L-76-213, Proposed Amendment to Facility Operating License DPR-67. Proposed Changes Modify Technical Specifications So They Are in Compliance with Appendix to 10 CFR 50

Proposed Amendment to Facility Operating License DPR-67. Proposed Changes Modify Technical Specifications So They Are in Compliance with Appendix to 10 CFR 50

ML18096B515
Person / Time
Site:	Saint Lucie
Issue date:	06/04/1976
From:	Robert E. Uhrig Florida Power & Light Co
To:	Stello V Office of Nuclear Reactor Regulation
References
L-76-213
Download: ML18096B515 (27)

Text

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12. /6) 50-335 FILL'UMIIEB NRC D)STRlBUTION Fon PART 60 DOCl(ET MATERIAl F IIOM; DATE OF DOCUI.IFNT FLORIDA POWER & LIGHT COMPANY 6/4/76 MRS VXCTOR STELLO MIAMI, FLORXDA DATE BECEIVEO MR ROBERT ED UHRIG 6/7/76

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LTD NOTORXZED 6/4/76 TRANS THE FOLLOWING: CONSISTING OF PROPOSED CHANGES TO MODIFY TECH SPECS SO THEY ARE IN COMPLIANCE-WITH APPENDIX I TO 10 CFR 50

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P gal%> FLORIDA POWER & LIGHT COMPANY June 4, 1976 L-76-213 Director of Nuclear act r Regulation Attn: Mr. Victor Stello, Director Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Stello:

Re: St. Lucie Plant Unit No. 1 Docket No. 50-335 Proposed Amendment to Facility 0 eratin License DPR-67 In accordance with 10 CFR 50.30, Florida Power Light S Company submits herewith forty (40) copies and three (3) signed originals of a request. to amend Appendix A of Facility Operating License DPR-67. The proposed changes are shown on the accompanying Technical Specification, pages bearing the date of this letter in the lower right,

,hand corner.

i The proposed changes modify the Technical, Specifications so that they are in',compliance with Appendix I to 10 CFR 50.

The modification was'made by applying a linear reduction to the data in our current Technical Specifications providing a practical implementation of compliance.

The proposed amendment has been reviewed and the conclusion reached that it does not involve a significant hazards con-sideration; therefore, prenoticing pursuant to 10 CFR 2.105 should not be required.

We would be pleased to meet with Staff members to discuss these changes. It would assist us to have such a meeting at a Florida site, since our technical staff members responsible for handling,our Specifications are involved in heavy operational commitments for the next several months.

Very truly yours, obert E. Uhrig Vice President REU/NR/hlc Attachment cc: Norman C. Moseley Jack R. Newman, Esq.

P EOP LE... SER V IN G P EOP LE

0 STATE OF FLORXDA ) ..

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COUNTY .OF DADE )

Robert E. Uhrig, being first duly, sworn, deposes and says:

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That'he is a Vice President Of Florida Power 1

& Light Company, the Licensee'erein>

That he has'xecuted the foregoing document; that 'the state-ments made in this said document age true and correct to the best of his knowledge, information and belief, and that he is authorized Co execute the document on behalf, of said Licensee.

Robert E. Uhr g Subscribed and sworn to before me this 8 -, day of 1976 otary Public, n an for the County of Dade,, State'of Florida .

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80IARY gINLIC STATE OF flORIDA Af LARQ QY, COMMISSION EXPIRES NOV. 30 1979 BONDED THRU CfNERAL INS. UNDERWRITERS gy commission expires:

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2. 4: RADIOACTIVE EFFLUENTS

. ~Ob ective

'o define the limits and conditions=for the controlled release of radioactive materials in liquid and gaseous effluents to the environs to ensure that 6/1/76

2-4 these releases are as low as reasonably achievable. These releases should not result in radiation exposures in unrestricted areas greater than a few percent of natural background exposures. The concentration of radioactive materials in effluents shall 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 reasonably achievable, the following design ob5ectives apply:

For liquid wastes:

a. The annual dose above background to the total body or any organ of an individual from each reactor at a site should not exceed 3 mrem or 10 mrem, respectively, 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'xceed 5 Ci.

