ML18096B448

Proposed Technical Specifications

ML18096B448
Person / Time
Site:	Saint Lucie
Issue date:	04/06/2018
From:	Florida Power & Light Co
To:	Office of Nuclear Reactor Regulation
References
Download: ML18096B448 (31)

Text

4 PROPOSl'.D TECllNICAI. SPECI: ICATIONS FOR ST I UCIE Pl ANT~ UNIT NO 1

~

Docket Number 50-335

2. 4 LliLlITING CONDITIONS FOP, OPERATION Radioactive Fffluents release of radioactive materials in liquid and gaseous effluents to the environs to ensure that these releases are as low 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 relea. es of radioactive material above background to unrestricted areas be as low as practicable as defined in Appendix

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

For liquid wastes:

a. The annual dose above background to the total body or any organ of an individual from all reactors at a site should not exceed 5 mrem in an unrestricted area.

b. The annual total quantity of radioactive materials in liquid waste, excluding tritium and dissolved gases, discharged from each reactor should not exceed 5 Ci.

2-For gaseous wastes:

c. The ~annual total quantity of noble gases above background dis-charged from the site should result in an air dose due to gamma 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 coul'd 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 forms with half-lives greater than'ight days, 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 excess of 1.5 mrem.

e. The annual total quantity of iodine-131 discharged from each reactor at a site should not exceed 1 Ci.

2.4.1 S ecifications for Li uid li>aste,Effluents

a. The concentration of radioactive materials released in liquid waste effluents from all reactors at the site shall not exceed the values specified in 10 CFR Part 20, Appendix B, Table IX, Column 2, for unrestricted areas.

b. The cumulative release of radioactive m. terials in liquid waste effluents, excluding tritium and dissolved gases, shall not exceed 10 Ci/reactor/calendar quarter.

c

c. The cumulative release of radioactive materials in liquid waste effluents, excluding tritium and dissolved gases, shall not exceed 20 Ci/reactor in any 12 consecutive months.

d. During release of radioactive wastes, the effluent control monitor shall be set to alarm and to initiate the automatic closure of each waste isolation valve prior to exceeding the limits specified in 2.4.1.a above.

e. The operability of each automatic isolation valve in the liquid radwaste discharge lines shall be demonstrated quarterly.

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

liquid wastes prior to their discharge'hen the projected 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 radwaste tank that can be discharged directly to the environs shall not exceed 10 Ci, excluding tritium and dissolved gases.

h. 'f the cumulative release of radioactive materials in I

liquid effluents, excluding tritium and dissolved gases, exceeds 2.5 Ci/reactor/calendar quarter, the licensee shall made 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 rcport these actions to the NRC in accordance with Specification 5.6.2.c(l).

i. An unplanned or uncontrolled offsite release of radioactive materials in liquid effluents in excess- of 0.5 curies requires notification. This notification shall be in accordance with Specification 5.6.2.c(3).

2.4.2 ~S ecifications for Li uid h'aste Samvlin . and .'ionitorin,

a. Plant reco'rds 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 which each discharge occurred. Sample analysis results and other reports shall bc submitted as required by Section 5.6.1 of these Specifications. L'stimates 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 discharged during the period of

'discharge.

c. Sampling and analysis of liquid radioactive waste shall be performed in accordance with Table 2.4-1. Prior to taking samples from a monitoring tank, at least two tank volumes shall be recirculated.'.

The radioactivity in liquid wastes shall be continuously monitored and 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

'xceeding 72 hours, no release from a .liquid waste tank shall be made and any release in progress shall be terminated.

e. The flow rate of liquid radioactive waste shall be continuously measured and recorded during release.

Al] 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,

g. The radioactivity in, steam generator blowdown shall be continuously monitored and recorded. [whenever these monitors are inoperable, the blowdown flow shall be diverted to the waste management system and the direct release to the environmlent terminated.

Bases: The release of radioactive materials in liquid waste affluents to unrestricted areas shall not exceed the concentration limits specified in 10 CFR Part'20 and should be as low as practicable in accordance with the requirements of 10 CFR Part 50.36a. These specifications provide reasonable assurance that the resulting 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

is provided a dependable source of power under unusual operating conditions which may temporarily result in releases higher than the design objective levels but still within the concentration limits specified in 10 CFR Part 20. It is expected that by using this operational flexibility under unusual operating conditions, and exerting every effort to keep levels of radioactive material in liquid wastes as low as practicable, 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-'nce taking into account a combination of variables including defective fuel, primary system leakage, primary to secondary system leakage, steam generator blowdown and the performance of the various waste systems, and are consistent with Appendix I to 10 CFR Part 50.

