG ko is the critical dose factor for radionuclide "i" and Q 4 is the activity of radionuclide "i" released in microcuries.

I Because of the assumptions about receptors, environment, and radionuclides and because of the regulations of 10CFR50 and 40CFR190, the lack of immediate restriction on plant operation, and the adherence to 10CFR20 concentrations (which limit public health consequences), a failure of Method I (i.e., the exposure of a real individual being underestimated) is improbable and the consequences of a failure are minimal.

I 7.2.7 Special Receptor Gaseous Release Dose Calculations i Technical Specification 6.9.1.6 requires that the doses to individuals involved in recreational activities within the site boundary are to be determined and reported in the annual Semiannual Effluent Report.

The gaseous dose calculations for the special receptors parallel the bases of the gaseous dose rates and doses in Sections 7.2.1 through 7.2.5.

Only the differences are presented here.

The special receptor XQs are given in Table B.7-5.

I Total Body Dose Rate From Noble Gases 7.2.7.1 Method I was derived from Regulatory Guide 1.109 as follows:

i DFB 4

(7-3) btb = 1E+06 [X/Q]T B . 7-14 I

L General Equaticn (7-3) is then multiplied by an Occupancy Facter (OF) to account for the time an individual will be at the on-site receptor locations during the year. For the Education Center, and the " Rocks", the OFs l

are:

E I $

= 0.14 Education Center 8 s u

6 The " Rocks" - = 0.0076 8 60 hr / r L

substituting y

L

[X/Q]T = 2.0E-06 sec/m3 (Education Center) b

= 5.9E-06 sec/m3 (The " Rocks")

multiplying by b 0F = 0.0014 (Education Center)

I l = 0.0076 (The " Rocks")

l gives I DFB (7-17) i g(mrem /yr) btbE = 0.0028 i l

DFB (mrem /yr) ( 7-18) btbR = 0.045 i i 4 l

I (I)Taken f rom Seabrook Station Technical Specifications (Figure 5.1-1).

I B . 7-15 I

L

r. where:

DtbE' and DtbR = Total body dose rates due to noble gases to an individual at the Education Center and the " Rocks" (recreational site), respectively.

L p Q4 = defined previously L

DFB = defined previously.

4 L

7.2.7.2 Skin Dose Rate From Noble Gases r

L Method I was derived from Equation (7-8):

e L

+ (7-8)

Q DF{

bskin = 1.11 1E+06 [X/Q]T i $

DFS 1E+06 X/Q i $

substituting

[X/Q]T = 2.0E-06 sec/m3 (Education Center) 5.9E-06 sec/m3 (The " Rocks")

X/Q = 6.7E-06 sec/m3 (Education Center) 2.3E-05 sec/m (The " Rocks")

multiplying by 0F = 0.0014 (Education Center)

[

= 0.0076 (The " Rocks")

! B .7-16

F giv;s i [2.22DF{+6.71DFS](mrem$ /yr)

, bskinE = 0.0014 L

M + M .5 g (mrem /yr)

[ bskinR = 0.0076 F

L then:

(7-I9) hi DFIE (mrem /yr) bskinE = 0.0014$

L p (7-20)

L hi DF'R i (mrem /yr) bskinR = 0.0076g where:

b skinE andbskinR = the skin dose rate due to noble gases to an individual at the Education Center and the " Rocks,"

respectively.

hg = defined previously.

DF'iE and DF'iR = the combined skin dose factors for radionuclide "i" for the Education Center, and the " Rocks",

respectively (see Table B.1-13.)

7.2.7.3 Critical Oraan Dose Rate From Iodines. Tritium and Particulates With Half-Lives Greater Than Eiaht Days

{

( Theequationsforb g are derived in the same manner as in Section 7.2.2, except that the occupancy factors are also included. Therefore:

[

DFG (7-21)

J bcoE = 0.0014 i icoE (mrem /yr)

DFG icoR (mrem /yr) (7-22) i bcoR = 0.0076 B .7-17 j

where:

b coE andbcoR = the critical organ dose rates to an individual at the L Education Center and the " Rocks", respectively.

