ML20037A698

Proposed ETS 2.4 Re Radwaste Treatment & Monitoring

ML20037A698
Person / Time
Site:	Crystal River
Issue date:	10/02/1974
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML20037A696	List: ML20037A694 ML20037A698
References
NUDOCS 8003260709
Download: ML20037A698 (26)

Text

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,S PROPOSED TECllNICAL S' ECIFICATIONS FOP.

P CRYSTAL RIVER, UNIT 3 Docket Number (s) 50-302 2.4 Ll!41 TING CONDITIONS FOR OPERATIE Radioactive Effluents Objective: To define the limits sad conditions for the controlled release of radioactive materials in liquid'and gaseous effluents to the envircus to ensure that these releases are as low as prac ticab).e. These releases should not result in radiation exposures in unrestricted areas greater than a few percent of natural back-ground exposures. The concentration of effluent discharges of radioactivities

. chall be within the limits specified in 10 CFR Part 20.

To ensure that.the releases of radioactive material above background to unrestricted areas be as low as practicable as defined in Appendix I I

to 10 CFR Part 50, the following design objectives apply:

For liquid wastes:

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The annual dose above backgreund to the total body or any organ a.

of an individual from'.a'11 reactors at a site should not exceed 5 mrem in an unrestricted area.

b.

The annual total quantity of radioactive e.aterials in liquid waste, excluding tritium and dissolyhd guaes, discharged from each reactor should not exceed 5 C1.

8003_260 Wf s.

- x u.b' 2-For ::aseous wastes:

The annual total quantity of noble gases above background dis-c.

charged from the site should result in an air dose due to gar:na radiation of less than 10 mrad, and an air dose due to beta radiation of less than 20 mrad, at any location near ground level which could be occupied by individuals at or beyond the boundary of the site.

d.

The annual total quantity of all radioiodines and radioactive material in particulate foms above background from all reactors at a site should not result in an annual dose to any organ of an individual in an unrestricted area from all pathways of exposure in c:: cess of 15 mrem.

I The annual total quantity of iodine-131 discharged from cach e.

reactor at a alte should not exceed 1 C1.

--y 2.4.1 bpecifications for Liquid Waste Effluents The concentration of radioactive materials released in liquid a.

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waste effluents from all reactora at the site shall not c::ceed the values specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for unrestricted areas.

b.

The cumulative release of radioactive mterints in liquid uasta ef fir 20ts.

excluding tritium and dissolved gases, shall not exceed 10 C1/ reactor / calendar quarter.

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'the cu.utlative release of radioactiva matarlais in li<:uid nar.to erfluynts, excluding tritium and dissolved gases, shall not execed 20 ci/ reactor in any 12 consecutive months.

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d.

During release of radioactive wastes, the ef fluent control monitor shall be set to alarm and to initiate the automatic -

closure of each vaste isolation va:Ve prior to exceeding the-limits specified in 2.4.1.a above.

The operability of each sutomatic isolation valve in the liquid radwaste e.

discharge lines shall be demonstrated quarterly.

f.

The equipment installed in the liquid radioactive vaste system shall be maintained and shall be operated to process radioactive.

Liquid wastes prior to their discharge when the projected s

cumulative release could exceed 1.25 Ci/ reactor / calendar quarter, excluding tritium and dissolved gases, g.

The maximum radioactivity to be contained in any liquid radw'aste tank that can be discharged directly to' the environs shall not -

exceed 10 Ci, excluding tritium and dissolved gases.

h.

If the cumulative release of radioactive materials in liquid effluents',

excluding tritium and dissolved gases', exceeds-2.5 ci/ reactor / calendar qua rcer, the licensee shall make an investigation to ide ne.tif the causes for such releases, define and initiate a a proncan of action to reduce such releases to the design objective icvelu itsted in Section 2.4, and report these actions to the Commission within 30 days from the end of the quarter during which the release occurred.

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^ s 2.4.9 Specifications for Liqui _d Waste Sampilng and Monitoring' a.