For gaseous wastes:

c. The annual total quantity of noble gases above .background discharged from the site should result in an annual air dose due to gamma radia-tion of less than 10 mrad, and an annual air dose due to beta radiation of less than 20 mrad, at any location near ground level which could be e occupied by individuals at or beyond the boundary of the site, and 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 this release quantity.

d. The annual total quantity of all radioiodines and radioactive material in particulate forms with half-lives greater than eight days, above background, from a reactor 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 15 mrem.

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

2.4.1

~ ~ Li uid Waste Effluents

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

b. The cumulative release of radioactive materials in liquid waste effluents, excluding tritium and dissolved gases, shall not exceed 6 Ci/reactor/

calendar quarter.

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

6/1/76

2-5

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, except as provided in 2.4.2.d below.

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

'ischarge lines shall be demonstrated quarterly.

The equipment installed in the liquid radioactive waste system shall be. maintained'nd shall be operated to process radio'active liquid wastes prior to their discharge when the projected cumulative release exceed 1.25 Ci/reactor/calendar quarter, excluding tritium and 'ould dissolved gases.

h go 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. If the, cumulative release of radioactive materials in liquid effluents, excluding. tritium .and dissolved gases, exceeds 2.5 Ci/reactor/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 in accordance with Specification

5. 6. 2. b (1) .

An unplanned or uncontrolled offsite release of radioactive materials in liquid effluents in excess of 0.5 curies requires notification.

'his notification shall be in accordance with Specification 5.6.2.b(3).

2.4.2 Li uid Waste Sam lin and Monitorin a0 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 which each discharge occurred. Sample analysis results and other reports shall be submitted as required by Section 5.6.1 of these Specifications. Estimates of the sampling and analytical errors associated with each reported value shall be included.

b. 'Prior to'elease 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.

6/1/76

2-6 .

P TABLE 2.4-1 RADIOACTIVE LIQUID SAMPLING AND ANALYSIS Liqui,d Sampling ,

Type of Detectable Source Frequency. Activity Analysis Concentrati~ns

( Ci/ml)

- 72 Monitor. Tank Each Batch Principal Gamma Emitters 6 Sx10 Releases 5

One. batch/month Dissolved Gases 10

'3 Weekly Composite Ba-La-140,I-131 10 Monthly H-3 Composite B. Steam Generator Blowdown Releases Quarterly Composite

~ 4,6,7 One Sample/

Gross e Sr-90, Sr-89 Principal Gamma Ba-La-140, I-131 Dissolved Gases 5'0 emitters Sx10 5

10 x 10 10 10 2

Month Monthly 10 Composite Gross a '10 Quarterly Sr-90, Sr-89 5 x 10 Composite 6/1/76

2.7 TABLE 2.4-1 (Cont'd) 1 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.

2 For certaih mixtures of gamma 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 appropriate to calculate the concentrations of such radionuclides using measured ratios with those radionuclides which are routinely identified and measured.

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

4 To be representative of the average quantities and concentrations of radioactive materials in liquid effluents, samples should be collected in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite should be thoroughly mixed in order for the composite sample to be representative average effluent release, of'he 5 -5 For dissolved noble gases in water, assume a MPC of 4 x 10 pCi/ml of water.

6 When operational or other type of limitations preclude specific gamma spectrum analysis of each tank, gross activity measurements shall be made to estimate the quantity and concentration of radioactive material released in the batch and a weekly sample composited from proportional aliquots from each batch released during the week shall be analyzed for the- pricipal gamma emitting radionuclides.

7 No sampling required when cold and drained.

6/1/76

"0 2-8

d. 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 independ-ently 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 waste tank shall be made and any release in progress shall be terminated.

es The flow rate of liquid radioactive waste shall be continuously measured and recorded during release. If the flow monitors are inoperable, the release flow shall be determined and recorded each hour based upon an estimate of the flow rate of the system.

h 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 instrument check prior to making a release.

gs The radioactivity in steam generator blowdown shall be continuously monitored and recorded. With one steam generator blowdown monitor inoperable, the sampling system shall be realigned so that the operable monitor is receiving flow from both steam generators. Whenever both monitors are inoperable, the blowdown flow shall be diverted to the waste management system and the direct release to the environment terminated.

h. The points of release to the environment to be monitored in this section 2.4.2 include all the monitored release points as provided for in Table 2.4-2.

i. Service water discharge shall be monitored for gross activity weekly if component cooling H20 monitor is out of service and the component '5 cooling water is >1 x 10 pCi/cc.

