Text

Rev 7 to Ei Hatch Nuclear Plant ODCM, Reflecting Proposed Changes,Per GL 89-01
	ML20106A650
Person / Time
Site:	Hatch
Issue date:	02/07/1992
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20106A644	List: HL-2390, Application for Amends to Licenses DPR-57 & NPF-5, Incorporating Programmatic Controls for RETS Into Administrative Controls Section of TS & Relocating Procedural Details for RETS to ODCM & Pcp,Per GL 89-01; ML20106A646; ML20106A650; ML20106A652;
References
	GL-89-01, GL-89-1, PROC-920207, NUDOCS 9209290256
	Download: ML20106A650 (117)
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e, PROPOSED CHANGES TO THE E. I. HATCH NUCLEAR PLANT OFFSITE DOSE CALCULATION MANUAL .

9 IN ACCORDANCE WITH NRC GENERIC LETTER 89-01

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i E. I. HATCH NUCLEAR PLANT' <

OFFSITE DOSE CALCULATION MANUAL !

INSERTION INSTRUCTIONS -i REVISION 7= 2/7/92-EAq. g Instruction Acknowledgment = Receipt Sign / Return Insertion Instructions Use/ Discard Distribution List Replace Title Page- Replace-i through iv Replace v through vii Add i Introduction Replace ,

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Add behind p. 6.1-3 6.2-1 through 6.2-3 Add behind Tab 6.2-Tab 6.3 Add behind p. 6.2-3 O 6.3 Tab 6.4 Add behind Tab 6.3; Add behind p. 6.3-l'

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INSERTION INSTRUCTIONS I REVISION 7 2/7/92 P_iLqg Instruction 6.4-1 Add behind Tab 6.4 Tab 7.0 Add behind p. 6.4-1 7.0 cover sheet Add behind Tab 7.0 Tab 7.1 Add behind cover sheet 7.1.1 Add behind Tab 7.1 Tab 7.2 Add behind p. 7.1-1 7.2-1 through 7.2-6 Add behind Tab 7.2 References (second page) Replace O

Ka\wp\ hatch \odem\tnstestr-O

E.1. liATCli NUCLEAR PLANT Offs 11E DOSE CALCULATION MANUAL DISlR11LQTION LIS1 PAGJShat lille/ Location Eppy No.

E. C. Sorrell Supervisor, Document Control 1-8 Plant Hatch D. M. Hopper Project Supervisor, Rad / Chem 11 Southern Nuclear Operating Co.

Inverness Center Building 40, 6th Floor Birmingham, Alabama M. C. Nichols Supervisor, En'.ironmental Center 14 & 21 Central Lab 5131 Maner Road Room 5131 Smyrna, Georgia 300B0 7

J. A. Wehrenberg Manager, Nuclear Plant Support - Haten 16 Southern Company Services inverness Center O Building 42 Birmingham, Alabama S. Hempstead Supervisor, Technical Wri+'ag 18 Southern Company Services inverness Center _

Building 42, 2nd floor Birmingham, Alabama G. W. Robson Supervisor, Document Control 22 Inverness Center Building 40, 5th Floor Birmingham, Alabama O

HATCH ODCM, REV 7 2/7/92 n w m u c swcas u e a

O OFFSITE DOSE CALCULA110tl MA!1 VAL ,

FOR GEORGIA PCWER COMPANY EDWIN 1. IIATCH NUCLEAR PLANT O

FEBRUARY 1992 O

l

. TABLE OF C0!iTEijJJ O ,

1.0 L10V10 EFFlV(({Il 1.0-1 1.1 LIQUID EFFLUENT HONITOR SL1 POINTS 1.1-1 1.1.1 Liquid Raowaste Effluent Radiation Monitors ..!-l 1.1.2 Plant Service Water Monitors 1.1-9 1.2 DOSE CALCULATION FOR LIQUID EFFLUENTS 1.2-1 ,

1.3 DOSE PROJECTIONS FOR LIQUID EFFLUENTS 1.3-1 1.3.1 Monthly Dose Projections 1,3-1 1.3.2 Dose Projections for Specific Releases 1.3-2 1.4 DEFINITIONS OF LlQUID EFFLVENT TERMS 1.4-1 1.S LIQUID RADWASTE TREATMENT SYSTEMS 1.5-1 1.6 MIXING 0F LIQUID WASTE TANKS 1.6-1 1.7 LIMITS OF OPERATION 1.7-1 1.7.1 Liquid Effluent Monitoring Instrumentation Control 1.7-1 1.7.2 Liquid Effluent Concentration Control 1.7-2 1.' 3 Liquid Effluent Dose Control 1.7-4 1.7.4 Liquid Radwaste Treatment System Control 1.7-6 2.C BASE 0VS EFFLVENTS 2.0-1 2.1 GASEOUS EFFLUENT MONITOR SETPOINTS 2.1-1 2.1.1 Unit 1 Rer.ctor Building Vent Stack, 2.1-1 Unit 2 Reactor Building Vent Stack, Unit 1 Recombiner Building Vent, and Building Exhaust Augmented Ventilat:or 2.1.2 Main Stack 2.1-4 2.1.3 Determination of Allocation Factor, AG 2.1-6 2.1.4 Unit 1 Condenser Offgas Pretreatment Monitor and !

Unit 2 Condenser Offgas Pretreatment Monitor 2.1-7 -

O HNP ODCM, REV 7 2/7/92 i l

TABLE OF CONTENTS JCONTINVED)

EAan 2.2 GASEOUS EFFLUENT DOSE RATE AND DOSE CALCULATIONS 2.2-1 2.2.1 SITE 80VHDARY Dose Rates 2.2-1 l 2.2.1.a Dose Rates Due to Noble Gases 2.2-1 I 2.2.1.b Dose Rates Due to lodine-131, Iodine-133, 2.2-2 l l Tritium, and Particulaies !

l 2.2.2 Air Dose and Dose to a HEMBER OF THE PUBLIC 2.2-4 in Areas at and beyond the SITE BOUNDARY l l

2.2.2.a Air Dose in Areas at and beyond the 2.2-4 !

SITE BOUNDARY 2.2.2.b Dose to a MEMBER OF THE PUBLIC 2.2-5 in Areas at and beyond the SITE BOUNDARY 2.2.2.c Dose Calculations to Support 2.2-17 Other Specific Technical Specifications 2.3 METEOLOLOGICAL MODEL 2.3-1 2.3.1 Atmospheric Dispersion 2.3-1 2.3.1.a Ground-Level Releases 2.3-1 2.3.1.b Elevated Releases 2,3-3 2.3.2 Relative Deposition 2.3-6 t

1 2.4 DEFINITIONS OF GASE0US EFFLUENT PARAMETERS 2.4-1 2.5 GASEOUS RADWASTE TREATMENT SYSTEM 2.5-1 lO l

HNP ODCM, REV 7 2/7/92 11 l L a

l TABLE OF CONTENTS (CONTI M DI h9e 2.6 LIMITS OF OPERATION 2.6-1 2.6.1 Gaseous Effluent Monitoring instrumentation Control 2.6-1 2.6.2 Gaseous Effluent Dose Rate Control 2.6-2 2.6.3 Gaseous Effluent Air Dose Control 2.6-4 2.6.4 Gaseous Effluent Dose to a MEMBER OF THE PUBLIC 2.6-6 Control 2.6.5 GASEOUS RADWASTE TREATMENT SYSTEM Control 2.6-9 3.0 RAD 10LOGlCA1_ ENVIRONMENTAL MONITORILG 3.0-1 3.1 LIMITS Of OPERATION 3.1-1 3.1.1 Radiological Environmental Monitoring Program 3.1-1 3.1.2 Land Use Census 3.1-3 3.1.3 Interlaboratory Comparison Program 3.1-4 4.0 TOTAL DOSE DETERMINATIONS 4.0-1 4.1 LIMITS OF OPERATION 4.1-1 4.1.1 Total Dose Control 4.1-1 S.0 POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR 5.0-1 ACTIVITIES INSIDE THE SITE BOUNDARY 6.0 REPORTS 6.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT 6,1,1 6.1.1 Report Contents 6.1-1 6.2 SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT b.2-1 6.2.1 Report Contents 6.2-1 O

HNP ODCM, REV 7 2/7/92 -111 l

TABLE OF CONTENTS (CONTINUED)

O P1911 6.3 MONTl!LY Ol'ERATING REPORT 6.3-1 6.4 SPECIAL REPORTS 6. ( l 7.0 GENERAL DEFINIT 101G 7.1 TERM TRANSFERRED TO THE ODCM FROM TECHNICAL SPECIFICATIONS 7.1-1 7.1.1 GASE0US RA0 WASTE TREATMENT SYSTEM 7.1-1 7.2 TERMS DEFINED IN TECilNICAL SPECIFICATIONS 7.2-1 7.2.1 ACTION 7.2-1 7.2.2 CHANNEL CAllBRATION 7.2-1 7.".3 CHANNEL CHECK 7.2-2 7.2.4 CHANNEL FUNCTIONAL TEST 7.2-2 7.2.5 00SE EQUIVALENT IODINE 7.2-2 7.2.6 FREQUENCY NOTATION ' . 2-3 7.2.7 MEMBER (S) 0F THE PUBLIC 7.2-3 7.2.8 HILK ANIMAL 7.2-3 7.2.9 OPERATIONAL CONDITION 7.2-4 7.2.10 OPERABLE - OPERABILITY 7.2-5 7.2.11 AEACTOR MODE 7.2-5 7.2.12 RATED THERMAL POWER 7.2-5 7.2.13 SITE B0UNDARY 7.2-6 7.2-14 SOURCE CHECK 7.2-6 7.2-15 THERMAL POWER 7.2-6 7.2-16 UNRESTRICTED AREA 7.2-6 O

HNP ODCM, REV 7 -2/7/92 iv- l I

LIST Of TABLES O

Table Iltle ELqe 1.2-1 Bioaccumulation factors 1.2-4 1.2-2 Adult Ingestion Dose factors 1.2-5 1.2-3 Site-Related Ingestion Dose Commitment Factors, 1.2-7 Ag (Fish Consumption) '

l.7-1 Radioactive Liquid Effluent Monitoring Instrumentation 1.7-8 ,

1.7-2 Radioactive Liquid Effluent Monitoring Instrumentation 1.7-10 .

Surveillance Requirements 1.7-3 Radioactive Liquid Effluent Sampling and Analysis Program 1.7-12 2.1-1 Dose factors for Exposure to a Semi-Infinite 2.1-3 Cloud of Noble Gases 2.1-2 Dose Factors for Exposure to Direct Radiation 2.1-9 from Noble Gases in the Elevated Finite Plume 2.2-1 inhalation Dose Factors for Infant 2.2-19 2.2-2 Inhalation Dose factors for Child 2.2-22 i 2.2-3 inhalation Dose Factors for Teenager 2.2-25 O- 2.2-4 Inhalation. Dose Factors for Adult 2.2-28 2.2 Ingestion Dose Factors For Infant 2.2-31 2.2-6 Ingsstion Dose Factors for Child 2.2-34 2.2-7 Ingestion Dose Factors for Teenager 2.2-37 2.2-8 Ingestion Dose Factors for Adult 2.2-40 .

2.2-9 L.'ternal Dose Factors for Standing on Contaminated Ground 2.2-43 2.2-10 Iniividt.41 bsage f actors 2.2-45 2.2-11 St.ble Element Transfer Data 2.2-46 2.2-12 Cestrolling Receptor 2.2-47 2.2-13 S,te-Specific (or Default)_ Values to be Used 2.2-48 in Pathway Factor Calculations 2.6-1 Radioactive Gaseous-Effluent Monitoring Instrutaentation 2.6-11 ,

2.6-2 Radioactive Gaseous Effluent Monitoring Instrumentation 2.6-15 Surveillance Requirements 2.6-3 Radioactive Gaseous Waste Sampling and Analysis Program 2.6-18 O .

HNP.0DCM, REV 7- 2/7/92 y l'

t LIST OF TABLES (CONTINUED) .

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Table Iltle hgg 3.0-1 Radiological Environmental Sampling Locations 3.0-3 3.1-1 Radiological Environmental Monitoring Program 3.1-6 3.1-2 Reporting levels for Radioactivity Concentrations 3.1-10 in Environmental Samples 3.1-3 Lower Limit of Detection 3.1-11 .

5.0-1 Location-Specific Parameters for the Roadside Park 5.0-3 -

5.0-2 Location-Specific Parameters for the Camping Area 5.0-4 5.0-3 Location-Specific Parameters for the Recreation Area- 5.0-5 5.0-4 Location-Specific Parameters for the Visitors Center 5.0-6 6.1-1 Environmental Radiological Monitoring Program Summary 6.1-3 l r

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O HNP ODCH, REV 7 2/7/92 vi l

LIST OF FIGURES O Ligur_e 11tle fian 1.1-1 Example Calibration Curve for Liquid Effluent Monitor 1.1-10 1.5-1 Liquid Radwaste Treatment System (Unit 1) 1.5-2 1.5-2 Liquid Radwaste Treatment System (Unit 2) 1.5-3 1.7-1 UNRESTRICTED AREA Boundary 1.7-14l 2.3-1 Vertical Standard Deviation of Material in a Plume (a,) 2.3-7 2.3-2 Open Terrain Recirculation Factor 2.3-8 2.3-3 Plume Depletion Effect for Ground-Level Releases 2.3-9 2.3-4 Plume Depletion Effect for Greater Than 100-m Releases 2.3-10 2.3-5 Relative Deposition for Grourd-Level Releases ?.3-11 2.3-6 Relative Deposition for Greater Than 100-m Releases 2.3 12 2.5-1 Condenser Offgas Treatment System 2.5-2 3.0-1 Radiological Environmental Sampling location Map 3.0-5 (Site Periphery) 3.0-2 Radiologcial Environmental Sampling Location Map 3.0-6 (Beyond Site Vicinity) 3.0-3 Location of Additional Control Station for TLDs 3.0-9 and Vegetation l

l l-O HNP 0DCH, REV 7 2/7/92 vii l

JNTROW[lL0ff O

The Offsite Dose Calculation Manual (0DCM) is a supporting document of the technical specifications which address radiological effluents and radiological environmental monitoring. As such, the ODCM describes the methodology and parameters to be used in the calculation of offsite doses due to radioactive liquid and gaseous effluents and in the calculation of liquid and gasecus ef fluent monitoring instrumentation alarm / trip setpoints. The ODCM contains a list and graphical description of the specific sampi locations for the radiological anut onmental monitoring program. Schematic configurations of the liquid and gaseous radwaste tuatment systems are also included. l The ODCM will be maintained at the plant for use as a reference guide and training document of accepted methodologies and calculations. Changes in the calculational methods or parameters will be incorporated into the ODCM in order to assure that the ODCM represents the present methodology in all applicable areas. Computer software to perform the described calculations will be maintained current with the ODCH.

As required by the revised definition of the ODCM presented in NRC Generic Letter 89-01, this manual also contains the Radioactive Effluent Controls

~

required by Technical Specification 6.18 and the Radiological Envircnmental Monitoring Program required by Technical Specification 6.19, along with the methods and parameters for conducting this program. This document also contains descriptions of information that should be included in the Annual Radiological Environmental Operating Report required by Technical Specification 6.9.1.6 and descriptions of information that should be included in the Semiannual Radioactive Effluent Release Report required by Technical Specification 6.9.1.8.

O HATCH ODCM, REV 7 2/7/92 l !

where:

C ,c - ti.e effluent concentration limit (Section 1.7.2) implementing l 10 CFR 20 for the site, corresponding to the specific mix of radionuclides in the waste tank being considered for discharge, in Ci/ml.

c = the setpoint, in Ci/ml, of the radioactivity monitor that measures the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint, which is inversely proportional to the volumetric flow of the effluent line and proportional to the volumetric flow of the dilution strecm plus the effluent stream, represents a value that, if exceeded, would result in concentrations exceeding the limits of 10 CFR 20 in the UNRESTRICTED AREA. l f - the flow setpoint as determined at the radiation monitor location in volume per unit time, but in the same units as F, below.