Plant records shall be maintained of the radioactive concentration and volume before dilution of Liquid waste intended for discharge and the average dilution flow and length of time over wh ich each '

discharge occurred.

Sample analysis results and other reports shall be submitted in accordance with Section 5.6.1 of c:iese Specifications. Estimates of the sampling and analytical errors

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associated with ench reported value shall be included.

b.

Prior to release of each batch of liquid waste, a sample shall be taken fran that batch s.-d analyzed for. the concecration of each significant gamra energy peak in accordance with Table 2.4-1 to demonstrate compliance with Specificatica 2.4.1 using the flow I

rate into which die waste is discharged during the period of, discharge, Sampling and analysis of liquid radioactive waste -shall be c.

performed in accordance with Table 2.4-1.

Prior to taking samples from a monitoring tank, at least two tank voluraes shall be recirculated.

d.

The radioactivity in liquid was tes shall' be continuously monitored and recorded during release. Whenever. these taenitors are inoperable for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, two independent samples of cach tank to be discharged shall be analyzed and two plant personnel shall independently check valving prior to the

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. discharge.

If these monitors are inoperable for a period exceeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, no release from a liquid waste tank shall be made and any release in progress shall be terminated.

The flow rate of liquid radioactive waste shall be continuously

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neasured and recorded during release.

f.

All liquid effluent radiation monitors shall be calibrated at least quarterly by means of a radioactive source which has been calibrated to a National Bureau of Standards source. Each monitor shall also have a functional test monthly and an instru-ment check prior to making a release, The radioactivity in steam generator blowdown shall be continuously g.

monitored and recorded. Wheneier th,ese monitors are inoparable, the b[owdown flow shall b ' divert'ed to the waste c.anagenent system and the direct; release to the environment terminated.

Tha r'elease of rad,igactive miterialr in liquis da'ste ef fluents to Bases:

unrestricted areas shall not'. exceed the" concentration lbsits

-c specified in 10 CFR Part 23 and should be as low as practicable in accordance with the requirements of 10 CFR Part 50.36a. These s

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specifications pro: ride reasonable-assurance that the resulting

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annual dose to the total body or any organ of an' individual in an unrestricted area will not exceed 5 mrem. At the same time, these specifications permit the flexibility of operation, compatible with considerations of health and safety, to assure that the public

S, t is provided a dependable source of power under unusual operating conditions which 'may temporarily result in releaces higher than the' design objective levels but still within the concentration limits specified in~10 CFR Part 20.

It is expected that by using:

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this operational. flexibility under unusual operating conditions, and exerting every effort to keep levels of radioactive material in liquid wastes as low as praccicable, the annual releases will-not exceed a small fraction of the concentration limits specified in 10 CFR Part 20.

The design objectives have been developed based on operating experi-ence taking into account a combination of variables including defective fuel, primary system Icakage, primary _.to se,condary system leakage, i

steam generator blowdown and the performance of the various waste treatment systems, and are consistent with Appendix 1 to 10 CFR Part 50.

Specification 2.4.1.a requires the licensee to limit the concentration of radioactive raterials in liquid waste effluents released from the site to levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for This specification p'ovides assurance that no unrestricted areas.

r member of the general public will be exposed to liquid containing radioactive materials in excess of limits considered permissible under the Commission's Rules and Regulations.

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Specifications 2.4.1.b and 2.4.1.c establish the upper limits for the release of radioactive materials in liquid effluents. The intent of these Specifications is to permit the licensee the ficxibility of

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operation to assure that the public is provided a dependable source a

of power under unusual operating conditions which may temporarily result-in releases higher than the levels normally achievable when the plant and the liquid waste treatment systems are functioning as designed. Releases of up to these levels }will result in concentrations of radioactive material in liquid waste effluents at small percentages of the limits specified in 10 CFR Part 20.

Specifications 2.4.1.d and 2.4.1.e require that suitable equipment to control and monitor the releases o_f, radioactive materials in liquid i

wastes are operating during any period these releases are taking place consistent with the requirements of 10 CFR Part 50, Appendix A, Design Criterion 64.