Bassa.:

The release of radioactive materials in liquid waste effluents to unre-stricted areas shall not exceed the concentration limits specified in 10 CFR Part 20 and should be as low as reasonably achievable in accordance with th requirements of Appendix I to 10 CFR Part 50. 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 3 mrem and 10 mrem, respectively,'per reactor.

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.

I ThbLE 2,4-2 ST+ LQCIE PLhHT LIQUID QhSTE SVSTEH LOChTIOM OF PROCESS AD EFFLUKHZ HQHITORS NHl ShHPLRRS Rs)UIRRD bY TECUUIChL SPECIPIChTIOHS High huto Control Crab Heasuremen Liquid Process Stream or Radiation to Continuous 'Sample Cross Dissolved Isotopic Level .

Release Point hiarm Isolation Valve Honitor Station 'hccivity I Cases hlpha . H-0 hnslysia hlarm ~

Hiscelleneous Paste Sample (Test) Tank (Masts snd Boric X X I '

hcid Condensate Tanks)

Detergent Masts ~

'

Collector Tank (Laundry Drain Tanks)

Primary Coolant System Lfqufd Radusste Discharge Pipe X Steam Cenerator bloM doun System .'X X

'

Outdoor Storage Tanks (potentially radio-active)

Component Cooling Systems Turbine Building Floor Drains (Storm Drains) x(') I(a)

(a) Grab samples to be taken and analyzed each S hours Mhen the gross activfty in S the socondary coolant system aacesda 10 pCf/ml>

'4

2-10 The design objectives have been developed based on operating experience 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 treatment systems, and are consistent with Appendix I to 10 CFR Part 50.

Specification 2.4.1.a requires the licensee to limit the concentration of radioactive materials in liquid waste effluents released from the site to levels specified in 10 CFR Part 20, Appendix B, Table II, Column 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 2.4.1.b and 2.4.l.c establish the upper limits for the release of radioactive materials in liquid effluents. The intent of these specifica-tions 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 of radioactive material in liquid waste effluents at small percentages of the limits specified in 10 CFR Part 20.

Consistent with the requirements of 10 CFR Part 50, Appendix A, Design Criterion 64, Specifications 2.4.1.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'icensee 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 reasonably achievable consistent with the requirements of Appendix I to 10 CFR Part 50. Normal use and maintenance of installed equipment in the liquid waste system provides reason-able assurance that the quantity releases will not exceed the design objective.

In order to keep releases of radioactive materials as low as reasonably achievable, the specification requires operation of equipment whenever it appears that the projected cumulative discharge rate will exceed one-fourth of this design objective annual quantity during any 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.

In addition to limiting conditions for operation listed under Specifications 2.4.l.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 cumula-tive release of radioactive materials in liquid waste effluents 6/1/76

2-11 exceeds one-half the design objective annual quantity 'dur'ing 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 spilla'ge 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 provide assurance that radioactive materials in liquid wastes are properly controlled 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 radio-active liquid,wastes released to the environment. Reports on the 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 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 2.4.2 include all the monitored release points as provided for in Table 2.4-2.

4 2.4.3 Gaseous Waste Effluents The terms used in these Specifications axe as follows:

I subscript"v, refers to vent releases subscript i, refers to individual noble gas nuclide (Refer to Table 2.4-3 for the noble gas nuclides, considered) g:--. the total noble gas release rate (Ci/sec)

,Zgi sum of the individual noble gas radionuclides determined to be pres'ent by isotopic anal'ysis K the average total body dose factor due, to gamma emission (rem/yr per Ci/sec) the average skin dose factor due to beta emissions (rem/yr per Ci/sec) the average air dose factor due to beta emissions (rad/yr per Ci/sec) the average air dose factor due to"gamma emissions (rad/yr,'er Ci/sec) 6/1/76-