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

As stated earlier, at Plant Hatch, each of the two units is served by its own independent liquid radwaste system; the two liquid radwaste systems discharge to separate dilution streams. If additional dilution flow is needed for either dilution stream, it is available from the plant service water system.

The two dilution streams release to the Altamaha River.

The sources of liquid radioactive effluents from Unit I are:

Waste sample tank A.

Waste sample tank B.

Chemical waste sample tank A.

Chemical waste sample tank B.

HATCH ODCH, REV 7 2/7/92 1.1-2 l

112R.1 O The radionuclide concentration for a waste tank to be released is obtained from the sum of measured concentrations as determined by the analyses required in Table 1.7-3: l fC3 = h C, + ( C, + C, + Cf + C,) (2) where:

C, - the concentrr. tion of each measured gama emitter observed by gamma-ray spectroscopy of the particular waste sample.

C, - the concentration of alpha emitters in liquid waste as measured in the MONTHLY composite sample. (NOTE: Sample is analyzed for gross a.)

C, - the measured concentrations of Sr-89 and Sr-90 in liquid waste as observed in the QUARTERLY composite sample.

C, - the measured concentrations of Fe-55 in liquid waste 2s observed in the QUARTERLY composite sample.

C.

- the measured concentration of H-3 in liquid waste as determined from analysis of the MONTHLY composite sample.

The C, term will be included in the analysis of each batch; terms for alpha, strontiums, iron, and tritium will be included in accordance with Table 1.7-3, as appropriate. l Sten 2 The measured radionuclide concentrations are used to calculate a dilution factor, DF, which is the ratio of total dilution flow rate to tank flow rate HATCH ODCM, REV 7 2/7/92 1.1-4 I l

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1.2 DOSE CALCULATION FOR LIQUID EFFLUE1TS For liquid release from Plant Hatch to the Altamaha River, the only human exposure pathway is due to consumption of fish taken from the river. No known drinking water is taken from the Altamaha River downstream from Plant Hatch. The methodology for calculating doses to a MEMBER OF Tile PUBLIC due to l

fish consumption is presented in this section.

The dose contribution to the maximum exposed MEMBER OF THE PUBLIC by way of l fish consumption from radionuclides identified in liquid effluent > released to UNRESTRICTED AREAS will be calculated for the purpose of implementing Section 1.7.3. In accordance with Appendix A of Reference 3, noble gases ari "ded from these dose calculations. Doses to a MEMBER OF THE PUBLIC are calc.s ..ed as follows:

-A U D, = f Aj, t=1b .6 c,l c3, c F, e (9) where:

D7 - the <umulative dose commitment to the total body or any organ, :, due to radioactivity in liquid effluents for the total time period b A t,, in mrem, (Reference 1).

f=1 Arg - the length of the fth time period over which Cgi and F, are averaged for all liquid releases, in hours.

C ig - the average concentration of radionuclide i, in undiluted liquid effluent during timo period Arg from any liquid release, in pCi/ml.

Ai = the decay. constant for radionuclide 1 (sec).

l l HATCH ODCM, REV 7 2/7/92 1.2-1 l

1.3 DOSE PROJECTIONS F0D. LIQUID EFFLUENTS r

O 1.3.1 Monthly Dose Pro.iectioni ,

In order to meet the requirements of Section 1.7.4, which pertains to l operation of the liquid radwaste trer.tment systems, dose projections must be made at least monthly during periods in which discharge of untreated liquid ;

effluents containing radioactive materials to UNRESTRICTED AREAS occurs er is j ixpected.

Projected 31-day doses to a MEMBER OF THE PUBLIC due to liquid effluents may l be determined as follows:

t '*'

Deg,,p = x 31 D ,,, p = x 31 O

where:

0 ,3 the cumulative total-body dose for the elapsed portion of the current quarter plus the release under consideration.

t - the number of days into the current quarter.

D,93 - the cumulative organ doses for the elapsed p rtion of the 4 current quarter plus the release under consideration.

, If activities planned during the remainder of the quarter are expected to contribute a significant dose and the determination can be reasonably made, this contribution _should be included in the equations: .

s.

'"*)

D y,,p = x 31 + D,x O

t

,( t j HATCH'00CM,-REV.7 2/7/92' l.3-1 l

D og ,,9 = '*' x 31 + D,3 where:

0,, - the expected dose dun to the particular planned activity.

1.3.2 Dose Proiections for: Soecific fieleases Dose projections may be performed for a particular release by performing a prerelease dose calculation, assuming that the planned release will proceed as anticipated. For individual dose projections due to liquid releases, follow the methodology presented in Section ).2, using sample analysis values 1

for the source to be released and parametric values expected to exist for the release period.

O O

HATCH ODCM, REV 7 2/7/92 1.3-2 l

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Section of Ista Definitiqn initial Use Cgi

- the average concentration of radionuclide i, in 1.2.1 undiluted liquid effluent during time period At g , in C1/ml.

CtAPc = the effluent concentration limit (Section 1.7.2) 1.1.1 l implementing 10 CFR 20 for the site in Ci/ml.

C3 - the concentration of Sr-89 or Sr-90 in liquid 1.1.1 wastes as determined by analysis of the QUARTERLY composite sample.

Ct = the measured concentration of H-3 in liquid waste 1.1.1 as determined by analysis of thL MONTHLY composite.

Dr - the cumulative dose commitment to the total body 1.2.1 or any organ, r, from the liquid effluents for O the total time period.

DF = the dilution factor, which is the ratio of the 1.1.1 total dilution flow rate to the effluent stream flow rate (s) required to assure that the limiting concentration of 10 CFR, Part 20, Appendix B, Table 11, Column 2 are met at the point of discharge to the unrestricted area.

DF,7 - a dose conversion factor for nuclide i, for adults 1.2.1 in preselected organ, y, in mrem /pC1, from Table 1.2-?.

f - the flow setpoint as determined for the rdiation 1.1.1 monitor location. (General expression for Equation 1.)

HATCH ODCH, REV 7 2/7/92 1,4-2 l

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l.7 LlHITS Of OPERATION O 1.7.1 U_qu'd Effluent Monitorina Instrumentation Control in accordance with Technical Specification 6.18(1), the radioactive liquid effluent monitoring instrumsntation channels shown in TeMe 1.7-1 shail be OPERABLE with their alarm / trip setpoints set to ensure that the limits specified in Section 1.7.2 are not exceeded. The alarm / trip setpoints of these channels shali be determined in accordance with Section 1.1.

1.7.1.1 Applicability As shown in Table 1.7-1.

1.7.1.2 Action 1.7.1.2.1 With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above control, without delay suspend the release of radioactive liquid effluents i

monitored by the affected channel, declare tlie channel inoperable, or' change the setpoint to a conservative value.

1.7.1.2.2 With the number of channels OPEpABLE less than the minimum channels required by Table 1.7-1, take the ACTION shown in Table 1.7-1.

1.7.1.2.3 For Unit 1: When the ACTION statement or other requirements of this control cannot be met, steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified condition may be made if, as a minimum, the requirenents of the ACTION statement are satisfied.

For Unit 2: The provisions of Technical Specifications 3.0.3 and 3,0.4 are not applicable.

O c

HATCH ODCM, REY 7 2/7/92 1.7-1 l

l.7.1.3 Surveillance Requirements Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CPANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION, and CHANNEL FUNCTION" TEST operations at the frequencies shown in Table 1.7-2.

1.7.1.4 Bases The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoint> for these instruments shall be calculated in accordance with the methods in Section 1.1 to ensure that the alarm / trip will occur prier to exceeding the limits of 10 CFR Part 20. The OPERABillTY and use of this instrumentation are consistent with the requirements of General Design Ciiteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

{ l.7.2 Liouid Efflu?nt Concanir311qn Control in accordance with Technical Specificctions 6.18(2) and (3), the concentration of radioactive material released at any time from the site to UNRESTRICTED AREAS Figure 1.7-1) shall be limited to the ct,;.centrations specified in 1

10 CFR Part 20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentrations shall be limited to 2E-4 Ci/ml total activity.

1.7.2.1 Applicability At all times, 1.7.2.2 Action 1.7.2.2.1 With the concentration of radioactiv:' material released from the site to UNRESTRICTED AREAS exceeding the abo"9 limits, without delay restore concentration within the above limits and provide notification to the Nuclear HATCH ODCM, REV 7 2/7/92 1.7-2 l l

J l____________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Regulatory Commission by including a discussion of the causes and corrective actions taken in the next semiannual radioactive etfluent release report, in accordance with Section 6.2.1.

t 1.7.2.2.2 for Unit 1: When the ACTIGN statement or other requirements of this control cannot be met, stepr. need not be taken to change the Operational Mode of the unit. Entry into an Operational Hode or other specified condition may be made if, as a minimum, the requirements of the ACTION statement are satisfied.

For Unit 2: The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable. >

1.7.2.3 Surveillance ReTairements 1.7.2.3.1 Radioactive liquid effluents shi..l be sampled and analyzed according to the samplir.g and an61ysis program of Table 1.7-3.

1.7.2.3.2 The results of radioactise analyse

shall be used in accordance with the method; in Section 1.1 to assure that the concentrations at the point of release are maintained within the limits of Section 1.7.2.

1.7.2.4 Bases This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to UNRESTRICTED AREAS will be 1955 than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. This limitation provides additional assuranc.e that the levele of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within the Section II.A design objectives of Appendix 1, 10 CFR Part 50, to a HEMBER OF THE FUBLIC, and within the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radionuclide, and its Maximum Permissible Concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

HATCH ODCH, REV 7 2/7/92 1.7-3 l

i 1.7.3 Liouid Effluent Dose Control i In accordance with Technical Specifications 6.18(4) and (5), the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid -

' i efflueats released, from each reactor unit, to UNRESTRICTED AREAS (Figure i 1.7-1) shall be limited to:

a. During any calendar quarter to less than or qual to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ,

b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ. !

1.7.3.1 Applicability ,

At all times.

1.7.3.2 Action 1.7.3.2.1 With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the limits specified above in Section 1.7.3, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies ,

the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents to ensure that subsequent releases will be in compliance with the limits specified above in Section 1.7.3. For Unit 2, this report.is in lieu of any-other report required by Technical Specification 6.9.1. (This report shall also include (1) the results of radiological 1 analyses of the drinking water source and (2) the radiological impact on finished drinking water supplies ,

with regard te the requirements of 40 CFR 141, Safe Drinking Water Act.) ,

l 1.7.3.2.2 -For Unit 1: When the ACTION statement or other requirements of this control cannot be met, steps need not be taken to change.the.0perational Mode of the unit. Entry into an Operational Mode or other specified condition u

HATCH ODCH,,REV 7 2/7/92 1.7-4 l l

- ~-

i may be made if, as a minimum, the requirements of the ACTION statement are satisfied.

For Unit 2: The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

1.7.3.3 Surveillance Requirements Dose Calculations: Cumulative dose contributions from liquid effluents shall be determined at least monthly in accordance with Section 1.2.

1.7.3.4 Bases This control is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The control implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time iraplement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive clerial in liquid effluents will be ke9t "as low as is reasonably achievable" (ALARA). The dose calculations in Section 1.2 implement the requirements in Section III.A of Appendix I, which state that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a HEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in Section 1.2 for calculating the doses due to the actual release rates of radioactive materials in liquid effluents will be ccc.sistent with the methodology provided in Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.113.

This control applies to the release of liquid effluents from each reactor at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared systems are proportioned-among the units sharing that system.

HATCH ODCH, REV 7 2/7/92 1.7-5 l

%w.-, ,---.,,,- . , _..g- py.,,,-~g ,,r, , _ . , , , , , g ,.,,, - y.. ____, , _ , , _ _

1.7.4 Liquid Radwaste Treatment System Control O in accordance with Technical Specification 6.18(6), the liquid radwaste treatment system, as described in Section 1.5, shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluents, tram each unit, to UNRESTHICTED AREAS (Figure 1.7-1) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ of a MEMBER OF THE PUBLIC in 31 days.

1.7.4.1 Applicability At all times.

1.7.4.2 Action 1.7.4.2.1 With radioactive liquid waste being discharged without treatment and in excess of the above limits, within 30 days, prepare and submit to the Nuclear Regulatory Commission, pursuant to Technical Specification 6.9.2, a

/ Special Report that includes the following information:

a. Identification of the inoperable equipment or subsystems and the l reason for inoperability,

b. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

c. Summary description of action (s) taken to prevent a recurrence.

1.7.4.2.2 for Unit 1:. When the ACTION statement or other requirements of this control cannot be met, steps need not be taken to change'the Operational Mode of the unit. Entry into an Operational Mode or other specified condition may be made if, as a minimum, the requirements of the ACTION statement-are satisfied.

For Unit 2: The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

HATCH ODCM, REV 7 2/7/92 1.7-6 l i

l 1.7.4.3 Surveillance Roquirements l O' Doses due to liquid releases shall be projected at least monthly in accordance l

I with Section 1.3, during periods in which discharge of entreated liquid effluent containing radioactive materials to UNRCSTRICTED AREAS occurs or is j expected to occur. ;

1.7.4.4 Bases The OPERABilliY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment ,

prior to release to UNRESTRICTFD AREAS. The requirements that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept ALARA.

This control implements the requi ements of 10 CFR Part 50.36(a), General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were '

specified as a suitable fraction of the guide set forth in Section II.A of Appendix 1,10 CFR'Part 50, for liquid effluents.

k 5

O HATCH ODCM, REV 7 2/7/92 1.7-7 l

TABLE 1.7-1 (SHEET 1 0F 2) ,

O RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION

Minimum Channels Instrument QPERABLE Aenlicability ACTION

1. Gross Radioactivity Monitors Providing Automatic Termination of Release liquid Radwaste Effluent Line 1 (a) 100

2. Gross Radioactivity Monitors not Providing Automatic Termination of Release Service Water System E# fluent Line 1 (b) 101

3. Flowrate Measurement Devices" Liquid Radwaste Effluent Line 1 (a) 102 Discharge Canal 1 (a) (b) 102 i 4. Service Water System to Closed l Cooling Water System Differential Pressure 1 At all times 103 l

" Applies to each unit.

" Pump curves may be utilized to estimate flow; in such cases, ACTION statement 102 is not required.

(a) Whtnever the radwaste discharge valves are not locked closed.

(b) V'tiever the service water system pressure is below the closed cooling water system pressure or AP indication is not available.

O HATCH 00CM, REV 7 2/7/92 1.7-8 l

TABLE 1.7-1 (SHEET 2 0F 2)

RADI0 ACTIVE LIQU10 EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATIONS ACTION 100 - With the number of channels OPERABLE less than required by the llinimum Channels OPERABLE requirement, effluent releases may be continued, provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Section 1.7.2.3.1.

b. At least two technically qualified individuals independently verify the release rate calculations and discharge valving.

Otherwise, suspend release of radioactive effluents via this pathway. If the channel remains inoperable for over 30 days, an explanation of the circumstances shall be included in the next semiannual radioactive effluent release report.

ACTION 101 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided that once per shift grab samples are collected and analyzed for grcss radioactivity (beta or gamma) at a lower limit of Detection of at least IE-7 Ci/ml. If th channel remains inoperable for over 30 days, an explanation of the circumstances shall be included in the next semiannual s radioactive effluent release report.

ACTION 102 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual rebases. If the channel remains inoperable for over 30 days, an explanation of the circumstances shall be included in the next semiannual radioactive effluent release report.

ACTION 103 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, assure that the service water system effluent monitor is OPERABLE.