Specification 2.4.1.f requires that the licensee maintain and operate the equipment installed in the liquid waste systems to reduce the release of radioactive materials in liquid effluents to as low as practicable consistent with the requirecents of 10 CFR Part 50.36a.

Normal use and maintenance of installed equipment in the liquid waste system provides reasonable assurance that the quantity released will not exceed the design objective.

In order to keep releases of radioactive

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-s 8-nat erials as low as practicahic,- the specification requires operatien -

of equipment whenever it appears t$5t' the p; ojected cumulative -dis-charge rate will exceed one-fourth of this design objective annual quantity during any calendar quarter._

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Specification 2.4.1.g 1Laits the amount of radioactive naterial that could.

be inadvertently released to the environment to an amount that will not exceed the Technical Specification 10mit.

In addition to limiting conditions for operation listed under Specifications 2.4.1.b and.:2.4.1.c, the reporting requirements of Specification 2.4.1.h delineate that the licensee shall. identify the cause whenever the cumulative release of radioactive natorials in i

Liquid waste effluents exceeds one-half the design objectiva annual quantity during any calendar quarter and describe the proposed program of.

action to reduce such releases to design objective levels on timely basis. This report must be filed within 30 days following the calendar quarter in which the release occurred.

The sampling and monitoring requirements given under Specification 2.4.2 provide assurance that radioactive materials in lignid vaates are properly controlled and monitored fu conformance with the requirements of Design Criteria 60 and 64.

These requirements provide the data for the licensee and the Commission to evaluate s'

the plant's performance relative to radioactive liquid wastes-I

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. released to the environment. -Reports on the quantities of radioactive materials released in liquid waste effluents nre furnished to the Ccemission according to Section 5.6.1 of these Technical Specifications in' conformance with Regulatory Guide 1.21.

On the basis of such reports j

and any additional information the Comission may from time to time require the licensee to take such action as the Commission deems appropriate.

The points of release to the environment to be monitored in Section 2.4.2 include all the monitored release points as provided for in Table 2.4-3.'

2.4.3 Specifications for Gaseous Waste Effluents 4

a.

(1) The release rate Ihnit of noble gases from the site shall be:

Q 6.0 E

+

21 E 6

L i

where:

Q = release rate from vents in Ci/sec (ground release) v i = the ich individual nuclide.

{ = the average gamma energy per disintegration for nuclida -1

[g=theaveragebetaenergyperdisintegrationforauclide1 Refer to Table 2.4-5 for I and E values to be used.

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sq f: (2) The release rate limit of all radiuiodincc and radioactive materials in particulate form with half-lives greater than eight days, released to the environs as part of the gaseous wastes from the site shall be:

4'.3 x 10 Q

11 where Q is defined above.

b.

(1) The average release rate of noble gases from the site during any calendar quarter shall be:

E 66 Q

1 iV

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and, 5

37 S

I iv 17 L

. i (2) The average release rate of noble gases from the site during any 12 consecutive months shall be:

132 Q

1 1

E ty gg

and, 2

5 75 Q

1 iy iv.

i (3) The average release rate of all iodines and radicactive materials in particulate form per site with half-lives greater than eight days during any calendar quarter shall be:

6 5.4 x 10 Q

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9 s (4) Theaveragereleaserateofalliodinesandradicaptive materials per site in particulate form with half-1kves creater t he eight days during any period of 12 consecutive months shall be:

7 1.08 x 10 q

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z (5) The amount of iodine-131 released during any calendar quarter shall not exceed 2 Ci/ reactor.

(6) The amount of iodine-131 released during any period of 12 consecutive months shall not exceed 4 C1/ reactor, Should any of the conditions of 2.4.3.c(1), (2) or (3) Listed below exist, c.

the licensee shall make an investigation to identify the _ causes of the release rates, define and initiate a program of action to reduce the release rates to design objective levels listed in Section 2.4 and report these actions to the Comission within 30 days from the end of the quarter during which the releases occurred.

