Text

Rev 18 to ODCM for Hope Creek Generating Station
	ML20206G228
Person / Time
Site:	Hope Creek
Issue date:	03/31/1999
From:	Ciarlante V, Russell J, Salvatoriello; Public Service Enterprise Group
To:	;
Shared Package
ML20206G068	List: ML20206G065; ML20206G211; ML20206G228;
References
	PROC-990331, NUDOCS 9905070133
	Download: ML20206G228 (200)
	v • d • e

I OFFSITE DOSE CALCULATION MANUAL FOR PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK GENERATING STATION Revision 18 March 1999 Prepared by:

Vincent Ciarlante J

M a

r-ODCM Coordinator - Radiation Protection Support Reviewed by:

Jchn Russell dd~ \\h A Station Qualified Reviewer-Radiation Protection Support Approved by:

Gregory Salvatoriello Chemistry Superintendent (

I Date: MzJ/75 Mtg.#

9 9-023 R

ha an:

9905070133 990329 PDR ADOCK 05000354 P

PDR

Hope Creek ODCM Rev 18 Revision Summary:

PARTI Added PART I, Radiological Effluent Controls in preparation for implementation of NRC Generic Letter 89-01. Part I includes the Radiological Efiluent Controls that will replace the Radiological Effluent Technical Specifications upon NRC approval of the License Amendment.

PARTII Revised previous revision of ODCM 'to become PART II, Calculation Methodologies.

There are no technical or editorial changes made to the previous revision of the ODCM.

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Hope Creek ODCM Rev 18 TABLE OF CONTENTS INTRODUCTION....................................................................................................................,...........1 PARTI RADIOLOGICAL EFFLUENT CONTROLS I

SECTION 1.0 DEFINITIONS Il 0 SECTION 3.0 AND 4.0 CONTROLS AND SURVEILLANCE REQUIREMENTS I 3 /4 0-i I

3 /4.0 Applicability 13/4 0-1 3 /4.3 Instrumentation 3 /4.3.7 Monitoring Instrumentation 3 /4.3.7.10 Radioactive Liquid Emuent Monitoring Instrumentation I 3 /4 3-91 3 /4.3.7.11 Radioactive Gaseous Emuent Monitoring Instrumentation

13/4 3-96 3/4.11 Radioactive Emuents 3 /4.11.1 Liquid Emuents 3 /4.11.1.1 Concentration 13/4 11-1 3 /4.11.1.2 Dose 13/4 11-5 3 /4.11.1.3 Liquid Waste Treatment System I 3 /4 11-6 3 /4.11.2 Gaseous Emuents.

3 /4.11.2.1 Dose Rate I 3 /4 11-8 3 /4.11.2.2 Dose -Noble Gases 13/4 11-12 3 /4.11.2.3 Dose - Iodine-131, lodine-133, Tritium, and Radionuclides in Particulate Form I 3 /4 11-13 3/4.11.2.4 Gaseous Radwaste Treatment System I 3 /4 11-14 3 /4.11.2.5 Ventilation Exhaust Treatment System I 3 /4 11-15 3 /4.11.2.8 Venting or Purging I 3 /4 11-18 3 /4.11.4 Total Dose I 3 /4 11-20

Hope Creek ODCM Rev 18 TABLE OF CONTENTS 3 /4.12 Radiological Environmental Monitoring 3 /4.12.1 Monitoring Program I 3 /4 12-1 3 /4.12.2 Land Use Census I 3 /4 12-13 3 /4.12.3 Interlaboratory Comparison Program 13/4 12-14 BASES FOR SECTIONS 3.0 AND 4.0 1 B 3 I4 - 0 3 /4.3 Instrumentation-3 /4.3.7 Monitoring Instrumentation 3 /4.3.7.10 Radioactive Liquid Emuent Monitoring Instrumentation I B 3 /4 3-6 3 /4.3.7.11 Radioactive Gaseous Emuent Monitoring Instrumentation I B 3 /4 3-6 3 /4.11 Radioactive Emuents 3 /4.11.1 Liquid Emuents I B 3 /411-1 3 /4.11.1.1 Concentration I B 3 /411-1 3 /4.11.1.2 Dose I B 3 /411-1 3 /4.11.1.3 Liquid Waste Treatment System I B 3 /411-2 3 /4.11.2 Gaseous Emuents I B 3 /411-2 3 /4.11.2.1 Dose Rate I B 3 /411-2 3 /4.11.2.2 Dose -Noble Gases I B 3 /411-3 3 /4.11.2.3 Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form I B 3 /411-3 3 /4.11.2.4 Gaseous Radwaste Treatment System I B 3 /411-4 l

3 /4.11.2.5 Ventilation Exhaust Treatment System I B 3 /411-4 3 /4.11.2.8 Venting or Purging I B 3 /411-5 3 /4.11.4 Total Dose I B 3 /411-5 ii

Hope Creek ODCM Rev 18 TABLE OF CONTENTS 3 /4.12 Radiological Environmental Monitoring i B 3 /412-1 3 /4.12.1 Monitoring Program I B 3 /412-1 3 /4.12.2 Land Use Census I B 3 /412-1 3 /4.12.3 Interlaboratory Comparison Program I B 3 /412-2 SECTION 5.0 DESIGN FEATURES I5-1 5.1 Site I5-1 5.1.1 Map Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents -

15-I SECTION 6.0 ADMINISTRATIVE CONTROLS 6.9 Reporting Requirements I6-18 6.9.1.6 Annual Radiological Environmental Operating Report I6-18 6.9.1.7 Radioactive Effluent Release Report I 6-19 i

6.15 Major Changes to Radioactive Liquid, Gaseous, and Solid Waste Treatment Systems I 6-24 i

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Hope Creek ODCM Rev 18 TABLE OF CONTENTS PART II - CALCULATIONAL METHODOLOGIES................................................

U l.0 LIQUID EFFLUENTS 1.1

' Radiation Monitoring Instrumentation and Controls................................................................................................ 2 1.2 Liquid Effluent Monitor Setpoint Determination........................................................................................................ 2 1.2.1 Liquid Emuent Monitors.................................................................................. 3 j

1.2.2 Conservative Default Values..................................................................................... 4 1

1.3 Liquid Emuent Concentration Limits - 1 0 C FR 20..............................................................................................

1.4 Liquid Emuent Dose Calculations - 10 CFR 50.................................................................................................. 5 1.4.1 Member of the Public Dose -

L iq uid E m uents......................................................................................... 5 1.4.2 Simplified Liquid Emuent Dose Calculation......................................................................................... 6

)

i 1.5 Liquid Emuent Dose Projections.................................................................................... 7 1.6 Representative S amples...................................................................................................... 8 2.0 GASEOUS EFFLUENTS 1

2.1 Radiation Monitoring Instrumentation and Controls............................................................................................... 8 2.2 Gaseous Emuent Monitor Setpoint Determination..................................................................................................... 9 1

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2.2.1 Plant Vent and FRVS Monitors.............................................................................. 9 2.2.2 Conservative De fault Values................................................................................... 10 2.3 Gaseous Emuent Instantaneous Dose Rate Calculation - 10 CFR 20.................................................................................... I 1 i

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Hope Creek ODCM Rev 18 TABLE OF CONTENTS 2.3.1 Site Boundary Dose Rate -

Noble Gase s..............................................................

2.3.2 Site Boundary Dose Rate -

Radioiodine and Particulates...................................................................... 12 2.4 Noble Gas Effluent Dose Calculations - 10 CFR 50.................................................................................................... 13 2.4.1 UNRESTRICTED AREA Dose -

Noble Gases.................................................................................

2.4.2... Simplified Dose Calculation for Noble Gases...................................................................................

2.5 Radioiodine and Particulate Dose Calculations - 10 CFR 50.................................................................................................... 14 2.5.1 UNRESTRICTED AREA Dose -

Radioiodine and Particulates..................................................................... 14 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates.................................................................. 15 2.6 Gaseous Effluent Dose Projection....................................................................................... I 5 3.0 S P E CI AL DOS E AN ALY S I S........................................................................................................ I 3.1 Doses Due to Activities Inside the SITE BOUNDARY.................................................................................................... 16 3.2 Total Dose to MEMBERS OF THE PUBLIC -

40CFR190...................................................................................................,......................I7 3.2.1 Efiluent Dose Calcul ati on...................................................................................... 17 3.2.2 Direct Exposure Determination............................................................................... I 7 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........................................ I 8 4.1 S ampl ing Program...............................................................................................................

4.2 Interlaboratory Comparison Program................................................................................ 18 5.0 HCGS EXPLOSIVE GAS MONITORING PROGRAM.................................................................I 8 V

Hope Creek ODCM Rev 18 TABLE OF CONTENTS TABLES 1-1 Parameters for Liquid Alarm Setpoint Determine. tion..................................................................................................... 2 1 1-2 Site Related Ingestion Dose Co mmitment Factors, Aio.............................................................................................. 22 1-3 B ioaccum ulation Factors (B Fi)...........................................................................

2-1 Dose Factors for Noble Gases.....................................................................

2-2 Parameters forGaseous Alarm Setpoint Determinati on...................................................................................................... 2 8 2-3 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Cal culations.........................................................................

2-4 Pathway Dose Factors -

Atmospheric Releases..................................................................................................... 3 0 A-1 Calculation o f Effective MPC....................................................................................... A-6 B-1 Adult Dose Contributions Fish and inverte brate Pathway s.................................................................................................... B-5 C-1 E ffec tive Dose Fac tors...................................................................................

D-1 Infant Dose Contribution Fraction of Dose....................................................................D-3 D-2 Fraction of Dose Contributions by Pathway..............................................................D-3 E-1 S ample Locatio ns....................................................................................... E-3 F-1 Maximum Permissible Concentrations................................................................... F-2 APPENDICES Appendix A - Evaluation of Conservative, Default MPC Value for Liquid Em uents..................................................................... A-1 Appendix B - Technical Basis for Effective Dose Factors - Liquid Radioactive Em uents........................................................................ B-1 Appendix C - Technical Basis for Effective Dose Factors - Gaseous Radioactive Emuents.................................................................C-1 Appendix D - Technical Basis for Effective Dose Parameters - Gaseous Radioactive Effluents.................................................................D-1 Appendix E - Radiological Environmental Monitoring Program - Sample Type, Location and Analysis........................................................ E-1 Appendix F - Maximum Permissible Concentration (MPC)

Values for Liquid Emuents..........

...................................................................F-1 Appendix G _ Controls for Releases from the Circulating Water Dewatering Sump..........

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Hope Creek ODCM Rev 18 HOPE CREEK GENERATING STATION OFFSITE DOSE CALCULATION MANUAL INTRODUCTION The Hope Creek Offsite Dose Calculation Manual (ODCM) is a supporting document to the Hope Creek Technical Specifications. The previous Limiting Conditions for Operations that were contained in the Radiological Emuent Technical Specifications (RETS) are now included in the ODCM as Radiological Effluent Controls (REC). The ODCM contains two parts: Part I - Radiological Effluent Controls, and Part II - Calculational Methodologies.

PartIincludes the foliowing:

The Radiological Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification 6.8.4 Descriptions of the information tim should be included in the Annual e

Radiological Environmental Operating Report and the Annual Radioactive Effluent Release Report reqdred by Technical Specifications 6.9.1.6 and 6.9.1.7, respectively.

Part II describes methodologies and parameters used for:

The calculation of radioactive liquid and gaseous effluent monitoring instrumentation alarm / trip setpoints, The calculation of radioactive liquid and gaseous concentrations, dose e

rates, cumulative quarterly and yearly doses, and projected doses.

Part II also contains a list and graphical description of the specific sample locations for the radiological environmental monitoring program, and the liquid and gaseous waste treatment systems Until the Hope Creek Technical Specifications are revised to delete the Radiological Effluent Technical Specifications, the ODCM Part I will only be used for information. The Technical Specifications with LCOs and Surveillance Requirements will remain the controlling document. Part I of the ODCM will become the controlling document for Radiological Effluent Control upon NRC approval of LCR H99-0002.

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PART I RADIOLOGICAL EFFLUENT CONTROLS 1

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6 PART I RADIOLOGICAL EFFLUENT CONTROLS SECTION 1.0 DEFINITIONS

1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these CONTROLS may be achieved. The defined terms appear in capitalized type and are applicable throughout these CONTROLS.

ACTION 1.1' ACTION sha'll be that part of a CONTROL which prescribes remedial measures required under designated conditions.

h CHANNEL CALIBRATION 1.4 A CHANNEL CEIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CEIBRATION shall encompass the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CEIBRATION may be perfcrmed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

CHANNEL CHECK 1.5 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 indications and/or status derived from independent instruir.ent channels measuring the same parameter.

CHANNEL FJNCTIONE TEST 1.6 A CHANNEL PUNCTIONE TEST shall be:

.a.

Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b.

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

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

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DEFINITIONS l

CONTROL 1.10 The present Limiting Conditions for Operation (LCOs) that are contained in the Radio?.ogical Effluent Technical Specifications are being transferred to the Offsite Dose Calculation Manual (ODCM) and are being renamed CONTROLS.

This is to distinguish between those LCOs which are being retained in the Technical specifications and those LCOs or CONTROLS that ars being transferred to the ODCM.'

DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

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DEFINITIONS FREQUENCY NOTATION 1.17 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

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I-DEFINITIONS MEMBER (S) OF THE PUBLIC 1.24 MEMBER (S) OF THE PUBLIC shall include all persons who are not f

occupationally associated with the plant. This category does not include l

employees of the utility, it contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

OFF-GAS RADWASTE TREA'IMENT SYSTEM (GASEOUS RADWASTE TREATMENT SYSTEM) 1.26 An OFF-GAS RADWASTE TREA'INENT SYSTEM (GASEOUS RADWASTE TREATMD'T SYSTEM)is any system designed and installed to reduce radioactive gaseous effluents by collecting reactor coolant system offgases from the main condenser evacuation system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

OFFSITE DOSE CALCULATION MANUAL 1.27 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip satpoints, and in the conduct of the radiological environmental monitoring program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification section 6.8.4 and (2) descriptions of the information that should be in the Annual Radiological Environmental Operating, Report and the Annual Radioactive Effluent Release Report required by Technical Specification sections 6.9.1.6 and 6.9.1.7.

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a DEFINITIONS OPER[J'.,P - OPERABILITY P

{

1.28 A rystem, subsystem, train, component or device shall be OPERABLE or ti. ave OPERABILITY when it is capable of performing its specified function (s) and when all necessary attendant instrumentation, controls, electrical power, 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 function (s).

OPERATIONAL CONDITION - CONDITION 1.29 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

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DEFINITIONS l

PURGE - PURGING l

1.34 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.35 RATED THERMAL POWER shall be a total reactor core heat transfer rate to l

the reactor coolant of 3293 MWT.

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REPORTABLE EVENT l

l 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 1

50.73 to 10 CFR Part 50.

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i DEFINITIONS

==================...................,,,,,,,,,,,,,,,==========..=............

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l SITE BOUNDARY i

1.42 The SITE BOUNDARY shall be that line beyond which the land is neither i

owned, nor leased, nor otherwise controlled, by the licensee.

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DEFINITIONS i'

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SOURCE CHECK 1.43 A SOURCE CHECK shall be the qualitative assessment of channel

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response when the channel sensor is exposed to a source of increased radioactivity.

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THERMAL POWER l

1.47 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

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DEFINITIONS j

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i UNRESTRICTED AREA 1.50 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY l

access to which is not controlled by the licensee for purposes of l

protection of individuals from exposure to radiation and radioactive j

materials, or any area within the SITE BOUNDARY used for residential i

quarters or for industrial, commercial, institutional, and/or l

recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.51 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents.

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.52 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

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I TABLE 1.1 SURVEILLANCE FREQUENCY NOTATION NOTATION FREQUENCY S

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l W

At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

l SA At least once per 184 days.

A At least once per 366 days.

R At least once per 18 months (550 days).

S/U Prior to each reactor startup.

P Prior to each radioactive release.

2 During startup, prior to exceeding 30% of I

RATED THERMAL POWER, if not performed within the previous 7 days l

N.A.

Not applicable.

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TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE 1.

POWER OPERATION Run Any temperature 2.

STARTUP Startup/ Hot Standby Any temperature b

3.

HOT SHUTDOWN Shutdown "*

> 2 00*F b

4.

COLD SHUTDOWN Shutdown "' *"

s 200*F*

l 5.

REFUELING *~

Shutdown or Refuel"

S 14 0'F I

"The reactor mode switch may be placed in the Run, Startup/ Hot Standby, or Refuel position to test the switch interlock functions and related instrumentation provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.

If the reactor mode switch is placed in the Refuel position, the one-rod-out interlock shall be OPERABLE.

"The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per Technical Specification 3.9.10.1.

" Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

"See Special Test Exceptions Technical Specification sections 3.10.1 and 3.10.3.

"'The reactor mode switch may be placed in the Refuel position while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE,

3ee Special Test Exception Technical Specification section 3.10.8.

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PART I RADIOLOGICAL EFFLUENT CONTROLS SECTIONS 3.0 and 4.0 CONTROLS AND SURVEILLANCE REQUIREMENTS I 3/4 0-1

3/4.0 APPLICABILITY CONTROLS 3.0.1 Compliance with the CONTROLS contained in the succeeding CONTROLS is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the CONTROLS, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a CONTROL shall 6xist when the requirements of the CONTROL and associated AQTION requirements are not met within the specified time intervals.

If the CONTROL is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required.

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3.0.3 When a CONTROL is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in I

an OPERATIONAL CONDITION in which the Specification does not apply by placing it, as applicable, in:

1.

At least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 3.

At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the CONTROL.

Exceptions to these requirements are stated in the individual CONTROLS.

This CONTROL is not applicable in OPERATIONAL CONDITIONS 4 or 5.

3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made when the conditions for the CONTROLS are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval.

Entry into an OPERATIONAL CONDITION or other specified condition may be made in accordance with the ACTION requirements when conformance to them permits continued operation of the facility for an i

unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements.

Exceptions to these requirements are stated in the individual CONTROLS.

3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to

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perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment.

This is an exception to CONTROL 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

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APPLICABILITY 90RVEffTANr'P DFOUTREMPMT9 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual CONTROLS unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within its specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

4.0.3 Failure to perform a Surveillance Requirc;aent within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute a failure to meet the OPERABILITY requirements for a CONTROLS. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance

]

when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Surveillance requirements do not have to be performed on inoperable j

equipment.

1 4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condition shall not be made unless the Surveillance Requirement (s) associated with the CONTROLS have been performed within the applicable surveillance interval or as otherwise specified.

This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements.

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l 3/ 4.3 INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION r

COMTROLS i

3.3.7.10 In accordance with Hope Creek Technical Specification 6.8.4.g.1),

the radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3.7.10-1 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of CONTROL 3.11.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).-

APPLICABILITY: At all times.

ACTION:

1 With a radioactive liquid effluent m,onitoring instrumentation a.

channel Alarm / Trip Setpoint less conservative than required by the above CONTROL, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b.

With less than the minimum number of radioactive liquid effluent-monitoring instrumentation. channels OPERABLE, take the ACTION shown in Table-3.3.7.10-1.

Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in 1

the next Radioactive Effluent Release Report pursuant to CONTROL 6.9.1.7 why this inoperability was not corrected in a timely manner, j

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The provisions of CONTROL 3.0.3 are not applicable.

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SURVEILLANCE REQUIREMENTS 4.3.7.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE j

CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3.7.10-1.

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N O

I T

N A

O T

I 0

1 2

2 N

T 1

1 1

1 E

C 1

1 1

1 M

A U

RTSN I

SE MLL G

UEB N

MNA 1

1 1

1 I

I NR 1

R NAE O

I HP 0

T MCO 1

I 2

N 9

7 O

M 3

3 g

T n

4 3

N i

/

E l

3 E

U o

T g

L L

o O

n I

B F

C NE i

A F

S l

T E

D e

TA o

N h

UE o

D A

t BL C

I E

U M

o MR o

Q R

t R

t I

A AF L

L e

LO e

AE n

A -

t n

E S

i N

n i

V GA L

GO e

L I

NE NI u

r T

I L e

I T l

e i

C DE g

DA f

S g

e A

I R r

I N f

E r

W O

V a

VI E

C a

I OF h

OM I

h n

D RO c

RR n

V c

w A

P se PE w

E se o

R T

N i n T

o D

i n d

N SO Di S

d Di w

E RI L

RC w

T L

o M

OT e

OI o

N e

l U

TA t n TT l

E t n B

R I N s w.

I A B

M sw T

NI ao NM E

ao r

S OM wd OO r

R wd e

N MR d w MT e

U d w w

I E

ao U

w S

ao o

YT Rl YA o

A Rl T

T B

T T

E B

I C d

I G M

d g

VI i r VN g

i r n

I T ue I I n

E ue i

TA qw TD i

T qw l

CM i o CI l

A i o o

AO LT AV o

R LT o

OT OO o

C I U I R C

W DA DP O

A A

L R

a R

a F

a b

1 2

3 e

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TABLE 3.3.7.10-1 (Continued)

[

TABLE NOTATION ACTION 110 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases l

via this pathway may continue provided that prior to initiating

'a release:

a.

At least two independent samples are analyzed in accordance with CONTROL 4.11.1.1.2, and b.

At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 111 -

With the number of channels OPERABLE less than required by I

the Minimum Channels OPERABLE requirement, effluent releases l

via this pathway may continue provided that, at least once per l

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gross radioactivity at a limit of detection of at least 10" microcuries/ml. Otherwise, suspend release of radioactive effluents via this pathway.

ISTION 112 -

With the number of channels OPERABLE less than required by j

the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

Pump performance curves generated in place may be used to estimate flow.

1 l

l l

I 1

I I 3/4 3-93

NO I

T A

TN L

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N T

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T A

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TABLE 4.3.7.10-1 (Continued)

TABLE NOTATIONS (1)

The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of.this pathway and control room alarm annunciation occur if any of the following conditions exists:

Instrument indicates measured levels.above the Alarm / Trip a.

Setpoint, or b.

Circuit failure, or c.

Instrument indicates a downscale failure.

(2)

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

Instrument indicates measured levels.above the Alarm a.

Setpoint, or b.

Circuit failure, or c.

Instrument indicates a downscale failure.

(3)

The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS)/ National Institute of Standards and Testing (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS/NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range.

For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration or are NBS/NIST traceable shall be used.

(4)

CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.

l I 3/4 3-95

INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION CONTROLS 3.3.7.11 In accordance with Hope Creek Technical Specification section 6.8.4.g.1), the radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3.7.11-1 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of CONTROLS 3.11.2.1 and 3.11.2.6 are not exceeded.

The Alarm / Trip Setpoints of these channels meeting CONTROLS 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: As shown in Table 3.3.7.11-1.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above CONTROL, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable,.or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3.7.11-1.

Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report pursuant to CONTROL 6.9.1.7 why this inoperability was not corrected in a timely manner.

c. The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3.7.11-1.

W I 3/4 3-96

NO 3

5 5

2 2

3 5

5 2

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T 2

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m t

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TABLE 3.3.7.11-1 tContinued)

TABLE NOTATION At all times.

- During' operation o'f the main condenser air ejector.

ACTION 122 - With the number of channels OPERABLE less than required by the Minimum Channels O?ERABLE requirement, effluent releases via this pathway may continue,provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Otherwise, suspend release of radioactive effluent's via this pathway.

ACTION 123 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases-via this pathway may' continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 124 - DELETED ACTION 125 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months samples are continuously collected with auxiliary sampling equipment as required in Table 4.11.2.1.2-1.

e 6

I 3/4 3-99

TABLE 3.3.7.11-1 (Continued)

TABLE NOTATION At all times.

- During operation of the main condenser air ejector.

ACTION 122 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 123 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may' continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 124 - DELETED ACTION 125 -

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months samples are continuously collected with auxiliary sampling equipment as required in Table 4.11.2.1.2-1.

I 3/4 3-99

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5 TABLE 4.3.7.11-1 (Continued)

TABLE NOTATION At all times.

- During operation of the main condenser air. ejector.

(1)

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

1.

Instrument indicates measured levels above the alarm setpoint.

2.

Circuit failure.

3.

Instrument indicates a downscale failure.

(2)

The initial CRANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS)/ National Institute of Standards and Testing (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS/NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range.

For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration or are NBS/NIST traceable shall be used.

I 3/4 3-102

3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION CONTROLS maammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmma 3.11.1.1 In accordance with Hope Creek Technical Specifications 6.8.4.g.2) and 3), the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1.1-1) chall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolygd or entrained noble gases, the concentration shall be limited to 2 x 10 microcuries/ml total activity.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS maammmm mmmmmmmmmmmmmmmmmmmm m mmmmmmmm mmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmma 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11.1.1.1-1.

4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of CONTROL 3.11.1.1.

W I 3/4 11-1

TABLE 4.11.1.1.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum Type of of Detection Liquid Release Sampling Analysis Activity (LLD)*

Type Frequency Frequency Analysis (pCi/ml)

A. Batch Waste P

P Principal Release

Each Batch Each Batch Gamma 5x10

Sample Emitters

Tanks I-131 1x10

P M

Dissolved and 1x10'*

One Batch /M Entrained Gases (Gamma Emitters)

P M

H-3 1x10

d Each Batch Composite Gross Alpha 1x10

P Q

Sr-89, Sr-90 5x10

d Each Batch Composite Fe-55 1x10

B.

