For CARDS 12 and 16, enter a 1 in space (80) if the tank to be failed is an
evaporator bottoms tank. Otherwise leave space (80) blank.

CARDS 13 and 17 for both waste streams contain the overall system "tank" fac-
tors. The "tank" factors indicate the type of processing the waste has under-
gone prior to its entry into the tank. The tank factors should be entered

as follows:

(1) For a collector tank in either waste stream, without demineralizers upstream of
the tank, values of TF = 1.0 should be entered in the appropriate spaces.

(2) For a collector tank in either waste stream with demineralizers upstream of the
tank, or for an evaporator bottoms tank in either waste stream, TFs based on the
description given in Section 2.b should be entered in the appropriate spaces.

The appropriate spaces for the TFs are:
(1) Enter in spaces (21-28) the TF for iodine.
(2) Enter in spaces (34-41) the TF for cesium and rubidium.

(3) Enter in spaces (47-54) the TF for other nuclides.

The following section explains the use of the parameters in this note and information
given in the SAR/ER to complete data input CARDS 10-17.

2. Explanation of Parameters used in Filling out Cards 10-17

Liquid Waste Flow Rates and Activities
Flow rates and activities are calculated using the waste volumes and activities given
in NUREG-0016, Table 1-2, "BWR Liquid Wastes" (Ref. 12).

These input flows are supplemented by the use of expected flows and activities more
specific to the plant design as given in the SAR/ER. The inlet streams are combined
to form the principal waste streams (drain wastes and regenerant wastes) considered
in this guide, based on the radwaste treatment system described in the SAR/ER.

Input activities are based on the weighted average activity of the composite stream
entering the waste collection tanks.

Tank Factors

The tank factors indicate the type of processing the waste stream has undergone prior
to its entry into the tank. The tank factors provide the capability to consider
radionuclide removal by demineralizers prior to the waste stream input into the tank.
For evaporator bottoms tanks, the tank factors provide the capability to consider the
effects of radionuclide concentration in the evaporator. Therefore, in determining


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Q. = The release rate of radionuclides, i,.in gaseous effluents from free-
standing stack, in uli/sec (per unit, unless otherwise specified.)

Wy = The release rate of radionuclides, i, in gaseous effluent from all vent
releases, in uCi/sec (per unit, unless otherwise specified).

(x7 s* sec/m?. For free-standing stack releases. The highest calculated
annual average relative concentration for any area at or beyond the
unrestricted area boundary.

GQ), = sec/m?. For all vent releases. The highest calculated annual
average relative concentration for any area at or beyond the unrestricted
area boundary.

We = The highest calculated annual average dispersion parameter for estimating
the dose to an individual at the controlling location due to all vent
releases:

Woe sec/m>, for the inhalation pathway. The location is the
unrestricted area boundary in the sector.

WF meters”, for the food and ground plane pathways. The location

is the unrestricted area boundary in the ___ sector.

W. = The highest calculated annual average dispersion parameter for estimating
the dose to an individual at the controlling location due to free-standing
stack releases:

We = sec/m3, for the inhalation pathway. The location is the
. Unrestricted area boundary in the sector.
We = meters 2, for the food and ground plane pathways. The location

is the unrestricted area boundary in the __. Sector."

SPECIAL NOTES: (1) If there are no free-standing stacks, the factors denoted by the sub-
script, s, need not be considered. (2) In all cases, the tritium releases use the first W
parameter, based on relative concentration (sec/m3). (3) All radioiodines are assumed to be
released in elemental form. If analysis includes the capability of determining elemental and
nonelemental forms in all releases, the food pathway parameters may be adjusted accordingly.

The Specification is applicable to the location (unrestricted area boundary or beyond),
characterized by the values of the parameters Vo B; or (x7@) which result in the maximum
total body or skin dose commitment. In the event that the analyses indicates a different
location for the total body and skin dose limitations, the location selected for considera-
tion shall be that which minimizes the allowable release values.

The factors K;, L;, and M; relate the radionuclide airborne concentrations to various
dose rates assuming a semi-infinite cloud. These factors may be taken directly from
Table B-1 of the Regulatory Guide 1.109 (Ref. 6), if the values therein are multiplied by
10® to convert picocuries ! to microcuriesTM! as used in the above equations. A tabulation
of these factors should be included in the ODCM.

23




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NOTICE
Availability of Reference Materials Cited in NRC Publications
Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room, 1717 H Street, N.W.
Washington, DC 20555

2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082,
Washington, DC 20013-7082

3. The National Technical Information Service, Springfield, VA 22161

Although the listing that follows represents the majority of documents cited in NRC publications,
it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Docu-
ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection
and Enforcement bulletins, circulars, information notices, inspection and investigation notices;
Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and
licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from the GPO Sales
Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and
NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of
Federal Regulations, and Nuclear Regulatory Commission Issuances.

Documents available from the National Technical Information Service include NUREG series
reports and technical reports prepared by other federal agencies and reports prepared by the Atomic
Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open {literature items,
such as books, journal and periodical articles, and transactions. Federal Register notices, federal and
state legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference
proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request
to the Division of Information Support Services, Distribution Section, U.S. Nuclear Regulatory
Commission, Washington, DC 20555.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process
are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available
there for reference use by the public. Codes and standards are usually copyrighted and may be
purchased from the originating organization or, if they are American National Standards, from the
American National Standards Institute, 1430 Broadway, New York, NY 10018.



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is calculated as follows:

F< 7 .5x107!2p yCi/m.

If F is measured at some value in excess of the limiting value (the limiting value
‘is 4x10® ml/sec in this example), then c may be established proportionately. If
F = 8x10® ml/sec, the alarm setpoint is:

c = 7.5x10°12F (yCi/ml per m1/sec)(ml/sec)
= 7.5x10712(8x106) = 6x10° yCi/m).

In this case, the alarm setpoint for the radioactive liquid effluent line monitor can be
established at 6x10°5 uCi/ml, provided that an automatic isolation/control trip action occurs
to satisfy the condition:

c/F < 7x5x107!2 yCi/ml per m1/sec.

Example 2

By using a constant capacity dilution pump (on the dilution stream prior to a mixing
box), the value of (F) is fixed; therefore, the setpoints to be given in the ODCM are:

f< ml/sec = CF/c
Fe m/sec (fixed)
ce x(1/f)uCi/ml < CF/F

If C = 3x1078 pCi/ml, F = 4x108 ml/sec and f < 4000 ml/sec, the radiation monitor setpoint
is calculated as follows:

c< CF/f

= (ant? f 4e)0°) = g.22(1/F) uci/m.

If f is measured at some value less than the limiting value (the limiting value is 4000 ml/sec
in this example), then c may be established proportionately. If f = 1000 ml/sec, the alarm
setpoint is:

¢ ® 0.12(1/f) (nCi/sec) (sec/m1)
5 pais = 1.2x107% yCi/ml.

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The alarm/trip setpoint for a liquid effluent radiation monitor should be determined
based on the instantaneous concentration limits of 19 CFR Part 20, Appendix B, Table II,
Column 2, and are to be applied at the point of discharge for that stream, pipe or conduit
into the unrestricted area, as defined by Figure 31-1." The alarm/trip setpoint should not
consider dilution, dispersion or decay in the unrestricted area beyond the release point. An
isolation control valve or system shall be provided on the discharge line prior to the release
point to permit termination of radioactive releases prior to exceeding the instantaneous
concentration limits of 10 CFR Part 20. The isolation and radiation monitoring points are
to be located upstream of the release point far enough to assure that the time lag between
the established alarm and isolation of the release will not permit a release exceeding these
limits. If the stream is diluted by non-radioactive effluents, and the stream dilution and
effluent isolation control system is in the exclusion area, the monitor's alarm/trip setpoint
may be determined by considering the known in-plant dilution. In-plant dilution is the ratio
of the total release rate at the release point into the unrestricted area to the release rate
of the undiluted stream, and should be based on continuous measurement of these liquid flows.
In such cases, alarm/trip setpoints should also be provided on the flow or level instrumenta~
tion with indicaticn in the main control room. The minimum or actual instantaneous in-plant
dilucion ratio on which the liquid effluent radiation monitor alarm/trip setpoint has been
based, should be continuously measured to aid prompt corrective action to satisfy Specifica-
tion 3.11.1.1.

Conservative assumptions may be included in establishing setpoints to account for such
system variables as the control and measurement system efficiency and detection capabilities
during normal and anticipated operating conditions, the effects of multiple release points
with common or shared in-plant dilution, variability of dilution flow and principal radio-
nuclide composition, and the time lag between alarm/trip action and final isolation of the
radioactive effluent. A record of analyses showing current spectra of radionuclides used to
calibrate radiation monitors should be maintained in the plant records.

The instruments listed in Table 3.3-11** should also be included in Table 4.3-11** to
provide the instrument surveillance requirements, such as calibration, source checking,
functional testing and channel checking.

4.2 REQUIREMENT FOR IMPLEMENTING 10 CFR PART 26

In preparing proposed Technical Specifications, Figure 3.11-1* should consist of a map
of the site area, showing the unrestricted area boundary for iiquid effluents, as defined
in 10 CFR 20.3(a)(17). Guidelines for preparing the figure are contained in Section 2.1.1
of Regulatory Guide 1.70 (Ref. 21).

Standard Technical Specification 3.11.1.1 specifies that:

*. .
This Figure is to be included in proposed Technical Specifications.
see footnote on page 12.

13


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b. During any calendar year:

5 ~ ~ ~ ~
3.17 x 10 } Ri TWO, + MS + WOiy * Mydiy] < 15 mrem
where:
~
Q = The releases of radionuclides, radioactive materials in particulate
form, and radionuclides other than noble gases in gaseous effluents,
i, for long term releases greater than 500 hrs/yr, in uCi. Releases
shall be cumulative over the calendar quarter or year as appropriate.
a, = The releases of radionuclides, radioactive materials in particulate

form and radionuclides other than noble gases in gaseous effluents, i,
for short term releases equal to or less than 500 hrs/yr, in uli.
Releases shall be cumulative over the calendar quarter or year as
appropriate,

W = The dispersion parameter for estimating the dose to an individual at
the controlling location for long term releases (greater than 500 hrs/yr):

W= (x7Q) for the inhalation pathway, in sec/m3.
W = (D7Q) for the food and ground plane pathways in meters 2.
w = The dispersion parameter for estimating the dose to an individual at

the controlling location for short term releases (equal to or less
than 500 hrs/yr):

w= (x/7q) for the inhalation pathway in sec/m?.
w= (D/q) for the food and ground plane pathway in meters 2.

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

R, = The dose factor for each identified radionuclide, i, in m2(mrem/yr) per
uCi/sec or mrem/yr per pCi/m>.

For the direction sectors with existing pathways within 5 miles from the unit, use
the values of R; for these pathways. If no real pathway exists within 5 miles from
the center of the building complex, use the cow-milk Ri assuming that this pathway
exists at the 4.5 to 5.0 mile distance in the worst sector. If the R; for an
existing pathway within 5 miles is less than a cow-milk Ry at 4.5 to 5.0 miles, then
use the value of the cow-milk R at 4.5 to 5.0 miles. The pathway values used for
calculating dose contributions shall be consistent with the results of the land use
census performed pursuant to Specification 3.12.2. The controlling value of R; for
each radionuclide shall be determined and provided in tabular form in the ODCM.

The parameters W and w shall correspond to the applicable pathway location.

SPECIAL NOTES: (1) If there is no free-standing stack, the factors denoted by the subscript, s,
need not be considered. (2) In all cases, the tritium releases use the first W or w parameter,
based on relative concentration (sec/m?). (3) All radioiodines are assumed to be released in
elemental form. If analysis includes the capability of determining the elemental and non-
elemental forms in all releases, the food pathway parameters may be adjusted accordingly.

The following information is provided to further clarify the application of these Speci-
fications and provide more information regarding the individual factors. The ODCM should
include a detailed presentation of the calculational model and a complete tabulation of all
values assigned to each parameter.

30


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This Specification provides impetus to maintain and use the gaseous radioactive waste
management system as required in 10 CFR 50.36a. Maintenance and use of the gaseous radio-
active waste management system components are discussed in Section 2.2 of this manual.
Since features; such as, free-standing stacks or elevated vents are not considered to be
treatment systems, the NRC staff considers that BWR offgas systems must be used without
bypassing normal treatment during reactor operation.

To determine if use of the installed equipment, other than in BWR offgas systems, is
necessary, the licensee should calculate the expected dose to an individual in the unrestricted
area by projecting the plant's cumulative gaseous effluent release over a 31-day period. These
releases should include all potentially radioactive plant gaseous effluents from all gaseous
radioactive waste management systems and ventilation exhaust treatment systems. Calculations
should include a margin, based on operating data, for anticipated operational occurrences and
should use the dose calculation models discussed in Section 5.3 of this manual. The dose
impact from the projected 31-day release should be compared to 0.2 mrad for gamma radiation,
0.4 mrad for beta radiation, or 0.3 mrem to any organ. If the projection indicates these
values will be exceeded, then installed radwaste treatment system components, which are capable
of reducing the quantities or concentrations of radioactive materials in gaseous effluents and
which are capable of reducing the projected impact to less than the values specified above,
must be used. The values for the projected impact, given above, corresponds to approximately
one forty-eighth of the annual design dose objective values of Appendix I, Section II.B and I1.C
of 10 CFR Part 50 in a month, and if continued for a year, these values would correspond to less
than one-fourth the values limited by Specifications 3.11.2.2.b and 3.11.2.3.b. The calculation
of projected cumulative dose impact that could result from the proposed operation may use the
methodology provided in Section 5.3 of this manual.

5.5 SPECIFICATION ON EXPLOSIVE GAS MIXTURE LIMITATION

The Standard Technical Specifications for BWRs and PWRs contain Specification 3.11.2.6A
and alternate Specification 3,11.2.6B for the limiting conditions for operation for systems
designed to treat and store radioactive gases which also contain quantities of uncombined
hydrogen and oxygen. Specification 3.11.2.6A applies to a system designed to withstand a
hydrogen explosion, If all components of the system, from containment to the release point,
are designed and tested to 20 times the normal operating pressure, the system is considered to
be designed to withstand a hydrogen explosion. Alternate Specification 3.11.2.6B applies to a
system not considered to be designed to withstand a hydrogen explosion.

The functional name, “waste gas holdup system," has been used in this manual to include
the various system designs found in BWRs and PWRs which serve the same basic purpose, i.e., to
remove radioactive waste gases from the reactor coolant, to treat and hold gases for radio-
active decay, and to monitor and control the radioactive materials in the gaseous waste prior

to final release.

The potentially explosive components of the waste gas holdup system may be effectively
inerted by nitrogen or steam, treated and re-used in the plant or stored and released after
delay. The treatment may involve hydrogen-oxygen recombiners, filters, holdup tanks, decay

37


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where:

a constant of unit conyersion, y0® pCi/uCi.
the cow's consumption rate, in kg/day (wet weight).
the receptor's milk consumption rate for age (a), in liters/yr.

ine agricultural Productivity by unit area of pasture feed grass, in
g/m.

the agricultural productivity by unit area of stored feed, in kg/m?.
the stable element transfer coefficients, in days/liter.
fraction of deposited activity retained on cow's feed grass.

the maximum organ ingestion dose factor for the ith radionuclide for
the receptor in age group (a), in mrem/pCi.
the decay constant for the ith radionuclide, in sec’! .

the decay constant for removal of activity on leaf and plant surfaces
by weathering, 5.73 x 107 sec (corresponding to a 14 day half-life).

the transport time from pasture to cow, to milk, to receptor, in sec.

the transport time from pasture, to harvest, to cow, to milk, to receptor,
nsec.

fraction of the year that the cow is on pasture (dimensionless) .

fraction of the cow feed that is pasture grass while the cow is on pasture
(dimensionless) .

SPECIAL NOTE: The above equation is applicable in the case that the milk animal is a goat.

Milk cattle are considered to be fed from two potential sources, pasture grass and stored
feeds. Following the development in Regulatory Guide 1.109 (Ref. 6), the values of f, and f,
will be considered unity, in lieu of site specific information provided in the annual land
census report by the licensee.