(1) If the average release rate of noble gases from the site curing any calendar quarter is:

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265 Q

g iyj or, 15 0 Q

1 iy 1y i

If the average release rate of all iodines and radioactive (2) materials in particulate per site form with half-lives greater than eightdaysduringanycaiendarquarteris:

>1 2.2 x 10 Q,,

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, 2 (3) If the amount of iodine-131 released during any calendar quarter is greater than 0.5 Ci/ reactor.

d.

During the release of gaseous vastes from the prieary system uaste gas holdup system the effluent uonitors listen in Table.

2.4-4'shall be operating and set to alarm and to initiate the C

automatically closure of the waste gas discharge valve prior to exceeding the limits specified in 2.4.3.a above. The operability of each automatic isolation valve shall-be demonstrated auarterly.-

e.

The maximum activity to be contained in one waste gas storaga tank shall not exceed 450,000 curies (considered as Xe-133).

2.4.4 Specifications for Gaseous Waste Sampling and Monitoring a.

Plant records shall be maintained and reports of the sampling I

and analyses results shall be submitted in accordance with Section 5.6.1 of these Specifications. Estimates of the aamoling and analytical error associated with each reported value should be included.

b.

Gaseous releases to the enviromment, except from the turbine building ventilation 'e::haust and as noted in Specification 2.4.4.c, shall be continuously monitored for gross radioactivity and the flow continuously measured and recorded. Whenever thasa monitors are inopernble, erab samoles shall be taken and analyzed daily for cross radioactivity. If these monitors are inoperable for more than seven days, these releases shall be terminated.

c.

During the release of gaseous wastes from the primary system waste gas holdup system, the gross activity monitor, the iodine collection device, and the particulate collection device shall be operating.

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- d.

All unste gas ef fluent monitor.v shall be calibrated at 'least Inarreriv by means of a known radioactive source' which' has been calibrated to a National Bureau of Standards source.- Each monitor shall have a. functional test at Icast monthly and instrument check at least daily.

Sampling and analysis of radioactive material in gaseous waste, e.

particulate form, and radioiodine shall be performed in accordance with Table 2.4-2.

Bases: The release of radioactive materials in gaseous waste effluents to unrestricted areas shall not exceed the concentration limits specified P

in 10 CFR Part 20, and in accordance with the requirements of.10 CFR Part 50.36a.

These specifications provide reasonable assurance that the resulting annutti air dose from the site due to gamma radiation will not exceed 10 mrad, an' nn annual air dose from the site due to beta radiation will not e::cead 20 erad from noble gases, and that the annual dose to any organ of an individual from iodines and particulates will not exceed 15 mrem per site.

At the.neue r.ime these specifications permit the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided with a dependable source of power under unusual operating conditions which may temporarily result in releases higher daan the design objective levals but still within the concentration -limits specified in 10 CFR Part 20.

It is expected that using this opera-tional ficxibility under unusual operating conditions, and by exerting

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. every ef fort to keep icvels of radioactivo material in ga:;eous vaste cifluents as lov as _ practicable, the annual releases will not exceed a small.

fraction of-the concentration ilmits specified in 10 CFR Part 20.

These efforts should include consideraticn of meteorological con-ditions during releases.

There -is a reduction factor of 243 by which the maximum permissible concentration of radioactive iodine in air should be reduced to allow. for the grass-cow-milk pathway.

(The factor is 1220 for the grass-goat-milk pathway). This factor has been derived for radioactive iodine, taking into account the milk pathway. It has been applied to radionuclides of. iodine and to all radionuclides in particulate form with a half-Life grea,ter than eight days. The g

factor is not appropriate for iodine where milk is not a pathway of exposure or for the other radionuclides.

The design objectives have been developed based on operatirg experience taking into account a combination of system variables including de'fective fuci, primary system leakage., primary to secondary system leakage, steam generator bicwdown and the performance of the varicus vaste treatment systems.