2-12 ThBLB 2 4>>3 GABE AND BETh DOSE PhCTORS PQR

'T UJCIE PUNT'NIT 1 X/Q ~ 2,1 pg 10 aecfio DOSE FACTORS FOR VEHT HOBL8 GAS iv -iv "iv iv Total Body Skin Beta Air Ganaaa hie BADIONUCLIDE

~rea r ~rad r

~rem

'i/sec

'adlarr Ci/sea Ci/sec Ci/Sec Xr-8Q 5.8 x 10 0.6 Oe 028 Xr-85m 0.88 3+1 Xx 85 0.m4 4el 0,015 Kr-&7 , 1,9 20 22 Kr-88 6,2 6.3 Kr-89 0,5 21 22 0 52 Xe-131m 0 4 1,0 2.3 0,5 Xe-133+ 0.3 3 ' 0.41

'e-133 0,36 Oe64 2.2 0,45 Ye-135ca p 64 15 ' 16 0,68 Xe-135 '1 5 3' 5.2 1 6 0

Xe-137 0, 072 27 0, 076 Xe-138 1e5 ,8.7 10 1e6 S/1/76

2-13 The values of K, L," M and N are to be determined each time isotopic analysis is required as delineated in Specification 2.4;4. Determine the following using the results of the noble gas radionuclide analysis:

K ~ (1/Q )ZQ K

- ( /Q )ZQ M ~ (1/g )ZQiLi

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N ~ (1/Q )ZQiNi where the values. of Ki-, L , M and Ni are provided in Table 2.4-3, and are site dependent gamma and keta dose factors

.Q ~ the measured release rate of the radioiodines and radioactive materials in particulate forms with half-lives greater than

. eight days.

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a. (1) The maximum release rate. limit of noble gases from the site shall be such that 20 (Q K) <1 0.33 ( (L + 1.1N )) ( 1 (2) The maximum 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 x 10 3 Q 1

b. (3;). The average release rate of noble gases from the site during any calendar quarter shall be such that 50 (QTvN )

and 25 (Q t

..M), <1 (2) The average release rate of gases from the site during any 12 consecutive months shall be 6/1/76

2-14 100 (Q N ) <.1 200,(Q K) < 1 and 50 (Q M) <1 . 67 (Q~L) <1 (3) The average release rate per site of all radioiodines and radio-active materials in particulate form with half-'lives greater than

. eight days, during any calendar quarter shall be such that

'50, (5.5x10 Q ) < 1, (4) The average release rate per site of all radioiodines and radio-active materials in particulate form with half-lives greater than eight days during any period of 12 consecutive months shall= be

'such that 3

100 (5.5x10, Q ) < 1 (5) The amount of iodine-131 released during any calendar quarter shall not exceed 0.5 Ci/reactor.

(6), The amount of iodine-131 'released during any period of the 12 conse-,

cutive months shall not exceed 1 Ci/reactor.

c. Should any of the conditions of 2.4.3.c(l), (2) 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 Sectio'n 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

'I average release rate of noble gases from the

/

site during any calendar quarter is such that 1 100 (Q N ), v-.l Tv v or 50 (Q M )

(2) If the average release rate per site of all radioiodines and radio-active material's in particulate form with half-lives greater than eight days during any calendar quarter is such that 100 (5.5xl0 3 Q ) > 1 (3) If the amount of iodine-131 released uring any calendar quarter is greater than 0.25 .Ci/reactor.

6/1/76

2-15 d0 During the release of gaseous wastes from the gas decay tanks, the gaseous discharge monitor shall be operating and set to alarm and to initiate the automatic closure of the waste gas discharge valve prior to exceeding the limits specified in 2.4.3.a above. Whenever this monitor is inoperable for a period not to exceed seven days, two independent. samples of each gas decay tank to be discharged shall be analyzed and two plant personnel shall independently check valving prior to the discharge. If this monitor is inoperable for a period exceeding seven days, no release from a gas decay tank shall be made

,. and any release in progress shall be .terminated., The operability of, each automatic isolation valve shall be demonstrated quarterly.

