Text

Corrected Rev 3 to Offsite Dose Calculation Manual
	ML20195D739
Person / Time
Site:	Seabrook
Issue date:	08/26/1986
From:	Leland W, Moody D, George Thomas; PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	;
Shared Package
ML20195D729	List: ML20195D725; ML20195D739;
References
	PROC-860826, NUDOCS 8806230182
	Download: ML20195D739 (146)
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OFFSITE DOSE CALCULATION MAXUAL Searoot swim New Hampshire Yankee FP 28 ass 8888jjp

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OFFSITE DOSE CALCULATION MANUAL (ODCM)

PREPARED BY:

M. STRUM, YANKEE ATOMIC ELECTRIC COMPANY b

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SUBMITTED BY W.B.LELAND, CHEMISTRY AND HEALTH PHYSICS MANAGER DATE SORC REVIEW COMPLETED DURING MEETING f/,, -T/) [

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APPROVED BY D.E. MOODY, STATI % MANAGER DATE

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APPROVED BY

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G.S. THOMAS, VICE PRESIDENT - NUCLEAR PRODUCTION DATE REVISION 3 -- EFFECTIVE: 8/26/86 DATE OF LAST PERIODIC REVIEW:

8/19/86 U

DATE NEXT PERIODIC REVIEW DUE: 8/19/88

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DISCLAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company The use of information contained in this document by anyone other

(' Yankee").

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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ABSTRACT l

The Seabrook Station 00CM (Off-Site Oose 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 environmentcl radiological monitoring program requirements (Part A); and (2) approved methods to determine ef fluent monitor setpoint values and estimates of doses and radionuclide concentrations occurring beyond the boundaries of the station resulting from normal station operation (Part B).

The sampling and analysis programs in Part A provide the inputs for the models of Part B in order to calculate off-site doses and radionuclide concentrations necessary to determine compliance with the dose and The concentration requirements of the Station Technical Specification 3/4.11.

radiological environmental monitoring program required by Technical Specification 3/4.12 and outlined within this mancal provides the means to determine that measurable concentrations of radioactive materials released as a result of the operation of Seabrook Station are not significantly higher than expected.

O

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SEABROOK STATION OFF-SITE DOSE CALCULATION HANUAL NEW HAMPSHIRE YANKEE JULY 19,1986

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TABLE OF CONTENTS (Continued)

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(

Page B. 3-1 3.0 0FF-SITE DOSE CALCULATION METN005...........................

B.3-2 3.1 Introductory Concepts..................................

3.2 Method to Calculate Total Body Dose from Liquid B.3-4 Releases...............................................

3.3 Method to Calculate Maximum Organ Dose from Liquid B.3-6 Releases...............................................

3.4 Method to Calculate the Total Body Dose Rate from 8.3-8 Noble Gases............................................

3.5 Method to Calculate the Skin Oose Rate from Noble 8.3-10 Gases..................................................

3.6 Method to Calculate the Critical Organ Oose Rate from Iodines, Trititi and Particulates with Tl/2 B. 3 -12 Greater Than 8 0ays....................................

3.7 Method to Calculate the Gamma Air Dose from Noble B.3-14 Gases..................................................

i 3.8 Method to Calculate the Beta Air Oose from Noble B. 3 -16 Gases..................................................

3.9 Method to Calculate the Critical Organ Oose from B.3-18

)

Tritium, Iodines and Particulates......................

3.10 Method to Calculate Direct Oose f rom Plant j

B.3-20 0peration..............................................

3.11 D o s e P roj e c t i o n s.......................................

.B.3-21

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8. 4 -1 4.0 ENVIRONMENTAL MONITORING PR0 GRAM............................

8.5-1 5.0 SETPOINT DETERMINATIONS.....................................

5.1 Liquid Effluent Instrumentation Setpoints..............

8.5-2 5.2 Gaseous Effluent Instrumentation Setpoints..............

8.5-8 6.0 L10010 AND GASE0US EFFLUENT STREAMS, RA01AT10N MONITORS AND RA0 WASTE TREATMENT SYSTEMS..............................

B.6-1 7.0 BASES FOR DOSE CALCULATION METN005..........................

8. 7 -1 8.0 BASES FOR LIQU10 AND GASEOUS MONITOR SETP0lNTS..............

B.8-1 R -1 4

REFERENCES..................................................

-vii-

TABLE OF CONTENTS Paae 11 REVISION RECOR0..................................................

iii LIST OF EFFECTIVE PAGES..........................................

iv DISCLAIMER OF RESPONSIBILITY.....................................

v ABSTRACT.........................................................

viii LIST OF FIGURES..................................................

ix LIST OF TABLES...................................................

PART A: RADIOLOGICAL EFFLUENT MONITORING PROGRAMS Section A.1 -1 1.0 Introduction................................................

A. 2 -1 2.0 Responsibilities for Part A.................................

A.3-1 3.0 Liquid Effluent Sampling and Analysis Program...............

4.0 Gaseous Effluent Sampling and Analysis Program..............

A.4-1 A.5-1 5.0 Radiological Environmental Monitoring.......................

A.5-1 5.1 Sampling and Analys i s Program..........................

A.5-2 5.2 Land Use Census........................................

PART 8: RADIOLOGICAL CALCULATIONAL METH005 AND PARAMETERS...........

B.1 -1 Sgetion 8.1 -1 1.0 INTR 00VCTION................................................

1.1 Responsibilities for Part 2............................

B.1 -1 1.2 Sumary of Methods, Oose Factors, Limits, Constants, Variables and Definitions...................

B.1 -2 2.0 METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS..........

3. 2 -1 ENG NG B. 2 -1 and C Method to Determine F 2.1 MethodtoDetermineR$dionuclide.......................

Concentration 2.2 for Each Liquid Effluent Pathway.......................

B.2-2 2.2.1 Waste Test Tanks Pathway........................

B.2-2 2.2.2 Turbine Building Sump Pathway...................

B.2-3 2.2.3 Steam Generator Blowdown Flash Tank Pathway.....

B.2-3

_vi.

l I

LIST OF TABLES P_A_RT A A

Title Pace Number A.3-2 A. 3-1 Radioactive Liquid Waste Sampling and Analysis Program A.4-2 A 4-1 Radioactive Gaseous Waste Sampling and Analysis Program A.5-3 A. 5-1 Radiological Environmental Monitoring Program Detection Capabilities for Environmental Sample Analysis A.5-7 A.5-2 A.5-3 Reporting Levels for Radioactivity Concentration in A. 5 -10 Environmental Samples PART 8 Title Page Number B.1 -1 Sumary of Radiological Effluent Technical Specifications B.1-3 and Implementing Equations B.1 -2 Sumary of Method I to Calculate Unrestricted Area B.1-6 Liquid Concentrations O

B.1 -3 Sumary of Method I to Calculate Of f-Site Ooses f rom B.1 -7 V

Liquid Releases B.1-8 B.1 -4 Sumary of Method I to Calculate Dose Rates B.1-5 Sumary of Method I to Calculate Ooses to Air f rom 8.1 -9 Noble Gases B.1-6 Sumary of Method I to Calculate Oose to an Individual f rom Tritium, Iodine and Particulates B.1-10 B.1 -7 Sumary of Methods for Setpoint Determinations B.1 -11 B.1 -12 B.1-8 Sumary of Variables B.1 -16 B.1 -9 Definition of Terms B.1 -10 Dose Factors Specific for Seabrook Station for Noble Gas B.1 -17 Releases B.1 -11 Oose Factors Specific for Seabrook Station for Liquid B.1 -18 l

Releases O

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LfST OF FfGURES Title Pace Number.

Radiological Environmental Monitoring Locations Within B.4-1 B.4-4 4 km of Seabrook Station 8.4-2 Radiological Environmental Monitoring Locations Between 4 km and 12 km from Seabrook Station B.4-5 B.4-3 Radiological Environmental Monitoring Locations Outside 8.4-6 12 km of Seabrook Station B.4-4 Direct Ra11ation Monitoring Locations Within 4 km B.4-7 of Seabrook Station B.4-5 Direct Radiation Monitoring sacations Between 4 km B.4-8 and 12 km from Seabrook Station 8.4-6 Direct Radiation Monitoring Locations Outside 12 km B.4-9 of Seabrook Station B.6-1 Liquid Effluent Streams, Radiation Monitors and Radwaste B.6-9 Treatment System at Seabrook Station 8.6-2 Gaseous Effluent Streams, Radiation Monitors and Radwaste B.6-10 Treatment System at Seabrook Station 9

e

,iii-

LieT OF TABLES (Continued)

Number Title g

B.1-12 Dose and Dose Rate Factors Specific for Seabrook Station for Tritium, Iodine and Particulate Releases B.1-19 t,,1 -13 Combined Skin Oose Factors Specific for Seabrook Station Special Receptors for Noble Gases B.1-20 B.1-14 Dose and Dose Rate Factors Specific for Seabrook Station.

Special Receptors for Iodines, Tritium and Particulstes B.1-21 B.4-1 Radiological Environmental Monitoring Stations B.4-2 B.7.1-1 Usage Factors for Various Liquid Pathways at B.7-4 Seabrook Station B.7.2-1 Environmental Param9ters for Gaseous Ef f' ts at Seabrook Station B.7-22 B.7.2-2 Usage Factors for Various Gaseous Pathways at Seabrook Station B.7-23 B.7.3-1 Seabrook Station Dilution Factors B.7-27 B.7.3-2 Seabrook Station Dilution Factors for Special Receptors B.7-28 W

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LIST OF EFFECTIVE PAGES Page Rev.

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O Page 1 of 2 ODCM Rev. 3 1

LIST OF EFr.".CTIVE PAGES Page Rev.

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B.5-1 6-30-86 B.8-4 6-30-86 B.5-2 6-30-86 B.8-5 6-30-86 B.5-3 6-30-86 B.8-6 6-30-86 B.5-4 6-30-86 B.8-7 6-30-86 B.5-5 6-30-86 B.8-8 6-30-86 B.5-6 6-30-86 B.8-9 6-30-86 B.5-7 6-30-85 B.5-8 6-30-86 R-1 6-30-86 B.5-9 6-30-86 B.5-10 6-30-86 B.5-il 6-30-86 B.5-12 6 30-86 B.5-13 6-30-86 B.6-1 6-30-86 B.6-2 6-30-86 B.6-3 6-30-86 B.7-1 6-30-86 B.7-2 6-30-86 B.7-3 6-30-86 B.7-4 6-30-86 B.7-5 6-30-86 B.7-6 6-30-86 B.7-7 6-30-86 B.7-8 6-30-86 B.7-9 6-30-86 B.7-10 6-30-86 B.7-11 6-30-86 B.7-12 6-30-86 B.7-13 6-30-86 B.7-14 6-30-86 B.7-15 6-30-86 B.7-16 6-30-86 B.?-17 6-30-86 B.7-18 6-30-86 B.7-19 6'30-86 B.7-20 6-30-86 B.7-21 6-30-86 B.7-22 6-30-86 3.7-23 6-30-86 B.7-24 6-30-86 B.7-25 6-30-86 B.7-26 6-30-86 B.7-27 6-30-86 B.7-28 6-30-86 B.8-1 6-30-86 B.8-2 6-30-86 B.8-3 6-30-86 O

Page 2 of 2 ODCM Rev. 3

Part A RADIOLOGICAL EFFLUENT MONITORING PROGRAMS

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1.0 IN'60 DUCTION The purpose of Part A of the ODCM (Of f-Site Dose Calculation Manual) is to describe the sampling and analysis programs conducted by the Station which provide input to the models in Part B for calculating liquid and gaseous 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.ll.

The Radiological Environmental Monitoring Program required as a minimum to be conducted (per Technical Specification 3/4.12) is described in Part A, with the identification of current locatione of sampling stations being utilized to meet the program requirements listed in Part B.

The information obtained f rom the conduct of the Radiological Environmental Monitoring Program provides data on measurable levels of radiation and radioactive materials in the environ-ment 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 confira

]

that measurable concentrations of radioactive materials released as a result of Seabrook Station operations are not significantly higher than expected based on the dose models in Part B.

1.

Pending review and resolution of the staff's concerns regarding applied dispersion parameters for Method I gaseous dose calculations, the

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limits shall be reduced by a f actor of 10.

This restriction will

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ensure compliance with 10CFR20 Appendix B and 10CFR50 Appendix 1.

Upon satisfactory review and resolution of the conceres regarding dispersion parameters with the staff, this restriction will be deleted from the O DCM.

2.

Method II gaseous and liquid dose calculations shall not be implemented until additional information is provided describing in sufficient detail the methodology used in Method II gaseous and liquid dose calcu-lations.

Upon review, approval and incorporation of this methodology detail in Part B of the ODCM, this restriction will be deleted from the O DCM.

G A.1-1

2.0 RESPONSIBTLITIES FOR PART A All changes to Part A of the 00CM shall be reviewed and approved by the Station Operations Review Comittee (SORC) and the Nuclear Regulatory Comission prior to implementation.

It shall be the responsibility of the Station Manage.' to ensure that the ODCM is used in the performance of the surveillance requirements and administrative controls of the appropriate portions of the Technical Specifications, O

A.2-1

3.0 LIOUID EFFLUENT SAMPLING AND ANALYSIS PROGRAM Radioactive liquid wastes shall be sampleo and analyzed in accordance The results of with the program specified in Table. A.3-1 for Seabrook Unit 1.

the radioactive analysis 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 system are maintained within the limits of Technical Specification 3.11.1.1 for Unit 1.

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.

O 8

A.3-1 1

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TABLE A.3-1 Radioactive Liquid Waste Sampling and Analysis Program Lower Limit Minimum Type of of Detection Liquid Release Sampling Analysis Activity (LLD) (1)

Type Frequency Frequency Analysis (uci/ml)

A.

Liquid P

P Principal Gama Radwaste Each Batch Each Batch Emitters (3) 5x10-7 Test Tanks I-131 1x10-6 (Batch Release)(2)

P M

Dissolved and 1 x10-5 Entrained Gases One Batch /M (Gamma Emitters)

T 1 x10-5 u

M(4)

H-3 b

P Each Batch Composite Gross Alpha lx10-7 P

Q(4)

Sr-89, Sr-90 5x10-8 Each Batch Cornposite Fe-55 1x10-6 B.

Turbine Building W

W Principal Gamma Emitters (3) 5x10-7

~ Sump Effluent (8)

Grab Sample 1-131 1x10-6 (Continuous Dissolved and Release (5)

W M

Entrained Gases lx10-5 Grab Sample (Gamma Emitters)

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TABLE A.3-1 Radioactive Liquid Waste Sampling and Analysis Program (continued)

Lower Limit Minimum Type of of Detection Liqrid Release Sampling Analysis Activity (LLD) (1).

Type Frequency Frequency Analysis (uci/ml)

D.

Service Water (7)

W W

Principal Gamma Emitters (3) 5x10-7 Grab Sample I-131 lx10-6 W

H Dissolved and Entrained Gases l x10-5 Grab Sample (Gamma Emitters) f lx10-5 W

M H-3 Grab Sample Gross Alpha lx10-7

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TABLE A.3-1 f

Notations (1)The LLO is defined, for purposes.of these specifications, as the smallest concentration of radioa:tive material in a sample that will yield a net count, above system background, 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 l

separation:

4.66 s b LLD =

6 E x V x 2.22 x 10 x Y x exp (-Ast)

Where:

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

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

j E = the counting efficiency (counts per disintegra' tion),

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

2.22 x 10-6 = the number of disintegrations per minute per microcurie, H

Y = the fractional radiochemical yield, when applicable, I

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'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 i oriori (before the fact) limit representing the capability of a measurement system and not as an a Dosteriori (af ter 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.