For Specification 2.4.3.a(1) dose calculations have been made for die critical sector. These calculations consider site meteorology, buoyancy characteristics, and radionuclide content of the effluent

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of each uni t.

Meteorological calculations for offsite locationn were performed, and die most critical one was selected to set the release rate. The controlling distance is 1450 meters to the ENE.

The gamma dose contribution was determined using Cquation 7.r3 in Section 7-5.2.5 of Meteorology and Atomic Energy - 1968. The releases from vents are considered to be ground. level releases which could result in a beta dose fran cloud submersion. The beta dose contribution was determined using Equation 7.21, as described in Section 7-4.1 of Meteorology and Atomic Energy - 1968. The beta dose contribution was determined on the basis of an infinite cloud:

passage with semi-infinite geometry for a ground level reicase (submersion dose). The beta and gamma components of the gross radio-e activity in gaseous effluents were combined to determine the allowable continuous release rate. Based on these calculations, a continuous release rate of gross radioactivity per site in the amount specified in 2.4.3.a(1) will not result in offsite annual doses above background in excess of the limits specified in 10 CFR Part 20.

The average gamma and beta energy per disintegration used in the equation of Specification 2.4.3.a(1) will be based on the average composition of gases determined from the plant vent and ventilation 4

exhausts. The average energy per beta or. gamma disintegration for those radioisotopes determined to be present from the isotopic-i

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. analyses are given in Table 2.4-5.

11here isotopes are identified that are not listed in Table 2.4-5, the gamma energy are determined from Table of Isotopes, C. M. Lederer, J. M. Hollander, and I. Perlman, Sixth Edition,1967 and the beta energy shall be as given in USNRDL-TR-802, II.

Spectra of Individual Negatron Emitters (Beta Spectra),

O. Hogan, P. E. Zigman, and J. L. Mackin.

For Specification 2.4.3.a(2), dose calculations have been made for the critical sectors and critical pathways for all radioiodines and radioactive material in particulate foam, with half-lives greater than eight days. The calculations consider site meteorology for these releases.

I For radiciodines and radioactive materials in particulate form, the controlling sector for unit vent releases is the ENE sector at a distance of 1450 meters (X/Q = 1.46 x lb" sec/m ) 'for the dose due to inhalation. The nearest milk cow is located in the ENEsector at a distance of 6400 meters. The applicable X/Q at the nearest milk cow is 1.78 x 10' sec/m. The grass-cow-milk-child thyroid chain is controlling.

The assumptions used for these calculations are:

(1) onsite wateoro-logical data for the most critical 22.5 degree sector; (2) credit for building wake; and (3) a reconcentration factor of 243and a grazing factor of 1.0was applied for possible ecological chain ef fects from radioactive iodine and particulate releases.

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- Specificafion 2.4.3.b establishes _ upper :;1te levels for the' rcicases of noble gases, iodine's and particulates with half-lives greater than

-eight days, a,nd iodine-131-at twice the design objective annual:

quantity during any calendar quarter, or four times the design c

objective ' annual quantity during any period of 12 consecutive months.

The intent of this specification is to permit the licensee the flexibility of operation to assure that the public is provided a-dependable source of power ~ under unusual operating conditions which p

may temporarily result in higher releases than the objectives.

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In addition to the Ibniting ' conditions for operation of Specifications 2.4,3.a and 2.4.3.b, the reporting requirements of'2.4.3.c delineata-i i

i that the cause be identified whenever the release of gaseous e,ffluents exceeds one-half the design objective annual quantity during any calendar quarter and describe the proposed program of action to reduce such release rates ' to the design objectives.

l Specification 2.4.3.d requires that suitable equipment to monitor and control the radioactive gaseous =rcicases are operating during any period these releases are taking place, i

l Specification 2.4.3.e limits the maximum offsite dose above background to below the limits of 10 CFR Part 20, postulating that the rupture i

of a waste gas storage tank holding the maximum activity releases all i

of the contents to the atmosphere.

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. The sampling and monitoring requirements given under _ Specification 2.4.4 provide assurance that radioactive materialf released in gaseous waste effluents are properly coatrolled-and monitored'in conformance with the requirements of Design Criteria 60 and 64.

