ML19345E463

Environ Protection Plan (Nonradiological) for Facility

ML19345E463
Person / Time
Site:	Monticello
Issue date:	01/09/1981
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML19345E464	List: ML19345E463 ML19354C415 ML20003A467
References
NUDOCS 8102040085
Download: ML19345E463 (27)

Text

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APPENDIX B TO FACILITY OPERATING LICENSE DPR-22 l

MONTICELLO N'JCLEAR GENERATING PLANT UNIT 1 NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 l

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TABLE OF CONTENTS Page 2.4 Radioactive Effluents TS B.2.4-1 4

l 2.4.1 Specification for Liquid Waste ~ Effluents TS B.2.4-2 2.4.2 Specifications for Liquid Waste Sa:spling

& Monitoring TS B.2.4-3 Bases-Liquid Wastes TS B.2.4-4 l

2.4.3 Specifications for Gaseous Waste Effluents TS B.2.4-7 2.4.4 Specifications for Caseous Waste Sampling & Monitoring TS B.2.4-12 j

Bases-Gaseous Wastes TS B.2.4-19 2.4.5 Specifications for Solid Waste Handling & Disposal TS B.2.4-19 l

-Bases-Solid Wastes TS B.2.4-19 4

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e TSB.2.4 f

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e LIST OF TABLES & FIGURES TABLE TS B.2.4-1 Radioactive Liquid Waste Sampling & Analysis TS B.2.4-2 Radioactive Gaseous Wasta Samplir;g & Analysis TS B.2.4-3 Loa.ation of Liquid Effluent Mon? tors & Samplers TS B.2.4-4 Locatien of Gaseous Process

& Efflue.'t Monitors & Sa=plers Gama & Beta Dose Factors TS 'd.2.4-5 FIGURE TS B.2.4-1 off Gas Storage Tank Gross Activity Limits e

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T58. 2. 4-11

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I 2.4 LIMITINC CONDITIONS FOR OPERATION i"

Radioactive Effluents Objective: To define the limits and :onditions for th,e controlled release of radioactive materials in liquid and gaseous effluents to the environs to ensure that these releases are as lov as practicable. These releases should not result in radiation exposures in unrestricted areas greater than a few percent of natural background exposures.

The concentrations of radioactive materials in effluents shall be within the limits specified in 10 CFR Part 20.

, To ensure that the releases of radioactive material are as low as practicable, the following design objectives apply until Technica1 Specifications are issued in accordance with the recently adopted Appendix I to 10 CFR Part 50:

I For liquid wastes:

The annual dose above background to the total body or any organ of an a.

individual should not exceed 5 mrem in an unrestricted area, b.

The annual total quantity of radioactive materials in liquid waste, excluding i

tritium and dissolved gases, discharged should not exceed 5 C1.

For gaseous wastes:

c.

The annual total quantity of noble gases above background discharged frem the site should result in an air dose due to gamma radiation of less than 10 mred, 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 radiciodines and radioactive material in

. particulate forms with half-lives greater than eight days above background, should not result in an annual dose to any organ of an individcal in an unrestricted area frem all pathways of exposure in excess of 15 erem.

e.

The annual total quantity of iodine-131 discharged should not exceed 1 Ci.

TS B. 2.4-1

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.4.1 Specifications for Liquid Waste Effluents The concentration of radiosceive materials released in liquid waste a.

effluents shall not exceed the value specified in 10 CFR Part 20, i

j Appendix B, Table II, Column 2, and Notes thereto, in the condens-r cooling water discharge canal.

b.

The cumulative release of radioactive materials in liquid waste effluent, excluding tritium and dissolved gases, shall not exceed 10 Ci in any calendar quarter.

The cumulative release of radioactive materials in liquid waste c.

effluents, excluding tritium and dissolved gases, shall not exceed 20 Ci in any 12 consecutive months.

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d.

The equipment installed in the liquid radioactive waste system shall be maintained and shall be operated to process radioactive liquid wastes prior to their discharge.

I The maxi:m.n radioactivity to be contained in any liquid radwaste tank e.

that can be discharged directly to the environs shall not exceed l

10 C1, excluding tritium and dissolved gases.

f.