E .

Q$ = defined previously.

DF'icoE and DF'icoR = the critical organ dose rate factors for radionuclide "i" for the Education Center and the " Rocks," respectively (see Table B.1-14.)

L 7.2.7.4 Gama Dose to Air From Noble Gases F

L Method I was derived from Equation (7-14):

Qj DF{ (7-14)

D{ir = 3.17E+04 [X/Q]Y substituting

[X/Q]T = 2.0E-06 sec/m3 (Education Center)

= 5.9E-06 sec/m3 (The " Rocks")

multiplying by 0F = 0.0014 (Education Center)

= 0.0076 (The " Rocks")

I and lE-06 pCi/Ci l

I 5

I B.7-18 l

giv5s Oi DF{ ( -23)

L. DairE = 8.88E-11 (mrad) 01 DF} (7-24)

D irR = 1.42E-09 (mrad) b where:

1

= the gamma air doses to an individual at the f D aire and D airR Education Center and the " Rocks," respectively.

h 1

Qg = total activity (yCi) released to the atmosphere via the station vents of each radionuclide "i".

DF{andDF}=definedpreviously.

7.2.7.5 Beta Dose to Air From Noble Gases Method I was derived f rom Equation (7-15):

0 01 DF ( 7-15) 0 air = 3.17E+04 X/0 i

{ substituting X/Q = 6.7E-06 sec/m3 (Education Center)

{

= 2.3E-05 sec/m3 (The " Rocks")

multiplying by 0F = 0.0014 (Education Center)

[ = 0.0076 (The " Rocks")

and 1E-06 pCi/Ci

[

B.7-19

gives Oi DFf (7-25)

OfirE = 2.97E-10 (mrad) 0 01 0 ( - 6) 0 irR = 5.54E-09 (mrad)

I where:

0 6 D

aire and DairR =Center the betaandairthe doses to anrespectively.

" Rocks," individual at the Education

! 09 = total activity (pCi) released to the atmosphere via the station vents of each radionuclide "i".

I DF and DF = defined previously.

I l

7.2.7.6 Critical Organ Dose From Iodines. Tritium and Particulates With Half-Lives Greater Than Eight Days Method I was derived in the same manner as Equation (3-18):

D cg

=

01 bco (3-18) multiplying by I

l OF = 0.0014 (Education Center)

I j = 0.0076 (The " Rocks")

and 1E-06 pCi/Ci gives 0 DFG i (7-27)

DcoE = 0.0014 icoE (mrem)

<>-ze>

,,,, = e.ee7, Ee i ,co,(m,em)

I e.7 20 I

where:

D and D = the critical organ doses of an individual at the coE coR Education Center and the " Rocks," respectively.

Q, = the total activity (uti) released to the atmosphere of radionuclide "i".

F DFG and DFG jcop = the critical organ dose factors (mrem /pCi) for the 1 oE Education Center and the " Rocks," respectively F for each radionuclide "i". The factors represent l

the age group and organ with the largest dose f actor (see Table B.1-14).

I The special receptor equations can be applied under the following f conditions (otherwise, justify Method I or consider Method II):

b

1. Normal operations (nonemergency event).

2. Applicable radionuclide releases via the station vents to the atmosphere.

If Method I cannot be applied, or if the Method I dose exceeds this limit, or if a more refined calculation is required, then Method 11 may be applied.

l l

1 l

1 i

B.7-21

MNWMNM U7 _ f~W U W~

TABLE B.7-2 Environmental Parameters for Gaseous Effluents at Seabrook Station (Derived from Reference A)*

Vegetables Cow Milk Goat Milk Meat Leafy Pasture Stored Pasture Stored Pasture Stored Variable Stored

2. 2. 0.75 2. 0.75 2. 0.75 2.