These. requirements provide the data for the licensee and the Commission to evaluate the plant's performance relative to radioactive waste effluents released. to the -

environment. Reports on the quantities of radioactive materials released in gaseous effluents are furnished to the Commission on the basis of Section 5.6.1 of these Technical Specifications and in conformance with Regulatory Guide 1.21.

On.the basis of such reports and any additional information the Commission may obtain from the licensee or others, the Cennission may from time to time require the licensee to take such action as the C3mmission deems appropriate.

points of r' leasa to the environ cent to be monitored in Section 2.4.4.

The e

include all the' monitored release points as provided for in Table 2.4-4.

Specification 2.4.4.b excludes monitoring the turbine building venti-Lation exhaust since this release is expected to be a negligible release point. Many PWR reactors do not have turbine building enclosures. To be consistent in this requirement for all PWR reactors, the monitoring of gaseous releases from turbine buildings is not required.

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N 1 2.4.5 Specifications for Solid Waste Handling and Disposal a.

Measurements shall be made to determine or estimate the total curie quantity and principle radionuclide composition of all radio-active solid waste shipped offsite.

b.

Solid wastes in storage and preparatory to shipment shall be monitored and packaged to assure compliance with 10 CFR Part 20, 10 CFR Part 71, and 49 CFR Parts 171-178.

c.

Reports of the radioactive solid waste shipments, volumes, principle radionuclides, and total curic quantity, shall be submitted in accordance with Section 5.6.1.

Bases: The requirenents for solid radioactive waste handling and disposal given under Specification 2.4.5 provide assurance that solid radioactive materials stored at the plant and shipped of fsite are packaged in conformance with 10 CFR Part 20,10 CFR Part 71, and 49 CFA Parts 171-178. These requirements provide the data for the licensee and the Co==ission to evaluate the handling and storage facilities for solid radwaste, and to evaluate the environmental impact of offsite shipment and storage. Reports on the quantities, principle isotopes and volumes of the shipments, are furnished to the Commission according to Section 5.6.1 of these Technical -Specifications. On the basis of such reports and any additional information the Commission may obtain from the licensee or others, the Commission may_ from time to time require the licensee to take such action as the Commission deems appropriate.

Tnbic 2.4-1 MDI_0,A,CT,J,V,E J.I,QU,7D,fiA}1 PLT,NG, A?m AMALYS,T,S 1.l tgu l d Snmplhin Typc 01 Dete s intele Sourco Frcquency ActIvlty Analynin CuncontentI nn

_ ____.. __.... __(l i:.Wi l.).f' A.

Monitor Tank Rolcases l'ach Bntch I'rincipal Canma Emi t tern Sx 10.7 (2)

Ono Batch / Month Disnolved cases 10

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Weekly Composite (1)

En-1.a-140, 1-131 10~0 Sr-89 5 x 10~

Monthly Composite ( )

11 - 3 10-5 o

-7 x

Crosn a 10 i

Quarterly Composite (1) Sr-90 5 x 10

-8 B.

Primary Coolant Weekly ( )

1-131, I-133' 10~0

' (2)

C.

Steam Generator Blowdown Principal Camma Emitters 5 x 10 Weekly (5)

-6 En-Ln-140, I-131 10 t

-5 One Sample / Month Dissolved Cases 10 t-Sr-89 5 x 10~

Monthly. Composite ( )

~0 11 - 3 10

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Cross a 10 Quarterly Composite (5) Sr-90 5 x 10-8 a

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Table 2.4-1 (Continued)

I NOT_ES_:

(1)

A composito sanple is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged.

J (2)

For certain mixtures of ga==a emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater concentrations.

Under these circumstances, it will be more appro-priate to calculate the concentrations of such radionuclides using.

measured ratios with those. radionuclides which are routinely identified and measured.

-(3)

The detectability limits for activity analysis are based on the technical feasibility and on the potential significance in the environment of the quantities released.. For some nuclides, lower detection limits may be readily achievable and when nuclides are measured below the stated limits, they should also be reported.