If the cumulativa release of radioactive materials.in liquid effluents, excluding tritium and dissolved gases, exceeds 2.5 Ci/ calendar quarter, the licensee shall make an investigation to identify the causes for such releases, define and initiate a program of action to reduce such releases to the design objective levels listed in Section 2.4, and report these actions to the NRC within 30 days from the end of the quarter f

during which the release occurred.

TS B. 2. 4-2 ge**

3 An unplanned or uncontrolled offsite release of radioactive materials

,b in liquid effluents in excess of 0.5 curies requires notification.

i This notification to the NRC shall be made within 30 days.

'2.4.2 Specifications for Liquid W ste Sampling and Monitoring Plant records shall be maintained of the radioactive concent.ation end a.

volume before dilution of liquid waste intended for discharge and the average dilution flow and length of time over which each discharge occurred. Summaries of the quantities of re'acases shall be included in the Sami-Annual Radioactive Effluent Report. Estimates of the sampling and analytical errors associated with each reported value shall be included.

b.

Prior to release of each batch of liquid waste, a sample shall be taken from that batch and analyzed for the concentration of each significant gamma energy peak in accordance with Table 2.4-1 to demonstrate compliance with Specification 2.4.1 using the flow rate into which the waste is

,v discharged during the period of discharge.

Sampling and analysis of liquid radioactive waste shall be performed c.

i in accordance with Table 2.4-1.

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

d.

The radioactivity in liquid wastes shall be continuously men'tored and i

recorded during release. Whenever these monitors 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 each tank to be discharged shall be analyzed and two plant personnel shall independently check valving prior to the 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 I

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waste tank shall be made and any release in progress shall be terminated.

TS B. 2. 4-3

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

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e.

The flow mte of liquid radioactive vaste shall be continuously measured and recorded during release, f.

The continuous monitors listed in Table 2.k.3 shr.11 be calibrated

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at least quarterly by means of a solid mdioaccive source which has been calibrated to a National Bureau of Standards source.

Each muitor chall also have a functional test monthly and an instrument check prior to taking a release.

g.

During each release of liquid radioactive vaste, the liquid radvaste i'

discharge monitor rendings shall be correlated with the results of analyses performed prior to release.

Bases: These Specifications are applicable until. Specifications prepared in accordance with Appendix I to 10 CFR Part 50 are issued by the Coi.unis sion. In some cases these Specifications may be more restrictive than l

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required by Appendix I.

In the event that plant availability is adversely

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affected by these Specifications, the licensee may apply to the Commission for appropriate Technical Specification changes on a case by case basis.

Specification 2.4.1.a requires the licensee to limit the concentration of radioactive materials in liquid vaste effluent's' released from the site to l

1evels specified in 10 CFR Part 20, Appendix B, Table II, Column 2, fcr unrestricted areas. This specification provides assurance that no member of the general public will be expose-to liquid containing radioactive materials in excess of limits considered permissible under the Cocenission's Regu lations.

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TS B. 2.4.

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N Specifications 2.4.1.b and 2.4.1.c establish the upper limits for the I

release of radioactive materials in liquid effluents. The intent of these j

Specifications is to permit the licensee the i;exibility of operation to assure that the public is provided a dependable source of power under unusual cperating conditions which may tempo;arily result in releases higher i

than the levels normally achievable when the plant and the licuid waste treatment systens are functioning as designed. R.eleases 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.

Specification 2.4.1.d requires that the licensee uaintain and operate the equipment installed in the liquid waste systens to reduce the release of i

radioactive materials in liquid effluents to as low as practicable con-i sistent with the require =ents of 10 CFR Part 50.36a. Nor=al use and maintenance of installed equipment in the liquid waste system provides i

j reasonable assurance that the quantity released will not exceed the design obj ective. In order to keep releases of radioactive materials as low as i

l practicable, the specification requires operation of equipment whenever a release of radioactive liquid waste is made.

l Specification 2.4.1.e restricts the amount of radioactive material that could be inadvertently released to the environment to an amount that will i

not exceed the Technical Specificatien limit.

s-TS B. 2. 4-5 l.

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In addition to 10miting conditions for, operation listed under Specifications 3

2.4.1.b and 2.4.1.c, the reporting requirements of Specification 2.4.1.f l

delineate that the licensee shall identify the cause whenever the cumulative I

release of radio.1ctive materials in if quid waste e'.fluents exceeds one-half 4

i the design objective annual quantity during any calendar quarter and l

describe the proposed program of action to reduce such releases to design objective levels on a timely basis.