O A.3-5

TABLE A.3-1 Notations (Continued)

(3)The principal gamma emitters for which the LLO specification applies include the following radionuclides:

Mn-54, Fe-59, Co-58, Co-60, Zn-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 ganna 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 levels 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 810wdown is directed to the discharge transition structure.

(7) Principal ganna 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 ganna emitters exceeds the LLO.

The following are additional sampling and analysis requirements:

PCCW sampled and analyzed weekly for principal gaana emitters, a.

b.

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

With the PCCW System radiation monitor inoperable, sample PCCW and SWS c.

daily f or principal gamma emitters.

Wi h a confirmed PCCW/SWS leak and PCCW activity in excess of 1x10-4 d.

uC/cc, sample SWS every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for principal ganna emitters, The setpoint on the PCCW head tank liquid rate-of-change alarm will be e.

set to ensure that its sensitivity to detect a PCCW/SWS leak is equal located in the to or greater than that of an SWS radiation monitor unit's combined SWS discharge, with an LLO of 1x10-b uC/cc.

If this sensitivity cannot be achieved, the SWS will be sampled once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(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 f or Iodine-131 and principal gamma emitters prior to release.

A.3-6 i

4.0 GASEOUS EFFLUENT SAMPLING AND ANALYSIS PROGRAM O'

Radioactive gaseous wastes shall be sampled and analyzed. accordance The results of with the program specified in Table A.4-1 for Seabrook Unit 1.

the radioactive analyses shall be used as appropriate with the methodologies of Part 8 of the 00CM 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 for Unit 1.

Radioactive effluent information for gaseous wastes obtained from this sampling and 0.*alysis 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 2.2, 3/4.11.2.3, 3/4.11.2.4, and 3/4.11.4.

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l y

p u

u u

u p

gc m

o o

m nn

)

a u

u u

u a

i e 4 S n

n n

n S

l u

(

i i

pq

) b t

t t

t

) b me 3 a n

n n

n 7 a ar

(

r o

o o

o

(

r SF MG C

C C

C MG ts ru ia Ah x t

rE n

e e

sl V

na se ev O

use t

d o oa p n

nm ee y a

oe sl T l

CR ae P

GR 2

1

. [

s s

L

't TABLE A.4-1 Radioactive Gaseous Waste Sampling ar.d Analysis Program (continued)

Minimum Type of Lower Limit Release Sampling Analysis Activity of Detection (l)

Gaseous Type Frequency Frequency Analysis (LLD) (uCi/cc) 3.

Gland Steam Continuous W

Principal Gamma Emitters (2) 1x10-Il Particulate Packing Exhauster Sample 1x10-12 Continuous W

I-131 Charcoal Sample Continuous M

Gross Alpha 1x10-Il w

2, Composite Particulate Sample Continuous Q

Sr-89, Sr-90 1x10-Il Composite Particulate Sample 4.

Containment P(3)

P Principal Gamma Emitters (2) 1x10-4 Purge Each Purge Grab Each Purge Sample H-3 (oxide) 1x10-6 l

TABLE A.4-1 Notations (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 count, above system background, 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 LLO =

6 E x V x 2.22 x 10 x Y x exp (-AAt)

Where:

LLO = the 'a priori' lower limit of detection (microcurie per unit mass or volume),

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

E = the counting efficiency (counts per disintegration),

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

2.22 x 10-6 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, h = 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.

l It should be recognized that the LLO is defined as an a oriori (before the fact) limit representing the capability of a measurement system and not as an a. Dosteriori (af ter the f act) limit for a particular measurement.

A.4-4 j

TABLE A.4-1 Oh Notations (Continued)

(2)The principal cama emitters for which the LLD specifications applies include the following radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-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 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 may be reported as "not detected." Values determined to be below detectable levels are not used in dose calculations.

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

For containment purge, requirements apply only when purge is in operation.

(4)1ritium grab samples shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.

(5)The 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 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 saven (7) days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months af ter changing, or af ter removal f rom I

sampler.

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least seven (7) days following each shutdown, startup, or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within a one-hour period and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding Lt0s may be increased by a factor of 10.

This requirement does not apply if 1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that ef fluent activity has not increased more than a f actor of 3.

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

l l

l 0

A.4-5 i

i

5.0 RA010 LOGICAL ENVIRONMENTAL MONITORING 5.1 Samolina 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 f rom station operation. This monitoring program is required by Technical Specification 3.12.1.

The monitoring program implementsSection IV.B.2 of Appendix 1 to 10CFR, Part 50, and thereby supplements the radiological ef fluent monitoring 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 8 of the 00CM.

The initially specified monitoring program will be ef fective for at least the first three years of comercial 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 8 to the 00CM.

Detection capability requirements, and reporting levels for radioactivity concentrations in environmental samples are shown on Tables A.5-2 and A.5-3, respectively, i

l 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

5.2 Land Use Census As part of the Radiological Environmental Monitoring Program, Technical Specification 3/4.12.2 requires that a land use census be conducted annually during growing season to identify within a distance of 8 km the location in 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 program and dose assessment models are made, if necessary.

This census satisfies the requirements of Section IV.3.3 of Appendix 1 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 from the census that yield a calculated dose of dose commitment 20 percent greater than at a location f rom 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.

OO A.5-2 J

TABLE A.5-1 Radiological Environmental Monitoring Program Number of Representative Sampling and Type and Frequency Exposure Pathway Samples and a

and/or Sample Sample Locations Collection Frequency of Analysis b

40 routine monitoring stations Quarterly.

Gamma dose quarterly.

1.

DIRECT RADIAT10N with two or more dosimeters placed as follows:

An inner ring of stations, one in each meteorological sector in the general area of the SITE BOUNDARY; An outer ring of stations, one in each meteorological sector, generally in the 6 to 8-km range 3

from the site; Y

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

2.

AIRBORNE Radiolodine and Samples from five locations :

Continuous sampler Radiolodine Cannister:

d operation with sample Particulates Three samples from close to the collection weekly, or I-131 analysis weekly.

three SITE B0UNDARY locations, more frequently if in different sectors, of high required by dust Particulate Sampler:

calculated long-tena average loading.

Gross beta radioactivity ground-level D/Q.

analysis following filter C

change -

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

ground-level D/Q.

O e

1

4 TABLE A.5-1 (Continued) 1 Number of Representative Sampling and Type and Frequency Exposure Pathway Samples and a

1 and/or Sample Sample Locations Collection Frequency of Analysis One sample from a control location.

as for example 15-30 km distant l

and in the least prevalent wind i

direction.

i j

p l

'RBORNE

. Surfaca One sample in th: discharge area.

Monthly grab sample.

Gasuna isotopic analysis' i

One sample from a control location.

monthly. Composite for tritium analysis quarterly.

i 5.

Sed %ent from One sample from area with existing Sem1 annually.

Ganuna isotopic analysis' f

from or potential recreational value.

semiannually.

1 L

shoreline l

"GES110N i

1 I

a.

Milk Samples from milking animals in Semimonthly when Gamma isotopic

and I-131 l

three locations within 5 km milking animals are on analysis on each sample.

j I

distaPce having the highest dose pasture, monthly at

nt eritial.

If their are none, other times.

then, one sample from milking animals in each of three areas 4

4 between 5 to 8 km distant where l

doses are calculated to be j

greater than 1 mres per yr.I One szJ.ste from milking animals at a control location, as for cxample,15-30 km distant and in the least prevalent wind a

4 direction.

i

TABLE A.5-1 (Continued)

Number of Representative Sampling and Type and Frequency Exposure Pathway Samples and a

and/or Sample Sample tocations Collection Frequency of Analysis e

b.

Fish and One sample of three commercia!!y Sair.ple in season, or Ganana isotopic analysis 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 Gansna isotopic 8 and 1-131 Products different kinds of broad leaf available.

analysis.

vegetation 9 grown nearest each of two different off-site locations of highest predicted long-term a

average ground-level D/Q if milk sampling is not performed.

One sample of each of the similar Monthly, when Gamma isotopic

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 perforteed.

O e

1 i

i i

TABLE A.5-1 (Continued) i v

Table Notation l

Specific parameters of distance and direction sector from the centerline of the Unit I reactor, and a) additional description where pertinent, shall be provided for each and every sample location in l

Deviations are permitted from the required sampling schedule if Table B.4-1 in the 00CM, Part B.

i specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability l

i If specimens are unobtainable due to sampling and malfunction of automatic sampling equipment.

}

equipment malfunction, effort shall be made to complete corrective action prior to the end of the next All deviations from the sampling schedule shall be documented in the Annual sampling period.

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 l

question and appropriate substitutions made within 30 days in the radiological environmental monitoring l

t j

Identify the cause of the unavailability of samples for that pathway and identify the new iocation(s), if available, for obtaining replacement samples in the next Sealannual Radioactive Effluent progras.

Release Report and also include in the report a revised figure (s) and table for the 00CM reflecting the new location (s).

A thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet l

Y b) i are considered as two or more dosimeters.

I q

Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more If gross beta activity in air particulate c) after sampling to allow for radon and thoron daughter decay.

f i

samples is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be j

j performed on the individual samples.

f Optimal air sampling locations are based not only on D/Q but on factors such as population in the area, l

d) year-round access to the site, and availability of power.

j Gaasna isotopic analysis means the identification and quantification of gasuna-emitting radionuclides that

}

i e) may be attributable to the effluents from the facility.

i 1

lhe dose shall be calculated for the maximum organ and age group, using the methodology and parameters 1

i f)

I in the 00CM, Part B.

1 if broad leaf vegetation is unavailable, other vegetation will be sampled.

g) i.

i i

TABLE A.5-2 Detection Capabilities for Environmental Sample Analysis,f,g a

Lower Limit of Detection (LLD)b Fish and Invertebrates Milk Food Products Sediment AirborneParticglate Water Analysis (pC1/kg) or Gas (pC1/m )

(pCi/kg. wet)

(pci/kg)

(pCi/kg wet)

(DC1/kg. dry)

Gross Beta 4

0.01 H-3 3,000 130 Mn-54 15 260 Fe-59 30 130 Co-58, 60 15 260 Zn-65 30 Zr-Nb-95 150 1

60' 1-131 15 0.07 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 15c d Ba-La-140 15c.d 9

O e

TABLE A.5-1 (Continued)

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

a)

Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed 6nd reported in the Annual Radiological Environmental Operating Report.

The LLO is defined, for purposes of these specifications, as the smallest b) concentration of radioactive material in a sample that wi*l yield a net count, above system background, that will be detected with 95%

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

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

4.66 s

" E

V

2. 22 ' Y

e x p ( -Mt )

Where:

LLO is the "a priori" lower limit of detection as defined above, as picoeuries per unit mass or volume; V

4.66 is a constant derived from the Kalpha and K eta values for b

i the 95% confidence level; is the standard deviation of the background counting rate or of sb 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 radicchemical 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.

J 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.5-8

TABLE A.5-2 (Continued) 94 It should be recognized that the LLO is defined as an g Driori (before the fact) limit representing the capability of a measurement system and not as This an a Dosteriori_ (after the fact) limit for a particular measurement.

does not preclude the calculation of an & Dosteriori LLO 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 interf ering 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.

c) Parent only.

The Ba-140 LLO and concentration can be determined by the analysis of its d) 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 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 time of collection is zero.

l h

e) Broad leaf vegetation only.

f)

If the measured concentration minus the three standard deviation uncertainty is found to exceed the specified LLO, the sample does not have to be analyzed to meet the specified LLD.

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

A.5-9

_. _ _ - _ _ _ = -

~ _.

I TABLE A.5-3 t

Reporting Levels for Radioactivity Concentrations in Environmental Samples Fish and Invertebrates Milk Food Products AirborneParticglate Water Analysis (pC1/kg) or Gas (pC1/m )

(pC1/kg. wet)

(DC1/kg)

(DCi/kg. wet)

H-3 30,000 30,000 Mn-54 1,000 i

10,000 Fe-59 400 30,000 Co-58 1,000 i

10,000 j

Co-60 300 20,000 l

2n-65 300 3

Zr-Nb-95 400*

3 100**

l I-131 100 0.9 Cs-134 30 10 1,000 60 1,000 1

Cs-137 50 20 2,000 70 2,000 I

300*

Ba-La-140 200*

i l

l l

i Parent only.

i l

- Broad leaf vegetation only.

I i

I l

I 1

1 1

1.0 INTRODUCTION

Part 8 of the 00CM (Of f-Site Oose Calculation Manual) provides formal and approved methods for the calculation of off-site concentration, off-site doses and effluent monitor setpoints, 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 00CM forms the basis for station procedures which document the off-site doses due to station operation which are used to show compliance with the numerical guides for design objectives of Section II of Appendix ! 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.

f 1.1 Responsibilities for Part B All changes to Part B of the 00CM shall be reviewed and approved by the Station Operations Review Comittee (SORC) in accordance with Technical i

Specification 6.13 prior to implementation.

Changes made to Part B shall be submitted to the Comission for their information in the Semiannual Radioactive Ef fluent 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 the ODCM is used in the performance of the surveillance requirements and administrate controls of the appropriate portions of the Technical Specifications.

O B.1-1 l

l

Sumarv of Methods. Oose Factors. Limits. Constants. Variables and 1.2 O

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

and setpoint methods are identified in Table 8.1-2 through Table 8.1-7.

Where more refined dose calculations are needed, the use of Method II dose determinations are described in Sections 3.2 through 3.9 and 3.ll.

The dose factors used in the equations are in Tables 8.*.-10 through 8.1-14 and the Regulatory Limits are summarized in Table 8.1-1.

The variables and special definitions used in this 00CM, Part 8, are in Tables B.1-8 and 8.l-9.

O l

8.I*2

m J

O TABLE B.1-1 Summary of Radiological Effluent Technical Specific.. ions and Implementing Equations Technical (1)

Specification Category Method I Limit l

3.11.1.1 Liquid Effluent Total Fraction of Eq. 2-1 51.0 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.

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

7

$ 10 mrem in a yr.

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

Treatment Operabili ty Organ Dose Eq. 3-2 5 0.2 mrem in a mo.

l 3.11.2.1 Gaseous Effluents Total Body Dose Rate Eq. 3-3 5 500 arem/yr.

Dose Rate from Noble Gases Skin Dose Rate Eq. 3-4 5 3000 mres/yr.

I from Noble. Gases i

Organ Dose Rate Eq. 3-5 5 1500 cres/yr.

f rom I-131, 1-133, Tritium and l

Particulates with Tjjp > 8 Days f

g 4

4

TABLE B.1-1 (continued)

Sunnary of Radiological Ef fluent Technical Specifications and Implementinq Equations Technical (1)

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

Dose from Noble Noble Gases 5 10 mrad in a yr.

Gases Beta Air Dose from Eq. 3-7 5 10 mrad in a qtr.

Noble Gases 5 20 mrad in a yr.

to 3.11.2.3 Gaseous Effluents Organ Dose f rom Eq. 3-8 5 7.5 mrem in a atr.

L Dose from I-131, lodines, Tritium and E

I-133, Tritium, Particulates with 5 15 mrem in a yr.

and Particulates 11/2 > 8 Days 3.11.2.4 Ventilation Organ Dose Eq. 3-8 5 0.3 mrem in a mo.

Exhaust Treatment 3.11.4 Total Dose (from Total Body Dose footnote (2).