'reatment Specification 2.4.1.a requires the licensee to limit the concentration of radioactive materials in liquid waste efftuents released from the ~ ite to levels specified in 10 CFR Part 20, Appendix 8, Table 1I,'olumn 2, for unrestricted areas. This specification provides assurance that no,.

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

Specifications 2.4;l.b and 2.4.l.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 flexibility of operation to assure that the public is provided a dependable source 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 I

P of radioactive material in liquid waste effluents at small percentages of the limits specified in 10 CPR Part 20.

Consistent with the requirements of 10 CFR Part 50, Appendix A, Design Critezion 64, Specifications 2.4.l.d and 2.4.l.e require operation of suitable equipment to control and monitor the releases of radioactive materials in liquid wastes during any period that these releases are taking place.

Specification 2.4.l.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 requirements of 10 CPR Part 50.36a. Normal use and maintenance of installed equipment in the liquid waste "ystem provides reasonable 'assurance that the quantity released will not exceed the design objective. In order to keep releases of radioactive

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materials as low as practicable, the specification requires operation of equipmcnt whenever it appears that the projected cumulative dis-a charge rate will exceed one-fourth of this design objective annual quantity duringany calendar quarter.

Specification 2.4.l.g restricts the amount of radioactive material that could be inadvertently released to the environment to an amount that will not exceed the Technical Specification limit.

Xn addition to limiting conditions for operation listed under Specifications 2.4.1.b and 2.4.l.c, the reporting requirements of Specification .2.4.l.h delineate that the licensee shall identify the cause whenever the cumulative release of radioactive materials in liquid waste effluents exceeds one-half the design objective annual quantity during any calendar quarter and 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 as .required by Specification 5.6.2 of these Technical Specifications.

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

The sampling and monitoring requirements given under Specification 2.4.2 provide assurance that radioactive materials in liquid wastes are properly controlled and monitored in conformance with the requirements of Design Criteria 60 and 64. 'hese requirements provide the data for the licensee and the Commission to evaluate the plant's performance

relative to radioactive liquid wastes released to the environment.

Reports on the quantities of radioactive materials released in liquid waste effluents 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 Commissi'on may obtain from the licensee or others, the Commission may from time to time require the licensee to taI:e such action as the Commission deems appropriate.

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

2.4.3 S ecifications for Gaseous Paste Fffluents The terms used in these Specifications are as follows:

subscripts v, refers to vent releases i, refers.to individual noble gas nuclide-(Refer to Table 2.4-5 for t1ie noble gal nuclides considered)

QT the total noble gas release rate (Ci/sec)

= ~iQi sum of the individual noble gas radionuclides determined to be present by isotopic analysis

K the average total body dose factor due to gamma emission 4

(rem/yr per Ci/sec)

L the average skin dose factor due to beta emissions (rem/yr pcr Ci/sec)

H = the average air dose factor due to beta emissions (rad/yr per Ci/sec)

N ~ the average air dose factor due to gamma emissions (rad/yr per Ci/sec)

The values of K, L, ."1 and N are to be determined each time isotopic'nalysis is required as delineated in Specification 2.4.4. Determine the following using the results of the noble gas radionuclide analysis:

(1/QT) jQ(K~

(1/QT) (Q 3.

)( = (z/QT) (Q,li,.

l.

N = ()./Qg) )QiNi 3.

where the values of lii, Li, Hi and Hi are provided in Table 2.4-5, li and are site dependent gamma and beta dose factors Q

= the measured release rate of the radioiodines and radioactive materials in particulate forms with half-lives greater than eight days.

a. (1) The release rate limit of noble gases from the site shall be such that 2.9 QT K and 0.S3 QT(L + 1.1N )

(2) The release rate limit of all radioiodines 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 such that 5.5 x10 Q

b. (1) The average release rate of noble gases from the site during any calendar quarter shall be such that 13Q N <1 and 6.3 Q rr (2) The average release rate of noble gases from the site during lt any 12 consecutive months shall be 25Q N and 13 Q lf (3) The average release rate per site of all radioiodines and radio-active materials in particulate form with half-lives greater I

than eight days during any calendar quarter shall be such that 13 5.= ) Q

.(4) The average release rate per site of all radioiodines and radioactive materials in particulate form with half-lives greater than eight days during any period of 12 consecutive months shall be such that I