This report must be filed within 30 days following the calendar quarter in which the release occurred, i

Specification 2.4.1.g provides for reporting spillage or release events which, while below the limits of 10 CFR Part 20, could result in releases higher than the design objectives.

The sampling and monitoring requirements given under Specification 2.4.2 dy provide assurance that radioactive materials in liquid wastes are properly controlled and monitored in conformance with the requirements of 2esign Criteria 60 and 64 of Appendix A to 10 CFR Part 50 and permit as licensee

. and the Joamission to evaluate the plant's performance relative to radio-active liquid wastes released to the environment. Reports of the quantiti'es '

of radiotetive materials released in liquid waste effluents are furnished to the Commission semi-annually.,On the basis of such reports and any additional information the Commission may obtain fram the licensee or i

others, the Commission may from time to time require the licensee to take such action as the Commission deer rno opriate.

The points of release to the environment to be monitored it. Section 2.4.2 include all the monitored release points listed in Table 2.4-3.

I TS B. 2.4-6

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1 4.3 Specifications for Caseous._ ste Ef fluents 4

The terms used in these Specifications are as follows:

subscripts v, i n s to vent releases s, refers to stack releases 1, refers to individual noble gas nuclide (Refer to Table 2.4-5 for the noble gas nuclides considered)

Q = the total noble gas release rate (Ci/sec) 7

= [Qi sum of the individual nobic gas radionuclides detemined to be present by isotopic analysis it = the average total body dose factor due to scrua emission 1

(re=/yr per Ci/sec)

L = the average skin dose factor due to beta emissions j

(rem /yr per Ci/sec) f( = the average air dose factor due to beta emissicas

[m (rad /yr per Ci/sec)

W = the average air dose factor due to ga:::::a emissions (rad /yr per Ci/sec)

The values of 5, E, E, and 3 for the vent releases are dete=ined frc: the isotopic analysis perfomed at the discharge of the stea= jet air ejectors as delineated in Specification 2.h.h.c.

Se values of E, E, E, and 5 for the stack are determined from the isotopic analysis perfor=ed at a point prior to dilution and discharge as delineated in Spec'.fication 2.k.4.c.

New values should be deter =ined each time isotopic analysis is required as follows:

E = (1/ar) Ecixi i

I = (1/ar) QSi i L

H = (1/or) ge1q R = (1/QT) EQ Nii 1

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vhere the values of K,11,g a ed L are rovided in Table 2.4-5, and are g

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site dependent garna and beta dose factors.

l Q = the men ared release rate (C1/sec) of the radiciodines and radioactive materials in particulate forms with half-lives. greater than eight days.

1.1 = a factor accounting for the increased L..>ppira power of tissue relative to air for beta radiation 1

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7 Should any of the conditions of 2.h.3.a(1) or (2) be exceeded, the licensee a.

j shall take appropriate corrective action to brin's the releases vi;hin these

limits, i

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(1) We release rate limit of noble gases from the site shall be such 4

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that 2.0 (Q K

+

Q K

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< 1 Ts s and Ts( s +.1.1E (I, + 1.15 )

< 1

+ Q 0.33 y

(2) n e release rate limit of i radiciodines and radioactive materialc i

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

such that 5

4 3.7 x 10 Q + 2.5 x 10 q <

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b.

Should any of the conditions of 2.4.3.b (1), (2), (3), (4), (5), or (6) be exceeded, the licensee shall take appropriate corrective action to 4

4 bring the releases within these limits.

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

13 (Q 3

+ Q3)

< 1 i

Tv v Ts s j

1 and 6.3(

Q

+

ETs s )

Tv v 1

(2) The average release rate of ncbie gases from the site during any 12 consecutive months shall be 25(Q

+

STs s) bI

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hv and 4

.'l 13( Q E

+

qts s} 5l gy s

(3) ne average release rate per site of all radiciodines and radio-3' active caterials in particulate form with half-lives greater than eight days during any calendar tuarter shall be such that l

5 4

13 (3.7 x10 Q

+

2.5 x10 Q, ) $ 1 y

(4) ne average release rate per site of all radiciodines and radioactive ma.larials in particulate for= with half-lives greater than eight days during any period of 12 conse:utive t

months shall be such that 4

5 4

25 (

3.7x10 Qv

+

2.5x10 Q, )

<. 1 1

l (5) ne amount of iodine-131 released durin; any calendse quarter 4

ahall not exceed 2 C1.