YV Agricultural (Kg/M2 )

Productivity 240. 240. 240. 240. 240. 240. 240.

l P Soil Surface Density (KG/M2) 240.

48. 48. 48. 48. 480. 480.

l T Transport Time to User (HRS)

~131400. 131400. 131400. 131400. 131400. 131400. 131400.

l TB Soil Exposure Time (HRS) 131400.

1440. 720. 720. 720. 720. 720. 720.

l TF Crop Exposure Time (HRS) 1440.

Tto Plume

24. O. 2160. O. 2160. O. 2160.

TH Holdup After Harvest (HRS) 1440.

50. 50. 6. 6. 50. 50. g QF Animals Daily Feed (KG/ DAY) 0.50 0.50 0.50 i FP Fraction of Year on Pasture

1. 1. 1.

FS Fraction Pasture when on Pasture FG Fraction of Stored 0.76 Veg. Grown in Garden FL Fraction of Leafy 1.0 Veg. Grown in Garden FI Fraction Elemental Iodine = 0.5 H Absolute (gm/M3 )

Humidity = 8.00**

• Regulatory Guide 1.109

• • Default value from NRC "GASPAR" Dose Code; K. F. Eckerman, revised December 2, 1975 s

I TABLE B.7-3 Usage Factors for Various Gaseous Pathways at Seabrook Station (from Reference A, Table E-5)*

I Maximum Receptor:

I Group Age Vegetables Leafy Vegetables Milk Meat Inhalation (kg/yr) (kg/yr) (1/yr) (kg/yr) (m3 /yr)

Adult 520.00 64.00 310.00 110.00 8000.00 Teen 630.00 42.00 400.00 65.00 8000.00 Child $20.00 26.00 330.00 41.00 3700.00 Infant 0.00 0.00 330.00 0.00 1400.00 I

The " Rocks" and Education Center:

I Age Leafy Group Vegetables Vegetables Milk Meat Inhalation I (kg/yr) (kg/yr) (1/yr) (kg/yr) (m /yr)

Adult 0.00 0.00 0.00 0.00 8000.0 Teen 0.00 0.00 0.00 0.00 8000.0 Child 0.00 0.00 0.00 0.00 3700.0 I Infant 0.00 0.00 0.00 0.00 1400.0 I

I I

• Regulatory Guide 1.109 I B.7-23 I

s 7.3 Receptor Points and Average Atmospheric Dispersion Factors for Important Exposure Pathways L

The gaseous effluent dose equations (Method I) have been simplified by c assuming an individual whose behavior and living habits inevitably lead to a L- higher dose than anyone else. The folicwing exposure pathways to gaseous effluents listed in Regulatory Guide 1.109 (Reference A) have been considered:

L

1. Direct exposure to contaminated air; P

L

2. Direct exposure to contaminated ground;

3. Inhalation of air; -

4. Ingestion of vegetables;

5. Ingestion of cow's and goat's milk; and l

l 6. Ingestion of meat.

I l

Section 7.3.1 details the selection of important off-site and on-site locations and receptors. Section 7.3.2 describes the atmospheric model used to convert meteorological data into atmospheric dispersion factors. Sectior.

7.3.3 presents the maximum atmospheric dispersion factors calculated at each of the off-site receptor locations.

l 7.3.1 Receptor Locations l

The most limiting site boundary location in which individuals are, or likely to be located as a place of residence was assumed to be the receptor for all the gaseous pathways considered. This provides a conservative estimate of the dose to an individual from existing and potential gaseous pathways for the Method I analysis.

This point is the N sector, 914 meters f rom the center of the reactor units.

B.7-24

Two other locations (on-site) were analyzed for direct ground plane exposure and inhalation only. They are the " Rocks" (recreational site) and L

the Education Center shown on Figure 5.1-1 of the Technical Specifications.