Tabulated below are the appropriate parameter values and their reference to Regulatory
Guide 1.109 (Ref. 6). In the case that the milk animal is a goat, rather than a cow, refer

to Regulatory Guide

1.109 for the appropriate parameter values.

Parameter Value Table (Ref. 6)
r (dimensionless) 1.Q for radiotodine E-15
0.2 for particulates E-15

Fa (days/liter) Each stable element E-1
Yap (liters/yr) - Infant 330 E-5

- Child 330 E-5

- Teen 400 E-5

- Adult 310 E-5
(DFL;) (mrem/pCi) Each radionuclide E-11 to E-14
Yp (kg/m) 0.7 E-15
Y, (kg/m?) 2.0 E-15
t, (seconds) 1.73 X 10° (2 days) E-15
t,, (seconds) 7.78 X 10® (90 days) E-15
9, (kg/day) 50 E-3

33


APPENDIX C
RATAFR LISTING

This appendix contains the program listing in FORTRAN for the RATAFR Code, applicable
to PWR's and BWR's using the input data cards from Appendix A or B of this manual. The
nuclear data library and subroutines are available in card deck form from the Effluent
Treatment Systems Branch, USNRC, (301)492-7775. The remainder of this appendix contains
the RATAFR program listing.

C-1


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D-28


19.

20.

ai.

22.

23,

24.

25.

Title 40, "Protection of Environment," Chapter I, Code of Federal Regulations, Part 190,
Federal Register, Vol. 42, No. 9, pages 2858 to 2861, Washington, be 20402,
January 13, 1977.

"SKYSHINE, A Computer Program for Evaluating Effects of Structure Design on N-16
Gamma-Ray Dose Rates," Radiation Research Associates, Inc., Report RRA-T7209,
Fort Worth, Texas, 76107, November 1972,

Regulatory Guide 4.15, "Quality Assurance for Radiological Monitoring Programs (Normal
Operation) - Effluent Streams and the Environment," U.S. Nuclear Regulatory Commission,
Washington, D.C. 20555, December 1977.

Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear
Power Plants, LWR Edition," Revision 2, USNRC Report NUREG-75/094, Washington, D.C.
20555, September 1975.

International Commission on Radiological Protection, Report of ICRP Committee II on
Permissible Dose for Internal Radiation, ICRP Publication 2, Pergamon Press, New York
70022, 1959.

U.S. Nuclear Regulatory Commission, “Short Term Diffusion Estimates," Section 2.3.4,

Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -
CWR Edition, USNRC Report NUREG-75/087, Washington, D.C. 20555, November 1975.

U.S. Nuclear Regulatory Commission, "Waste Gas System Failure," Section 15.7.1 (Draft
Revision) Standard Review Plan for the Review of Safety Analysis Reports for Nuclear
Power Plants - LWR Edition, USNRC Report NUREG-75/087, Washington, D.C. 20555, 1978.

Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to

Assess Plant Conditions During and Following an Accident," Revision 1, U.S. Nuclear
Regulatory Commission, Washington, D.C. 20555, August 1977.

43


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C-21


For purposes of implementation, the definition of UNRESTRICTED AREA has been expanded as
follows: “any area at or beyond the site boundary access to which is not controlled by the
licensee for purposes of protection of individuals from exposure to radiation and radioactive
materials, and any area within the site bourdary used for residential quarters or industrial,
commercial, institutional and recreational facilities". The UNRESTRICTED AREA boundary may
coincide with the exclusion (fenced) area boundary, as defined in 10 CFR 100.3(a) (Ref. 15),
may include land areas owned by the licensee, provided that occupancy is controlled by the
licensee for the purposes of meeting the requirements of 10 CFR Part 20, but does not include
areas over water bodies.

To assure that the UNRESTRICTED AREA boundary is defined in each license, the Standard
Technical Specifications require two maps, one for liquid effluents (Figure 3.11-1)" and one
for gaseous effluents (Figure 5.1-1)" locating all points surrounding the facility at which
the licensee shall comply with the expanded definition, given above. These boundaries shall
be consistent with those established in the Safety Analysis Report, or the Final Hazard
Summary for the facility. The UNRESTRICTED AREAS, established at or beyond these boundaries,
are also considered in the LIMITING CONDITIONS FOR OPERATION to keep levels of radioactive
materials in effluents as low as is reasonably achievable, pursuant to 10 CFR 50.36a.

RELEASE RATE used in this manual is defined as, "the discharge of radioactive materials
‘in liquid or gaseous effluents per unit time." The "second" is used as the practical report-
‘ing time unit for establishing release rates to show compliance with the requirements of
10 CFR Part 20 and for establishing instantaneous limitations based on potentially radioactive
releases. The "hour" is used as the practical reporting time unit in establishing average
release rates to show conformance with the requirements of 10 CFR Part 50 for noble gas
releases, for gaseous radioactive effluents other than nodle gases (radioiodines and particu-
lates) and for radioactive materials released in liquid effluents. Liquid releases are
further subdivided into batch and continuous releases. Gaseous releases are subdivided into
short- and long-term releases. These gaseous release subdivisions classify cumulative
releases as being either less than or greater than 500 hrs/year, respectively, for caseous
effluents. Further discussion is provided in Sections 3.3, 4.2 and 5.2 of this manual.

RADIOACTIVE WASTE MANAGEMENT SYSTEMS are defined as all process and control equipment
provided to reduce the amount or concentration of radioactive materials (in any form)
released from the facility. The overall systems may be divided into subsystems to handle
the radioactive materials contained in liquid and gaseous streams and in solid waste. Tie
Standard Technical Specifications have adopted nomenclature for systems and components which
are in common use in the industry. In preparing proposed Technical Specifications, the
system and component names may be changed to correspond to the terminology used in the Final
Safety Analysis Report (FSAR) or the Final Hazard Summary, if applicable. Engineered
Safety Feature (ESF) atmospheric cleanup systems are not considered to be radioactive waste
management system components and, therefore, are not within the scope of this manual.

“This Figure is to be included in proposed Technical Specifications.


TABLE OF CONTENTS

Chapter Page
1 INTRODUCTION. .....scecceceeeeeeeceeeeseesteteeseneneeees eer rer bee eeeeee 1
1.1 Purpose..... 1
1.2 Background... 1
2 DEFINITIONS AND STAFF POSITIONS........seeeeeeeseeeeees dee aeeeeeee eererrereess 5
2.1 Definitions 5
2.2 Staff Positions Augmenting Standard Definitions: 5
3 SPECIAL CONSIDERATIONS 7
3.1 Multi-Unit Sites with Shared Radioactive Waste Management Systems 7
3.2 Population Exposure. 7
3.3 Meteorological Data. 8
3.4 National Interim Primary Drinking Water Regulations. 8
3.5 Solid Waste Management System - Process Control Program Q
3.6 Offsite Dose Calculation Manual (ODCM) oe 9
3.7 Identification of Radionuclides in Effluents.. 9
3.8 Environmental Radiation Protection Standards for
Nuclear Power Operations.........ccceeeesecceeseeseceeee reer eeereeey 10
4 — LIQUID EFFLUENTS..........eeeeeeeeeceneeeeeeeeener eens, reer eee ee erereeeeesy 12
4.1 Instrumentation. 12
4.2 Requirement for in 13
4.3. Requirement for Implementing 10 CFR Part 50. 4
4.4 Specification on Radioactivity Contents in Liqu 0 Inks . 7
4.5 Specification on the Use of Liquid Radioactive Waste Wanagenent System.... 18
5 GASEOUS EFFLUENTS........ cece cee eee ec eee eee e ee etee ene seeeeeeneteeeeeneeseneee + 20
5.1 Instrumentation 20
5.2 Dose Limit for Implementing i0 CFR Part 20 21
5.3 Dose Limit for Implementing 10 CFR Part 50.. 27
5.4 Spectfication on the Use of Gaseous Radtoactive Waste Management System. 36
5.5 Specification on Explosive Gas Mixture Limitation. 37
5.6 Specification Unique to LWR Design Features 38

REFERENCES...

APPENDIX A

APPENDIX Bu... cece ce eee ee eect rete n eet e nee e cee ee een e eee en enn eet eee ee eenee teeeeees B-1
APPENDIX C....... PPereeeeeee SOOO eS SST eee eee S eee eee eee ee eer e eee e eee e rere eereeererersy C-1
APPENDIX D.........-4- eer eee ee eerrery erereeereedy de eeeeee see eeeeee seeeeeeeeee D-1
ADDENDUM. .........sseeeeeeseceeeeeene peeeeeeee deeeeeeee beeen eens eee eee eeeeee seeees AA-1


D-52

os MHSTs WNd AVHE on NULLLGGY NT NOTLANG ON DUNTEG


NOBLE GAS SOURCE TERM

Source Term, nCi/sec,
at the main condenser

Approx... evacuation system (SJAE)

Isotope Half-Life Q Decay
Xe-140. 13.7 s 1.1 x 106
Kr-90 33s 9.8 x 105
Xe-139 410s 9.8 x 105
Kr-89 3.2m 4.6 x 105
Xe-137 3.8m 5.3 x 105
Xe-138 14.0 m 3.1 x 105
Xe-135m 15.6 m 9.1 x 10%
Kr-87 76 m 7.0 x 104
Kr-83m. 1.86 hr 1.2 x 10%
Kr-88 2.8 br 7.0 x 10%
Kr-85m, 4.4 br 1.1 x 104%
Xe-135 9.2 hr 7.7 x 10%
Xe-133m 2.3d 1.0 x 103
Xe-133 5.27 d 2.9 x 10"
Xe-131m W.9d 5.2 x 107
Kr-85 10.76 y 7.0 x 102

Total “4.7 x 108
b. For systems which are detonation-resistant (i.e., rupture discs have been

isolated and loop seals are vented back to the main condenser), release from
the main condenser evacuation system exits via the normal release point.

Activity release into the system continues for one hour following the
accident at the Technical Specification limit unless positive means (such
as automatic isolation] are provided to limit the releases from this source.
Radioiodines and activation gases may be ignored.

No deposition during downwind transport occurs.

The total radioactive inyentory (neglecting radioiodines and activation
gases) in any delay lines and from the process equipment is released within
a two-hour period with no decay.

The total noble gas content of all charcoal delay beds is released over a
period of two hours. A fractional release of the particulate inventory on
the charcoal beds should be assumed. The rate of release is equal to the

rate of absorption using the information contained in NUREG-0016 (Ref. 12).

The main condenser air inleakage is 6 scfm for three shell main condensers.

40


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NUREG-0133

Preparation of Radiological
Effluent Technical Specifications
for Nuclear Power Plants

A Guidance Manual for Users of
Standard Technical Specifications

U.S. Nuclear Regulatory
Commission

Office of Nuclear Reactor Regulation

J. S. Boegli, R. R. Bellamy,
W. L. Britz, R. L. Waterfield

Mn s
kk

Reprinted October 1987


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5.3.1 REQUIREMENTS FOR IMPLEMENTING 10 CFR PART 50

The Standard Technical Specifications 3.11.2.2 and 3.11.2.3 implement 10 CFR Part 50,
Appendix I, as follows:

"The air dose in unrestricted areas (see Figure 5.1-1)* due to noble gases
released in gaseous effluents shall be limited to the following:

a. During any calendar quarter, to < 5 mrad for gamma radiation and
< 10 mrad for beta radiation;

b. During any calendar year, to < 10 mrad for gamma radiation and < 20 mrad
for beta radiation;

(The dose design objectives may be reduced based on expected public
occupancy of areas, e.g., beaches and yisitor centers within the
unrestricted area boundary. (For PWRs only) the dose design objectives
may be reduced based on predicted noble gas releases from the turbine
building, if effluent sampling is not provided.)*
“The dose to an individual from radioiodines, radioactive materials in
particulate form, and radionuclides other than noble gases with half-lives
greater than 8 days in gaseous effluents released to unrestricted areas (see
Figure 5.1-1)* shall be limited to the following:
a. During any calendar quarter to < 7.5 mrem; and

b. During any calendar year to < 15 mrem."

The ODCM should provide the mathematical relationships used to implement the Specifica-
tions. The relationships presented below are acceptable for inclusion in the ODCM. If other
methods are selected to implement these Specifications, it is expected that the alternative
method will include the same general features considered below.

The air dose in unrestricted area (see Figure 5.1-1)* due to noble gases released in
gaseous effluents should be determined by the following expressions:

a. During any calendar quarter, for gamma radiation:
ty ny wv ~
=)
3.17 x 108} [mM LG7D,Q4y + GTA) ayy} + £8,955 + 4444] <5 mrad, and

During any calendar quarter, for beta radiation:

oY ~ ~ ~
3.17 x wor} Ny LOD) + G79) aay + (x70 sis * GID 445] < 10 mrad, and

b. During any calendar year, for gamma radiation:

~ ~ ~ ~
3.17 x 1078} [m, LTD, 0;, + GTDyajy] + 18045 + | < 10 mrad, and
During any calendar year, for beta radiation:

~ ~ ~ “
3.17 x 10} Ny LOAD Qy + G7D yay + 7%, + G79),9;,1 < 20 mrad
where:

M. = The air dose factor due to gamma emmissions for each identified noble gas
radionculide, in mrad/yr per pCi/m?.

¥See footnote on page 21.
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dose must be identified. Tables 4.11-1 and 4.11-2** of the Standard Technical Specifications
contain the sampling and analysis programs required for identifying principal radionuclides in
effluents. These tables were compiled using the guidelines of Regulatory Guide 1.21 (Ref. 17)
and reflect current radiochemical analytical methods. Other methods may be necessary to
enhance identification and analysis, as provided by the footnotes to Tables 4.11-1 and 4.11-2.
In lieu of sample-analysis, if the applicant does not consider that the collection, radio-
chemical separation, and analytical methods are technically feasible or practical at the speci-
fied LLD, then the dose limitations in Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3 should
be proportionally reduced by assuming the continued presence and release concentrations of
those radionuclides as determined by the source term (GALE Code, Ref. 11 or 12). For example,
the dose LCO may be reduced based on predicted radioactive materials in gaseous effuents from
PWR turbine buildings if sampling is not provided. For BWRs and PWRs it may be reduced if
carbon-14 analysis is not provided.

3.8 ENVIRONMENTAL RADIATION PROTECTION STANDARDS FOR NUCLEAR POWER OPERATIONS

Standard Technical Specification 3.11.2.5 specifies in the Action that when the cal-
culated doses associated with the effluent releases exceed twice* the limits of any one of the
Specifications 3.11.1.2, 3.11.2.2 or 3.11.2.3, the licensee shall prepare and submit a Special
Report to the Commission and limit subsequent releases such that the dose or dose commitment to
a real individual from all uranium fuel cycle sources is limited to < 25 mrem to the total body
or any organ (except the thyroid, which is limited to < 75 mrem) over 12 consecutive months.
This Special Report shall include an analysis which demonstrates that radiation exposures to
all real individuals from a1] uranium fuel cycle sources (including all liquid and gaseous
effluent pathways and direct radiation) are less than the standards in 40 CFR Part 190,
Environmental Radiation Protection Standards for Nuclear Power Operations (Ref. 18). If
analysis indicates that releases resulting in doses that exceed the 40 CFR 190 Standard could
occur, obtain a variance from the Commission to permit such releases. The Standard Technical
Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3 consider doses to a real individual and apply
to each reactor but do not include any other portion of the uranium fuel cycle or direct
shine from the reactor.

The “Uranium fuel cycle" is defined in 40 CFR Part 190.02(b) as:

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

The following general gujdelines are presented for preparation of the Special Report:

1) determine which uranium fuel cycle facilities or operations, in addition to the
nuclear power reactor units at the site, contribute to the annual dose to the
maximum exposed member of the public. The maximum exposed member of the public for

"This value may be reduced for multi-unit sites depending on staff analysis.

“These Tables are to be included in proposed Technical Specifications.