Continuous M

Principal Gamma 5x10" Releases' Composite' Emitters

Station Service NA Water System I-131 1x10

(SSWS) (If contaminated W

M Dissolved and 1x10

as indicated Grab Sample Entrained Gases by SACS or (Gamma Emitters)

RACS system.

i NA M

H-3 1x10

d Composite Gross Alpha 1x10

NA Q

Sr-89, Sr-90 5x10

Composite' i

Fe-55 1x10

1 I 3/4 11-2 i

i TABLE 4.11.1.1.1-1 (Continued)

TABLE NOTATION

'The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 s, LLD =

E

V e 2.22 x 10'

Y

exp (-A A t)

Where:

LLD is the "a priori" lower limit of detection as defined above, as v.icrecuries per unit mass or volume, is the standa.d deviation of the background counting rate or sb or the counting rate of a blank sample as appropriate, as counts per minute,

)

E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 10' is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, 1 is the radioactive decay constant for the particular radionuclide (sec), and at for plant effluents in the elapsed time between the midpcint of sample collection and time of counting (sec).

Typical values of E, V, Y, and at should be used in the calculation.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

"A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated,' and then thoroughly mixed by a method described in the ODCM to assure representative sampling.

I 3/4 11-3

TABLE 4.11.1.1.1-1 (Continued)

TABLE NOTATION

The principal gamma emitters for which the LLD specification applies exclusively are: Mn-54, Fe-59, Cc-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137,.agd Ce-141. Ce 144 shall also be measured, but with an LLD of 5 x 10 This does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to CONTROL 6.9.1.7 d A Composite aaNple is one in which the quantity of liquid samples is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g;, from a volume of a system that has an input flow during the continuous release, j

l I 3/4 11-4

RADIOACTIVE EFFLUENTS DOSE CONTROL

...................................................................c...........

3.11.1.2 In acebrdance with Hope Creek. Technical Specifications 6.8.4.g.4) and 5), the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each reactor unit, to UNRESTRICTED AREAS (see Figure 5.1.1-1) shall be limited:

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

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

APPLICABILITY:

At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits,

b. The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

I 3/4 11-5

RADIOACTIVE EFFLUENTS LIQUID WASTE TREATMENT SYSTEM CONTROLS

................................................................s 3.11.1.3InaccordancewithHopeCreekTechnicalSpecifications6.8.4.6),

the liquid radwaste treatment system shall be OPERABLE and appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected' doses due to the liquid effluent, from each :;eactor unit, to UNRESTRICTED AREAS (see Figure 5.1.1-1) would exceed 0.06 mrem to the total body or 0.2 mrom to any organ in any 31-day period.

APPLICABILITY: At all times.

ACTION:

l

a. With radioactive liquid waste being discharged and in excess of

{

the above limits and any portion of the liquid radwaste treatment system not in operation, prepare and submit to the commission within 30 days pursuant to Technical Specification 6.9.2 a Special Report that includes the following information:

1.

Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability, 2.

Action (s) taken to restore the inoperable equipment to OPERABLE status, and 3.

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

b. The provisions of CONTROLS 3.0.3 are not applicable.

l SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases from each reactor unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM.

4.11.1.3.2 The installed liquid radwaste treatment system shall be demonstrated OPERABLE by meeting CONTROLS 3.11.1.1 and 3.11.1.2.

I 3/4 11-6

l RADIOACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE CONTROLS 3.11.2.1 In accordance with Hope Creek Technical Specifications 6.8.4.g.3) and 7), the dose rate due to radioactive materials released in gaseous effluents from the site to, areas at and beyond the SITE BOUNDARY (see Figure 5.1.1-1) shall be listed to the following:

a.

For noble gases: Less than or equal to 500 mroms/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and i

b.

For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.

APPLICABILITY:

At all times.

ACTION:

With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s).

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM.

4.11.2.1.2 The dose rates due to iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11.2.1.2-1.

I 3/4 11-8

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pq i

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a a

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r r

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C C

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tt e

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ee yA T

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n e

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TABLE 4.11.2.1.2-1 (Continued)

TABLE NOTATION (a)The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a not count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

i I

4.66 s.

LLD =

E

V

2.22 x 10'

Y

exp (-A At)

Where:

6 LLD is the "a priori" lower limit of detection as defined above, as microcuries per unit mass or volume, s, is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 10' is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, 1 is the radioact{3e decay constant for the particular radionuclide (sec

),

and i

at for plant effluents is the elapsed time between the midpoint of sample collection and time of counting (sec).

Typical values of E, V, Y, and A t should be used in the calculation.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a

]

particular measurement.

I 3/4 11-10

TABLE 4.11.2.1.2-1 (Continued)

TABLE NOTATIONS (b)

The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141 and Ce-144 in iodine and particulete releases. This list does not mean that only these nuclides are te be considered. Other gamma peal:s that are identifiable, together with thcae of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to. CONTROL

6. 9.1. 7.,

(c)

Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor l

shows that effluent activity has not increased more than a factor of 3.

(d)

Tritium grab samples shall be taken at least once per 7 days from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

(e)

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with CONTROL 3.11.2.1, 3.11.2.2, and j

3.11.2.3.

(f)

Samples shall be changed at least once per 7 days 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 also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 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 LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

I 3/4 11-11

E RADIOACTIVE EFFLUENTS DOSE - NOBLE GASES CONTROLS 3.11.2.2 In accordance with Hope Creek Technical Specification 6.8.4.g.5).and s), the air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1.1-1) shall be limited to the following:

a.

During any calendar quarter: Less than or equal to 5 mrads for gasuna radiation and less than or equal to 10 mrads for beta radiation and, b.

During any calendar years Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

APPLICABILITY:

At all times.

ACTION With the calculated air dose from radioactive noble gases in a.

gaseous effluents exceeding any of the above limits, prepare and submit to the commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

J l

i b.

The provisions of CONTROL 3.0.3 are not applicable.

i i

SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and j

current calendar year for noble gases shall be determined in accordance with.

the methodology and parameters in the ODCM at least once per 31 days, i

l 9

I 3/4 11-12 I

1 RADIOACTIVE EFFLUENTS DOSE - IODINE-131, IODINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM l

CONTROLS J

3.11.2.3 In accordance with Hope Creek Technical Specifications 6.8.4.g.5) and 9), the dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1.1-1) shall be limited to the following:

a.

During any calendar quarter: Less than or equal to 7.5 mrems to any organ and, b.

During any calendar year: Less than or equal to 15 mrems to my organ.

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of iodine-131, a.

iodine-133, tritium, and 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 Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits, b.

The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

I 3/4 11-13

RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT SYSTEM CONTROLS 3.11.2.4 The GASEOUS RADWASTE TREA'INENT SYSTEM shall be in operation.

APPLICABILITY: Whenever the main condenser steam jet air ejector system is in operation.

ACTION:

With gaseous radwaste from the main condenser air ejector system a.

being discharged without treatment for more than 7 days, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Spedal Report that includes the following information:

1.

Identification of the inoperable equipment or subsystems and the reason for the inoperability, 2.

Action (s) taken to restore the inoperable equipment to OPERABLE status, and 3.

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

The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

..............................................................=........===....

4.11.2.4 The readings of the relevant instruments shall be checked every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the main condenser air ejector is in use to ensure that the gaseous radwaste treatment system is functioning.

1 4

(

D I 3/4 11-14

l RADIOACTIVE EFFLUENTS VENTILATION EXHAUST TREA'! MENT SYSTEM CONTROLS 3.11.2.5 In accordance with Hope Creek Technical Specifications 6.8.4.g.6),

the VENTILATION EXHAUST TREA'INENT SYSTEM for-the Reactor Building and the Service and Radwaste Building shall be OPERABLE and appropriate portions of this system'shall be used to reduce release of radioactivity when the projected doses in 31 days due to gaseous effluent releases from each unit to areas at and beyond the SITE BOUNDARY (see Figure 5.1.1-1) would exceed:

a.

0.2 mrad to air from gamma radiation, or i

b.

0.4 mrad to air from beta radiation, or c.

0.3 mrem to any, organ of a MEMBER OF THE PUBLIC j

APPLICABILITY: At all times.

i I

ACTION:

4 j

With radioactive ventilation exhaust being discharged without a.

treatment and in excess of the above limits, prepare and submit to i

the comission within 30 days pursuant to Technical Specification 1

6.9.2 a'Special Report that includes the following information:

1.

Identification of any inoperable equipment or subsystems, and the reason for the inoperability, 2.

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

3.

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

I b.

The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5.1 Doses due to gaseous releases from each unit to areas at and beyond

'the SITE BOUNDARY shall'be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM, when the VENTILATION EXHAUST TREA'IMENT SYSTEM is, not being fully utilized.

4.11.2.5.2 The installed VELTILATION EXHAUST TREATMENT SYSTEM shall be considered OPERABLE by. meeting CONTROLS 3.11.2.1 and 3.11.2.2 and 3.11.2.3.

I 3/4 11-15

i RADIOACTIVE EFFLUENTS VENTING OR PURGING CONTROLS 3.11.2.8 VENTIbiG or PURGING of the Mark I containment drywell shall be through either the reactor building vgntilation system or the filtration, recirculation and ventilation system.

APPLICABILITY: Whenever,the containment is vented or purged.

ACTION:

With the requirements of the above specification not satisfied, a.

suspend all VENTING and PURGING of the drywell, b.

The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.8'The containment shall be determined to be a*Aigned for VENTING or PURGING through either the reactor building ventilation system, the filtration, recirculation and ventilation system, or the hardened torus vent within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING or PURGING of the drywell.

4

Following Type A Integrated Leakage Rate Testing, the Mark I containment drywell may be vented through the hardened torus vent.

I 3/4 11-18

i RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE CONTROLS 3.11.4 In accordance with Hope Creek Technical Specification 6.8.4.g.11, the annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mroms.

APPLICABILITY: At all times.

ACTION:

a.

With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of CONTROLS 3.11.'1.2a.,

3.11.1.2b., 3.11.2.2a., 3.11.2.2b.,

3.11.2.3a., or 3.11.2.3b., calculations should be made including direct radiation contributions from the units including outside storage tanks, etc. to determine whether the above limits of CONTROLS 3.11.4 have been exceeded.

If such'is the case, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent' releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits.

This Special Report, as defined in 10 CFR 20.405c, shall include an analysis that estimates 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 or concentrations.

If the estimated dose (s) exceeds the above limits, and if the release condition resulting 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 request, and a variance is granted until staff action.on the request is complete.

l b.

_The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS i

4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with CONTROLS 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM.

4.11.4.2 Cumulative dose contributions from direct radiation from the units including outside storage tanks, etc. shall be determined in accordance with the methodology and parameters in the ODCM.

This requirement is applicable only under conditions set forth in CONTROL 3.11.4, ACTION a.

I 3/4 11-20 p

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM CouTROLs 3.12.1 In accordance with, Hope Creek Technical Specification 6.8.4.h.1), the radiological environmental acnitoring program shall be conducted as specified in. Table 3.12.1-1.

APPLICABILITY: At all times.

ACTION:

,i With the radiological envirewunantal monitoring program not being a.

conducted as specified in Table 3.12.1-1, prepare and submit to the Comunission, in the Annual Radiological Environmental Operating Report required by Technical Specification 6.9.1.6, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence, b.

With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12.1-2 when averaged over any calendar quarter, prepare and submit to the commission within 30 days, pursuant to Technical Specification 6.9.2, a special Report that identifies the cause(s) for exceeding the 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 of CONTROLS 3.11.1.2, 3.11.2.2, and 3.11.2.3. When more than one of the radionuclides in Table 3.12.1-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2)

+

...> 1.0 reporting level (1) reporting level (2)

When radionuclides other than those in Table 3.12-2 are detected i

and are the result.of plant effluents, this report shall be i

submitted if the potential annual dose

to A MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year

-limits of CONTROLS 3.11.1.2, 3.11.2.2, and 3.11.2.3. 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 reported and described in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.6.

c.

With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12.1-1, identify specific locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days.

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

I 3/4 12-1

6 RADIOLOGICAL ENVIRONMENTAL MONITORING CONTROLS (Continued) sa m mm mmmmmmmmmm m m m m mm mm m m m m mmm mmmmm mm un g am m mmm m m m mmmm mm m m m m m m m m m m m mm m m m m m m ACTION:

(Continued)

The specific locations from which samples were unavailable may then be deleted from the monitoring program.

Pursuant to CONTROL 6.9.1.8, identify the cause of the unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Radioactive Effluent Release Report pursuant to CONTROL 6.9.1.8 and also include in.the report a revised figure (s) and table for the ODCM reflecting the,new location (s).

d.

The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS maammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmma 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12.1-1 from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12.1-1 and the detection capabilities required by Table 4.12.1-1.

I 3/4 12-2

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r r

o ooo mnn ec8 ce ol O

v i

f at a r

rrr oif l a l p r

I i

t oeah f

ff f mivI peoa fl D

td a nh c ot a m

t cm o

A anc e

t s e

eee rah.an e

er R

f t ao l e i

l l l l f c l si5 er l t on L

ph sd pn ppp oea rm pn es mt e mo mmm sl ciI sna mot rsee a

i s ai aaa e

nt l e 1

acn eel l sel t st sss l ean aes s

a b rpp epi a

pet enmwen at mpmm erp ec eee mrst eit sa e

s ueaa nh uy no nnn ahioh neoh nti NRSSOt sb Ol OOo Stdpt abdt Oad yae wl g

t h p n

n t m i

e N

aa k

m O

PS n

i Ik i

d Tl er r

e Si ro D

S EM u/

G sd N

on c

d I a pa x

E 4

y s

c n

n n

o o

es i

c i

ui c

t i

t qs i

r p

r ey p

o o

o rl 6

p t

p Fa n t '"

o' s"s l o

e e

dA ssl i

n iib ib af si asi o

a yd n yd u

e nl e ul e u

p n a aa y

ann Gnn T

Gao ao y'

ry oe c

t n

,h t

s e

n l

e du of a v

nq sin r

ae a

o a

r eys h

gF sl a n

l e f

i n nas o

l o i u pi nt e

mt eno m

ac l an i

S e pi t

M l

mme A

l aer t

R o

S sa A

G C

O RP 6

)

d G

2 1

e N

y yr u

I l

ne 4

n R

l nt l at

/

i O

att i n a

as 3

t T

i nn a

pmwe n

I caa sl i o t

I o

N rtl ep crys C

O erp i

nfb a M

mo cy i

w

(

mpf eb rsd 1

L omo p

pt et.

A ci sd ct nd 1

T h yt ec cd gl g y

e huaae N

"s 2

E cli mn aoipr 1

M al n ae err a

N n

eai su prdh 3

O e

o nca l

f iic R

v i

f oie f f od us E

I i

t oivr on o s qi L

V tda t

a i

eoiid B

N anc ean e

l f l

A E

f t ao l eie l t e p

t n

T on L

pr g

pog mf ahe L

es mcsr mnr aoh ce A

rsee aeea a

a s

tib C

eel l srih ssh s

h I

b rpp cc ac sawe G

mpmm ed es ees eae v

O ueaa n n pi nri nl rna L

NRSS Oasd Oad Ocaih O

)

I d

D e

A u

s R

n e

i t

t a

n r

y o

b s

ae C

e t

wl dt c

(

h p nr u

t m N

ae d

aa O

v o

P S_ I hn dr T

sI oP er S

i o

ro E

F F

u/

G sd N

on I

b c

pa x

E 4

0

l TABLE 3.12.1-1 (Continued)

TABLE NOTATIONS (1)

Specific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be provided for each and every sample location in Table 3.12.1-1 in a table i

and figure (s) in the ODCM. Refer to NUREG-0133, " Preparation of i

Radiological Effluent Technical Specifications for Nuclear Power Plants,"

i October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due l

to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.6.

It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media I

and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program given in the ODCM. Pursuant to CONTROL 6.9.1.7, submit in the next Radioactive Effluent Release Report l

documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflecting the new location (s) with supporting information identifying the cause of the unavailability of samples for I

that pathway and justifying the selection of the new location (s) for obtaining samples.

(2)

One or more instruments, such as a pressurized ion chamber, for measuring l

and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.

Film badges shall not be used as dosimeters for measuring direct radiation. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

(3)

Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

(4)

Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

(5)

The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an l

area beyond but near the mixing zone.

" Upstream" samples in an estuary must be taken far enough upstream to be beyond the plant influence.

Salt I

water shall be sampled only when the receiving water is utilized for recreational activities.

I 3/4 12-7 L

TABLE 3.12.1-1 (Continued) i TABLE NOTATION l

(6) A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which.the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short relative to the compositing period in order to assure obtaining a representative sample.

(7)

Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulie gradient or recharge properties are suitable for contamination.

-( 8 )

The dose shall be calculated for the maximum organ and age l

group, using the methodology and. parameters in the ODCM.

(9)

If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.

l l

1 l

j, I 3/4 12-8

s t )t e

c u

e u w 0

0 0

a l

d 0

0 0

v o

1 0

r g, 0,

a P k 1

2

/

d i

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F (p s

S i

E x

L e

P M

y A

a S

wh L

k )1 t

3 0

0 a

A l /

6 7

0 p

R i

i 3

M C r

I E

p e

(

t tO a

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I V

g N

n E

ik N

n I

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d

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t N

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0 0

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0 0

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F 0,

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a T O I

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n n

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oa f

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7 5

6 I

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)

1 r

r/

4 0

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5 8

5 e

ei 0

1 3

1 1

t tc 0

a ap 3

w W(

gn i

kn ir a

d t

0 0

s e

6 5

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i b

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4 7

f y

s 4

9 8

5 b

1 3

3 a

l s

5 5

5 6

N 3

1 1

L D

a o

3 1

L n

r n

e o

n r

s s

a L

A g

H M

F C

Z Z

I C

C B

. i

,i.

TABLE 4.12.1-1 (Continued)

TABLE NOTATIONS (1)

This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual. Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.6.

(2) Required detection capabilities for thermoluminescent dosimeters used for environmental mea,surements shall be in accordance with the recommendations of Regulatory Guide 4.13.

(3) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a not count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 s.

LLD =

E

V

2.22

Y

exp(- 1 At)

Where LLD is the "a priori" lower limit of detection as defined above, as picoeuries per unit mass or volume, s, is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute, i

E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2 22 is the number of disintegrations per minute per picoeurie, Y is the fractional radiochemical yield, when applicable, 1 is the radioactive decay constant for the particular radionuclide (sec'*), and At for environmental samples is the elapsed time between sample collection, or end of the sample collectibn period, and time of counting (sec)

Typical valves of E, V, Y, and A t should be used in the calculation.

I 3/4 12-11

TABLE 4.12.1-1,' Continued)

TABLE NO7/QJONS It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (af ter the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.

Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may rander these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Enviranmental Operating Report pursuant to CONTROL 6.9.1.6.

i l

I 3/4 12-12

g

- RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS CONTROLS 3.12.2 In accordance with Hope Creek Technical Specifications 6.8.4.h.2), a land use census shall be conducted and shall identify within a distance of 8 km (5 miles).the location in each of the is meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden

of greater than 50 m (500 ft') producing broad leaf 8

vegetation.

- APPLICABILITY: At all times.

)

ACTION:

With a land use census identifying a location (s)- that yields a a.

calculated dose or dose commitment greater than the values currently being calculated in CONTROL 4.11.2.3, identify the new location (s) in the next Radioactive Effluent Release Report, pursuant to CONTROL 6.9.1.7.

b.

With a land use census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with CONTROL 3.12.1, add the new location (s) to the radiological environmental monitoring program within 30 days. The sampling location (s), excluding the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

Pursuant to CONTROL 6.9.1.7, identify the new location (s) in the next Radioactive Effluent Release Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new location (s).

c.

The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.1 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, visual survey, aerial survey, or by consulting local agriculture cuthorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.6.

Broad leaf. vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census.

Specifications for broad leaf vegetation sampling in Table 3.12.1-1, Part 4.c.,

shall be followed, including analysis of control samples.

I 3/4 12-13 1

L.

RADIOLOGICAL ENVIRO!OMNTAL MONITORING 3/4.12.3 INTERIABORATORY COMPARISON PROGRAM CONTROL 3.12.3 In accordance wit'h Hope Creek Technical Specifications 6.8.4.h.3), analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Conunission.

APPLICABILITY: At all times.

ACTION:

Nith analyses not being performed as required above, report the a.

corrective actions taken to prevent a recurrence to the Conunission in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.6.

b.

The provisions of CONTROL 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM. A summary of the results obtained as part of the above required Inter-laboratory Comparison Program chall be included in the Annual Radiological Environmental Operating Report pursuant to CONTROL 6.9.1.6.

l l

l 1

e4 I 3/4 12-14

l I

i 1

PART I RADIOLOGICAL EFFLUENT CONTROLS j

BASES FOR SECTIONS 3.0 AND 4.0 CONTROLS AND SURVEILIANCE REQUIREMENTS NOTE The BASES contained in the succeeding pages summarize the reasons for the CONTROLS in Section 3.0 and 4.0, but are not part of these CONTROLS

t INSTRUMENTATION BASES j

3/4.3.7.10 RADIOACTIVE LIQUID EFFLUENT MONI'!ORING INSTRUMENTATION l

The radioactive liquid effluent luonitoring 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 points for these instruments shall be calculated and adjusted in I

accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the lic.its of 10 CFR Part 20.

The i

OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3/4.3.7.11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INST _R_UMENTATION The radioactive gaseous effluent monitoring instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous affluents during actual or potential releases of gaseous affluents.

The alarm / trip setpoints for these instruments shall be calculated and J

adjusted in accordance with the methodology and parameters in the ODCM.

This will ensure the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

This instrumentation also includes provisions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the main condenser offgas treatment system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR part 50.

l i

I B 3/4 3-6 I

3/4.11 RADIOACTIVE EFFLUENTS RASES 3/4.11.1 LIOtED EFFLUENTS 3/4.11.1.1 CONCENTRATION This CONTROL is provided to ensure that the concentration of radioactive materials released in liquid wasta effluents to UNRESTRICTED AREAS will be, less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in IBGLESTRICTED AREAS will result in exp'sures within (1) the Section II.A design objectives of Appendix o

I,10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limito of 10 CFR Part 20.106 (e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs).

Detailed discussion of the LLD, and other detection limits can be found in Currie, L.

A., " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"

NUREG/CR-4007 (September 1984), and in the NASL Procedures Manual, NASL-300 (revised annually).

3/4.11.1.2 DOSE This specification 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 l

provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actum1 exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are 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 1.113,

" Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

I B 3/4 11-1 4

j RADIOACTIVE EFFLUENTS BASES 3/4.11.1.3 LIQUID RADWASTE TREA'INENT SYSTEM The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to their release to the environment. The requirement 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 "as low as is reasonably achievable." This specification implements the requir==ents of General Design criterion 60 of S pand4w A to 10 CFR Part 50 and the de' sign objective given in Section II.D of AppanM v 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 dore deslign objectives set forth in Section II. A of Appendix I, 10 CFR Part 50, for liquid effluents.

3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This CONTROL is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS.

The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODOt. The specified release rate limits restrict, at all times, the corresponding gansna and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the total body or to less than or equal to 3000 mrems/ year to the skin.

I B 3/4 11-2

1

)

{

RADIOACTIVE EFFLUENTS BASES DOSE RATE (Continued)

These release fate 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 mrams/ year.

l l

'; a required detection capabilitism for radioactive materials in gaseous waste samples are tabulated in termr of the icwer limits of detection (LLDs).

Detailed discussion of the LLD,-r.ad other detection limits can be found in Currie, L.

A., " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements "

NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

3/4.11.2.2 DOSE - NOBLE GASES This CONTROL is provided to implement the requirements of Sections II.B.

III.A, and IV.A of Appendix I, 10 CFR Fart 50. The CONTROL implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide j

the required operating flexibility and at the same time implement the guides i

set forth in Section IV.A of Appendix I to assure that the release-of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept l

"as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I 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 MEMBER OF THE PUBLIC through l

appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in l

gaseous effluents are 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 l

for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The l

ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

3/4.11.2.3 DOSE - IODINE-131, IODINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM This CONTROL is provided to implement the requirements of Sections II.C, l

III.A and IV.A of Appendix I, 10 CFR Part 50. The CONTROLS are the guides set l

forth in Section II.C of Appendix I. The ACTION statements provide the l

required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as g',

is reasonably achievable." The ODCM calculational methods specified in the surveillance Requirements implement I B 3/4 11-3 l

I 1

l MDlOACTIVE EFFLUENTS j

RASES DOSE - IODINE-131, IODINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM (Continued) the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposu.e Sf a MEMBER OF THE PUBLIC through appropriate pathways is unlikely tr. be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the 1

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 j

1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-i Cooled Reactors," Revision 1, 721y 1977.