5 25 mrem in a yr.

All Sources) 5 25 mrem in a yr.

Organ Dose 1 75 mrem in a yr.

Thyraid Dose O

e O

%/

TABLE B.1-1 (continued)

Summary of Radiological Ef fluent Technical Specifications and Implementing Equations 4

(1)

Technical 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 Tank Monitor 3.3.3.10 Geseous Effluent Monitor Setpoint T

Plant Vent Alarm / Trip Setpoint Eq. 5-9 T.S. 3.11.2.1 Wide Range Gas for Total Body Dose (Total Body)

Monitors Rate Alarm / Trip Setpoint Eq. 5-10 T.S. 3.11.2.1 (Skin) for Skin Dose Rate (1)

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

(2)

Technical Specification 3.ll.4.a requires this evaluation only if twice the limit of equations 3-1, If this occurs a Method II calculation, using actual release point 3-2, 3-12, 3-15 or 3-18 is reached.

parameters with annual average or concurrent meteorology and identified pathways for a real individual, shall be made.

a

TABLE B.1-2 Summary of Method _I Eauations to Calculate Unrestricted Area Liauid Concentrations.

Equation Ecuation Number Cateaorv ENG,{NPCi 2-1 Total Fraction of MPC in p

< )

1 Liquids, Except Noble Gases g

g NG(ml)={g NG d

C 2-2 Total Activity of Dissolved C I i

and Entrained Noble Gases from all Station Sources

$ 2E-04 O

B.1-6

TABLE B.1-3 O

Sumary of Method I Eauations to Calculate Off-Site Ooses from Liauid Releases.

Equation Eauation Number Category 3-1 Total Body 0tb(mrem) = k Qg 0Fl itb Dose j

3-2 Maximum 0mo(mrem) = k Q OFL,3 g

Organ Dose j

O B.1-7

TABLE 8.1-4 Surrrnary of Method I Eauations to Calculate Oose Rates Equation Ecuation Number __

Cateaory 3-3 Total Body Dose Rate ftb (mrem) = 0.62 h DFB i

i from Noble Gases yr 4

skin (mrem) " {O 3 -4 Skin Oose Rate 0F, i

4 yr from Noble Gases h

I yr ) "

i

$c, 0FG Ra rom dn co i

Tritium and Particulates with T 1/2 Greater Than Eight Days O

i l

TABLE 8,1-5 Summarv of Method I Ecuations to Calculate O

Ooses to Air from Noble Gases Equation Eauation Number Categorv Y

T 3-6 Gamma Dose to Air Oair (mrad) = 2.0E-08 0 DF 9

i from Noble Gases 3

0 0

3-7 Beta Dose to Air 0 air (mrad) = 4.4E-08 0 0F 9

i from Noble Gases 3

O 8.1-9

i TABLE B.1-6 Summary of Nethod I Ecuations to Calculate Dose to an Individual from Tritium. Iodine and Particulates Equation Ecuation Number Category 3-8 Dose to Critical 0co (mrem) =

Q OFG i

ico Organ from Iodines, 3

Tritium and Particulates

)

O e

TABLE 8.1-7 Sumary of Methods for Setooint Oeterminations Equation Eauation Number Cateaorv l

5-1 1.iouid Effluents:

_.0 F_,__

Liould Waste Test N

Tank Monitor setpoint (

)" f l 0F min 1

(RM-6509) hng k$a~

RCset(gph) = 1x108, 39p,

~

Gaseous Effluents:

l Plant Vent Wide Range Gas Monitors (RM-6528-1, 2, 3) 5-5 Total Body Rtb (uC1/sec) = 806 0FBc f

1 skin (vCi/sec) = 3000 1

5-6 Skin R

0F'

)

B.1 -11

TABLE B.1-8 Sumary of Variables Definition Units Variable

= Concentration at point of discharge of pCi/ml C

dissolved and entrained noble gas 'i' in liquid pathways from all station sources G

= Total activity of all dissolved and entrained

fC_i, C

mi noble gases in liquid pathways from all station sources

= Concentration of radionuclide 'i' at the point uti C

ml di of liquid discharge

= Concentration of radionuclide "i" uCi/mi C,

= Concentration, exclusive of noble gases, of uti C pg radionuclide "i" from tank "p' at point of ml discharge

= Concentration of radionuclide 'i' in mixture pCi/ml C,4 at the monitor mrad D

= Beta dose to air ar mrad D

= Beta dose for air at Education Center ir E mrad D

= Beta dose to air at "Rocks' ir R Y

mrad 0

= Gama dose to air air D

= Gamma dose to air at Education Center mrad T

ir E D

= Gama dose to air at "Rocks" mrad air R D

= Dose to the critical organ mrem mrem O

= Direct dose d

Dfinite

= Gama dose to air, corrected for finite cloud mrad 8.1-12

T ABLE B.1-8 (continued)

Sunnary of Variables Units Definition Variable mrem

= Dose to the maximum organ 0,

mrem

= Dose to skin f rom beta and gama 0

t mrem

= Dose to the total body Otb ratio OF

= Dilution factor ratio

= Minimum allowable dilution factor 0Fmin mrem-sec OF'

= Composite skin dose factor pCi-yr 3

""'*~*-

1

= Total body gama dose f actor f or nuclide 'i' pCi-yr 0F8 9 (Table 8.1-10) 3

= Composite total body dose factor C

D OFBc mrem

= Site-specific, total body dose factor for a 0Fl ith liquid release of nuclide 'i' (Table B.1-11) uCi

= Site-specific, maximum organ dose f actor for. a mrem DFL,*C liquid release of nuclide 'i' (Table B.1-11) 9C1

= Site-specific, critical organ dose f actor for a mrem OFG uCi gaseous release of nuclide 'i' (Table B.1-12) co mrem-sec_

= Site-specific, critical organ dose rate f actor 0FG'CO for a gaseous release of nuclide 'i' pCi-yr I

(Table B.1-12)

""'*~*3

= Beta skin dose factor for nuclide 'i' pCi-yr 0FS q (Table B.1-10) mrem-sec

= Combined skin dose f actor f or nuclide 'i' uti-yr DFj (Table B.1-10)

"8d'*

DF}

= Gama air dose f actor f or nuclide "i" pCi-yr (Table 8.1-10)

B.1 -13 i

T ABLE 8.1-8 (continued)

Sunvnary of Variables Units Definition Variable mead-m 0

= Beta air dose factor for nuclide "i" pCi-yr 0F I (Table 8.1-10) b

=driticalorgandoserateduetoiodines yr C

and particulates f

mrem h

= Skin dose rate due to noble gases yr skin mrem b

= Total body dose rate due to noble gases yr tb 1

= Deposition factor for dry deposition of D/0 elemental radioiodines and other particulates

,2 gpm or

= Flow rate out of discharge tunnel Fd ft /sec

= Flow rate past liquid waste test tank monitor gpm F,

cc j

= Flow rate past plant vent monitor F

sec Dimensionless aedon of total m associaW wm f); f3f

=

2 3

Paths 1, 2, and 3 Oimensionless

= Total fraction of MPC in liquid pathways f

F (excluding noble gases)

= Maximum permissible concentration for gC_i_

MPC cc g

radionuclide "i" (10CFR20, Appendix 8, Table 2, Column 2) curies, or

= Release to the environment for pcuries O g radionuclide 'i'

= Release rate to the environment for pCi/sec j

Q radionuclide 'i'

= Liquid monitor response for the limiting pCi/mi Rsetpoint concentration at the point of discharge cpm, or

= Response of the noble gas monitor at the uti/sec R g limiting skin dose rate B.1-14

l

_ TABLE B.1-B (continued) l s

Sunenary of Variables Variable Definition Units i

R

= Response of the noble gas monitor to cpm, or tb limiting total body dose rate pCi/sec Dimensionless l

S

= Shielding factor p

I F

= Detector counting ef ficiency f rom the com mR/hr or S

9 gas monitor calibration pCi/cc uCi/cc

= Detector counting efficiency for noble com mR/hr i

pCi/cc Ci/cc r

S 9

gas "i"

= Detector counting efficiency from the cos S) liquid monitor calibration pC1/ml S

= Detector counting efficiency for ces y$

radionuclide "i" uCi/ml X/Q

= Average undepleted atmospheric T

dispersion factoc (Tables B.7-4 and B.7-5) m 5'c (X/Q)T

= Ef fective average ganrna atmospheric 3

dispersion factor (Tables B.7-4 and 8.7-5) m SWF

= Service Water System flow rate gph PCC

= Primary component cooling water measured uCi/ml (decay corrected) gross radioactivity concentration i

B.1-15

TABLE 8.1-9 Definition of Terms 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.

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 Commission on Radiological Units and Measurements When applied to the evaluation of internal deposition or (ICRU).

radioactivity, the term "dose," as used here, includes the prospective dose component arising from retention in the body beyond the period of The dose commitment is environmental exposure, i.e., the dose comitment.

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 4

dose accumulation.

Liauid Radwaste Treatment System - The components or subsystens which comprise the available treatment system as shown in Figure B.6-1.

O B.1 -16

i TABLE 8.1-10 Dose Factors Specific for Seabrook Station for Noble Gas Releases Gamma Total Body Beta Skin Combined Skin Beta Air Garna Air Oose Factor Dose Factor Oose Factor Oose Factor Dose Factor 3

3 3

3 DF8 (mead-m ) 0F}(mead-m) g(mrem-m)DFj(mrem-sec) 3 (mrem-m )

DFS g

oci-vr DCi-vr Radionuclide OF8 oci-vr uti-vr DCi-vr Ar-41 8.84E-03*

2.69E-03 1.01E-02 3.28E-03 9.30E-03 1.33E-05 2.88E-04 1.93E-05 Kr-83m 7.56E-08 Kr-85m 1.17E-03 1.46E-03 2.86E-03 1.97E-03 1.23E-03 Kr-85

1. 61 E-05 1.34E-03 1.86E-03 1.95E-03 1.72E-05 Kr-87 5.92E-03 9.73E-03 1.77E-02 1.03E-02 6.17E-03 Kr-88 1.47E-02 2.37E-03 1.38E-02 2.93E-03 1.52E-02 Kr-89 1.66E-02 1.01E-02 2.59E-02 1.06E-02 1.73E-02 1.56E-02 7.29E-03 2.13E-02 7.83E-03 1.63E-02

(

) Kr-90 Xe-131m 9.15E-05 4.76E-04 7.65E-04 1.1?E-03 1.56E-04 Xe-133m 2.51E-04 9.94E-04 1.60E-03 1.48E-03 3.27E-04 Xe-133 2.94E-04 3.06E-04 6.66E-04 1.05E-03 3.53E-04 Xe-135m 3.12E-03 7.11E-04 3.30E-03 7.39E-04 3.36E-03 Xe-135

1. 81 E-03 1.86E-03 3.89E-03 2.46E-03 1.92E-03 Xe-137 1.42E-03 1.22E-02 1.79E-02 1.27E-02 1.51E-03 Xe-138 8.83E-03 4.13E-03 1.21E-02 4.75E-03 9.21E-03

8.84E-03 = 8.84 x 10-3 B.1 -17

TABLE B.1-Il G,

Dose Factors Specific For Seabrook Station For Liquid Releases Total Body Maximum 03gan Dose Factor Dose Factor DFL (mrem)

DFL;

[myem) uCi uCi Radionuclide H-3 3.02E-13 3.02E-13 Cr-51 1.83E-ll 1.48E-09 M n - 5.'s 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-08 9.21E-08 Zn-65 2.73E-07 5.49E-07 Br-83 1.31E-14 1.89E-14 Rb-86 4.18E-10 6.96E-10 Sr-89 2.17E-10 7.59E-09 l

Sr-90 3.22E-08 1.31E-07 Mo-99 3.10E-Il 2.62E-10 Tc-99m 4.95E-Il 7.16E-Il 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 6.97E-14 9.66E-14 l

Te-131m 3.16E-08 2.94E-06 Te-132 9.05E-08 3.80E-06 I-130 2.77E-ll 3.20E-09 I-131 2.21E-10 1.00E-07 I-132 3.30E-12 4.03E-12 1-133 2.55E-ll 1.15E-08 I-134 1.18E-12 1.40E-12 I-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 1.64E-10 3.48E-09 La-140 5.13E-Il 4.13E-08 Ce-141 3.67E-Il 9.31E-09 Ce-144 1.95E-10 6.46E-08 Np-239 4.55E-12 5.71E-10 O

B.1-18

TABLE 8.1-12 O

Dose and Dose Rate Factors Specific for Seabrook Station for Iodines. Tritium and Particulate Releases Critical Organ Critical Organ Dose Factor Dose Rate Factor ico (mrem) ico (mrem-sec}

uCi vr-uCi Radionuclide 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 Zn-65 1.19E-06 3.75E+01 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 I-132 7.67E-09 2.42E-01 1-133 6.25E-07 1.97E+01 1-134 2.01E-09 6.34E-02 I

I-135 3.24E-08 1.02E+00 Cs-134 1.31E-05 4.13E+02 O

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

8.1-19 i

,--..-,_.m_._,

TABLE 8.1-13 Combined Skin Oose Factors Specific for Seabrook Station SDecial Receptors (l) for Noble Gas Release Education Center The "Rocks' Combined Skin Combined Skin Oose Factor Dose Factor OF'E(m em-sec)

DF'R(mrem-sec) i uCi-vr i

uCi-yr Radionuclide Ar-41 3.85E-02 1.21E-01 Kr-83m 4.25E-05 1.25E-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 4.93E-02 1.52E-01 Kr-89 1.06E-01 3.40E-01 Kr-90 8.48E-02 2.70E-01 Xe-131m 3.54E-03 1.17E-02 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 5.42E-02 Xe-137 8.52E-02 2.84E-01 Xe-138 4.80E-02 1.53E-01 O

(1)

See Seabrook Station Unit 1 Technical Specification Figure 5.1-1.

1 l

O B.1-20

TABLE 8.1-14 Dose and Dose Rate Factors Soecific for Seabrook Station SDecial Receptorsll) for lodine.

Tritium and Particulate Releases The "Rocks" Education Center Critical Organ Critical Organ Critical Organ Critical Organ Dose Factor Dose Rate Factor Oose Factor Oose Rate Factor icoR uCi )

OfG Imrem-sec)

$ mrem icoE uci )

DFGjcoEImrem-sec)

DM mrem coR uCi-vr I

uCi-vr OFG Radionuclide H-3 2.71E-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 Fe-59 2.98E-07 9.40E+00 9.93E-07 3.13E+01 Co-58 2.62E-07 8.26E+00 8.73E-07 2.75E+01 Co-60 8.99E-06 2.84E+02 4.01E-05 1.26E+03 2n-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 Sr-90 2.11E-05 6.65E+02 7.02E-05

2. 21 E+03 Mo-99 5.25E-08 1.66E+00 1.75E-07 5.52E+00 I-130 3.61E-07 1.14E+01 1.20E-06 3.78E+01 1-131 3.17E-06 1.00E+02 1.06E-05 3.34E+02 1-132 3.78E-08 1.19E+00 1.26E-07 3.97E+00 1-133

7. 51 E-07 2.37E+01 2.50E-06 7.89E+01 7

I-134 9.90E-09 3.12E-01 3.29E-08 1.04E+00 I-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 C s-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 i

Ce-144 2.61E-06 8.23E+01 8.68E-06 2.74E+02 j

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

O 8.1-21

2.0 METHOD TO CALCULATE OFF-SITE LIOUID 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 total f raction of MPC in liquid pathways, other then noble gases, denoted here as F

, at the point of discharge from the statior, to the environment NG (see Figure B.6-1).