10


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The concentration of tritium in milk is based on the airborne concentration rather than
the deposition. Therefore, the Re is based on [x/Q]:

RELA] = KKM FO USG(DFL;), [0.75(0.5/H)] (mrem/yr per uCi/n?)

where:

K'" = a constant of unit conversion, 103 gm/kg.
H = absolute humidity of the atmosphere, in g/m?
0.75 = the fraction of total feed that is water.

0.5 = the ratio of the specific activity of the feed grass water to the atmos-
pheric water.

and other parameters and values are given above, The value of H may be considered as 8 grams/
meter, in lieu of site specific information (Ref. 6).

5.3.1.4 Grass-Cow-Meat Pathway Factor, allto/q]

The integrated concentration in meat follows in a similar manner to the development for
the milk pathway, therefore:
oa, ith

O(U.,) f, (-f, fee o t
Mi = ap s if
Ry[D/Q] = K* GF hy Fe(r)(DFL;) 4 [ ~t Ep |e
(m2 ‘mrem/yr per pCi/sec)
where:

Fe = the stable element transfer coefficients, in days/kg.

Uap = the receptor's meat consumption rate for age (a), in kg/yr.

te = the transport time from pasture to receptor, in sec.

th = the transport time from crop field to receptor, in sec.

Tabulated below are the appropriate parameter values and their reference to Regula-
tory Guide 1.109 (Ref. 6).

Parameter Value Table (Ref. 6)
r (dimensionless) 1.0 for radioiodine E-15
0.2 for particulates E-15

Fe (days/kg) Each stable element E-1
Yap (kg/yr) - Infant 0 E-5

- Child 41 E-5

- Teen 65 E-5

- Adult 110 E-5
(DFL) 5 (mrem/pCi) Each radionuclide E-11 to E-14
Yp (ka/m2y 0.7 E-15

34


In each of the processing systems considered, the Code can calculate the tank concentra-
tions in either the collector tank or the evaporator bottoms tank. However, separate
computer runs must be made for these two tank classifications. If it is desired, the Code
can calculate concentrations in one tank of each of the processing systems in one computer
run by making the appropriate entries in CARDS 8-16. If a tank in only one of the pro-
cessing systems is to be considered, then the appropriate entries need only be made in
CARDS 8-12 or 13-16, depending on which system contains the tank of interest.

Five input data cards define the major parameters for the failed tank in the shim bleed
processing system. The first two cards (CARDS 8 and 9) describe the inputs to the tank.
The shim bleed wastes are entered on CARD 8. For reactor designs which combine the shim
bleed with other reactor grade wastes prior to processing, the other wastes are entered as
equipment drain wastes on CARD 9.

Four input data cards define the major parameters for the failed tank in the waste drain
processing system. The first card (CARD 13) describes the inputs to the tank. Essentially
the same information is required on CARDS 8, 9 and 13, on CARDS 10 and 14, on CARDS 11 and
15, and on CARDS 12 and 16.

The following information explains the use of the parameters in this Appendix and informa-
tion given in the SAR/ER to complete the input data cards.

For CARD 8, enter in spaces (42-49) the flow rate (gpd) of the inlet stream. The value of
the shim bleed rate must always be entered, even if the tank being evaluated is not in
this system, since it is used in determining the primary coolant concentrations. Do not
enter inlet waste activity for the shim bleed since the activity for this waste stream is
calculated by the Code.

The following information is required on CARDS 9 and 13 for both of the processing systems
considered in the Code.

(1) Enter in spaces (17-40) the name of the waste stream (e.g., clean wastes).

(2) Enter in spaces (42-49) the flow rate (gpd) of the inlet stream.

(3) Enter in spaces (57-61) the activity of the inlet stream expressed as a
fraction of primary coolant activity (PCA).

For CARDS 10 and 14, the following information is required:

(1) Enter in spaces (27-50) the name of the tank to be failed (e.g., shim bleed
collector).

(2) Enter in spaces (65-73) the volume (gallons) of the tank to be failed. If a
tank in either of the two processing streams is not being considered, leave

spaces (65-73) blank for that tank.

For CARDS 11 and 15, enter a 1 in space (80) if the tank to be failed is an evaporator
bottoms tank. Otherwise leave space (80) blank.

A-2


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CARDS 12 and 16 for both waste streams contain the overall system "tank" factors. The
"tank" factors indicate the type of processing the waste has undergone prior to its entry
‘into the tank. The tank factors should be entered as follows:

a)

(2)

For a collector tank in either waste stream, without demineralizers upstream of
the tank, values of TF=1.0 should be entered in the appropriate spaces.

For a collector tank in either waste stream with demineralizers upstream of the
tank, or for an evaporator bottoms tank in either waste stream, TF's based on
the description given in Section 2.a(3) should be entered in the appropriate
spaces,

The appropriate spaces for the TFs are:

qa)
(2)
(3)

Enter in spaces (21-28) the TF for iodine.
Enter in spaces (34-41) the TF for cesium and rubidium.
Enter in spaces (47-54) the TF for other nuclides.

The following section explains the use of the parameters in this note and information
given in the SAR/ER to complete data input CARDS 8-16.

Explanation of Parameters used in Filling out CARDS 8-16

a. Liquid Waste Flow Rates and Activities

qa)

(2)

Shim Bleed Wastes and Equipment Drain Wastes
The flow rates of the shim bleed waste stream and equipment drain wastes pro-

cessed with the shim bleed and the activity of the equipment drain wastes are
based on information given in the SAR/ER. The activity of the shim bleed wastes
is based on the primary coolant letdown system effluent activity and is calcu-
lated in the Code.

The activity of the combined inlet stream is calculated by the Code based on the
weighted average of the composite stream entering the tanks.

Waste Drain Tank Wastes
Flow rates and activities are calculated using the waste volumes and activities
given in NUREG-0017, Table 1-2, "PWR Liquid Wastes" (Ref. 11).

These input flow rates are supplemented by the use of expected flows and activi-
ties more specific to the plant design as given in the SAR/ER. The individual
streams are combined based on the radwaste treatment system described in the
SAR/ER. Input activities are based on the weighted average activity of the
composite stream entering the waste collection tanks. The input flow rates and
activities are entered in units of gpd and fractions of PCA.


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3. The relative concentration, (379) pga» to be used is described in Section 5.6.1 of
this manual.

Based on these assumptions, the applicant should backcalculate from a whole body dose of

2.5 rem to an individual at the exclusion area boundary to obtain the main condenser evacuation
system rate limit, in yCi/sec, having a noble gas isotopic distribution proportional to the
noble gas source term, given aboye. For licensed BWR facilities that have a Technical Speci-
fication limit, in wCi/sec, based on a 5 rem consequence criteria, a reevaluation of the
specified limit is not required. Guidance on the radiation monitoring instrumentation is given
in Regulatory Guide 1.97 (Ref. 25).

5.6.3 PWR Monitoring of Steam Generator Blowdown Flask Tank Vent

Standard Technical Specification 3.3.3.9, including Tables 3.3-12* and 4.3-12,* requires
that PWRs continuously monitor the steam generator blowdown tank vent for gross noble gas
radioactivity and continuously sample for radioactive iodines and particulates. Many PWRs with
U-tube steam generators direct their blowdown to a blowdown treatment system without venting,
so that item f in Tables 3.3-12 and 4.3-12 is not applicable. Others vent their flash tank to
the main condenser, where the airborne radionuclides are either removed by condensing steam or
drawn into the main condenser evacuation system, where they are then monitored prior to release
to the atmosphere; therefore, this design provision makes item f not applicable. However,
there are several operating PWR’s that direct their btowdown to a flash tank which is vented
directly to the atmosphere. Monitoring these releases presents serious difficulties, due to
the presence of steam in the exhaust. In lieu of a flash tank vent radiation monitor, a
determination of the release of radioiodine-131 via the flash tank yent can be made by calcu-
lating from a measured concentration in the secondary water by the following equation:

W = Clog lfpp(1-SOpry)

Q, 5 The release rate of radioiodine-131, y, from the steam generator
y flash tank vent, in uCi/sec.

Cc = The concentration of radjoiodine-131, y, in the secondary coolant water
averaged over not more than one week, in uCi/ml.

Roop = The steam generator blowdown rate to the flash tank, in ml/sec.

fey = The fraction of blowdown flashed in the falsh tank determined from
a heat balance taken around the flash tank at the applicable reactor
power level.

SQrty = The measured steam quality in the flash tank vent; or an assumed value of
0.85, based on NUREG-0017 (Ref. 11).

If this option is chosen, the applicant shall perform this calcuation every time measure-

ments of secondary water radioiodine concentrations are required by Technical Specifications,

. and the calculated release shall be assumed at this calculated level until the next secondary
water analysis is completed. These calculations shall be provided by the applicant in his

semiannual effluent release report.
¥See footnote on page 20.
yy


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ADDENDUM

Setpoint Calculations

The radiological effluent Technical Specifications require alarm/trip setpoints for
radiation monitors and flow measurement devices for each effluent line. Setpoint values are
to be calculated to assure that alarm and trip actions occur prior to exceeding the limits
of 10 CFR 20 at the release point to the unrestricted area. The calculated alarm and trip
action setpoints to be specified in the ODCM for each radioactive liquid effluent line monitor
and flow measurement device must satisfy the following equation:

where:

C = the effluent concentration limit (Specification 3.11.1) implementing 10 CFR 20 for
the site, in pCi/ml

c= the setpoint, in pCi/ml, of the radioactivity monitor measuring the radioactivity
concentration in the effluent line prior to dilution and subsequent release; the
setpoint, which is proportional to the volumetric flow of the effluent line and
inversely proportional to the volumetric flow of the dilution stream plus the
effluent stream, represents a value which, if exceeded, would result in concentra~
tions exceeding the limits of 10 CFR 20 in the unrestricted area

f= the flow setpoint as measured at the radiation monitor location, in volume per unit
time, but in the same units as F, below

F = the dilution water flow setpoint as measured prior to the release point, in volume
per unit time.

{Note that if no dilution is provided, c < C. Also, note that when (F) is large compared
to (f), then F+f = F.]

The equation is satisfied when the following alarm/trip setpoints are provided for each
effluent line in the ODCM:

f of (in ml/sec; for example).

F ef (in ml/sec; for example).
c <f (in pCi/ml; for example) .

Some plants may be operated using a fixed value for one or more of these three variables,
c, f or F.

Example 1

By using a constant capacity radwaste system discharge pump (on the undiluted stream) the
value of (f) is fixed; therefore, the setpoints to be given in the ODCM are:


1. REPORT NUMBER (Assigned by DOC!
NRC Form 338 U.S. NUCLEAR REGULATORY COMMISSION REPORT NUMBER (Assigned by DOC

BIBLIOGRAPHIC DATA SHEET NUREG-0133

4 TITLE AND SUBTITLE (Add Volume No, if appropriate} 2. (Leave blank}

"Preparation of Radiological Effluent Technical Specifications

for Nuclear Power Plants" A Guidance Manual for Users of 3 RECIPIENT ACCESSION NO

cifications
?, AUTHOR(S) 5. DATE REPORT COMPLETED
7 MONTH YEAR
2. 5. Boegli, R. R. Bellamy, W. L. Britz, ‘October 1978
9 PERFORMING ORGANIZATION NAME AND MAILING ADORESS (Include Zip Code) DATE REPORT tSSUED
Division of Site Safety and Environmental Analysis MONTH [Tvean
U.S. Nuclear Regulatory Commission October 1978
Office of Nuclear Reactor Regulation 6 (Leave blank)

Washington, D.C. 20555

8 (Leave blank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADORESS (Include Zip Code}
10 PROJECT/TASKANORK UNIT NO

N/A
N/A 11, CONTRACT NO
N/A
13. TYPE OF REPORT PERIOD COVERED finclusive dates)
N/A N/A
15 SUPPLEMENTARY NOTES 14. (Leave blank}

16. ABSTRACT (200 words or less}
This guidance manual provides the NRC staff methodology for calculating parameters for
limiting conditions of operation required in the radiological effluent Technical Speci-
fications for light-water-cooled nuclear power plants. It provides guidance in using
the model specifications reported in NUREG-0472 (Rev. 1)*and NUREG-0473 (Rev. 1),*
applicable to operating PWR and BWR licensees, and users of the Standard Technical
Specifications packages available for various vendor designs. The manual addresses the
implementation of the Regulations and current NRC staff positions as related to the
radioactive waste management systems, effluent control and radiological monitoring
programs and provides equations, references, computer codes, and guidelines pertinent
to these limiting conditions for operation.

*0ctober_ 1978
17, KEY WORDS AND DOCUMENT ANALYSIS 17a DESCRIPTORS.

N/A

17b, IDENTIFIERS/OPEN-ENDED TERMS

18 AVAILABILITY STATEMENT 19, SECURITY CLASS (This report] }21 NO. OF PAGES)

Unlimited 20 SECURITY CLASS (This page) | 22 PRICE
$

#U,S.GOVERNMENT PRINTING OFF 1CE:1987~202-292:60279


NUREG-0133

PREPARATION OF
RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS
FOR NUCLEAR POWER PLANTS

A Guidance Manual for Users of Standard Technical Specifications
Editors

S. Boegli

R. Bellamy

L

L

. Britz
. Waterfield

PERG

Principal Investigators

W. C. Burke
F. J. Congel

Other Participating Staff

J. E. Fairobent K, F. Eckerman
P. G. Stoddart J. H. Osloond

L. G. Bell F. M. Akstolewicz
F. P. Cardile D. L. Ondish

J. T. Collins W. E. Kreger

Manuscript Completed: October 1978
Date Published: October 1978

Division of Site Safety and Environmental Analysis
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission


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The B; and V; factors for the radionuclides are based on the finite plume model of
Regulatory Guide 1.109 (Ref. 6). From Equation 6 of Section 6.2 of this Regulatory Guide,
8, can be expressed as:

ek
"4

i

Where:

fa Ay, uw, E 1
ik i Ya “2” pmrad/yr.
My uy Gi a7 sec)

the results of numerical integration over the plume spatial distribu-
tion of the airborne activity as defined by the meteorological condition
of wind speed (uj) and atmospheric stability class k for a particular
wind direction.

a numerical constant representing unit conversions, presented below.
the distance from the release point to the receptor location, in meters.
the mean wind speed assigned to the jth wind speed class, in meters/sec.

the joint frequency of occurrence of the jth wind speed class and kth
stability class (dimensionless).

the number of photons of energy corresponding to the 2th energy group
emitted per transformation of the ith radionuclide, in number/
transformation.

the energy assigned to the 2th energy group, in Mev.

the energy absorption coefficient in air for photon energy E,, in
meters”), %

The constant K follows from Equation 6 of Section C.2.a of Regulatory Guide 1.109 (Ref. 6),

as:

"

The V; factor is

14

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260 mrad(radians)(m3) (transformation) 16 sectors);,,-¢ Ci sec
sec(Mev) (C1) Garpadtans) (10° seq) (3.15x107 yr

2.1x10" mrad (m3) (transformation) /yr(Mev) (uCi) .

computed with conversion from air dose to tissue depth dose, thus:

faA, su, E
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the tissue energy absorption coefficient for photons of energy Ee. in
em? /gm.

the tissue density thickness taken to represent the total body dose
(Sgm/cm?) .

the ratio of the tissue to air absorption coefficients over the energy
range of photons of interest. This ratio converts dose (mrad) to dose
equivalent (mrem) .

The parameter, Pye contained in the radioiodine and particulates Specification 3.11.2.1.b
includes pathway transport parameters of the ith radionuclide, the receptor's usage of the

24


CHAPTER 5
GASEOUS EFFLUENTS
5.1 INSTRUMENTATION
Standard Technical Specification 3.3.3.9 requires that:

"The radioactive gaseous process and effluent monitoring instrumentation channels shown
in Table 3.3-12* shall be OPERABLE with their alarm/trip setpoints set to ensure that
the limits of Specification 3.11.2.1 are not exceeded. The setpoints shall be deter-
mined in accordance with procedures as described in the ODCM, ard shall be recorded in
the station log.”

