ML20094C277

Proposed Tech Specs Adding Section 3.0 Limiting Conditions for Operation Applicability,Deleting Section 3.9, Effluent Releases & Adding New Section 4.24, Radiological Effluent Releases

ML20094C277
Person / Time
Site:	Palisades
Issue date:	07/31/1984
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18051B015	List: ML20094C277
References
NUDOCS 8408070438
Download: ML20094C277 (175)

Text

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T ch Spic Chinga Requsst 2

'Policad2a Plcnt RETS, ODCM~& PCP July 31, 1984 I. CHANGES'(Continued)

M. Change the second sentence of S'ection 5.1 Site to read " Figure 5-1 shows the plan of site."

N. Add Figure'5-1 " Palisades Nuclear Plant Site Boundary" at the end of Section 5.

0. ' Change.Section 6.5.2.8.1 e to read:

"e. The results of actions taken to correct deficiencies identified by the audit program specified in Specifications 6.5.2.8.2 and 6.5.2.8.3 at least once every six months."

P. Add Section 6.5.2.8.3 (pages attached).

Q. . Delete paragraph 4 Solid Wastes, and paragraph 5 Radiological Impact on Man of existing Section 6.9.3.1.A.

R. Add new paragraphs 4 Solid Wastes and 5 Radiological Impact on Man, to Section 6.9.3.1. A (pages attached) .

S. Add new paragraph 6 PCP and ODCM changes to Section 6.9.3.1.A (pages attached)

T. Delete existing Section 6.9.3.1.B " Radiological Environmental Monitoring Program".

U. Add new Section 6.9.3.1.B entitled " Annual Radiological Environ-mental Operating Report" (pages attached).

V. Replace existing Table 6.9-1 with revised Table 6.9-1 (attached).

W. Add new Section 6.9.3.3 d (pages attached).

X. Add new Section 6.18 "Offsite Dose Calculation Manual ODCM" (pages attached).

'Y. Add new Section 6.19 " Process Control Program PCP" (pages attached).

Z. Add new Section 6.20 " Major Modifications to Radioactive Liquid Gaseous and Solid Waste Treatment Systems" (pages attached).

AA. Delete Special Technical Specifications Pursuant to Agreement in its entirety.

BB. Remove and replace Table of Contents in its entirety.

II. DISCUSSION The'above proposed Technical Specifications changes are requested in order to satisfy NUREG-0472 and staff guidance in the development of Radiological Effluent Technical Specifications.

TS0784-0011-NLO2 #

8408070438 840731 PDR ADOCK 05000255 P PDR

Tech Spec Change Request 3

[ Palisades Plant RETS, ODCM & PCP ES July 31, 1984 g

(( II. DISCUSSION (continued) gg A. These definitions are being added in response to NUREG-0472

- guidance in the development of RETS.

Ei E_ B. These sections are being added in response to NUREG-0472 guidance 9- and supercede our August 21, 1980 proposed Sections 3.0.1, 3.0.2, L_ 3.0.3 and 3.0.4.

r C. This page is being added for administrative purposes, k D. This page is being changed and renumbered for administrative

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[ purposes.

W E. and F. These pages are being renumbered for administrative purposes.

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G. and H. Information contained in existing Section 3.9 has been expanded, revised and reformatted in accordance with NUREG-0472 guidance and 17 is now located in Section 3.24 " Radiological Effluent Releases" of this proposal.

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EF :I. All process monitor discussion has been relocated to proposed l-L Tables 4.24-1 " Radioactive Liquid Effluent Monitoring Instrumenta-

- tion Surveillance Requirements" and 4.24-2 " Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements".

EE

(( J. and K. Information contained in existing Section 4.11 " Environmental Monitoring Program" has been reformatted and revised in at ordance j,

with NUREG-0472 and staff guidance and is now located in v.ew

proposed Section 4.11 " Radiological Environmental Monitoring".

h7 L. This section is being added in response to NUREG-0472 guidance in WT the development of RETS. The Section is numbered in keeping with fi the Standard Technical Specifications methodology whereby surveil-lance requirements are the same subsection number as the LCO.