(4)

The power level and elsanup or purification flow rate at the sample ti=e shall also be reported.

(5)

To be representative of the average quantities and concentrations of radioactive caterials in liquid effluents, ca=ples should be collected in proportion to the rate of flow of the effuent. stream.

Prior to analyses, all t.;acples taken for the composite should be throughly mixed in order for the composite sacple to be repre-sentativ of the average efflueiit releasec.

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Tchic 2.4 '

RADIOACTIVE CASEOUS WAS_TE SAMPLING AND ANA13SI3 Cnocous Sampling

. Type of D7t$ctable

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Source Frcquency Activity Analynin Concentratgs

. 6c.c.!In11

~4 ()

A.

Waste cas Decay Each Tank Principal Canma Emittern 10 Tank Ec1casca

~0

'H-3 10 B.

Containment Purgo Releases Each Purge Principal Comma Emitters 10 - ()

~0 H-3 10

-4

( ) (3).

C.

Condenser Air Ejector Monthly Principal Camma Emitters 10

~0 g'

H-3 10

2) (3)

N D.

Environmental Release Points Monthly Principal Gamma Emitters 10 (Gas Samples)

-6

-12 Weekly (Charcoal Sample)

I-131 10

-10 Monthly (Charcoal Sample)

I-13$,I-135 10 Weekly (Particulates)

Princ'ipal Camma Emitters 71 (at Icast for Ba-La-140, 1-131 10 Monthly Composite ( )

Srl89 10

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(Particulates)

'e

.gg Cross a 10 Quarterly Composite ( )

Sr-9,0 10-11 (particulat_es)

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Table 2.4-2 (Continued)

I:0TES:

(1)

The above detectability limits for activity analysis are based on technical f aasibility and on the potential significance in the environ =ent of the quantities released. For some nuclides, lovar detection limits may be readily achievable and when nuclidas are measured below the stated limits, they should also be reported.

(2)

Analyses shall also be performed following each refueling, startup' or simiinr operational occurrence which could alter the mixture of radionuclides.

(3)

For certain mixtures of ga=ma emitters, it may not be possible to ceasure radionuclides at levels near their sensitivity limits when other nuclides are present in the sample at much higher levels.

Under these circumstances, it will be more appropriate to calculate the levels of such radionuclides using observed ratios with thosa radionuclides which are measurable.

(4)

To be representative of the average quantities and concentrations of radioactive caterials in particulate form released in ganous effluents, sa:ples should be collected in proportion to the rate of flew-of the effluent stream.

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h Table 2.4-3 IWR-fJ.QUfD VA*iTR SYSTDt 14CAf t0N OF PROCESS AND EFFLUENT HONITOR% AN) SWPlJR9 REQUIRPD BT TECHNICAL SptCtFICArf 0:e!t Crab I

/

Auto Control to Continuous Son.ple Cross Dianolved '

Isotopic

,P,rocess Stream or Delen9e Point Alarm.Taolation Valve _ h nitor Efftins Aertvity 1

_Csaen LQh3 M

Acaly,is Hiscellaneous Waste Sample (Test) Tank X

X X

X X

X x

chemical Vaste Sample (Test) Tank I

X X

X X

X X

Detergent Waste Collector Tanke X

X X

X X

X X

Frisury Coolant System X

X X

e Liquid Radvaste Discharge Pipe X

X X

X r4 Steam Cenerator Blowdown Systenea X

X X

.X X

X X

X X

Service Water Discharte Pipe X

X X

Outdoor Storace Tanks - Dikes or gerentien Fonds X

X X

Emergency Core Cooling System I

X X

X Nuclear Closed Cooling System X

X X

Turbine tullding Sumps (Floor Drains)

X X

X X

=

\\

• in most PWs the centents of the detergent waste collector tank are sampled, analysed and then (1.tered prior to release through the 11guid raduaste discharga pipe. *The detorcent waste systca should be designed with either a split tank or two separata collection or sample (test) tanks to permit isolatian of the tar.ks for nixing, sampling and analyste prior to release.