1 HATCH ODCH, REV 7 2/7/92 1.7-9 l

l TABLE 1.7-2 (SHEET 1 0F 2)

O RADIDACTIVE LIQUID EFFLUENT M0!11TORING INSTRUMEllTAT10fl SURVEILLANCE REQUIREMEliTS, CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL lttstrument QJtLCX M CX CALIBRATION IISI

1. Gross Gamma Radioactivity Monitors Providing Alarm and Automatic isolation -_

Liquid Radwaste Eif,1uent Line 0" P(3) R Q(1)

2. Gress Gamma Radioactivity Monitors Providing Alarm but not Providing Automatic Isolation Service Water Sy', tem Effluent Line D" M R Q(4)

3. Flowrate Measurement Devices liquid Radwaste Effluent Line D(2)" NA R Q Discharge Canal D(2)" NA R Q

4. Service Water System to Cicsed Cooling Water System Differential Pressure D NA R NA O

HATCH ODCM, REV 7 2/7/92 1.7-10 l

TABLE 1.7-2(SHEET 20F2)

O- RADIOACTIVE LIQUID EFFLUENT MONITORING 1

j INSTRUMENTATION SURVEllLANCE PrQUIREMENTS*

TABLE NOTA 710NS

Applies to each unit.

- 0uring releases via this pathw y.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation !

of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the alarndtrip setpoint. !

b. Instrument indicates an isolation on high ularm. ;

c. Instrument controls are not set in operate mode.

(2) CHANNEL CHECK shall cons'st of verifying indication of flow during periods l of release. CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch eleases are made. !

(3) The SOURCE CHECK prior to release shall consist of verifying that the )

instrument is reading onscale.

(4) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the alarm setpoint,

b. Instrument indicates a downscale failure. .

c. Instrument controls are not set in operate mode.

I I

l l

i lO HATCH ODCM, REV 7 2/7/92 1.7-11 l

h f

r TABLE 1.7-3 (SHEET 1 0F 2) :

P RADI0 ACTIVE LIQUID EFFLUENT SAMPLING AND ANALYSIS PROGRAM

l 8

Liquid Minimum Type of Lower Limit of i Release Sampling Analysis Activity Detection' d

- Tvoe freauencv freauency Analysis ,_(yCi/ml)

, Batch Waste P P Principal b '

Release Tanks Each Batch Each Batch Gamma SE-7 Emitters' 1-131 lE-6 P M Dissolved and One Batch /M Entrained Gases lE-5 .;

P M H-3 lE-5 .

Each Batch Composite * ---------------------------- l Gross-Alpha IE-7 P Q Sr-89, Sr-90 SE-8 Each Batch Composite * ------------------------------

fe-55 2E-6 .

. . _ _ _ m _ en es . 48 4tp Ele . . . . . GP SG - e em . . em . w . . . . . es . . ep e. _ go es es . . _ m em . . . . W m . _ e9 - . W W e @ We.% -G...._

?

i i

+

l e

i HATCH ODCM, REV 7. 2/7/92 1.7-12 l

TABLE 1.7-3 (SHEET 2 0F 2) t Y RADI0 ACTIVE LIQUID EFFLUENT SAMPLING AND ANAL.YSIS PROGRAM

TABLE NOTAT',0NS

Applies to each unit.

a. The Lower Limit of Detection is definel in Table Notation (a) of Table 3.2-3.

b. For certsin radionuclides with low gamma yield or low energies or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the Lower Limit of Detection. Under these circumstances, the Lower Limit of Detection may be increased inversely proportional to the magnitude of the gamma yielr' .e., SE-7/1, where: I - photon abundance expres!,ed as a decimal fract but in no case shall the lower Limit of Detection, as calculated i., .s manner for a specific radionuclide, be greater than 10 percent of the naximum Permissible Concentration value specified in 10 CFR 20, Appendix B, Table II, Column 2.

c. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and ln which the method of sampling employed results in a specimen that is representative of the liquids released.

O V d. A batch release is the discharge of liquid wastes of a discrete volume.

Prior to samplf ag for analysis, each batch shall be isolated and then thoroughly mixed, as described in Section 1.6, to assure representative sampling.

e. The principal gamma emitters for which the Lower Limit of Detection

requirements will apply are exclusively the following radionuclides

l Mn-54, Fe-59, Cc-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144. This list does not mean that only these nuclides are to be

.ietected and reported. Other measursble and identifiable peaks, together with the above nuc191es, shall also be identified and reported.

]

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2.0 GASEOUS EFFLUENTS At Plant Hatch there are four points where radioactivity normally is released to the atmosphere in gaseous discharges. These four release points are:

The main stack which serves both units.

Unit I reactor building vent stack.

Unit 2 reactor building vent stack.

Unit I recombiner building vent, in addition to these four release points, releases also may be made from building exhaust augmented ventilation system (s) provided sitch system (s) has (have) been included in Tables ?.6-1, 2.6-2, and 2.6-3.

The main stack serves as the discharge point for the following release sources from each unit:

Mechanical vacuum pumps.

Off-gas system.

Gland seal exhaust.

Standby gas treatment system through which drywell purges are discharged.

The waste gas treatment building ventilation also discharges through the main stack.

Each reactor building vent stack serves as the discharge point for the following (of each respectise unit):

,

Reactor building.

Refueling floor ventilation.

Turbine building.

Radwaste building.

The Unit I recombiner building vent discharges directly to the atmosphere.

HATCH ODCM, REV 7 2/7/92 2.0-1 l

Gaseous effluent monitor setpoints are required only for noble gas muitors n serving the release points; the methodology for cale'llating noble gas monitor V setpoints is presented in Section 2.1. Although setpoint calculations are not required for radioiodine and particulate monitors, the methodology for assuring that the potential organ dose rates due to Iodine-131, Iodine-133, tritium, and particulates with half-lives greater than 8 days in gaseous releases frc: the site to areas at and beyond the SITE B0UNDARY do not exceed the limits of Section 2.6.2 is presented in the note following Section 2.2.1.b.

O O

HATCH 00CM, REV 7 2/7/92 2.0-2 l

2.1 GASE0US EFFLUENT MONITOR SETPOINTS O The gaseous monitor setpoint values determined in the following sections will be regarded as upper bounds for the actual setpoint adjustments. That is, setpoint adjustments are not required to be perfc.rm-i if the existing setpoint level corresponds to a lower count rate than the calculated value.

Setpoints may be established at values lower than the calculated values, if desired.

If no release is planned for a particular pathway, or if there is no detectable activity in the planned release, the monitor setpoint should be established as close to background as practical to prevent spurious alarms and yet alarm sheuld an inadvertent releasa occur.

If a calculated setpoint is less than the monitor reading associated with the particular release pathway, no release may be made under current conditions.

Under such circumstances, the number of simultaneous release pathways may be ceduced or contributing source terms may be reduced and the setpoint recalculated.

2.1.1 !Lnit 1 Reactor Building Vent Stack. Unit 2 Reactor Buildino Vent Stack.

Unit 1 Recombiner Buildina Vent. and Buildina Exhaust Auamented Venti-lation Monitors: Dll-K619 A and B, 2011-K636 A and B, Dll-P003 A and B (Monitor identification for building exhaust augmented ventilation will be determined prior to making releases via this pathway.)

For the purpose of implementing Section 2.6.1, the alarm setpoint level for j these noble gas monitors t ill be calculated as follows:

C, - the monitor reading of the noble gas monitor at the alarm setpoint concentration.

O HATCH 00CM, REV 7 2/7/92 2.1-1 l

(AG x SF) x R., x Dr. (1)

C, - the lesser of or (AG x SF) x R,, x 0,, (2)

SF - the safety factor; a conservative factor applied to each noble gas monitor to compensate for statistical fluctuations and errors of measuremant. (For example, SF - 0.5 corresponds to a 100-percent variation.)

AG - an administrative allocation factor applied to apportion the release setpoints among all gaseous release discharge pathways (normally four) to assure that release limits will not be exceeded by simultaneous releases. The tillocation factor for a particular discharge pathway may be assigned any desired value between 0 and I under the condition that the sum of the allocation factors for all simultaneous release pathways does not exceed 1. For ease of implementation, AG nay be set equal to 1/n, where: n - the number of simultaneous final gaseous release points. For a more exact determination of a'ilocation factors, see Section 2.1.3.

D,, - the dose rate limit to the total body of a MEMBER OF THE PUBLIC, which is 500 mrem /yr.

R, - the monitor reading per mrem /yr to the tota' ody for vent releases.

C + ((x/g),Eg K, Q,,) . (S)

O HATCH ODCM, REV 7 2/7/92 2.1-2

where:

O V C - the mon'itor reading of a noble gas monitor corresponding to the grab sample radionuclide concentrations taken in accordance with Table 2.6-3. The monitor response l corresponding to the measured concentration is determined from the monitor calibration curve for the particular monitor.

(X/Q), - the highest annual average relative concentration at and beyond the SITE BOUNDARY. (If decired, the annual average relative concentration at and beyond the SITE B0UNDARY for the particular release point may be used.) The release points addressed in this section are ground-level releases.

(X/Q), -

8.37 x 10-* sec/m' in the ENE sector.

K, - the total-body dose factor due to gamma emissions from O radionuclide 1 (mrem /yr per pCi/m') from Table 2.1-1.

V Q,, - the rate of release of noble gas radionuclide 1 ( Ci/sec) from the vent release pathway under consideration, which is the product of X,, and F , where: X,, - the concentration of radionuclide i for the particular release, and F, - the maximum expected release flow rate for this release point (X,, in Ci/ml and F, in ml/sec).

D,, - the dose rate limit to the skin of the body of a MEMBER OF THE PUBLIC in areas at and beyond the SITE B0UNDARY, which is 3000 mrem / year.

R,, - the monitor reading per mrem /yr to the skin.

[

t HATCH ODCM, REV 7 2/7/92 2.1-3 l

n'* . - - e e >. -

t V

(3 R,, -

C + ((X/Q),j E (L, + 1.1 M ) Q,,) (4)

.( ,

where:

L, - the skin dose factor due to beta emissions from radio-auclide i (mrem /yr per pCi/m') from Table 2.1-1, 1.1 - the mrem skin _ dose per mrad air dose.

M, - the air dose factor due to gamma emissions from radio-nuclide 1 (mrad /yr per pCi/m') from Table 2.1-1.

2.1.2 Main Start Monitor: Dll-K600 A and B A

'D) For the purpose of implementing Section 2.6.1, the alarm setpoint level for -l the main stack noble _ gas monitor will be calculated as follows:

C, - the monitor reading of the noble gas monitor at the alarm i setpoint concentration.

l.

l l ( AG x SF) x R., x - 0,, (5) l the lesser of or (AG x SF) x R,, x D., (6) l R,, - C+ V, Q,, (7)_

O HATCH ODCM, REV 7 2/7/92 2.1-4 l

-s

1 R. , - C+ (L, (X/Q), + 1.1 B,)Q,, '(8)

O l

where:

V, - the constant, which includes the dose factor, for each identified noble ga; radionuclide accounting for the gamm6 radiation from the elevated finite plume resulting from the main stack release, in mrem /yr per Ci/sec, from Table 2.1-2.

l B, - the constant, which includes the air dose factor, for each identified noble gas radionuclide accounting for the gama radiation from the elevated finite plume j

.1 resulting from the main stack release, in mrad /yr per pCi/sec, from Table 2.1-2.

Q,, - the rate of release of noble gas radionuclide 1 (pCi/sec) from the main stack, which is equal to the

( product of X,, and F., where: X,, - the concentration of radionuclide i for the uain stack release, and F, = the maximum expected main stack release flow rate (X,, in pC1/ml and F, in ml/sec).

i (X/Q), - the highest annual average relative concentration in areas at and beyond the SITE BOUNDARY associated with l releases from the main stack. The main stack is an elevated release.

the 4.10 x 10-* sec/m' in the ENE sector.

l All other terms were identifled previously in Section 2.1.1.

O l- HATCH ODCM, REV 7 2/7/92 2.1-5 l

2.1.3 Q3 termination of Allocation Factor. AG Q

V dhen simultaneous gaseous releases are made to the environment, an (administrative) allocation factor must be applied to each discharge pathway. This is to ensure that simultaneous gaseous releases from the site to areas at and beyond the SITE BOUNDARY will not exceed the dose rate limits specified in Section 2.6.2. For Plant Hatch, final discharge pathways w'n ich may be released simultaneously are:

The main stack.

Unit I reactor building vent stack.

Unit 2 reactor building vent stack.

Unit I recombiner building vent.

Building exhaust augmented ventilation.

The allocation factor for each discharge pathway must be between 0 and 1, and the sum of the allocation factors for the simultaneous releases must not exceed 1.

There are three methods by which allocation factors may be determined:

1 1. The allocation factor for a particular release pathway may be administratively selected based on an estimate of the fraction of the total dose rate (from all simultaneous releases) which is contributed by the particular release pathway.

2. The allocation factor may be calculated using the expression:

AG =

1/n where: n - the number of release pathways to be released simultaneously.

3. The allocatien factor may be determined for a particular discharge pathway by calculating the ratio of the total-body dose rate due to noble gases released from the particular discharge pathway under consideration to the total-body dose rate due to noble gases in all simultaneous releases, v

HATCH ODCM, REV 7 2/7/92 2.1-6 l

i for the main stack:

E I ' '

AG = - (9)

E V,Q,, + E(X/Q),E X,Q..

i n i where: n - the number of simultaneous releases.

For vent releases:

(X[Q), E K, Q,, g I

AG = (10) i E V, Q,, + E (X/Q), E K, Q,,

i n i k where: n - the number of simultaneous vent releases.

(r) - the particular discharge pathway number for which an allocation factor is being determined.

p 2.1.4 Unit 1 Condenser Offoas Pretreatment Monitor and Unit 2 Condenser d Offaas Pretreatment Monitor Monitors: 1Dll-K601, 2Dll-K601, IDll-K602, and 2011-K602 C

For the purpose of implementing Section 2.6.1, the alarm setpoint level for these noble gas monitors will be calculated as follows:

C, - 240,000 / (CF,

F,,,) (11) where: 240,000 - the release rate limit for pretreatment condenser offgas specified in Technical Specifications 3.15.2.7 (Unit 1) and 3.11.2.7 (Unit 2) ( Ci/sec).

C. = the conitor reading of the condenser offgas pretreatment monitor at the alarm setpoint (mR/hr).

C F, - the calibration factor for the condenser affgas pretreatment monitor ((pCi/sec)/((mR/hr)*(cfm))).

F,,, = the condenser offgas flow rate (cfm).

HA1CH ODCM, REV 7 2/7/92 2.1-7 l

2.2 GASEOUS EFFLUENT DOSE RATE AND DOSE CALCULATIONS 2.2.1 SITE B0UNDARY Dose Rates l 2.2.1.a Dose Rates Due To Noble Gases for the purpose of implen,enting Section 2.6.2.a, the dose rate in areas at and beyond the SITE B0UNDARY due to noble gases shall be calculated as follows:

D.

- the total body dose rate at time of release (mrem /yr).

. I(M).IKi Q.

V i . IV 0 V

i 4 (9)

D, - the skin dose rate at time of release (mrem /yr).

I (T/Q), I (L. + 1.1 H ) Q i , . I (L,(X/Q), + 1.1 B,) Q'.

i (10)

V I _ _ .

Terms were defined previously in Sections 2.1.1 and 2.1.2.

The dose rate limits are site limits at any point in time; therefore, dose rates are summed over all releases occurring simultaneously. For Plant Haten the vent releases are: _

Unit I reactor building vent stack.

Unit 2 reactor building vent stack.

Unit I recombiner building vent.

Building exhaust augmented ventilation.

The only elevated release is the main stack which serves both units.

Simultaneous releases may include any cambination of these four release points.

O HATCH ODCM, REV 7 2/7/92 2.2-1 g

l I

2.2.1.b Dose Rates Due to Iodine-131, Iodine-133, Tritium, and Particulates l j

( l For-the purpose of implementing Section 2.6.2.b, organ dose rates due to !