I (6) The amount of iodine-i31 released during any period of 12 consecutive months shall not exceed 4 Ci.

I TS B. 2. 4-9

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Should any of the conditions of 2.4.3.c(1), (2), or (3) below exist, the c.

licensee shall make an investigation to identify the causes of the release rates, define and initiate a prog am of action to reduce the release j

j rates to design objective levels listed in Section 2.4 and report these actions to the NRC within 30 days from the end of the quarter during which the releases occurred.

(1) If the average release rate of noble gases frcan the site l

during any calendar quarter is such that

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> 1 50 (Q [v Ts s T

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+

Q M

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25 (Qgy Ts s (2) If the averat;e release rate per site of all radioiodines and radioactive materials in particulate form with half-lives greater than eight days during any calendar quarter is such that 2.5x10'Q,) > l 5

50 (3.7x10 Q +

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(3) If the amount of iodine-131 relea' sed during any calendar quarter is grester than 0.5 C1.

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d d.

Whenever gaseous vastes are being releassd from the offgas treatment system, at least one Main Stack monitor shall be operable and set to alam and initiatc automatic terminat. ion of offgas discharge prior to exceeding the limits of Specification 2.4.3.a above. The capability of each automatic Isolation valve shall be demonstrated quarterly.

If no Main Stack monitor is operating, releases from the offgas system shall be terminated within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

e.

During power operation the Reactor Building Ventilation System monitoring system shall be operable and set to alam end initiate automatic temina-tica of Reactor kilding ventilation air discharge prior to exceeding the limits of Specification 2.4.3.a above. The capability of auto =ati_c isolation of the ventilation system shall be demonstrated quarterly.

If the Reactor Nilding Ventilation System monitoring system is not operating, releases frce the Reactor Nilding Ventilation System shall 4

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be teminated within 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

f.

If the gross radioactivity release rate cf noble gases at the steam jet air ejector monitors exceeds, for a period greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the equivalent of 260,000 uci/sce following a 30-minute decay, notify the NRC within ten days.

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g.

The drywell shall be purged through the standby gas treatment *ystem.

h.

Except as specified in Specification 2.4 31 below, at least two hydrogen monitors in the offgas line downstream of each operating recombiner shall be opernble during pyer operntion.

If the hydrogen concentration reaches a set point of four percent by volune, the offgas flow chall be stopped auto =atically by closing the valves upstrea= of the affected recombiner.

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Whenever the required hydrogen monitors are not operable, effgas flow to the compressed storage subsystem shall be terminated.

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'Ihe maximum gross radioactivity contained in,one gas decay tank after

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12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> holdup that can be discharged directly to the environs shall be less than 22,000 curies of Xc-133 dose equivalent.

k.

The mechanical condenser vacuum pump shall be capable of being isolated and secured on a signal of high radioactivity whenever the main steam line isolation valves are open or it shall be isolated.

1.

At least once during each operating cycle automatic isolation of the mechanical condenser vacuum pump shall be verified.

m.

An unplanned or uncontrolled offsite release of radioactive materials in Easeous effluents in excess of 5 curies of noble gas or 0.02 curie of radiciodine in any one hour period shall be reported within 30 days.

).4.4 Specifications for Caseous Waste Samoling and Monitorine a.

Plant records shall be maintained of the sampling and analyses results. Sum = aries of the quantities of releases shall be included in the Effluent and Waste Disposal Semiannual Report. Estimates of the sampling and analytical error associated with each reported value

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should be included.

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b. Prior to opening a Turbine Building roof vent, except in the event of fire, and daily thereaf ter, air samples shall be taken in three locations shown to be representative of radiation levels on the operating floor and analyzed for radioactive iodines and particulates with half-lives greater than eight days.

In addttion, when a vent is open a continitous radioactive iodine and particulate air monitor shall be in operation on the turbine floor.

The average cor. centration of radiciodine and particulate matter neasured in the Turbine Building shall be used in conjunction with the design flow rate of the roof exhausters in determing Turbine Building release rates.