Table 3.3-12* provides a list of the radioactive gaseous effluent monitoring instru-
mentation needed to comply with the requirements of General Design Criteria (GDC) 60, 63,
and 64 of Appendix A to 10 CFR Part 50. The list includes instrumentation such as radio-
activity monitoring and sampling devices, automatic control devices and essential flow and
level devices that are components of the monitoring systems. The list uses common nomenclature
for the effluent streams; however, the names may be revised as necessary, to conform with a
particular plant's nomenclature. The list may include effluent streams which are not applic-
able to a given plant, or which may be duplicated and, therefore, should be tailored for the
proposed Technical Specifications. Clarification of proposed Technical Specifications should
be provided by a simple drawinc or sketch, showing stream intersections, instrumentation and
control features. Duplicate instrumentation (i.e., instruments that measure different sensor
parameters or ranges) should be listed separately in Tables 3.3-12* and 4.3-12.* The channei
logic should assure that the alarmed trip action is not negated by switching.

The plant procedures should contain a quality assurance program for instruments as
recommended in Regulatory Guide 4.15 (Ref. 20).

5.1.1 Setpoint Determination to be Provided in the ODCM

The alarm/trip setpoint or automatic control trip setpoint for each instrument channel
listed in Table 3.3-12* should be provided and should correspond to a value(s) which repre-
sents a safe margin of assurance that the instantaneous gaseous release limit of Specifica-
tion 3.11.2.1(a) will not be exceeded. For channels with separate alarm and automatic
control trips, the setpoint for the automatic control trip should be the established value
referenced above; the corresponding setpoint for alarm/trip should be established such that
an alarm trip will occur either in advance of the automatic control trip or simultaneously
with the automatic control trip. For channels with alarm trips only, tne setpoint for the
alarm/trip should be the established value referenced above, provided that the manual or
procedural response to the alarm represents a safe margin of assurance that the instantaneous
gaseous release limit of 10 CFR Part 20 will not be exceeded. The alarm/trip setpoint in
the ODCM should list the alarm setpoint for those channels where any trip actions are by

Rape a . . saga
This Table is to be included in the proposed Technical Specifications.
20


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(BR), the breathing rate of the receptor of age group (a), in m/yr.

(DFA; ) 4 = the maximum organ inhalation dose factor for the receptor of age group (a)
for the ith radionuclide, in Hi The total body is considered as

an organ in the selection of (DFA;),.

The breathing rates (BR). for the yarious age groups are tabulated below, as given in
Table E-5 of the Regulatory Guide 1.109 (Ref. 6).

Age Group (a) Breathing Rate (m?/yr
Infant 1400
Child 3700
Teen 8000
Adult 8000

Inhalation dose factors (DFA;) 4 for the various age groups are given in Tables E-7
through E-10 of Regulatory Guide 1.109 (Ref. 6).

5.3.1.2 Ground Plane Pathway Factor, CH [p/q]

aSco/q] = k*K(SF)DEG,[(1-e%4°)/a,] (mn? -mrem/yr per uCi/sec)
where :
K' = a constant of unit conversion, 108 pci/pci.

K" = a constant of unit conversion, 8760 hr/year.
4 = the decay constant for the ith radionuclide, sec”).
t = the exposure time, 4.73x108 sec (15 years).

DFG, = the ground plane dose conversion factor for the ith radionuclide (mrem/hr
per pCi/m?).

SF = the shielding factor (dimensionless) .
A shielding factor of 0.7 is suggested in Table E-15 of Regulatory Guide 1.109
(Ref. 6). A tabulation of OFG; values is presented in Table £-6 of Regulatory Guide 1.109

(Ref. 6).

5.3.1.3 Grass-Cow-Milk Pathway Factor, ne 1p/q]

what
¢, , Oey) fof, (I-fyfgle th] -aite
RIDA] = R'E Flr (OFLA) a +P — |e

(m? -mrem/yr per uCi/sec)

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j. CARD 10-17: Liquid Tank Parameters
Tanks in two separate processing systems are considered in the BWR-RATAFR Code:

(1) Waste Drain Processing System, CARDS 10-13.
(2) Regenerant Solutions Processing System, CARDS 14-17.

In each of the processing systems considered, the Code can calculate the tank concentra-
tions in either the collector tank or the evaporator bottoms tank. However, separate
computer runs must be made for these two tank classifications. If it is desired, the
Code can calculate concentrations in one tank of each of the processing systems in one
computer run by making the appropriate entries in CARDS 10-17. If a tank in only one
of the processing systems is to be considered, then the appropriate entries need

only be made in CARDS 10-13 or 14-17 depending on which system the tank of

interest is in.

Four input data cards are used to define the major parameters for the failed tank in
each of the processing systems. Essentially, the same information is required on the
four input data cards used for each of the processing systems. The instructions given
in this section are applicable to both processing streams, with the following exception.
The inlet waste activity is not entered on CARD 14 for the regenerant solutions wastes
for systems using regenerable condensate demineralizers, since the activity is calculated
by the Code.

The following information explains the use of the parameters in this Appendix and
information given in the SAR/ER to complete the input data cards.

The following information is required on CARDS 10 and 14 for both of the streams con-
sidered in the Code.

(1) Enter in spaces (18-40) the name of the waste stream (e.g., high-purity wastes).

(2) Enter in spaces (42-49) the flow rate (gpd) of the inlet stream.

(3) Enter in spaces (57-61) the activity of the inlet stream, expressed as a
fraction of the primary coolant activity (PCA).

On CARD 14, do not enter the activity of the regenerant solutions waste inlet stream in
spaces (57-61).

For CARDS 11 and 15, the following information is required:

(1) Enter in spaces (27-50) the name of the tank to be failed (e.g., High Purity
Collector) .

(2) Enter in spaces (65-73) the yolume (gallons) of the tank to be failed. If a tank

in either of the two processing streams is not being considered, leave spaces
(65-73) blank for that tank.

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5.2.1.2 Calculation of P; (Ground Plane)

K'K"DFG; (e/a, (m2-mrem/yr per uCi/sec)

J
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K' = a constant of unit conversion, 10° pCi/uCi.

K" = a constant of unit conversion, 8760 hr/year.

a; = the decay constant for the ith radionuclide, sec}.

t = the exposure period, 3.15x107 sec (1 year).
DFG. = the ground plane dose conversion factor for the ith radionuclide
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The deposition rate onto the ground plane results in a ground plane concentration that
is assumed to persist over a year with radiological decay the only operating removal mech-
anism for each radionuclide. The ground plane dose conversion factors for the ith radionuclide,
OFG; are presented in Table E-6 of Regulatory Guide 1.109 (Ref. 6), in units of mrem/hr per
pCi /m?.

Resolution of the units yields:
P; (Ground) = 8.76x109 OFG, (e/a;
5.2.1.3 Calculation of P; (Food)

Q(U,,) -
K'r Veta) Fi, DFL; fe \ytey (m2-mrem/yr per pCi/sec)

=
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K' = a constant of unit conversion, 10° pCi/uCi.

a = the cow's consumption rate, in kg/day (wet weight).

U. = the infant's milk consumption rate, in liters/yr.

Y. = the agricultural productivity by unit area, in kg/m?

F_ = the stable element transfer coefficients, in days/liter.

= fraction of deposited activity retained on cow's feed grass.

OFL; = the maximum organ ingestion dose factor for the ith radionuclide, in
mrem/pCi.
a, = the decay constant for the ith radionuclide, in secTM}.
My = the decay constant for removal of activity on leaf and plant surfaces

by weathering, 5.73x10"7 sec”! (corresponding to a 14 day half-time).

te = the transport time from pasture to cow, to milk, to infant, in sec.

26


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C-28


CHAPTER 3
SPECIAL CONSIDERATIONS

3.1 MULTI-UNIT SITES WITH SHARED RADIOACTIVE WASTE MANAGEMENT SYSTEMS

The Standard Technical Specifications are written on a “per unit" basis, since this is
the format in which operating licenses are issued. When shared radioactive waste management
systems are used by more than one reactor unit on a site, the wastes from all units are mixed
for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to
a specific reactor unit. The licensee should estimate the contributions from each unit based
on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable,
the treated effluent releases may be allocated equally to each of the radioactive waste
producing reactors sharing the treatment system. For determining conformance to LCOs, these
allocations from shared treatment systems are to be added to the releases specifically attrib-
uted to each unit to obtain the total releases per unit.

It is preferred that discharge lines leading to common release points be separately
controlled and monitored, and the licensee should consider instrumentation which will provide
volumetric recording and radiological effluent monitoring, sampling and analyses on a unit
basis. This has been accomplished for some units by continuous representative sampling and
analysis of the streams prior to mixing and continuously monitored and controlled at the
common release point. Multi-unit sites with common release points, but without shared radio-
active waste management systems,* are also required to determine the alarm/trip setpoints in
Sections 4.1 and 5.1 of this manual to assure compliance with the requirements of 10 CFR
Part 20 (Ref. 14) at each release point.

In preparing Technical Specifications for units at adjacent sites (multi-unit stations
with a common boundary), the sites should be considered as a multi-unit site.

3.2 POPULATION EXPOSURE

The Standard Technical Specifications 3.11.1.3 and 3.11.2.4 require the use of equipment
installed in the radioactive waste management system to meet the requirements of 10 CFR 50.36a
(Ref. 1).

To assure that the Technical Specifications consider the population radiation doses,
the use of the installed radioactive waste management system is required when the projected
cumulative doses exceed an appropriate fraction of the individual dose limitations. This
method of establishing use of the radwaste equipment assures that the staff's Appendix I
cost-benefit analysis in the safety evaluation is not invalidated. Sections 4.5 and 5.4 of
this manual provide these values for the Standard Technical Specifications. Guidance on
reporting population exposures is provided in Regulatory Guide 1.21 (Ref. 17).


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the radionuclide concentration in a tank, the type of processing upstream of the tank
must be considered and entered as tank factors on CARDS 13 and 17.

The following factors are considered in calculating overall tank factors for the
systems :

a)

(2)

(3)

(4)

TFs are categorized by radionuclides.
Halogens

Cs, Rb

Other Nuclides

Note: TF of 1 is assumed for tritium.

The system TF is the product of the individual equipment TF in each of the
systems, e.g., the effect of the demineralizer removal, if any, is multiplied
by the effect of the evaporator concentration, if any.

Tank Factors for Demineralizers

The tank factors for demineralizers are entered in the same manner as decon-
tamination factors (DFs) are entered in the BWR-GALE Code. Therefore, the
values used for TFs for demineralizers are the same as those given in NUREG-0016,
Table 1-3, “Decontamination Factors for BWR Liquid Waste Treatment Systems"
(Ref. 11).

Tank Factors for Evaporators
The tank factors for evaporators express the increase in concentration of

radionuclides in the evaporator bottoms resulting from evaporator operation.
The values entered on the cards are the ratio of the evaporator bottoms stream
flow to the evaporator inlet stream flow. Therefore, the TFs for evaporators
are entered as follows:

Evaporator All Nuclides
Waste Drain Stream 0.01
Regenerant Waste Stream 0.05

B-4




Fo = liquid radioactive waste flow per unit
2 discharge structure exit flow per unit x applicable factor

The liquid radioactive waste flow is the maximum flow from all continuous and batch
radioactive effluent releases specified in Table 4.11-1, from all liquid radioactive waste
management systems, per unit. The discharge structure exit flow is the average flow during
disposal from the discharge structure release point into the receiving water body (in an
unrestricted area) per unit. The definition of FY also requires a value to be included in
Specification 3.11.1.2 for the dilution as a result of mixing effects in the near field of the
discharge structure. For plants with once through cooling, the applicable factor is set equal
to one, i.e., nd additional dilution is considered. For plants with cooling towers, onsite
ponds, or lagoons, the factor shall be a number such that the product of the average blowdown
flow to the receiving water body, in cfs and the applicable factor, is 1000 cfs or less. The
1000 cfs figure was selected to correspond to a typical flow for a unit with once-through
cooling water and agrees with the staff method for determining compliance with Appendix I
(Ref. 1) at the OL stage. The value of this applicable factor is to be included in the blank
provided for the term Foe The actual dilution factor value is dependent upon the dilution
available in the near field of the receiving water body; however, the applicable factor is
limited, as stated above.

4.3.1 Dose Factor Related to Liquid Effluents

The above equation for calculating the dose contributions requires the use of a dose
factor A; for each nuclide, i, which embodies the dose factors, pathway transfer factors
(e.g., bioaccumulation factors), pathway usage factors, and dilution factors for the points of
pathway origin. The adult total body dose factor and the maximum adult organ dose factor for
each radionuclide will be used from Table E-11 of Regulatory Guide 1.109 (Ref. 6); thus the
list should contain critical organ dose factors for various organs. The dose factor may be
written:

A, = ky(U,/0, + UpBF, + UyB1,)DF

Ay = composite dose parameter for the total body or critical organ of an adult for
TM nuclide, i, for all appropriate pathways, mrem/hr per uCi/ml.

k, = units conversion factor, 1.14x10° = 10° pci/yci x 105m /kg + 8760 hr/yr.
U. = 730 kg/yr, adult water consumption (fresh water site only).

Up = 21 kg/yr, adult fish consumption (a11 sites).

Uy = 5 kg/yr, adult invertebrate consumption (salt water site only).

BF, = Bioaccumulation factor for nuclide, i, in fish (fresh or salt water site, as
1 applicable), pCi/kg per pCi/1, from Table A-1 of Regulatory Guide 1.109 (Ref. 4).

BI, Bioaccumulation factor for nuclide, i, in invertebrates (salt water only), pCi/kg

per pCi/1, from Table A-1 of Regulatory Guide 1.109 (Ref. 4).

DF; = Dose conversion factor for nuclide, i, for adults in pre-selected organ, t, in
mrem/pCi, from Table E-11 of Regulatory Guide 1.109 (Ref. 4).


APPENDIX A

RATAFR CODE FOR PRESSURIZED WATER REACTORS
Input Cards and Sample Calculation

Parameters Required for the PWR-RATAFR Code

Complete the cards from information given in the Applicant's Safety Analysis and
Environmental Reports and in Section 2 of this Appendix.

CARD 1: Name of Reactor
Enter in spaces (33-60) the name of the reactor.

CARD 2: Thermal Power Level
Enter in spaces (73-80) the maximum thermal power level (MWt) evaluated for safety
considerations in the Safety Analysis Report.

CARD 3: Mass of Coolant in Primary System
Enter in spaces (73-80) the mass of coolant (103 lbs) in the primary system at
gperating temperature and pressure.

CARD 4: Primary System Letdown Rate
Enter in spaces (73-80) the average letdown rate (gpm) from the primary system to
the purification demineralizers.

CARD 5: Letdown Cation Demineralizer Flow Rate

Enter in spaces (73-80) the annual average flow rate (gpm) through the cation
demineralizers for the control of cesium in the primary coolant. The average flow
rate is determined by multiplying the average letdown rate (value entered on Card 4)
by the fraction of time cation demineralizers are in service to obtain the average
cation demineralizer flow rate.

CARD 6: Hydrological Travel Time

Enter in spaces (73-80) the travel time (days) it takes for the liquid waste of a
failed tank to reach the nearest potable water supply or nearest surface water in
an unrestricted area.

CARD 7: Hydrological Dilution Factor

Enter in spaces (73-80) the dimensionless value of:

The annual volume of water flowing past the potable water supply divided by the
total volume of liquid waste in the failed tank (80% of design capacity).

CARDS 8-16: Liquid Tank Parameters

Tanks in two separate processing systems are considered in the PWR-RATAFR Code:
(1) Shim Bleed Processing System, Cards 8-12;

(2) Waste Drain Processing System, Cards 13-16.

A-1


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this evaluation may or may not correspond to the individual considered in the
Technical Specification;

2) determine the total annual dose to this person from all existing pathways and
sources of radioactivity and radiation using the methodologies described in this
NUREG document and applicable references. Where additional information on pathways
and nuclides is needed, the best available information should be used and documented;

3) include direct radiation from the site in the dose determination. An acceptable
method for calculating radiation from the N-16 component of direct radiation is:
SKYSHINE, A Computer Procedure for Evaluating Effects of Structure Design on N-16
Gamma-Ray Dose Rates, Radiation Research Associates, Inc. Report RRA-17209,
November 1972 (Ref. 19).