'I t

e. The .maximum activity to be contained in one waste gas storage tank shall not exceed 110,000 curies I

(considered as Xe-133).

f." An unplanned'r uncontrolled offsite release of radioactive materials

,in gaseous effluents in excess of 5 curies of noble gas or 0.02 curie of radioiodin'e in gaseous form requires notification. This notifica-tion shall be in accordance with Specification 5.6.2.b(3).

2.4.4 Gaseous Waste Sam 1

lin and Monitorin 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 with each reported value should be included. 'ssociated

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. Whenever these monitors are inoperable; grab samples shall be taken and analyzed daily for gross radioactivity. If these monitors are inoperable for more than seven days, these releases shall be terminated.

During the release of gaseous wastes from the primary system waste gas holdup system,- the iodine collection device, and the particulate col-lection device shall be operating, except as noted in 2.4.3.d'bove.

All waste gas effluent monitors shall be calibrated at least quarterly by means of a known radioactive source which has been calibrated to a National Bureau of Standards source. Each monitor shall have a func-tional test at least monthly and instrument check at least daily.

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

.accordance with Table 2.4-2. The points of release to the environ-ment to be monitored in this Section 2.4.4 include all the monitored release points as provided for in Table 2.4-4.

6/1/76

2-16 TABLE 2. 4-4 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS'aseous Detectable Sampling Type of Concentration Source Frequency Activity Analysis (PCi/ml) 2 A. Waste Gas Decay Tank ',Each Tank - Principal Gamma Emitters 10 6 Releases H-3 10 3

B. Containment Purge Each Purge Principal Gamma Emitters 10 Releases H-3 5 42 7 3 C." Condenser Air Ejector Monthly Principal Gamma Emitters 10 6

H-3 10'273 D. Enviionmental Release 'onthly Principal Gamma Emitters 10 6 10 Points (Gas Samples) H-3 12

'eekly (Charcoal I-131 10 Sample) 10 Monthly (Charcoal I-133,I-135 1~

Sample)

Weekly (Particulates) Principal Gamma Emitters (Ba-La-140, I-131, and 4

others)

Monthly Composite (Particulates) 10-11 Gross a Quarterly Composite 4 Sr-90, Sr-89 10-11 (Particulates)

The above detectability limits for activity'and analysis are based on technical feasi-bility 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 2are measured below the stated limits, they should also be reported.

i:Ear certain mixtures of gamma emitters, it may not be possible to measure radionuclides

'at-levels'"near their sensitivity limits when other nuclides are present in the sample it

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at much higher levels. Under these circumstances, wiU. be more appropriate to cal-culate the levels of such radionuclides using observed ratios with those radionuclides 3which are measurable.

Analyses shall also be performed following each refueling or similar operational oc-4currence which could alter the mixture of radionuclides.

To be representative of the average quantities and concentrations of radioactive materials in particulate form released in gaseous effluents, samples should be collected'n pro-5portion to the rate of flow of the effluent stream.

Required when the gross activity in the secondary coolant system, as re~uired to be determined in Appendix A of these technical specifications, exceeds 10 pCi/ml.

6/1/76

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. TABLE 2,4>>$

I STg LUCIE RLhttT CASEOIIS MASTS SYSTEH fACAT10tt OF PROCESS AHD EPPLUEtIT HOHITORS hMD SAHPLERS R+UIRED ttY TECllttICAL SPECIFIChTIOttg Grab huto Control to Continuous Saaple Hoasureaent Process Screaa or Release Point hlara .Isolation Valve Honicor Station Hoblo Cas I ~ ParCfculaCo tl-3 . hlpha Masc4 Caa Storage Tanks X X X X X. X X Condenser hir E)actor X. X X .X X Building Ventilation Systeas Reactor ContafnaenC Buildfng fuhenevcr there is flou co Plant Vent) .. X -X X hurf1 fary Building (Co Plant Venc) X Fuel ltandling h SCorags Buildings X Radiance brea (Co Plant Vent) X X X - X Stean Generator Bloudoun Tank Vent or Condenser Ventb x(') x(d X X X X Turbine Gland Seal Condenser 4 X ~ X X 'X X Masts Evaporator Condenser Ventc '~

I 4

X X X X

'i If any or ell ot the process acreaas or building ventilatfon sysceas aro routed to a single roleasa point, tha need tor ~ continuous aonitor. at tha individual discharge point co che aafn exhaust duct is eliafnated. One continuous aonfcor at the final reloase point ia sufficient.

b In eoae PMRs the steaa generator bloudoun tank vent la routed Co Cho aafn turbine condenser, and the need for a continuous aonftor at this release point fe elfalnuted.

c Por PMRs fn uhfch che uasce evaporator condenser fs vented directly co tha acaoephere.