• • 13 sone F'-1's processed liquid from the steam generator blowdown system is returned directly to the secondary system and the need for continuous ronitoring at this rc1 case point is eliminated.

0 a

9 e

=e.

9 s.

f TsSte'3.4-4 IVR-CWGt? W.nTE SYSTCH 1ACATTOM OF fMCI'$9 PN3 FITT.UFNT MSNilw.$ *'.3_Syli't"s llMQlyD BY_ TI:f'fP!f t:Alg yriffjfgT),qtaj H

4 4

(

Crab Auto Control to Continusue 1 ample P?rseure g 3 Iy pe*, sey.c, a c M *1csae Faf a h

fantatfrin valve jpgtjan gget3g Jp, I

fyri M

g3 f

b'aste Cas Storage Tanks X

X X

X 3

X X

X X

Condenser Air Ejector X'

E

,X E

X X

X X

Vent Peader Systee*

I 4

I I

I X

X X

g Bu11 ding ventilatica Systems X

O Xeactor Contairswnt Sutiding (vbenever N

X X

X X

X X

X X

X there le ficw) 1 Austliary Sutidinge X

X X

X i I

X X

X Fuel Mandling & Storage Building

• I I

I X

X X

.X X

Xadweste Building

• X X

N 4

X X

X X

e Steae Cenerator 31swdmn Tank Vent or Condenser Vent **

I X

X X

X X

X X

Turbine Cinnd Seal Condenser X

X X

X X

X X

X Mechasical Vacuum Pep X

X X

X X

X X

X

.tiaate Evn erator C,ondenser vent X

X X

X X

X X

X

• • ! av or att of the procesa streams or butidinC ventilation systems are routed to a sint.le release point. the need for o cont t.PJouS eCF.1 tor At 1:w trJavidual discharr,e potet to the main eshaust duct to ellatnated. One teatinuous teatter at the flaat retcane teolut in pufficient.

• I :* w Iw's alw caran r.enerator blow. town tank vent in routed to tt.c main tur!.f ar can.lr==cr and the n.v4 for a s ent ine=*oes wesitor at t hle a

rr itwW p sleL la eliminal ae*for l'A's where Llie van porntor cordenner in wonted directly to the atsonphere.

1. '@p 6

a

m

- 26 Table 2.4-5 AVERACE ENERGY PER DISINTECRATION Isotope E, Mev/ dis (Ref)

E, Itev/ dis ( }

(Ref) g Kr-83m 0.00248 (1) 0.0371 (1)

Kr-85 0.0022 (1) 0.250 (1)

Kr-85m 0.159 (1) 0.253 (1)

Kr-87 0.793 (1) 1.32 (1)

Kr-88 1.95 (1) 0.377 (1)

Kr-89 2.22 (2) 1.37 (2)

Kr-90 2.10 (2) 1.01

('2)

Xe-131n 0.0201 (1)

O.J$3 (1)

Xc-133 0.0454

..(1) 0.135

,(1)

Xc-133:::

0.042 (1) 0.19 (1)

Xc-135 0.247 (1) 0.317 (1)

Xe-135m 0.432 (1) 0.095 (1)

e-137 0.194 (1) 1.64 (1)

Xe-138 1.18 (1) 0.611 (1)

(1) ORNI.-4923, Radioactive Ato:s - Surplement I, M. S. Martin, Nov2=bac i

1973.

(2) NEDO-12037, " Sus =ary of Gacna and Deta Emitters and Intcaa U:y Data,"

M. E. Meek, R. S. Gilbert, January 1970.

(The average 6 caergy was not co=puted using the 1/3 value assumption as used in this reference.

It was computed from the maximum energy using the equation in the Report of Committee II on Permissible Dose for Internal Radiation (1959), ICRP Publication 2, Pergamon Press, 1960,)'

(3) The average S energy includes conversion electroiis.

O UV gt gg6%,'

-