Ff"O is limited to less than or equal to one, i.e.,

F NG < ),

1 The total concentration of all dissolved and entrained noble gases at the point of discharge from the multiport diffuser from all station sources combined, denoted C

, is limited to 2E-04 pC1/m1, i.e.,

f 0

Cf12E-04pCi/ml.

Method to Determine F "O and C"O 2.1 O

First, determine the total f raction of MPC (excluding noble gases), at the point of discharge from the station from all significant liquid sources denoted F 0; and then separately determine the total concentration at the point of discharge of all dissolved and entrained noble gases from all stationsources,denotedCfasfollows:

G,{MPCo.i < ),

{2,1)

F 1

9 3

IuCi/ml}

uct/ml and:

B.2-1 O

C 5 2E-04 (2-2)

Cf

=

(yCi/ml)

(uCi/ml)

(uCi/ml) where:

Total fraction of HPC in liquids, excluding noble F

=

gases, at the point of discharge from the multiport diffuser Concentration at point of discharge from the multiport C,j

=

p dif fuser of radionuclide "i", except for dissolved and entrained noble gases, from all tanks and other significant sources, p, from which a discharge may be made (including the waste test tanks and any other significant source from which a discharge can be made) (vCi/ml)

Maximum permissible concentration of radionuclide 'i' except HPCj

=

for dissolved and entrained noble gases f rom 10CFR20, Appendix B, Table II, Column 2 (pCi/ml)

Cf

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

Concentration at point of discharge of dissolved and entrained C

=

noble gas 'i' in liquids from all station sources (uCi/ml) 2.2 Method to Determine Radionuclide Concentration for Each Liouid Effluent Source 2.2.1 Waste Test Tanks C

is determined for each radionuclide detected from the activity p,g 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 f rom the floor drain tanks when that liquid does not require processing in the evaporator, distillate f rom the boron recovery evaporator B.2-2 8

1 when the BRS evaporator is substituting for the waste evaporator, and distillate from the Steam Generator 810wdown System evaporators and flash f

i 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 j

gamma emitters in accordance with the liquid sample and analysis program outlined in Part A to the 00CM.

2.2.2 Turbine 8911dina Sumo 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 t,he blowdown liquid is being discharged to the environment instead of the normal recycling process (see Table A.3-1).

B.2-3

3.0 0FF-SITE 00SE CALCULATION METH00S Chapter 3 provides the basis for station procedures required to meet the Radiological Ef fluent Technical Specifications (RETS) dose or dose rate requirements contained in Section 3/4.11 of the station operating Technical Specifications. A simple, conservative method (called Method 1) 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 equations are given in Chapter 7.0.

The Semiannual Radioactive Effluent Release Report, to be filed after January 1 each year per Technical Specification 6.8.1.4, requirs: that meteorological conditions concurrent with the time of release of radioactive materials in gaseous ef fluents, as determined by sampling frequency and measurement, be used for determining the gaseous pathway doses.

For continuous release sources (i.e., plant vent, condenser air removal exhaust, y/

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 activitics (i.e., containment purge or venting to atmosphere of the Waste Gas 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 from each unit; otherwise quarterly average meteorology will be applied.

Quarterly average meteorology will also be applied to batch releases if the hourly met data for the period of batch release is unavailable.

O V

B.3-1

3.1 Introductorv Concepts _

In part, the Radiological Effluent Technical Specifications (RETS) limit dose or dose rate. The term "dose" for ingested or inhaled radioactivity means the dose comitment, measured in mrem, which results f rom 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 the dose comitment is evaluated is 50 years.

The phrases "annual dose" or dose in one year" then refers to the 50-year dose comitment resulting from exposure to one year's worth of releases.

' Dose in a quarter' similarly means the 50-year dose comitment resulting f rom exposure to one quarter's releases. The term "dose,' with respect to external exposures, such as to u ble ga? 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 acct.mulation to the dose comitment.

"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 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 of 10 mrem / year, even though the actual dose received in the year of exposure may be less than 10 mrem.

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 boundary at any time are limited to the equiv31ent annual dose limits of 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 assurance that members 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)).

O B.3-2

Thequantitiesa0andbareintroducedtoprovidecalculable m

quantities, related to of f-site doses or dose rates that demonstrate compliance with the RETS.

i Delta 0, denoted AD, is the quantity calculated by the Chapter 3, Method I dose equations.

It represents the conservative increment in dose.

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 release because of the conservative margin built into the dose factors and the I

selection and definition of critical receptors.

The radionuclide specific dose factors in each Method I dose equation represent the greatest dose to any organ of any ags up.

(Organ dose is a function of age because organ mass and intake are functions of age.) The critical receptor assumed by ' Method I" equations is then generally a hypothetical individual whose behavior - in 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.

Odot, den:tedb,isthequantitycalculatedintheChapter3doserate equations.

It is calculated using the station's ef fluent 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 from the plant stack continued for one entire year.

Since peak release rates, or resulting dose rates, are usually of short time duration on I

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 f actors and variable parameters to be selected to best fit the actual release.

Guidance is provided, but the p

appropriate margin and depth of analysis are oetermined in each instance at the time of analysis under Method II.

B.3-3

Method to Calculate the Total Body Dose from Liauid Releases 3.2 Technical Specification 3.11.1.2 limits the total body dose comitment to a member of the public from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year per unit. Technical Specification 3.11.*

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 the total body dose comitment to any real member of the public f rom all station sources (including liquids) to 25 mrem in a year.

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 t more refined calculation of total body dose is needed, i.e., Method I ir.dicates the dose might be greater than the Technical Specification limits.

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.

3.2.1 Method I The increment.n total body dose from a liquid release is:

O

=k Q 0Fl

( 3-1 )

tb 4

itb i

(mrem)

( ) (uCi) (*C I

where:

Site-specific total body dose f actor (mrem /pci) for a DFlitb

=

liquid release.

It is the highest of the four age groups.

See Table 8.1-11.

Total activity (uCi) released for radionuclide "i".

(For Qi

=

strontiums, use the most recent measurement available.)

O B.3-4

is the average (typically monthly 918/F ; where Fd X

d

=

average) dilution flow of the Circulating Water System at

)

the point of discharge from the multiport diffuser (in ft /sec).

For normal operations with a cooling water flow 3

/

of 918 f t /sec, K is equal to 1.

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

1.

Liquid releases via the multiport diffuser a unrestricted areas (at the edge of the initial misitg or prompt dilution zone that corresponds to a factor of 10 d' ution), 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 appear to exceed a Technical Specification limit or if a more exact calculation is required, then Method 11 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 from Regulatory Guide 1.109, and used in the derivation of the simplifieo 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.

O 8.3-5

Method to Calculate Maximum Organ Dose from Licuid Releases 3.3 Technical Specification 3.11.1.2 limits the maximum organ dose comitment to a Meraber of the Public. f rom radioactive material in liquid Technical ef fluents to 5 mrem per quarter and 10 mrem per year per unit.

Specification 3.11.1.3 requires liouid radwaste treatment when the maximum organ dose projected exceeds 0.2 mrem in any 31 dbys (see Subsection 3.11 for dose projections).

Technical Specification 3.11.4 limits t'ie maximum organ dose comitment to any real member of the public f rom all station sources (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 and more conservative than Method II.

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.

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

O

=k 0

0Fl (3-2) gg 4

imo i

(mrem) ()(uC1)(*[)

where:

Site-specific maximum organ dose factor (mrem /uti) for a OFl mo i

=

liquid release.

It is the highest of the four age groups.

See Table B.1-11.

Total activity (uCi) released for radionuclide "i".

(For j

Qi

=

stronti_ums, use the most recent maa;urement available.)

i O

B.3-6

is the average (typically monthly 918/F ; where Fd K

d

=

average) dilution flow of the Circulati'g Water System at iO the point of discharge from the multiport diffuser (in ft /sec).

For normal operations with a enoling '.ater flow 3

of 918 ft /sec, K is equal to 1.

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

Liquid releases via the multiport dif fuser to unrestricted areas 1.

(at the edge of the initial mixing or prompt dilution zone tnat corresponds to a factor of 10 dilution), and 2.

Any continuous or batch release over any time period.

3.3.2 Method II 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 calculation is required, then Method 11 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 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 i

coincides conditions at the time of release, such as identified pathways of exposure and dilution values associate / with the receptor.

j O

B.3-7

1 Method to Calculate the Total Body Dose Rate From Noble Gases 9!

3.4 limits the dose rate at any time to Technical Specification 3.11.2.1 the total body from noble gases at any location at or beyond the site boundary j

The Technical Specification indirectly limits peak release to 500 mrem / year.

rates by limiting the dose rate that is predicted from continued release at the to a rate equivalent to no more than 500 By limiting Otb

)

peak rate.

mrem / year, he 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 f

release rate via the station vents 'I.

Method I applies at all release I

rates.

is desired by the UseMethod11ifamorerefinedcalculationofb tb

)

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 Sea Section 7.2.1 for actually been exceeded during a short time interval.

basis.

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

Specifications are performed when the ef fluent monitor alarm setpoint is exceeded, or as required by the Action Statement (Technical Specification 3.3.3.10, Table 3.3-10) when the monitor is inoperable.

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

station vent stack, and, any potent *<* ground level releases.

O B.3-8 i

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

0.62 Q

OFB (3-3)

Dtb j

j

=

3 (mrem)

DCi-sec) fi) mrem-m 3

sec pCi-yr yr CM where:

The release rate at the station vents (uCi/sec), for each Q

=

noble gas radionuclide, "i', shown in Table 8.1-10.

Total body gamma dose f actor (see Table B.1-10).

OFB

=

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

1.

Nonnal operations (nonemergency event), and 2.

Noble gas releases via any station vent to the atmosphere.

3.4.2 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 11 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 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

i Method to Calculate the Skin Oose Rate from Noble Gases 3.5 Technical Specification 3.11.2.1 limits the dose rate at any tiae to the skin from noble gases at any location at or beyond the site boundary to The Technical Specification indirectly limits peak release 3,000 mrem / year.

rates by limiting the dose rate that is predicted from continued release at to a rate equivalent to no more than the peak rate.

By limiting Oskin 3,000 mrem / year, we assure that the skin dose accrued in any one year by any Since it can be member of the general public is less than 3,000 mrem.

is derived would not be expected that the peak release rate on which Oskin exceeded without corrective attion being taken to lower it, the resultant 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 Oose Rate from the peak

)

release rate via the station vents Method I applies at all release rates.

UseMethodIIifamorerefinedcalculationofb is desired by the skin 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 7.2.2 for basis.

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

Determinations of dose rate for compliance with Technical Specifications are performed when the effluent monitor alarm setpoint is I

exceeded.

(I)

The Turbine Building vent ground level release X/Qs are used in the j

ODCM Method I equations.

This is to conservatively account for the station vent stack, and, any potential ground level releases.

l B.3-10

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

b 0Fj (3-4) skin "

i i

gmrem) di) mrem-sec) yr sec pCi-yr where:

h The release rate at the station vents (pCi/sec) for each

=

4 radionuclide, 'i', shown in Table B.1-10.

OFj combined skin dose f actor (see Table 8.1-10).

=

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

1.

Normal operations (nonet:

gency 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 II may be applied. Method 11 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 i

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.

l B.3-11

3.6 Method to Calculate the Critical Organ Oose Rate from Iodines. Tritium and Particulates with Tl/2 Greater Than 8 Days Technical Specification 3.11.2.1 limits the dose rate at any time to any organ from I, I I,

H and radionuclides in particulate form with I

half lives greater than 8 days to 1500 mrem / year to any organ.

The Technical Specification indirectly limits peak release rates by limiting the dose rate that is predicted from continued release at the peak rate.

Bylimitingb 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 peak release rate via the station vents Method I applies at all release rates.

Use Method II if a more refined calculation of D is desired by the co 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 7.2.3 for basis.

3.6.1 Method i The Critical Organ Dose Rate can be determined as follows:

h 0FG (3-5)

O,

=

g g

ico mrem) uCi) mrem-sec) yr sec pCi-yr (I)

The Turbine Building vent ground level release X/Os are used in the 00CM Method I equations.

This is to conservatively account for the station vent stack, and, any potential ground level relesses.

B.3-12

O where: OFGjg, = Site-specific critical organ dose rate f uCi-yr for a gaseous release.

See Table 8.1-12.

h

= The activity release rate at the station vents of 4

radionuclide "i" in pCi/sec (i.e., total activity measured of radionuclide 'i' averaged over the Ome period for which the filter / charcoal sample collector was in the effluent 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, justify Method I or consider Method II):

1.

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

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 11 may be applied. Method 11 consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev.1 (Ref erence 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.

I O

B. 3-13

Method tc Calculate the Gama Air Oose f rom Noble Gases O\\

3.7 Technical Specification 3.11.2.2 limits the gamma dose to air from noble gases at any location at or beyond the site boundary to 5 mrao in any quarter and 10 mrad in any year per unit.

Oose evaluation is required at least once per 31 days, f

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

vent releases during the period.

Use Method II if a more refined calculation is needed (i.e., use of actual release point parameter with annual or actual meteorology to obtain release-specific X/0s), or if Method I predicts a dose greater than the Technical Specification limit to determine if it had actually been exceeded.

See Section 7.2.4 for basis.

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

D

= 2.0E-08 O

0F}

(3-6) ar g

1 3

(mrad) (DCi-vr)

( Ci) (mrad-m )

pCi-m pCi-yr where:

O

= total activity (pCi) released to the atmosphere via station g

vents of each radionuclide 'i' during the period of interest.

See Table S.1-10 OF}=gammadosefactortoairforradionuclide"i".

(I)

The Turbine Building vent ground level release X/Os 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.3-14

Equation 3-6 can be applied under the following conditions (otherwise O

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 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.

O V

B. 3-15

Nethod to Calculate the Beta Air Dose from Noble Gases O

3.8 Technical Specification 3.11.2.2 limits the beta dose to air f rom noble 4

gases at any. location at or beyond the site boundary to 10 mrad in any quarter and 20 mrad in any year per unit. Dose evaluation is required at least once per 31 days.

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

U;e 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 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 O

DF (3-7) 0

=

ir g

i 3

(mrad)

(oCi-vr)

Cl) (mrad-m 3

pCi-yr pCi-m where:

0 OF = beta dose factor to air for radionuclide "i".

See Table B.1-10 0

= total activity (pCi) released to the atmosphere via station 9

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 Nethod I equstions.

This is to conservatively account for the station vent stack, and, any potential ground level releases.

B.3-16

Equation 3-7 can be appliv under the following conditions (otherwise O

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 If 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 !! 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.

O 8.3-17

Method to Calculate the Critical Organ Oose f rom Iodines. Tritium and 3.9 Particulates Technical Specification 3.11.2.3 limits the critical organ dose to a 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 f rom all station sources (including gaseous ef fluents) 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 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.

3.9.1 Method i 0

OFG (3-8)

D,

=

C ge, (mrem)

(uCi)(]*)

Q

= Total activity (9C1) released to the atmosphere of radionuclide j

"1" during the period of interest.

For strontiums, use the most recent measurement.

For each je, = Site-specific critical organ dose factor (mrem /uC1).

OFG radionuclide it is the age group and organ with the largest dose factor.

See Table 8.1-12.

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

2.

Iodine, tritium, and particulate releases via station vents to the O\\

atmosphere, and Any continuous or batch release over any time period.