Honfcorfng syscea uill be installed by Harch I, l976, if chfs syatea is eclll opcratfonal.

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Honltored via Condcnsor hlr E)cctor Syscca.

2-18 Bases The release of radioactive materials in gaseous waste effluents to un-restricted areas shall not exceed the concentration limits specified in 10 CFR Part 20 and should be as low as reasonably achievable in accordance with the requirements of Appendix I to 10 CFR Part 50. 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-lives greater than eight days will not exceed 15 mrem per site.

1 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 with 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, Even with this operational flexibility under unusual operating conditions, if the licensee exerts every effort to keep levels of radioactive material in gaseous waste effluents as low as reasonably achievable, 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 experience taking into 'account a combination of system variables including defective fuel, primary system leakage, primary to secondary system leakage, steam generator blowdown and the performance of the various waste treatment systems.

Specification 2.4.3.a(l) limits the release rate of noble gases from the site so that the corresponding annual gamma 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 compliance with 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 2.4-3. The expressions used to calculate these dose factors are based on dose models derived in Section 7 of Meteorolo and Atomic Ener -1968 an'd model techniques provided in Draft Regulatory Guide 1.AA.

Dose calculations have been made to determine the site boundary location with the highest anticipated dose rate from noble gases using onsite 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.

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~ 4 2-19 The offsite location with the highest anticipated annual dose from released noble gases is 1600 meters in the North direction.

The release rate Specifications for a 'radioiodine 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 a'e: 1) individual inhalation of airborne radio-nuclides, 2) deposition 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 milk by man. Methods 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 onsite 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 the corresponding annual thyroid dose via the most restrictive pathway is less than 1500 mrem.

For zadioiodines and 'radioactive material in particulate form with half-lives greater than eight days, the most restrictive location is a residence located 3,000 meters in the WSW direction (vent X/Q 5.5x10 7 sec/m ).

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 half the design ob]ective annual quantity during any calendar quarter, or the design objective annual quantity during any period of 12 consecutive months. Xn 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-fourth 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.

Specification 2.4.3.e limits the maximum quantity of radioactive gas that can .

be contained in a waste gas .storage tank. The calculation of this quantity should assume instantaneous ground release, a X/Q based 5 percent meteorology, the average gross energy is 0.19 Mev per disintegration (considering Xe-133

. to be the principal emitter) and exposure occurring at the minimum site boundary radius using a semiinfinite cloud model. The calculated quantity will limit the offsite dose above background to 0.5 rem or less, consistent with Commission guidelines.

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2-20 1'

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 given under Specification 2.4.4 provide assurance that radioactive materials released in gaseous are properly controlled and monitored in conformance with the waste'ffluents requirements of Design Criteria 60 and 64. These requirements provide the

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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'thers,. the Commission may from time to time require the licensee,to take such action as the Commission

'deems. appropriate.

Specification 2.4.4.b excludes monitoring the turbine building ventilation ezhause since this release is expected to be a negligible release point.

Many PMR reactors do not have turbine building enclosures. To be consistent in this requirement for all PMR reactors, the monitoring .of gaseous releases from turbine buildings. is not required.

2:4.5 Solid Paste Handlin and Dis osal

a. Measurements shall be made to determine or estimate the total curie quant'ity and prin'ciple radionuclide composition of all radioactive solid waste shipped offsite.

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b. Reports of the radioactive solid waste shipments, volumes, principal radionuclides, and total curie quantity, shall be submitted in accord-

'ance with Section 5.6.1. I ~

Bases The requirements 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 offsite are packaged in con-formance with 10 CFR Part 20, 10 CFR Part 71, and 49 CFR Parts 170-178.

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