These equations also provide for

)

determining the actual doses based upon the historical average atmospheric conditions. The release rate controls for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways'to man, in the areas at and l

beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were:

(1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the niilk and meat by man, and (4) deposition on the ground with subsequent exposure of man.

3/4.11.2.4 AND 3/4.11.2.5 GASEOUS RADWASTE TREAhENT AND VENTILATION EXHAUST TREAMENT The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATICN FIf!AUST 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 these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I,-10 CFR Part 50, fer gaseous effluents.

I B 3/4 11-4

RADIOACTIVE EFFLUENTS BASES 3/4.11.2.8 VENTING OR PURGING This CQtrTROL provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10 CFR Part 20 for UNRESTRICTED AREAS.

3/4.11.4 TOTAL DOSE This CONTROL is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the proptration and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrems to the total body or any orgac, except the thyroid, which shall be limited to less than or equal to 75 mrems.

For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of

)

Appendix 1, and if direct radiation doses from the reactor units including outside storage tanks, etc. are kept small. The Special Report will describe

)

a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 lin.its.

For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities j

at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in CONTROLS 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE FUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle, j

3 I B 3/4 11-5

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological environmental monitoring program required by this CONTROL provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the station operation. This monicoring program implements section IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological

~

effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.12.1-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for i

a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found in Currie, L.

A., " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"

NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, KASL-300

.(revised annually).

3/4.12.2 LAND USE CENSUS I

This CONTROL is provided to ensure that changes in the use of' areas at and beyond the SITE BOUNDARY are identified and that modifications to the l

radiological environmental monitoring program are made if required by the results of this census. The best inforumtion from the door-to-door survey, l

from aerial survey, from visual survey or from consulting with local I

agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to j

gardens of greater than 50 m* provides assurance that significant exposure l

pathways via leafy vegetables will be identified and monitored since a garden I

of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.

To determine this minimum garden size, the following assumptions were made:

(1) 20% of the garden was used for growing broad leaf vegetation (i.e.,

I similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m*.

i I B 3/4 12-1

3/4.12 RADIOLOGICAL ENVIRONIENTAL MONITORING BAssa

.......=0,....................................................................

3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requi'rement for participation in an approved 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 perfomed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of W adix'I to 10 CFR Part 50.

I 1

i j

i l

i 9

I B 3/4 12-2

e D

PART I RADIOLOGICAL EFFLUENT CONTROLS SECTION 5.0 DESIGN FEATURES l

5.0 DESIGN FEATURES

.=.......................................................,,,,,,,,,,,,,,,,,,,,,

5.1 SITE MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.1 Information regarding radioactive gaseous and liquid effluents which will allow identification of structures and release points as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown in Figure 5.1.1-1.

N e.

we O

4 e.*

=

A e-1 I 5-1 j

9 e

P

~

!i id hI E00I lifi!I

~

,1 s

11 -

l;"

'IT li ~'l 3

g a

~

g e

e l

ens J

. eram

/

4.,

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1 L \\

,4

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)

9

\\%

' A' v

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~5 i19 -(k;5f'iI f!b lg j(!

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, g Q~~ a.

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=

-]

G-:_;::::HusHse t=ESTRICTED A8835 I

AND SITE ""^=Y FOR RADIDAChVE M550U3 AND LIQUID EFFl@#$

I FIGURE 5.1.1-1 HOPE CREEK I 5-2 I

e P

4 PART I RADIOLOGICAL EFFLUENT CONTROLS SECTION 6.0 ADMINISTRATIVE CONTROLS

ADMINISTRATIVE CONTROLS i

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT j

1 6.9.1.6 Routine radiological environmental operating reporting covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be subuIitted prior to May 1 of the year following initial criticality.

The annual radiological environmental operating reports shall include sununacies, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a conqparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an i

assessnent of,the observed impacts of the plant operation on the environment.

i The re; orts shall also include the results of land use censuses required by CONTROL 3.12.2. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the' problem and a planned course of action to alleviate the problem.

The annual radiological environmental operating reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period. Deviations from the sampling program identified in CONTROL 3.12.1 shall be reported. In the event that some results are not available for j

inclusion with the reports, the report shall be submitted noting and-explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the followings a summary description of the I

radiological environmental monitoring program; at least two legible maps, one covering sampling locations near the SITE BOUNDARY, and a second covering the more distant locations, all keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the i

Interlaboratory Comparison Program, as required by CONTROL 3.12.3.

i The report shall also include the results of specific activity analysts in which the primary coolant exceeded the limits of Technical Specification 3.4.5. The following informatica shall be incl W d (11 Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit.

Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the. duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

I 6-18

ADMINISTRATIVE CONTROLS m es sa mme s o amm am e na n o ma am m a m e n e m o m m e ne smas sesses s e m eOmos s e s s e m a s e m o s e s s e s s a RADIOACTIVE EFFLUENT RELEASE REPORT 6.9.1.7 Radioactive release rep ets covering the operation of the unit shall be submitted by May 1 of each year and in accordance with the requirements of 10CFR50.36a.

The radioactive effluent release report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, ' Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials'in Liquid and Gaseous Effluents from Light-Water-cooled Nuclear Power, Plants," Revision 1, June 1974, with data suunnarized on a quarterly basis following the format of Appendix 5 thereof.

The radioactive effluent release report shall include an annual summary of hourly meteorological data collected over the previous year. This annual sununary may be either in the form of an hour-by-hour listing of wind speed, wind direction, and atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. The 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.

The report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1.1-1) during the report period.

All assumptions used in making these assessments, i.e.,

specific activity, exposure time and location, shall be included in these reports. The historical annual average meteorology or the meteorological conditions concurrent 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 pathway doses. The assessment of radiation doses shall be perforined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCM).

The Radioactive Effluent Release Report shall identify those radiological environmental sample parameters and locations where it is not possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In addition, the cause of the unavailability of samples for the pathway and the new location (s) for obtaining replacement samples should be identified. The report should also include a revised figure (s) and table (s) for the ODCM reflecting the new location (s).

i I 6-19

L ADMINISTRATIVE CONTROLS RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)

The radioactive affluent release report shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months'to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.

The radioactive effluent release report shall include the following 1

information for each class of solid waste (as defined by 10 CFR 61) shipped offsite during the report periods a.

Container volume, b.

Total curie quantity (specify whether determined by measurement or estimate),

c.

Principal radionuclide (specify whether determined by measurement or estimate),

d.

Type of waste (e.g., spent resin, compact dry waste, evaporator bottoms),

Type of container (e.g., LSA, Type A, Type B, Large e.

Quantity), and f.

Solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release report shall include unplanned releases from the site to the UNRESTRICTED AREA of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The radioactive effluent release report shall include any changes to the PROCESS CONTROL PROGRAM (PCP), OFFSITE DOSE CALCULATION MANUAL (ODCM) or radioactive waste systems made during the reporting period.

Est 1

I 6-20

ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREA'DiENT SYSTEMS 6.15.1 Licensee initiated major changes to the radioactive waste system (liquid, gaseous and solid) :

\\

1.

Shall be reported to the Commission in the UFSAR for the period in which the evaluation was reviewed by SORC.

The discussion of each change shall contains a..

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR i

50.59; b.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; c.

A detailed description of the aquipment, components and processes involved and the interfaces with other plant systems; d.

An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; e.

An evaluation of the change, which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto; f.

A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; g.

An estimate of the exposure to plant operating personnel as a result of the change; and h.

Documentation of the fact that the change was reviewed and found acceptable by the SORC.

2.

Shall become effective upon review and acceptance by the SORC.

I 6-24

p-Hope Creek ODCM Rev 18 i

l i

1 i

l PARTII CALCULATIONAL METHODOLOGIES

Hope Creek ODCM Rev 18 1.6 LIQUID EFFLUENTS 1.1 Radiation Monitorinz Instrumentation and Controls

'Ihe liquid efDuent monitoring instrumentation and controls at Hope Creek for controlling and monitoring normal radioactive material releases in accordance with the Hope Creek Radiological Emuent Technical Specifications arr. sum'marized as follows:

(1)

Alarm (and Automatic Termination)- Liquid Radwaste Discharge Line Monitor provides the i

alarm and adamatic termination ofliquid (RE4861) radioactive material releases from the liquid waste management system as requimd by Technical SM~ don 3.3.7.10.

Circulating Water Dcmiug Sump DM se Monitor (RE4557) provides alann and automatic termination of liquid radioactive releases from the circulating dewatering sump.

Cond-don drains from certain supply ventilation units and liquids from the fill and venting of the circulating water side of the condenser waterboxes are directed to this sump.

Automatic termination is performed by trip of the sump pumps on high gamma rarliation signal. Controls for releases from the dewatering sump are identified in Appendix G.

(2)

Alarm (Only)

'Ihe Cooling-Tower Blowdown Emuent Monitor (RE8817) provides an Alarm function only for releases into the environment as required by Technical Specification 3.3.7.10.

Liquid radioactive waste flow diagrams with the applicable, associated radiation monitoring instrumentation and controls are presented in Figure 1-1.

1.2 Liquid Emuent Monitor Setpoint Determination Per the requirements of Technical Specification 3.3.7.10, alarm setpoints shall be established for the liquid monitoring instamentation to ensure that the release concentration limits of Specification 3.11.1.1 are met (i.e., the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR 20 Appendix B. Table II, Column 2, (Appendix F) for radionuclides and 2.0E-04 uCi/ml for dissolved or entrained noble gases). The following equation (adopted from NUREG-0133) must be satisfied to meet the liquid emuent restrictions:

C (F + f) es (1.1) f where:

C=

the effluent concentration limit of Technical Specification (3.11.1.1) implementing the 10 CFR 20 MPC (Appendix F) for the site, in uCi/ml.

]

the setpoint, in uCi/ml, of the radioactivity monitor measuring the radioactivity c=

concentration in the effluent line prior to dilution and subsequent release; the setpoint, 2

I i

i l

Hope Creek ODCM Rev 18 represents a value which, if exceeded, would result in concentrations exceeding the limits of 10 CFR 20 in the UNRESTRICTED AREA.

f=

the flow rate at the radiation monitor location, in volume per unit time, but the same units as F, below.

F=

the dilution water flow rate as measured prior to the release point, in volume per unit time.

[ Note that if no dilution is provided, c < C. Also, note that when (F) is large compared to (f),

then (F + f) = F.]

1.2.1 Liemid EfRuent Monitors The setpoints for the liquid efBuent monitors at the Hope Creek Generating Station are determined by the following equation:

MPC,

CTBD * [1 -CF]

SP s

+ bkg (1.2)

RR with:

I C,(gamma)

MPC,=

(1.3)

C, (gamma)

I

MPC, where:

SP alarm setpoint corresponding to the maximum allowable release rate (uCi/ml).

=

MPC, =

an effective MPC value for the mixture of radionuclides in the effluent stream, (uCi/ml).

C, the concentration of radionuclide in the liquid effluent (uCi/ml).

=

MPC, =

the MPC value corresponding to radionuclide i from (Appendix F) 10 CFR 20, Appendix B, Table 11, Column 2 (uCi/ml).

CTBD =

the Cooling-Tower Blowdown Discharge rate at the time of release (gal / min).

RR the liquid effluent release rate (gal / min) at the monitor location (i.e., at the

=

liquid radwaste monitor, at the TBCW monitor, or at the CTBD monitor).

bkg the background of the monitor (uCi/ml).

=

CF Correction factor to account for non-gamma emitting nuclides and radiation

=

monitor inaccuracies.

The radioactivity monitor setpoint equation (1.2) remains valid during outages when the Cooling-Tower Blowdown discharge is potentially as its lowest value. Reduction of the waste stream flow (RR) may be necessary during these periods to meet the discharge criteria.

Procedural restrictions prevent simultaneous liquid releases.

3

Hope Creek ODCM Rev 18 1.2.2 Conservative Default Values Conservative alarm setpoints for liquid radwaste radiation monitors may be determined through the use of default parameters. Table 1-1 summarized all current default values in use for Hope Cree 1 They are based upon the following:

9 (a) substitution of the effective MPC value with a' default value of 7.92E-05 uCi/ml for radwaste releases (Refer to Appendix A forjustification);

(b) substitutions of the Cooling-Tower Blowdown discharge rate with the minimum average flow,in gal / min; and, (c) substitutions of the effluent release rate with the highest allowed rate, in gal / min.

(d) substitution of a 0.8 correction factor (CF) to account for monitor int uracies and non-gamma emitting radionuclides.

The use of the conservative alarm setpoint, or a setpoint below the conservative value, is acceptable provided that the value used is at least as conservative as the release specific setpoint calculated in accordance with Equation 1.2 above. Procedural controls exist to verify the setpoint utilized is at or below what is required.

1.3 Liquid Effluent Concentration Limits - 10 CFR 20 Technical Specification 3.11.1.1 limits the concentration of radioactive material in liquid effluents (after dilution in the Cooling-Tower Blowdown Discharge System) to less than the concentrations as specified in 10 CI R 20, Appendix B, Table II, Column 2 (Appendix F) for radionuclides other than. noble gases. Noble gases are limited to a diluted concentration of 2.0E-04 uCi/ml. Release rates are controlled and radiation monitor alarm setpoints are established as addressed above to ensure that these concentration limits are not exceeded.

4 However, in the event any liquid release results in an alarm setpoint being exceeded, an evaluation of compliance with the concentration limits of Technical Specification 3.11.1.1 may be performed using the following equation:

C, RR 5: 1 (1.4)

MPC, CTBD + RR C

actual concentration of radionuclide i as measured in the undiluted liquid

=

i effluent (uCi/ml).

MPC =

the MPC value corresponding to radionuclide i from 10 CFR 20, Appendix B, i

Table II, Column 2 (Appendix F) (uCi/ml).

2Eo4 uCi/ml for dissolved or entrained noble gases.

=

the actual liquid effluent release rate (gal / min)

RR

=

4

Hope Creek ODCM Rev 18 CTBD =

the actual Cooling-Tower Blowdown discharge at the time of release (gal / min).

I i

l 1.4 Liquid Effluent Dose Calculation - 10 CFR N 1.4.1 MEMBER OF THE PUBLIC Dose - Liquid Emuents Technical Specification 3.11.1.2 limits the dose or dose commitment to MEMBERS OF THE PUBLIC from radioactive materials in liquid effluents from Hope Creek Generating Station to:

- during any ealendar quarter:

s 1.5 mrem to total body s 5.0 mrem to any organ 1

- during any calendaryear:

s 3.0 mrem to total body s 10.0 mrem to any organ Per the surveillance requirements to Technical Specification 4.11.1.2, the following calculation methods shall be used for determining the dose or dose commitment due to the liquid radioactive effluents from Hope Creek.

8.35E-04

VOL D, =

I (C,

A o)

(1.5) i CTBD where:

D, dose or dose commitment to organ o, including total body (mrem).

=

A,o site-related ingestion dose commitment factor to the total body or any organ o

=

for radionuclide i (mrem /hr per uCi/ml).

C, average concentration of radionuclide i, in undiluted liquid effluent

=

representative of volume VOL (uCi/ml).

VOL volume ofliquid effluent released (gal).

=

CTBD =

Average Cooling-Tower Blowdown discharge rate during release period (gal / min).

8.35E-04 =

conversion factor (1.67E-2 hr/ min) and a near field dilution factor of 0.05 (refer to Appendix B, Page B-4 for definition).

5

)

Hope Creek ODCM Rev 18

'Itc site related ingestion dose / dose commitment factors (A,) are presented in Table 1-2 and have been derived in accordance with NUREG-0133 by the equation:

A., = 1.14E+05 [(UI

BIJ + (UF

BF)] Df; (1.6) i where:

A.

composite dose parameter for the total liody or critical organ o of an adult for

=

radionuclide i, for the fish and invertebrate ingestion pathways (mrem /hr per uCi/ml).

1.14E+05=

conversion factor (pCi/uCi

ml/kg per br/yr).

UI adult invertebrate consumption (5 kg/yr).

=

Bi, bioaccumulation factor for radionuclide i in invertebrates from Table 1-3

=

(PCi/kg per pCi/1).

UF adult Ssh consumption (21 kg/yr).

=

Bfi bioaccumulation factor for nuclide i in fish from Table 1-4 (pCi/kg per pCi/l).

=

i Dfi dose conversion factor for nuclide i for adults in preselected organ, o, from

=

Table E-11 of Regulatory Cuide 1.109 (mrem /pCi).

The radionuclides included in the periodic dose assessment per the requirements of Technical Specification 3/4.11.1.2 are those as identified by gamma spectral analysis of the liquid waste samples collected and analyzed per the requirements of Technical Specification 3/4.11.1.1, Table 4.11.1.1.1-1.

Radionuclides requiring radiochemical analysis (e.g., Sr-89 and Sr-90) will be added to the dose analysis at a frequency consistent with the required minimum analysis frequency of Technical Specification Table 4.11.1.1.1-1.

1.4.2 Simplified Liquid Effluent Dose Calculation In lieu of the individual radionuclide dose assessment as presented in Section 1.4.1, the following simplified dose calculation equation may be used for demonstrating compliance with the dose limits of Technical Specification 3.11.1.2. (Refer to Appendix B for the derivation and justification for this simplified method.)

Total Body D. =

1.94E442

VOL

I C, (1.7)

CTBD Maximum Organ D,=

4.28E+42

VOL

I C, (1.8)

CTBD 6

Hope Creek ODCM Rev 18 where:

i D.

= conservatively evaluated total body dose (mrem).

D.,

= evaluated maximum organ dose (mrem).

i C,

= average concentration of radionuclide i, in undiluted liquid effluent representative of the volume VOL (uci/ml).

VOL = volume ofliquid effluent released (gal).

CTBD = average Cooling-Tower Blowdown discharge rate during release period (gal / min).

1.94E+02

= couversion factor (1.67E-2 hr/ min) the ingestion dose commitment factor (Zn-65, total body - 2.32E5 mrem /hr per uCi/ml), and the near field dilution factor of 0.05 (See AWir B).

4.28E+02

= conversion factor (1.67E-2 hr/ min) the conversion==rimum organ dose j

conversion factor (Zn-65, Liver - 5.13E5 mrem /hr per uCi/ml), and the nearfield dilution factor of 0.05 (See Appendix B).

1.5 Liquid Emment Dose Projections Technical Specification 3.11.1.3 requires that the liquid radioactive waste processing system be used to reduce the radioactive material levels in the liquid waste prior to release when the 31-day projected doses exceed:

1

- 0.06 mrem to the total body, or

- 0.2 mrem to any organ.

The applicable liquid waste processing system for maintaining radioactive material releases ALARA are the drain Siters and deminaralizers as delineated in Figure 1-1.

Dose projections are made at least once per 31-days by the following equations:

D, = (D./ d)

31d (1.9)

D, = (D./ d)

31d (1.10) where:

D, the total body dose projection for current 31-day period (mrem).

=

D.

the total body dose to date for current calendar quarter as determined by

=

equation (1.5) or(1.7).

D.

the maximum organ dose to date for current calendar quarter as determined by

=

equation (1.5 or(1.8)(mrem).

d the number of days in current calendar quarter at the end of the release.

=

31d the ntunber of days of concern.

=

1

Hope Creek ODCM Rev 18 1.6 Representative Samples A sample should be representative of the bulk stream or volume of emuent from which it is taken. Prior to sampling, large volumes ofliquid waste should be mixed in as short a time interval as practicable to assure that any sediments or particulate solids are distributed uniformly in the waste mixture. Recirculation pumps'for liquid waste tanks (collection or sample test-tanks) should be capable of recirculating at a rate of not less than two tank volumes in eight hours. Minimum recirculation times and methods of recirculation are controlled by specific plant procedures.

2.0 GASEOUS EFFLUENTS 2.1 Radiation Monitorias Instrumentation and Controls

'Ihe gaseous emuent monitoring instrumentation and controls at Hope Creek for controlling and monitoring normal radioactive material releases in accordance with the Radiological Emuent Technical Specifications are summarized as follows:

(1)

Filtration, Recirculation, and Ventilation System -

The FRVS is maintained in a standby condition. Upon reactor building isolation, the FRVS recirculation system recirculates the reactor building air through HEPA and chamoal filters. Re. leases are made to the atmosphere via a reactor building vent or the South Plant Vent depending on mode of operation. Noble gas monitoring is provided by RE-4811 A.

(2)

South Plant Vent -

The SPV receives discharge from the radwaste evaporator, reactor building purge, auxiliary building radwaste area, condensate demineralizer, pipe chase,'feedwater heater, and untreated ventilation' sources. Emuents are monitored (for noble gas) by the RE-4875B monitor.

(3)

North Plant Vent -

The NPV receives discharge from the gaseous radwaste treatment system (Offgas system) and untreated ventilation air sources. Emuents are monitored (for noble gases) by the RE-4573B monitor.

Gaseous radioactive waste flow diagrams with the applicable, associated radiation monitoring instrumentation controls are presented in Figures 2-1 and 2-2.

8

~

Hope Creek ODCM Rev 18 2.2 Gaseous Effluent Monitor Setpoint Determination 2.2.1 Plant Vent, FRVS Per the requirements of Technical Specification 3.3.7.11, alarm setpoints shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed the limits of Specification 3.11.2.1, which corresponds to a dose rate at the SITE BOUNDARY of 500 mrem / year to the total body or 3000 mrem / year to the skin.

Based on a grab sample analysis of the applicable release (i.e., of FRVS, pipe chase, gaseous radwaste treatment system air, etc.), the radiation monitoring alarm setpoints may be established by the following calculation method. 'Ibe measured radionuclide concentrations.

and release rate are used to calculate the fraction of the allowable release rate, as limited by Specification 3.11.2.1, by the equation:

FRAC = [4.72E+02

X/Q

VF * (C,

KJ) / 500 (2.1)

FRAC = [4.72EM2

X/Q

VF * (C, * (L, + 1.1M))) / 3000 (2.2) where:

FRAC = fraction of the allowable release rate based on the identified radionuclide concentrations and the release flow rate.

X/Q

= annual average meteorological dispersion to the controlling site boundary location (sec/m3).

VF

= ventilation system flow rate for the applicable release point and monitor (ft3/ min).

C,

= concentration of noble gas radionuclide i as determined by radioanalysis of grab sample (uCi/cm3)

K,

= total body dose conversion factor for noble gas radionuclide i (mrem /yr per uCi/m3), from Table 2-1 L,

= beta skin dose conversion factor for noble gas radionuclide i (mrem /yr per uCi/m3), from Table 2-1 M,

= gamma air dose conversion factor for noble gas radionuclide i (mradlyr per uCi/m3), from Table 2-1 1.1

= mrem skin dose per mrad gamma air dose (mrem / mrad) 4.72E+02 = conversion factor (cm3/ft3

min /sec) 500

= total body dose rate limit (mrem /yr) 3000

= skin dose ratelimit(mrem /yr) 9

Hope Creek ODCM Rev 18 Based on the more limiting FRAC (i.e., higher value) as determined above, the alarm l

setpoints for the applicable monitors may be calculated by the equation:

SP.= [AF

I C / FRAC] + bkg (2.3) i where:

SP

= alarm setpoint corresponding to the maximum allowable release rate (uCi/cc).

FRAC = highest fraction of the allowable release rate u determined in equation (2.2).

bkg

= background of the monitor (uCi/cc).

AF

= maministrative allocation factor for the speci6c monitor (0.2 NPV,0.2 SPV,0.1 FRVS).

The allocation factor (AF) is an administrative control imposed to ensure that combined releases from Salem Units 1 and 2 and Hope Creek will not exceed the regulatory limits on release rate from the site (i.e., the release rate limits of Technical Specification 3.11.2.1).

Normally, the combined AF value for Salem Units 1 and 2 is 0.5 (0.25 per unit), with the remainder 0.5 allocated to Hope Creek. Any increase in AF above 0.5 for the Hope Creek Generating Station will be coordinated with the Salem Generating Station to ensure that the combined allocation factors for all units do not exceed 1.0.

2.2.2 Conservative Default Values A conservative alarm setpoint can be established, in lieu of the individual radionuclide evaluation based on the grab sample analysis, to eliminate the potential of periodically having to adjust the setpoint to reflect minor changes in radionuclide distribution and variations in release flow rate. The alarm setpoint may be conservatively determined by the default values presented in Table 2-2.

These values are based upon:

- the maximum ventilation (or purge) flow rate;

- a radionuclide distribution adopted from ANSI N237-1976/ANS 18.1 " Source Term Specifications", Table 5 and;

- an administrative allocation factor of 0.5 to conservatively ensure that any releases from Hope Creek do not exceed the maximum allowable release rate.

For the noble gas radionuclide distribution from ANSI N237-1976/ANS 18.1 (Note Table C-1), the alarm setpoint based on the total body dose rate is more restrictive than the-corresponding setpoint based on the skin dose rate. The resulting conservative, default setpoints are presented in Table 2-2.

10

Hope Creek ODCM Rev 18 l

2.3 Gaseous Effluent Instantaneous Dose Rate Calculations - 10 CFR 20 2.3.1 Site Boundary Dose Rate-Noble Gases Technical Specification 3.11.2.la limits the dose rate at the SITE BOUNDARY due to noble gas releases to 5 500 mrem /yr, total body and 5 3000 mrem /yr, skin. Radiation monitor alarm setpoints are established to ensure that these release limits are not exceeded. In the event any gaseous releases from the station results in an alarm setpoint (as determined in Section 2.2.1) being exceeded, an evaluation of the SITE BOUNDARY dose rate resulting from the release shall be performed using the following equations:

D.,= X/Q

I (K,

Q)

(2.4)

D, = X/Q

E ((L, + 1.1M,)

Q, (2.5) g where:

D.