These release rates shall be included in the "Qv" terms in Specifications 2.4.3.a.

2.4.3.b, and 2.4.3.c.

The position of the Turbine Building roof

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vents shell be logged once each shif t.

TSB. 2. k -12 5.. 90

r An isotopic analysis shall be uude.of a representative sample of gaseous e.

j activity at the discharge of the steam jet air ejectors and at a r/oint e

l prior to dilution and discharge.

(1) at least monthly, and (2) following each refueling outage, and (3) if the steam jet air ejector monitors indicate an increase of greater i

than 50% in the steady state fission gas release after factoring out increases due to power changes.

Following each analysis, steam jet air ejector monitor and stack monitor readings shall be correlated with the results of the analysis.

d.

The continuous monitors listed in Table 2.4-4 shall be calibrated at s

least quarterly by means of a known solid radioactive source which has Each monitor been calibrated to a National Bureau of Standards source.

shall have a functional test at least monthly and an instrument check at least daily.

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

particulate foams and radiciodines shall be performed in accordance with Table 2.4-2.

The hydrogen monitors shall be functionally. tested monthly and calibrated f.

Each monitor shall quarterly with an appropriate gas mintdre source.

have a sensor check at least daily.

Condenser air inleakage shall be evaluated weekly and used in conjunction g.

with steam jet air ejector offEas isotopic analyses and Figure 2.4-1 to to determine that the limit of Specification 2.4.3.j is not exceeded.

1 TSB.2.4-1).

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t Bases: The release of radioactive materials in gaseous waste effluents to unrestricted dreat shall not exceed the concentr: tion limits specified in i

10 CFR Part 20 and should be as low ~as practical in accordance with the re-t

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quirements of 10 CFR Part 50.3oa.

Thet ecifications provide reasonable essurance that the resultirg annual air dose fro:a the site due to ga::rna radiation will not exceed 10 mrad, the annual air dose from the site due to beta radiation will not exceed 20 mrad frco noble gases, that no individual in an unrestricted area will receive an annual dose to the total body gr ater than 5 crem or an annual skin dose to the total body creater than 15 mrem from ficsion p:cduct noble gases, and that the annual dc.se to any organ of an i

individual fro:s radioiodines and radioactive material in pa.ticulate form with half-lives greater than eight days will not exceed 15 mrem frva the site.

At the sa:ne time these specifications permit the flexibility of operation, ccupatible with consideratiens of health and safety, to assure that the public is provided with a dependable source of power under unusual operating con-L ditions which may tempoiarily result in releases higher than the design

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objective levels but still within the concentration li=its specified in

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10 CFR Part 20.

Even with this operational flexibility, the annual releases" l

i will not exceed a small fraction of the concentration' 11mits specified in L

10 CFR Part 20.

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The design _ objectives have been developed based on operating experience taking into account a combination of system variables including defective fuel, primary system les) Sage, and the performance of the various waste treatment systems.

4 I

r TS B. 2. 4-14 i

e Specification 2.4.3.a(1) 10mita the release rate of noble gases from i.

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the site so that the corresponding annual gamaa and beta dose rate above back-ground to an individual in an unrestricted area will not exceed 500 mrem to the total body or 3000 mrem to the skin in cocpliance with the limits of 10 CFR Part 20.

For Specification 2.4.3.a(1), gn==a and beta dose factors for the individual noble gas radionuclides have been calculated for the plant gaseous release points and are provided in Table 2.4-5.

The expressions used to calculate these dose factors are based on dose models derived in Section 7 of i

Meteorology and Atomic Encrev-1968 and model techniques provided in Draf t Regulatory Guide 1.AA.

4 I

Dose calculations have been made to determine the site boundary location with the highest anticipated dose rate fras noble gases using on-site meteorological s

data and the dose expressions provided in Draf t Regulatory Guide 1. AA.

The dose expression considers the release point location, building wake effects, and the physical characteristics of the radionuclides.

l The offsite location with the highest anticipated annual dose from released i

noble gases is 700 meters in the SSE directioni l

The release rate Specifications for radiciodine and radioactive material in particulate form with half-lives greater than eight days are dependent on existing radionuclide pathways to man.