F L

E u

[

[

[

[

[

[

[

[

[

( B.7-25 he mmmm ensm um

t 7.3.2 Seabrook Station Atmospheric Dispersion Model The time average atmospheric dispersion factors are computed for u

routine (long-term) ground level releases using the AEOLUS Computer Code F (Reference B). AE0LUS is based, in part, on the straight-line airflow model L discussed in Regulatory Guide 1.111 (Reference C).

I AEOLUS produces the following average atmospheric dispersion factors for each location:

1. Undepleted X/Q dispersion factors for evaluating ground level concentrations of noble gases;

(

2. Depleted X/4 dispersion factors for evaluating ground level

{ concentrations of iodines and particulates; f

I

3. Gamma X/Q dispersion factors for evaluating gamma dose rates from a sector averaged finite noble gas cloud (multiple energy undepleted l source); and I

l 4. D/Q deposition factors for evaluating dry deposition of elemental radioiodines and other particulates.

l Gamma dose rate is calculated throughout this ODCM using the finite cloud model presented in " Meteorology and Atomic Energy - 1968" (Reference E, Section 7-5.2.5. That model is implemented through the definition of an I effective gamma atmospheric dispersion factor, [X/Q ] (Reference B, Section T

6), and the replacement of X/Q in infinite cloud dose equations by the T

[X/Q ].

7.3.3 Long-Term Average Atmospheric Dispersion Factors for Receptors Actual measured meteorological data for the two-year period, April-1979 through June-1981, were analyzed to determine the locations of the maximum off-site average atmospheric dispersion factors. Each dose and dose rate calculation incorporates the maximum applicable off-site long-term average atmospheric dispersion factor. The values used and their locations are summarized in Tables B.7-4 and B.7-5.

B.7-26

. . _ . .. 1

m nn _m r~~\ v v m M T_ -

CR O F- L-. I 3 Im - T~U- - L J 7

-TABLE B.7-4 Seabrcok Station Dilution Factors

• Dose to Critical Dose Rate to Individual Dose to Air Orean Total Body Skin Critical Organ Gamma Beta Thyroid X/Q depleted (5'C) - -

1.3E-06 - -

1.3E-06 m

X/Q undepleted (sec) -

1.4E-06 - -

1.4E-06 -

m 3.lE-09 - -

3.lE-09 D/0(h)

Y X/QT (5'C) 6.2E-07 6.2E-07 -

6.2E-07 - -

m

• North site boundary, 916 meters from Containment Building l

l r

m W W M M M M M M M M M M M M em W W TABLE B.7-5 Seabrook Station Dilution Factors f or Special (On-Site) Receptors  :

Dose to Critical Dose Rate to Individual Dose to Air Organ Total Body Skin Critical Organ Gansna Beta Thyroid Education Center:

(WSW - 335 meters)

X/Q depleted (S'3) -

~

6.2E-06 - -

6.2E-06 m

c 6.7E-06 X/Q undepleted (**3 ) -

6.7E-06 - - -

m os y D/0 ( )

1.lE-08 - - -

2.0E-06 2.0E-06 -

2.0E-06 - -

X/QT m(5'3)

The " Rocks" (ENE - 318 meters)

X/Q depleted (*'3 ) - -

2.1E-05 - -

2.1E-05 m

X/Qundepleted(S$C) -

2.3E-05 - -

2.3E-05 -

m D/Q ( )

5.0E-08 - - -

m X/QT (se 5.9E-06 5.9E-06 5.9E-06 - -

3 )

m

I 8.0 BASES FOR LIOUID AND GASEOUS MONITOR SETPOINTS I 8.1 Basis for the Liauid Waste Test Tank Monitor Setooint The liquid waste test tank monitor setpoint must ensure that Specification 3.3.3.9 is not exceeded for the appropriate in-plant pathways.