= Total body dose rate (mrem /yr).

D,

= skin dose rate (mrem /yr).

X/Q

= atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m3).

Q,

= average release rate of radionuclide i over the release period under evaluation (uCi/sec).

K.

= total body dose conversion factor for noble gas radionuclide i (mrem /yr per uCi/m3), from Table 2-1 L

= beta skin dose conversion factor for noble gas radionuclide i (mradlyr per uCi/m3),

i from Table 2-1 M

= gamma air dose conversion factor for noble gas radionuclide i (mradlyr per i

uCi/m3, from Table 2-1.

1.1

= mrem skin dose per mrad gamma air dose (mrem / mrad)

As appropriate, simultaneous releases from Salem Units 1 and 2 and Hope Creek will be considered in evaluating compliance with the release rate limits of Specification 3.11.2.1a, following any releases exceeding the above prescribed alarm setpoints. Monitor indications (readings) may be averaged over a time period not to exceed 15 minutes when determining noble gas release rate based on correlation of the monitor reading and monitor sensitivity.

~

The 15-minute averaging is needed to allow for reasonable monitor response to potentially changing radioactive material concentrations and to exclude potential electronic spikes in monitor readings that may be unrelated to radioactive material releases. As identified, any electronic spiking monitor responses may be excluded from the analysis.

NOTE: For administrative purposes, more conservative alarm setpoints than those as prescribed above may be imposed. However, conditions exceeding these more limiting alarm setpoints do not necessarily indicate radioactive material release rates exceeding the dose limits of Technical Specification 3.11.2.la. Provided actual releases do not result in radiation 11

)

Hope Creek ODCM Rev 18 monitor indications exce-Aing alarm setpoint values based on the above criteria, no further analyses are required for demonstrating compliance with the limits of Specification 3.11.2.1a.

Actual meteorological conditions concurrent with the release period or the default, annual average dispersion parameters as presented in Table '2-3 may be used for evaluating the gaseous effluent dose rate.

2.3.2 Site Boundary Dose Rate - Radiolodine and Particulates Technical Specification 3.11.2.lb limits the dose rate to 51500 mrem /yr to any organ for I-131, 1-133, tritium and particulates with half-lives greater than 8 days. To demonstrate compliance with this limit, an evaluation is performed at a frequency no greater than that corrhia to the sampling and analysis time period (e.g., nominally once per 7 days).

The following equation shall be used for the dose rate evaluation:

D, = X/Q

I (R,

Qj (2.6) where:

I D,

= average organ dose rate over the sampling time period (mrem /yr).

X/Q

= atmospheric dispersion to the controlling SITE BOUNDARY location for the i

inhalation pathway (sec/m3).

R,,

= dose parameter for radionuclide i (mrem /yr per uCi/m3) and organ o for the child J

inhalation pathway from Table 2-4.

Q,

= average release rate over the appropriate sampling period and analysis frequency for radionuclide i - - I-131, I-133, tritium or other radionuclide in particulate form with half-life greater than 8 days (uCi/sec).

By substituting 1500 mrem /yr for D, and solving for Q, an allowable release rate for I-131 can be determined. Based on the annual average meteorological dispersion (See Table 2-3) and the most limiting potential pathway, age group and organ (inhalation, child, thyroid - Ri

= 1.62E+07 mrem /yr per uCi/m3), the allowable release rate for I-131 is 34.7 uCi/sec.

Reducing this release rate by a factor of 2 to account for potential dose contributions from other radioactive particulate material and other release points (e.g., Salem), the corresponding release rate allocated to Hope Creek is 17.4 uCi/sec. For a 7-day period, which is the nominal sampling and analysis frequency for I-131, the cumulative release is 10.5 Ci.

Therefore, as long as the 1-131 release in any 7-day period do not exceed 10.5 Ci, no I

additional analyses are needed for verifying compliance with the Technical Specification 3.11.2.1.b limits on allowable release rate.

12

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l l

Hope Creek ODCM Rev 18 2.4 Noble Gas Effluent Dose Calculations - 10 CFR 50 2.4.1 UNRESTRICTED AREA Dose-Noble Gases Technical Specification 3.11.2.2 requires a periodic assessment of releases of noble gases to evaluate complianse with the quarterly dose limits of 5 5 mrad, gamma-air and 510 mrad, beta-air and the calendar year limits $ 10 mrad, gamma-air and 5 20 mrad, beta-air.

The limits are applicable separately to each generating station and are not combined site l

limits. The following equations shall be used to calculate the gamma-air and beta-air doses.

l D, =.3.17E-08

X/Q

I (M,

Qi)

(2.7) l l

l D. = 3.17E-08

X/Q

I (N,

Q,)

(2.8) where:

l D,

= air dose due to gamma emissions for noble gas radionuclides (mrad).

D.

= air dose due to beta emissions for noble gas radionuclides (mrad).

X/Q

= atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m3).

Q,

= cumulative release of noble gas radionuclide i over the period ofinterest (uCi).

M,

= air dose factor due to gamma emission from noble gas radionuclide i (mrad /yr per uCi/m3, from Table 2-1.

I N,

= air dose factor due to beta emissions from noble gas radionuclide i (mrad /yr per uCi/m3, Table 2-1).

3.17E-08

= conversion factor (yr/sec).

l 2.4.2 Simplified Dose Calculation for Noble Gases

?

In lieu of the individual noble gas radionuclide dose assessment as presented

above, the following simplified dose calculation equations may be used for verifying compliance with the dose limits of Technical Specification 3.11.2.2 (Refer to Appendix C for the derivation and justification of this simplified method).

3.17E-08 D, =

X/Q

M,

I Q, (2.9) 0.50 3.17E-08 D. =

X/Q

N,

I Q.

(2.10) 0.50 l

l where:

13 I'

Hope Creek ODCM Rev 18 i

M,

= 8.1E+03, effective gamma-air dose factor (mrad /yr per uCi/m3).

I N,,

= 8.5E+03, effective beta-air dose factor (mrad /yr per uCi/m3).

Q,

= cumulative release for all noble gas radionuclides (uCi).

l 0.50

= conservatism factor to account for potential variability in the radionuclide distribution.

Actual meteorological conditions concurrent with the release period or the default, annual average dispersion parameters as presented in Table 2'-3, may be used for the evaluation of the gamma-air and beta-air doses.

j 2.5 Radioiodine and Padiculate Dose Calculations - 10 CFR 50 2.5.1 UNRESTRICTED AREA Dose - Radioiodine and Particulates In accordance -with the requirements of Technical Specificatica 3.11.2.3, a periodic assessment shall be performed to evaluate compliance with the quarterly dose limit < 15 mrem to any organ. 'Ihe following equation shall be used to evaluate the maximum organ dose due to release of I-131, I-133, tritium and particulates with half-lives greater than 8 days:

D,,, = 3.17E-08

W

SF,

E (R,

QJ (2.11) where:

D,

= dose or dose commitment via all pathways p and age group a (as identified in Table 2-3) to organ o, including the total body (mrem).

W

= atmospheric dispersion parameter to the controlling location (s) as identified in Table 2 3.

X/Q

= atmospaeric dispersion for inhalation pathway and H-3 dose contribution via other pathwr.ys (sec/m3).

D/Q

= atmospheric deposition for vegetation, milk and ground plane exposure pathy:ays (1/m2).

R,

= dose factor for radionuclide i (mrem /yr per uCi/m3 or m2 - mrem /yr per uCi'sec)

L and organ o from Table 2-4 for each age group a and the applicable pathway p as identified in Table 2-3.

Values for R were derived in accordance with the 4

methods described in NUREG-0133.

Q,

= cumulative release over the period ofinterest for radionuclide i - I-131, I-133, H-3 or radioactive material in particulate form with half-life greater than 8 days (uCi).

- Sf,

= annual seasonal correction factor to account for fraction of the year that the applicable exposure pathway does not exist.

(1) For milk and vegetation expo.rure pathways:

= A six month fresh vegetation and grazing season (May through October)= 0.5 (2) For inhalation andgroundplane exposure pathways:

= 1.0 14

Hope Creek ODCM Rev 18 For evaluating the maximum exposed individual, the infant age group is controlling for the milk pathway. Only the controlling age group as identified in Table 2-3 need be evaluated for compliance with Technical Specification 3.11.2.3.

2.5.~2 Simplified Dose Calculation for Radioiodines and Particulates l

- In lieu of the individual radionuclide (I-131, I-133 and particulates) dose assessment as presented above, the following simplified dose calculation equation may be used for verifying compliance with the dose limits of Technical Specification 3.11.2.3 (Refer to Appendix D for the derivation andjustification of this simplified method).

D.

= 3.17E-08

W

SF,

RI-131

I Q, (2.12) where:

1 i

D_

= maximum organ dose (mrem).

RI-131 = I-131 dose parameter for the thyroid for the identified controlling pathway.

= 1.05E+12, infant thyroid dose parameter with the cow-milk pathway controlling (m2 - mrem /yr per uCi/sec).

j W

= D/Q for radioiodine,2.87E-101/m2.

Qi

= cumulative release over the period ofinterest for radionuclide i - I-131 or 1

radioactive material in particulate form with half-life greater than 8 days (uCi).

The location of exposure pathways and the maximum organ dose calculation may be based on the available pathways in the surrounding environment of Hope Creek as identified by the annual land-use census (Technical Specification 3.12.2).

Otherwise, the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-3.

2.6 Gaseous Effluent Dose Projection Technical Specification 3.11.2.4 requires that the VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce radioactive material levels prior to discharge when projected doses in 31-days exceed:

- 0.2 mrad to air from gamma radiation, or

- 0.4 mrad to air from beta radiation, or

- 0.3 mrad to any organ of a MEMBER OF THE PUBLIC l

He applicable gaseow processing systems for maintaining radioactive material releases ALARA are the Gaseous Radwaste Treatment System and Exhaust Treatment System as delineated in Figures 2-1 and 2-2.

15 i

Hope Creek ODCM Rev 18 Dose projection are performed at least once per 31-days by the following equations:

D, = (D,/ d)

31d (2.13)

D, = (D,/ d)

31d (2.14)

D, = (D

/ d)

31d (2.15) where:

D,

= gamma air dose projection for cunent 31-day period (mrad)'.

D,

= gamma air dose to date for current calendar quarter as determined by equation (2.7) or(2.9)(mrad).

l D,

= beta air dose projection for current 31-day period (mrad).

D,

= beta air dose to date for current calendar quarter as determined by equation (2.8) or (2.10)(mrad).

D,

= maximum organ dose projection for current 31-day period (mrem).

D.,

= maximum organ dose to date for current calendar quarter as dete. mined by equation (2.11)or(2.12)(mrem).

d

= number ofdays in current calendar quarter at the end of the release.

31d

= the number of days of concem.

l 3.0 SPECIAL DOSE ANALYSIS l

3.1 Doses Due to Activities Inside the SITE BOUNDARY l

In accordance with Technical Specification 6.9.1.7, the Radioactive Emuent Release Report i

(RERR) submitted by May 1st of each year shall include an assessment of radiation doses from l

radioactive liquid and gaseous emuents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY.

l The calculation methods as presented in Sections 2.4 and 2.5 may be used for determining the maximum potential dose to a MEMBER OF THE PUBLIC based on the parameters from Table 2-3 and 2-hours per visit per year. The default value for the meteorological dispersion data as presented in Table 2-3 may be used if cunent year meteorology is unavailable at the time of NRC spordug. However, a follow-up evaluation shall be performed when the data becomes available.

j

~

\\

l 16

I Hope Creek ODCM Rev 18 3.2 Total Dose to MEMBERS OF THE PUBLIC - 40 CFR 190 The Radioactive Effluent Release Report (RERR) submitted by May 1st of each year shall also include an assessment of the radiation dose to the likely most exposed MEMBER OF THE PUBLIC for reactor releases and other nearby uranium fuel cycle courses (including dose contributions from effluents and direct radiation from on-site sources). For the likely most exposed MEMBER OF THE PUBLIC in the vicinity of Artificial Island, the sources of exposure need only consider the Salem Generating statio' and the Hope Creek Generating n

Station: No other fuel cycle facilities contribute to the MEMBER OF THE PUBLIC dose for the ArtificialIsland vicinity.

I

)

The dose contribution from the operation of Salem Generating Stations will be estimawl based on the methods as wed in the Salem Offsite Dose Calculation Manual (SGS ODCM).

l As appropriate for demonstratmg/ evaluating compliance with the limits of Technical Specification 3.11.4 (40 CFR 190), the results of the environmental monitoring program may be used for providing data on actual measured levels ofradioactive material in the actual l

pathways ofexposure.

3.2.1 Emuent Dose Calculations For purposes ofimplementing the surveillance requirements of Technical Specification 3/4.11.4 and the reporting requirements of 6.9.1.7 (RERR), dose calculations for the Hope Creek Generating Station may be performed using the calculation methods contained within the ODCM; the conservation controlling pathways and locations of Table 2-4 or the actual pathways and locations as identified by the land use census (Technical Specification 3/4.12.1) may be used. Averar annual meteorological dispersion parameters or meteorological conditions concunert with the release period under evaluation may be used.

3.2.2 Direct Exposure Dose Determination Any potentially significant direct exposure contribution to off-site individual doses may be evaluated based on the results of the environmental measurements (e.g., TLD, ion chamber l

measurements) and/or by the use of a radiation transport and shielding calculation method.

l Only during a non-typical condition will there exist any potential for significant on-site sources I

at Hope Creek that would yield potentially significant off-site doses (i.e.., in excess of 1 mrem per year to a MEMBER OF THE PUBLIC), that would require detailed evaluation for demonstrating compliance with 40 CFR 190. However, should a situation exist whereby the direct exposure contribution is potentially significant, on-site measurements, off-site measurements and/or calculational techniques will be used for determination of dose for assessing 40 CFR 190 compliance.

17 L-

Hope Creek ODCM Rev 18 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4.1 Sampling Program he operational phase of the Radiological Environmental Monitoring Program (REMP) is conducted in accordance with the requirements of Appendix A Technical Specification 3.12.

Le objectives of the program are:

- To determine whether any significant increases occur in the concentration of radionuclides in the critical pathways of exposure in the v%nity of Artificist Island;

- To determine if the operation of the Hope Creek Generating Station has resulted in any increase in the inventory oflong lived radionuclides in the environment;

- To detect any changes in the ambient gamma radiation levels; and

- To verify that HCGS operations have no detrimental effects on the health and safe? of the public or on the environment..

The sampling requirements (type of samples, collection frequency and analysis) and sample locations are presented in Appendix E.

NOTE: No public drinking water samples or irrigation water samples are taken as these pathways are not directly effected by liquid effluents discharged from Hope Creek Generating Station.

4.2 Interlabo.? tory Comparison Program Technical Specification 3.12.3 requires analyses be performed on radioactive material supplied as part of an Interlaboratory Comparison. Participation in an approved Interlaboratory Comparison Program provides a check on the preciseness ofmeasurements of radioactive materials in environmental samples. A summary of the Interlaboratory Comparison Program results will be provided in the Annual Radiological Environrnental Operating Report pursuant to Technical Specifications 6.9.1.7.

5.0 HCGS EXPLOSIVE GAS MONITORING PROGRAM The Hope Creek Explosive Gas Monitoring program was moved within the Hope Creek Technical Specifications to section 6.8.4.d. This was performed in Technical Specification Amendment 91.

Details of the Hope Creek Explosive Gas Monitoring program are maintained in station implementing procedures and are controlled by the 50.59 safety evaluation and procedure processes.

4 18

Hope Creek ODCM Rev i8 l

FIGURE 1-1 l

LIQUID RADWASTE TREATMENT AND MONITORING SYSTEM lare water omiert :

l P "l l

Cooling Tower Supply Vest Drains Sumo Bash ColledorTanks (2)

Equipment Equipment Equipment Drain Drain Filter Drain Domin SampleTanks (2)

Waste Surge Tank l

Condensate Storage Tank 1

Floor Drain Floor Floor Floor Drain Collector Tanks (2)

Drain Filter Drain Domin Sample Tanks (2)

See Note 1

\\\\\\\\\\\\\\ \\\\\\Y V0eank N

a s\\\\\\\\\\\\

N N N NE t N s

NNNNN N N N sel' e\\

i\\\\\\\\\\\\\\\\\\\\\\

l Detergent Drain Detergent l

Tanks (2)

Drain Filter l

RE 4861 Log: RE = Radiation Ef!Iuent Monitor

~

4861= Uquid Radwaste Monitor RE 8817 = Cooling tower Blowdown Monitor 8817 l

qsri: cwa 69 Mw Note 1: Specified equipmen% W.*a. beh3 t isinstalled but notin use. Equipment is pending To Delaware River abandonment per DCP 4EC-3634.

19

Hope Creek ODCM Rev 18 FIGURE 1-2 SOLID RADWASTE PROCESSING SYSTEM See Note 1 N ' ' 'ex'x\\NNNN N'

\\N

,\\'NN N\\\\\\

N N N

\\

\\ N N %

Note 1: Specified equipment is installed but not in use

\\ \\

D 4EC-34 w'Nx xx x

x xx x x'sxxs N Nwhy Agb;g NR p

\\ \\d N

6

\\ \\ \\ \\ \\

\\

a s *D' w'xNNXN W sw N's D

xxx xxxxwtw r

1r r in Tanks v

e v ng T nk Weste Sludge Phase Seperator JL

)

See Note 1 ir

\\\\\\ \\

\\\\

'fCNek'dQiQN

c1'"pf0"'

s

\\N N N N\\N N N N N

ir k

Spent Rosin Tank (Fgure 1-1) u 20

Hope Creek ODCM Rev 18 TABLE l-1 PARAMETERS FOR LIQUID ALARM SETPOINT DETERMINATION Actual Default Parameter Value Value Units Comments MPCe Calc 7.92E-5*

uCi/ml Calculated for each batch to be released MPC I-131 3.0E-07 N/A uCi/ml Taken from 10 CFR 20, Appendix B, Table 11 Column 2(Appendix F)

C,

' Measured N/A uCi/ml Taken from gamma spectralanalysis ofliquid effluent

MPC, Measured N/A uci/ml Taken from 10 CFR 20, Appendix B. Table II, Column 2 (Appendix F)

CTBD Measured 12000 gpm Cooling tower blowdown discharge RR Measured 176 gpm or Determined prior to release, release rate can be adjusted for Technical Specification compliance 1300 gpm(CST)

Estimated 100 gpm(TBCW)

Maximum flow rate with both pumps running

{

(50 gpm each)

SP (Setpoints)

A)RE4861 Cale 1.08E-03 uCi/mi Default alarm setpoints; more conservative values may be used as appropriate and desirable 4

for ensuring regulatory compliance and for maintaining releases ALARA RE8817 Cale 1.58E-05 uCUml RE4557 Calc 2.40E-06 uCFml Maximum alarm setpoint for continuous release; more conservative value may be established by plant procedure B)RE4861 Calc 1.46E-04 uCi/mi These setpoints are for condensate storage tank releases

[

RE8817 Calc 1.58E-05 uCi/ml

See Appendix A for basis 21

l Hope Creek ODCM Rev 18 TABLE l-2 SITE RELATED INGESTION DOSE COMMITMENT FACTOR, A.

(FISH AND INVERTEBRATE CONSUMPTION)

(mrem /hr per uCi/ml)

%d litic lioni-1.n cr 1.Itmh l in roid Aidn n 1.und

(. i-1.1.1 H-3 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 C-14 1.45E+4 2.90E+3 2.90E+3 2.90E+3 2.90E+3 2.90E+3 2.90E+3 Na-24 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 P-32 4.69E+6 2.91E+5 1.81E+5 5.27E+5 Cr-51 5.58E+0 3.34E+0 1.23E+0 7.40EM 1.40E+3 Mn-54 7.%E+3 1.35E+3 2.10E+3 2.16E+4 Mn-56 1.78E+2 3.15E+1 2.26E+2 5.67E+3 Fe-55 5.l lE+4 3.53E+4 8.23E+3 1.97E+4 2.03E+4 Fe-59 8.06E+4 1.90E+5 7.27E+4 5.30E+4 6.32E+5 Co-57 1.42E+2 2.36E+2 3.59E+3 Co-58 6.03E+2 1.35E+3 1.22E+4 Co-60 1.73E+3 3.82E+3 3.25E+4 Ni-63 4.96E+4 3.44E+3 1.67E+3 7.18E+2 Ni-65 2.02E+2 2.62E+1 1.20E+1 6.65E+2 Cu-64 2.14E+2 1.01 E+2 5.40E+2 1.83E+4 Zn-65 1.61E+5 5.13E+5 2.32E+5 3.43E+5 3.23E+5 Zn-69m 5.66E+3 1.36E+4 1.24E+3 8.22E+3 8.29E+5 Br-82 4.07E+0 4.67E+0 l

Br-83 7.25E-2 1.04E-1 Br-84 9.39E-2 7.37E-7 Br-85 3.86E-3 Rb-86 6.24E+2 2.91 E+2 1.23E+2 Rb-88 1.79E+0 9.49E-1 2.47E-11 Rb-89 1.19E+0 8.34E-1 6.89E-14 Sr-89 4.99E+3 1.43E+2 8.00E+2 Sr-90 1.23E+5 3.01 E+4 3.55E+3 Sr-91 9.18E+1 3.71 E+0 4.37E+2 Sr-92 3.48E+1 1.51 E+0 6.90E+2 Y-90 6.06E+0 1.63E-1 6.42E+4 Y-91m 5.73E-2 2.22E-3 1.68E-1 Y-91 8.88E+1 2.37E+0 4.89E+4 Y-92 5.32E-1 1.56E-2 9.32E+3 Y-93 1.69E+0 4.66E-2 5.35E+4 Zr-95 1.59E+1 5.11EM 3.46E+0 8.02EM 1.62E+4 Zr-97 8.81 E-1 1.78E-1 8.13E-2 2.68E-1 5.51E+4 Nb-95 4.47E+2 2.49E+2 1.34E+2 2.46E+2 1.51 E+6 Nb-97 3.75E%

9.49E-1 3.46E-1 1.11E+0 3.50E+3 Mo-99 1.28E+2 2.43E+1 2.89E+2 2.96E+2 Tc-99m 1.30E-2 3.66E-2 4.66E-1 5.56E-1 1.79E-2 2.17E+1 Tc-101 1.33E-2 1.92E-2 1.88E-1 3.46E-1 9.81 E-3 5.77E-14 As-76 4.38E+2 1.16E+3 5.14E+3 3.42E+2 1.39E+3 3.58E+2 4.30E+4 22

Hope Creek ODCM Rev 18 TABLE l-2 (cont'd)

SITE RELATED INGESTION DOSE COMMITMENT FACTOR, A, (FISH AND INVERTEBRATE CONSUMPTION)

(mrem /hr per uCi/ml)

%n lide lione 1.is er 1.llmh

'.I h roid hidnes 1.uug GI-l.11 3

Ru-103 1.07E+2 4.60E+1 4.07E+2 1.25E+4 Ru-105 8.89E+0 3.51 E+0 1.15E+2 5.44E+3 Ru-l%

1.59E+3 2.01 E+2 3.06E+3 1.03E+5 Rh-103m Rh-106 Ag-110m 1.56E+3 1.45E+3 8.60E+2 2.85E+3 5.91 E+5 Sb-124 2.77E+2 5.23 E+0 1.10E+2 6.71 E-1 2.15E+2 7.86E+3 Sb-125 1.77E+2 1.98E+0 4.21 E+1 1.80E-1 1.36E+2 1.95E+3 Te-125m 2.17E+2 7.86E+1 2.91 E+1 6.52E+1 8.82E+2 8.66E+2 Te-127m 5.48E+2 i.96E+2 6.68E+1 1.40E+2 2.23E+3 1.84E+3 Te-127 8.90E+0 3.20E+0 1.93EM 6.60E@

3.63E+1 7.03E+2 Te-129m 9.31E+2 3.47E+2 1.47E+2 3.20E+2 3.89E+3 4.69E+3 Te-129 2.54E+0 9.55E-1 6.19E-1 1.95E+0 1.07E+1 1.92E+0 Te-131m 1.40E+2 6.85E+1 5.71 E+1 1.08E+2 6.94E+2 6.80E+3 Te-131 L59E+0 6.66E-1 5.03 E-1 1.31 E+0 6.99E+0 2.26E-1 Te-132 2.04E+2 1.32E+2 1.24E+2 1.46E+2 1.27E+3 6.24E+3 1-130 3.96E+1 1.17E+2 4.61 E+1 9.91 E+3 1.82E+2 1.01E+2 1-131 2.18E+2 3.12E+2 1.79E+2 1.02E+5 5.35E+2 8.23 E+1 1-132 1.06E+1 2.85E+1 9.96E+0 9.96E+2 4.54E+1 5.35E+0 1-133 7.45E+1 1.30E+2 3.95E+1 1.90E+4 2.26E+2 1.16E+2 I-134 5.56E+0 1.51 E+1 5.40E+0 2.62E+2 2.40E+1 1.32E-2 I-135 2.32E+1 6.08E+1 2.24E+1 4.01 E+3 9.75E+1 6.87E+1 Cs-134 6.84E+3 1.63E+4 1.33E+4 5.27E+3 1.75E+3 2.85E+2 Cs-136 7.16E+2 2.83 E+3 2.04E+3 1.57E+3 2.16E+2 3.21 E+2 Cs-137 8.77E+3 1.20E+4 7.85E+3 4.07E+3 1.35E+3 2.32E+2 Cs-138 6.07E+0 1.20E+1 5.94E+0 8.81E+0 8.70E-1 5.12E-5 Ba-139 7.85E4 5.59E-3 2.30E-1 5.23 E-3 3.17E-3 1.39E+1 Ba-140 1.64E+3 2.06E+0 1.08E+2 7.02E-1 1.18E+0 3.38E+3 Ba-141 3.81 E+0 2.88E-3 1.29E-1 2.68E-3 1.63 E-3 1.80E-9 Ba-142 1.72E+0 1.77E-3 1.08E-1 1.50E-3 1.00E-3 2.43E-18 La-140 1.57E+0 7.94E-1 2.10E-1 5.83E+4 La-141 8.06E-2 3.67E-2 9.13E-3 2.68E+2 Ce-141 3.43E+0 2.32E+0 2.63E-1 1.08E+0 8.86E+3 Ce-143 6.04E-1 4.46E+2 4.94E-2 1.97E-1 1.67E+4 Ce-144 1.79E+2 7.47E+1 9.59E+0 4.43E+1 6.04E+4 Pr-143 5.79E+0 2.32E+0 2.87E-1 1.34E+0 2.54E+4 Pr-144 1.90E-2 7.87E-3 9.64E-4 4.44E-3 2.73E-9 Nd-147 3.96E+0 4.58E+0 2.74E-1 2.68E+0 2.20E+4 W-187 9.16E+0 7.66E+0 2.68E+0 2.51 E+3 Np-239 3.53E-2 3.47E-3 1.91 E-3 1.08E-2 7.l l E+2 23

Hope Creek ODCM Rev 18 TABLE l-3 BIOACCUMULATION FACTORS (pCi/kg per pCi/ liter)*

1.1 \\11 N I s \\1-l\\\\ \\iiitIIsil s \\l I \\\\ \\ lIKit IN\\ l it! Eltl< \\ l E.