The pathways which were examined for these Specifications are:

1) individual inhalation of airborne radionuclides,

2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, and,3) deposition onto grassy areas where milch antsals graze with consumption of the milk by man, which was determined to be the most limiting pathway.

Methods for estimating doses to the thyroid via these pathways are described in Draf t Regulatory Guide 1.AA.

TS B.2.4-15

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The offsite location with the highest anticipated thyroid dose rate from radiolodines and radioactive material in particulate form with half-lives i

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greater than eight days was detensined using on-site =eteorological data and the exprescions described in Draft Regulatory Guide 1. AA.

Specification 2.4.3.a(2) linits the release rate of radiciodines and radioactive material in particulate form with half-lives greater than eight days so that the corresponding atnual thyroid dose via the most restrictive pathway is less than 1500 mrem.

For radiciodines and radioactive material in particulate form with half-lives greater than eight days, the most restrictive location is a dairy fans located 3700 meters in the NNE direction (vent X/Q = 4.3x10-7sec/m ; stack 3

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X/Q = 2.5x10 sec/m ).

Specification 2.4.3.b establishes upper effsite levels for the releases of A

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noble gases and radioiodines and radioactive rasterial in particulate fons

\\s with half-lives greater than eight days at twice the design objective annual quantity during any calendar quarter, or four times the design objective annual quantity during any period Of 12 consecutive months.

In additien to the limiting conditions for operation of Specifications 2.4.3.a and 2.4.3.b",'

the reporting require =ents of 2.4.3.c provide that the cause shall be identified whenever the release of gaseous effluents exceeds one-half the design objective annual quantity during any calendar quarter and that the proposed program of action to reduce such release rates to the design object'ives shall be described.

Specification 2.4.3.d and 2.4.3.e are in accordance with Design Criterion 64 of Appendix A to 10 CFR Part 50.

A lV)

TS B. 2. 4-16

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4 Specification 2.4.3.f is intended to monitor the perfonnance of the core. An j

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increase in the detivity levels of gaseous releases nay be the result of defective fueld'gsnun+e4 p.d Since core performance is of utmost importance in the re; ult-1 ing doses from accidents, a report must be filed within 10 days following the specified increase in activity level at the steem jet air ejector.

,1 Specification 2.4.3.g requires that the drywell atmosphere receive treatment for the removal of gaseous iodine and particulates during purging.

4 Specification 2.h. 3.h requires that hydrogen concentration upstream of the compressed radioactive gaseous storage tanks shall be monitored whenever the co= pressed storage subeystem is in use.

Specification 2.h.31 requires offgas flow to the compressed storage tanks to be terminated in the event that the hydrogen monitors downstraa= of the recombiners are inoperable.

This prevents the possible acewsulation of an i

-,s W

explosive mixture in portions of the offgas system which are not designed to fully withstand a hydrogen detonrJ..lon.

i Specification 2.4.3.j limits the maximum gross activity in one decay tank i

~ on the basis that accidental release of its contents to the environs by cperator error after 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> decay should not result in exceeding the dose I

equivalent to the maximum quarterly release rate specified in Specificetion i

l 2.4.3.c.1.

Staff analysis of an elevated,ralease under accident meteorology for a minimum release period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> indicated a release of 22,000 curies of Xe-133 or the dose equivalent would result in an air dose from the site of 20 mrad from noble gases.

\\

Calculations have been performed to determine the relationship between steam jet air ejector offgas activity and co= position and condenser air inleakage.

These calculations were used to determine die curves presented in Figure 2.4-1.

The results of the measurement of condenser air inleakage and the average air ejector offgas release rate are used in conjunction with the TSB.2.4-17

most recent offgas isotopic analysis to determine if the maximum pemitted f) s,/

Xe-133 dose equivalent tank radioactivity contents may be exceeded.

This analysis is adequate to initiate corr,ective action in the unlikely event that the tank radioactivity limit is being approached.

Specifications 2.4.3.K and 2.4.3.1 require that the mechanical vacuum pu=p be provided with autematic isolation capability to limit the release of activity from the main condenser during an accident.

Specification 2.4.?.m provides for reporting release events which, while below the 1bsits of 10 CFR Part 20, could result in releases higher than the design objectives.

The sa=pling and monitoring requirenents given under Specification 2.4.4 provide assurance that radioactive materials released in gaseous vaste (A) effluents are properly controlled and monitored in confor:ance with the s_ -

require =ents of Design Criteria 60 and 64.