The liquid waste test tank monitor is placed upstream of the major source of dilution flow.

I The derivation of Equation 5-1 begins with the general equation for the response of a radiation monitor:

R =

C,g S jg (8-1)

I (cps) =

( ) (co ml) where:

R = Response of the monitor (cps)

I S), = Detector counting ef ficiency for radionuclide "i" (cps /(pCi/ml))

= Activity concentration of radionuclide "i" in mixture at I

C g

the monitor (pCi/ml)

The detector calibration procedure for the liquid waste test tank monitor at Seabrook Station establishes a counting efficiency by use of a known calibration source standard and a linearity response check. Therefore, in Equation 8-1 one may substitute 3) for Sjg, where Sj is the detector counting efficiency determined from the calibration procedure. Therefore, Equation 8-1 becomes:

R = S j C g (8-2) i (cps) = (C *I) ( )

'I B.8-1 I

L The MPC for a given radionuclide must not be exceeded at the point of discharge. When a mixture of radionuclides is present, 10CFR20 specifies that p the concentration at the point of discharge shall be limited as follows:

L

$1 (8-3)

PC

• - 1 i e uti-ml ml pCi)

L I where:

L r C = Activity concentration of radionuclide "i" in the mixture at di L the point of discharge (uCi/ml)

MPC, = MPC for radionuclide "i" from 10CFR20, Appendix B Table II, Column 2 (vCi/ml)

L The activity concentration of radionuclide "i" at the point of discharge is related to the activity concentration of radionuclide "i" at the monitor as

{ follows:

F C =

C ,g (8-4) di L uci uti (ml ) , ml ) gjLm) gpm where:

F L C di = Activity concentration of radionuclide "i" in the mixture at the point of discharge (pCi/ml)

F, = Flow rate past monitor (gpm)

F = Flow rate out of discharge tunnel (gpm) d B.8-2

I Substituting the right half of Equation 8-4 for Cdi in Equation 8-3 and solving for F d/F , yields the minimum dilution factor needed to comply with Equation 8-3:

I DF 2 (8-5) min 5 I

q uCi-m1 Ispm) Imi yCi) gpm I where:

I F d

= rate out of d scharge tunnel (gpm)

I F, = Flow rate past monitor (gpm)

Cg = Activity concentration of radionuclide "i" in mixture at the monitor (uti/ml)

I MPC

= MPC for radionuclide "1" from 10CFR20, Appendix B, Table II, Column 2 (yCi/ml)

If F d/F ,is less than DFmin, then the tank may not be discharged until either Fdor F, or both are adjusted such that:

F d (8-5)

F 1 DF min m

I (gpm)

I Usually F d m s gmater Dan DF min .e., ee s m re mudon han necessary to comply with Equation 8-3). The response of the liquid waste test tank monitor at the setpoint is therefore:

R setpoint " Il D in I uti ml *I)() Icos-m1) pCi IuCi) m1 I B.8-3 I

where f) is equal to the fraction of the total contribution of MPC at the discharge point to the environment to be associated with the test tank effluent pathway, such that the total sum of the fractions for the three liquid discharge pathways is equal to or less than one (f) +f2*I3 I II

• The monitoring system is designed to incorporate the detector efficiency, S), into its software. This results in an automatic readout in pCi/cc or pCi/mi for the monitor response. Since this procedure for converting cps to pCi/ml is inherently done by the system software, the monitor response setpoint can be calculated in terms of the total waste test tank activity concentration in pCi/ml determined by the laboratory analysis.

Therefore, the setpoint calculation for the liquid waste test tank is:

R setpoint "f l C ,3 U-U DF min i

(" ) ()() (" ))

8.2 Basis for the Plant Vent Wide Range Gas Monitor Setpoints The setpoints of the plant vent wide range gas monitors must ensure that Technical Specification 3.11.2.1.a is not exceeded. Sections 3.4 and 3.5 show that Equations 3-3 and 3-4 are acceptable methods for determining compliance with that Technical Specification. Which equation (i.e., dose to total body or skin) is more limiting depends on the noble gas mixture.