3.

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 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.

O 8.3-19

3.10 Method to Calculate Direct Oose from Plant Operation 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 It should De noted year (except the thyroid, which is limited to 75 mrem).

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 f rom TLD locations situated on-site near potential sources of direct radiation, as well as those TL0s 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 liquia and gaseous effluents.

O B.3-20

3.11 Dose Proiections b

Technical Specifications 3.11.1.3 and 3.11.2.4 require that appropriate portions of liquid ar,d gaseous radwaste treatment systems, respectively, be 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 The reasonably achievable" criteria of Appendix 1 to 10CFR Part 50.

surveillance requirements of these Technical Specifications state that dose projections be performed at least once per 31 days when the liquid radwaste treatment systems or gaseous radwaste treatment systems are not being fully utilized.

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 limit for use of radwaste equipment, adjusted if appropriate for known or expected dif ferences between past operational parameters and those anticipated for the next 31 days.

O 3.11.1 liauid Dose Proiections The 31-day liquid dose projections are calculated by the following:

and organ dose Dmo (Equations 3-1 (a) Determine the total body Dtb and 3-2, respectively) for tne last typical cot.:pleted 31-day period. The last typical 31-day period should be one without significant identified operational differences from the period being projected to, such as full power operation vs. periods when the pl4nt is shut down.

(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.

O B.3-21

(c) Calculate the ratio (R )

f the estimated gross primary coolant 2

activity for the projected period to the average value in the reference period. Use the most recent value of primary coolant activity as the projected value if no trend in decreasing or increasing levels can be determined.

(d) Determine the projected dose from:

.R Total Body:

O

=Og. R) g p

Max. Organ:

D

=0

. R).R m pr 2

3.11.2 Gaseous Dose Proiections For the gaseous radwaste treatment system, the 31-day dose projections are calculated by the following:

(a) Determine the gama air dose D {r (Equation 3-6), and the beta air a

dose 0, (Equation 3-7) f rom the last typical 31-day operating period.

(b) Calculate the ratio (R ) of anticipated number of curies of noble 3

gas to be released from the hydrogen surge tank to the atmosphere 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.

If no differences between the reference period and the next 31 days can be identified, set R 3

(c) Determine the projected dose from:

D }r pr = Dar.R3 Gama Air:

a Beta Air:

O

=D r. R, r pr O

8 B.3-22

For the ventilation exhaust treatment system, the critical organ dose f rom iodines, tritium, and particulates are projected for the next 31 days by the following:

(a) Determine the critical organ dose 0,(Equation 3-8) from the last typical 31-day operating period, i

(b) Calculate the ratio (R ) of anticipated primary coolant dose 4

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 )

f an cipated pH ma y system leakage S

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) Detertnine the projected dose f rom:

l Critical Organ:

O

=O,.R4.R5 co pr c

9 O

B.3-23

4 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM O

.0 The radiologi al 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.

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 f

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

Pursuant to Specification 4.12.2, the land use census will be conducted

'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 tensus 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 O

authorities.

Technical Specification 6.8.1.3 requires that the results of the 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 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 i

measurements only, as specified in the NRC Branch Technical Position, but upon all measurements.

This will prevent the positive bias associated with the calculatioi of the mean and range based upon detectable measurements only.

Secondly, the Lower Limit of Detection column will specify the LLO required by 00CM Table A.5-2 for that radionuclide and sample medium.

O B. 4 -1 4

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 1.

AIRBORNE (Particulate and Radioiodine)

AP/CF-01 PSNH Barge 2.7 ESE Landing Area AP/CF-02 Hampton Marina 2.7 E

AP/CF-03 SW Boundary 0.8 SW 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-Discharg,* Area 5.3 E

WS-51 Ipswich Bay (Control) 16.9 SSE b.

Sediment SE-02 Mampton-Discharge Area (b) 5.3 E

SE-07 Hampton Beach 3.1 E

SE-0B 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 E.2 SW TM-0B 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

1 Area j

HA-54 Ipswich Bay (Control) 17.2 SSE MU-06 Hampton - Discharge 5.2 E

Area MU-56 Ipswich Bay (Control) 17.4 SSE O

B.4-2

TABLE B.4-1 (continued)

RadioloaicalEnvironmentalMonitorinaStationka)

\\s /

Distance From Exposure Pathway Sample Location Unit 1 Direction From

.ontainment (km) the Plant C

and/or Samole and Desianated Code 4.

DIRECT RADIATION TL-1 Brimmer's Lane, 1.1 N

f Hampton Falls TL-2 Landing Rd., Hampton 3.2 NNE TL-3 Glade Path, Hampton 3.1 NE Beach TL-4 Island Path, Hampton 2.4 ENE Beach 1L-5 Harbor Rd., Hampton 2.7 E

Beach TL-6 PSNH Barge Landing 2.7 ESE Area TL-7 Cross Rd., Seabrook 2.6 SE Beach TL-B 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 4

J O

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 Brinner'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, SeabrooX 2.7 SE Beach TL-22 Cable Ave.,

7.6 SSE Salisbury Beach TL-23 Ferry Rd., Salisbury B.1 S

TL-24 Ferry Lots Lane, 7.2 SSW Salisbury TL-25 Elm St., Amesbury 7.6 SW TL-26 Route 107A Amesbury 8.1 WSW B.4-3

TABLE B.4-1 (continued)

Radioloaical Environmental Monitorina Stationf a)

Distance From Exposure Pathway Sample Location Unit 1 Direction Frcm and/or Samole and Desionated Code Containment (kml the Plant TL-27 Highland St.,

7.6 W

S. Hampton TL-28 Route 150, Kensington 7.9 WNW TL-29 Frying Pan Lane, 7.4 NW Hampton Falls TL-30 Route 101C, Hampton 7.9 NNW TL-31 Alumni Drive, Hampton 4.0 NNE TL-32 Seabrook Elementary 1.9 S

School TL-33 Dock Area, Newburyport 9.7 S

TL-34 Bow St., Exeter 12.1 NW TL-35 Lincoln Ackerman 2.4 NNW School TL-36 Route 97, Georgetown 22 SSW (Control)

TL-37 Plaistow, NH (Control) 26 WSW TL-38 Hampstead, NH (Control) 29 W

TL-39 Epping, NH (Control) 27 NW TL-40 Newmarket, NH (Control) 24 NNW TL-41 Portsmouth NH 21 NNE (Control)(b)

TL-42 Ipswich, MA (Contral)(b) 27 SSE (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 00CM; program requirements specified in Part A do not apply to samples taken at this location.

(c) Samples will be collected pursuant to 00CM Table A.5-1.

Samples are not required f rom all stations listed during any sampling interval (FH = Fish; HA = Lobsters; MU = Mussels).

Table A.5-1 specifies that "one sample of three 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 ef fort has been made (e.g., yellowtail flounder and winter flounder).

B.4-4

i 9

PM

\\

t

~~

r b4 s,

8 ROWS RIVER I

f E.ASR00K 1 M/cr 2@

/

M/CF=04 @

S&EASR00K 2 a

H N NM m urs plu s cRe u M/Cr.o1 st ca i

i E

j s*

o o

500 1000 4,

HETEns 5

Figure B'.4-1 Radiological F.nviron:nental Monitoring Locations Within 4 Kilometers of Seabrook Station B.4-5

i i

5 I

l 0

S i

i__..i i

g EIlt*tt Ti a 4 g

8i N

)

RYE EZACH AP/CT*C5 SEC EMEARCENENT !W T! CURE 2.1 T

D.H.03e.__. s......

's 4

I tw.to @

! [ NAMFTC# SEM H i

l. St-07 SEA 8R00Kl$TATl0N' 0

G

' 5C***

5 3 u 01 MJ 06 FH 03 KA 04 SE 01

{

SEABROCK EEACH

,/

e

/

g-, s...

h IM 04 w MLISillRY ELACH t'Ri m e' C ATUA7tc oegAu O

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

\\V Y

'l t llNu.Tt.as d

TORK e

DURRAM e y

's #4f PORT 5 MOUTH e

.Os.

NE) MARK (i e

\\

RAYtCND e gppig,

ql

.,i f.

x, s

ltrertR e l

N l

l i

SEE CRLAMCDGr? IM TECURE 3.2 l

1 HAW TOM e e

i l

SEASA00K STATION mi m ST0s. l SEA 8R00Ke ( "* P *

1 i

e'.

l

}

l N.-.

Ol)CHARGESITE

,/ *',

y,e SALI5 BURY l

i g,

ruiSTow.

/

l l

l m nc oc w

, /. -.

'y.J 8

i NEWBURYPORT e Ng HAVERHILLe I

f e St 57 i

FLtM 13LNQ s.

'I TM 20 FH-53 W5-ICTHUEM e

-u

~

m

@*,h@

d IP5WICH4 b

d e GLOUCESTER O

Figure B.4-3 Radiological Environmental Monitoring 1,ocations Outside 12 Kilometers of Seabrook Station B. 4-7 l

NNE

/

i

\\

"NW N

/

,g

\\:

8

/

TL 35 e TL.3 l

NW h

@TL4 ENE TL 5 TL 6

@ TL 1 WNW TL 14 @

gnov atvtR i

l s h er k

g)g W

TL13@

[ E)

SEMR -

2 /

'N NAMPTOW NAR$d2 f

j TL 12@

[.

/

'/#* * '

n.n e 4

lt 10

? TL-6 WSW 9

n8 i

e ESE

\\

TL 7 @

O TL-32@

gN SW g

g E

g k

5 o

500 1000 TL.21 g h

SSW

Tens b

3 s

FiEure B.4-4 Direct Radiation Monitoring Locations Within 4 Milomet*fS of Seabrook Station B. 4-8

l

/

(V y

g<

NNW t-r L - ""i-J u t.inu ti e i

h NNE J

H NE NW N

f

@ TL 34

\\

s.,,,,

ty! 8E ACH TL.17 TL 30 ft 18 e.TL.19 l

@ TL 29 NO SEE ENL4RC IN FlCURE 2.4 ENE 3@..

g' TL.20 NMPTOW SEACH W

@ TL.27 E

DISCMM Slit

.s h" stA8R00t SEAuf

/'e TL 26 ej,

s.....'

1 f aq ESE f

WSW

\\

@ TL 25

!At,rstety cu TL 24' TL.t TL 23 V

b Sluy 0 1

SE Art wTre oeg w Ti.33@

g

)

1 SSW SSE 1

S O

Figure B.4-5 nirect Radiation Monitoring Locations Between 4 Kilometers and 12 Kilometers from Seabrook Station B. 4-9

~

w

-~

m

h J 0

S._

'O

'I

' Pfl l

E llJ tW 18. A %

NNW o

,m yy DURit 't

TL.40.

NE ng u aggy.

PORT Util e '.

@ W.

@ TL.39 TL.41 RAYNND

10 Miles g,, g g,

k

*\\.

Y N

ENE e

i N

Ititt

l i

WDUT 15 !! Cunt 8.3 i

l S F. A BROOK TI I

I KIN 35f; Ne g.

W e

$tABR0 e

b

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l j.

y PLAISTOW e a

l AI N IC OCfAN TL.37%,h.y.)

NI URYPORT j

ESE N

/

g y....................

\\

PLUM LAHO e.~.

,/ mu.

LAwRtact

,,,,,, cy 3g IP5WICM 8

. TL 42 SW 0

CR SSW SSE Figure 3.4-6 Direct Radiation Monitoring Locations Outside 12 Kilometers of Seabrook Station B.4-lO

.. -. _ _.... -. -. ~.. -.

5.0 SETPOINT DETERMINATIONS Chapter 5 contains the plant procedures that the plant operator 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.

J

.i l

l 4

l i

l 1

a l

4

~

l 1

i 8.5-1 1

Liouid Effluent _.nstrumentation setpoints t

5.1 Technical Specification 3.3.3.9 requires that the radioactive liquid ef fluent 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 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.

5.1.1 Liauid Waste Test Tank Monitor (RM-6509)

The liquid waste test tank effluent monitor provides alarm and automatic tertnination of release prior to exceeding the concentration limits It specified in 10CFR20, Appendix B, Table II, Column 2 to the environment.

is also used to monitor discharge, from various waste sumps to the environment.

5.1.1.1 Method to Determine the Setooint of the Liacid Waste Test Tank Monitor (RM-65M1 The instrument response (uCi/ml) for the limiting concentration at the point of disc'.arge is the setpoint, denoted Rsetpoint, and is determined as follows:

Rsetpoint " 1 C

I in (uti/ml)

()/)

(

)

where-1

= Dilution f actor (dimensionless)

(5-2)

DF

=

m F,

= Flow rate past monitor (gpm)

F

= Flow rate out of discharge tunnel (gpm) d OF

= Minimum allowable dilution factor (dimensionless) min O

l l

B.5-2 l

l

is the fraction of the total 2

I ); where f) 1 - (f f

=

3

[

contribution of MPC at the discharge point to be associated and f are the with the test tank effluent pathway and, f2 3

similar f ractions for. Turbine Building sump and steam generator I I)*

blowdown pathways, respectively:

(f) +f2+f3 1

(~}

0Fmin "

l 4

l MPC f or radionuclide "i" f rom 10CFR20, Appendix B, Table II, MPC

=

4 Column 2 (>Ci/ml).

In the event that no activity is expected to be 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 8, notes.

Activity concentration of radionuclide "i" in mixture at the C,9

=

monitor (uci/ml) 5.1.1. 2 liouid Waste Test Tank Monitor Setooint Example The activity concentration of each radionuclide, Cg, in tfie 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 folicwing data:

i C,4 (vCi/ml)

MPC4 (vCi/ml)

Cs-134 2.15E-05 9E-06 C s-137 7.48E-05 2E-05 Co-60 2.56E-05 3E-05 C,4 = 2.15E-05 + 7.48E-05 + 2.56E-05 i

Ip_C i)

IuCi I C

gCis uci (ml '

(mi )

ml ml

= 1.22E-04 Ig01) m1 8.5-3

O mi (5-3) 0F g MPC) min-vCi-m1 Imi pCi) 2.15E-05 + 7.48E-05 + 2.56E-05 9E-06 2E-05 3E-05 uCi-m1 uCi-m1 uCi-ml Imi pCi)

Imi pCi)

Iml pCi)

DFmin "

The minimum dilution factor, OFmin, needed to discharge the mixture of radionuclides in th example is 7.

The release rate of the waste test tank is between 10 and gpm. The circulating water discharge flow can vary f rom 10,500 to 412,^N gpu,of dilution water. With the dilution flow taken as 412,000 gpm ari; the release rate f rom the waste test tank taken as.150 gpm, the OF 1s:

d DF p

=

m (gpm)

(5-4)

(gpm) 412.000 com~

150 gpm

= 2750 0

B.5-4 1

l l

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

Rsetpoint " 1 0

in

( )( )

(ml )

ml 0

= 0.6 1.22E-04 7

()()

(

)

= 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 atrove background, s

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

The Turbine Building drains liquid ef fluent monitor continuously monitors the Turbine Building sump effluent line.

The only sources to the Sump Ef fluent System are f rom 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 those radionuclides found in the secondary system, with reduced activity f rom decay and dilution.

The Turbine Building drains liquid ef fluent monitor provides alarm arni 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 O

B.5-5

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:

(5-21) 2 (DF') (1.0E-07 pCi/ml)

High Trip Monitor

=f Setpoint (uCi/ml) where:

Circulating water flow rate (comi OF' =

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

1 + f ); where the f values are described

" I - (f f2 3

above.