Iodine-131, Iodine-133,. tritium, and all radioactive materials in particulate l form with half-lives greater than 8 days are required to be calculated for the I 1

inhalation pathway for the child age group. The child age group would .j experience the highest potential dose rate via the inhalation pathway. In- !

accordance with Appendix C to Reference 3, noble gases are excluded from these I calculations. These dose rates are calculated as follows:

i D. - the organ dose rate at time of release (mrem /yr). l

. I (@). I P,, Q,' , . (X/Q), I P,, Q,',

33) v i . .

i .

where:

(X/Q), - defined in Section 2.1.1.

(X/Q), - defined in Section 2.1.2.

Q,', - the release rate ( Ci/ cec) of Iodine-131, Iodine-133, tritium, and particulates with half-lives greater than 8 days (required by Section 2.6.2.3.2) from the Unit I reactor building vent stack, the Unit 2 reactor building vent stack, the Unit I recombiner building vent, and building exhaust-augmanted ventilation.

Q,', - the release rate (pCi/sec) of Iodine-131, Iodine-133, tritium,_ and particulates with half-lives greater than 8 days (required by Section 2.6.2.3.2) from the main stack.

P, , = the organ dose parameter for organ o and radionuclide 1 (mrem /yr per Ci/m') for inhalation determined as follows:

HATCH ODCM, REV 7 2/7/92 2.2-2 l

P,4

= K (BR) DF,, (12) and where:

(

K = the constant of unit conversion, 10' pCi/ Ci . .

BR = the breathing rate for child age group (3700 m*/ year), Table 2.2-10, from Reference 3.

DF,, - the inhalation pathway dose factor for child age group for organ o and radionuclide 1, Table 2.2-2, from Reference 3.

NOTE: In order to assure that potential dose rates (prerelease) to an ,

organ due to Iodine-131. Iodine-133,. tritium, and particulates with half-lives greater than 8 days in simultaneous gaseous releases from the site to areas at and beyond the SITE BOUNDARY O do not exceed 1500 mrem /yr as specified in Section 2.6.2,b, the potential organ dose rate D, must be limited as follows:

l-D, +-(AG x SF) s 1500 mrein/yr (13) where: AG and SF are assigned the same values as were used in Section 2.1 for tia release source pathway under consideration.

To further ensure that dose rate limits were not exceeded, (post-release) dose rates from simultaneous releases shall be summed, as shown above.

~

4 O

HATCH ODCH, REV 7 2/7/92 2.2-3 l

2.2.2 Air Dose and Dose to a MEMBER OF THE PUBLIC in Areas et and Beyond the 111E BOUNDARY 2.2.2.a Air Dose in Areas at and Beyond the SITE B0UNDARY l For the purpose cf implementing Section 2.6.3, the air dose in areas at and beyond the SITE BOUNDARY shall be determined as follows:

D, - the air dose due to gamma emissions from noble gas radionuclides (mrad).

- 3.17 x 10* ( Ill ) I "' 0"J + i "

V

)O ..

where:

3.17 x 10 - the fraction of 1 yr per 1 sec, i),, - the cumulative release of noble gas radionuclide i over the period of interest ( Ci) from the vent release under consideration.

i),, - the cumulative release of noble gas radionuclide i -

over the period of interest (pCi) from the main stack.

M, - defined previously in Section 2.1.1.

B, - defined previously in Section 2.1.2.

1 (X/Q), - defined previously in Section 2.1.1.

2.2-4 !

HAiCH ODCM REV 7 2/7/92 1

D, - the air dose due to beta emissions from noble gas radionuclides (mrad).

. 3.17 x 10 4 I (U). I N, D,[ fS), I N, h,,Y V i . .

I _ _

where:

N, = the air dose factor due to beta einissicns from noble gas radionuclide i (mrad /yr per 141/m') from Table 2.1 l.

(X/Q), - defined previously in Section 2.1.2.

2.2.2.b Dose to a MEMBER OF THE PUBLIC in Areas at and beyond the SITE BOUNDARY Dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and radioactive .Taterial in particulate form with half-lives greater than 8 days will be calculated for the purpose of implementing Section 2.6.4. In accordance with Appendix C of Reference 3, noble cases are excluded from these dose calculations. Doses to a MEMBER OF THE PUBLIC are calculated as follows: l NOTE: For Plant datch, the controlling receptor f:,r which doses mt... be caiculated, the dispersion and deposition values at the location of the

~

controlling receptor, and the applicable exposure pathways are presented in Table 2.2-12.

D, - dose to an organ j of a MEMBER OF THE PUBLIC in age group a from Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days (mrem). ,

I R,,,, f,,@,, + W,,@,,]

. 3.17 x 10 4 (16) p$

HATCH 00CM, REV 7 2/7/92 2.2-5 l s

1 where:

3.17 x 10 - the fraction of 1 yr per 1 sec.

W ', - the pathway-t.vendent relative dispersion or deposition in the areas at and beyond the SITE BOUNDARY at the location of the controlling receptor, associated with plant vent releases.

(X/Q'),, - the annual average relative dispersion parameter for location of controlling (critical) receptor for plant vent releases. ;

(T/Q'),, applies only to inhalation and '

all tritium pathways. (for all 6 'J out.

pathways, the (' source term is 1. "ted to tritium.) See Table 2.2-12 for value. f' W ,', -

(D7Q' ),, - the annual average deposition parameter for the location of control 11ag (critical) receptor for plant vent releases. (67Q'),,

applies to ground-plane and all ingestion pathways, with the exception of tritium. i See Table 2.2-12 for value.

)

O HATCH ODG, REV 7 2/7/92 2.2-6

l W,', - the pathway-dependent relative dispersion i or deposition in the area at and beyond the L SITE BOUi10ARY at the lor,ation of the controlling receptor, associated with main stack releases.

(T/Q ' ),,, - the annual average relative dispersion parameter for location of controlling (critical) receptor for main stack rel ea ta.c . (T[Q'),, applies only to inhalation and all tritium pathways. (For all tritium pathways, the (' source term is limited to tritium.) See Table 2.2-12 for value.

Wl, -

(UTQ'),, - the annual average deposition parameter for the location of controlling (critical) receptor for main stack releases. (UTQ'),,

applies to ground-plane and all ingestion pathways, with the exception of tritium.

See Table 2.2-12 for value.

The selectiUn of the dispersicn or deposition parameter, X/Q or D/Q, is dependent upon the pathway being considered. The dispersion parameter, X/Q, is required for the inhalation pathway. The deposition parameter, D/Q, is required for the ground-plane pathway and all ingestion pathways, with the exception of tritium. Since tritiua is taken up by vegetation directly from surrounding air, X/Q is required for tritium contributions from the ingestion pathways.

O HATCH ODCM, REV 7 2/7/92 2.2-7 g

Ql, - the cumulative release ( C1), from plant vent _ releases, of radionuclide i as required by Section 2.6.4.3 over the period of interest. Dose determinations required by Section 2.6.4.3 i are on a per reactor basis; therefore, cumulative release quantities must also be reactor-specific. (For dose contribution due to tritium from the ingestion pathways, the 6,',termislimitedtotritium.)

Ql, - the cumulative release ( Ci), from the main stack releases, of radionuclide i as required by Section 2.6.4.3 over the l period of interest. Dose determinations required by Section 2.6.4.3 are on a per reactor basis; therefore, l l

cumulative release quantities must also be reactor-specific.

Since the main stack serves both reactors, release quantities l

must be apportioned between the two units. In absence of i evidence that one reactor contributes a greater quantity of !

radioactivity than the other over the period of interest, Os release quantities may be apportioned equally between the two units. (Fur dose contributions due to tritium from ingestion pathways, the Q,', term is limited to tritium.) ;

R , ,, = the pathway-specific, individual age-specific organ dose factor for radionuclide-i, pathway p, organ j, and individual age group a. Routine dose caiculations for a MEMBER OF THE PUBLIC address the inhalation, ground-plane, l grass-cow (or goat)-milk, grass-cow-meat . and garden vegetation pathways. However, the dose pathways actually present at the controlling location, as well as tha controlling age group for a MEMBER OF THE PUBLIC are determined through the Land Use Survey for the site.

Pathway factors R ,,, are determined as. shown in the following subsections.

HATCH ODCM, REV 7 2/7/92 2.2-8 l.

R. , ,, - K'KF,Q,U,,(DFL,,), (0.75(0.5/H)] (mrem /yr per Ci/m') (26) i V where:

K'" = a constant of unit conversion,10' gm/kg.

H = absolute humidity of the atmosphere, in gm/m'.

0.75 = the fraction of_ total feed that is water.

0.5 = the ratio of the specific activity of the feed grass water to the atmospheric water.

Other parameters and values are given above.

2.2.2.c Dose Calculations To Support Other Specific Technical Specifications In the event radiological impact assessment becomes necessary to implement RETS 6.9.1.12 or 6.9.1.13, dose calculations will be performed using the

( equations in Section 2.2.2.b, with the substitution of average meteorological >

t- parameters for the period of the report and the appropriate pathway receptor dose f actors (R,,,,) .

For the purpose of implement Section 3.1.2, dose calculations may be performed l using the equations in Section 2.2.2.b, substituting the appropriate pathway l

receptor dose factor (R ,,3) and the appropriate dispersion parameters for the

location (s) of interest. Annual average diversion parameters may be used for these calculations.

1 l The receptor for which dose calculations may be required in order to support Technical Specifications 6.6, 6.9.1, 6.9.2, or Section 3.1.2 may not be the l previously identified critical receptor. The receptor age group and exposure L pathways present (and applicable) at the location of interest must be determined. The equations for calculating the pathway factors R.,,, were presented in Section 2.2.2.b. Plant Hatch site-specific values, or appropriate default values, required in the pathway factor determinations are

, presented u Table 2.2-13.

HATCH ODCM, FW 7 2/7/92 2.2-17 _l

TABLE 2.2-11 (n) STABLE ELEMENT TRANSFER DATA

F, - Milk F, - Milk Element (Cow) (Goat) E, - Meal H 1.0E-02 1.7E-01 1.2E-02 C 1.2E-02 1.0E-01 3.lE-02 Na 4.0E-02 4.0E-02 3.0E-02 P 2.5E-02 2.5E-01 4.6E-02 Cr 2.2E-03 2.2E-03 2.4E-03 Mn 2.5E-04 2.5E-04 8.0E-04 Fe 1.2E-03 1.3E-04 4.0E-02 E Co 1.0E-03 1.0E-03 1.3E-02 Ni 0.7E-03 6.7F-03 5.3E-02 Cu 1.4E-02 1.3E-02 8.0E-03 Zn 3.9E-02 3.9E-02 3.0E-02 Rb 3.0E-02 3.0E-02 3.lE-02 '

Sr 8.0E-04 1.4E-02 6.0E-04 Y 1.0E-05 1.0E-05 4.6E-03 Zr 5.0E-06 5.0E-06 3.4E-02 .

Nb 2.5E-03 2.5E-03 2.8E-01 Mo 7.5E-03 7.5E-03 8.0E-03 Tc 2.5E-02 2.5E-02 4.0E-01 p Ru 1.0E-06 1.0E-05 4.0E-01 Q Rh Ag 1.0E-02 5.0E-02 1.0E-02 5.0E-02 1.5E-03 1.7E-02 Te 1.0E 03 1.0E-03 7.7E-02 I 6.0E-03 6.0E-02 2.9E-03 Cs 1.2E-02 3.0E-01 4.0E-03 Ba 4.0E-04 4.0E-04 3.2E-03 _

La 5.CE-06 5.0E-06 2.0E-04 Cu 1.0E-04 1.0E-04 1.2E-03 Pr 5.0E-06 5.0E-06 4.7E-03 Nd 5.0E-06 5.0E-06 3.3E-03 W 5.0E-04 5.0E-04 1.3E-03 Np 5.0E-06 5.0E-06 2.0E-04 F, units: days / liter F, units: days /kg

Reference 3, Table E-1, 1

HATCH ODCH, REV 7 2/7/92 2.2-46 l

c ._. -_ _

E.3.1.b Elevated Releases O

b - the sector-averaged annual average relative concentratio1 at X/Q any distance in the given sector for radionuclides other than noble gase; 2.032 ( I 6* ""

=

Jk NU,,6,,

where:

6, ,

the plume depletion factor taken from figure 2.3-4. For an elevated release, this factor is stability dependent.

h - effective release height, in meters a h, - h .

NOTE: Effective release height may be further aojusiad for plume rise in accordance with section E,4.3.2 of Appendix E of Reference 5.

h, - the height of the main stack (120 m),

h, - terrain height at location of interest, in meters (obtained

! from Figure 2.3-12 of Reference 6). i n - the number of hours the wind of wind speed class j is directed 1 l 1012 the given sector during the time atmospheric stibility l category k existed. These values may be obtained from Table 2 l -_ of Reference 23.

i b3 - the wind speed (mid-point of wind speed class j) at the height of release h (m sec") during atmospheric stability k.

N - the total hours of valid meteorological data recorded for all sectors, wind speed classes, atid stability categories frc'a Table E.4-7 of Reference 5.

l The t'.naining symbols are the same as. those previously defined.

I When considering the direct gamma radiation from an elevated' finite plume, the

- constants B, and V, defined in Section 2.1.2 for each identified noble gas radionuclide are calculated using the following:

5 III "3 ^" #*

M" 8 -

~J vjk2 N U, pC1/ sec HATCH ODCM, REV 7 2/7/92 2.3-3 l

P Section of IEm Definition Initial Use V

D, - the dose to an organ of a MEMBER OF THE PUBLIC 2.2.2.b from Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with ;

half-lives greater than 8 days (mrem).

D, - the organ dose rate at time of release (mrem /yr). 2.2.1.b D, - the skin dose rate at time of release (mrem /yr). 2.2.1.a D,, - the 1tmiting dose rate to the skin of the body of 2.1.1 a MEMBER OF THE PUBLIC in areas at and beyond the SITE B0UNDARY, which is 3000 mrem /yr.

D, - the total body dose rate at time of release 2.2.1.a (mrem /yr).

0,, - the limiting dose rate to the total body of a 2.1.1 MEMBER OF THE PUBLIC in areas at and beyond the SITE BOUNDARY, which is 500 mrem /yr.

D, - the air dose due to beta emissions from noble 2.2.2.a gases (mrad).

D, - the air dose due to gamma emissions from noble 2.2.2.a gases (mrad).

D/Q - the sector-averaged relative deposition for kny 2.3.2 distance in a given sector, t

(D/Q'),, - the annual average deposition parameter for.the 2.2.2.b locstion of controlling (critical) receptor for f

plant vent releases.

O j HATCH ODCM, REV 7 2/7/92 2.4-2 l ,

l Section of Igre Definition Initial Use Q,, - the rate of release of noble gas radionuclide i 2.1.1 l

( Ci/sec) from the vent release pathway under ;

consideration.

Q,, - the rate of release of noble gas radionuclide i 2.1.2 l

( Ci/sec) from the main stack.

Q,, - the cumulative release of noble gas radionuclide i 2.2.2.a over the period of interest (pCi) from the vent release under consideration.

Q,, - the cumulative release of noble gas radionuclide i- 2.2.2.a over the period of interest ( Ci) from the main stack.

Q ,', - the cumulative release of Iodine-131, Iodine-133, 2.2.2.b tritium, and radionuclides in particulate form with half-lives greater than 8 days from plant vent

\

releases over the period of interest ( Ci).

Q ,', - the cumulative release of Iodine-131, Iodine-133, 2.2.2.b tritium, and radionuclides in particulate form with half-lives greater than 8 days from the main stack over the period of interest (pCi).

r - the distance from the point of release to the 2.3.1 receptor of interest for dispersion calculations (meters).

R. ,,3 - - the pathway-specific, individual age-specific 2.2.2.b organ dose factor for radionticlide i, pathway p, organ j, and age group a (mrem /yr per Ci/m') or (m'-mrem /yr per pCi/sec).

l l

L s l

HATCH ODCM, REV 7 2/7/92 2.4-5 l l

l'

Section of Term Definition Initial Utg O - the total absorbtion coefficient for air (m). 2.3.1

- the energy absorbtion coefficient for air (m'*). 2.3.1 p, - the tissue energy absorbtion coefficient for 2.3.1 photons of energy Eg (cm' gm) .

u, - the wind speed (midpoint of windspeed class j) 2.3.1 at the height of release, h.

u3 - the wind speed (midpoint of windspeed class j) 2.3.1 at ground level (m/sec) during atmospheric stability class k.