Therefore, each equation must be considered separately. The derivations of Equations 5-5 and 5-6 begin with the general equation for the response R of a radiation monitor:

R = S gg C

g (8-7)

(cpm) = (CD* C I I" I cm B.8-4 l

l

I where:

I R = Response of the instrutcent (cpm) 3 Sg g = Detector counting ef ficiency for noble gas "i" (cpm /(yCi/cm ))

C ,$ = Activity concentration of noble gas "i" in the mixture at the noble gas activity monitor (pCi/cm3)

I C g , the activity concentration of noble gas "1" at the noble gas activity monitor, may be expressed in terms of gQ by dividing by F, the appropriate flow I rate. In the case of the plant vent noble gas activity monitors the appropriate flow rate is the plant vent flow rate.

C g = h$ h (8-8)

(uti)3

,guci) sec (sec) cm c ,3 where:

h = The release rate of noble gas "i" in the mixture, for each noble 4

gas listed in Table B.1-10.

F = Appropriate flow rate (cm /sec)

Substituting the right half of Equation 8-8 into Equation 8-7 for Cg yields:

R = S gg h$ h (8-9)

(cpm) (CDm-cm ) ( ) (sec) cm$

As in the case before, for the liquid waste test tank monitor, the plant vent wide range gas monitor establishes the detector counting efficiency by use of a calibration source. Therefore, Sgcan be substituted for S gg I B.8-5 I 1

L in Equatien 8-9, where S is the detector ccunting efficiency determined

[ g from the calibration procedure. Therefore, Equation 8-9 becomes:

T 1 E hj (8-10)

R = S g p L, x

3 (cpm) = ('D* C )I'I I"e I cm The total body dose rate due to noble gases is determined with Equation 3-3:

b = 0.62 hg DFB (3-3) tb $

3 mrem-m Imrem) yr " IDCi-sec) IuCi) sec I pci-yr I Ci +3 where:

l b = total body dose rate (mrem /yr) tb f 3 0.62 = (1.0E+06) x (6.2E-07) (pCi-sec/pti-m )

l 1E+06 = number.of pCi per pCi (pCi/pci) 6.2E-07 = [X/Q]T, maximum annual average gamma atmospheric 3

dispersion factor (sec/m )

l I h4 = As defined above.

I DFB = total body dose factor (see Table B.1-10) 9 3

(mrem-m /pci-yr)

^'""'"'""~'"'''"""'"'""'"'"'"'"'"

E E B.8-6

u L DFB c

Q4 - Q DFB q (8-11) j 5 4

3

[ mrem-m pCi-yr 3

uti) sec (uci) sec mrem-m pCi-yr )

Solving Equation 8-11 for DFB c yields:

h3 0FB q DFB = (5-7)

[ b3 i

Technical Specification 3.ll.2.1.a limits the dose rate to the total body f rom noble gases at any location at or beyond tree site boundary to 500 mrem /yr. Bysettingb equal to 500 mrem /yr and substituting DFB for DFB $

[ tb in Equation 3-5, one may solve for [ Q$ at the limiting whole body noble gas I

{ dose rate:

h= 4 806 DFB c

3 Iuti) sec "Imrem-uCi-m yr-pCi-sec ) IDCi-vr3) l mrem-m l Substituting this result for [ h in gEquation 8-11 yields Rtb, the response of the monitor at the limiting noble gas total body dose rate:

R tb

= 806 S g h DFB c

3 _c ,3 c

(cpm) = (mrem-uCi-m -pCi-su I I ,,pCi )Isec) 3 IDCi-vr3) cm mrem-m The skin dose rate due to noble gases is determined with Equation 3-6:

skin i DFj (3-6) mrem) , uCi) yr sec mrem-sec) pCi-yr I B.8-7 I

.. )

L.