High Trip (5-22)

Warning Alarm

=IMonitor Setpoint) (0.25)

Monitor Setpoint (pCi/ml) 5.1.3 Steam Genera +or Blowdown Liauid 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.

I For any liquid monitor, in the event that no activity is expected to be discharged, or can be measured in the Lystem, the liquid monitor setpoint should be based on the most restrictive MPC for an "unidentified" mixture given in 10CFR20, Appendix B notes.

9;

~

8.5-6

l l

5.1. 4 PCCW Head Tank Rate-of-Change Alarm Setooint 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 f rom the PCCW System.

For the rate-of-change alarm, a setpoint is selected based on detection of an activity level equivalent to 10 pCi/mi in the discharge of the Service Water System. The activity in

-0 the PCCW is determined in accordance with the liquid sampling and analysis program described in Part A, Table A.3-1 of the OOCH and is used to determine the setpoint.

The rate-of-change alarm setpoint is calculated from:

-8, 397, PCC (5-23)

RC

= 1x10 set hr ), (M)

(M)

(ml )

al I

mi hr pCi O

G where:

RCset

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

lx10-8

= the minimum detectable activity level in the Service Water System due to a PCCW to SWS leak (pCi/ml).

SWF

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

PCC

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

-5 As an example, assume a PCCW activity concentration of 1x10 uti/mi with a service water flow rate of only 80 percent of the normal flow of 21,000 gpm. The rate-of-change setpoint is then:

~0 6

RCset = 1x10 1.0x10 g h (1/1x10-

)

O B.5-7

1 I

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

-8 1x10 pCi/mi can be found from:

l

-5 uCi/mi 1000 aph_

($_g) g set, 1x10 PCC O

M O

B.5-8

Gaseous Effluent Instrumentation Setooints 5.2 Technical Specification 3.3.3.10 requires that the radioactive gaseous ef fluent instrumentation in Table 3.3-13 of the Technical !secific their alann setpoints set to insure that Technical Specification 3.11.2.1 is not exceeded, i

Plant Vent Wide-Rance Gas Monitors (RMQ9-1.2 and 3)_

5.2.1 The plant vent wide-range gas monitors are shown on Figure 8.6-2.

r Method to Oetermine the Setooint of the Plant Vent Wide Rance Ga 5.2.1.1 Monitors (RM-6528-1.2 and 3)

The setpoint for the plant vent wide-range gas monitor (readout response in pCi/sec) is set by limiting the off-site noble gas dose rate to N

is the the total body or to the skin, and is denoted Rsetpoint*

setpoint lesser of:

(5-5) 806 R

=

tb 0FB g 3

pCi/sec (mrem-uci-m )

I oCi-vr}

3 yr-pCi-sec mrem-m and:

(5-6)

R

= 3,000 0F.

skin C

I uCi-pCi/sec

("y'r*) (mrem vr )

see where:

= Response of the monitor at the limiting total body dose R

rate (pCi/sec)

B.5-9

.,.--,-,n,.,,

, = - - _, - -,. - -., -., ~,,,,,,.. -,

500 mrem-uci-m

)

yr-pci-sec (lE+06)

(6.2E-07) 500

- Limiting total body dose rate (mrem /yr) lE46

= Number of pCi per pCi (pCi/pCi) 6.2E-07

= [X/Q]T, maximum annual average gamma atmospheric. dispersion 3

factor (sec/m )

= Composite total body dose f actor (mrem-m /pCi-yr)

DFBc h OFB j

j i

(5-7)

=

T-o ci i

h

= The release rate of noble gas "i" in the mixture, f or each 4

noble gas identified in the off-gas (vCi/sec) 3 DFB

= Total body dose f actor (see Table B.1-10) (mrem-m / Ci-yr) q R

= Response of the monitor at the limiting skin dose rate skin (pCi/sec) 3,000

= Limiting skin dose rate (mrem /yr)

DF'

= Composite skin dose f actor (mrem-sec/pci-yr) 0F 1

j (5-8)

=

b3 i

O 8.5-10

l

= Combined skin dose f actor (see Table 8.1-10) 0F' p

i (mrem-sec/pCi-yr)

Plant Vent Wide Ranae Gas Monitor Setooint Example _

5. 2.1. 2 The following setpoint example for the plant vent wide range gas monitors demonstrates the use of equations 5-5 and 5-6 for deternining setpoints.

This setpoint example is based on the following data (see Table B.1-10 for DFB and 0F ):

q h

0FB 0Fj 4

3 3

gj.

(mrem-m )

mrem-sec)

_..,suc) oCi-vr uCi-vr g

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 h 0FB 3

4 (5-7) 0FB

=-

c Oi h 0FB9 = (1.03E+04)(8.83E-03) + (4.73E+02)(5.92E-03) 4 i

+ (2.57E+02)(1.47E-02) + (1.20E+02)(1.17E-03)

+ (3.70E+02)(1.81E-03) + (1.97E+01)(2.94E-04)

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

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

i B. 5 -11

+ 1.20E+02 + 3.70E+02 + 1.97E+01 9)

- 1.15E+04 uC1/sec 0F8

= 1.15E+04 c

3

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

= 806 tb 0F8c

= (806)

(8.52E-03)

= 9.46E+04 uCi/sec and next; j OFj (5-8)

OF' =

hg i

h 0Fj = (1.03E+04)(1.21E-02) + (4.73E+02)(1.77E-02) 4 i

+ (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 (uci-mrem-sec/sec uci-yr)

Oy = 1.38E+02 1.15E+04 1

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

(5-6) O Rskin = 3,000 0F, c

8.5-12 i

= (3,000)

(1.20E-02)

= 2.50E+05 uti/sec and R For the The setpoint, R,, p g, is the lesser of R skin.

"9 8

is less than Rskin' noble gas mixture in this example Rtb Therefore, in this example the the total body dose rate is more restrictive.

plant vert wide-range gas monitors should each be set at 9.46E+04 vCi/sec above backgr9und, 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 in the system to be vented, the gaseous monitor setpoint should be based on Xe-133.

O O

8.5-13

-T'"9'Y y*w-y m,9

6.0 LIOUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RA0 WASTE TREATMENT SYSTEMS Figure 8.6-1 shows the liquid ef fluent streams, radiation monitors and the appropriate Liquid Radwaste Treatment System.

Figure 8.6-2 shows the gaseous effluent streams, radiation monitors and the appropriate Gaseous Radwaste Treatment Sy: tem.

For more detailed infonation concerning the above, refer to the Seabrook Station Final Safety Analysis Report, Sections 11.2 (Liquid Waste System),11.3 (Gaseous Waste System) and 11.5 (Process and Ef fluent Radiological Monitoring and Sampling System).

The turbine gland seal condenser exhaust is an unmonitored release 1

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 O

exhaust by use of the flow rates.

O B.6-1

I O

O MAMEUP T*'""

STOR AG g s ORAGE PAS TANK TANK PAS UNIT 1 UNIT 2 UNIT 2 UNIT 2

)

UNIT 1 UNif 1 1

I I

I l

W j

i i

6,m

,'__G_g t _

_q)__ _ i I

I BORON I

RECOVERY l

r f

SYSTEM i

i

't l o

_ _ _.q)_..

...g

'l 1

[EN LWP$

TUR8INE TUR8tNE I

LWPS 4

NON RECYCLE SulLOING BUILOING RECYCLE 8 LOEv00WN PORTION UNIT 2 g

POR TION SYSTEM UNIT 1 I

l l

I

's h

u _ _t i

'i

=

1 3

l l

u

~

-EM-6S21) ( FM-6S19) f~

l

( RM-6SO9 )

o CIRCULATING u

y

....--e=

WATER SYSTEM g

O6

{

(2-RM-6521)

PCCW

_]l(*

SWS SYSTEM SYSTEM RELEASE

,, Future,,

CCLT-2172-1 CvCS LETOOWN OlVERSION EQUIPMENT OR AIN AGE CCLT-2272-1 EOuiPMENT LEAK AGE h

PA8 FLOOR OR AINS NON.RECYCLASLE AND MISC.

LEGENO CONT AINMENT SUMPS CONDENSATE LE AK AG E LABOR ATORY OR AINS DECOW AM ATION W AT E R RECYCLA8LE DE AE R ATED RECYCLABLE AERE ATED h

TUR81NE BLOG SUMP NON RECYCLA8LE AND MISC. Q TRITIUM CONTROL AELE ASE STE AM C EN. 8LOWOOWN SECOND ARY SIDE STM. G EN.8LOWOQWN RM Radiation Monitor 0

Service Water System 6

CCLT Level Transmitter v

Figure B.6-1 Liquid Ef fluent Streats, Radiation MOnit TS, and Radwaste Treatment System at SeabrOok Station B.6-2

[

ftt8tNE CLAND

\\

SEAL C0C D5t1

\\

EXMAUST k

h (DCl!NC HOCCINC MODE CafLY)