In addition to N-16, all direct radiation from the plant and storage facilities
should be considered in the dose determination. The direct dose component (including
N-16) may be determined by calculation or actual measurement (e.g., high pressure
ionization chamber). The calculation or actual measurement must be documented in
this Special Report.

The 25 mrem and 75 mrem dose standards are effective December 1, 1979, except for doses
arising from operations associated with the milling of uranium ore which is effective
December 1, 1980.

Further information on the method of implementation of 40 CFR Part 190 is being developed
by the NRC staff.

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pathway media and the dosimetry of the exposure. Pathway usage rates and the internal dosim-
etry are functions of the receptor's age; however, the youngest age group, the infant, will
always receive the maximum dose under the exposure conditions for Specification 3.11.2.1. For
the infant exposure, separate values of Py may be calculated using the PARTS computer program
given in Appendix D of this manual for the inhalation pathway which uses a W parameter based on
(x7@), and the food (milk) and ground pathway which uses a W parameter normally based on (D/Q),
except for tritium, for application in the ODCM. The following sections provide detail on
calculating these Py values for inciusion in the OOCM.

The values of P; are independent of vent and stack release elevation. In the case of
tritium, (x7@) is the W parameter for the food (milk) pathway as well as the inhalation
pathway. As tritium is a weak beta emitter, the ground plane contribution is zero for
tritium. (NOTE: The value for the P, (food) for tritium is 2.4x103 mrem/yr per uCi/m?.) If
the controlling locations for vent and stack releases are different, the controlling location
for vent releases should be used in Specification 3.11.2.1.

Omitting the subscripts for vent and stack releases, the dose rate from the ith radio-
nuclide (except tritium) is:

Pi (inhalation) (x/Q) Q + [P, (food) + P;(ground plane)] (0/Q)Q (mrem/yr)
and for tritium, is:
P, (inhalation) (570Q + P; (food) T7@)Q = 3.0 x 10T7Q)Q (mrem/yr)

5.2.1.1 Calculation of P, (Inhalation)

P, = K'(BR) DFA; (mrem/yr per yCi/m}
where:
k' = a constant of unit conversion, 10° pCi/uCi.
BR = the breathing rate of the infant age group, in m3/yr.
DFA, = the maximum organ inhalation dose factor for the infant age group for

the ith radionuclide, in mrem/pCt. The total body is considered as an
organ in the selection of DFA, «

The age group considered is the infant group. The infant's breathing rate is taken

as 1400 m3/yr from Table E-5 of Regulatory Guide 1.109 (Ref. 6). The inhalation dose factors
for the infant, OFA, are presented in Table E-10 of Regulatory Guide 1.109, in units of

mrem/pCi .
Resolution of the units yields:

P; (inhalation) = 1.4x109 OFA,

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B-7


For those tanks that are determined to be included in Specification 3.11.1.4 and
Tables 3.3-11 and 4.3-11, a curie limit should be determined based on the methodology pre-
sented in Appendices A or B of this manual, using the PWR-RATAFR Computer Code for pressurized
water reactor plants, or the BWR-RATAFR Computer Code for boiling water reactor plants,
respectively. The methodology is based on the calculated radionuclide inventory in the tank
at 80% capacity using a design basis fission product source term of (1) 1% of the operating
fission product inventory in the core being released to the primary coolant for a PWR, or
(2) a fission product release consistent with a noble gas release rate of 100 wCi/MWt-sec
at 30 minutes decay for a BWR. These Computer Codes determine the radionuclide inventory
in a tank that would result in concentrations equal to the limits of 10 CFR Part 20, Appen-
dix B, Table II, Column 2, at (1) the nearest potable water supply and (2) the nearest
surface water supply in an unrestricted area.

By excluding tritium and dissolved or entrained noble gases from the surveillance analy-
ses, since these can be estimated for any licensee event report, Specification 3.11.1.4 should
include the lowest curie quantity of activation and mixed fission products determined for any
tank listed in Specification 3.11.1.4 as the curie limit for all tanks included in that
Specification.

Most operating reactors have required the use of temporary process and storage tanks
during maintenance and service periods, or when temporary solidification equipment is used at
the facility; therefore, the Specification 3.11.1.4 should indicate a "temporary tank." The
curie limit for a temporary tank may be calculated by the above method, but should be limited
to < 10 curies, excluding tritium and dissolved or entrained gases. If the temporary tank is
mobile and not used for more than a calendar quarter, it need not be included in Tables 3.3-11*
and 4.3-11.*

4.5 SPECIFICATION ON THE USE OF LIQUID RADIOACTIVE WASTE MANAGEMENT SYSTEM

Standard Technical Specification 3.11.1.3 specifies that:

"The liquid radwaste treatment system shall be OPERABLE. The appropriate
subsystems shall be used to reduce the radioactive materials in liquid wastes
prior to their discharge when the projected doses due to the liquid effluent
releases to unrestricted areas (see Figure 3.11-1)* when averaged over 31 days,
exceeds 0.06 mrem to the total body or 0.2 mrem to any organ."

The operability of the liquid radioactive waste management system ensures that this
system will be available for use whenever liquid effluents require treatment prior to release
to the environment. The term “liquid radwaste treatment system" involves all of the installed
and available liquid radioactive waste management system equipment, as well as their controls,
power instrumentation, and services that make the system functional. Equipment that is con-
sidered standby or redundant is also included, since their function is to assure operability.
The specification also permits alternate treatment paths using alternate subsystems and equip-
ment to be used in the event that the norma: treatment is inoperable.

*see footnote on page 13.
18


Parameter Value Table (Ref. 6

Y, (ka/n?) 2.0 E15
t, (seconds) 1.73 x 10® (20 days) E-15
t, (seconds) 7.78 x 108 (90 days) E-15
Qe (kg/day) 50 £3

The concentration of tritium in meat is based on its airborne concentration rather than
the deposition, Therefore, the a is based on |y/Q]:

aMEx/0 = KK F QeU, 5 (DFLS) [0.75(0.5/H)] (mrem/yr per uCi/m?)
where all terms are defined above and Section 5.3.1.3 of this manual.

5.3.1.5 Vegetation Pathway Factor, RYID/QI

The integrated concentration in vegetation consumed by man follows the expression
developed in the derivation of the milk factor. Man is considered to consume two types of
vegetation (fresh and stored) that differ only in the time period between harvest and
consumption, therefore:

Yepoy = kK" r Le PHL ySp gith
ina) = x [nim | crip, [of + uSte
(mn? «mren/yr per nCi/sec)
where:
K' = a constant of unit conyersion, 108 pci/nci.

Uy = the consumption rate of fresh leafy vegetation by the receptor in age
group (a), in kg/yr.

ue? = the consumption rate of stored vegetation by the receptor in age group (a),
in kg/yr.

f, = the fraction of the annual intake of fresh leafy vegetation grown locally.
f, = the fraction of the annual intake of stored vegetation grown locally.

t = the average time between harvest of leafy vegetation and its consumption,
‘in seconds.

t, = the average time between harvest of stored vegetation and its consumption,
‘in seconds.
Yy = the vegetation areal density, in kg/m?.

and all other factors are defined in Section 5.3.1.3 of this manual.

Tabulated below are the appropriate parameter values and their reference to Regulatory
Guide 1.109 (Ref. 6).

35


APPENDIX B

RATAFR CODE FOR BOILING WATER REACTORS
Input Cards and Sample Calculation

Parameters Required for the BWR-RATAFR Code

Complete the cards from information given in the Applicant's Safety Analysis and Environ-
mental Reports and in Section 2 of this Appendix.

CARD 1: Name or Reactor (SAR/ER)
Enter in spaces (33-60) the name of the reactor.

CARD 2: Thermal Power Level (SAR/ER)
Enter in spaces (73-80) the maximum thermal power level (MWt) evaluated for safety
considerations in the Safety Analysis Report.

CARD 3: Total Steam Flow Rate (SAR/ER)
Enter in spaces (73-80) the total steam flow rate from the reactor (oe 1bs/hr).

CARD 4: Mass of Coolant in Reactor Vessel (SAR/ER)
Enter in spaces (73-80) the mass of water in the reactor vessel (108 Ibs).

CARD 5: Cleanup Demineralizer Flow (SAR/ER)
Enter in spaces (73-80) the primary coolant flow rate (10° lbs/hr) through the reactor
coolant cleanup system demineralizers.

CARD 6: Condensate Demineralizer Regeneration Time

For deep bed condensate demineralizers, use 3.5 day regeneration frequency per
demineralizer. If ultrasonic resin cleaning is used, assume 7-day regeneration
frequency per demineralizer. Multiply the frequency by the number of demineral-
izers, and enter the calculated number of days in spaces (73-80). For filter/
demineralizers (Powdex), enter zero in spaces (73-80).

CARD 7: Fraction of Feed Water Through Condensate Demineralizer (SAR/ER)
Enter in spaces (73-80) the fraction of feedwater processed through the condensate
demineralizers.

CARD 8: Hydrological Travel Time

Enter in spaces (73-80) the travel time (days) it takes for the liquid waste of a
failed tank to reach the nearest potable water supply or nearest surface water in
an unrestricted area.

CARD 9: Hydrological Dilution Factor

Enter in spaces (73-80) the dimensionless value of:

The annual volume of water flowing past the potable water supply divided by the
total volume of liquid waste in the failed tank (80% of design capacity).

B-1


manual initiation. The method for calculating fixed or adjustable setpoints shall be pro-
vided in the ODCM and auditable records shall be maintained indicating the actual setpoints
used at all times.

The alarm/trip setpoint for any gaseous effluent radiation monitor should be determined
based on the instantaneous (see RELEASE RATE fundamental time units in Section 2.2 of this
manual) concentration limits of 10 CFR Part 20, Appendix B, Table II, Column 1, and are to
be applied at the point at which the discharge leaves the restricted area boundary into an
unrestricted area, as defined by Figure 5.1-1.** The bases for each setpoint should
consider the type of release at the monitor's location, e.g., long-term releases using
the long-term atmospheric dispersion conditions or, if applicable (see Section 3.3 of this
manual), short-term releases using the short-term atmospheric dispersion conditions. An
isolation control valve or system shall be provided on the discharge line upstream of the
release point to permit isolation prior to exceeding the specified release limits.

If the alarm/trip setpoints are based on predetermined factors accounting for atmospheric
conditions, the elevation of the release point should be considered. The symbols used in
the equations in this manual use a subscript (s) for a free-standing stack for elevated
releases (or vents that take the plume out of the building wake) and a subscript (v) for vent
for releases that are not completely elevated. Guidance on the staff's method for estimating
atmospheric transport and dispersion of gaseous effluents in routine releases is provided
in Regulatory Guide 1.111 (Ref. 8). The radiation monitor alarm/trip setpcints for each
release point should be based on the radioactive noble gases in gaseous effluents. It is not
considered to be practicable to apply instantaneous alarm/trip setpoints to integrating radia-
tion monitors sensitive to radioiodines, radioactive materials in particulate form and radio-
nuclides other than noble gases. Alarm/trip setpoints should also be provided in the main con-
trol room for flow measurement devices which are part of continuous monitoring or sampling
systems and should alarm on loss-of-flow or departure from an established flow range. In all
cases, conservative assumptions may be necessary in establishing these setpoints to account
for system variables, such as the control and measurement system efficiency and detection capa-
bilities during normal, anticipated, and unusual operating conditions, che variability in
release flow and principal radionuclides, and the time lag between alarm/trip action and the
final isolation of the radioactive effluent. The current spectrum of radionuclides used to
calibrate radiation monitors should be maintained in the plant records. The instruments
listed in Table 3.3-12* should also be included in Table 4.3-12,* to provide the instrument
surveillance requirements, such as calibration, source checking, functional testing, and
channel checking.

5.2 DOSE LIMIT FOR IMPLEMENTING 10 CFR PART 20

In preparing proposed Technical Specifications, Figure 5.1-1** should consist of a map
of the site area showing the exclusion boundary, as defined in 10 CFR 100.3(a) (Ref. 15)
and the unrestrictea area beundary, as deftned in 10 CFR 20.3(a)(17) (Ref. 14). Guidelines
for this figure are contained in Section 2.1.1 of Regulatory Guide 1.70 (Ref, 21). Details
on the release point locations and significant elevations should be given in Figure 5.1-1.**

a
See footnote on page 20.
“This Figure is to be included in oroposed Technical Specifications.

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13.

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15.

7.

REFERENCES.

Title 10, "Energy," Chapter I, Code of Federal Regulations; Part 50, pages 250 to 327,
U.S. Government Printing Office, Washington, D.C. 20402, January 1, 1977.

U.S. Nuclear Regulatory Commission, "Draft Radiological Effluent Technical Specifications
for PWR's," USNRC NUREG-0472, Revision 1, Washington, D.C. 20555, October 1978.

U.S. Nuclear Regulatory Commission, "Draft Radiological Effluent Technical Specifications
for PWR's," USNRC NUREG-0473, Revision 1, Washington, D.C. 20555, October 1978.

Regulatory Guide 1.16, "Reporting of Operating Information - Appendix A Technical
Speci Fications," Revision 4, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555,
ugust 1975.

Regulatory Guide 10.1, "Compilation of Reporting Requirements for Persons Subject to
NRC Regulations," Revision 3, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555,
May .

Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of
Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appen-
dix I," Revision 1, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555,
October 1977.

Regulatory Guide 1.110, "Cost-Benefit Analysis for Radwaste Systems for Light-Water-
Cooled Nuclear Power Reactors," U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, March 1976.

Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of
Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1,
U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, July 1977.

Regulatory Guide 1.112, "Calculation of Releases of Radioactive Materials in Gaseous and
Liquid Effluents from Light-Water-Cooled Power Reactors," U.S. Nuclear Regulatory Com-
mission, Washington, D.C. 20555, April 1976.

Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and
Routine Reactor Releases for the Purpose of Implementing Appendix I," Revision 1, U.S.
Nuclear Regulatory Commission, Washington, D.C. 20555, April 1977.

U.S. Nuclear Regulatory Commission, “Calculation of Releases of Radioactive Materials in
Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code) ," USNRC
Report NUREG-0017, Washington, D.C. 20555, April 1976.

U.S. Nuclear Regulatory Commission, "Calculation of Releases of Radioactive Materials in
Gaseous and Liquid Effluents from Boiling Water Reactors (BWR-GALE Code) ," USNRC Report
NUREG-0016, Washington, D.C. 20555, April 1976.

U.S. Nuclear Regulatory Commission, "X0QD0Q, Program for the Meteorological Evaluation

of Routine Effluent Releases at Nuclear Power Stations," USNRC Report NUREG-0324,
Washington, D.C. 20555, September 1977.

Title 10, "Energy," Chaper 1, Code of Federal Regulations; Part 20, pages 144 to 172,
U.S. Government Printing Office, Washington, 20402, January 1, 1977.

Title 10, "Energy," Chapter 1, Code of Federal Regulations; Part 100, pages 409 to 421,
U.S. Government Printing Office, Washington, D.C. 20402, January 1, 1977.

Title 40, "Protection of Environment," Chapter 1, Code of Federal Regulations, Part 141,
pages 169 to 182, U.S. Government Printing Office, Washington, D.C. $0402, January 1,
1977.

Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid
Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-
Water-Cooled Nuclear Power Plants," Revision 1, U.S. Nuclear Regulatory Commission,
Washington, D.C, 20555, dune 1974,

42


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N, = The air dose factor due to beta emissions for each identified noble gas
radionuclide, in mrad/yr per »Ci/m?,

3
<
a

sec/m?, For yent releases. The highest calculated annual
average relative concentration for area at or beyond the unrestricted
area boundary for long term releases (greater than 500 hrs/year).

wa. * sec/m?. For yent releases. The relative concentration for
Gx7a v areas at or beyond the unrestricted area boundary for short term releases
(equal to or less than 500 hrs/year).