.25 5.5xl0 Qv

- 12 (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 Ci/reactor.

c. Should any of the conditions of 2.4.3.c(1), g2) or (3) listed below exist, 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 KPiC 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 during any calendar quarter is such that 50 Q~N or 25 Q~~1 (2) If the average re>ease 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 50 5. 5 xlG4 Av (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 wastes from the primary system waste gas holdup system the effluent monitors li ted in Table 2.4-4 shall be operating and set to alarm and to initiate the

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'I automatically closure of the waste gas discharge va'lve prior to exceeding the limits specified in 2.4.3.a'above. The operability of each automatic isolation valve shall be demonstrated C

quarterly.

e. The maximum activity to be contained in one Waste gas storage tanl: shall not exceed 110,000 'uries (considercd ..s Xe-133).

f. An unplanned or uncontrolled offsite release of radioactive materials in gaseous effluents in excess of 5 curios of noble gas or 0.02 curie of radioiodine in gaseous form requires notification.

This notification shall be in accordance ~~ith Specification 5.6.2.c(3).

S ecifications for Gaseous 3laste Sam lin and ';fonitorina

a. Plant records shall be maintained and reports of the sampling and analyses results shall be submitted in accordance with Section 5.6 of these Specifications. Estimates of the sampling and analytical error .associated ~rith each-reported value should be included.

b. Gaseous releases to the environment, except from the turbine building ventilation exhaust and as noted in Specification 2.4.4.c, shall be continuously monitored for gross radioactivity and the floe continuously measured and recorded. lG>enever these monitors are inoperable, grab samples shall be taken and analyzed daily for gross radioactivity. If these monitors are inoperable for more than seveg days, these releases shall be terminated.

c. During the release of gaseous wastes from the primary syst: em activity monitor,

~ waste gas holdup system, thc gross the iodine collection device, and the particulate collection device shall be operating.

d. 'All waste gas effluent monitors, shall be calibrated at least quarterly by means of a known radioactive source which has been calibrated'o a Tfational Bureau of Standards source. Hach monitor sh'all have a functional test at least monthly and instrument check at least daily.

e. Sampling and analysis of radioactive material in gaseous waste, including particulate forms and radioiodines shall be performed.

in accordance with T ble 2.4-2.

Bases: The release of radioactive materials in gaseous waste effluents to unrestricted areas shall not exceed the concentration limits specified in 10 CFR Part,20 and should be as low as practical in accordance with the requirements of 10 CPR Part 50.36a. These specifications provide reasonable assurance that the resulting annual air dose from the site due to gamma radiation will not exceed 10 mrad, and an annual air dose from the site due to beta radiation will not exceed .20 mrad from noble gases, that no individual in an unrestricted area will receive" an annual dose to the total body greater than 5 mrem or, an annual skin dose greater than 15 mrem from fission product noble gases, and that the annual dose to any organ of an individual from radioiodines and radioactive material in particulate form with half-livbs greater than eight days will not exceed 15 mrem per site.

II I h

15 At the same time those specifications permit the flexibility of operation, compatible vith considerations of health and safety, to assure that the public is provided ~~ith a dependable source of power under unusual operating conditions i"hich may temporarily result in releases higher than the design objective levels but still vithin the conccntratien limits specified in 10 CFR Part 20. Even ~"ith this operational flexibility under unusual operating conditions, if the licensee exerts every effort to keep levels of radioactive material in gaseous ~aste effluents as lov as. practicable, the annual releases vill not exceed a small fraction of the concentration limits specified in 10 CFR Part 20.

The design objectives have been developed based on operating experience taking into account a combination of system variables including defective fuel, primary system leakage, primary to secondary system leakage, steam generator blovdo~m and the performance of the various ~;aste treatment systems.

Specification 2.4.3.a(1) limits the release rate of noble gases .from the site so that the corresponding annual gamma and beta dose rate above background to an individual in an unrestricted area vil3 not exceed 500 mrem to the total body or 3000 mrem to the skin in compliance vith the limits of 10 CFR Part 20.

For Specification 2.4.3.a(1), gamma 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 techniques provided in Draft Regulatory Guide 1.M.

Dose. calculations have been made to determine the site boundary location with the highest anticipated dose rate from noble gases using on-site meteorological data and the dose expressions provided in Draft Regulatory Guide 1.AA. The dose expression considers the release point

,location, building wake effects, and the physical characteristics of the radionuclides.