CO NT AINM E NT SUILDING VENTIL ATORS

'j, n '!. "...E *. +, [.", i. *,

TU R 81N E il iL

l f

/

b

.. i

/

at..scs

(

r CONT AINMENT PURGE AIRg g

s y

s ir

~

H w l

BLOWOOWN WASTE h

F LASH T ANK BUILDING

.'W REACTOR AIR Tfe

I-l COOLANT G ASEOUS W ASTE PROCESSING SYSTEM h JY'
'.'s.' ?
t*
.. ;4.,
lA,'
A 7
4
- TYP& CAL OP 3 db:
P RIM AR Y di L' di
.a J
i GUARD d I I
VENT S ED.
<It!
idi.
ir t..
STACK Lgm l
l
.. r.
4 m
SURGE' at65281,2.2T
- CH ARCOAL SEDS DRYER l
31 6530-1,2 )
I COWRESSOR PRIMARY
***i'
AUXILIARY i
BUILDING i
2.1 M
.y at4 Sos at-6504 f
6 OONTROL
),,
,.T A N K ','
j DEG ASIFIE R L
o b
ir AUXILIARY BUILDING VENT AIR c
E LEGEND i
[
[
H HEPA FILTER I'
c C - CHARCOAL FILTER w -
RM-RADIATION BUILDING
$3 MONITOR Figure B.6-2 Gaseous Effluent Streams, hadiation Monitors, and Radwaste Treatment System at Seabrook Station B.6-3
l
\\
7.0 BASES FOR DOSE CALCULATION METH005 (v
7.1 Liouid Release Oose Calculations This section serves:
(1) to document the development and conserystive nature of Method I equations to provide background information to Method !
users, and (2) to identify the general equations, parameters and approaches to Method Il-type dose assessments.
Method I may be used to show that the Techni al 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 j
limits in Technical Specification 3.11.1.2 are based on the ALARA design The minimum dose values objectives in 10CFR50, Appendix 1 Subsection II A.
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 1). The definition, below, of a single "critical receptor" (a hypothetical or real individual whose behavior results in a maximum potential dose) provides part of the conservative margin to the calculation of tot 31 body dose in Method I.
Method 11 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 11 analysis fo. 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 1.
The general equations used in the base case analysis are also used as the starting point in Method 11 evaluations.
The base case, the method of reduction, and the assumptions and data used are presented below.
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 g
The base case analysis uses the general equations, effluents was derived.
methods, data and assumptions in Regulatory Guide 1.109 (Equations A-3 and B. 7 -1
A-7, Reference A).
The liquid pathways contributing to an incividual dose are due +.o consumption of fish and invertebrates, shoreline activities, and swiming and boating near the discharge point.
A normal operating plant discharge flow rate of 918 f t /sec was used with a mixing ratio of 0.10.
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 11 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and for shoreline activity 0.0 hours0 days 0 hours 0 weeks 0 months . Table B.7.1-1 outlines the human consumption and environmental parameters used in the analysis.
The resulting, site-specific, 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:
a0
=k 0
0Fl
( 7 -1 )
tb 4
itb (mrem) ( ) (9C1) (*C )
where:
Site-specific total body dose factor (mrem /pCi) for a OFLitb
=
liquid release.
It is the highest of the f'ur age groups.
o See Table B.1-11.
Total activity (uCi) released for radionuclide "i".
Qi
=
918/F ; where Fd is the average dilution flow of the K
d
=
Circulating Water System at the point of discharge from the 3
multiport diffuser (in ft /sec).
B.7 2 1
Method 1 is more conservative than Method II in the region of the used in Technical Specification limits because the dose factors DFLitb Method I were chosen for the base case to be the highest of the four age In effect each groups (adult, teen, child and infant) for that radionuclide.
radionuclide is conservatively represented by its own critical age group.
7.1.2 Dose to the Critical Oraan The methods to calculate maximum organ dose parallel to the total body dose methods (see Section 7.1.1).
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 the maximum organ dose factor (OFL
) for that radionuclide.
DFL g also includes the external dose contribution to the critical organ.
For any liquid release, during any period, the increment in dose f rom radionuclide 'i' to the maximum organ is:
(7-2) 60
=k Q
OFL,
g j
(mrem) ( ) (uCi) (*C
)
where:
Site-specific maximum organ dose factor (mrem /uCi) for a OFlimo
=
liquid release.
See Table 8.1-11.
Total activity (uCi) released for radionuclide 'i".
Og
=
918/F ; where Fd is the average dilution flow of the f
K d
=
Circulating Water System at the point of discharge f rom the 3
multiport diffuser (in ft /sec).
O B.7-3
TABLE 8.7-1 Usage Factors for Various Liquid Pathways at Seabrook Station Zero where no pathway exists)
(From Reference A, Table E-5*, except ?.s noted.
AGE VEG.
LEAFY MILK MEAT FISH INVERT.
POTABLE SHOREllNE SWIMMING *** BOATING ***
WATER (KG/YR)
(KG/YR)
(LITER /YR)
(KG/YR)
(KG/YR)
(KG/YR)
(LITER /YR)
(HR/YR)
(HR/YR)
(HR/YR)
VEG.
Adult 0.00 0.00 0.00 0.00 21.00 5.00 0.00 334.00**
8.00 29.00 Teen 0.00 0.00 0.00 0.00 16.00 3.80 0.00 67.00 45.00 52.00 Child 0.00 0.00 0.00 0.00 6.90 1.70 0.00 14.00 28.00 52.00 Infant 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Y
Y.
v
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 Ef fects f rom Nuclear Power Industry,"
HEDL December 1971 O
e
~
7.2 Gaseous Release Oose Calculations O.
7. 2.1 Total Body 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.
Method I may be used to show that the Technical Specification which limits total body dose rate f rom 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 f rom general equation B-B in Regulatory Guide 1.109 as follows:
b
= 1E+06 (X/Q]
h 0FB (7-3) tb 4
4 3
(mrem)
DCi)
( ) (sec.)
(uci) (mrem-m )
ec sec pCi-yr yr pCi
,3 where:
[X/Q]T Maximum receptor location long-term average garrrna atmospheric
=
dispersion factor.
3 6.2E-07 (sec/m ),
=
h Release rate to the environment of noble gas "i" (uCi/sec).
=
3 Gama total body dose f actor, (
).
See Table B.1-10.
0FB
=
4 (Regulatory Guide 1.109 Table B-1).
Equation 7-3 reduces to:
0.62 h
0FB
=
tb 4
i (3-3)
O B.7-5
3 mrem oci-sec fji I
I"I I
bec.)Imrem-m )
3 pd -yr yr Ci-m The selection of critical receptor, outlined in Section 7.3 is inherent in the derived Method I, since the maximum expected off-site long-term average atmospheric dispersion factors were used.
All noble gases in Table B.1-10 should be considered.
A Methcd 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 Oose 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 erluations 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 f rom 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 0, pages 5 and 6, is 30 rem /10).
The f actor 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 average meteorology and that the release rate reading remains constant over the entire year.
Method I was derived from the general equation B-9 in Regulatory Guide 1.109 as follows:
0{ir + 3.17 E+04 Q$ (X/Q) 0FS (7-4)
D
= 1.11 g
01 1
B.7-6
i 3
Ci sec mrem-m Imrem) " ( ) Imrad)
IDCi-vr)
% (,T ' IpCi-yr I yr yr Ci-sec where:
Average ratio of tissue to air absorption coefficients (will 1.11
=
convert mrad in air to mrem in tissue).
Beta skin dose factor for a semi-infinite cloud of OFSj
=
radionuclide "i" which includes the attenuation by the outer
' dead' layer of the skin.
T (7-5) 0}ir = 3.17E+04 Q
(X/Q) 0F 3
4 i
3 see d
Ci
,3 ) mra -m Imrad)
IDCi-vr)
IE) I pCi-yr yr Ci-sec DFf = Gamma air dose factor for a uniform semi-infinite cloud of radionuclide "i".
Now it is assumed for the definition of (X/0 ) for Reference 8 that:
D (X/0]T [X/0)
(7-6)
/
O
=
finite air 3
mrad mead see m
I sec)
I I
I II,3 )
yr yr I
31.54 h
(7_7) and Q
=
4 4
g ")
Ci-sec) uti)
C1 y
pCi-yr sec i
b
= 1.11 1E+06 (X/Q]Y h
0F{
(7-8) so:
3 skin i
I 3
uCi) mrad-m (mrem),( )
DCi)
,3 )
sec yr pCi see pCi-yr
+.1E+06 X/0 h
0FS 4
i 3
fjt mrem-m )
( p_C_i)
C see Ci 3
sec pCi-yr O
g B.7-7
l l
l substituting 6.2E-07 sec/m3
[X/Q]Y
=
1.4E-06 sec/m3 X/Q
=
(7-9) 0.62 h
DF{+
1.40 h
0FS gives b 4
=
skin 4
1 i
3 3
mrem oci-see-mremquci q Ci-yr ) DCi-sec)guci)g Ci-yr )
mrem-m mrem-m g
y g 3
sec p 3
sec p yr uCi-m -mrad uCi-m h [0.62 0FJ + 1.40 0FS ]
(7-10)
=
4 i
define DFj=0.620F{+1.400FS
( 7 -11) g "70 1
DFj (3-4) then:
Dskin g
mrem) uCi) mrem-sec) yr 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 of f-site atmospheric dispersion factors at the most limiting location.
All noble gases in Table 8.1-10 must be considered.
7.2.3 Critical Organ Oose Rate From lodines. 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.
1 i
B.7-8
Method I may be used to show that the Technical Specification which
)
limits organ dose rate f rom iodines, tritium and radionuclides in particulate (j
form with half lives greater than 8 days released to the atmosphere (Technical Specification 3.11.2.1) has been met-for the peak above-mentioned release rates.
The annual organ dose limit is 1500 mrem (f rom NBS Handbook 69, Reference 0, pages 5 and 6).
It is evaluated by looking at the critical organ dose comitment to the most limiting of f-site receptor assuming long-term site average meteorology.
Theequationforb,isderivedbymodifyingEquation3-8from e
Section 3.9 as follows:
(3-8)
O 0FG O
co i
$e, (mrem)
(pCi)
(*C
)
applying the conversion f actor, 3.154E+07 (sec/yr) and converting Q to hinvCi/secyields 3.154E+07 0
DFG
( 7-I 2 )
=
co 4
ico
(*y*r* ) (E)
( U ) (mrem) yr sec pCi Eq. 3-8 is rewritten in the form:
(3-5) co i
DFGjco (mrem (uCi) (mrem-sec.
yr sec pCi-yr #
where 3.154E+07 0FG
( 7-13) kco
=
ico (mrem-sec) " (see)
(mremI pCi-yr yr pCi O
B.7-9
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 f actors may be l
performed with concurrent meteorology and latest land use census data to I
identify existing pathways.
Because of the choice of atmospheric dispersion factors and pathways, it is expected that Method I results always will exceed Method 11 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 Gama 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, parametcrs and approaches to Method II-type dose assessments.
Method I may be used to show that the Technical Specification which limits of f-site gama 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 1, 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 from Equations B-4 and B-5 of Regulatory Guide 1.109 with the added assumptionthatOfinite = O D Q h (UQ]:
60,{r = 3.17E+04 [X/0)Y Q
OFf (7-14) i (mrad) = (C ec) (sec/m )
(Ci) (mrad
)
3
_pC O
B. 7 -10
where:
3.17E+04 = number of pCi per Ci divided by the number of seconds per year.
[X/Q)T = maximum annual average ganc.a atmospheric dispersion factor
= 6.2E-07 (sec/m )
Q
= number of curies of noble gas "i" relea*ed 4
DF}
= Gansna air dose f actor for a uniform semi-infinite cloud of radionuclide "i".
which leads to:
2.0E-08 O
0F{
(3-6) 0,jr g
l
=
i rd (mrad)
(
3)
(pCi) ("C _
)
The majcr dif ference between Method I and Method 11 is that Method II would use actual or concurrent meteorology with a specific noble gas release spectrum to determine (X/Q)T rather than use the most limiting meteorological dispersion value obtained for the years 1979 to 1981.
7.2.5 Beta 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 beta 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 8.1, 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 from Equations B-4 and B-5 of Regulatory Guide 1.109:
8.7-11
a0
= 3.17E+04 X/0 0
OF
( 7-15 )
$7 3
(mrad) = (DCi-sec)(E)
(pCi) (mra6m )
~YI 3
pd-F m
where: OF = Beta air dose f actors for a uniform semi-infinite cloud of 0
radionuclide "i".
j substituting X/Q = Maximum long-term average undepleted atmospheric dispersion factor 3
= 1.4E-06 sec/m,
We have 0
4.4E-08 Q
OF (3-7)
D
=
ir rd (mrad) = (D -r)
( Ci) (*C _
)
3 7.2.6 Dose to Critical Organ From Icdines. Tritium and Particulates With Half-Lives Greater Than Eight Days 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 11-type dose assessments.
Method I may be used to show that the Technical Specifications which limit of f-site organ dose f rom gases (3.11.2.3 and 3.11.4) have been met for releases over the appropriate periods.
Technical Specification 3.11.2.3 is based on the ALARA Objectives in 10CFR50, Appendix I, Subsection 11 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.
B. 7 -12
Method I was developed such that "the actual exposure of an 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 11 allows that actual individuals, associated with Method 1.
identifiable exposure pathways, be taken into account for any given release.
In fact, Method I was based on a Method 11 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.
The steps performed in the Method I derivation follow.
First, the dose impact to the critical receptor (in the form of dose factors OFGico (mrem /uC1)] for a unit activity release of each iodine, tritium, and particulate radionuclide with half lives greater than eight days to gaseous effluents was derived. Seven exposure pathways (ground plane, inhalation, stored vegetables, leafy vegetables, cow milk, goat milk, and meat ingestion) ere assumed to exist at the site boundary (not over water or marsh areas)
O 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 whole body and skin for four age groups (adult, teenager, child, and infant) due to the seven combined exposure pathways.
For each radionuclide, the highest dose per unit activity release for any organ (or whole body) and age group was then selected to become the M6thoo 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 f rom direct exposure to contaminated ground plane; Equation C 4 for doses associated with inhalation of all radionuclides to different organs of individuals of dif ferent 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 8.7-2 and B.7-3 outline human consumption and environmental parameters used in the anclysis.
It is conservatively assumed that the critical receptor lives at the "maximum off-site atmospheric dispersion factor location" as defined in Section 7.3.
B.7-13
The resulting site-se.i ic dose factors are for the maximum organ which combine the limiting e group with the highest Jose f actor for any s
organ with each nuclide. These critical organ, critical age dose factors are given in Table 8.1-12.
I For any iodine, tritium, and particulate gas release, during any period, the increment in dose from radionuclide 'i' is:
(7-16) oDico " O OFG4c, i
where DFG is the critical dose f actor for radionuclide 'i' and Q is ico the activity of radionuclide 'i' released in microcuries.
Because of the assumptions about receptors, environment, and radionuclides and because of the rey lations of 10CFR50 and 40CFR190, the lack of imediate 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.
O 7.2.7 SDecial Receptor Gaseous Release Dose Calculations 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.
1.2.7.1 Total Body Dose Rate From Noble Gases Method I was derived from Regulatory Guide 1.109 cs follows:
0FB (7-3) tb = 1E+06 [X/Q]T i
O B.7-14
General Equation (7-3) is then multiplied by an Occupancy Factor (OF) to account for the time an individual will be at the on-site receptor O-locations during the year.
For the Education Center, and the "Rocks", the OFs are:
)
h /
= 0.14 Education Center -
h p h"
= 0.0076 The "Rocks" -
60
/r I
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")
r gives C"8 i(mrem /yr)
(7-17 )
htti = 0.0028 i
PFB (mre9/yr)
(7-18) h p - 0,0c j
(I)Taken from Seabrook Station Technical Specifications (Figure 5.1-1).
i O
8.7-15
where:
tbR = Total body dose rates due to noble gases to an btbE,andh individual at the Education Center and the "Rocks" l
(recreational site), respectively.
h = defined previously 4
DFB, = defined previously.
7.2.7.2 Skin Oose Rate From Noble Gases Method I was derived from Equation (7-8):
(7-8) bskin = 1.111E+06 (X/Q]Y h
0F{
+
4 i
DFS 1E+06 X/Q i
substituting (X/Q]Y = 2.0E-06 sec/m3 (Education Center) 5.9E-06 sec/m3 (The "Rocks")
X/0 = 6.7E-06 sec/m3 (Education Center) 1 2.3E-05 sec/m (The "Rocks")
multiplying by J
OF = 0.0014 (Education Center)
= 0.0076 (The ' Rocks")
9' B.7-16
gives (2.22 0F{ + 6.7) 0FS ] (mrem /yr) g bskinE = 0.0014i i
[6.490F}+22.50FS)(mrem /yr) bskinR = 0.0076i i
then:
i II-I9) h DF'E (mrem /yr) bskinE = 0.0014$
i i
(7-l)
{
b b,R (mrem /yr) bskinR = 0.0076 where:
b andbskinR = the skin dose rate due to noble gases to an skinE individual at the Education Center and the "Rocks,"
respectively.
Q4 = defined previously.
CF'iE and DF'iR = the combined skin dose factors for radionuclide "i for the Education Center, and the "Rocks",
j respectively (see Table B.1-13.)
Critical Organ Dose Rate From lodines. Tritium and Particulates With f
7.2.7.3 Half-lives Greater Than Eicht Days Theequationsforb are derived in the same manner as in Section 7.2.2, except that the occupancy factors are also included.
Therefore:
0FG II-2I) bcoE = 0.0014 i
icoE (mrem /yr) o (7-22) i 0FGjcoR (mrem /yr) bcoR = 0.0076 8.7-17
where:
coR = the critical organ dose rates to an individual at the h
8" Education Center and the "Rocks", respectively, coE h = defined previously.
4