GI”), = sec/m3. For free-standing stack releases. The highest calcu-
ated annual average relative concentration for areas at or beyond the
unvestricted area boundary for long term releases (greater than 500 hrs/
year).

(x/q is > sec/m?. For free-standing stack releases. The relative concen-
Eration for areas at or beyond the unrestricted area boundary for short
term releases (equal to or less than 500 hrs/year).

Gs = The average release of noble gas radionuclides in gaseous effluents, i,
for short term releases (equal to or less than 500 hrs/year) from the
free-standing stack, in pCi. Releases shall be cumulative over the calen-
dar quarter or year as appropriate.

Gy = The average release of noble gas radionuclides in gaseous effluents, i, for

short term releases (equal to or less than 500 hrs/year) from all vents,
in uli. Releases shall be cumulative over the calendar quarter or year as
appropriate.

~

Qj, = The average release of noble gas radionuclides in gaseous releases, i,
for long term releases (greater than 500 hrs/year) from the free-
standing stack, in uCi. Release shall be cumulative over the calendar
quarter or year as appropriate.

~

Gy = The average release of noble gas radionuclides in gaseous effluents,
i, for long term releases (greater than 500 hrs/yr) from all vents, in
uCi. Releases shall be cumulative over the calendar quarter or year
as appropriate.

B. = The constant for long term releases (greater than 500 hrs/yr) for each
identified noble gas radionuclide accounting for the gamma radiation
from the elevated finite plume, derived in accordance with the dose
methodology in Regulatory Guide 1.109, Appendix B, Section 1, in
mrad/yr per »Ci/sec.

b. = The constant for short term releases (equal to or less than 500 hrs/yr)
for each identified noble gas radionuclide accounting for the gamma
radiation from the elevated finite plume, derived in accordance with

the dose methodology in Regulatory Guide 1.109, Appendix B, Section 1,
in mrad/yr per nCi/sec.

3.17 x 108 = The inverse of the number of seconds in a year.

The dose to an individual from radiotodines, radioactive materials in particulate
form, and radionuclides other than noble gases with half-lives greater than 8 days
in gaseous effluents released to unrestricted areas (see Figure 5.1-1)* should be
determined by the following expressions:
a. During any calendar quarter:

~ ~

~ ~
3.17 x 10 *} R [W,Q,, + waist WOiy + Wid <.7.5 mrem, and

*See footnote on page 21.

29




CHAPTER 2

DEFINITIONS AND STAFF POSITIONS

2.1 DEFINITIONS

Section 1.0, DEFINITIONS, of the Standard Technical Specifications provides standard
definitions of terms and phrases which appear capitalized throughout the specifications.
Standard definitions are provided to assure licensing consistency. When a term or phrase is
used in a limited subject area of the Standard Technical Specifications, it is defined in
the limited subject area and referenced by Specification number, table, figure or footnote.

2.2 STAFF POSITIONS AUGMENTING STANDARD DEFINITIONS

In certain circumstances, terms used in the Standard Technical Specifications are
defined or specified in applicable regulations, such as 10 CFR Part 50 (Ref. 1). Staff
Positions clarifying certain of these definitions are as discussed below.

LIMITING CONDITIONS FOR OPERATION (LCO) is defined in 10 CFR 50.36(c)(2) as "the
lowest functional capability or performance levels of equipment required for safe operation
of the facility." This definition is applicable to the components of the radioactive waste
management systems during normal reactor operations, including anticipated operational
occurrences. When an LCO for a nuclear power reactor is not met, the licensee shall either
shut down the reactor or follow any remedial action permitted by the Technical Specifica-
tions until the LCO can be met. Remedial action by the licensee may include processing by
normal or alternate modes of operation for the control of radioactive effluents using such
existing equipment as may be installed in the radioactive waste management systems.

MAINTENANCE AND USE of the equipment installed in the radioactive waste management
systems is required in 10 CFR 50.34a(c) and in 10 CFR 50.36a(a)(1). The term, MAINTENANCE
AND USE, is applicable to the installed components of the liquid, gaseous, and solid radio-
active waste management systems and to instrumentation installed for the monitoring and
control of potentially radioactive effluents. MAINTENANCE AND USE does not require the
installation of fully redundant systems; however, prudent management procedures, such as
scheduled standby and maintenance periods should be employed. The Standard Technical Speci-
fications specify levels or values above which equipment installed in the radioactive waste
management systems shall be used to enable the licensee to show that he is exerting his best
efforts to maintain levels of radioactive effluents "as low as is reasonably achievable," in
accordance with 10 CFR 50.36a.

UNRESTRICTED AREA is defined in 10 CFR 20.3(a)(17) (Ref. 14), as "any area access to
which is not controlled by the licensee for purposes of protection of individuals from
exposure to radiation and radioactive materials, and any area used for residential quarters."

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In this case, the alarm setpoint for the radioactive liquid effluent line monitor can be
established at 1.2x10TM% pCi/ml, provided that an automatic isolation/control trip action
occurs to satisfy the condition:

cf > 0.12 yCi/sec.
Value of ¢

A detailed description of the method to be used to obtain the value of (c) should be
provided in the ODCM. Since (c) is dependent on the radionuclide distribution, yields,
calibration and the monitor's parameters, each of these variables should be considered and
the fixed or adjustable setpoint method of determination described in the ODCM for each
effluent monitor. This may be accomplished by tabulation. Changes to the ODCM shall be
provided in the SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT.

AA-3




Regulatory Guide 1.113 - "Estimating Aquatic Dispersion of Effluents from Accidental and

Routine Reactor Releases for the Purpose of Implementing Appendix I"
(Revision 1), April 1977. (Ref. 10)

Computer codes used to determine certain values and parameters used with the Regulatory
Guides listed above are described in the following documents:

NUREG-00Q17 - "Calculation of Releases of Radioactive Materials in Gaseous and Liquid
Effluents from Pressurized Water Reactors (PWR-GALE Code)," April 1976
(Ref. 11)

NUREG-0016 - "Calculation of Releases of Radioactive Materials in Gaseous and Liquid

Effluents from Boiling Water Reactors (BWR-GALE Code) " A-rii 1976 (Ref. 12)

NUREG-0324 - "XOQDOQ, Program for the Met2orological Evaluation of Roucine Effluent
Releases at Nucléar Power Stations," September 1977 (Ref. 13)

These guides and computer codes provide acceptable methods of complying with the Com-
mission's regulations; however, conformance with the staff's guidelines on Standard Technical
Specifications, Regulatory Guides, and dose calcuiation methodology is not required. If the
proposed Technical Specifications include mathematical models and parameters that differ
from the methodology used by the staff to calculate set points and release rates or estimate
doses due to releases of radioactive materials in effluents, the parameters and calculations
used shall be substantiated in the Offzite Dose Calculation Manual.

Based on the findings of the Environmental and Safety Hearings, the Commission may
impose certain additional or alternative limiting conditions for operation based on values
that are more restrictive than those determined using this manual or provided in the Standard
Technical Specifications. In such cases, the limiting coiditions will be basez on the decisiong
of the Atomic Safety and Licensing Board rather than the limiting values determined by this
manual or included in the Standard Technical Specifications.


Pipes, charcoal adsorbers, and cryogenic stills. The Specification is provided to ensure
that the concentrations of potentially explosive gases contained in the system are maintained
outside the explosive envelope for hydrogen and oxygen ({.e., less than 4% Hy by volume or
less than 4% 0, by volume). The alternate Specification 3,11.2,6B provides an additional
setpoint limitation to ensure that the automatic dilution, inerting or recombiner control is
functioning to maintain the relative concentration of components of potentially explosive
gas mixtures outside one-half the above flammability limits (j.e., 2% Hy and/or 05). Based
on the design, the licensee should specify the gas to be measured: hydrogen, oxygen or both
hydrogen and oxygen.

5.6 SPECIFICATIONS UNIQUE TO LWR DESIGN FEATURES

The Standard Technical Specifications contain several Specifications unique to certain
design features of PWRs and BWRs; in general, these Specifications contain limiting condi-
tions for operation. The following Sections describe these limitations and the method for
determining the limiting values.

5.6.1 PWR Gas Storage Tank Specification 3.11.2.7

Specification 3.11.2.7 requires that the quantity of radioactive gas in each gas storage
tank at a PWR be limited to a predetermined curie content. It is not applicable to PWRs that
use adsorption units for gas holdup, but is applicable for compressed gas storage and for
cryogenic storage systems. The purpose of this Specification is to assure that, in the event
of an uncontrolled release of the tank contents, the resulting total body exposure to an indi-
vidual at the nearest exclusion area boundary will not exceed 0.5 rem.

Determination of the curie limit should consider the following expression;

500 mrem (3.15 x 107 sec/year)
108 wCi/Ci } 5K, (X70) pga (mrem-sec/uCi -yr)

17 z

where:

17 = The sum quantity of all noble gas nuclides (i) in a gas storage tank
based on a gas mixture resulting from gaseous wastes, in curies.

K = The total body dose factor due to gamma emissions for each identified
noble gas radionuclide, in mrem/yr per uCi/m>. (See Section 5.2.1 of
this manual.)

G7 pga = The relative concentration at the exclusion area boundary used for
evaluation of design basis accidents for ground release conditions, in
sec/m3. Guidelines are provided in Standard Review Plan 2.3.4 (Ref. 23).

Normally the major radioactive nuclide constituent in PWR waste gas storage tanks is
Xe-133. Radiation monitoring and sampling of these tanks should consider the Xe-133
(equivalent) concentration. Plant procedures shall not permit operation with communication
between tanks.

38


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Dy = Dilution factor from the near field area within one-quarter mile of the release
point(s) to the potable water intake for the adult water consumption (fresh
water site oniy).

Inserting the usage factors of Regulatory Guide 1.109 (Ref. 6) as appropriate into the
equation gives the following expressions:

For Fresh Water sites: AY, = 1.14x10°(730/0, + 21BF; )DF ;

For Salt Water sites: AS, = 1.14x105(218F, + SBI, )DF ,

As noted, ail the factors required to calculate the values of A;, should be contained in
the ODCM. The staff's method of calculating dilution factors for aquatic dispersion is pro-
vided in Regulatory Guide 1.113 (Ref. 10). The ODCM should include a detailed presentation
of the calculation model and a tabulation of all values assigned to the parameters in expres-
sions used to implement the Specification 3.11.1.2.

4.3.2 Special Reports

The Standard Technical Specifications 3.11.1.2, 3.11.1.3, 3.11.2.2, 3.11.2.3, 3.11.2.4 and
3.11.3.1 require that action be taken when the Specificaticn cannot be met. The action is in
the form of special reports, in lieu of licensee event reports, indicating the corrective
action to be taken to reduce tne dose impact due to the release of radicactive materials in
liquid effluents.

These special reports should be prepared using the methodology provided in this manual to
determine the dose impact. Such information in the special reports will be used by the staff
in determining if the corrective action proposed by the licensee is adequate to bring the
releases within the design objectives of Appendix I to 10 CFR Part 50. These special reports
may also require submitting additional information as described in Section 3.4 of this manual.

4.4 SPECIFICATION ON RADIOACTIVITY CONTENTS IN LIQUID-CONTAINING TANKS

Standard Technical Specification 3.11.1.4 and Tables 3.3-11* and 4.3-11* list liquid-
containing tanks outside containment that are to be analyzed periodically to verify tnat ihe
radioactivity content (in curies, excluding tritium and dissolved or entrained noble gases) is
below the specified value. Tanks included in this Specification are those that are not sur-
rounded by liners, dikes or walls capable of holding the tank contents and do not have tank
overflow and drains connected to the liquid radioactive waste management system. Indoor tanks
are uot included unless an analysis based on design basis fission product leakage from the
fuel results in radionuclide concentrations in excess of the limits of 10 CFR Part 20,
Appendix B, Table II, Column 2, where leaked fluid is capable of affecting the nearest exist-
ing or known future water supply** in an unrestricted area.

a as
“See footnote on page 12.

+e
"Supply" means a well or surface water intake that is used as a water source for direct

human consumption or indirectly through animals, crops or food processing. "Known future"
water supply means potential wells or surface water intakes which are identified, or may be
reasonably deduced from available information.

V7


CHAPTER 4

LIQUID EFFLUENTS

4.1 INSTRUMENTATION
Standard Technical Specification 3.3.3.8 requires that:

“The radioactive liquid effluent monitoring instrumentation channels shown in

Table 3.3-11* shall be OPERABLE with their alarm/trip setpoints set to ensure that the

limits of Specification 3.11.1.1 are not exceeded. The setvoints shall be determined in

aacoraance with procedures as described in the ODCM, and shall be recorded in the sta-
lon log.

Table 3.3-11* provides a list of radioactive liquid effluent monitoring instrumentation
needed to comply with the requirements of General Design Criteria (GDC) 60, 63, and 64 of
Appendix A to 10 CFR Part 50 (Ref. 1). The list includes instrumentation such as radio-
activity monitoring and sampling devices, automatic control devices, and essential flow and
level devices which are components of the monitoring channels. The list uses common nomen-
clature for the effluent streams; however, the names may be revised, as necessary, to conform
with a particular plant's nomenclature. Deletion of any item listed should be justified.
Clarification of proposed Technical Specifications should be provided by a simple drawing or
sketch showing stream intersections, instrumentation, and control features. Duplicate instru-
mentation (i.e., instruments that measure different sensor parameters or ranges) should be
listed separately. The channel logic should assure that the alarmed trip action is not
negated by switching.

The plant procedures should contain a quality assurance program for instruments as
recommended in Regulatory Guide 4.15 (Ref. 20).

4.1.1 Setpoint Determination to be Provided in the ODCM

The alarm and trip setpoint(s) for each instrument channel listed in Table 3.3-11*
should be provided and should correspond to a value(s) which represents a safe margin of
assurance that the instantaneous liquid release limit of 10 CFR Part 20 is not exceeded. If
the alarm and the automatic control trip are separate devices, the alarm/trip setpoint
in the ODCM should list the separate trip setpoints. The alarm/trip setpoint in the ODCM
should list the alarm setpoint where any trip actions are by manual initiation. The method
for calculating fixed and adjustable setpoints shall be provided in the ODCM and auditable
records shall be maintained indicating the actual setpoints used at all times. For setpoint
calculations, see the Addendum to this manual.

—_———
This Table is to be included in proposed Technical Specifications.

12


CHAPTER 1

INTRODUCTION

1.1 PURPOSE

The purpose of this manual is to describe methods found acceptable to the staff of the
U.S. Nuclear Regulatory Commission (NRC) for the calculation of certain key values required
in the preparation of proposed radiological effluent Technical Specifications using the
Standard Technical Specifications for light-water-cooled nuclear power plants. This manual
also provides guidance to applicants for operating licenses for nuclear power plants in the
preparation of proposed radiological effluent Technical Specifications or in preparing
requests for changes to existing radiological effluent Technical Specifications for operating
licenses. The manual additionally describes current staff positions on the methodology for
estimating radiation exposure due to the release of radioactive materials in effluents and on
the administrative control of radioactive waste treatment systems.

1.2 BACKGROUND

Section 50.36, "Technical Specifications," of 10 CFR Part 50, "Licensing of Production
and Utilization Facilities" (Ref. 1), requires that each nuclear power reactor operating
license issued by the NRC contain Technical Specifications that set forth limits, operating
conditions, and other regulatory requirements imposed on the facility operation for the
protection of the health and safety of the public. Conditions and limitations corresponding
to certain key values which are system-dependent and site-related are to be incorporated in
these Technical Specifications for compliance with 10 CFR 50.36a, “Technical Specifications
on Effluents from Nuclear Power Reactors." Under the provisions of Section 50.36, each
applicant for an operating license is required to submit proposed Technical Specifications
for his facility, including the supporting bases. These are reviewed by the NRC staff to
assure that the proposed Technical Specifications contain such conditions and limitations as
deemed appropriate and necessary; approved Technical Specifications are then included as
Appendix A of the operating license.