H 9.0E-01 9.3 E-01 C

1.8E+03 1.4E+03 Na 6.7E-02 1.9E-01 P

3.0E+03 3.0E+04 Cr 4.0E+02 2.0E+03 Mn 5.5E+02 4.0E402 I

Fe 3.0E+03 2.0E+04 Co 1.0E+02 1.0E+03 Ni 1.0E+02 2.5E+02 i

Cu 6.7E+02 1.7E+03 Zn 2.0E+03 5.0E404 Br 1.5E-02 3.1 E+00 Rb 8.3 E+00 1.7E+01 Sr 2.0E+00 2.0E+01 Y

2.5E+01 1.0E+03 Zr 2.0E+02 8.0E+01 Nb 3.0E+04 1.0E+02 Mo 1.0E+01 1.0E+01 Tc 1.0E+01 5.0E+01 Ru 3.0E+00 1.0E+03 Rh 1.0E+01 2.0E+03 Ag 3.3E+03 3.3E+03 Sb 4.0E+01 5.4E+00 Te 1.0E+01 1.0E+02 1

1.0E+01 5.0E+01 Cs 4.0E+01 2.5E+01 Ba 1.0E+01 1.0E+02 La 2.5E+01 1.0E+03 Ce 1.0E+01 6.0E+02 Pr 2.5E+01 1.0E+03 Nd 2.5E+01 1.0E+03 W

3.0E+01 3.0E+01 Np 1.0E+01 1.0E+01 As 3.3E+02 3.3E+02 Values in this table are taken from Regt$latory Guide 1.109 except for phosphoms (fish) which is adapted from NUREG/CR-1336 and silver, arsenic and antimony which are taken from UCRL 50564, Rev.1, October 1972.

24

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Hope Creek ODCM Rev 18 FIGURE 2-2 VENTILATION EXHAUST TREATMENT SYSTEM RE Gessous Rahme Tresenent Symem (from Figure 2-1).

INorth Plant unosened vensienon vent At Sou't:es M

I f See Note 1 Auxikary Bud 6ng Radweste Ares R

H 48 vertianon (Trpcalof 3) i r------ - ------------ -------

Condensate Derruneralizer R

H C

H mSouth Plant Room Ar j

Vent i

Compartmert Exhaust System (TBCE) u..............................l Pipe Chase Ar Feedwater Hester Room Air Turtune Budding Exhaust TB Od Storage Room Exhaust Giand !eal Exhaust MettlancalVacuurn Purre Exhaust MF 2-1)