= the critical organ dose rate factors for OF' co and DF'icoR radionuclide 'i' for the Education Center and the ' Rocks,' respectively (see Table 8.1-14.)
7.2.7.4 Gantna Dose to Air From Noble Gases Method I was derived f rom Equation (7-14):
(7-14) 0 0F{
D{ir = 3.17E44 [X/0]T 1
substituting
[X/Q)T = 2.0E-06 sec/m3 (Education Center)
= 5.9E-06 ser/m3 (The ' Rocks')
multiplying by 0F = 0.0014 (Education Center)
= 0.0076 (The ' Rocks')
and 1E-06 Ci/uci 9
8.7-18
l l
gives (7-23)
O 0F{
(mrad)
T i
0, ire = 8.88E-11 (7-24) 0 0F{
(mrad)
T 1
0,irR = 1.42E-09 where:
= the garrma air doses to an individual at the T
Education Center and the ' Rocks,' respectively, 0
and O ire irR g = total activity (pCi) released to the atmosphere via the statio O
vents of each radionuclide "i".
OFj and 0F{ = defined previously.
7.2.7.5 Reta Dose to Air From Noble Gases Method I was derived f rom Equation (7-15):
(7-15) l 0 0F 0
4 air = 3.17E+04 X/Q D
substituting X/0 = 6.7E-06 sec/m3 (Education Center)
= 2.3E-05 sec/m3 (The ' Rocks")
P91tiplying by 0F = 0.0014 (Education Center)
= 0.0076 (The ' Rocks')
and 1E-06 Ci/vci 8.7-19
gives (7-25) 0F Oi (mrad) 0 aire = 2.97E-10 0
i
( -26)
O 0F i
(mrad) 0 airR = 5.54E-09 D
where:
= the beta tir doses to an individual at the Education 0
0 0
and 0 aire airR Center and the "Rocks," respectively.
g = total activity (uti) released to the atmosphere via the O
station vents of each radionuclide "i".
OF and DF
= defined previously.
Critical Oraan Oose From lodines. Tritium and Particulates With 7.2.7.6 Half-l.ives Greater Than Eiaht Days Methed I was derived in the same manner as Equation (3-8):
O 0FG (3-0) g ico O,=
c multiplying by 0F = 0.0014 (Education Center)
= 0.0076 (The ' Rocks')
and 1E-06 Ci/uti; plus substituting the location specific 0FGs gives
(~
I 0 0FG 0
= 0.0014 1
icoE coE i
E L
O DFG Wed (7-28) g
$ cog DcoR = 0.0076 i
D.7-20
where:
0 and D
= the critical organ doses of an individual at the coE coR Education Center and the "Rocks,' respectively.
4 = the total activity (uCi) released to the atmosphere of radionuclide Q
"i".
icoR = the critical organ dose factors (mrem /uC1) for the OFG and DFG Education Center and the "Rocks,' respectively icoE for each radionuclide 'i". The factors represent the age group and organ with the largest dose f actor (see Table B.1-14).
The soecial receptor equations can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1.
Nonnal 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 O
limit, or if a more refined calculation is required, then Method Il may be j
applied, I
I l
8.7-21
-_.___.____._____________,,.._.__w
TABLE 8.7-2 Environmental Parameters for Gaseous Ef fluents at Seabrook Station (Derived from Reference A)*
Vegetables Cow Milk Goat Milk Meat Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV Agricultural (Kg/M )
2.
2.
0.75 2.
0.75 2.
0.75 2.
2 Productivity P
Soil Surface Density (KG/M )
240.
240.
240.
240.
240.
240.
240.
240.
2 T
Transport Time to User (HRS) 48.
48.
48.
48.
480.
480.
T8 Soil Exposure Time (HRS) 131400.
131400.
131400.
131400.
131400.
131400.
131400.
131400.
IF Crop Exposure Time (HRS) 1440.
1440.
720.
720.
720.
720.
720.
720.
Ito Plume TH Holdup After Harvest (HRS) 1440.
24.
O.
2160.
O.
2160.
O.
2160.
[$ QF Animals Daily Feed (KG/ DAY) 50.
50.
6.
6.
50.
50.
0.50 0.50 0.50 kj FP Fraction of Year on Pasture 1.
1.
1.
FS Fraction Pasture when on Pasture FG Fraction of Stored 0.76 Veg. Grown in Jarden FL Fraction of teafy 1.0 Veg. Grown in Garden F1 Fraction Elemental lodine = 0.5 3
H Absolute (gm/M )
Humidity = 8.00**
Regulatory Guide 1.109
Default value from NRC "GASPAR" Lase Code; K. F. Eckerman, revised December 2, 1975 9
9 e
TABLE 8.7-3 O
Usaae Factors for Various Gaseous Pathways at Seabrook Stition (from Reference A, Table E-5)*
Maximum Receptor:
Age Leafy Group Veoetables Vecetables Milk Meat Inhalation i
3 (kg/yr)
(kg/yr)
(1/yr)
(kg/yr)
(m /yr)
Adult 520.00 64.00 310.00 110.00 8000.00 Teen 630.00 42.00 400.00 65.00 8000.00 Child 520.00 26.00 330.00 41.00 3700.00 Infant 0.00 0.00 330.00 0.00 1400.00 The "Rocks' and Education Center:
Age Leafy 9fr_qug Vecetables Veaetables Milk Meat Inhalation 3
(kg/yr)
(kg/yr)
(1/yr)
(kg/yr)
(m /yr)
Adult 0.00 0.00 0.00 0.00 8000.0 f
Teen 0.00 0.00 0.00 0.00 8000.0 Child 0.00 0.00 0.00 0.00 3700.0 Infant 0.00 0.00 0.00 0.00 1400.0
Regulatory Guide 1.109 8.7-23
Receptor Points and Averace Atmospheric Dispersion Factors for 7.3 In1ortant Exoosure Pathways The gaseous effluent dose equations (Method 1) have been simplified by assuming an individual whose behavior and living habits inevitably lead to a higher dose than anyone else.
The following exposure pathways to gaseons effluents listed in Regulatory Guide 1.109 (Reference A) have been consid6 red:
1.
Direct exposure to contaminated air; 2.
Direct exposure to contaminated ground; 3.
Inhalation of air; 4.
Ingestion of vegetables; 5.
Ir.gestion of cow's and goat's milk; and 6,
Ingestion of meat.
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.
Section 7.3.3 presents the maximum atmospheric dispersion factors calculated at each of the off-site receptor locations.
i 7.3.1 Receptor I.ocations The most limiting site boundary location in which individuals are, or l
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 f
estimate of the dose to an individual from existing and potential gaseous pathways for the Method I analysis.
This point is the H sector, 914 meters f rom the center of the reactor units.
B.7-24
l l
Two other locations (on-site) were analyzed for direct' ground plane l
exposure and inhalation only. They are the ' Rocks' (recreational site) and the Education Center shown on Figure 5.1-1 of the Technical Specifications, i
4 l
l i
i i
1 i
1 J
b i
1 i
8.7-25 i
1
. -. --. _-,._._ -- -.., _.,, _.. m
-.._._.-__,,_.-.m.__._
Seabrook Station Atmospheric Dispersion Model 7.3.2 The time average atmospheric dispersion factors are computed for routine (long-terni) ground level releases using the AEOLUS Computer Code AEOLUS is based, in part, on the straight-line airflow model (Reference B).
discussed in Regulatory Guide 1.111 (Ref erence C).
AEOLUS produces the following average atmospheric o'spersion factors for each location:
Undepleted X/Q dispersion factors for evaluating ground level 1.
concentrations of noble gases; Depletsd X/Q dispersion factors for evaluating ground level 2.
concentrations of iodines and particulates; 3.
Gamma X/Q dispersion factors for evaluating gamma dose rates from a sector averaged finite noble gas cloud (multiple energy undepleted source); and 4.
0/0 deposition factors for evaluating dry deposition of elemental radioiodines and other particulates.
Gama 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 imp'emented through the definition of an T
ef f ective gama atmospheric dispersion f actor, (X/Q ) (Ref erence 8, Section 6), and the replacement of X/Q in infinite cloud dose equstions by the T
(X/Q ].
7.3.3 Lona-Term Average Atmospheric Oispersion Factors for Receptors Actual measured meteorological data for a two-year period, April-1979 through March-1980, and June-1980 through May-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 sumarized in Tables 8.7-4 and 8.7-5.
8.7-2b
I i
TABLE B.7-4 Scabrook Station Dilution Factors
Dose to Critical Dose Rate to Individual Dose to Air Organ Total Body Skin Critical Organ Gamma Beta Thyroid i
r X/Q depleted (**C) 1.3E-06 1.3E-06 a
X/Qundepleted(s{c) 1.4E-06 1.4E-06 t
l D/Q (
)
3.lE-09 3.lE-09 l
l r
I u
l X/QT ()
6.2E-07 6.2E-07 6.2E-07 3
I I
4 j
North site boundary, 916 meters from Containment Building i
.i I
3 i
l I
i
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 Gamma Beta Thyroid Education Center:
(W5W - 335 meters)
X/Q depleted ( * )
6.2E-06 6.2E-06 m
X/0 undepleted (5)
6.7E-06 6.7E-06 m
w D/0 (
)
1.1E-08
~
~
h X/QY {sec) 2.0E-06 2.0E-06 2.0E-06 m
The "Rocks" (ENE - 318 meters)
X/Q depleted ( " )
2.lE-05 2.1E-05 m
X/Q undepleted ( * )
2.3E-05 2.3E-05 m
D/0 (
)
5.0E-08 m
5.9E-06 5.9E-06 5.9E-06 (y)
T X/0 m
9 O
O
8.0 BASES FOR LIOUID AND GASEOUS MONITOR SETPOINTS b
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.
1 The derivation of Equation 5-1 begins with the general equation for the response of a radiation monitor:
( 8-1)
C,g S)j R
=
1 (ml ) (CDS-"I)
E (cps) =
pCi where:
R
= Response of the monitor (cps)
S)g
= Detector counting ef ficiency for radionuclide "i" (eps/(pCi/ml))
l C,$
= Activity concentration of radionuclide "i' in mixture at the monitor (uci/mi)
The detector calibration procedure for the liquid waste test tank monitor at Seabrook Station establishes a counting ef ficiency by use of a known calibration source standard and a linearity response check. Therefore, in Equation 8-1 one may substitute S; for S)4, where S) is the detector counting efficiency determined from the calibration procedure. Therefore.
Equation 8-1 becomes:
S)
C (8-2)
R
=
mi 1
uci (cps) = (cos-ml) gl) 901 m
B.8-1
The MPC for a given radionuclide must not be exceeded at the point of discharge. When a mixture of radionuclides is present, 10CFR20 specifics that the concentration at the point of discharge shall be limited as follows:
pf 11 (8-3) i i
uti-ml Iml uC1) where:
C
= Activity concentration of radionuclide 'i' in the mixture at di the point of discharge (pCi/ml) g = MPC for radionuclide 'i' f rom 10CFR20 Appendix B, Table II, MPC Column 2 (uci/ml)
The activity concentration of raa1onuclide 'i' at the point of discharge is related to the activity concentration of radionuclide 'i' at the monitor as follows:
[F (8-4)
C C
=
di mg d
U uci (mi ) * (ml ) {my gpm where:
di = Activity concentration of radionuclide 'i' in the mixture at the C
point of discharge (uti/ml)
F,
= Flow rate past monitor (gpm)
F
= Flow rate out of discharge tunnel (gpm) d 9
B.8-2
in Equation 8-3 and Substituting the right half of Equation 8-4 for Cdi O
solving for F /F, yields the minimum dilution f actor needed to comply with d
Equation 8-3:
ain I 1
(8-5)
DF g
bd IE)
Iml vCi) gpm where:
F
= Flow rate out of discharge tunnel (gpm) 4 F,
= Flow rate past monitor (gpm)
C,3
= Activity concentration of radionuclide "i' in mixture at the monitor (vCi/ml)
MPC
= MPC for radionuclide "i" from 10CFR20, Appendix B, Table II, 4
Column 2 (uCi/ml)
If F /F,is less than DFmin' then the tank may not be discharged until d
either F or F,or both are adjusted such that:
d F d DF (8-5)
I min m
(E) gpm Usually F /F,is greater than DFmin (i.e., there is more dilution than d
necessary to comply with Equation 8-3).
The response of the liquid waste test tank monitor at the setpoint is therefore:
Rsetpoint " 1 0
"I in d.j.
uti ml
,"()()
(cos-ml)
(ml )
O vCi 8.8-3 t
is equal to the fraction of the total contribution of MPC at the where f) 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*f3 I The monitoring system is designed to incorporate the detector This results in an automatic readout in efficiency, 5), into its software.
Since this procedure for uCi/cc or uCi/ml for the monitor response.
converting cps to uCi/mi 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/mi determined by the laboratory analysis.
Therefore, the setpoint calculation for the liquid waste test tank is:
R setpoint 1O in (d.1)
(
II
}
(9Cil mi n.i Basis for the Plant Vent Wide Range Gas Monitor Setooints 8.2 The setooints of the plant vent wide range gas monitors must ensure that Technical Specification 3.II.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:
(8-7)
S C
R
=
g (cpm) =
(C D}C * )(
)
3
)
8.8-4
where:
A R
= Response of the instrument (cpm) 3 94 = Detector counting efficiency for noble gas 'i' (cpm /(vCi/cm ))
5 C,9 = Activity concentration of noble gas 'i' in the mixture at the 3
noble gas activity monitor (vCi/cm )
g, the activity concentration of noble gas 'i' at the noble gas activity C
by dividing by F, the appropriate flow monitor, may be expressed in terms of Q4 In the case of the plant vent noble gas activity monitors the rate.
appropriate flow rate is the plant vent flow rate.
h h
(8-8)
C,g
=
i
( U ) = ( S ) (5'C) 3 5"
3 cm cm O
where:
i I
h = The release rate of noble gas "i' in the mixture, for each noble g
gas listed in Table 8.1-10.
3 F = Appropriate flow rate (cm /sec)
Substituting the right half of Equation 8-8 into Equation 8-7 for C,g yields:
S h
h (8 9)
R
=
gg g
(cpm)
(#U *)(
)(
)
em 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, S can be substituted for S g
gg 8.8-5 j
1 in Equation 8-9, where 5 is the detector counting efficiency determined 9
from the calibration procedure. Therefore, Equation 8-9 becomes:
O h
h
( B-10)
S R
g
=
g 3
(h)
(cpm) = (U
)(
)
Cm The total body dose rate due to noble gases is determined with Equation 3-3:
0.62 b
0F8 (3-3) htb g
4
=
3 Iy_Ci)
Imrem-m )
C Imeem)
IDCi-see) 3 sec pCi-yr yr Cim where:
b
= total body dose rate (mrem /yr) tb 3
0.62
= (1.0E+06) x (6.2E-07) (pCi-sec/uci-m )
1E+06
= number of pCi per 9C1 (pCi/uci) 6.2E-07
= (X/Q)T, maximum annual average gamma atmospheric 3
dispersion factor (sec/m )
h
= As defined above.
g
OFB,
= total body dose f actor (see Table B.1-10) 3 (mrem-m /pci-yr)
A composite total body gama dose f actor, OFB, may be defined such that:
9 B.8-6
=
h=
h 0F8 (8-11) g 3
g g
OF8 3
i 3
mrem-m -
fi
{gy) { mrem-m )
pCi-yr see sec pCi-yr i
Solving Equation 8-11 for 0F8 yields:
g h 0F8 g
g (5-7)
Of8
=
bg i
Technical Specification 3.11.2.1.a limits the dose rate to the total body from noble gases at any location at er beyond the site boundary to 500 mrem /yr. By cetting D equal to 500 mres/yr and substituting 0F8 for DF8 g
4 tb in Equation 3-3, one may solve for [ Qg at the limiting whole body noble gas l
I dose rate:
IO~I2) b=
806 g
0FB, f1) mrem uci-m ) oci-vr )
sec yr-pCi-sec 3
mrem-m Substituting this result f or [ h in Equation 8-10 yields Rtb, the response g
I of the monitor at the limiting noble gas total body dose rate:
h (8-13) 806 5
R DF8
=
tb 9
g ICP*I " Imrem-uti-m ) Icom-cm ) Isee) IDCi-vr )
yr-pC1-sec 901 3
3 em mrem-m The skin dose rate due to noble gases is determined with. Equation 3-4:
)
(3-4) skin i
0Fj O
i O
uti) mrem-sec) mrem yr sec pCi-yr 8.8-7
where:
skin = Skin dose rate (mrem /yr) h,
= As defined above.
= Combined skin dose f actor (see Table 8.1-10) (mrem-sec/pci-yr)
OFj A composite combined skin dose factor, OF', may be defined such that:
(8-14) hg 0Fj OFj h,
=
i i
I IuCi)
(vCi) Imrem-sec) mrem-sec I pC1-yr sec sec pCi-yr Solving Equation 8-14 for DF' yields:
b 0Fj g
i (5-8)
OFj=
i Tecnnical Specification 3.11.2.1.a limits the dose rate to the skin from noble gases at any location at or beyond the site boundary to 3,000 mrem /yr.
equal to 3,000 mrem /yr and substituting 0F' for DFj in By setting Oskin Equation 3-4 one may solve for [ O at the limiting skin noble gas dose rate:
g i
h = 3,000 (8-15) g 0F' i
c uti-vr uti) mrem) mrem-sec) see yr Substituting this result for h in Equation 8-10 yields Rskin, the response g
[
of the monitor at the limiting noble gas skin dose rate:
r Ol i
8.8-8 i
1 1
( 8-16)
R
= 3.000 S
skin g
F 0F'e U
3 uCi-vr com-cm I Iges) gmrem-sec)
(cpm)
(mrem) yr uti c,3 i
As with the liquid monitoring system, the gaseous monitoring system is also designed to incorporate the detector efficiency, S, into its The monitor also converts the response output to a release rate software.
Therefore, (pCi/sec) by using a real time stack flow rate measurement input.
multiplying by the stack flow rate measurement (F), the Equations 8-15 and 8-16 become:
1 (5-5) 806 E
=
tb 0F8e l
fjg, mrem-uci-m )
goCi-vr )
see yr-pCi-sec 3
mrem-m M
R
= 3000 skin uti-vr uti), mrem) mrem-sec) see yr 8.3 Basis for PCCW Head Tank Rate-of-Chance Alarm Setooint 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 from the PCCW System.
For the rate-of-change
-8 alarm, a setpoint based on detection of an activity level of 10 vCi/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 inf ormation obtained f rom the liquid sampling and analysis comitments described in Table A.3-1 of Part A ensure that potential releases f rom the Service Water Syster are known.
Sampling and analysis requirements for the Service Water System extend over various operating ranges with increased 4
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.
8.8-9
1 REFERENCES s/
A.
Regulatory Guide 1.109, ' Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I", U.S. Nuclear Regulatory Commission, Revision 1, October 1977.
B.
Hamawi, J.
N., 'AEOLUS - A Computer Code for Determining Hourly and Long-Term Atmospheric Dispersion of Power Plant Ef fluents and for Computing Statistical Distributions of Dose Intensity From Accidental Releases', Yankee Atomic Electric Company, Technical Report, YAEC-1120, January 1977.
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.
D.
National Bureau of Standards, "Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and in Water for Occupatic9a1 Exposure", Handbook 69. June 5, 1959.
E.
Slade, D.
H., "Meteorology and Atomic Energy - 1968', USAEC, July 1968.
F.
Seabrook Station Technical Specifications.
i D
R-1 I
l
_ _ _ _ _, _ _, ~. _ _ _ _ _ _.

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1.0 INTRODUCTION

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