Standard Technical Specifications have been developed by the staff for each appropriate
nuclear steam supply system (NSSS) vendor to provide guidance to applicants for the prepara-
tion of proposed Technical Specifications. These are as follows:

NUREG-0212 - “Standard Technical Specifications for Combustion Engineering Pressurized Water
Reactors"

NUREG-0103 - "Standard Technical Specifications for Babcock and Wilcox Pressurized Water
Reactors"


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5.2.1 Implementation of 10 CFR Part 20 - Airborne Releases.

The Standard Technical Specification 3.11.2.1 implements 10 CFR Part 20 as follows:

"The instantaneous dose rate in unrestricted areas (see Figure 5.1-1)** due to radioactive
materials released jn gaseous effluents from the site shall be limited to the foliowing
values:

a. Tne dose rate limit for noble gases shal] be < 500 mrem/yr to the total
body and < 3000 mrem/yr to the skin, and

b. The dose rate limit for all radioiodines and for all radioactive materiais

in particulate form and radionuclides other than noble gases with half
lives greater than 8 days shall be < 1500 mrem/yr to any organ."

The ODCM should provide the mathematical relationships usec to implement the above
specification. The relationships presented below are acceptable for inclusion in the ODCM.
If other methods are selected to implement the specification, it is expected that the alter-
native method will include the same general features considered below.

a. Release rate limit for noble gases:

t Lv, Qs + K (GAD, Qy) < 500 mrem/yr, and
£ Uby GAD, + 1-1 By) Ugg * (Ly + 1-1 My MGTDy Qy)d < 3000 mrem/yr

where the terms are defined below.

b. Release rate limit for all radionuclides and radioactive materials in particu-
Jate form and radionuclides other than noble gases:

g P, [W, Gst Wy Gy) < 1500 mrem/yr
where:

K = The total body dose factor due to gamma emissions for each identified
noble gas radionuclide, in mrem/yr per uCi/m?,

u

4 The skin dose factor due to beta emissions for each identified noble
gas radionuclide, in mrem/yr per wCi/m3,

Me = The air dose factor due to gamma emissions for each identified noble
gas radionuclide, in mrad/yr per uCi/m3 (unit conversion constant
of 1.1 mrem/mrad converts atr dose to skin dose),

P. = The dose parameter for radionuclides other than noble gases for the
inhalation pathway, in mrem/yr per yCi/m? and for food and ground plane
pathways, in m2(mrem/yr per »Ci/sec). The dose factors are based on
the critical individual organ and most restrictive age group (infant) .

Vv. = The constant for each identified noble gas radionuclide accounting for
the gamma radiation from the elevated finite plume, derived in accordance
with the'dose methodology in Regulatory Guide 1.109, Appendix B, Sec-
tion 1, in mrem/yr per uCi/sec.

By = The constant for long-term releases (greater than 500 hrs/yr) for each
identified noble gas radionuclide accounting for the ganma radiation
from the elevated finite plume, derived in accordance with the dose
methodology in Regulatory Guide 1.109, Appendix B, Section 1, in mrad/yr
per uCi/sec.

a
* See footnote on page 21.

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3.3 METEOROLOGICAL DATA

The Standard Technical Specifications consider the historical annual average atmospheric
dispersion condition rather than real time dispersion conditions in determining the LCO for
radioactive materials in gaseous effluents.

Releases are characterized as “long" or "short" cerm, depending on the frequency and
duration of the releases. This characterization permits the matching of the releases to more
appropriate atmospheric diffusion, dispersion and decay conditions.

"Long-term" refers to releases that are generally continuous and stable in release rate
with some anticipated variation (i.e., <50%, based on a running monthly average) in release
rate, such as is experienced in normal ventilation system effluents at nuclear power plants.
Determination of doses due to long-term releases should use the historical annual average
relative concentration (x/%) based on meteorological data summarized, as recommended in
Regulatory Guide 1.111 (Ref. 8).

"Short-term" refers to releases that are intermittent in radionuclide concentrations or
flow, such as releases from PWR gas storage tanks, PWR containment ventings and purges, BWR
drywell purges (See Standard Technical Specification 3.11.3.8), BWR mechanical vacuum pump
exhausts, and systems or components with infrequent use. Short-term releases may be due to
operational variations which result in radioactive releases greater than 50% of the releases
normally considered as long-term. Short-term releases from these sources during normal
operation, including anticipated operational occurrences, are defined as those which occur
for a total of 500 hours or less in a calendar year but not more than 150 hours in any quarter.
Determination of doses due to short-term releases can use the annual average relative concen-
tration (long-term) if it can be demonstrated that past short-term releases were sufficiently
random in both time of day and duration (e.g., the short-term release periods were not depen-
dent solely on atmospheric conditions or time of day) to be represented by the annual average
dispersion conditions. Otherwise, the short-term relative concentration value should be
calculated in accordance with the guidelines provided in NUREG-0324 (Ref. 13) for short-term
release.

Even though "annual average" atmospheric dispersion conditions are used as basis for the
Standard Technical Specifications, “real time" meteorological data should be summarized hour-
by-hour and coupled with the corresponding releases, and the summary should be included in
the SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT.

3.4 NATIONAL INTERIM PRIMARY DRINKING WATER REGULATIONS

Operators of nuclear power plants located on fresh water bodies which are used as sources
of water for drinking water supply systems, are required to make a special report concern-
‘ing the impact on the water supply system due to liquid effluent releases into the water
bodies which are above the value(s) permitted in Specification 3.11.1.2 of the Standard
Technical Specifications. The NRC has no legal responsibility to implement 40 CFR Part 141,
"National Interim Primary Drinking ‘iater Regulations" (Ref. 16), or to assure routine conform
ance to the Act since this is the responsibility of the Environmental Protection Agency and the


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A fraction of the airborne deposition is captured by the ground plant vegetation caver.
The captured material is removed from the yegetation (grass) by both radiological decay
and weathering processes.

The values of a, Yap> and Y are provided in Regulatory Guide 1.109 (Ref. 6), Tables E-3,
E-5, and E-15, as 50 kg/day, 330 liters/day and 0.7 kg/m?, respectively. The value te is
provided in Regulatory Guide 1.109 (Ref. 6), Table E-15, as 2 days (1.73x105 seconds). The
fraction, r, has a value of 1.0 for radioiodines and 0.2 for particulates, as presented in
Regulatory Guide 1.109 (Ref. 6), Table E-15.

Table E-1 of Regulatory Guide 1.109 (Ref. 4) provides the stable element transfer
coefficients, Fue and Table E-14 provides the ingestion dose factors, OFL;» for the infant's

organs. The organ with the maximum value of OFL, is to be used.

Resolution of the units yields:

vf,
P; (food) = 2.4x1020 a DFL, fe7ite] (m? -mrem/yr per yCi/sec)
W

for all radionuclides, except tritium.

The concentration of tritium in milk is based on its airborne concentration rather than
the deposition rate.

Py os KOK SF Q.U, QDFL; [0.75(0.5/H)} (mrem/yr per yCi/m3)

where:
K'" = a constant of unit conversion, 103 gm/kg.
H = absolute humidity of the atmosphere, in gm/m?.
0.75 = the fraction of total feed that is water.
0.5 = the ratio of the specific activity of the feed grass water to the atmos-
pheric water.

From Table E-1 and E-14 of Regulatory Guide 1.109 (Ref. 6), the values of Fn and DFL;
for tritium are 1.0x1072 day/liter and 3.08x10°7 mrem per pCi, respectively. Assuming an
average absolute humidity of 8 grams/meter3, the resolution of units yields:

P, (food) = 2.4x103 mrem/yr per »Ci/m?

for tritium, only.

5.3 DOSE LIMIT FOR IMPLEMENTING 10 CFR PART 50

In preparing proposed Technical Specifications, Figure 5.1-1* should consist of a map
of the site area showing the unrestricted area boundary for gaseous effluents as defined in

Section 5.2 of this manual. Guidelines for this figure are contained in Regulatory Guide 1.70,
Section 2.1.1 (Ref. 21).

¥See footnote on page 21.
27




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water plant operator. This special report is intended for public information and as a tool
to assure awareness by the licensee of the impact of radioactive liquid releases on the
community's water supply system. The impact within the water supply system is dependent on
treatment given to the water taken into the system. The water plant operator is responsible
for providing appronriate treatment to assure that 40 CFR Part 141 requirements are met.
While the operator of the nuclear power plant is not responsible for meeting the requirements
of 40 CFR Part 141 in the water supply system, his success in meeting the requirements of
Specification 3.11.1.2 will assure an environmentally acceptable impact on the water supply
system. The non-radioiogical impact is separately considered in the Appendix B Technical
Specifications.

3.5 SOLID WASTE MANAGEMENT SYSTEM - PROCESS CONTROL PROGRAM

Standard Technical Specification 3.11.3.1 requires the operator of each nuclear power
plant to establish a PROCESS CONTROL PROGRAM for the solid radioactive waste management
system. The purpose of the PROCESS CONTROL PROGRAM is to provide reasonable assurance of the
complete SOLIDIFICATION of processed wastes and of the absence of free water in the processed
waste. At the time the applicant submits proposed Technical Specifications, he should submit
the PROCESS CONTROL PROGRAM for NRC review and approval prior to implementation. The PROCESS
CONTROL PROGRAM should consist of the processing steps and a set of established process param-
eters, which include but are not limited to pH, oi] content, ratio of solidification agent to
influent waste, water content, and ratio of solidification agent to chemical additive for each
type of anticipated waste (filter sludges, spent resins, evaporator bottoms, boric acid solu-
tions, sodium sulfate solutions and filter media). The surveillance requirements in the
Standard Technical Specifications provide the steps to be taken to assure that operation is
within the parameters established by the PROCESS CONTROL PROGRAM. Packaging procedures should
demonstrate conformance with Specification 3.11.3.1. The PROCESS CONTROL PROGRAM required by
the Standard Technical Specifications is to be documented in the operating procedures for each
reactor and available for review by the NRC inspector. A summary of changes to the PROCESS
CONTROL PROGRAM shall be provided in the SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT .

3.6 OFFSITE DOSE CALCULATION MANUAL (ODCM)

Standard Technical Specifications 3.3.3.8 and 3.3.3.9 require the operator of each
nuclear power plant to establish alarm and trip action setpoints for each radioactive liquid
and gaseous effluent release point in maintained, auditable records, determined in accordance
with the OFFSITE DOSE CALCULATION MANUAL (ODCM). The ODCM shall contain the methodology and
parameters to be used in the calculation of offsite doses due to radioactive liquid and gaseous
effluents pursuant to Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3, and the established
limits of Specifications 3.11.1.1 and 3.11.2.1. The ODCM shall be submitted to the NRC with
the proposed Technical Specifications for review and approval by the NRC. Changes to the
ODCM shal1 be provided in the SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT.

3.7 IDENTIFICATION OF RADIONUCLIDES IN EFFLUENTS

In order to determine the radiological impact associated with the release of radioactive
materials in liquid and gaseous effluents, the principal radionuclides contributing to the


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“The concentration of radioactive material released from the site to unrestricted
areas (see Figure 3.11-1)* shall be limited to the concentrations specified in
10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than noble
gases and 2 x 10°" pCi/ml total activity concentration for alt dissolved or
entrained noble gases."

The concentration limits provided in 10 CFR Part 20, Appendix B, Table If, Column 2,
do not include an MPC for noble gases dissolved or entrained in liquid effluents. An MPC
of 2 x 107* uCi/ml has been established, based on the assumption that xenon-135 is the con-
trolling radionuclide; the Xe-135 MPC in air (submersion) was converted to an equivalent.
concentration in water using tne method described in International Commission on Radiological
Protection (ICRP) Publication 2 (Ref. 22). The value of 2 x 10° uCi/ml shall be used
for a mixture of dissolved or entrained noble gases, not otherwise identified in liquid
releases.

To demonstrate that the Specifications are being met, the surveillance requirements
specify that a sampling and analysis program be implemented according to Table 4.11-1 ”
There are two general types of releases: batch and continuous. A batch release is the
discharge of liquid waste of a discrete volume. A continuous release is the discharge of
liquid wastes of a nondiscrete volume; e.g., from a volume or system that has an input flow
during the continuous release. For example, releases from sample monitor tanks are batch,
and releases from steam generator blowdown are continuous. The sampiing and analysis fre-
quency and the type of analysis required by the Standard Technical Specifications is given in
Table 4.11-1 for each type of release. The lower limit of detection is also specified in
Table 4.11-1 for typical in-plant radiochemical analysis equipment. This program meets the
requirements of 10 CFR Part 50, GDC 64, and conforms to the guidelines gives in regulatory
Guides 1.21 (Ref. 17) and 4.15 (Ref. 20).

4.3 REQUIREMENT FOR IMPLEMENTING 10 CFR PART 50

The Stanaard Technical Specification 3.11.1.2 requires that the cumulative dose con-
tributions be determined in accordance with the ODCM at least once per 31 days. The cumula-
tive dose contributions should consider the dose contributions from the maximum exposed indi-
vidual's consumption of fish, invertebrates and potable water, as appropriate. Normally,
the adult is the maximum exposed individual. All of these pathways should be considered in
the calculation unless demonstrated not to be present. For many plant sites, the dose calcu-
lations may be performed assuming conservative dilution factors and receptor locations to
show compliance with the Technical Specification rather than a more rigorous determination.
The relationships presented below are acceptable for inclusion in the ODCM. If other methods
are selected to implement this Specification, it is expected that the alternate method will
‘include the same general features considered below.

*see footnote on page 13.
“See footnote on page 12.


APPENDIX D
COMPUTER PROGRAMS FOR DOSE PARAMETERS

The following computer programs provide the NRC staff method for calculating various
parameters used in the Technical Specifications.

PARTS, a computer program to calculate technical specification dose parameters for
the iodine and particulate portions of gaseous effluents; available from the
Radiological Assessment Branch of the Nuclear Regulatory Commission, Washington,
D.C. 20555.

RABFIN, a computer program to calculate technical specification dose parameters for
the noble gas portion of gaseous effluents; available from the Radiological
Assessment Branch of the Nuclear Regulatory Commission, Washington, D.C. 20555.

LADTAP, a computer program to calculate the doses from radioactive effluents released
to the hydrosphere; the program and a modification to calculate the technical
specification dose parameters for radionuclides in liquid effluents is available
from the Radiological Assessment Branch of the Nuclear Regulatory Commission,
Washington, D.C. 20555,

The remainder of this Appendix contains the PARTS program listing, and modifying the
routines to the LADTAP Code.

LADTAP MODIFICATIONS

An option has been added to the staff's code LADTAP to tabulate the AG factors of
Section 4.3.1 of this manual. Listed below are the routines which have been modified,
the changes are indicated by the ‘CHANGE 1' notation in columns 73-80.

To execute this option the standard LADTAP input is prepared with the following
departures:

1. The 50 mile population (card 3 of the LADTAP input deck) should be set
negative.
2, The release of all nuclides should be set to one Ci/yr.

Only the input data defining the ALARA determination of LADTAP is necessary. Following
the input data deck structure of Enclosure 1 to the LADTAP program, data beyond card
number 7 need not be prepared for the option.


This Specification requires maintenance and use of the liquid radioactive waste manage-
ment system for conformance to 10 CFR Part 50.36a. Maintenance and use of the radioactive
waste management system components are discussed in Section 2.2 of this manual.

To determine if use of the installed equipment is necessary, the licensee must project
the cumulative liquid effluent releases over the ensuing 31 days. These releases should
include all plant effluents from all liquid radioactive waste management and liquid waste
disposal system components that are planned to be operated at the projected capacity and
performance of each component used during the specified time. These releases should include
a margin, based on operating data, for anticipated and unplanned operational occurrences and
should use the methodology discussed in Section 4.3 of this manual. The impact from this
projected cumulative release is to be compared to 0.06 mrem for the total body or 0.2 mrem
for any organ. If the projection indicates these values will be exceeded, then the installed
liquid radioactive waste management system components that will reduce those radioactive
materials in liquid effluents and the projected impact, must be used.