These requirements provide the data for the licensee and the Co=nission to evaluate the plant's performance relative to radioactive vaste effluents released to the environment.

l Reports en the quantities cf radioactive materials released in gaseous l

effluents are furnished to the Co=nission semiannually.

On the I

basis of such reports and any additional information the Co=sission may l

l obtain from the licensee or others, the Commission nsy from time to ti=e l

require the licensee to take such action as the Co= mission dee=s appropriate.

The points of release to the environment to be monitcred in Section 2.4.4,

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

These Specifications cre epplicable for the interim period until the date

{q,/

that Specifications prepared in accordance with new Appendix I become effective. In some cases these Specifications may be more restrictive than TS B. 2. 4-18 t

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

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

l required by Appendix I.

In the event that plant availability is

[

adversely affected by these specifications, the licensee may apply l

l to the Cocenission for appropriate Technical Specification changes i

on a case by case basis.

]

j 2.4.5 Specificatiens for nelid Waste Handlinst and Disposal s.

Measurements shall be made to determine or estimate the total curia quantity and principle radionuclide composition of all radioactive solid waste shipped offsite.

b.

j Sunnaries of radioactive solid waste shipments, volumes, principal radionuclides, and total curie quantity, shall be included in the Effluent and Waste Disposal Sectiannual Reporti 1

Bases: The require:nents for solid radioactive vaste handling and disposal i

given under Specification 2.4.5 provide assurance that solid radioactive materials stored at the plant and shipped offsite are packaged in I

conformance with 10 CFR Part 20, 10 CFR Part 71, and 49 CFR Parts 170-178.

l j

i TS B.2.4-19

TABLE TS B.2.4-1 i

RADICACTIVE LIQUID WASTE SAMPLING AND ANALYSIS Detectable Liquid Sampling Type of Concentrations S ource Frequency Activity An.!ysis (uci/ml)*

b A.

Monitor Tank Releases Each Batch Principal Comma Emitters 5 x 10 7 One Batch / Month Dissolved Gases

• 10~0 e

Weekly Compositec Ba-La-140,. I-131 10-6 Monthly Compositec Sr-89 5 x 10-5 E-3 10-5 Gross Alpha 10-7

\\

C Quarterly Composite Sr-90 5 x 10-8 i

l d

B.

Primary Coolant Weekly 1-131, 1-133 10-6 "The detectability limits for activity analysis are based an the technical feasibility and on the potential significance in the environment of the quantities released.

For some nuclides, lower detec, tion ifmits may be readily achievable, and when nuclides are-measured below the stated limits, they should also be reported.

b For certain mixtures of ga=ma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the same in much greater concentrations. Under these circumstances, it will be more '

appropriate to calculate the concentrations of such radionuclides using measured ratios with those radionuclides which are routinely identified and measured.

cA composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged.

d

-~g The power level and cleanup or purification flow rate at the sample time shall also 1be reported.

'For dissolved noble gases in water, assume a NPC of 4 x 10-5uci/mi of water.

e TABLE TS B.2.4-1

TABLE TS B.2.4-2 p

/

I RADI0 ACTIVE GASSOUS WASTE SAMPLIIU AND ANALYSIS V

Caseous Sampling Type of Detectable Source Frequency Activity Analysis Concentrations (uCi/ml)'

A. (bntain=ent Purges Each Purge Principal Gam =a E=itters 10-4 b H-3 10-6

-4 b,e B. Environ = ental Release Monthly Principal Ga:n=a Emitters 10 Points (Gas Se_.ples)

H-3 10-6 Weekly I-131 10-12 d (Charcoal Sa=ples),

O Monthly I.133, I-135 10-10 (Charcoal Sa=ples)

Weekly Principal Ga-a Emitters 10-11 d (Particulates )

(at least for Ec-In-li.O and I-131) e Monthly Co=posite g7,gg 10-11 (Particulates )

y Gross Alpha 10 -

Qtrly Co=posite' (Particulates)

Sr-90 10-11 U 'Ibe above detectability limits for activity analysis are based on technical feasibility anc on the potential significance in the enviren=ent of the quar.tities released.

For sete nuclides, lower detection limits may be readily achievable, and when nuclides are ceasured balov the stated li=its, they should also be reported.