Up - the wind speed (midpoint of windspeed clas j) 2.3.1 at the height of release, h, of an elevated release during atmospheric sta'ility class k.

V, - a constant, which includes the dose factor, for 2.1.2 each identified noble gas radionuclide accounting _

for the gamma radiation from the elevated finite

. plume resulting from the main stack release (mrem /yr per Ci/sec) from Table 2.1-2.

W'. , - the pathway-dependent relative dispersion or 2.2.2.b deposition in areas at and beyond the SITE BOUNDARY at the location of the controlling receptor associated with plant vent releases.

O HATCH ODCH, REV 7 2/7/92 2.4-7 l

. . . . . -. ~ . ._

l Section of. l Igr_r3 j)efinition Initial Use -l J

]

- W,', - the pathway-dependent relative dispersion or 2.2.2.b deposition in areas at and beyond the SITE B0VNDARY at the location of the controlling receptor associated with stack releases.

1 i

X/Q = the sector-averaged annual average relative 2.3.1 concentration at any distance in the given l sector.

XTQ. x the highest annual average relative concentra- 2.1.1 tion in areas at and beyond the SITE B0UNDARY l associated with ground-level releases.

l i

i iTQ, - the highest annual average relative concentra- 2.1.1 l tion in areas at and beyond the SITE BOUNDARY l l associated with releases from the main stack.

l 1.

(X/Q'),, = 'the annual average relative dispersion parameter 2.2.2.b for the location of the controlling receptor for plant vent releases.

1 (X/Q'),, - the annual average relative dispersion parameter 2.2.2.b for the location of the controlling receptor for main stack releases.

l O

HATCH ODCM, REV 7 2/7/92 2.4-8 l l

l

Section of Igra Definition Initial Usg f ,q b

z - the vertical-distance from a ground-level receptor 2.3.1 to the volume element considered as a point souce in the evaluation of I ,,3t.

g C, - the monitor reading of the condenser offgas 2.1.4 pretreatment monitor at the alarm setpoint (mR/hr).

C F, - the calibration factor for the condenser offgas 2.1.4 pretreatment monitor (( Ci/sec)/((mR/hr)*(cfm))).

F,,, - the condenser offgas flow rate (cfm), 2.1.4 O

l ks\=t\betch\odem\edem2 4 U- .

HATCH'00CM, REV 7 2/7/92 2.4-9 l l

2.5 CASE 0VS-RADWASTE TREATMENT SYSTEM O Figure 2.5-1 is a schematic of the condenser offgas treatment system showing the release paints to areas at and beyond the SITE BOUNDARY. -This schematic l is representative of Unit I and Unit 2.

O l

l l

l

(

J HATCH ODCM, REV 7 2/7/92 2.5-1 l v

s -. -. ..- - - _--, - - - ---~_--.- -- - _ . _ - - - - - - ,- - - _ _- - - - _ - _ _ - . _ _ - - - - - - - - - - - _ _ - -

F 2.6 1.lHITS Of OPERATION i

O 2.6.1 Gaseous Effluent Monitorino Instrumentation Control In accordance with Technical Specification 6.18(1), the radioactive gaseous effluent monitoring instrumentation channels shown in Table 2.6-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Section 2.6.2 are not aceeded. The alarm / trip setpoints of these channels shall be determined in accordance with Section 2.1.

2.6.1.4 Applicability As shown in Table 2.6-1.

2.6.1.2 Action 2.6.1.2.1 With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value that will ensure that the limits of Section 2.6.2 are met, without delay restore the setpoint w a value that will ensure that the-limits of section 2.6.2 are met or declare the channel inoperable.

2.6.1.2.2 With the number of channels OPERABLE less than the minimum ck.nnels required by Table 2.6-1 take the ACTION shown in '.?ble 2.6-1, 2.6.1 2.3 for Unit 1: When the ACTION statement or other requirements of this control cannot be met, steps need not be taken to change the Operational Hodi of the unit. Entry into an Operational Mode or other specified condition may be made if, as a minimum, the requirements of the ACTION statement are satisfied.

For Unit 2: The provisicas of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

O HATCH 00CM, REV 7 2/7/92- 2.6-1 l-a

~

2.6.1.3. Surveillance Requirements O Each rtdioactive gaseous of fluent monitoring instrumentation channel shall be uemonstrated OPERABLE by performaace of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and J.HANNEL FUNCTIONAL TEST optrations at the .':equencies shown in Table 2.6-2.

2.6.1.4 Bases The radioactive gaseous effluent instrumentation i t provided , aonitor and control, as applicable, the releases of radioactive materials in gaseous offluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the methods in Section ?-i to ensura that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILI1Y and use of this instrumentation are consistent with the requirements of Gencral Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

2.6.2 Qtseous Effluent Dose Rate Contro)

In accordance with Technical Specifications 6.18(3) and (7), the dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE B0UNDARY (Figure 1.7-1) shall be limited to the following:

a. The dose rate limit for noble gases shall be less than or equal to 500 mrem / year to the total body and less than or equal to 3000 mrem / year to the skin.

b. The dose rate limit for 1-131, 1-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days shall be less than or equal to 1500 mrem / year to any, organ.

O HATCH ODCH, REV 7 2/7/92 2.6-2 ]

1 2.6.2.1 Applicability O At all times.

2.6.2.2 Action With the dose rate (s) exceed'ng the above limits, without delay . crease the .

re h a;e rate to comply with the limit (s) stated in Section 2.6.2.

2.6.2.3 Surveillance Requirements 2.6.2.3.1 The dose rata due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with methode, and procedures described in Sections 2.1 and 2.2.1.a.

2.6.2.3.2 The dose rate due to radioactive materials other than noble gases in gaseous effluents shall be deter:.ined to be within limits stated in Section 2.6.2 in accordance with the methods and procedures described in Section 2.2.1.b by obtaining representative samples and performing analyres in accordance with the sampling and analysis program specified in Table 2.6-3.

2.6.2.4 Bases This control is provided to ensure that at all times the dose rate at and beyond the SITE B0UNDARY from gaseous effluents from all onsite units will be within the annual dose limits of 10 CFR Part 20 for UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table 11. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER C.i THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual avera s e .;ncentrations exceeding the limits specified in Appenu.s. B, Table II of 10 CFR Part 20 0 .

HATCH ODCM, REV 7 2/7/92 2.6-3 l-

I (10 CFR Part 20.106(b)). For HEMBERS OF THE PUBLIC who may at times be within f the SITE B0UNDARY, the xwpancy of that HEMBER OF THE PUBLIC will usually be i sufficiently low to compensate for any increase in the atmospheric diffusion j factor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the carresponding gamma and beta dose rates above ,

background to a HEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem / year to the total body or to less than or equal to 3000 mrem / year to the skin. These release rate limits also restrict, at all ,

times, the corresponding thyroid dose rate above background to a child via the !

inhalation pathway to less than or equal to 1500 mrem / year.

This control applies to the release of gaseous effluents from all reactors at j the site. For units with shared radwaste treatment systems, the gaseous

~

s effluents from the shared system are proportioned among the units sharing that I system.

}

7 2.6.3 Qaseous Effluent Air Dose Control In accordance with Technical Specifications 6.18(5) and (8), the air dose due to noble gases eleased in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (figure 1.7-1) shall be limited to the following:

a. During any calendar quarter, to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation.

b. During any calendar year, to less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation. >

2.6.3.1 Applicability !

At all times, O

HATCH ODCM, REV 7 2/7/92- 2.6-4 l .,

_ _ -__._...__.__..__.__.____._____..__w

i 2.6.3.2 Action O 2.6.3.2.1 With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report identifying the cause(s) for exceeding the limit (s) and defining the corrective actions taken to reducs the releases and ptoposed corrective actions to be taken to assure that subsequent releases will be in compliance with Section 2.6.3.

2.6.3.2.2 for Unit 1: When the ACTION statement or other requiren,ents of this control cannot be met, steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified condition may be made if, as a minimum, the requirements of the ACTION statement are satisfied.

For Unit 2: The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

2.6.3.3 Surveillance Requirements s

Dose Calculations: Cumulative air dose contributions for the current calendar -

quarter and current calendar year for noble gases shall be determined at ' east

~

monthly in accordance with Section 2.2.2.a.

2.6.3.4 Bases 1 This control is provided to implement the iequirements of Sections II.B.

Ill.A, and IV.A of Appendix 1, 10 CFR Part 50. The control implements the guides set forth in Sectin 11.B of Appendix 1; the ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix 1, assuring that the releases of O

HATCH ODCM, REV 7 2/7/92 2.6-5 1

l radioactive material in gaseous effluents will be kept ALARA. The Surveillance Requirements implement the requirements in Section III.A of Appendix 1, which state that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of a HEMBER OF THE PUBLIC through appropriate pathways is ,

unlikely to be substantially underestimated. The dose calculations established in Section 2.2.2.a for calculating the doses due to the actual releases of radioactive noble gates in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977; and Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, Revision 1, July 1977. The equations in Section 2.2.2.a provided for determining the air doses at and beyond the SITE BOUNDARY will be based upon historical annual average atmospheric conditions. NUREG-0133 provides methods for dose calculations consistent with Regulatory Guides 1.109 and 1.111.

O 2.6.4 Gaseous Effluent Dose to a MEMBER OF THE PUBLIC Control In accordance with Technical Specifications 6.18(v) and (ix), the dose to any organ of a MEMBER OF THE PUBLIC from I-131, I-133, tritium, and all 1 radionuclides in particulate form with half-lives gmter thu 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE ,

! BOUNDARY (Figure 1.7-1) shall be limited to the following:

a. During any calendar quarter to less than or equal to 7.5 mrem to any organ,

b. During any calendar year to less than or equal to 15 mrem to any organ.

[

HATCH ODCM, REV 7 2/7/92 2.6-6 l .

c . _ _

2.6.4.1 Applicability At all times.

2.6.4.2 Action 1

2.6.4.2.1 With the calculated dose from the release of 1-131, 1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the fluclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Repcrt identifying the cause(s) for exceeding the limits and defining the corrective actions taken to reduce releases and proposed corrective actions to be taken to assure that subsequent releases will be in compliance with Section 2.6.4.

2.6.4.2.2 for Unit 1: When the ACTI0f1 statement or other requirements of this control cannot be met, steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified condition ma be made if, as a minimum, the requirecents of the ACT10t1 statement are satisfied.

i'or Unit 2: The provisions of Technical Specifications 3.0.3 and 3.0.4 are _

not applicable.

2.6.4.3 Surveillance Requirements Cumulative organ dose contributions to a MEMBER OF THE PUBLIC from 1-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater

~

than 8 days in gasecus effluents released to areas at and beyond the SITE BOUtl0ARY, from each unit, for the current calendar quarter and the current calendar year shall be determined at least monthly in accordance with Section 2.2.2.b.

O HATCH ODCM, REV 7 2/7/92 2.6-7 -l

l 2.6.4.4 Bases O This control is provided to implement the requirements of Sections 11.C, Ill.A, and IV.A of Appendix 1, 10 CFR Part 50. The statements in Section 2.6.4.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix 1, assuring that the releases of radioactive materials in gaseous effluents will be kept ALARA.

The calculational methods specified in Section 2.6.4.3 implement the requirements of Section Ill. A of Appendix 1, which state that conformance with the guides of Appendix ! be shown by calculational procedures based c, models _

and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to t'e substantially underestimated. The calculational methods approved by the llRC for calculating the doses due to the actual releases of the subject materials are required to be consistent with Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Reinses of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1, Revision 1, October 1977; and Regulatery Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine t' leases from Light-Water-Cooled Reactors, Revision 1 July 1977. These equations also provide for determining the actual doses based upon historical annual average atmospheric conditions. The release requirements for the radionuclides stated in the control are dependent upon ~

the existing radionuclide pathways to man H areas at and beyond the Si1E BOUNDARY. The pathsays examined in the levelopment of these calculations are:

a. Individual inhalation of airborne radionuclides,

b. Deposition of radionuclides onto green, leafy vegetation with subsequent consumption by man.

c. Deposition onto grassy areas where MILK AfilMALS and meat-producing animals graze with consumption of the milk and meat by man.

d. Deposition on the ground with subsequent exposure of man.

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HATCH ODCM, REV 7 2/7/92 2.6-8 l

2.6.5 GASEOUS RADWASTE TREATMENT SYSTEM Control In accordance with Technical Specification 6.18(6), the GASE0US RADWASTE TREATMENT SYSTEM as described in Section 2.5 shall be in operation.

2.6.5.1 Applicability Whenever the main condenser air eje '.or system is in operation.

2.6.5.2 Action 2.6.5.2.1 With the GASE0US RADWASTE TREATi4ENT SYSTEM inoperable for more than 7 days, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which includes the following information:

a. Identification of the inoperable equipment or subsystems and the reason for inoperabil Ry.

O b. Action (s) taken to restore the inoperable equipment to OPERABLE status.

c. Summary description of action (s) taken to prevent a recurrence.

2.6.5.2.2 for Unit 1: When the ACTION statement or other requirements of this control cannot be met, steps need not be taken to change the Operational Mode of the unit. Entry into an Operational Mode or other specified condition may be made if, as a minimum, the requirements of the ACTION statement are satisfied.

For Unit 2: The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

O HATCH ODCM, REV 7 2/7/92 2.6-9 l

- y -.- --

m,

2.6.5.3 Surveillance Requirements O GASE0US RADWASTE TREATMENT SYSTEM operability shall be demonstrated by administrative controls which assure that the offgas treatment system is not >

bypassed, v 2.6.5.4 Bases j The OPERABILITY of the GASE0US RADWASTE TREATMENT SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides reasonable assurance that-the releases of radicactive materials in gaseous effluents will be kept 3 ALARA. This control implements the requirements of 10 CFR Part 50.36(a),

General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section ll D of Appendix 1 to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the system were specified as a ;

suitable fraction of the guides set forth in Sections II.B and ll.C of :

Appendix I,10 CFR Part 50, for gaseous effluents. .

t O

HATCH ODCM, REV 7 2/7/92 2.6-10 l

- . - - - - . . - ..-. -.. - . ... - -- - - . - . . .-. - - . = - -

TABLE 2.6-1 (SHEET 1 0F 4)

RADI0 ACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION Minimum Channels Applic-Instrument OPERA 53LE ability P_arameter Action

1. Reactor Building Vent Stack Monitoring System (Each Unit)

a. Noble Gas Activity Radioactivity Rate Monitor 1 Measurement 105

b. Iodine Sampler Verify presence of I Cartridge 1

Cartridge 107

c. Particulate Sampler Verify presence of Filter 1 Filter 107

d. Effluent System Flowrate System flowrate :

l Measurement Device 1 Measurement 104

e. Sampler Flowrate Sampler Flowrate Measurement Device 1 Measitrement 104

2. Recombiner Building Ven*,ilation Monitoring System l

a. Noble Gas Activity Radioactivity Rate Monitor 1 Measurement 105 l

b. Iodine Sampler Verify Presence of Cartridge 1 of Cartridge 107 >

c. Particulate Sampler Verify Presence of Filter 1 F: - Jcr 107-

d. Sampler flowrate Sampler Flowrate Measurement Device 1 Measurement 104 l

!O ,

- HATCH ODCM, REV 7 2/7/92- 2.6-11 _

l ,

e- -

.gw weere*w-ye *

. - . . . . -.- . . . - - _ . . . . . . . - . . . = . . - . . _ - _ _ _ -

TABLE 2.6-1 (SHEET 2 0F 4)

O RADICACTIVE GASEOUS EFFLUENT MONITORING INSTRUPENTATION Minimum Channels Applic-Instrunni OPERABLE ability Parameter Action

3. Main Stack Monitoring System

a. Noble Gas Activity Radioactivity Rate Monitor 1 Measurement' 105

b. Iodine Sampler Verify Presence of Cartridge 1 Cartridge 107 l c. Particulate Sampier Verify Presence of Filter

1 Filter 107

d. Effluent System Flowrate System Flowrate Measurement Device 1

Measurement 104

e. Sampler Flowrate Sampler Flowrate O Measurement Device 1 Measurement 104

4. Condenser Offgas Pretreatment lionitor (Each Unit)

Noble Gas A:tivity Radioactivity Rate Monitor 1

- Measurement 108 O

HATCH ODCM, REV 7 2/7/92 2.6-12 l 3

T TABLE 2.6-1 (SHEET 3 0F 4)

O RADI0 ACTIVE GASEOUS EFFLUENT MONITORING IhSTRUMENTATION TABLE NOTATIONS

' Monitor must be capable of responding to a lower Limit of Detection of IE-4 C1/ml.