The offsite location with the liighest anticipated annual dose from released noble gases is 1600 meters in the ~%orth direction.

The release rate Specifications for a radioiodine and radioactive material in particulate form with half-lives greater than eight day' are dependent on existing radionuclide pathways to man. The pathways which were examined for these Specifications are: 1) individual inhalation of airborne radionuclides, 2) depo ition of radionuclides onto green leafy vegetation with subsequent consumption by man, and 3) deposition onto grassy areas where milch animals graze with consumption of the milL. by man. hlethods for estimating doses to the thyroid via these pathways are described in Draft Regulatory Guide 1.AA. The offsite location with the highest anticipated thyroid dose rate from radioiodines and radioactive material in particulate form with half-lives greater than eight days was determined using on"site meteorological data and the expressions described in Draft Regulatory Guide 1.AA.

Specification 2.4.3.a(2) limits the release rate of radioiodines and radioactive material in particulate form with half:-lives greater than eight days so that thc corresponding annual thyroid dose via the most restrictive pathway is less than 1500 mrcm.

For radioiodines and radioactive material in particulate form with hali-lives greater than eight days, the most restrictive location is a dairy farm located'2,000 meters in the SSM direction (vent X/Q= 3.5x 10 sec/m 3 ),

Specification 2.4.3.b establishes upper offsite levels for the releases of noble gases and radioiodines and radioactive material in particulate form 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 addition to the limiting conditions for operation of Specifications 2.4;3.a and 2.4.3.b, the reporting requirements 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 objectives shall be described.

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

r Specification 2.4.3.e limits the maximum quantity of radioactive gas U

that can be contained in a waste gas storage tanl;. The calculation of this quantity should assume instantaneous ground release, a X/Q based

5 percent meteorology, the average gross energy is 0.19 Hev per disintegration (considering Xc-133 to be the principal emitter) and exposure occurring at thc minimum site boundary radius using a semi-infinite cloud model. The calculated quantity will limit the offsite dose above background to 0.5 rem or less, consistent with Commission guidelines.

Specification 2.4.3.f provides for reporting 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 give'n under Specification 2.4.4 provide assurance that radioactive materials released in gaseous waste effluents are properly controlled and monitored in conformance with the requirements of Design Criteria 60 and 64. Thyrse N

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. 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.

The points of release to the environment to be monitored in Section 2.4.4 include all the monitored release points as provided for in Table 2.4-4.

Specification 2.4.4.b excludes monitoring the turbine building ventilation exhaust since this release is expected to be a negligible release point. ."fany PVR reactors do not have turbine building enclosures.

To be consistent in this requirement for all PHR reactors, the monitoring of gaseous releases from turbine buildings is not required; 2.4.5 Specifications for Solid Haste Handlin, and Dis osal

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

b. Reports of the radioactive solid waste shipments, volumes, principle radionuclides, and total curie quantity, shall be submitted in accordance srith Section 5.6.1, Bases: The requirements for solid. radioactive waste handling and disposal given under Specigication 2.4.5 provide assurance that solid radioactive materials stored at the plant and shipped offsite are packaged in conformance with10 CFR Part 20, 10 CFR Part 71, and 49 CFR Parts 170-178.

TABLF 2.4.1 RADIOACTIVE LIQUID SAi'IPLING AND ANALYSIS Dclcct JI)fr.

Littuid $ antnl Inc Type of Concrnttations Source Frcqurncy Activity AnalysiS (ti C i/r nil a

~b A. Monitor Tank Releases Each Batch Principal Gamma Emi:ters 5xIO One Batch/Moltth Dissolved Gases f 10 6

V/eekly Compositcc Ba La 14iO, I.131 10 Sr 89 5 x 10" Monthly Composite'3 10 Gross a. 10 Quarterly Composite Sr 90 5x10 B. Primary Coolant Neeklyc 1.131, I.133 10 b

C. Steam Generator Blowdown Principal Gamma Erni;ters Sx10 Weekly'ne 6 Ba La.140, I 131 10 Sample,/Month Dissolved Gases f 10 Sr-89 5x10 Composite'.3 10 'onthly Gross ct 10 Quarterly Composite Sr 90 5x 10 The detectability limits for activity analysis are based on the technical feasibility and on the potential sirnificance in .

the environment of the quantities released. For some nuclides, lower detection limi;s may be readily achievablc,'and when nuclides are measured below the sta;ed limits, they should also be reported.

b For certain mixtures of gamma emitters, it Jnay not be possible to measure raJ'.onuclicles in co,tcentrations near their sensitivity limits when other nuclides are pre:.ent in the sample in much gtqater concentrations. Under thes". circum.

stances, it will be more appropriate to calculate the concentrations of such tatlionuclidcs using nleaSured ratios with those radionuclidcs which are routinely identified and measured.