The values for the projected impact, given above, correspond to approximately one
forty-eighth of the design objective values of Appendix I, Section II.A of 10 CFR Part 50 in a
month, and if continued at this rate for a year, they would correspond to less than one-fourth
the values limited by Specification 3.11.1.2.b. The calculations of projected cumulative
dose impact that could result from the proposed operation should use the methodology provided
in Section 4.3 of this manual.


(3) Tank Factors

The tank factors indicate the type of processing the waste ha$ undergone prior
to its entry into the tank. The tank factors provide the capability to consider
radionuclide removal by demineralizers or other treatment equipment prior to the
tank. For evaporator bottoms tanks, the tank factors provide the capability to
consider the effects of radionuclide concentration in the evaporator. Therefore,
in determining the radionuclide concentration in the tank, the type of processing
upstream of the tank must be considered and entered as tank factors on CARDS 12
and 16.

The following factors are considered in calculating overall tank factors for
the systems.

(a) TFs are categorized by radionuclides.
Halogens
Cs, Rb
Other Nuclides
Note: TF of 1 is assumed for tritium.

(b) The system TF is the product of the individual equipment TF in each of the
systems, e.g., the effect of the demineralizer removal, if any, is multi-
ptied by the effect of the evaporator concentration, if any.

(c) Tank Factors for Demineralizers
The tank factors for demineralizers are entered in the same manner as
decontamination factors (DFs) are entered in the PWR-GALE Code. Therefore,
the values used for TFs for demineralizers are the same as those given in
NUREG-0017, Table 1-3, "Decontamination Factors for PWR Liquid Waste
Treatment Systems" (Ref. 11).

(d) Tank Failures for Evaporators
The tank factors for evaporators express the increase in concentration of

radionuclides in the evaporator bottoms resulting from evaporator operation.
The values entered on the CARDs are the ratio of the evaporator bottom
stream flow to the evaporator inlet stream flow. Therefore, the TFs for
evaporators are as follows:

Evaporator All Nuclides
Waste Drain Stream 0.02
Shim Bleed Stream 0.02

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The noble gas Specification 3.11.2,2 is to be evaluated at the location in the unrestric-
ted area where analyses of annual average air doses were found to be maximum. In the event
that the analyses indicate different locations for the beta and gamma limitations, the loca-
tion selected for consideration shall be that which minimizes the allowable release values
due to gamma radiation.

The radioiodine and particulate Specification 3.11.2.3 is applicable to the location in
the unrestricted area where the combination of existing pathways and receptor age groups
indicates the maximum potential exposures. The inhalation and ground plane exposure pathways
shall be considered to exist at all locations. The grass-cow-milk, grass-cow-meat, and vegeta-
tion pathways are considered based on their existence at the various locations. Of the various
age groups, the infant or child receives the largest dose; thus, only these two age groups will
be discussed. It is the intent, however, that a licensee undertake annual surveys of the age
groups and the land use at the various locations about the site, reports these results accord-
ing to Specification 3.12.2 and determines the applicable parameters of R; for tabulation in
the ODCM. The new parameters shall be submitted to the NRC for review and approval prior to
‘implementation.

The My and Ni factors of the noble gas relationship relate the airborne concentration of
the noble gas to the air dose rates assuming a semi-infinite cloud. These factors may be taken
directly from Table B-1 of the Regulatory Guide 1.109 (Ref. 6), and the values therein have
been multiplied by 10€ to convert picocuriesTM to microcuries 1},

The factor, B., is defined in Section 5.2.1 of this manual. The corresponding short-
term factor, b is computed following the same procedure replacing the meteorological variables,
j,k, and 2, for long-term releases with variables for short-term releases using the method-
ology provided in Regulatory Guide 1.111 (Ref. 8), NUREG-0324 (Ref. 13), and NUREG-75/087
(Ref. 23). Such information should be provided in tabular form in the ODCM.

In developing the R yalues, separate expressions are written for each of the potential
pathways. These expressions are similar to those developed in Section 5.2.1 of this manual for
P;, and are denoted by aSpval, Ra, RE[D/QI, alt[o/a) and RY[D/QI, where the superscripts G,
1, C, M, and V refer to ground plane, inhalation, cow's milk, meat and vegetation, respectively.
The ‘argument notation, [ ], indicates the appropriate dispersion parameter, W, to be applied
with the R; factor. Note that the argument is not included in the following expressions. In
the case of tritium, the dispersion parameter, W, is always taken as (x70). The R parameter
is independent of long-term or short-term releases and should be provided in tabular form in
the ODCM.

5.3.1.1 Inhalation Pathway Factor, RiLx/Q]

Riix/ = K'(BR), (DFA;), (mrem/yr per uCi/m3)
where:

K! = a constant of unit conversion, 10°pci/uci.

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The dose contributions for the total time period z 4t, should be determined by
calculation at least once per 31 days and a cumulative summation of these total
body and any organ doses should be maintained for each calendar quarter. These
dose contributions should be calculated for all radionuclides identified in liquid

effluents released to unrestricted areas using the following expression:

m
D = ¢ [A z at, Cz, FoJ
1 q UT gay & Eek

where:

t = the cumulative dose commitment to the total body or any organ, t, from the
m

liquid effluents for the total time period x at,, in mrem.
221

at, = the length of the zth time period over which Cy, and FY are averaged for

all liquid releases, in hours.

C,. = the average concentration of radionuclide, i, in undiluted liquid effluent

during time period At, from any liquid release, in yCi/ml.

‘it = the site related ingestion dose commitment factor to the total body
or any organ t for each identified principai gamma and beta emitter
listed in Table 4.11-1,* in mrem-ml per hr-uCi.
F = the near field average dilution factor for Cy during any liquid
effluent release. Defined as the ratio of the maximum undiluted liquid
waste flow during release to the product of the average flow from the
site discharge structure to unrestricted receiving waters times.
(___ is the site specific applicable factor for the mixing effect of

the discharge structure.)

The term Cay is the composite undiluted concentration of radioactive material in liquid
waste at the common release point determined from the Radioactive Liquid Waste Sampling and
Analysis Program, Table an" in the Standard Technical Specifications. All dilution
factors beyond the sample point(s) are to be included in the FY and A terms .

The term FY is a near field average dilution factor, considering the combined liquid
releases from each unit even if there is more than one release point to the unrestricted area
per unit within one-quarter mile of each other. As described in Section 3.1 of this manual,
multi-unit sites with shared radioactive waste management systems should calculate the total
continuous and batch liquid release concentrations for each reactor. The value of the term Fe
should be determined as:

*see footnote on page 12,

15


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NUREG-0452 - "Standard Technical Specifications for Westinghouse Pressurized Water
Reactors"

NUREG-0123 - "Standard Technical Specifications for General Electric Boiling Water
Reactors"

Current Standard Technical Specifications are available from the National Technical
Information Service (NTIS), Springfield, Virginia, 22161. These Standard Technical Specifi-
cations will contain the radiological effluent Technical Specifications to be used by the
applicant for an operating license. In the interim, model radioactive effluent Technical
Specifications have been provided in NUREG-0472 (Ref. 2) for pressurized water reactors,
and NUREG-0473 (Ref. 3) for boiling water reactors. Table 1.1 provides a summary of those
applicable sections in the Standard Technical Specifications which contain conditions and
limitations relative to the radiological effluent Technical Specifications to be discussed
in this manual.

The Standard Technical Specifications contain the limiting conditions for operation
necessary for complying with the Commission's regulations and are in a format that is accept-
able to the NRC staff. The reporting requirements reflect the guidelines provided in Regu-
latory Guide 1.16, "Reporting of Operating Information - Appendix A Technical Specifications,"
Revision 4, August 1975 (Ref. 4) and Regulatory Guide 10.1, "Compilation of Reporting Require~
ments for Persons Subject to NRC Regulations," Revision 3, May 1977 (Ref. 5).

The methodology discussed in this manual and used to implement the requirements of 10 CFR
Part 50, Appendix I, "Numerical Guides for Design Objectives and Limiting Conditions for
Operation to Meet the Criterion 'As Low As Practicable' for Radioactive Material in Light-
Water-Cooled Nuclear Power Reactor Effluents” (Ref. 1), is consistent with the Regulatory
Guides used in the staff's safety evaluations pursuant to 10 CFR 50.34a(c). These guides
are as follows:

Regulatory Guide 1.109 - "Calculation of Annual Doses to Man from Routine Releases of
Reactor Effluents for the Purpose of Evaluating Compliance with
10 CFR Part 50, Appendix I" (Revision 1), October 1977. (Ref. 6)

Regulatory Guide 1.110 - "Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled
Nuclear Power Reactors," March 1976. (Ref. 7)

Regulatory Guide 1.111 - "Methods for Estimating Atmospheric Transport and Dispersion of
Gaseous Effluents in Routine Releases from Light-Water-Cooled
Reactors" (Revision 1), July 1977. (Ref. 8)

"Calculation of Releases of Radioactive Materials in Gaseous and
Liquid Effluents from Light-Water-Cooled Power Reactors," April
1976. (Ref. 9)

'

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Parameter Value Table (Ref. 6)

r (dimensionless) 1.0 for radioiodines E-1
0.2 for particulates E-1

(DFL;) 4 (mrem/pCi) Each radionuclide E-11 to E-14
uk (kg/yr) ~ Infant 0 £5

> Child 26 E-5

~ Teen 42 E-5

~ Adult 64 E-5
US (kg/yr) ~ Infant 0 E-5

- Child 520 E-5

> Teen 630 E-5

~ Adult 520 E-5
f, (dimensionless) site specific (default = 1.0)
fy (dimensionless) site specific (default = 0.76) (see Ref. 6, page 28)
t (seconds) 8.6 x 104 (1 day) E-15
th (seconds) 5.18 X i0® (60 days) £-15
yy (kg/m) 2.0 E-15

The concentration of tritium in vegetation is based on the airborne concentration
rather than the deposition. Therefore, the Ry is based on [x/Q]:

aYEy/01 = KIKI" ube + ust, (DFL;), [0.75(0.5/H)] (mrem/yr per uci/m?).
where all terms have been defined above and in Section 5.3.1.3 of this manual.

The staff has developed a computer code PARTS for calculating the Ri parameters, which is
described in Appendix D of this manual.

5.4 SPECIFICATION ON THE USE OF GASEOUS RADIOACTIVE WASTE MANAGEMENT SYSTEM

Standard Technical Specification 3.11.2.4 specifies that:

"The gaseous radwaste treatment system shall be OPERABLE. The appropriate subsystems
shall be used to reduce radioactive materials in gaseous waste prior to their dis-
charge when the projected gaseous effluent releases from all release points to
unrestricted areas (see Figure 5.1-1)* would result in a dose in any period of

31 days that exceeds 0.2 mrad for gamma radiation, 0.4 mrad for beta radiation, or
0.3 mrem to any organ for that same 31 day period."

The operability of the gaseous radioactive waste management system ensures that this
system will be available for use whenever gaseous effluents require treatment prior to
release to the environment. The term "gaseous radwaste treatment system" includes all of the
installed and available gaseous radioactive waste management system equipment, as well as
their controls, power, instrumentation and services that make the system functional. Equip-
ment that is considered standby or redundant is also included, since the function is to assure
operability. The action also permits alternate treatment paths using alternate subsystems
and equipment to be used in the event that the normal treatment is inoperable.

*See footnote on page 21.
36


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PARTS INPUT DECK TABLE

Card

No. Format Variable Columns Description
1 20 A4 Name 1-80 Plant Title Card, Name, Docket No. and Plant Type
2 F 5.2 H 1-5 Humidity absolute (default value = 8.0 gr/m?)

F 5.2 YL 6-10 Yield of leafy vegetables for human consumption
(default value = 2.0 Kg/m2)

F 5.2 YV 11-15 Yield of produce other thay leafy vegetables
(default value = 2.0 Kg/m*)

F 5.2 YP 16-20 Agricultural productivity of animal pasture feed
(default value = 0.70 Kg/m¢ wet weight)

F 5.2 Yc 21-25 Agricultural crop productivity of animal feed other
than pasture grass (default value = 2.0 Kg/m2 wet
weight)

F 5.2 Qc 26-30 Milk cow and beef cattle consumption rate for feed
or forage (default value = 50 Kg/day wet weight)

F 5.2 QG 31-35 Goat consumption rate for feed or forage (default
value = 6 Kg/day wet weight)

E 8.2 boq 41-48 Annual average relative deposition rate (D/Q),
determined for a specific plant airborne release
and site location (meters-¢) (Optional Use For
Reference & Information Only)

3 Il TAGE 1 Identification of Controlling Age Group (1 is Infant,
2 is Child, 3 is Teen and 4 is Adult)
Il TORG 2 Identification of Controlling Organ (1 is Thyroid,
2 is the critical organ)

6A4 ZLOC 3-26 Receptor Location identification, Name, Compass
Sector and Distance

F 5.0 GF 27-31 Fraction of yr. humans are exposed to ground surface
radiation (default value = 1.0)

F 5.0 ZIN 32-36 Annual occupancy factor for the inhalation pathway
(default value = 1.0)

F 5.0 FV 37-41 Fraction of yr. leafy vegetables are grown (default
value = 1.0 .

F 5.0 FP 42-46 Fraction of yr. cows are on pasture (default
value = 1.0

F 5.0 FG 47-51 Fraction of produce from local garden (default
value = 0.76)

F 5.0 FPF 52-56 Fraction of daily intake of cows derived from pasture
while on pasture (default value - 1.0)

F 5.0 FeT 57-61 Fraction of yr. goats are on pasture (default
value = 1.0),

F 5.0 FPG 62-66 Fraction of daily intake of goat from pasture while
on pasture (default value = 1.0)

F 5.0 FB 67-71 Fraction of yr. beef cattle are on pasture (default
value 1.0)

F 5.0 FBF 72-76 Fraction of daily intake of beef cattle derived from

pasture while on pasture (default value = 1.0)


Alternate Specification 3.11.2.7, "PWR Waste Gas Processing System," should be used for
plants that use adsorption units for gas holdup prior to release. This Specification requires
that the gross radioactivity in noble gases removed from the waste gas system by means ef steam
jet air ejectors {or other devices) and as measured prior to entering the adsorption systems at
PWR plants shall be limited by a release rate alarm setpoint with indication in the main con-
trol room. The purpose of this pretreatment continuous radiation monitor setpoint is to
provide reasonable assurance that the potential accident total body dose to an individual at
the exclusion area boundary will not exceed a small fraction of the limits specified in 10 CFR
Part 100 in the event this effluent is inadvertently discharged directly to the environment
without treatment. Guidelines for determining the release rate limit are provided in Standard
Review Plan 15.7.1 (Ref. 24), and guidance on the radiation monitoring instrumentation is given
in Regulatory Guide 1.97 (Ref. 25).

5.6.2 BWR Main Condenser Evacuation System Specification 3.11.2.7

This Specification requires that the gross radioactivity in noble gases removed from the
main condenser by means of steam jet air ejectors (or other devices) and as measured prior to
entering the treatment, adsorption and delay systems at BWR plants shall be limited by a
release rate alarm setpoint with indication in the main control room. The purpose of this
pretreatment continuous radiation monitor setpoint is to provide reasonable assurance that the
potential total body accident dose to an individual at the exclusion area boundary will not
exceed a small fraction of the limits specified in 10 CFR Part 100 in the event this effluent
including the radioactivity accumulated in the treatment system is inadvertently discharged
directly to the environment without treatment.

The method for determining the specified rate limit, in yCi/sec, should use the following
assumptions:

1. The release of radioactive material from the fuel is postulated to have an
isotopic composition of noble gases determined from noble gas source term dis-
tribution for a 3500 MWt reactor. These yalues should be scaled linearly for
reactors of higher and lower powers.

2. The assumptions related to the release of radioactive material from the system
following the accident are:

a. For systems which are not fully detonation-resistant or for those systems which
are equipped with rupture discs that have not been isolated or loop seals which
do not yent back to the main condenser, release occurs from a break just down-
stream of the main condenser eyacuation system. Release from the main condenser
evacuation system is assumed to be at ground level and a delay of five minutes
is assumed to account for radioactive decay during transit from the release
point to the exclusion area boundary.

39