During releases via this pathway.

- During operation of the main condenser air ejector.

ACTION 104 - With the number of channels OPERABLE less than required by the !

Minimum Channels OPERABLE requirement, effluent releases vir. this pathwty may continue, provided the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the number of channels cidtABLE remains less than required by the Minimum.

Channels OPERABLE requirement for over 30 days, an explanation of the circumstances shall be included in the next Semiannual Radioactive Effluent.

l Release Report.

ACTION 105 - With the number of chaiiels OPERABLE less than required by the i Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided grab samples are taken daily and analyzed daily for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With the number of main stack monitoring system l channels OPERABLE less than required by the Minimum Channels OPERABLE l requirement, without delay suspend drywell purge.

If the number of channels OPERABLE remains less than required by the Minimum Channels OPERABLE requirement for over 30 days, an explanation of the circumstances shall be included in the next Semiannual Radioactivt Effluent Release Report.

I

\

L HAfCH ODCM, REV 7 2/7/92 2.6-13 l t

TABLE 2.6-1 (SHEET 4 0F 4)

RADIGACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATIONS ACTION 107 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue, provided samples are continuously collected with auxiliary sampling equipment for periods on the order of 7 days and analyzed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months i after the end of the sampling period.

l If the number of channels OPERABLE remains less than required by the Minimum l Channels OPEt1ABLE requirement for over 30 days, an explanation of the !

circumt tances shall be ir.cluded in the next Semiannual Radioactive Effluent !

Release Report. 1 ACTION 108 - With the number of channels OPERABt E less than required by the Minimum Channels OPERABLE requirement, release to the environment may continue ,

for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided: j

a. The offgas system is not bypassed, and D. The offgas post-treatment monitor (Dll-K615) or the main stack monitor (Dll-K600) is OPERABLE.

Otherwise, be in at least HOT STANDBY within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If the number of channels OPERABLE remains less than required by the Minimum i Channels OPERABLE requirement for over 30 days, an explanation of the i circumstances shall be included in the next Semiannual Radioactive Effluent l Release Report.

i l

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l l

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HATCH ODCM, REV 7 2/7/92 2.6-14 l

TABLE 2.6 ? (SilEET10F3)

O RADI0 ACTIVE GASEOUS EFFLUENT A0NITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL Instrument QJEg Q1EM CALIBRATION ILSI

1. Reactor Building Vent stack Monitoring Systen.

(Each Unit)

a. Noble has Activity Monitor D* M R Q(1)

b. Iodine Sampler Cartridge W'(3) NA NA NA

c. Particulate Sampler Filter W'(3) NA NA NA

d. Effluent System Flowrate Measaring Device D* NA R Q

e. Sampler Flowrate

( Measuring Device D* NA R Q

2. Recombiner Building Ventilation Monitoring System

a. Noble Gas Activity Monitor D* H R Q(1)

b. Iodine Sampler Cartridge W'(3) NA NA NA

c. Particulate Sampler Filter W'(3) NA NA NA'

d. Sampler Flowrate Measuring Device D* NA R Q O

HATCH ODCM, REV 7 2/7/92 2.6-15 l

r i

TABLE 2.6-2 (SHEET 2 0F 3)

O RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL Instrument QiL(f QiECK CALIBRAT10h IEjil

3. Main Stack Monitoring System

a. Noble Gas Activity Monitor D' M R Q(1)

b. Ioulpe Sampler W'(3) NA NA NA

c. Particulate Sampler W'(3) NA NA NA

d. Flowrate Monitor 0* NA R Q

e. Sampler Flowrate Monitor D* NA R Q

4. Condenser Offgas Pretreatment Monitor O (Each Unit)

Noble Gas Activity Monitor D** M R Q(1) 10 HATCH ODCM, REV 7 2/7/92 2.6-16 l L .

TABLE 2.6 2 (SHEET 3 Of 3)

RADIDACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS

0uring releases via this pathway.

"During operation of the main condenser air ejector.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm "

antaciation occurs if any of the following conditions exists:

n. Instrument indicates measured levels above the alarm / trip setpoint.

b. Circuit failure occurs. ,

c. Intrumt et indicates a down.cale failure.

(3) The CHANNEL CHECK shall consist of verifying the presence of a filter element and sampler flow at the weekly filter changeout.

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HATCH ODCH, REV 7 2/7/92 2.6-17 l

l TABLE 2.6 3 (SHEET 1 0F 3)

RADIOACTIVE ' GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM :

r E

Lower Limit I of '

Sampling Analysis Type of Detection

Gaseous Release Tvoe Freauency Freauency Activity Analysis .fuCi/ml)

A. Environmental M' M' Principal' Gamma Release Points Grab Sample Emitters IE-4 b

. 1. Main Stack H-3 IE-6

2. Reector Building e Vent (Each Unit) ;

3. Recombiner i Building Vent d

B. All Release Types Continuous' W 1-131 lE-12

~(as listed in A Charcoal --- -----.------------.------

above) Sample 1-133 1E.10 d

Continuous' _W Irincipal, Gamma Particulate L'itters ,

Sample (1-M1,Others) lE-Il ,

Continuou: M Gross-Alpha IE-Il Composite Particulate Sample Continuous' Q Sr-89, Sr-90 1E-Il Composite .

Particulate '

Sample V

HATCH ODCH, REV 7 2/7/92 2.6-18 l

- . . - - .. - . _ _ _ _ - - . = _ _ _ = -_ .-- --- - . . . . -

TABLE 2.6-3 (SHEET 2 0F 3)

RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS

a. Lower Limit of Detection is defined in Table IW.ation (a) of Table 3.2-3.

b. For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possiale to measure radionuclides in enneentrations near the Lower Limit of Detection. Under these circumstances, the Lower Limit of Detection may be increased inversely proportional to the magnitude of the gamma yield (i.e., IE-4/1, where ! - photon abundance expressed as a decimal fraction), but in no i case shall the Lower Limit of Detection, as calculated in this manner for-I a specified radionuclide, be greater than 10 percent of the Maximum Permissible Concentration value specified in 10 CFR 20, Appendix B, Table II, Column 1.

I

c. Sampling and analyses for principal gamma emitters shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15

percent of the RATED THERMAL POWER within a 1-hour period if analysis

, hows that the DOSE EQUIVALENT l-131 concentration in the primary coolant and the Main Stack Noble Gas Activity Monitor reading have increased by a factor of 3.

i

d. Sampling shall be performed weekly, and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler). Sampling shall L also be performed once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15 percent RATED THERMAL POWER in I hour and analyses completed within-48 hours of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding Lower Limits of Detection may be increased by a factor of 10. The more frequent sampling and analysis requirement applies only if analysis shows that the DOSE EQUIVALENT l-131 cancentration in the primary coolant and the Main Stack Noble Gas Activity Mwitor reading have-increased by a factor of 3.

O :

HATCH ODCM, REV 7 2/7/92 2.6-19 l

TABLE 2.6-3 (SHEET 3 0F 3)

RADI0 ACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS

e. The ratio of the sample flowrate to the sampled stream flowrate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Sections 2.6.2, 2.6.3, and 2.6.4.

~~

f. The principal gamma amitters for which the Lower Limit of Detection requirement will apply are exclusively the following radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-138 for noble gas releases; and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Co-144 for particulate releases. This list does not mean that only these nuclides are to be detected and reported. Other measurable and identifiable peaks, together with the above nuclides, shall also be identified and reported. Nuclides below the Lower Limit of Detection for the analyses should not be reported as being present at the lower Limit of Detection level for that nuclide. When unusual circumstances result in a Lower Limit of Detection higher than required, the reasons shall be documented in the Semiannual Radioactive Effluent Release Report.

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HATCH 00CH, REV 7 2/7/92 2.6-20 l

3.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM O Sampling locations, as required by Section 3.1.1, are described in Table 3.0-1 l and shown on maps in Figures 3.0-1 through 3.0-3.

There are no known u inking water users downstream of Plant Hatch. Therefore, the LLO for I-131 in water need not be as stringent as that for milk.

The census of MILK ANIMALS is based on the requirement in Appendix I to 10 CFR l-Part 50 that the licensee " Identify changes in the use of UNRESTRICTED AREAS (e.g., for agricultural purposes) to permit modifications in monitoring-programs for evaluating doses to individuals from principal pathways of exposure." The consumption of milk from animals grazing on contaminated

pasture and the consumption of vegetation contaminated by airborne radiolodine hre major potential sources of exposure. Samples from MILK ANIMALS are considered a better indicator of radioiodine in the environment than "egetation. Because sufficient milk samples frequently are not available within 5 miles, vegetation samples will be collected also.

O ~

Grass is available almost year-round, whereas leafy vegetation is available only for 8 months of the year at best. The sampling stations for grass are located near the SITE B0UNDARY in two sectors with high offsite D/Q values l where it might be practical to establish a vegetation plot. The highest l offsite D/Q for each individual sector occurs =approximately at the SITE

! B0UNDARY.

l L Although either fish or clam samples may be collected from the river, fish samples are preferred, because the maximum dose commitment to a MEMBER OF THE l

PUBLIC as a result of liquid effluents is through the fish , ansumption '

pathwry.

Sediment will be collected annually, because shoreline recteational areas are under water and, thert re, not in use approximately half of the year. -

l l

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HATCH ODCM, REV 7 2/7/92 3.0-1 l l

4 r

Allowing deviations from the sampling schedule is based on the recognition of unavoidable practical difficulties which, in the absence of the allowed O' deviations, would result in violatloc of Technical Specification 6.19(1). I l

l l

l 1

l j

I i

1 i

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n v

HATCH ODCM, REV 7 2/7/92 3,0-2

-ll c .

3.1 LlHITS OF OPERATION 3.1.1 Radioloaical Environmental Monitorino Proaram in accordance with Technical Specification 6.19(1), the Radiological

]

Environmental Monitoring Program shall be conducted as specified in lable 3.1-1, 3.1.1.1 Applicability l i

At all times. '

l 3.1.1.2 Action ;

3.1.1.2.1 Any deviations in conducting the Radiological Environmental Monitoring Program' from that as specified in Table 3.1-1 shall be documented in the Annual Radiological Environmental Surveillance Report; the reasons for these deviations and any appropriate plans for preventing a recurrence shall:

be stated. (Deviations are permitted from the required sampling schedule if- l specimens are unobtainable due to hazardous conditions, unavailability, inclement weather, malfunction of equipment, or other just reasons. If deviations are due,to equipment malfunction, strenuous efforts shall be made to complete corrective action prior to the end of the next sampling period.)

3.1.1.2.2 With the confirmedb , measured level of radioactivity as a result of a plant effluents in an environm9ntal sampling medium as specified in Table 3.1-1 exceeding the reporting levels of Table 3.1-2 when averaged over any calendar quarter, submit to the Nuclear Regulatory Commission within 30 days or after confirmation, whichever is later, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the

The requirements for radiological environmental menitoring are the same for both units at the site. Thus, a single program including monitoring, land use census, and quality assurance serves both units, b

Defined as a Confirmatory reanalysis of the original, e duplicate, or a riew sample, as-appropriate. The results of the confirm 9 tory analysis shall. be completed at .the earliest time consistent with the analvsis.

a HATCH ODCH, REV 7 2/7/92 3.1-1 l

1 4

3.1.1.2.2 (Continued)

O limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose

to a MEMBER OF THE PUBLIC is less ;

than the calendar year limits specified in Sections 1.7.3, 2.6.3, and 2.6.4.

When more than one of the radionuclides in Table 3.1-2 are detected in the sampling medium, this report shall be submitted if:

(concentration (l)/liiait level (l)) + [ concentration (2)/ limit level (2)) +...>1.0 i L When radionuclides other than those in Table 3.1-2 are detected and are the :

result of ,)lant effluents, this report shall be submitted if the calculated annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the annual limits stated in Sections 1.7.3, 2.6.3, and 2.6.4. This report is not

required if the measured level of radioactivity was not the result of plant ,

effluents; however, in such an event, the condition shall be described in the '

Annual Radiological Environmental S rveillance Report. The levels of naturally occurring radionuclides need not be reported.

3.1.1.2.3 If adequate samples of milk, grass or leafy vegetation (during the growing season) from any of the sample locations required by Table 3.1-1 can no longer be obtained or the availability is frequently or persistently wanting, efforts shall be made to find replacement locations. The cause of the unavailability and identification of the affected locations and the locations (if any) for obtaining replacement samples shall be submitted 'to the g Nuclet.r Regulatory Commission in the next Semiannual Radioactive Effluent 4 Release Report. The locations from which samples became unavailable may be deleted; however, any locations from which suitable replacement samples are available shall be added to the program. *

The methodology and parameters used to estimate the potential annual dose to a-MEMBER OF THE-PUBLIC shall be indicated in this report.

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HATCH ODCM, REV.7 2/7/92 3.1-2 l ,

3.1.1.3 Sur'eillance Requirements The radiological environmental monitoring samples shall be collected, pursuant to Table 3.1-1, frm the locations described in Section 3.0, and shall be analyzed pursuant to the requirements of Tables 3.1-1 and 3.1-3, 3.1.1.4 Bases The Radiological Environmental Monitoring Program required by this control provides measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides leading to the highest potential radiation exposures of MEMBERS OF THE PUBLIC, resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring program by measuring concentrations of radioactive materials and levels of radiation that may then be compared with those expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

3.1.2 Land Use Census in accordance with Technical Specification S.i.(2), a land use census shall be conducted to identify the location of the ner. rest MILK ANIMAL and the nearest permanent residence in each of the 16 meteorological sectors within a distance of 5 miles and the locations of all MILK ANIMALS within a distance of 3 miles.

3.1.2.1 Applicability At all times.

3.1.2.2 Action 3.1.2.2.1 With a land use census identifying a location (s) yielding a calculated thyroid dose or dose cond , scat greater than the values currently being calculated in accordance with 5 W.J.an 2.6.4, tubmit the new location (s) ,

to the Nuclear Regulatory Commission in the next Semiannual Radioactive Effluent Release Report.

O MATCH ODCM, REV 7 2/7/92 3.1-3 l a

i 3.1.2.2.2 With a land use census ide-tifying a location (s) yielding a calculated thyroid dose or dose commitment (via the same exposure pathway) 20 O0 percent greater than at a location from which samples are currently being obtained, add the new location (s) to the program within 30 days if sampler are available. The sampling location having the lower calculated thyroid dose may then be deleted from the program.

3.1.2.3 Surveillance Requirements The land use census shall be conducted once per 12 months by door-to-door survey, by visual survey from automcbile or aircraft, by consulting ic.al ~

agriculture authorities, or by a combination of these methods as feasible using the information to provide a good census. Results of the annual census, as well as any chenges in sampling locations, shall be discussed in the Annual Radiological Environmental Surveillance Report.