I where:

w Dskin = Skin dose rate (mrem /yr)

E Qg = As defined above.

DFj

= Combined skin dose factor (see Table B.1-10) (mrem-sec/yCi-yr)

A composite combined skin dose factor, DF', may be defined such that:

I l DF' h4 = h$ DFj (8-14) i i j Imrem-sec) pCi-yr IuCi) sec (uCi) sec (mrem-sec) pCi-yr I Solving Equation 8-14 for DF' yields:

hj DFj DF' = ,

(5-8)

O l i i I Technical Specification 3.ll.2.1.a limits the dose rate to the skin l

f rom roble gases at any location at or beycad the site boundary to 3,000 mrem /yr.

By setting D skin equal to 3,000 mrem /yr and substituting DF' for DFj in Equation 3-6 one may solve for [ Qg at the limiting skin noble gas dose rate:

I 4 h = 3,000 DF, (8-15) 3 i c j uCi-vr uCi) sec mrem) yr mrem-sec)

Substituting this result for [ h in $Equation 8-11 yields Rskin, the response of the monitor at the limiting noble i gas skin dose rate:

3.

I I B.8-8 I

R skin 3,000 S g h DF'

(-

3 P

(cpm) (mrem) Icom-cm pCi )c Isec) I uti-vr )

mrem-sec yr ,3 I As with the liquid monitoring system, the gaseous monitoring system is also designed to incorporate the detector efficiency, S , into its software. This results in an automatic readout in pCi/cm for the monitor response. Therefore, Equations 8-13 and 8-16 become:

R tb

= 806 h DFB c

3 sec yr-pCi-sec ) I ,3 IIDCi-vr IuCi "Imrem-uCi-m c,3 mrem-m 3) i l

R skin

= 3000 fh (5-6)

(DCi) , (mrem) DCi-vr sec) mrem-sec)

( ,3 yr ,3 8.3 Basis for PCCW Head Tank Rate-of-Change Alarm Setpoint The PCCW head tank rate-of-change alarm will work in conjunction with the PCCW radiation monitor to alert the operator in the Main Control Room of a leak to the Service Water System f rom the PCCW System. For the rate-of-change alarm, a setpoint based on detection of an activity level of 10- pCi/cc in the discharge of the Service Water System has been selected. This activity level was chosen because it is the minimum detectable level of a service water monitor if such a monitor were installed. The use of rate-of-change alarm with information obtained from the liquid sampling and analysis commitments described in Table A.3-1 of Part A ensure that potential releases f rom the Service Water System are known. Sampling and analysis requirements for the Service Water System extend over various operating ranges with increased I sampling and analysis at times when leakage from the PCCW to the service water is occurring and/or the activity level in the PCCW is high.

B.8-9 I

w REFERENCES r A. Regulatory Guide 1.109, " Calculation of Annual Doses to Man From Routine L Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I", U.S. Nuclear Regulatory Commission, Revision 1, October 1977.

D. Hamawi, J. N., "AEOLUS - A Computer Code for Determining Hourly and Long-Term Atmospheric Dispersion of Power Plant Effluents and for r Computing Statistical Distributions of Dose Intensity From Accidental L Releases". Yankee Atomic Electric Company, Technical Report, YAEC-1120, January 1977.

I C. Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases From Light-Water Cooled Reactors", U.S. Nuclear Regulatory Commission, March 1976.

E' L D. National Bureau of Standards, " Maximum Permissible Body Burdens and Maximum Pennissible Concentrations of Radionuclides in Air and in Water for Occupational Exposure", H...dbook 69, June 5, 1959.

E. Slade D. H., " Meteorology and Atomic Energy - 1968", USAEC, July 1968.

[ F. Seabrook Station Technical Specifications.

E

[

E E

F R-1

- __- .___ _ _ - . _ - _ _ _ _ .