Reanor Budding Ve"nelation Exhaust Systern '

~~~~

= ********************

R H

Reactor ung RE (Typcalof 3) 4811A Alf FRVS Recrc 9ymem l

l FRVS Vent l

H C

H

> FRVS Systern Vent (Typcalof B) l (Typcal of 2 )

L.

.)

R = Roughing Fdter H = HEPA F Aer C s Charcool Faer RE a Ra$ anon Effluent Morutor Note 1: spectned equpmott is instened but not h use.

Egidpment perwung abandonment por DCP 4EC-3634, 26

f l

Hope Creek ODCM Rev 18 TABLE 2-1 DOSE FACTORS FORNOBLE GASES l

Total Body -

Skin Gamma Air Beta Air Dose Factor Dose Factor Dose Factor Dose Factor Ki Li Mi Ni Q Radionuclide uCi/m3) uCi/m3) uCi/m3) uCi/m3 l

Kr-83m 7.56E-02 1.93E+01 2.88E402 l

l Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 l

l Kr-85 1.61E+01 1.34E+03 1.72E401 1.95E+03 I

l Kr-87 5.92EM3 9.73E+03 6.17E+03 1.03E+04 i

Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 l.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E%3 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11EM2 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E403 1.92E+03 2.46E+03 Xe-137

_.L42E+03

.1.22E+04 1.51E+03

._.1.27E+04 Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2.69E@3 9.30E+03 3.28E+03 27 e

p j

l l

Hope Creek ODCM Rev 18 TABLE 2-2

. PARAMETERS FOR GASEOUS ALARM SETPOINT DETERMINATION HOPE CREEK Actual Default Parameter Value Value Units Comments X/Q Calculated 2.67E-6 sec/m3 From FSAR Table 2.3-31, i

0.5 mile, N VF(NPV)

Measured 41900 ft'/ min Maximum Operation VF(SPV)

Measured 440,180 ft'/ min Maximum Operation VF(FRVS)

Measured 9000 ft'/ min Maximum Operation AF(NPV)

Coordinated 0.2 Unitiess Administrative with SGS allocation factor to ensure release do not exceed AF(SPV) 0.2 Unitiess release rate limit AF(FRVS) 0.1 Unitiess C; Measured N/A uCi/cm' K Nuclide N/A mrem /yr Table 2-1 4

Specific per uCi/m' L Nuclide N/A mrem /yr Table 2-1 i

Specific per uCi/m')

M,Nuclide N/A mradlyr Table 2-1 Specific per uCi/m' SP:

NPV Calculated 2.43E-4 uCi/cc Default alarm Setpoints; i

SPV Calculated 2.31E-5 uCi/cc more conservative values FRVS Calculated 5.65E-4 uCi/cc may be used as deemed appropriate for ensuring ALARA & regulatory compliance 28

Hope Creek ODCM Rev 18 TABLE 2-3 CONTROLLING LOCATIONS, PATHWAYS AND ATMOSPHERIC DISPERSION FOR DOSE CALCULATIONS

Tech Spec Location Pathway (s)

Age Group (sec/m3)

(1/m2) 3.11.2.la Site Boundary Noble Gases N/A 2.67E-06 N/A 0.5 Mile, N direct exposure 3.11.2.1b Site Boundary Inhalation Child 2.67E-06 N/A 0.5 Mile, N 3.11.2.2 Site Boundary Gamma-Air N/A 2.67E-06 N/A 0.5 Mile, N Beta-Air 3.11.2.3 Residence /

Milk, ground Infant 7.2E-08 2.87F-10 Dairy - 4.9 plane and Miles, W inhalation The identified controlling locations, pathways and atmospheric dispersion are from the Artificial Island Radiological Monitoring Program and the Hope Creek FSAR.

29

Hope Creek ODCM Rev 18 Table 2-4 Pathway Dose Factors - Atmospheric Releases R(io), Inhalation Pathway Dose Factors - ADULT (mrem /yr per uCi/m3)

E E

in rui.

E suclidi

~.tlod H-3 1.26E+3 1.26E+3 1.26E+3 l 1.26E+3 1.26E+3 1.26E+3 C-14 1.82E+4 3.41E+3 3.41E+3 3.41E+3 l 3.41E+3 3.41E+3 3.41 E+3 P-32 1.32E+6 7.71E+4 l

8.64E+4 5.01E+4 Cr-51 5.95E+1 2.28E+1 l 1.44E+4 3.32E+3 1.00E+2 Mn-54 3.96E+4 9.84E+3 l 1.40E+6 7.74E+4 6.30E+3 Fe-55 2.46E+4 1.70E+4 l 7.21E+4 6.03E+3 3.94E+3 Fe 59 1.18E+4 2.78E+4 1.02E+6 1.88E+5 1.06E+4 Co-57 6.92E+2 3.70E+5 3.14E+4 6.71 E+2 Co-58 1.58E+3 9.28E+5 1.06E+5 2.07E+3 Co-60 1.15E+4 5.97E+6 2.85E+5 1.48E+4 Ni-63 4.32E+5 3.14E+4 1.78E+5 1.34 E+4 1.45E+4 Zn-65 3.24E+4 1.03E+5 6.90E+4 8.64E+5 5.34E+4 4.66E+4 Rb-86 1.35E+5 1.66E+4 5.90E+4 Sr-89 3.04E+5 1.40E+6 3.50E+5 8.72E+3 Sr-90 9.92E+7 9.60E+6 7.22E+5 6.10E+6 Y-91 4.62E+5 1.70E+6 3.85E+5 1.24E+4 Zr-95 1.07E+5 3.44E+4 5.42E+4 1.77E+6 1.50E+5 2.33E+4 Nb-95 1.41 E+4 7.82E+3 7.74E+3 5.05E+5 1.04E+5 4.21 E+3 Ru-103 1.53E+3 5.83E+3 5.05E+5 1.10E+5 6.58E+2 Ru-106 6.91E+4 1.34E+5 9.36E '-6 9.12E+5 8.72E+3 Ag-110m 1.08E+4 1.00E+4 1.97E+4 4.63E+6 3.02E+5 5.94E+3 Sb-124 3.12E+4 5.89E+2 7.55E+1 2.48E+6 4.06E+5 1.24E+4 Sb-125 5.34E+4 5.95E+2 5.40E+1 1.74E+6 1.01E+5 1.26E+4 Te-125m 3.42E+3 1.58E+3 1.05E+3 1.24E+4 3.14E+5 7.06E+4 4.67E+2 Te-127m 1.26E+4 5.77E+3 3.29E+3 4.58E+4 9.60E+5 1.50E+5 1.57E+3 Te-129m 9.76E+3 4.67E+3 3.44E+3 3.66E+4 1.16E+6 3.83E+5 1.58E+3 I-131 2.52E+4 3.58E+4 1.19E+7 6.13 E+4 6.28E+3 2.05E+4 Cs-134 3.73E+5 8.48E+5 2.87E+5 9.76E+4 1.04E+4 7.28E+5 Cs-136 3.90E+4

-1.46E+5 8.56E+4 1:20E+4 1.17E+4 1.10E+5 Cs-137 4.78E+5 6.21 E+5 2.22E+5 7.52E+4 8.40E+3 4.28E+5 Ba-140 3.90E+4 4.90E+1 1.67E+1 1.27E+6 2.18E+5 2.57E+3 Ce-141 1.99E+4 1.35E+4 6.26E+3 3.62E+5 1.20E+5 1.53E+3 Ce-144 3.43E+6 1.43E+6 8.48E+5 7.78E+6 8.16E+5 1.84E+5 Pr-143 9.36E+3 3.75E+3 2.16E+3 2.81 E+5 2.00E+5 4.64E+2 Nd-147 5.27E+3 6.10E+3 3.56E+3 2.21E+5 1.73E+5 3.65E+2 30

1 1

Hope Creek ODCM Rev 18 Table 2-4 (cont'd) l Pathway Dose Factors - Atmospheric Releases R(io), Inhal: tion Pathway Dose Factors - TEENAGER (mrem /yr per uCi/m3) s uctig g

E 19 nin sidne

. uni

.1 - 1.1.

' llod; H-3 1.27E+3 1.27E+3 1.27E+3 1.27E+3 1.27E+3 1.27E+3 C-14 2.60E+4 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 P-32 1.89E+6 1.10E+5 9.28E+4 7.16E+4 Cr-51 7.50E+1 3.07E+1 2.10E+4 3.00E+3.

1.35E+2 Mn-54 5.11E+4 1.27E+4 1.98E+6 6.68E+4 8.40E+3 i

Fe-55 3.34E+4 2.38E+4 1.24E+5 6.39E+3 5.54E+3 Fe-59 1.59E+4 3.70E+4 1.53E+6 1.78E+5 1.43E+4 Co-57 6.92E+2 5.86E+5 3.14E+4 9.20E+2 Co-58 2.07E+3 1.34E+6 9.52E+4 2.78E+3 Co-60 1.51 E+4 8.72E+6 2.59E+5 1.98E+4 Ni-63 5.80E+5 4.34E+4 3.07E+5 1.42E+4 1.98E+4 Zn-65 3.86E+4 1.34E+5 8.64E+4 1.24E+6 4.66E+4 6.24E+4 Rb-86 1.90E+5 1.77E+4 8.40E+4 Sr-89 4.34E+5 2.42E+6 3.71E+5 1.25E+4 Sr-90 1.08E+8 1.65E+7 7.65E+5 6.68E+6 Y-91 6.61 E+5 2.94E+6 4.09E+5 1.77E+4 Zr-95 1.46E+5 4.58E+4 6.74E+4 2.69E+6 1.49E+5 3.15E+4 Nb-95 1.86E+4 1.03E+4 1.00E+4 7.51E+5 9.68E+4 5.66E+3 Ru-103 2.10E+3 7.43E+3 7.83E+5 1.09E+5 8.96E+2 Ru-106 9.84E+4 1.90E+5 1.61 E+7 9.60E+5 1.24E+4 Ag-110m 1.38E+4 1.31 E+4 2.50E+4 6.75E+6 2.73E+5 7.99E+3 Sb-124 4.30E+4 7.94E+2 9.76E+1 3.85E+6 3.98E+5 1.68E+4 Sb-125 7.38E+4 8.08E+2 7.04E+1 2.74E+6 9.92E+4 1.72E+4 Te-125m 4.88E+3 2.24E+3 1.40E+3 5.36E+5 7.50E+4 6.67E+2 Te-127m 1.80E+4 8.16E+3 4.38E+3 6.54E+4 1.66E+6 1.59E+5 2.18E+3 Te-129m 1.39E+4 6.58E+3 4.58E+3 5.19E+4 1.98E+6 4.05E+5 2.25E+3 1-131 3.54E+4 4.91E+4 1.46E+7 8.40E+4 6.49E+3 2.64E+4 Cs-134 5.02E+5 1.13E+6 3.75E+5 1.46E+5 9.76E+3 5.49E+5 Cs-136 5.15E+4 - 1.94E+5 1.10E+5 1.78E+4 1.09E+4 1.37E+5 Cs-137 6.70E+5 8.48E+5 3.04E+5 1.21E+5 8.48E+3 3.l l E+5 Ba-140 5.47E+4 6.70E+1 2.28E+1 2.03E+6 2.29E+5 3.52E+3 Cc-141 2.84E+4 1.90E+4 8.88E+3 6.14E+5 1.26E+5 2.17E+3 Ce-144 4.89E+6 2.02E+6 1.21E+6 1.34E+7 8.64E+5 2.62E+5 Pr-143 1.34E+4 5.31E+3 3.09E+3 4.83E+5 2.14E+5 6.62E+2 Nd-147 7.86E+3 8.56E+3 5.02E+3 3.72E+5 1.82E+5 5.13E+2 31

l Hope Creek ODCM Rev 18 Table 2-4 (cont'd)

Pathway Dose Factors - Atmospheric Releases R(io), Inhalation Pathway Dose Factors - CHILD (mrem /yr per uCi/m3) s uylul<

tom ne tipon E

E

.1 - 1.1.

.itad H-3 1.12E+3 1.12E+3 1.12E+3 1.12E+3 1.12E+3 1.12E+3 C-14 3.59E+4 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3 P-32 2.60E+6 1.14E+5 4.22E+4 9.88E+4 Cr-51 8.55E+1 2.43E+1 1.70E+4 1.08E+3 1.54E+2 Mn-54 4.29E+4 1.00E+4 1.58E%

2.29E+4 9.51E+3 Fe-55 4.74E+4 2.52E+4 1.11E+5 2.87E+3 7.77E+3 Fe-59 2.07E+4 3.34E+4 1.27E+6 7.07E+4 1.67E+4 Co-57 9.03E+2 5.07E+5 1.32E+4 1.07E+3 Co-58 1.11E+6 3.44E+4 3.16E+3 1.77E+3 Co-60 1.31E+4 7.07E+6 9.62E+4 2.26E+4 Ni-63 8.21E+5 4.63E+4 2.75E+5 6.33E+3 2.80E+4 Zn-65 4.26E+4 1.13E+5 7.14E+4 9.95E+5 1.63E+4 7.03E+4 Rb-86 1.98E+5 7.99E+3 1.14E+5 Sr-89 5.99E+5 2.16E+6 1.67E+5 1.72E+4 Sr-90 1.01 E+8 1.48E+7 3.43E+5 6.44E+6 Y-91 9.14E+5 2.63E+6 1.84E+5 2.44E+4 Zr-95 1.90E+5 4.18E+4 5.96E+4 2.23E+6 6.11 E+4 3.70E+4 Nb-95 2.35E+4 9.18E+3 8.62E+3 6.14E+5 3.70E+4 6.55E+3 Ru-103 2.79E+3 7.03E+3 6,62E+5 4.48E+4 1.07E+3 Ru-106 1.36E+5 1.84E+5 1.43E+7 4.29E+5 1.69E+4 Ag-110m 1.69E+4 1.14E+4 2.12E+4 5.48E+6 1.00E+5 9.14E+3 Sb-124 5.74E+4 7.40E+2 1.26E+2 3.24E+6 1.64E+5 2.00E+4 Sb-125 9.84E+4 7.59E+2 9.10E+1 2.32E+6 4.03E+4 2.07E+4 Te-125m 6.73E+3 2.33E+3 1.92E+3 4.77E+5 3.38E+4 9.14E+2 Te-127m 2.49E+4 8.55E+3 6.07E+3 6.36E+4 1.48E+6 7.14E+4 3.02E+3 Te-129m 1.92E+4 6.85E+3 6.33E+3 5.03E+4 1.76E+6 1.82E+5 3.04E+3 1-131 4.81E+4 4.81E+4 1.62E+7 7.88E+4 2.84E+3 2.73E+4 Cs-134 6.51E+5 1.01E+6 3.30E+5 1.21E+5 3.85E+3 2.25E+5 Cs-136 6.51E+4 t71E+5 9.55E+4 - 1.45E+4-4.18E+3 -

1.16E+5 Cs-137 9.07E+5 8.25E+5 2.82E+5 1.04E+5 3.62E+3 1.28E+5 Ba-140 7.40E+4 6.48E+1 2.11 E+1 1.74E+6 1.02E+5 4.33E+3 Cc-141 3.92E+4 1.95E+4 8.55E+3 5.44E+5 5.66E+4 2.90E+3 Ce-144 6.77E+6 2.12E+6 1.17E+6 1.20E+7 3.89E+5 3.61E+5 Pr-143 1.85E+4 5.55E+3 3.00E+3 4.33E+5 9.73E+4 9.14E+2 Nd-147 1.08E+4 8.73 E+3 4.81E+3 3.28E+5 8.21 E+4 6.81E+2 32

f l

l Hope Creek ODCM Rev 18 Table 2-4 (cont'd)

Pathway Dose Factors - Atmospheric Releases R(io), Inhalation Pathway Dose Factors - INFANT (mrem /yr per uCi/m3)

E g

JJney Lun;

' 1 -1.1.

(uant<

hen E

H-3 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 C-14 2.65E+4 5.31E+3 5.31E+3 5.31E+3 5.31E+3 5.31E+3 5.31E+3 P-32 2.03E+6 1.12E+5 1.61E+4 7.74E+4 Cr-51 5.75E+1 1.32E+1 1.28E4 3.57E+2 8.95E+1 Mn-54 2.53E+4 4.98E+3 1.00E+6 7 06E+3 4.98E+3 Fe-55 1.97E+4 1.17E+4 8.69E+4 1.09E+3 3.33E+3 Fe-59 1.36E+4 2.35E+4 1.02E+6 2.48E+4 9.48E+3 Co-57 6.51E+2 3.79E+5 4.86E+3 6.41E+2 Co-58 1.22E+3 7.77E+5 1.11 E+4 1.82E+3 Co-60 8.02E+3 4.51E+6 3.19E+4 1.18E+4 Ni-63 3.39E+5 2.04E+4 2.09E+5 2.42E+3 1.16E+4 Zn-65 1.93E+4 6.26E+4 3.25E+4 6.47E+5 5.14E+4 3.l lE+4 Rb-86 1.90E+5 3.04E+3 8.82E+4 Sr-89 3.98E+5 2.03E+6 6.40E+4 1.14E+4 1

Sr-90 4.09E+7 1.12E+7 1.31E+5 2.59E+6 Y-91 5.88E+5 2.45E+6 7.03E+4 1.57E+4 Zr-95 1.15E+5 2.79E+4 3.11 E+4 1.75E+6 2.17E+4 2.03E+4 Nb-95 1.57E+4 6.43E+3 4.72E+3 4.79E+5 1.27E+4 3.78E+3 Ru-103 2.02E+3 4.24E+3 5.52E+5 1.61E+4 6.79E+2 Ru-106 8.68E+4 1.07E+5 1.16E+7 1.64E+5 1.09E+4 Ag-110m 9.98E+3 7.22E+3 1.09E+4 3.67E+6 3.30E+4 5.00E+3 Sb-124 3.79E+4 5.56E+2 1.01 E+2 2.65E+6 5.91 E+4 1.20E+4 Sb-125 5.17E+4 4.77E+2 6.23 E+1 1.64E+6 1.47E+4 1.09E+4 Te-125m 4.76E+3 1.99E+3 1.62E+3 4.47E+5 1.29E+4 6.58E+2 Te-127m 1.67E+4 6.90E+3 4.87E+3 3.75E+4 1.31 E+6 2.73E+4 2.07E+3 Te-129m 1.41 E+4 6.09E+3 5.47E+3 3.18E+4 1.68E+6 6.90E+4 2.23E+3 1-131 3.79E+4 4.44E+4 1.48E+7 5.18E+4 1.06E+3 1.96E+4 Cs-134 3.%E+5 7.03E+5 1.90E+5 7.97E+4 1.33E+3 7.45E+4 Cs-136 4.83E+4 1.35E+5 5.64E+4 1.18E+4 1.43E+3 5.29E+4 Cs-137 5.49E+5 6.12E+5 1.72E+5 7.13E+4 1.33E+3 4.55E+4 Ba-140 5.60E+4 5.60E+1 1.34E+1 1.60E+6 3.84E+4 2.90E+3 Ce-141 2.77E+4 1.67E+4 5.25E+3 5.17E+5 2.16E+4 1.99E+3 Cc-144 3.19E+6 1.21 E+6 5.38E+5 9.84E+6 1.48E+5 1.76E+5 Pr-143 1.40E+4 5.24E+3 1.97E+3 4.33E+5 3.72E+4 6.99E+2 Nd-147 7.94E+3 8.13 E+3 3.15E+3 3.22E+5 3.12E+4 5.00E+2 33

Hope Creek ODCM Rev 18 Table 2-4 (cont'd)

Pathway Dose Factors - Atmospheric Releases R(io), Grass-Cow-Milk Pathway Dose Factors - ADULT (mrem /yr per uCi/m3)for H-3 and C-14 (m2

mrem /yr per uCi/sec)for others sudidi h,h

.n c hs n,n s idne

.un:

H-3 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 C-14 3.63E+5 7.26E+4 7.26E+4 7.26E+4 7.26E+4 7.26E+4 l 7.26E+4 P-32 1.71E+10 1.06E+9 1.92E+9 l 6.60E+8 Cr-51 1.71E+4 6.30E+3 3.80E+4 7.20EM l 2.86E+4 Mn-54 8.40E+6 2.50E+6 2.57E+7 l 1.60E+6 Fe-55 2.51E+7 1.73E+7 9.67E+6 9.95E+6 l 4.04E+6 Fe-59 2.98E+7 7.00E+7 1.95E+7 2.33E+8 l 2.68E+7 Co-57 1.28E+6 3.25E+7 l 2.13E+6 Co-58 4.72E+6 9.57E+7 1.06E+7 Co-60 1.64E+7 3.08E+8 3.62E+7 Ni-63 6.73E+9 4.66E+8 9.73E+7 2.26E+8 Zn-65 1.37E+9 4.36E+9 2.92E+9 2.75E+9 1.97E+9 Rb-86 2.59E+9 5.11E+8 1.21E+9 Sr-89 1.45E+9 2.33E+8 4.16E+7 Sr-90 4.68E+10 1.35E+9 1.15E+10 Y-91 8.60E+3 4.73E+6 2.30E+2 Zr-95 9.46E+2 3.03E+2 4.76E+2 9.62E+5 2.05E+2 Nb-95 8.25E+4 4.59E+4 4.54E+4 2.79E+8 2.47E+4 Ru-103 1.02E+3 3.89E+3 1.19E+5 4.39E+2 Ru-106 2.04E+4 3.94E+4 1.32E+6 2.58E+3 Ag-110m 5.83E+7 5.39E+7 1.06E+8 2.20E+10 3.20E+7 Sb-124 2.57E+7 4.86E+5 6.24E+4 2.00E+7 7.31E+8 1.02E+7 Sb-125 2.04E+7 2.28E+5 2.08E+4 1.58E+7 2.25E+8 4.86E+6 Te-125m 1.63E+7 5.90E+6 4.90E+6 6.63E+7 6.50E+7 2.18E+6 Te-127m 4.58E+7 1.64E+7 1.17E+7 1.86E+8 1.54E+8 5.58E+6 Te-129m 6.04E+7 2.25E+7 2.08E+7 2.52E+8 3.04E+8 9.57E+6 1-131 2.%E+8 4.24E+8 1.39E+11 7.27E+8 1.12E+8 2.43E+8 Cs-134 5.65E+9 1.34E+10 4.35E+9 1.44E+9 2.35E+8 1.10E+10 Cs-136 2.61E+8 - 't.03E+9 5.74E+8 7.~87E+7 1.17E+8 7.42E+8 Cs-137 7.38E+9 1.01E+10 3.43E+9 1.14E+9 1.95E+8 6.61E+9 Ba-140 2.69E+7 3.38E+4 1.15E+4 1.93E+4 5.54E+7 1.76E+6 Cc-141 4.84E+3 3.27E+3 1.52E+3 1.25E+7 3.71 E+2 Cc-144 3.58E+5 1.50E+5 8.87E+4 1.21E+8 1,92E+4 Pr-143 1.59E+2 6.37E+1 3.68E+1 6.96E+5 7.88E+0 Nd-147 9.42E+1 1.09E+2 6.37E+1 5.23E+5 6.52E+0 34

Hope Creek ODCM Rev 18 Table 2-4 (cont'd)

Pathway Dose Factors - Atmospheric Releases R(io), Grass-Cow-Milk Pathway Dose Factors - TEENAGER (mrem /yr per uCi/m3) for H-3 and C-14 (m2

mrem /yr per uCi/sec) for others uclitic lione ne h rois s id m-

.un:

.1-1.1.1 T.llod, 3

H-3 l

9.94E+2 9.94E+2 9.94E+2 9.94E+2 9.94E+2 9.94E+2 C-14 l 6.70E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 P-32 l 3.15E+10 1.95E+9 2.65E+9 1.22E+9 Cr-51 l

2.78E+4 1.10E+4 7.13E+4 8.40E+6 5.00E+4 l

Mn-54 l

1.40E+7 4.17E+6 2.87E+7 2.78E+6 Fe-55 l 4.45E+7 3.16E+7 2.00E+7 1.37E+7 7.36E+6 Fe-59 l 5.20E+7 1.21E+8 3.82E+7 2.87E+8 4.68E+7 Co-57 l

2.25E+6 4.19E+7 3.76E+6 l

Co-58 l

7.95E+6 1.10E+8 1.83E+7 Co-60 l

2.78E+7 3.62E+8 6.26E+7 Ni-63 l 1.18E+10 8.35E+8 1.33E+8 4.01E+8 Zn-65 l 2.11E+9 7.31E+9 4.68E+9 3.10E+9 3.41E+9 Rb-86 l

4.73E+9 7.00E+8 2.22E+9 I

Sr-89 l 2.67E+9 3.18E+8 7.66E+7 Sr-90 9.92E+7 9.60E+6 7.22E+5 6.10E+6 Y-91 1.58E+4 6.48E+6 4.24E+2 1.20E+6 3.59E+2 Zr-95 1.65E+3 5.22E+2 7.67E+2 Nb-95 1.41E+5 7.80E+4 7.57E+4 3.34E+8 4.30E+4 Ru-103 1.81E+3 6.40E+3 1.52E+5 7.75E+2 Ru-106 3.75E+4 7.23E+4 1.80E+6 4.73E+3 Ag-110m 9.63E+7 9.11 E+7 1.74E+8 2.56E+10 5.54E+7 Sb-124 4.59E+7 8.46E+5 1.04E+5 4.01E+7 9.25E+8 1.79E+7 Sb-125 3.65E+7 3.99E+5 3.49E+4 3.21E+7 2.84E+8 8.54E+6 4

Te-125m 3.00E+7 1.08E+7 8.39E+6 8.86E+7 4.02E+6 Te-127m 8.44E+7 2.99E+7 2.01 E+7 3.42E+8 2.10E+8 1.00E+7 Te-129m 1.1 IE+8 4.10E+7 3.57E+7 4.62E+8 4.15E+8 1.75E+7 1-131 5.38E+8 7.53 E+8 2.20E+11 1.30E+9 1.49E+8 4.04E+8 Cs-134 9.81 E+9 2.31E+10 7.34E+9 2.80E+9 2.87E+8 1.07E+10 Cs-136 4.45E+8 1.75E+9 9.53E+8 1.50E+8 1.41E+8 1.18E+9 Cs-137 1.34E+10 1.78E+10 6.06E+9 2.35E+9 2.53E+8 6.20E+9 Ba-140 4.85E+7 5.95E+4 2.02E+4 4.00E+4 7.49E+7 3.13E+6 Ce-141 8.87E+3 1.35E+4 2.79E+3 1.69E+7 6.81 E+2 Ce-144 6.58E+5 2.72E+5 1.63E+5 1.66E+8 3.54E+4 Pr-143 2.92E+2 1.17E+2 6.77E+1 9.61 E+5 1.45E+1 Nd-147 1.81E+2 1.97E+2 1.16E+2 7.l lE+5 1.18E+1 35

l l

l Hope Creek ODCM Rev 18 Table 2-4(cont'd)

Pathway Dose Factors - Atmospheric Releases R(io), Grass-Cow-Milk Pathway Dose Factors - CHILD (mrem /yr per uCi/m3) for H-3 and C-14 (m2

mrem /yr per uCi/sec) for others Nuclid ton

.n e h ron sidne

.un;

.1-1.11 T.llod, l

3 H-3 1.57E+3 1.57E+3 1.57E+3 -

1.57E+3 1.57E+3 1.57E+3 C-14 1.65E+6 3.29E+5 3.29E+5 3.29E+5 3.29E+5 3.29E+5 3.29E+5 P-32 7.77E+10 3.64E+9 2.15E+9 3.00E+9 Cr-51 5.66E+4 1.55E44 1.03E+5 5.41 E+6 1.02E+5 Mn-54 2.09E+7 5.87E+6 1.76E+7 5.58E+6 Fe-55 1.12E+8 5.93E+7 3.35E+7 1.10E+7 1.84E+7 Fe-59 1.20E+8 1.95E+8 5.65E+7 2.03E+8 9.71 E+7 Co-57 3.84E+6 3.14E+7 7.77E+6 Co-58 1.21E+7 7.08E+7 3.72E+7 Co-60 4.32E+7 2.39E+8 1.27E+8 Ni-63 2.96E+10 1.59E+9 1.07E+8 1.01E+9 Zn-65 4.13E+9 1.10E+10 6.94E+9 1.93E+9 6.85E+9 Rb-86 8.77E+9 5.64E+8 5.39E+9 Sr-89 6.62E+9 2.56E+8 1.89E+8 Sr-90 1.12E+11 1.51 E+9 2.83E+10 Y-91 3.91 E+4 5.21 E+6 1.04E+3 Zr-95 3.84E+3 8.45E+2 1.21E+3 8.81E+5 7.52E+2 Nb-95 3.18E+5 1.24E+5 1.16E+5 2.29E+8 8.84E+4 Ru-103 4.29E+3 1.08E+4 1.11E+5 1.65E+3 Ru-106 9.24E+4 1.25E+5 1.44E+6 1.15E+4 Ag-110m 2.09E+8 1.41E+8 2.63E+8 1.68E+10 1.13E+8 Sb-124 1.09E+8 1.41E+8 2.40E+5 6.03E+7 6.79E+8 3.81 E+7 Sb-125 8.70E+7 1.41E+6 8.%E+4 4.85E+7 2.08E+8 1.82E+7 Te-125m 7.38E+7 2.00E47 2.07E+7 7.12E+7 9.84E+6 Te-127m 2.08E+8 5.60E+7 4.97E+7 5.93E+8 l.68E+8 2.47E+7 Te-129m 2.72E+8 7.61 E+7 8.78E+7 8.00E+8 3.32E+8 4.23 E+7 I-131 1.30E+9 1.31E+9 4.34E+11 2.15E+9 1.17E+8 7.46E+8 Cs-134 2.26E+10 3.71E+10 1.15E+10 4.13E+9 2.00E+8 7.83E+9 Cs-136 1.00E+9 2.76E49 1.47E+9 2.19E+8 9.70E+7 1.79E+9 Cs-137 3.22E+10 3.09E+10 1.01E+10 3.62E+9 1.93E+8 4.55E+9 Ba-140 1.17E+8 1.03E+5 3.34E+4 6.12E+4 5.94E+7 6.84E+6 Ce-141 2.19E+4 1.09E+4 4.78E+3 1.36E+7 1.62E+3 Ce-144 1.62E+6 5.09E+5 2.82E+5 1.33E+8 8.66E+4 Pr-143 7.23E+2 2.17E+2 1.17E+2 7.80E+5 3.59E+1 Nd-147 4.45E+2 3.60E+2 1.98E+2 5.71 E+5 2.79E+1 36

I l

l Hope Creek ODCM Rev 18 Table 2-4 (cont'd) l Pathway Dose Factors - Atmospheric Releases

, R(io), Grass-Cow-Milk Pathway Dose Factors - INFANT (mrem /yr per uCi/m3) for H-3 and C-14 (m2

mrem /yr per uCi/sec) for others suchde toni he hm.a udne

.unt a-l.i.

.itad H-3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 C-14 3.23E+6 6.89E+5 6.89E+5 6.89E+5 6.89E+5 6.89E+5 6.89E+5 P-32 1.60E+11 9.42E+9 2.17E+9 6.21E+9 Cr-51 1.05E+5 2.30E+4 2.05E+5 4.71E+6 1.61E+5 l

Mn-54 3.89E+7 8.63E+6 1.43E+7 8.83E+6 Fe-55 1.35E+8 8.72E4-7 4.27E+7 1.11E+7 2.33E+7 Fe-59 2.25E+8 3.93E+8 1.16E+8 1.88E+8 1.55E+8 Co-57 8.95E+6 3.05E+7 1.46E+7 Co-58 2.43E+7 6.05E+7 6.06E+7 Co-60 8.81E+7 2.10E+8 2.08E+8 Ni-63 3.49E+10 2.16E+9 1.07E+8 1.21E+9

_Zn-65 5.55E+9 1.90E+10 9.23E+9 1.61E+10 8.78E+9 Rb-86 2.22E+10 5.69E+8 1.10E+10 2.59E+8 331E+8 Sr-89 1.26E+10 Sr'-90 1.22E+11 1.52E+9 3.10E+10 Y-91 7.33E+4 5.26E+6 1.95E+3 Zr-95 6.83E+3 1.66E+3 1.79E+3 8.28E+5 1.18E+3 Nb-95 5.93E+5 2.44E+5 1.75E+5 2.06E+8 1.41E+5 Ru-103 8.69E+3 1.81E+4 1.06E+5 2.91 E+3 Ru-106 1.90E+5 2.25E+5 1.44E+6 2.38E+4 Ag-110m 3.86E+8 2.82E+8 4.03E+8 1.46E+10 1.86E+8 Sb-124 2.09E+8 3.08E+6 5.56E+5 1.31E+8 6.46E+8 6.49E+7 Sb-125 1.49E+8 1.45E+6 1.87E+5 9.38E+7 1.99E+8 3.07E+7 Te-125m 1.51E+8 5.04E+7 5.07E+7 7.18E+7 2.04E+7 Te-127m 4.21E+8 1.40E+8 1.22E+8 1.04E+9 1.70E+8 5.10E+7 i

Te-129m 5.59E+8 1.92E+8 2.15E+8 1.40E+9 3.34E+8 8.62E+7 1-131 2.72E+9 3.21E+9 1.05E+12 3.75E+9 1.15E+8 1.41E+9 l

Cs-134 3.65E+10 6.80E+10 1.75E+10 7.18E+9 1.85E+8 6.87E+9 Cs-136 1.96E+9

- -5.77E+9 2.30E+9 4.70E+8 8.76E+7 2.15E+9 Cs-137 5.15E+10 6.02E+10 1.62E+10 6.55E+9 1.88E+8 4.27E+9 Ba-140 2.41 E+8 2.41E+5 5.73E+4 1.48E+5 5.92E+7 1.24E+7 Ce-141 4.33E+4 2.64E+4 8.15E+3 1.37E+7 3.llE+3 Ce-144 2.33E+6 9.52E+5 3.85E+5 1.33E+8 1.30E+5 Pr-143 1.49E+3 5.59E+2 2.08E+2 7.89E+5 7.41E+1 Nd-147 8.82E+2 9.06E+2 3.49E+2 5.74E+5 5.55E+1 37

1 l

l Hope Creek ODCM Rev 18 1

i Table 2-4 (cont'd)

Pathway Dose Factors - Atmospheric Releases R(io), Vegetation Pathway Dose Factors - ADULT (mrem /yr per uCi/m3) for H-3 and C-14 (m2

mrem /yr per uCi/sec) for others w lule' lom h ron Udne

.ung 3

H-3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 l 2.26E+3 C-14 8.97E+5 1.79E+5 1.79E+5 1.79E+5 1.79E+5 1.79E+5 l1.79E+5 P-32 1.40E+9 8.73E+7 1.58E+8 l 5.42E+7 Cr-51 2.79E+4 1.03E+4 6.19E+4 1.17E+7 l 4.66E+4 Mn-54 3.11 E+8 9.27E+7 9.54E+8 l 5.94E+7 Fe-55 2.09E+8 1.45E+8 8.06E+7 8.29E+7 l 3.37E+7 Fe-59 1.27E+8 2.99E+8 8.35E+7 9.96E+8 l 1.14E+8 Co-57 1.17E+7 2.97E+8 l 1.95E+7 Co-58 3.09E+7 6.26E+8 l 6.92E+7 Co-60 1.67E+8 3.14E+9 l 3.69E+8 Ni-63 1.04E+10 7.21 E+8 1.50E+8 l 3.49E+8 Zn-65 3.17E+8 1.01E+9 6.75E+8 6.36E+8 l 4.56E+8 Rb-86 2.19E+8 4.32E+7 l1.02E+8 Sr-89 9.96E+9 1.60E+9 l 2.86E+8 Sr-90 6.05E+11 1.75E+10 l 1.48E+10 Y-91 5.13E+6 2.82E+9 l 1.37E+5 Zr-95 1.19E+6 3.81 E+5 5.97E+5 1.21E+9 I 2.58E+5 Nb-95 1.42E+5 7.91 E+4 7.81E+4 4.80E+8 4.25E+4 Ru-103 4.80E+6 1.83E+7 5.61 E+8 2.07E+6 Ru-106 1.93E+8 3.72E+8 1.25E+10 2.44E+7 Ag-110m 1.06E+7 9.76E+6 1.92E+7 3.98E+9 5.80E+6 Sb-124 1.04E+8 1.96E+6 2.52E+5 8.08E+7 2.95E+9 4.11 E+7 Sb-125 1.36E+8 1.52E+6 1.39E+5 1.05E+8 1.50E+9 3.25E+7 Te-125m 9.66E+7 3.50E+7 2.90E+7 3.93E+8 3.86E+8 1.29E+7 Te-127m 3.49E+8 1.25E+8 8.92E+7 1.42E+9 1.17E+9 4.26E+7 Te-129m 2.55E+8 9.50E+7 8.75E+7 1.06E+9 1.28E+9 4.03E+7 l-131 8.09E+7 1.16E+8 3.79E+10 1.98E+8 3.05E+7 6.63 E+7 Cs-134 4.66E+9 1.1 lE+10 3.59E+9 1.19E+9 1.94E+8 9.07E+9 l

Cs-136 4.20E+7 1.66E+8 9.24E+7 1.27E+7 1.89E+7 1.19E+8 Cs-137 6.36E+9 8.70E+9 2.95E+9 9.81 E+8 1.68E+8 5.70E+9 i

Ba-140 1.29E+8 1.62E+5 5.49E+4 9.25E+4 2.65E+8 8.43E+6 Ce-141 1.96E+5 1.33E+5 6.17E+4 5.08E+8 1.51 E+4 Ce-144 3.29E+7 1.38E+7 8.16E+6 1.llE+10 1.77E+6 Pr-143 6.34E+4 2.54E+4 1.47E+4 2.78E+8 3.14E+3 Nd-147 3.34E+4 3.86E+4 2.25E+4 1.85E+8 2.31 E+3 38

l 1

l l

l Hope Creek ODCM Rev i8 Table 2-4 (cont'd)

Pathway Dose Factors - Atmospheric Releases R(io), Vegetation Pathway Dose Factors - TEENAGER (mrem /yr per uCi/m3) for H-3 and C-14 (m2

mrem /yr per uCi/sec) for others s uWid

ton,

.n i hyoie uant

.un:

.t-1.1.

.isoa H-3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 C-14 1.45E%

2.91E+5 2.91E+5 2.91E+5 2.91E+5 2.91E+5 2.91E+5 P-32 1.61E+9 9.%E+7 1.35E+8 6.23E+7 Cr-51 3.44E+4 1.36E+4 8.85E+4 1.04E+7 6.20EM Mn-54 4.52E+8 1.35E+8 9.27E+8 8.97E+7 Fe-55 3.25E+8 2.31E+8 1.46E+8 9.98E+7 5.38E+7 Fe-59 1.81E+8 4.22E+8 1.33E+8 9.98E+8 1.63E+8 Co-57 1.79E+7 3.34E+8 3.00E+7 Co-58 4.38E+7 6.04E+8 1.01E+8 Co-60 2.49E+8 3.24E+9 5.60E+8 Ni-63 1.61E+10 1.13E+9 1.81E+8 5.45E+8 Zn-65 4.24E+8 1.47E+9 9.41E+8 6.23E+8 6.86E+8 Rb-86 2.73E+8 4.05E+7 1.28E+8 Sr-89 1.51E+10 1.80E+9 4.33E+8 Sr-90 7.51E+11 2.11E+10 1.85E+11 Y-91 7.87E+6 3.23E+9 2.l lE+5 Zr-95 1.74E+6 5.49E+5 8.07E+5 1.27E+9 3.78E+5 Nb-95 1.92E+5 1.06E+5 1.03E+5 4.55E+8 5.86E+4 Ru-103 6.87E+6 2.42E+7 5.74E+8 2.94E+6 Ru-l%

3.09E+8 5.97E+8 1.48E+10 3.90E+7 Ag-110m 1.52E+7 1.44E+7 2.74E+7 4.04E+9 8.74E+6 Sb-124 1.55E+8 2.85E+6 3.51E+5 1.35E+8 3.l lE+9 6.03E+7 Sb-125 2.14E+8 2.34E+6 2.04E+5 1.88E+8 1.66E+9 5.00E+7 Te-125m 1.48E+8 5.34E+7 4.14E+7 4.37E+8 1.98E+7 Te-127m 5.51E+8 1.96E+8 1.31E+8 2.24E+9 1.37E+9 6.56E+7 Te-129m 3.67E+8 1.36E+8 1.18E+8 1.54E+9 1.38E+9 5.81E+7 I-131 7.70E+7 1.08E+8 3.14E+10 1.85E+8 2.13E+7 5.79E+7 Cs-134 7.09E+9 1.67E+10 5.30E+9 2.02E+9 2.08E+8 7.74E+9 Cs-136 4.29E+7