A composite sample is one in which the quantity of liquid sampled is proportional to thc quantity of liquid Yrastc discharged.

d Thc power level and cleanup or purification flow rate at thc sample titnc shall also be reported.

c To be representative of the average quJntities and concentrations of radioactive materials in liquid efflucnts, samples should bc collected in proportion to thc rate of flow of the effluent stream. Prior to analyses, all samples takeit fOr the composite sltould be thoroughly mixed in order for the contposite sample to be representative of the average effluent release.

f of 4 x pCi/ml of water.

For dissolved noble gases in water, assume a MPC 10

TABLE 2A.2 RADIOACTIVE GASLOUS WASTE SAI lf'Lli'G AND ANALYSIS Detcctct>tc Gaseous Sc:nnling Tyne of Concenltat>ons Source Frequency Activity Analysis fpCt/ml)~

A. Waste Gas Decay Tank Releases Fach 1 ank Principal Gamma Emit ters 10" ~b 10 B. Containmcnt Purge Releases Each Purge Principal Gamnta Emittcts 10 H3 10 I

C. Condenser Air Ejector Monthly Principal Gamma Emitter s 10 10 O. C D. Environmental Release Points Monthly Principal Gamma F.mit ters 10 (Gas Samples)

H.3 6 10 Weekly (Charcoal Sample) I ~ 131 10 (Charcoal Sample) I.133, I 135 10-t o

'onthly Weekly (Particulates) Principal Gamma Emitters (at least for Ba La.140,1-131) 10 II Monthly Compositcu Sr 89 1P (Particulates)

II Gross tt 1P Quarterly Composite Sr.90 10-"

(Par ticulates)

The above detectability limits for activity analysis are based on technical feasibility and on tile pcftcntial sicnificance in the environntcnt of thc quantities released. For some nuclides, lower detection limits may bc readily achicvablc, and when nuclides are measured below the stated limits, tltcy should also be reported.

For certain mixtures ol gamma emitters, it may not bc possible to measure radionuclides at levels near their sensitiv:

ity lifnits when other nuclides are present in tltc sample at much higher levels. Under tltese circumstances, it tvill be more appropriate to calculate the levels of such radionuclides using obs';vcd ratios with those radionuclides which are measurable.

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

To bc rcpresentativc of the average quantities and concentrations of radioactive materials in particulate form released in gaseous effluents, samples should bc collected in proportion to the rate of flow of the effluent stream.

TABLE 2.4-3 PRESSURIZED WATER REACTOR LIQUID WASTE SYSTEM LOCATION OF PROCESS AND EFFLUEi JT MONITORS AND SAMiPLERS REQUIRED BY TECHNICAL SPECIFICATIONS GIau Measurement }I11th Siinil)le Isotopic L ino id Radiation AU to Cont Iol to C 0 I i l I I IV0 U S GIOSS Dissolved Level Process Stream or Release Point Alarm isolation Valve Momtor S 'iil 1 011 ACtivily Gilses Alpha H.3 AnalySiS Ala IITI Miscellaneous Waste Sample (Test) Tank X Chemical Waste Sample (Test) Tank X X Detergent Waste Collccto'ank' X X Primary Coolant System Liquid Radwaste.Discharge Pipe Steam Generator Blowdown Systemb X X Service 'Water Discharge Pipe Outdoor Storage Tanks (potentially radioactive) X X x ~

Component Cooling Systems X X Turbine Building Sumps (Floor Dr ins)

In most PWRs the con',crts of thc detergent waste collector tank arc sampled, analy.-.cd, and thc>> filtered prior to release through thc liquid radtva:te discharge pipe. The detcrlent waste system should bc designed with either a split tank or two scpa. atc collection or sample (tcs',) tanks to permit isolation of thc tanks for mixing, sampling, and a>>alysis prior to relcilsc.

In soma PWRs processed liI;ttid from the stcam ge>>crator blovrdown system is returned directly to thc scen>>tlary system, and .Ihc need for co>>tinuous nlOnitOr-ing at this rclcasl: point is climinalcll.