3.1.2.4 Bases y

This control is provided to ensure that changes in the use of UNRESTRICTED AREAS are identified and that modifications to the monitoring program are made, if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix 1 to 10 CFR Part 50, 3.1.3 I n t e rl a b e r alp r y C omp ariLQalf ou r a m In accorriance with Technical Specification 6.19(3), analyses shall be -

performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Nuclear Regulatory Commission. Analyses need to be performed only where the type analysis and sample are the same as that required in Table 3.1-1.

3.1.3.1 Applicability At all times.

O liATCH ODCM, REV 7 2/7/92 3.1-4 l

.~

3.1.3.2 Action With analyses not being performed as required above, report the corrective actions taken (to prevent a recurrence) in the Annual Radiological Environmental Surveillance Report.

3.1.3.3 Surveillance Requirements A summary of results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Surveillance Report. _

3.1.3.4 Dases The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental monitoring to demonstrate that the results are reasonably valid.

O n i \.e\m.te m\.ac=\.as=2.*

HATCH ODCM, REV 7 2/7/92 3.1-5 l

TABLE 3.1-1 (SHEET I 0F 4)

W

{ RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 8 Approximate

~

9 Number of Sample Sampling and Type of Analysis E Exposuse Pathway Collection Fre<1uency and Freauency and/or Sample locations *

]

~ 1. Airborne Continuous operation of Radiciod .a canister.

b" a. Radiciodine sampler with sample collection I-131 weekly.

and Particulates 5 weekly. J l

Particulate sampler.

Analyze for gross teta radio-activity not less than 24 w hours following filter change

_. and analyze for I-131 weekly.

a Perform gamma isotopic analysis on affected sample when gross beta activity is 10 times the yearly mean of control samples. Composite (by location) for gamma isotopic analysis quarterly.

Quarterly Ca::ma dose quarteiy.t

2. Direct Radiation 35

e~

_ TABLE 3.1-1 (381 2 0F 4) s F1 RADIOLOGICAL ENVIRONMENTAL MONIf0 RING PR0 ORAM z

O

~

Sc Approximate Number EI Exposure Pathway cf Sample Sampling ar.d Type of Analysis

_ ano/or Samole Locations

Collection Frecuency and Frecuency

[ ,

e e3 Ingestion

~

23

a. Milk 4' Bi-weekly u. .a isotopG and I-131 analyses bi- weekly.

b. Fish

or clams 2 Semi-annually Gamma isotopic analysis on edible portions semi-annually.

c. Grars or Leafy 3 Monthly during growing Ga s;a isotopic analysis i ;. Vegetation season. monthly.d 1,

4. Waterborne

a. Surface 2 Composite

sample .sna isotopic analysis monthly.

collected monthly. .:omposite (by location) for tritium analysis quarterly.

l b . .w e 9t 1 Yearly Gaana isotopic analysis sar.1ple yearly.

j i

l i

1 m

d' . ;.

7

/ i i Jg : ,

.a. .o g

en .

O O 9 TABLE 3.1-1 (SHEET 3 0F 4)

D RADIOLCGICAL ENVIRONMENTAL MONITORING PROGRAM E

~ o 8

2 Approximate

~

rNmber of Sample Sampling and Type of Analysis

I Exposure Pathway and Frecuency Locations

Ccllection Freauency and/or Samole

[

4. Waterborne (Continued)

River Water collected near the I-131 analysis on each sample D c. Drinking Water 1*

when bi-weekly collections are

1** intake will be a composite sample; the finished water will required. Gross beta .ind gamma be a grab sample. These samples isotopic analyses on each sample; will be collected monthly unless composite (by location) for the calculated dose due to con- tritium quartcrly.

sumption of the water is greater w than 1 mrem / year; then the

collection will bc 'oi-weekly.

a The collection may revert to monthly should the calculated doses become less than 1 mrem /yaar.

1

0ne sample of' river water near the intake.

- 0ne sample of finished water frem each of one to three of nearest water supplies which could be affected by HNP discharge.

i

. _ *

j

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TABLE 3.1-1 (SHEET 4 0F 4)

E y RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM TABLE NOTATIONS 8

? R y a. Sample locations are shown in Table 3.0-1 and Fiiures 3.0-1 through 3.0-4.

In

b. Up to three sampling locations within 5 miles ani in different sectors will be used as available.

addition, one or more control locations beyord 10 miles will be used.

Clams will be sampled if difficulties D c. Commercially or recreationally important fish my be sampled.

" are encountered in obtaining sufficient fish samples.

d. If gamma isotopic analysis is not sensitive enough to meet the Lower Limit of Detection, a separate analysis for I 131 may be performed.

m e. Composite samples shall be collected by collecting an aliquot at intervals not exceeding a few hours.

If it found that river water downstream of HNP is used for drinking, water samples will be collected l 1 f.

and analyzed as specified herein.

.>se who

g. A survey shall be conducted annually at least 50 river r<:'es downstream of HNP to identify use Altamaha River water for drinking.

O O O ..

TABLE 3.1-2 y REPORTING LEVELS FOR RAD 10ACTIVITS CONCENTRATIONS IN ENVIRONMENTAL SAMPLES

-8 9 Water Airb rne Particulate 3

Fish Milk Grass

~

Analysis (oCi/1)' _g .ases (oci/m ) (DCi/ko. wet) (oci/1) (oCi/ka yet.1 M.

' - [' H-3 3E4*

q Mn i.E3 3L4 Fe-59 4E2 lE4-Co-58 IE3 3E4 Co-60 3E2 1E4 y Zn-65 3E2 2E4 h Zr-95 4E2 Nb-95 6E2 I-131 2E0 9E-1 3E0 1E2 Cs-134 3El IEl IE3 6El IE3 Cs-137 SE1 2E1 2E3 7El 2E3

'Ba-140 2E2 3E2 La-140 2E2 4E2

For drinking water samples, the reporting level is 2E4 pCi/ liter.

I m

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--3 O O O TABLE 3.1-3 (SHEET I 0F 3) i LOWER LIMIT OF DETECTION.b

]n 8

9 Airborne

~

Particulate Grass Sediment Water orGay Fish Milk M

foci /1) foci /ko. wet) foci /ka. dry)

Analysis (cci/1) f oCi/m ) 19Ci/ko. wet) gross beta 4E0 IE-2 d

H-3 2E3 2E0 IE2 Hn-54 3El 3E2 fe-59 w Co-58 2El IE2 IE2 j- 00-60 2El 3El 3E2 Zn-65 Zr-95 3El Nb-95 2El IE0 6El

.I-131* IE0 7E-2 IE2 2El 6El 2E2 Cs-134 ZEl SE-2 2E1 8E1 2E2 Cs-137 2E1 6E-2 2E2 6El 6El Ba-140 2E1 La-140 2EI i

i

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O O O TABLE 3.1-3 ' SHEET 2 0F 3) 5 Mx LOWER LIMIT OF DETECTION l

y TABLE NOTATIONS :

51 a. The Lower Limit of Detection (LLD) is the smallest concentration of radioactive material in a sample that will be detected with 95 percent probability with 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, wnich may include radiochemical separation:

LLD - (4.66

3s ) / U

V

2.N

Y

up(-M)) ,

L where:

LLD = the "a priori" Lower Limit of Detection (defined as pCi per unit mass or volume). '

w sb = the standard deviation of'the background counting rate or of the counting rate of a blank sample !

p as appropriate (as counts per minute).

E = the counting efficiency (as counts per disintegration).

V - the sampin size (in units of mass or volume).

2.22 = the-number of disintegrations per minute per picocurie, r Y = the fractional radiochemical yield (when applicable). t i

A = the radioactive decay constant for the particular radionuclide.

AT = the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples).

AT = _the elapsed time. between midpoint of sample collection period and time of counting (for plant effluents).

j

O O O TABLE 3.1-3 (SHEET 3 0F 3)

E LOWER LIMIT OF DETECTION M

x TABLE NOTATIONS 8

2

~

The value of s 3 used in the calculation of the LLD for a detection system shall be based on the actual 5 observed variance of the background counting rate or of the conting rate of the blank samples (as

appropriate) rather than on an unverified thecretically predic W variance. Typical values of E, V,

" Y, and AT should be used in the evaluation.

m Other D b. This does not mean that only the radionuclides in Table 3.1-3 are to be detected and reported.

D measurable and identifiable peaks, together with the above nuclides, shall bc identified and reported.

" Only manmade radionuclides need be reported.

c. LLD for drir. king water samples. If no drinking water pathway exists, the LLD for gama isotopic analysis may be used.

w d. If no drinking water pathway exists, a value of 3E3 pCi/ liter may be used.

7 w

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4.0 10TAL DOSE DETERMINATIONS O For the purpose of implementing Section 4.1, total dose determinations will be made by:

Calculating doses due to liquid effluents in acccrdance with Section 1.7.3. l Calculating doses due to gaseous effluents in accordance with Section 2.C.4. [ .

Combining direct radiation doses based on direct radiation measurements with these effluent doses to determine total dose to a MEMBER OF THE PUBLIC. l Tne methodology for calculating doses to a MEMBER OF THE PUBLIC due to liquid l effluents is presented in Section 1.2, and the methodology for calcula. ting doses to a MEMBER OF THE PUBLIC due to gaseous effluents is presented in l Section 2.2.2.b.

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O HATCH ODCH, REV 7 2/7/92 4.0-1 l

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4.1 LIMITS OF OPERATION 4.1.1 Total Dose Control In accordance with Technical Specification 6.18(10), the annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of i radioactivity t.nd to radiation from uranium fuel cycle sources shall be limited to less than or equsi to 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less t%n or equal to 75 mrem.

4.1.1.1 Applicability "

At all times.

4.1.1.2 Actions 4.1.1.2.1 With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Sections 1.7.3, 2.6.3, or 2.6.4, calculations shall be made including direct radiatior.

contributions from the reattor units and from outside storage tanks to determine whether the limits stated above in_Section 4.1.1 have been exceeded.

If such is the case, in lieu of a Licensee Event Report, prepare and submit to l the Nuclear Regulatory Commission within 30 days, pursuant to Technical-

, Specification S.9.2, a Special Report that defines the corrective action (s) to be taken to reduce subsequent releases to prevent recurrence of exceedir.g the

! above timits and include the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.405c, shalliinclude an-analysis estimating the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent _ pathways and direct radiation, for the calendar year. that includes the release (s) covered by this report. It.shall also describe levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels er concentrations. If the estimated dose (s) exceeds the above limits, and if the releue condition resul. ting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely reouest, and a variance is granted until staff action on the request is complete.

l l

LO

-HATCH 00CM, REV 7 2/7/92 4.1 l

P 4.1.1.2.2 For Unit 1: When the ACTION statement or other requirements of this control cannot be met, steps need not be taken to change the Operational Mode O of the Unit. Entry into an Operational Mode or other specified condition may_

be made if, as a minimum, the requirements of the ACTION statament are satisfied. .

For Unit 2: The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

4.1.1.3 Surveillance Requirements , 4.1.1.3.1 Cumulative dose contributions fro,n liquid and gaseous effluents snall be determined in accordance with Sections l'.7.3.3, 2.6.3.3, and 2.6.4.3 and in accordance with the methodology and parameters described in Sections 1.2, 2.2.2.a, and 2.2.2.b.

4.1.1.3.2 Cumulative dcse contributions from direct radiation from the reactor units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters described in Section 4.0. This requirement is applicable only under conditions set forth :.bove in Section

~

4.1.1.2.1.

l-4.1.1.4 Bases l This control is provided to meet the reporting requirements of 40 CFR 190.

l.

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l HATCH ODCM, REV 7 2/7/92- 4.1-2 l

.= _____ _ _ __._ _ _._____ _ _ . _ -. ~

J

5.0 POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE A TO THEIR ACTIVITIES INSICE THE SIILjlpVNDARY U

for the purpose of implementing Section 6.2.1.2, an assessment of potential 8 doses to MEMBERS OF THE PUBLIC due to their activities within the SITE BOUNDARY will be performed if circumstances have changed such that any of the limits of Sections 2.6.3 or 2.6.4 are exceeded. The locations of concern. l within the SITE BOUNDARY are the roadside park, the camping area, the l recreation area, and the Visitors Center. Historical annual average dispersion and deposition values, and elevated plume dose factors for these locations are presented in Tables 5.0-1 through 5.0-4, along with the estimated occupancy factors. (Estimated occupancy factors are for a MEMBER OF THE PUBLIC during a year.)

In the event that any limit of Section 2.6.3 is exceeded, an assessment will be performed considering direct radiation dose to a MEMBER OF THE PUBLIC resulting from noble gases in the plume. This assessment will take into consideration the annual average dispersion _ parameters and the estimated occupancy factors stated in Tables 5.0-1 through 5.0-4, or more precise values if available, for the locations of interest.

I In the event that any limit of Section 2.6.4 is exceeded, an assessment will j be performed considering the dose to a MEMBER 0.i THE PUBLIC due _to inhalation !

of airborne radioactive material <. suspended in the plume and due to radioactive material de?>sitec on the ground. This asse;sment will take into consideration the annual average dispersion and deposition parameters and the i estimated occupancy factors stated in Tables 5.0-1 through 5.0-4, or more orecise values if available, for the locations of interest.

O HATCH ODCid, C 7 2/7/92 5.0-1 l

If none of the limits discussed above is exceeded, potential annual doses tn a O MEMBER OF THE PUBl.IC at locations of concern within the SITE B0UNDARY are as follows:

l Potential doses to a MEMBER OF THE PUBLIC at the Visitors Center are not l expected to exceed 0.03 mrem due to inhalation and ground-plane exposure, and 0,02 v em due to direct radiation from the plume.

Potential doses to a MEMBUt 0F THE PUBLIC at the roadside park are not l expected to exceed 0.005 mrem due to inhalation and ground-plane exposure, and 0.003 mrem due to direct radiation frem the plume.

Potential doses to a MEMBER OF THE PUBLIC at the camping area are not l expected to exceed 0.04 mrem due to inhalation and ground-plane exposure, and 0.06 mrem due to direct radiation from the plume.

i Potential doses to a HEMBER OF THE PUBLIC at the recreation area are not l expected to exceed 0.6 mrem due to inhalation and ground-plane exposure, and 0.3 mrem due to direct radiation from the plume.

These values are based on annual average dispersion and deposition parameters and the estimated occupancy factors referenced above. Estimated occupancy factors for the recreation area, Visitors Center, the roadside park, and the camping area are based on activities observed at these locations over the las_t several years and continued anticipated usage of these areas.

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SECTION 6.0 REPORTS 0

O HATCH ODCM, REV 7 2/7/92 l

6.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT O ~In accordance with Technical Specification 6.9.1.6, the Annual Radiological Environmental Surveillanca Report covering the radiological environmental surveillance activities related to the plant during the previous calendar year shall be submitted before May 1 of each year. Tha report shall include summaries, interpreictions, and analysas of trends of the results of the ,

Radiological Environmental Monitoring Program for the reporting-period. The material provided shall be consistent with the objectives outlined in this ODCM and Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

6.1.1 Reoort Cantents v.l.l.1 Semmaries, interpretations, and statistical evaluations of the results of the radiological environmental surveillance activities for the reporting period, including (as appropriate) a compariso.: with the preoperational studies, operational controls, previous environmental surveillance reports, ano an assessment d any observed impacts of the plant cperation on the environment shall ~ o e included in the report. The report shall also include the results of the land use census required by Section 3.1.2 and the results of licensee participation in the Interlaboratory Comparison Program required by Section 3.1.3 6.1.1.2 The report shall include summarized and tabulated results in thc format of Table 6.1-1 of all raciological environment :1 samples taken during _

the report period, with the exception of naturally occurring-radionuclides which need not be reported. In the event that some results are not available for inclusion with the report, the raport shall be submitted, noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as practicable in a supplementary report.