- 1.69E+8-9.19E+7 1.45E+7 1.36E+7 1.13E+8 Cs-137 1.01E+10 1.35E+10 4.59E+9 1.78E+9 1.92E+8 4.69E+9 Ba-140 1.38E+8 1.69E+5 5.75E+4 1.14E+5 2.13E+8 8.91 E+6 Ce-141 2.82E+5 1.88E+5 8.86E+4 5.38E+8 2.16E+4 Ce-144 5.27E+7 2.18E+7 1.30E+7 1.33E+10 2.83E+6 Pr-143 7.12E+4 2.84E+4 2.34E+8 3.55E+3 1.65E+4 Nd-147 1 3.63E+4 3.94E+4 2.32E+4 1.42E+8 2.36E+3 39

I I

Hope Creek ODCM Rev 18 Table 2-4 (cont'd)

Pathway Dose Factors - Atmospheric Releases R(io), Vegetation Pathway Dose Factors - CHILD (mrem /yrper uCi/m3) for H-3 and C-14 (m2

mrem /yr per uCi/sec) for others g

M

.itod l

suaid be H-3 4.01E+3 4.01E+3 a.u C 2 2.01E+3 4.01E+3 4.01 E+3 C-14 3.50E+6 7.01'E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 P-32 3.37E+9 1.58E+8 9.30E+7 1.30E+8 Cr-51 6.54E44 1.79E+4 1.19E+5 6.25E+6 1.18E+5 Mn-54 6.61E+8 1.85E+8 5.55E+8 1.76E+8 Fe-55 8.00E+8 4.24E+8 2.40E+8 7.86E+7 1.31E+8 Fe-59 4.01E+8 6.49E+8 1.88E+8 6.76E+8 3.23E+8 Co-57 2.99E+7 2.45E+8 6.04E+7 Co-58 6.47E+7 3.77E+8 1.98E+8 Co-60 3.78E+8 2.10E+9 1.12E+9 Ni-63 3.95E+10 2.11 E+9 1.42E+8 1.34E+9 Zn-65 8.12E+8 2.16E+9 1.36E+9 3.80E+8 1.35E+9 Rb-86 4.52E+8 2.91 E+7 2.78E+8 Sr-89 3.59E+10 1.39E+9 1.03 E+9 i

Sr-90 1.24E+12 1.67E+10 3.15E+11 Y-91 1.87E+7 2.49E+9 5.01 E+5 Zr-95 3.90E+6 8.58E+5 1.23E+6 8.95E+8 7.64E+5 Nb-95 4.10E+5 1.59E+5 1.50E+5 2.95E+8 1.14E+5 Ru-103 1.55E+7 3.89E+7 3.99E+8 5.94E+6 Ru-106 7.45E+8 1.01E+9 1.16E+10 9.30E+7 Ag-110m 3.22E+7 2.17E+7 4.05E+7 2.58E+9 1.74E+7 Sb-124 3.52E+8 4.57E+6 7.78E+5 1.96E+8 2.20E+9 1.23E+8 Sb-125 4.99E+8 3.85E+6 4.62E+5 2.78E+8 1.19E+9 1.05E+8 Te-125m 3.51E+8 9.50E+7 9.84E+7 3.38E+8 4.67E+7 Te-127m 1.32E+9 3.56E+8 3.16E+8 3.77E+9 1.07E+9 1.57E+8 Te-129m 8.54E+8 2.39E+8 2.75E+8 2.51E+9 1.04E+9 1.33E+8 1-131 1.43 E+8 1.44E+8 4.76E+10 2.36E+8 1.28E+7 8.18E+7 Cs-134 1.60E+10 2.63E+10 8.14 E+9 2.92E+9 1.42E+8 5.54E+9 Cs-136 8.06E+7 2.22E+8 1.18E+8 1.76E+7 7.79E+6 1.43E+8 Cs-137 2.39E+10 2.29E+10 7.46E+9 2.68E+9 1.43E+8 3.38E+9 Ba-140 2.77E+8 2.43E+5 7.90E+4 1.45E+5 1.40E+8 1.62E+7 Cc-141 6.35E+5 3.26E+5 1.43E+5 4.07E+8 4.84E+4 Ce-144 1.27E+8 3.98E+7 2.21E+7 1.04E+10 6.78E+6 Pr-143 1.48E+5 4.46E+4 2.41E+4 1.60E+8 7.37E+3 Nd-147 7.16E+4 5.80E+4 3.18E+4 9.18E+7 4.49E+3 40

Hope Creek ODCM Rev 18 l

Table 2-4 (cont'd) i Pathway Dose Factors - Atmospheric Releases R(io), Ground Plane Pathway Dose Factors (m2

mrem /yr per uCi/sec) 1 Nuclide Any Organ H-3 C-14 P-32 Cr-51 4.68E46 Mn-54 l.34E+9 Fe-55 i

Fe-59 2.75E+8 l

Co-58 3.82E+8 Co-60 2.16E+10 Ni-63 Zn-65 7.45E+8 Rb-86 8.98E+6 Sr-89 2.16E+4 Sr-90 Y-91 1.08E+6 Zr-95 2.48E+8 Nb-95 1.36E+8 Ru-103 1.09E+8 Ru-105 4.21E+8 Ag-110m 3.47E+9 Te-125m 1.55E+6 Te-127m 9.17E+4 Te-129m 2.00E+7 I-131 1.72E+7 Cs-134 6.75E+9 Cs-136 1.49E+8 Cs-137 1.04E+10 Ba-140 2.05E+7 Ce-141 1.36E+7 Ce-144 6.95E+7 Pr-143 Nd-147 8.40E+6 41

i Hope Creek ODCM Rev 18 l

l l

APPENDIX A EVALUATION OF DEFAULT MPC VALUES FOR LIQUID EFFLUENTS l

A-1

i Hope Creek ODCM Rev 18 1

l APPENDIX A Evaluation of Default MPC Value for Liquid Radwaste Effluent Radiation Monitors In acconiance with the requirements of Technical Specification 3.3.7.10 the radioactive effluent monitors shall be operable with alarm setpoints established to ensure that the concentration of radioactive material at the discharge point does not exceed the MPC value of 10 CFR 20, Appendix B, Table II, Column 2 (Appendix F). The determination of allowable radionuclide concentration and corresponding alarm setpoint is a function of the individual monitor.

In order to limit the need for routinely having to reestablish the alarm setpoints as a function of changing radionuclide distributions, a default alarm setpoint can be established. This default 1

setpoint can be based on an evaluation of the radionuclide distribution from the 1993 to 1995 release data of the liquid effluents from Hope Creek and the effective MPC value for this distribution.

The effective MPC value for a radionuclide distribution is calculated by the equation:

IC (gamma)

(A.1) i

MPC,

=

C (gamma) i I

MPC, where:

MPC,

= an effective MPC value for a mixture of radionuclides (uCi/ml)

C

= concentration of radionuclide i in the mixture i

MPC,

= the 10 CFR 20, Appendix B, Table II, Column II MPC value for radionuclide i (uCi/ml) Appendix F Considering the average effective MPC values from 1993 thru 1995 releases it is reasonable to select an MPC value of 7.92E-05 uCi/ml as typical ofliquid radwaste discharges. This value will be reviewed and. adjusted as necessary based on the distribution history of effluents from Hope Creek. Using the value of 7.92E-5 uCi/ml to calculate the default alarm setpoint, results in a setpoint that:

(1).Will not require frequent re-adjustment due to minor variations in the nuclide distribution which are typical of routine plant operations, and; A-2

l Hope Creek ODCM Rev 18 (2) Will provide for a liquid radwaste discharge rate (as evaluated for each batch release) that is compatible with plant operations (Refer to Table 1-1).

1.0 Default Setpoint Determination:

Conservative alarm setpoints can be determined through the use of default pararneters. Table 1-1 summarizes all current default values in use for Hope Creek.

A. Liquid Radwaste Monitor (RE4861)

MPC,

CTBD * [1 -CF]

SP <

+ bkg (1.2)

RR Default values from Table 1-1:

MPC,

= 7.92E-5 uci/ml CTBD

= 12000 gpm RR

= 176 gpm (1300 CST)

Bkg

= 0 uci/ml CF

= 0.8 7.92E-5

12000

0.2 N

SP <

+0 176 SP < l.08E-3 ucVmi Correction Factor:

A correction factor.nust be applied to i default setpoint calculation in order to account for radiation monitc uncertaintier and the contribution of non-gamma emitting radionuclides such as H-3, Sr, and Fe.

a. Radiation MonitorInaccuracies:

Hope Creek PSBP 311649 lists a total loop accuracy of 30% for the liquid radwaste radiation monitors. A factor of 0.30 is applied to the defaIllt setpoint to ensure the trip setpoint is reached before the analytical limit is obtained.

A-3

Hope Creek ODCM Rev 18

b. Non-Gamma Emitting Radionuclides:

Non-gamma emitting radionuclides are analyzed on a monthly and quarterly basis from composite samples ofliquid radwaste releases.

Nuclide MPC (uci.ml)

Activity (uci.ml)

Activity / MPC l

H-3 3E-3 1.0E-1 33.3 Fe-55 8E-4 4.7E-4 0.59 l

Sr-89 3E-6 1.6E-6 0.53 Sr-90 3E-7 2.0E-8 0.07 Total 34.5 The values in the table above represent the historical maximum reactor coolant values for non-gamma emitting nuclides (H3 is an assumed maximum). Reactor coolant values were chosen to represent the maximum concentration of non-gamma emitting radionuclides that could be released from Hope Creek station in liquid effluent. The activity values in the table is further diluted by a minimum factor of 68 prior to release to the Delaware River. The minimum dilution factor is obtained by using the minimum cooling tower blowdown flowrate of 12,000 gpm and the maximum release rate of176 gpm.

A conservative correction factor for non-gamma emitting radionuclides can be obtained by using the highest Activity / MPC fraction and the minimum dilution factor as follows:

Correction Factor (non-gamma) = 34.5 / 68 = 0.5 An overall correction factor can be obtained by adding the correction factor for radiation monitor inaccuracies and non-gamma emitting radionuclides as follows:

Overall Correction factor = 0.3 + 0.5 = 0.8 B. Cooling Tower Blowdown Radiation Monitor (RE8817)

The cooling tower blowdown radiation monitor provides an Alarm only function for releases into the environment. The cooling tower blowdown is the final release point for liquid effluents from Hope Creek station to the Delaware River.

SP < MPC,

0.2 SP < 7.92E-5 uci/ml

0.2 SP < 1.58E-5 uci/mi (RE8817)

A4

i Hope Creek ODCM Rev 18 C. Turbine Building Circulating Water Dewatering Sump Radiation Monitor (RE4557)

The Turbine Building Circulating Water Dewatedng Sump Radiation Monitor (RE4557) provides automatic termination ofliquid radioactive releases from the Circulating Water Dewatering Sump.

l The sump pumps discharge to the circulating water system to the cooling tower. Radioactive materials other than tritium are not normally expected to be discharged through this pathway. Plant design and procedures maintain the setpoint at <2 times background rLdiation levels. Releases i

fmm the sump at gamma activity concentrations less than the monitor setpoint are considered continuous releases since inputs to the sump would occur during disc %rge. Releases of activity above the setpoint may be perfonned on a batch basis following sampling and analysis of the sump contents. Hope Creek calculation SP-0004 established a serpoint for the monitor at 1.4E-02 uCi/ml based on a postulated release ofreactor steam into the sump. Using the MPCe determined for Liquid Radwaste and Cooling Tower Blowdown monitors, a more conservative maximum default value for batch releases can be determined:

MPC,

CTBD * [1 -CF]

SP <

+ bkg (1.2)

RR Default values from Table 1-1:

MPC,

= 7.92E-5 uci/ml CTBD

= 12000 gpm RR

= 100 gpm Bkg

= 0 uci/mi CF

= 0.8 7.92E-5

12000

0.2 SP <

+0 100 SP < 1.90E-3 sci /ml (batch releases only)

For continuous releases, the maximum setpoint should be less than 2.4E-6 uci/ml above background to limit dose consequences from this pathway.

A-5 9

Hope Creek ODCM Rev 18 TABLE A-1 CALCULATION OF EFFECTIVE MPC HOPE CREEK

^

1993 ACTIVITY 1994 ACTIVITY 1995 ACTIVITY NUCLIDE MPC RELEASED (Ci)

RELEASED (Ci) RELEASED (Ci)

Na-24 3.0E-05 1.20E-04 N/D 2.15E-05 Cr-51 2.0E 03 1.57E41 7.91E-02 1.29E-01

' 1.0E-04 630E-02 1.10E-01 4.12E-01 Mn-54 As-76 2.0E-05 1.03E-05 1.62E-04 8.42E-05 Co-58 9.0E-05 1.35E-03 3.04E-03 2.68E-02 Fe-59 5.0E-05 8.74E-03 1.14E-02 1.12E-01 Co-60 3.0E-05 1.15E-02 6.49E-02 8.48E-02 Zn-65 1.0E-04 7.01E-02 7.79E-02 8.77E-02 Ag-110m 3.0E-05 2.16E-03 1.90E-03 7.42E-03 Zr-97 2.0E-05 1.45E-05 N/D 3.45E-05 Tc-99m 3.0E-03 1.79E-03 1.45E-03 8.68E-04 Cs-134 9.0E-06 N/D 1.11E-06 1.51E-04 Cs-137 2.0E-05 4.85E-05 1.33E-05 2.97E-04 Zn-69m 6.0E-05 2.33E-05 7.22E-05 N/D Nb-97 9.0E-04 4.33E-04 2.32E-04 N/D In-ll5m 4.0E-04 N/D N/D 2.90E-05 H-3 3.0E-03 6.17E+01 3.33E+01 4.61E+01 Fe-55 8.0E-04 4.51E-02 1.86E-01 5.44E-01 Total Curies 3.16E-01 3.50E-01 8.61E-01 (Gamma)

SUM (Ci/MPCi) 2.%E+03 4.42E+03 1.07E+04 (Gamma)

SUM (Ci/MPCi) 2.06E+04 1.13E+04 1.60E404 (Non-Gamma)

MPCe(uCi/ml) 1.53E-04 7.92E-05 8.04E-05 l

N/D=Not detected A-6

Hope Creek ODCM Rev 17 APPENDIX B TECHNICAL BASIS FOR EFFECTIVE DOSE FACTORS LIQUID RADIOACTIVE EFFLUENTS

Hope Creek ODCM Rev 18 APPENDIX B i

Technical Basis for Effective Dose Factors -

Liquid EfRuent Releases lhe radioactive liquid efBuents from Hope Creek from 1993 through 1995 were evaluated to determine the dose contribution of the radionuclide distribution. This analysis was performed to evaluate the use of a limited dose analysis for u mining environmental doses, pmviding a simplified method of-determining compliance with the dose limits of Technical Speci6 cation 3.11.1.2. For the expected radionuclide distribution of effluent from Hope Creek during 1993 to 1995, the contmiling organ is the liver. The calculated liver dose is predominately a function of the Zn45 and Fe-55 releases. The radionuclides, Zn45 and Fe-55 also contribute the large majority of the calculated total body dose. The results of this evaluation are presented in Table B-1.

For purposes of simplifying the details of the dose calculation pmcess, it is conservative to identify a controlling, dose significant radionuclide and limit the calculation process to the use of the dose conversion factor for this nuclide. Multiplication of the total release (i.e., cumulative activity for all radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.

For the evaluation of the maximum organ dose, it is conservative to use the Zn45 dose conversion factor (5.13E5 mrem /hr per uci/ml).

By this approach, the maximum organ dose will be overestimated since this nuclide has the highest organ dose fraction of all the radionuclides evaluated. For the total body calculation, the Zn45 dose factor (2.32E5 mrem /hr per uCi/ml, total body) is the highest among the identified dominant nuclides.

i For evaluating compliance with the dose limits of technical Specification 3.11.1.2, the following simplified equations may be used:

Total Body 8.35E-04

Vol E. =

A,,.

C, (B.1)

CTBD B-2

Hope Creek ODCM Rev 18 where:

D.

= dose to the total body (mrem)

A,,.

= 2.32ES, total body ingestion dose conversion factor for Zn-65 where A is dose conversion factor, i is isotope which is Zn-65, and TB is the total body (mrem /hr per uCi/ml) l VOL

= elume ofliquid effluent released (gal)

C,-

= total concentration of all radionuclides (uCi/ml)

CTBD

= avenge cooling tower blowdown diessige rate during release period (gal / min) 8.35E-04 = conversion factor (1.67E-2 hr/ min) and the near field dilution factor 0.05 Substituting the value for the Zn-65 total body dose conversion factor, the equation simplified to:

1.94E+ 2

VOL D. =

I C,

@l)

CTBD Maximum Organ 835E-4

VOL

Aio, Liver Dmax =

I C, (B3)

CTBD Where:

Dmax = maximum organ dose (mrem)

Aio, 5.13ES, liver ingestion dose conversion factor for Zn-65 where A is dose conversion

=

Liver factor, i is isotope which is Zn-65 and O is maximum organ which is the liver (mrem /hr per uCi/ml).

Substituting the value for Aio the equation simplifies to:

4.28E2

VOL Dmax =

I C, (B.4)

CTBD Tritium is not included in the limited analysis dose assessment for liquid releases, because the potential dose resulting from normal reactor releases is relatively negligible.

B-3

Hope Creek ODCM Rev 18 -

Near Field Dilution Factor i

The near field dilution factor stems from NUREG-0133, Section 4.1.

For plants with cooling towers, such as Hope Creek, a dilution factor is applicable so that the product of the average blowdown flow (in CFS) and the dilution factor is 1000 cfs or less. The average minimum cooling tower blowdown for Hope Creek is 1.90E4 GPM (from FSAR 11.2). This converts to 42 CFS, for conservatism a dilution factor of 20 will be used, giving a dilution flow of 880 CFS. This near field dilution factor of 20 is inverted to a multiple of 0.05, multiplied times the liquid effluent dose equations.

t 4

B-4

Hope Creek ODCM Rev 18 !

TABLE B-1 Adult Dose Contributions Fish and Invertebrate Pathways Hope Greek Nuclide Release TB Dose GI LLI Dose

' Liver Dose Year (Ci)

Frac.

Fe-55 4.51E-2 0.02 0.03 0.04 1993 Fe-55 1.86E-2 0.07 0.08 0.13 1994 i

i Fe-55 5.44E-1 0.15 0.10 0.24 1995 Mn-54 6.30E-2 0.04 0.01 1993 Mn-54 1.10E-1 0.01 0.06 0.02 1994 Mn-54 4.12E-1 0.02 0.07 0.03 1995 Co-58 1.35E-3 1993 Co-58 3.04E-3 1994 Co-58 2.68E-2 1995 Fe-59 8.74E-3 0.04 0.18 0.04 1993 Fe-59 1.14E-2 0.04 0.17 0.04 1994 Fe-59 1.12E-1 0.22 0.53 0.22 1995 Co-60 1.15E-2 0.01 1993 Co-60 6.49E-1 0.01 0.05 1994 Co-60 8.48E-2 0.01 0.02 1995 Zn-65 7.01E-2 0.93 0.69 0.90 1993 Zn-65 7.79E-2 0.87 0.61 0.81 1994 Zn-65 8.77E-2 0.60 0.23 0.51 1995

= Less than 0.01 B-5

Hope Creek ODCM Rev 18 l

l APPENDIX C TECHNICAL BASIS FOR EFFECTIVE DOSE FACTORS GASEOUS RADIOACTIVE EFFLUENTS i

1 Y

1 a

1 Hope Creek ODCM Rev i8 APPENDIX C Teebnical Basis for Effective Dose Factors -

Gaseous Radioactive Emments Overview 1he evaluation of doses due to releases of radioactive material to the atmosphere can be simplified by the use of effective dose transfer factors instead of using dose factors which are radionuclide specific, These effective factors, which are based on typical radionuclide distributions of releases, can be applied i

to the total radioactivity releases to approximate the dose in the environment. Instead of having to j

perform individual radionuclide dose analysis only a single multiplication (i.e., Keff, Meff, or Neff times I

the total quantity of radioactive material releases) would be needed. The approach provides a reasonable estimate of the actual dose while eliminating the need for a detailed calculation technique.

Determination of Effective Dose Factors Effective dose transfer factors are calculated by the following equations:

4 K,, = I (K

f)

(C.1)

Where:

K,,

= the effective total body factor due to gamma emissions from all noble gases released.

K,

= the total body dose factor due to gamma emissions from each noble gas radionuclide i released.

f,

= the fractional abundance of noble gas radionuclide i relative to the total noble gas activity.

( L + 1.1M,,) = I (( L, + 1.1M, )

f, )

(C.2) where:

l l

(L + 1.lM,,) = the effective skin dose factor due to beta and gamma emissions from all noble gases released.

(L + 1.1 M ) = the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i

i released.

I C-1

r 1

l Hope Creek ODCM Rev 18 M, = I ( M

f)

(C.3) where:

M,

= the effective air dose factor due to gamma emissions from all noble gases released.

M,

= the air dose factor due to gamma emissions from each noble gas radionuclide i released.

N, = I (N,

f)

(C.4) where:

4 N,

= the effective air dose factor due to beta emissions from all noble gases released.

I N,

= the air dose factor due to beta emissions from each noble gas radionuclide i released.

Normally, it would be expected that past radioactive effluent data would be used for the determination of the effective dose factors. However, the noble gas releases from Hope Creek have a short history and with continued excellent fuel performance, has hampered efforts in collecting and detecting appreciable noble gas mixes of radionuclides. So, to provide a reasonable basis for the derivation of the effective noble gas dose factors, the source terms from ANSI N237-1976/ANS-18.1, " Source Term Specifications", Table 5 has been used as representing a typical distribution. The effective dose factors as derived are presented in Table C-1.

Application To provide an additional degree of conservatism, a factor of 0.50 is introduced into the dose calculation process when the effective dose transfer factor is used. This conservatism provides additional assurance that the evaluation of doses by the use of a single effective factor will not significantly underestimate any actual doses in the environment.

For evaluating compliance with the dose limits of Technical Specification 3.11.2.2, the following simplified equations may be used:

3.17E-08 D, =

X/Q

M,* I Q, (C.5) 0.50 3.17E-08 D,=

X/Q

N,

I Q, (C.6) 0.50 C-2

Hope Creek ODCM Rev 18 Where:

D,

= air dose due to gamma emissions for the cumulative release of all noble gases (mrad)

D,

= air dose due to beta emissions for the cumulative release of all noble gases (mrad)

X/Q

= atmospheric dispersion to the controlling site boundary (sec/m')

M,

= 8.1E3, effective ganana-air dose factor (mrad /yr per uCi/m')

N,

= 8.5E3, effective beta air dose factor (mrad /yr per uCi/m')

Q,

= csmulative release for all noble gas radionuclides (uCi) 3.17E-08 = conversion factor (yr/sec) l 0.50

= conservatism factor to account for the variability in the effluent data Combining the constants, the dose calculation equations simplify to:

D,

= 5.14E-4

X/Q

I Q, (C.7)

D.

= 5.39E-4

X/Q

I Q, (C.8)

The effective dose factors are to be used on a limited basis for the purpose of facilitating the timely assessment of radioactive effluent releases, panicularly during periods of computer malfunction where a detailed dose assessment may be unavailable.

i C-3 l

l 1

I l

l i

i Hope Creek ODC.M Rev 18 l

l TABLE C-1 Effective Dose Factors I

Noble Gases - Total Body and Skin

{

Total Body Effective Skin Effective K,

(L + 1.1 M, )

Radionuclide f

(mrem /yr per uci/m')

Kr83m 0.01 Kr85m

'0.01 1.0El 2.8E1 Kr87 0.04 2.4E2 6.6E2 Kr88 0.04 5.9E2 7.6E2 Kr89 0.27 4.5E3 7.9E3 Xe133 0.02 5.9E0 1.4El Xel35 0.05 9.0El 2.0E2 Xel35m 0.06 1.9E2 2.6E2 Xel37 0.31 4.4E2 4.3E3 Xel38 0.19 1.7E3 2.7E3 Total 7.8E3 1.7E4 Noble Gases - Air Total Body Effective Skin Effective K.,

(L + 1.1 M,y)

Radionuclide J

(mrem /yr per uci/m')

i Kr83m 0.01 3.0E0 Kr85m 0.01 1.2El 2.0El Kr87 0.04 2.5E2 4.lE2 Kr88 0.04 6.1E2 1.2E2 Kr89 0.27 4.7E3 2.9E3 Xel33 0.02 7.0E0 2.lEl Xel35 0.05 9.6El 1.2E2 Xel35m 0.06 2.0E2 4.4El Xe137 0.31 4.7E2 3.9E3 Xel38 0.19 1.8E3 9.0E2 Total 8.lE3 8.4E3 1

1

Based on noble gas distribution from ANSI N237-1976/ANS-18.1," Source Term l

Specification".

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C-4 l.

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Hope Creek ODCM Rev 18 1

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APPENDIX D l

TECHNICAL BASIS FOR EFFECTIVE DOSE PARAMETERS GASEOUS RADIOACTIVE EFFLUENTS l

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1 Hope Creek ODCM Rev 18 APPENDIX D Technical Basis for Effective Dose Parameters Gaseous Radioactive Effluent Releases The pathway dose factors for the controlling infant age group were evaluated to determine the controlling pathway, organ and radionuclide. This analysis was performed to provide a simplified method for determining compliance with Technical Specification 3.11.2.3. For the infant age group, the controlling pathway is the grass - cow - milk (g/c/m) pathway. An infant receives a greater radiation dose from the g/c/m pathway than any other pathway. Of this g/c/m pathway, the maximum exposed organ including the total body, is the thyroid, and the highest dose contributor is radionuclide I-131. The results of this evaluation are presented in Table D-1.

1 For purposes of simplifying the details of the dose calculation process, it is conservative to identify a controlling, dose significant organ and radionuclide and limit the calculation process to the use of the dose conversion factor for the organ and radionuclide. Multiplication of the total release (i.e., cumulative activity for all radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.

For the evaluation of the dose commitment via a controlling pathway and age coup, it is conservative to use the infant, g/c/m, thyroid, I-131 pathway dose factor (1.67E12m ' mrem /yr per uCi/sec). By this 2