TABLE 2A-4 PRESSURIZED WATER REACTOR GASEOUS WASTE SYSTEM LOCATION OF PROCESS AND EFFLUENT MONITORS AND SAMPLFRS REQUIRED BY TECHNICAL SPECIFICATIONS Grnlr Auto Corrirol to COII<inu(lus S ' l I i I le 1

h1< isurr rnc'nt Process Stream or Release Point lsoliltloll Valve 4".onitoi S<,I<rorr N<rl>le Gas Par t I cula'ie H.3 Alpha Waste Gas Storage Tanks x

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Condenser System'larm Air Ejector Vent Header Buildir.g Ventilation Systems Reactor Containment Building (whenever there is flow) X Auxiliary Building' Fuel Handling It< Storage Building" X Radwaste Building'team Generator Blowdown Tank Vent or Condcnscr Vent Turbine Gland Seal Condenser X X X Mechanical Vacuum Pump Waste Evaporator Condenser Vent X X X lf any or all of the process streams or building ventilation sys'.cms are routed to a single re!case point, thc nccd for a continuous monitor at the individual dis-charge point to thc main exhaust duct is eliminated. Qnc continuous monitor at thc fiital rclca:c point is sufficient.

In some PWRs thc stcam generator blowdown tank vent is routed to the main turbi>>c condc>>scr. and the nccd for a continuous monitor at this re!case point is eliminated.

For PWRs in which thc wast<. evaporator cortdcnscr is vc>>tcd directly to thc atmosphere.

Table 2.4-5 GAMMA AND BETA DOSE FACTORS FOR St. Lucio Plant, Unit 1 X/Q = 2.1 x 10 sec/m3 Dose Factors for Vent Noble Ki L ~

Miv Niv Gas'adionuclide Total Body Slcin Beta Air Gamma Air ramlrr ((~rem/ r (~rad/ ~rad/ rX

( ) KCi/scc/ r'Ci/sec Ci/sec I Kr-83m 5.8 x 10 5 0 0.6 0:028 Kr-85m Oa38 3.1 4.1 0.92 Kr-85 0.014 2.8 4.1 0.015 Kr-87 1.9 20 22 2.0 Kr-SS 6.0 5.0 6.2 6.3 Kr-89 0.5 21 22 0.52 Xe-131m 0.4 1.0 2-3 0.5 Xe-133m 0.3 F 1 3.1 0.41 Xe-133 0 '6 0. 64 2:2 0. 45 Xe-135 0.64 1.5 -l. 6 0.68 Xe-135 1.5 3.9 5.2 1.6 Xe-137 U.u/2 26 27 0.076 Xe-.138 1.5 8 7 10 1.6

5.6.2 c. Nonroutine Radioactive Effluent Reports (1) PWR Liquid Radioactive Wastes Report. If the cummula" ive releases of radioac'rive materials in liquid effluents, excluding tritium and dissolved gases, should exceed one-half the design objective annual quantity during any calendar quarter, the licensee shall make an investigation to identify the causes of such releases and define and initiate a program of action to reduce such releases to the design objective levels. A written report of these actions sh'all be sub-mitted to the NRC within 30 days from the end of the quarter during which the release occurred.

(2) PWR Gaseous Radioactive Wastes Report. Should the condit'ions (a),

(b), or (c) listed below exist, the licensee shall make an invest-igation to identify the causes of the release rates and define and initiate a program of action to reduce the release rates 'to design objective levels. A written report of these actions shall be sub-mitted to the NRC within 30 days from the end of the quarter during which the releases occurred.

(a) If the average release rate of noble gases for the site during t

any calendar quarter exceeds one-half the design obejctive annual quanity.

(b) 'f the average release rate per site of all radioiodines and radioactive materials in par"iculate form with half-lives greater than eight days during any calendar. quarter exceeds one-half the design objective annual quantity.

I ~

2 (c) If the amount of iodine-131 released during any ca}.endar quarter is greater than 0.5 Ci/reactor.

(3) pMR Unplanned or Uncontrolled Release Report. Any unplanned or un-controlled offsite release of radioactive materials in excess of 0.5 Curie in liquid or in excess of 5 Curies of noble gases or 0.02 Curie of radioiodines in gaseous form requires notification.

This notification must be made by a written report within 30 days to the NRC. The report shall describe the event, identify the causes of the unplanned or uncontrolled release and report actions taken to prevent recurrence.