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

c For certain mixtures of ge=ca e=itters, it =ay not be possible to measure radionuclides at n [Mvels near their sensitivity limits when other nuclides are present in the se=ple at

'2h higher levels.

Under these circu= stances, it vill be more appropriate to calculate

( le levels of such radionuclides using observed ratios with those radionuclides which are measurable.

TABLE TS B.2.4-2 (Page 1 of 2)

)

s_ /

3 TABLE TS B.2.4-2 Notes (continued) i O When the average daily gross radicactivity release rate exceeds that given in 2.4.3.c(1) or where the steady-state gross radioactivity release rate increases by 50% over the previous corresponding power level steady-state release rate, the iodine and particulate collection' devices for the release point whose coniribution exceeds 50% of these rates shall be removed and analyzed to determine the change in iodine-131 and particulate release rate. The analyses for this release point chall be done daily following such change until it is shown that a pattern exists which can be used to predict the release rate after which it may revert to weekly sampling.

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

f Isotopic analysis performed in accordance r'th Specification 2.4.4.c at the discharge of the steca jet air ejectors and at a point prior to dilution and discharge of gaseous vaste from the offgas system.

\\

Concentraticna of individual Atr.na emittere in the 3eactor L:tidine vent

,)

are expected to be below the eininum detectable lovels with the exist-ing analytical equipment.

Therefore, isotopic analyses of smaples from the vent will not norcally be perforced and the teotopie content will be assumed to be that existing at the steam jet air ejector.

1 e

'~'s I.

TABLE TS B.2.4-2 (Page 2 of 2) ee

-w-

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e D

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TABLE TS B.2.4-4 i

IDCATICIE OF CASEOUS FROCESS AND EFFLUENT MONITORS AND SAMPLERS Grab Radiation Auto Control to Continuous Sample Measure:nent Procesa Stream or Belease Point

'Alaru Isolation Valve Monitor Station Noble Gas I

Particulate H-3 Alpha Condenser / Air Ejector (before X

X X

X X

j ges treatment system)

M f p Treatment System (before X

X 4

dilutionanddischarge) i Mata Stack X

X X

X X

X X

X X

1 Riactor kilding X

X X

X X

X X

X X

2 y:stilat, ion System

\\

brbinekilding X

X X

X beerating rioor kebanical Vacuum

• X Pump

" Isolation on main steam line high radiation.

W TABLE TS B.2.4-4 I

i

p V

((q.

n

)

TABLE TS ( v -5 4

N-4 CAHMA AND BETA DOSE FACTORS Monticello Doese Factors for Vent Dose Factors For Stack N:ble Caa iv iv "iv iv is is is is b

R:dionuclide Total Body Skin

• Beta Air Gama Air Total Body Skin Beta Air Cama Air rem /yr rem /yr rad /yr rad /yr rem /yr rem /yr rad /yr rad /yr Ci/sec Ci/sec Ci/sec Ci/sec Ci/sec C1/sec C1/sec C1/sec

-4

-5 Kr-83m 2.0 x 10 0

1.6 0.13 2.0 x 10 0

0.063 S.3 x 10

~

Kr-85m 2.0 8.0 11 2.1 0.59 0.32 0.43 0.6 Kr-85 0.023 7.4 11 0.024 8.6 x 10-3 0.29 0.43

'9.1 x 10-3 Kr-87 6.6 54 57 6.9 2.5 2.1 2.3 2.7 Kr-88 15 13 16 16 6.3 0.52 0.64 6.6 Kr-89 9.4

'56 58 9.9 2.1 2.2 2.3 2.2 X2-131m 0.69 2.6 6.1 0.89 0.15 0.10 0.24 0.18 Xo-133m 0.54 5.5 8.1 0.75 0.12 0.22 0.33 0.14 X:-133 0.63 1.7 5.8 0.79 0.13 0.067 0.23 0.14

.Xs-135m 3.8 2.9 4.1 4,1

'1.2 0.16 0.16 1.3 X2-135 2.9 10 14 3.0 0.94 0.41 0.54 0.99 X:-137 1.1 67 70 1.2 0.25 2.7 2.8 0.26 Xe-138 93 23 26 9.8 31 0 91

1. 0 32 TABLE TS B.2.4-5 i

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