approach, the maximum dose commitment will be overestimated since I-131 has the highest pathway dose factor of all radionuclides evaluated.

For evaluating compliance with the dose limits of Technical Specification 3.11.2.3, the following simplified equation may be'used:

D.,

= 3.17E-8

W

RI-131

IQ (D.1)

Where:

D.,

= maximum organ dose (mrem)

W

= atmospheric dispersion parameter to the controlling location (s) as identified in Table 2-3.

X/Q

= Atmospheric dispersion for inhalation pathway (sec/m')

D/Q

= atmospheric disposition for vegetation, milk and ground plane exposure pathways (m)

Q3

= cumulative release over the period ofinterest for radioiodines and particulates (uCi).

3.17E-S

= conversion factor (yr/sec)

RI-131

= I-131 dose parameter for the thyroid for the identified controlling pathway.

= 1.05E12, infant thyroid dose parameter with the grass - cow - milk pathway controlling 2

(m mrem /yr per uCi/sec)

D-1 4

T Hope Creek ODCM Rev 18 The ground plane exposure and inhalation pathways need not be considered when the above simplifie calculational method is used because of the overall negligible contribution of these pathways to the total thyroid dose. It is recognized that for some particulate radionuclides (e.g., Co-60 and Cs-137), the ground exposure pathway.may represent a higher dose contribution than either the vegetation or milk pathway.

However, use of the I-131 thyroid dose parameter for all radionuclides will maximize the organ dose calculation, especially considering that no other radionuclides has a higher dose parameter for any organ via any pathway than I-131 for the thyroid via the milk pathway.

The location of exposure pathways and the maximum organ dose calculation may be based on the available pathways in the surrounding environment of Hope Creek as identified by the annual land-use census (Technical Specification 3.12.2). Otherwise, the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-3.

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D-2

Hope Creek ODCM Rev 18 TABLE D-1 Infant Dose Contributions Fraction of Total Organ and Body Dose PATHWAYS Target Organs Grass - Cow - Milk Ground Plane Total Body

~

0.02 0.15 Bone 0.23 0.14 Liver 0.09 0.15 Thyroid 0.59 0.15 4

Kidney 0.02 0.15 Lung 0.01 0.14 GI LLI 0.02 0.15 TABLE D-2 Fraction of Dose Contribution by Pathway Pathway Frac Grass-Cow-Milk 0.92 Ground Plane 0.08 Inhalation N/A D-3

Hope Creek ODCM Rev 18 APPENDIX E RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM -

SAMPLE TYPE, LOCATION AND ANALYSIS l

I D-4

Hope Creek ODCM Rev 18 APPENDIX E SAMPLE DESIGNATION Samples are identified by a three part code. The first two letters are the power station identification code, in this case "SA". The next three letters are for the media sampled.

AIO m AirIodine IDM = Immersion Dose (TLD)

APT = Air Particulates MLK = Milk ECH = Hard Shell Blue Crab PWR = Potable Water (Raw)

ESF = Edible Fish PWT = Potable Water (Treated)

ESS = Sediment SWA = Surface Water WWA= Well Water The last four symbols are a location code based on direction and distance from the site. Of these, the first two represent each of the sixteen angular sectors of 22.5 degrees centered about the reactor site.

Sector one is divided evenly by the north axis and other sectors are numbered in a clockwise direction; i.e.,2=NNE,3=NE,4=ENG, etc. The next digit is a letter which represents the radial distance from the plant:

S

= On-site location E

= 4-5 miles off site l

A

= 0-1 miles off-site F

= 5-10 miles off-site B

= l-2 miles off-site G

= 10-20 miles off-site C

= 2-3 miles off-site H

= > 20 miles off-site D

= 3-4 miles off-site The last number is the station numerical designation within each sector and zone; e.g.,1,2,3,... For example; the designation SA-WWA-5DI would indicate a sample in the SGS program (SA), consisting of well water (WWA), which had been collected in sector number 5, centered at 90' (due east) with respect to the reactor site at a radical distance of 3 to 4 miles off-site,(therefore, radial distance D). The number 1 indicated e. t thk is sampling station #1 in that particular sector.

The Radiological Environmental Monitoring Program is a common program for both Salem and Hope Creek Generating Stations.

E-2

Hope Creek ODCM Rev 18 SAMPLING LOCATIONS All sampling locations and specific information about the individual locations are given in Table E-1.

Maps E-1 and E-2 show the locations of sampling stations with respect to the site.

TABLE E-1 A. Direct Radiation Monitoring Locations (IDM)

STATION CODE STATION LOCATION ISI 0.55 mi. N of vent 2S2 0.4 mi. NNE of vent 2S4' O.59 mi. NNE of vent 3S1 0.58 mi. NE of vent 4S1 0.60 mi ENE ofvent SSI 1.0 mi. E of vent; site access road 6S2 0.21 mi. ESE of vent; observation building

7S1 0.12 mi. SE of vent; station personnel gate

10S1 0.14 mi. SSW of vent; cire water bldg.

11S1 0.09 mi. SW of vent; service water bldg.

ISSI 0.57 mi. NW of vent 16S1 0.54 mi. NNW of vent 4D2 3.7 mi. ENE of vent; Alloway Creek Neck Road SD1 3.5 mi. E of vent; local farm 10D1 3.9 mi. SSW of vent; Taylor's Bridge Spur 14D1 3.4 mi. WNW of vent; Bay View, Delaware 15D1 3.8 mi. NW of vent; Rt 9, Augustine Beach, DE.

2El 4.4 mi. NNE of vent; local farm 3El 4.1 mi. NE of vent; local farm 9El 4.2 mi. S of vent i1E2 5.0 mi. SW of vent 12El 4.4 mi. WSW of vent; Thomas Landing 13El 4.2 mi. W of vent; Diehl House Lab 16El 4.1 mi. NNW of vent; Port Penn IFl 5.8 mi.N of vent; Fort Elfsborg 2F2 8.7 mi. NNE of vent; Salem Substation 2FS 7.4 mi.NNE of vent; Salem High School 2F6 7.3 mi.'NNE of vent; PSE&G Training Center E I

Hope Creek ODCM Rev 18 TABLE E-1(Continued)

A. Direct Radiation Monitoring Locations (IDM)

STATION CODE STATION LOCATION 3F2 5.1 mi. NE of vent; Hancocks Bridge Munc. Bldg 3F3 8.6 mi. NE of vent; Quinton Township School 4F2 6.0 mi. ENE of vent; Mays Lane, Harmersville SF1 6.5 mi. E of vent; Canton 6F1 6.4 mi. ESE of vent; Stow Neck Road 7F2 9.1 mi. SE of vent; Bayside, NJ 10F2 5.8 mi. SSW of vent; Rt. 9 1IFl 6.2 mi SW of vent; Taylors Bridge, DE.

12F1 9.4 mi. WSW of vent; Townsend Elementary School 13F2 6.5 mi. W of vent; Odessa, DE.

13F3 9.3 mi. W of vent; Redding Middle School 13F4 9.8 mi. W of vent; Middletown, DE.

14F2 6.6 mi. WNW of vent; Boyds Corner 15F3 5.4 mi. NW of vent 16F2 8.1 mi. NNW of vent; Delaware City Public School 1G3 19 mi. N of vent; N. Church St. Wilmington, DE 3G1 17 mi. NE of vent; local farm 10G1 12 mi SSW of vent; Smyrna, Delaware 16G1 15 mi. NNW of vent; Wilmington Airport 3HI 32 mi.NE of vent; National Park, NJ 14G1 11.8 mi. WNW of vent; Rte 286; Bethel Church Road; Delaware

TLD locations will be maintained by site area monitoring program B. Air Sampling Locations (AIO, APT)

STATION CODE STATION LOCATION SSI 1.0 mi. E of vent; site access road SDI 3.5 mi. E of vent; local farm 16El 4.1 mi. NNW of vent; Port Penn 1F1 5.8 mi. N of vent; Fort Elfsborg 2F6 7.3 mi. NNE of vent; PSE&G Training Center 14G1 11.8 mi. WNW of vent; Rte 286; Bethel Church Road; Delaware

~

E-4

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Hope Creek ODCM Rev 18 TABLE E-1(Continued)

C. Surface Water Locations (SWA) - Delaware River STATION CODE STATION LOCATION 11A1 0.2 mi. SW of vent; Salem Outfall Area

'12C1 2.5 mi. WSW of vent; West bank of Delaware River 7El 4.5 mi. SE of vent; Delaware River 16F1 6.9 mi.NNW of vent; C&D Canal D. Ground Water Locations (WWA)

STATION CODE STATION LOCATION -

No public drinking water samples or irrigation water samples are taken as these pathways are not directly affected by liquid efYluents discharged from Hope Creek or Salem Generating Stations.

E. Drinking Water Locations (PWR,PWT)

STATION CODE STATION LOCATION No public drinking water samples or irrigation water samples are taken as these pathways are not directly affected by liquid effluents discharged from Hope Creek or Salem Generating Stations.

F. Water Sediment Locations (ESS)

STATION CODE STATION LOCATION 11Al 0.2 mi. SW of vent; Salem outfall area 15A1 0.3 mi. NW of vent; Hope Creek outfall area 16Al 0.7 mi. NNW of vent; South Storm Drain outfall 12Cl 2.5 mi. WSW of vent; West bank of Delaware river 7El 4.5 mi. SE of vent; I mi West of Mad Horse River 16F1 6.9 mi. NNW of vent; C&D Canal 6S2 0.2 mi. ESE of vent; observation building O Milk Sampling Locations (MLK)

STATION CODE STATION LOCATION 2F7 5.7 mi.NNE of vent; local farm 1IF3 5.3 mi. SW of vent; Townsend DE.

14F4 7.6 mi. WNW of vent; local farm 3G1 17 mi. NE of vent; local farm E-5

Hope Creek ODCM Rev 18 TABLE E-1 (Continued)

H. Fish and Invertebrate Locations (ESF,ECH) -

STATION CODE,

STATION LOCATION 11Al 0.2 mi. SW of vent; Salem outfall area 12Cl 2.5 mi. WSW of vent; West bank of Delaware River q

7El 4.5 mi. SE of vent; I mi West of Mad Horse Creek

1. Food Product. Locations STATION CODE STATION LOCATION The Delaware River at the location of Salem and Hope Creek Nuclear Power Plants is a brackish water source. No irrigation of food products is perfortned using water in the vicinity from which liquid plant wastes have been discharged.

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Hope Creek ODCM Rev 18 SAMPLES COLLECTION AND ANALYSIS Sample Collection Method Analysis Air Particulate.

Continuous low volume Gross Beta analysis air sampler. Sample on each weekly collected every week sample. Gamma along with the filter spectrometry shall change.

be performed if gross beta exceeds 10 times the yearly mean of the control station value. Samples shall be analyzed 24 hrs or more after collection to allow for radon and thorium daughter decay. Gamma isotopic analysis on quarterly composites.

AirIodine A TEDA impregnated Iodine 131 analysis charcoal cartridge is are performed on connected to air each weekly sample.

particulate air sampler and is collected weekly at filter change.

Crab and Fish Two batch samples are Gamma isotopic scaled in a plastic analysis of edible bag orjar and frozen portion on collection.

semi-annually or when in season.

Sediment A sediment sample is Gamma isotopic taken semi annually.

analysis semi-annually.

Direct 2 TLD's will be Gamma dose quarterly.

collected from each location quarterly.

E-7

Hope Creek ODCM Rev 18 SAMPLE COLLECTION AND ANALYSIS (Cont'd) l l

l Sample Collection Method Analysis Milk Sample of fresh milk Gamma isotopic is collected for each analysis and I-131 farm semi-monthly when analysis on each cows are in pasture, sample on collection.

I monthly at other times.

Water Sample to be collected Gammaisotopic

Potable, monthly pmviding winter monthly H 3 on l

Surface) icing conditions allow.

quarterly surface sample, monthly on ground water sample.

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Hope Creek ODCM Rev 18 Figure E-1 ONSITE SAMPLING LOCATIONS I

16 C/

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MININUM CXCLUS{

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UM EXCLUS!

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DELA RE Ri ER 6

11 7

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E-9

Hope Creek ODCM Rev 18 Figure E-2 sALEu t.no HorE & aeneurne srArea MDMOGCAL ENVm0MHENTAL MONfiDRn@ Ml00MM OFF-3ITE sAMPLh6 LOCATION

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Hope Creek ODCM Rev 18 9

APPENDIX F MAXIMUM PERMISSIBLE CONCENTRATIONS LIQUID EFFLUENTS

\\

I F-1

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i 2

Hope Creek ODCM Rev 18 The following radionuclide concentrations were obtained from 10 CFR 20 Appendix B, Table II, Column 2 as revised January 1,1991.

Table F-1 Maximum Permissible Concentrations Element Isotope Soluble Conc Insoluble Conc.

(uci/ml)

Actinium (89)

Ac-227 2E-6 3E-4 Ac-228 9E-5 9E-5 Americium (95)

Am-241 4E-6 3E-5 Am-242m 4E-6 9E-5 Am-242 1E-4 1E-4 Am-243 4E-6 3E-5 Am-244 SE-3 SE-3 Antimony (51)

Sb-122 3E-5 3E-5 Sb-124 2E-5 2E ~

Sb-125 1E-4 Arsenic (33)

As-73 SE-4 SE-4 As-74 SE-5 SE-5 As-76 2E-5 2E-5 As-77 8E-5 8E-5 Astatine (85)

At-211 2E-6 7E-5 Barium (56)

Ba-131 2E-4 2E-4 Ba-140 3E-5 2E-5 Berkelium (97)

Bk-249 6E-4 6E-4 Bk-250 2E-4 2E-4 Beryllium (4)

Be-7 2E-3 2E-3 Bismuth (83)

Bi-2%

4E-5 4E-5 Bi-207 6E-5 6E-5 Bi-210 4E-5 4E-5 Bi-212 4E-4 4E-4 Bromine (35)

Br-82 3E-4 4E-5 Cadmium (48)

Cd-109 2E-4 2E-4 Cd-l1Sm 3E-5 3E-5 Cd-l15 3E-5 4E-5 Calcium (20)

Ca-45 9E-6 2E-4 Ca-47 SE-5 3E-5 F-2

Hope Creek ODCM Rev 18 Table F-1 (Continued)

Element Isotope Soluble Conc.

Insoluble Conc.

(uci/ml)

(

Californium (98)

Cf-249 -

4E-6 2E-5 i

Cf-250 lE-5 3E-5 Cf-251 4E-6 3E-5 Cf-252 7E-6 7E-6 Cf-253 1E-4 1E-4 Cf-254 IE-7 IE-7 Carbon (6)

C-14 8E-4 Cerium (58)

Ce-141 9E-5 9E-5 Cc-143 4E-5 4E-5 Ce-144 IE-5 IE-5 Cesium (55)

Cs-131 2E-3 9E-4 Cs-134m 6E-3 1E-3 Cs-134 9E-6 4E-5 Cs-135 1E-4 2E-4 Cs-136 9E-5 6E-5 Cs-137 2E-5 4E-5 Chlorine (17)

Cl-36 8E-5 6E-5 Cl-38 4E-4 4E-4 Chromium (24)

Cr-51 2E-3 2E-3 Cobalt (27)

Co-57 5E-4 4E-4 Co-58m 3E-3 2E-3 Co-58 1E-4 9E-5 Co-60 SE-5 3E-5 Copper (29)

Cu-64 3E-4 2E-4 Curium (%)

Cm-242 2E-5 2E-5 Cm-243 5E-6 2E-5 Cm-244 7E-6 3E-5 j

Cm-245 4E-6 3E-5 Cm-246 4E-6 3E-5 Cm-247 4E-6 2E-5 Cm-248 4E-7 1E-6 Cm-249 2E-3 2E-3 Dyspre:;ium (66)

Dy-165 4E-4 4E-4 i

Dy-166 4E-5 4E-5 i

'~

F-3

Hope Creek ODCM Rev 18 Table F-1 (Continued)

Element Isotope Soluble Conc.

Insoluble Conc.

(uci/ml)

Dysprosium (66) 1-135 4E-6 7E-5 Iridium (77)

Ir-190 2E-4 2E-4 Ir-192 4E-5 4E-5 Ir-194 3E-5 3E-5 Iron (26)

Fe-55 8E-4 2E-3 Fe-59 6E-5 SE-5 Lanthanum (57)

La-140 2E-5 2E-5 Einstemium(99)

Es-253 2E-5 2E-5 Es-254m 2E-5 2E-5 Es-254 IE-5 IE-5 Es-255 3E-5 3E-5 Erbium (68)

Er-169 9E-5 9E-5 Er-171 1E-4 1E-4 Europium (63)

Eu-l52 (9.2 hrs) 6E-5 6E-5 Eu-152 (13 yrs) 8E-5 8E-5 Eu-154 2E-5 2E-5 Eu-155 2E-4 2E-4 Fermium (100)

Fm-254 1E-4 1E-4 Fm-255 3E-5 3E-5 Fm-256 9E-7 9E-7 Fluorine (9)

F-18 8E-4 SE-4 Gadolinium (64)

Gd-153 2E-4 2E-4 Gd-159 8E-5 8E-5 Gallium (31)

Ga-72 4E-5 4E-5 Germanium (32)

Ge-71 2E-3 2E-3 Gold (79)

Au-l%

2E-4 IE-4 Au-198 SE-5 5E-5 Au-199 2E-4 2E-4 Hafnium (72)

Hf-181 7E-5 7E-5 Holmium (67)

Ho-166 3E-5 3E-5 Hydrogen (3)

H-3 3E-3 3E-3 Indium (49)

In-113m 1E-3 1E-3 in-114m 2E-5 2E-5 In-115m 4E-4 4E-4 In-115 9E-5 9E-5 F-4

1 l

Hope Creek ODCM Rev 18 Table F-1 (Continued)

Element Isotope Soluble Conc.

Insoluble Conc.

(uci/ml)

Iodine (53) 1-125 2E-7 2E-4 l

I-126 3E-7 9E-5 j

l I-129 6E-8 2E-4 1

I-131 3E-7 6E-5 I-132 8E-6 2E-4 I-133 1E-6 4E-5 I-134 2E-5 6E-4 Lead (82)

Pb-203 4E-4 4E-4 i

Pb-210 1E-7 2E-4 Pb-212 2E-5 2E-5 Lutetium (71)

Lu-177 1E-4 1E-4 Manganese (25)

Mn-52 3E-5 3E-5 Mn-54 1E-4 1E-4 Mn 56 1E-4 1E-4 Mercury (80)

Hg-197m 2E-4 2E-4 Hg-197 3E-4 SE-4 Hg-203 2E-5 1E-4 Molybdenum (42)

Mo-99 2E-4 4E-5 Neodymium (60)

Nd-144 7E-5 8E-5 Nd-147 6E-5 6E-5 Nd-149 3E-4 3E-4 Neptunium (93)

Np-237 3E-6 3E-5 Np-239 1E-4 1E-4 Nickel (28)

Ni-59 2E-4 2E-3 Ni-63 3E-5 7E-4 Ni-65 1E-4 1E-4 Niobium (41)

Nb-93m 4E-4 4E-4 Nb-95 1E-4 1E-4 Nb-97 9E-4 9E-4 Osmium (76)

Os-185 7E-5 7E-5 Os-191m 3E-3 2E-3 Os-191 2E-4 2E-4 Os-193 6E-5 5E-5 Palladium (46)

Pd-103 3E-4 3E-4 Pd-109 9E-5 7E-5 i

'~

F-5

Hope Creek ODCM Rev 18 Table F-1 (Continued)

[

Element Isotope Soluble Conc.

Insoluble Conc.

(uci/ml)

Phosphorus (15)

P-32 2E-5 2E-5 Platinum (78)

Pt-191 1E-4 1E-4 Pt-193m lE-3 1E-3 Pt-193 9E-4 2E-3 Pt-197m lE-3 9E-4 Pt-197 1E-4 1E-4 Plutonium (94)

Pu-238 SE-6 3E-5 Pu-239 SE-6 3E-5 Pu-240 SE-6 3E-5 Pu-241 2E-4 1E-3 l

Pu-242 SE-6 3E-5 i

Pu-243 3E-4 3E-4 Polonium (84)

Po-210 7E-7 3E-5 Potassium (19)

K-42 3E-4 2E-5 Praseodymium (59)

Pr-142 3E-5 3E-5 Pr-143 SE-5 SE-5 Promethium (61)

Pm-147 2E-4 2E-4 Pm-149 4E-5 4E-5 Protactinium (91)

Pa-230 2E-4 2E-4 Pa-231 9E-7 2E-5 Pa-233 1E-4 1E-4 i

l Radium (88)

Ra-223 7E-7 4E-6 Ra-224 2E-6 5E-6 Ra-226 3E-8 3E-5 l

Ra-228 3E-8 3E-5

~

Rhenium (75)

Re-183 6E-4 3E-4 L

Re-186 9E-5 5E-5 Re-187 3E-3 2E-3 Re-188 6E-5 3E-5 Rhodium (45)

Rh-103m 1E-2 1E-2 l

Rh-105 1E-4 1E-4

~

L Rubidium (37)

Rb-86 7E-5 2E-5 Rb-87 1E-4 2E-4 Ruthenium (44)

Ru-97 4E-4 3E-4 l

Ru-103 8E-5 8E-5 l

l l

F-6

1 Hope Creek ODCM Rev 18 1

Table F-1 (Continued)

Element Isotope Soluble Conc.

Insoluble Conc.

(uci/ml)

Ru-105 1E-4 1E-4 Ru-106 IE-5 IE-5 Samarium (62)

Sm-147 6E-5 7E-5 Sm-151 4E-4 4E-4 Sm-153 8E-5 8E-5 Scandium (21)

Sc-46 4E-5 4E-5 Sc-47 9E-5 9E-5 l

Sc-48 3E-5 3E-5 l

Selenium (34)

Se-75 3E-4 3E-4 Silicon (14)

Si-31 9E-4 2E-4 Silver (47)

Ag-105 IE-4 IE-4 Ag-110m 3E-5 3E-5 Ag-111 4E-5 4E-5 Sodium (11)

Na-22 4E-5 3E-5 Na-24 2E-4 3E-5 Strontium (38)

Sr-85m 7E-3 7E-3 Sr-85 1E-4 2E-4 Sr-89 3E-6 3E-5 l

Sr-90 3E-7 4E-5 Sr-91 7E-5 5E-5 Sr-92 7E-5 6E-5 Sulfur (16)

S-35 6E-5 3E-4 Tantalum (73)

Ta-182 4E-5 4E-5 i

Technetium (43)

Tc-96m lE-2 IE-2 Tc-%

1E-4 SE-5 l

Tc-97m 4E-4 2E-4 l

Tc-97 2E-3 8E-4 Tc-99m 6E-3 3E-3 Tc-99 3E-4 2E-4 Tellurium (52)

Te-125m 2E-4 1E-4 Te-127m 6E-5 5E-5 Te-127 3E-4 2E-4 Te-129m 3E-5 2E-5 Te-129 8E-4 8E-4 Te-131m 6E-5 4E-5

~

F-7

Hope Creek ODCM Rev 18 Table F-1 (Continued)

Element Isotope Soluble Conc.

Insoluble Conc.

(uci/ml)

(uci/ml) 3 Te-132 3E-5 2E-5 Terbium (65)

Tb-160 4E-5 4E-5 Thallium (81)

TI-200 4E-4 2E-4 TI-201 3E-4 2E-4 TI-202 1E-4 7E-5 TI-204 1E-4 6E-5 I

Thorium (90)

Th-227 2E-5 2E-5 Th-228 7E-6 1E-5 Th-230 2E-6 3E-5 Th-231 2E-4 2E-4 Th-232 2E-6 4E-5 Th-natural 2E-6 2E-5 Th-234 2E-5 2E-5 Thulium (69)

Tm-170 SE-5 5E-5 Tm-171 SE-4 SE-4 J

Tin (50)

Sn-113 9E-5 8E-5 I

Sn-124 2E-5 2E-5 a

Tungsten (74)

W-181 4E-4 3E-4 W-185 1E-4 1E-4 W-187 7E-5 6E-5 Uranium (92)

U-230 5E-6 SE-6 U-232 3E-5 3E-5 U-233 3E-5 3E-5 U-234 3E-5 3E-5 U-235 3E-5 3E-5 U-236 3E-5 3E-5 U-238 4E-5 4E-5 U-240 3E-5 3E-5 U-natural 3E-5 3E-5 Vanadium (23)

V-48 '

3E-5 3E-5 Ytterbium (70)

Yb-175 1E-4 IE-4 F-8 i

Hope Creek ODCM Rev 18 Table F-1 (Continued)

Element Isotope Soluble Conc.

Insoluble Conc.

(uci/ml)

Yttrium Y-90 2E-5 2E-5 Y-91m 3E-3 3E-3 Y-91 3E-5 3E-5 Y-92 6E-5 6E-5 Y-93 3E-5 3E-5 Zine (30)

Zn-65 1E-4 2E-4 Zn-69m 7E-5 6E-5 Zn-69 2E-3 2E-3 Zirconium (40)

Zr-93 8E-4 8E-4 Zr-95 6E-5 6E-5 Zr-97 2E-5 2E-5 Any single radio-3E-6 3E-6 nuclide not listed above with decay mode other than alpha emission or spontaneous fission and with radio -

active half-life greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Any single radio-3E-8 3 E nuclide not listed above, which decays by alpha emission or spontaneous fission.

Notes:

1. If the identity of any radionuclide is not known, the limiting values for purposes of this table shall be: 3E-8 uci/ml.

2. If the identity and concentration of each radionuclide are known, the limiting values should be derived as follows: Determine, for each radionuclide in the mixture, the ratio between the quantity present in the mixture and the limit otherwise established in Appendix B for the specific radionuclide not in a mixture. The sum of such ratios for all the radionuclides in the mixture may not exceed "1" (i.e. " unity").

F-9

Hope Creek ODCM Rev 18 APPENDIX G CONTROLS FOR RELEASES

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FROM THE CIRCULATING WATER DEWATERING SUMP G-1

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Hope Creek ODCM Rev 18 APPENDIX G l

CONTROLS FOR RELEASES FROM THE j

CIRCULATING WATER DEWATERING SUMP The radiation monitoring instrumentation for the circulating water dewatering sump includes an offline gamma scintillation detection (Nal) with a sample flow indicator. The system does not include any discharge flow control device or measurement ofdischarge flow. Also, the radiation i

monitor is not included in the current Technical Specifications. As indicated in Appendix A, the setpoint for the radiation monitor is set by design and station procedures to be less than 2x be4 pound radiation levels. Releases through the sump at levels below the setpoint are

~

considered continuous releases. In order to control and quantify continuous releases from this point, the following sampling and analysis schedule is required:

Sampling Minimum

, Type of LLD Frequency Analysis Activity (uCi/ml)

Frequency Analysis Continuous M

Principal Gamma Emitters SE-7 Composite M

I-131 1E-6 M

H-3 1E-5 M

Gross Alpha 1E-7 Grab sangle M

Dissolved and 1E-5 M

Entrained Noble Gases (Gamma emitters)

Continuous Q

Sr-89, St-90 SE-8 Composite Fe-55 1E-6 i

Sampling and_ analysis for batch releases from this point shall be performed to meet the requirements of Technical Specification Table 4.11.1.1.1-1.

The following controls apply to the effluent radiation monitoring system (RY-4557):

MINUMUM OPERABILITY - with the radiation monitoring system inoperable, suspend continuous release of radioactive effluents via this pathway. Batch releases may be performed provided ACTION 110 of Technical Specification 3.3.7.10 is implemented.

G-2 k.

i Hope Creek ODCM Rev 18 SURVEILLANCE REQUIREMENTS-

~ A. CHANNEL CHECK - Daily, including verification of sample flow through the radiation monitor during sump pump operation B. SOURCE CHECK - Monthly -accomplished by automatic check source operation C. CHANNEL CALIBRATION - Every 18 months D. CHANNEL FUNCTIONAL TEST-Quarterly

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E. FLOW RATE MEASUREMENT DEVICES - Since there are no discharge process flow rate measurement devices, there are no surveillance requirements for discharge process flow rate.

Conservative assumptions should be made for release rates. The maximum release rate from the sump is 100 gpm. This value should be used to calculate radiation monitor setpoints and for calculation of concentation to determine compliance with Technical Specification 3.11.1.1.

i More realistic values may be used to calculate total activity released and dose consequences.

j Actual values should be used if process flow measurement devices are installed.

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