6.1.1.3 Also to be included _in the report are the following: a summary description of the Raatological Environmental Monitoring Program, a map of all sampling locations as keyed to a table indicating distances and directions from the main ' stack, and results of the licensee participation in the O

5 Interlaboratory Comparison Program.

' HATCH CDCM, REV 7 2/7/92 6.1-1 1

6.1.1.4 Any deviations in conducting the Radiological Environmental Monitoring Program from that specified in Taule 3.1-1 shall be documented in the report, in accordance with Section 3.1.1.2.1.

6.1.1.5 If the measared level of radioactivity in an environmental sampling medium is not the result of plant effluents, the cor.lition shall be reported as required by Section 3.1.1.2.2.

6.1.1.6 In addition to the radionuclides listed in Table 3.1-3, other measurable and identifiable peaks shall be identified and reported. Only manmade radionuclides need be reported.

6.1.1.7 If Interlaboratory Comparison Program analyses are not performad as rte ired by Section 3.1.3, the corrective actions t?. ken to prevent a c recurrence must be included in the report. -

O O

HATCH ODCM, REV 7 '/7/92 6.1-2 l

. . . m. . .

1 O O O

L.

TABLE 6.1-1 '

sE F4: ENVIRONMENTAL RADIOLOGICAL M3NITORING PROGRAM

SUMMARY

L :n ,

lj Name of Facility Edwin I. Hatch Nuclear Plant Docket No. 50-321. 50-365

=c.

~

Location of Facility Applina County. Georaic Reporting Period C2.

l Medium or Type and na Pathway: Sampled Total Number Lower Limit- All Indicator Name, Control Number of 23 (Unit of of' Analyses of Locations Distance, Mean Locations REPORTABl.E jg Mea;urement) Performed Detection

Mean Rance and Direction Range b Mean Rangg EVENTS I

T w

1-4 i

Lower Limit of Detection is defined in Table Notation a of Table 3.I-3.

b Mean and range based upon detectabic measurements cnly. Fraction of detectable measurements at specified 'i locations is. indicated-in parentheses.

<nov..unwu.we - ,

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6.2 SEMIANNUAL RADIDACTIVE EFFLUENT RELEASE REPORT In accordance witt. Technical Specification 6.9.1.8, the Semiannual Radioactive Ef fluent Release Report covering the operation of the plant during the previous 6 months of operation shall be submitted withi;. 60 days after January 1 and July 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant.

~

. material prov'.ded shall be consistent with the objectives outlined in this ODCM and tha Process Control Program (PCP) and in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendi) I to 10 CFR Part 50.

6.2.1 Report Contents 6.2.1.1 The report shall include a summary of the quantities of r dioactive liquid and 91seous effluents and solid waste released from each unit as outlined in Regulatory Guide 1.21, "Mer uring, Evaluating, and Reporting Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

6.2.1.2 The report to be submi+.ted 60 days after January 1 of each year shall include and annual summary of meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, atme:pheric stability and precipitation (if measured) on magnetic tape or in the form of joint frequency distributions cf wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report rhall include an assessment of the radiation doses from liquid and gaseous cffluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BO'iNDARY (Figure 1.7-1) during the reporting period if circumstances have changed such that the potential doses are significantly greater than expected at onsite locations as discussed in Section 5.0. All assumptions used in making th;se assesrments, i.e., specific V

HATCH ODCM, REV 7 2/7/92 6.2-1 l )

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

activity, (xposure time, and location, shall be included __in these reports.

Historical innual average meteorological conditions or meteorological conoitions cencurrent 'with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be-used for determining the gaseous pathaay doses, lhe atsessment of radiation doses shall be performed in accordance with Sections i.7.3, 2.2.2, and 5,0.

6.2.1.3 For each type of solid waste shipped offsite during the report period, the following information shall be included in the report:

a. Container volume, ,

b. Total curie quantity (specify wh3ther determined by measurement or est imate).

c. Principal radionuclides <;pecify whether determined by measurement or estimate). ,

d. Type of waste, e.g., spent resin, compacted dry whste, evaporator be '.t om s .

, c. Type of container, e.g., LSA, type A, type B, large quantity.

f. Solidification agent, e.g., cement.

6.2.1.4 The report shall include (on a quarterly basis) unpinned releases from the site-to UNRESTRICTED AREAS of radioactive materials in gaseous anJ liquid effluents that were in excess of I Ci, excluding dissolved and entrained gases and tritium for liquid effluents, or those in excess of 150 Ci of noble gases ot 0.02 Ci of radioiodines for gaseous releares.

1 O

HATCH ODCM, REV 7 2/7/92 6.2-2 g

, s

6.2.1.5 Any changes to the ODCH made during the reporting period shall be included in the raport.

6.2.1.6 If a radioactive liquid effluent monitoring instrumentation channel requireu by Tab'ic -1.7-1 remains inoperable for over 30 days, an explanation of the circumstances shall be included i') the next report. This requirement does ,

not include the Service Water System to Closed Cooling Water System ;

Differential Piessure channel.

6.2.1.7 If the concentration of radioactive material released frem the site to UNRESTRICTED AREAS exceeds the limits of Section 1.7.?, a discussion of the causes and enrrective actions taken must be included ir, the next report.

6.2.1.8 If a raoloactive gaseous effluent monitoring instrumentation channel required by Table 2.6-1 remains inoperable for over 30 days, an explanation of the circumstances shall be included in the next report.

6.2.1.9 When unusual circumstances result in a lower Limit of Detection higher than required by Table 2.6-3 (which addresses radioactive gaseous sample analysis), the reasons shall be documented in the report.

6.2.1.10 If adequate samples of milk, grass or leafy vegetation (during the growing season) from any of the sample locations required by Table 3.1-1 are unavailable, the cause of the unavailability and identification of the affected location (s) and the location (s), if any, for obtaining replacement samples shall be submitted in the next report.

4 6.2.1.11 With a land u;e census identifying a locatim(s) yielding a calc.ulated thyroid dose or dose commitment greater than the values currently being calculated in accordance with Section 2.6.4, submit the new location (s) in-the next report.

6.2.1.12 If the contents within any outside temporary tank exceed the timits of Technical SpecificM sc>. 3.15.1.4 (Unit 1) or 3.11.1.4 (Unit 2),

notification shall bo included in the report.

HATCH ODCM, REV 7 2/7/92 6.2-3 l

6.3 MONTHLY OPERATING RE?0RT O A report of hny major changes to the radioactive wa:;te treatment systems shall be submitted to the Nuclear Rcgulatory Commission with the Monthly Operating Report (which must be suomitted in accordance with Tect.nical Specification 6.9.1.10) for the period in which the 2 valuation was reviewed at.d accepted by the Plant Review Board.

G 4

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O HATCH ODCM, REV 7 2/7/92 6.3-1 l

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6.4 (DECIAL REPORTS Special reports shall be submitted to the Nuclear Regulatory Connission in accordance with Technical Specification 6.9.2 as required by Sections 1.7.3.2.1, 1.7.4.2.1, 2.6.3.2.1, 2.6.4.2.1, 2.6.5.2.1, 3.1.1.2.2, and 4.1.1.2.1.

This section addresses only reporting requirements included in the ODCH; special reports aisc may be required under circumstances described in Plant Hatch Technical Specifications. -

O O

HATCH ODCM, REV 7 2/7/92 6.4-1 l J

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SECTION 7,0 GENERAL DEFINITIONS i

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1 HATCH ODCM, REV 7 2/7/92 l=

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7.1 TERM TRANSFERRED TO THE ODCM FROM TECHNICAL SPECIFICATIONS O The term defined in this secticn was transferred from the T9chnical Specifications to the ODCM in accordance with NRC Generic Letter 89-01.

Wherever this term appears in the text of the Limits of Operation Sections of the ODCM. it is presented in all capital letters to indicate ~ that it is specifically defined, t 7.1.1 GASEOUS RADWASTE TREATMENT SYcTDj The GASEOUS RADWASTE TREATMENT SYSTEM is the offgas holdup system designed and q installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or -

holdup for the purpose of reducing the total radioactivity prior to release to ,

the environment.

O O

HATCH ODCM, REV 7 2/7/92 7.1-1 -l lj

7.2 TERMS DEFINED IN TECHNICAL SPECIFICATIONS O Tne following definitions are contained in the Technical Specifications Section 1.0 " Definitions." Because these terms are used extensively "

throughout the ODCH, they are also included in this section for convenience.

Throughout the text of the Limits of Operation Sections of the ODCM, these terms are presented in all capital letters to indicate that they are specifically defined. In some cases, the definition of a particular term is not exactly the same for both units; in that case, both definitions are presented along with an indication of the unit to which the definition applies.

7.2.1 ACTION for Unit 1: ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.

For Unit 2: ACTIONS shall be those additional requirements specified as corollary statements to each specification and shall be part of the specifications.

7.2.2 CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel :

output such that it responds with the necessary range and accuracy to known vtlues of the parameter which the channel monitors The CHANNEL CALIBRATION shall encompass the entire channel including th2 sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be perfor.med by any series of-sequential, overlapping or total channel steps such that the entire channel is calibrated.

HATCH ODCM, REV 7 2/7/92 7.2-1 l

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7.2.3 CHANNEL CHECK i.a A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indicetions and/or status derived from indepondent instrument channels measuring the same parameter.

7.2.4 CHANNEL FlgtTIONAL TEST 3

A CHANNEL FUNCTIONAL TEST shall be:

For Unit 1:

c. Analog channels - the injection of a simulated signal into the channel

as close to the primary sensor as practicable to verify PPERABILITY including alarm and/or trip functions,

b. Bistable channels - the injection of a simulated signal into the channel tj sensor to verify OPERABILITY including alarm and/or trip functions.

For Unit 2:

The definition is the same as for Unit I with the exception that "and channel failure trips" is added to item a.

l.2.5. DOSE EQUIVALENT IODINE The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (mic.rocurie/ gram), which alone would produce the same thyroid dose as the quantity and iat apic neixture of I-131, I-13 1-133, I-134, and I-135 a:tually prese n . The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 or those in NRC Regulatory Guide 1.109, Revision 1, October 1977.

D r V HATCH ODCH, REV 7 2/7/92 7.2-2 l

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

I 7.2.6 ~FRE0VENCY NOTATION

, The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals as follows:-

i

).

NOTATIOP( DEFINITION FRE0VENCY

S Once per shift 0..ce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l- D Daily Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

? W Weekly Once per 7 days y M Monthly Once per 31 days Q Quarterly Once per 92 days

, SA Semi-annually Once per 184 days ;

J R REFUELING .Once per 18 months S/U STARTUP Prior to each reactor startup

1. P Prior Completed prior to each release NA Not applicable Not' applicable 7.2.7 MEMBER (S) 0F THE PUBLIC i

MEMBER (S) 0F_ THE PUBLIC shall include all persons who are not occupationally

. associated with the plant. This category does not include employees of the e utility, its contractors, or its vendors. Also excluded from thf t category are persons who enter the site to service equipment or to make deliveries.

l This category does- include persons who use portions of the site for recreational, occupational, .or other purposes. not associated with the plant.

7.2.8 MILK-ANIMAL l A MILK ANIMAL is a cow or goat-that is producing milk for human consumption.

i l

L t

Lv

HATCH 00CM, REV 7 2/7/92 7.2-3 l 1

y , - -

-.3 + ... y .. +- c ,. 9 4 ..,..,r ...~o..v_. y 7m. ,,,,,- #y . .g w% + me.w g _ % ,. , + .,--g Mo v = r- w ww e"'V * =

h 7.2.9 OPERATIONAL CONDITION l

'O For Unit 1: This term is not specifically defined. See the definition of REACTOR MODE below.

j For Unit 2: An OPERATIONAL CONDITION shall be any one inclusive combination j of mode switch position and average reactor coolant temperature indicated'as follows:

i

! HODE SWITCH AVERAGE REACTOR

CONDITION POSITION COOLANT TEMPERATURE i

1. POWER OPERATION Run Any Temperature i 2. STARTUP Startup/ Hot Standby Any Temperature l 3. HOT SHUTOOWN Shutdown > 212" F***

4. COLD SHUTDOWN Shutdown < or - 212 F***

5. REFUELING

Refuel ** < or = 212' F i

O Reactor vessel head unbolted or removed and fuel in the vessel.

See Special Test Exception 3.10.3.

f During the performance of inservice hydrostatic or leak testing with all control rods fully inserted and reactor coolant temperatures above 212 F, the reactor may be considered to be in the COLD SHUTDOWN condition for the purpose of determining Limiting Condition for Operation applicability. However, compliance with an ACTION requiring COLD SHUTDOWN shall require a reactor coolant temperature < or - 212 F. In addition, compliance with the following specifications is required when performing the hydrostatic and leak testing under the identified conditions: 3.6.5,1, 3.6.5.2, 3.6.6.1, and 3.7.1.1.

HATLH ODCM, REV 7 2/7/92 7.2-4 l

I h

7.2.10 OPERABLE - OPERABILITY t

A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary

! attendant instrumentation, controls, normal and emergency electrical power f sources, cooling or seal water, lubrication or other auxiliary equipment that

{ are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support j function (s).

7.2.11 REACTOR MODE l For Unit 1: The reactor mode is established by the Mode Switch position. The switch positions are REFUEL, SHUTDOWN, START & HOT STANDBY and RUN; thus the

{ four possible reactor modes are: Refuel Mode, Shutdown Mode, Start & Hot Standby Mode, and Run Mode. (See Unit 1 Technical Specifications Section 1.0 3 " Definitions" for definitions of these terms.)

O For Unit 2: This term is not specifically defined. See the definition of f

OPERATIONAL CONDITION above.

i ,

! 7.2.12 RATED THERMAL POWER For Unit 1: Rated thermal power means the reactor it operating, at a steady i state power of 2436 megawatts thermal. This is also referred to as 100

{ percent thermal power.

1

For Unit 2: RATED THERMAL POWER shall be a total reactor core heat transfer to the reactor coolant of 2436 MW.

O HATCH ODCM, REV 7 2/7/92 7.2-5 l

7.2.13 SITE B0UNDARY O The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by Georgia Power Company, as shown in Figure 1.7-1.

7.2.14 SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

7.2.15 THERMAt POWER For Unit 1: This term is not defined.

For Unit 2: THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

7.2.16 UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY to which access for purposes of protection of MEMBERS OF THE PUBLIC from exposure to ,

radiation and radioactive materials is not controlled by the licensee. This includes any area within the SITE BOUNDARY used for residential quarters or for -long term industrial, commercial, institutional, and/or recreational purposes.

3 O

HATCH ODCM, REV 7 2/7/92 7.2-6 l

REFERENCES (Continued)

14. Hatch Nuclear Plant land Use Survev - 1987, Georgia Power Company, February 1987,

15. Letter to Georaia Power Company from Pickard. Lowe. and Garrick. Inc..

Washington, D.C., May 11, 1987.

16. Letter to Georaia Power Comoany from Pickard. Lowe. and Garrick. Inc..

Washington, D.C., June 3, 1987.

17. Letter to Georaia Power Comoany frca Pickard. Lowe and Garrick. Inc..

Washington, D.C., June 11, 1987.

18. Internal Memorandum. W.H. Ollinoer to D.M. Hooper, Georgia Power Company, June 9, 1987.

19. Letter to Georaia Power Company from Guantum Technoloav. Inc.,

Marietta, Georgia, June 17, 1987,

20. Hatch Nuclear Plant land Use Survey, Georgia Power Company, November 1987.

21. Letter to Georaia Power Comoany from Pickard. Lowe. and Garrick. inc.,

Washington, D.C., November 30, 1987.

22. Hatch Nuclear Plant land Use Survey, Georgia Power Company, O November 10, 1988, and December 20, 1988.

23. Letter to Georaia Power Comoany from J. H. Davis. Health Physics Consultant, Lilburn, Georgia, September 17, 1990.

24. Meinke, W.W. and T.H. Essig, "Offsite Dose Calculation Manual Guidance: r Standard Radiological Effluent Controls for Boiling Water Reactors,"

Generic letter 89-01, Supplement No. 1, NUREG-1302, April 1991.

O HNP ODCM, REV 7 2/7/92 l

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ML20106A650

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