= (e.g. LCO 3.24.3 has as its surveillance requirement Section Ei 4.24.3.) Blank Sections 4.20, 4.21, 4.22 and 4.23 are created (and

[_ left intentionally blank) for administrative continuity.

t M. and N. A new figure was created for this submittal that clearly shows the P= site boundary.

E 0 and P. Section 6.5.2.8.3 was added and referenced in 6.5.2.8.1.e to t delineate NSB responsibilities for ODCM, PCP and radiological

- environmental monitoring program.

I Q., R., S., T. and U.

[ Changes being made in response to NUREG-0472 guidance.

[- V. Example Table 6.9-1 " Environmental Radiological Monitoring Program f Summary" has been clarified.  :

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!E EP TS0784-0011-NLO2

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OT:ch1Spac Chrugs.Requsst-

~ 4 P0licadas Plant

-RETS, ODCM & PCP

-July-31, 1984 .

W. A new section concerning the submittal of Special Reports

" - referenced in these proposed specifications has been created.

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X., Y..and Z.

These-Sections were added in response to NUREG-0472 and staff

-guidance. ,..

~AA. These Sections are being deleted for the reasons specified in the cover letter.

Analysis of No Significant Hazards Consideration The changes proposed herein constitute additional-limitations, restrictions and controls not presently included in the Technical Specifications.

. -The operation of the facility in'accordance with the above change will not:

1) involve an increase to the probability or consequences of an accident previously evaluated; 2) create' the possibility a new or different kind of accident from those previously evaluated; and 3) involve a significant reduction in the margin of safety.

III. CONCLUSION The' Palisades Plant Review Committee in Special Meeting Number 84-048 on July-16, 1984 has reviewed this Technical Specification Change Regt.est and has determined that this change does not involve an unreviewed safety question and therefore involves no significant hazards consideration.

This change has~also been reviewed under the cognizance of the Nuclear Safety Board. A copy of this Technical Specification Change Request has

.been-sent to the Stata of Michigan official designated to receive such Amendments, to the Operating License.

. CONSUMERS POWER COMPANY ff -

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By \A/ VW )

- M W Reynolds

-Executive Vice President, Energy Supply

-Sworn and subscribed to before me this 31st day of July 1984.

4 W& vu lAA L-SherryLhrfey,Notarybahc JacksM County, MichigaU My commission expires November 5, 1986.'

SHERRY LYNN DURFEY

- Notary Fehlic, Jackson County, MietL My Commission Expires Nov. 5,1986

, TS0784-0011-NLO2

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Consumers Power Company Palisades-Plant - Docket 50-255

- - . . ~ TECHN(CAL SPECIFICATION CHANGE REQUEST

~ PROPOSED PAGES

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TS0784-0032-NLO2 s

1. 1

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS - APPENDIX A SECTION- DESCRIPTION PAGE NO

'1.0 DEFINITIONS 1-1

' 1.1 . Reactor Operating Conditions 1-1 1.2 Protective Systems 1-3 1.3 Instrumentation Surveillance 1-3

,1.'4 Miscellaneous Definitions 1-4

'2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1

~ 2.1: Safety' Limits - Reactor Core 2-1 2.2 Safety Limits - Primary Coolant System Pressure 2-3

.2.3 Limiting Safety System Settings - Reactor Protective System 2-4

. 3. 0. LIMITING CONDITIONS FOR OPERATION 3-1 t _ ' '3. 0 : Applicability 3-1

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3.1 _ Primary Coolant System 3-lb 3.1.1 Operable Components 3-lb 3.1.2 .Heatup and Cooldown Rate 3-4 3.1.3 Minimum conditions for Criticality 3-15 3.1.4 Maximum Primary Coolant Radioactivity 3-17 3.1.5 Primary Coolant System Leakage Limits 3-20 3.1,.6 Maximum Primary Coolant Oxygen and Halogens Concentrations 3-23 13.1.7 Primary and Secondary Safety Valves 3-25 3.1.8 Overpressure Protection Systems 3-25a 3.1.9 Leakage Detection Systems 3-25b 3.2 . Chemical and Volume Control System 3-26 3.3f Emergency Core Cooling System 3-29 3.4 Containment Cooling 3-34

-3.5- Steam and Feed-Water Systems 3-38

3. 6 '- Containment System' 3-40 3.7 Electrical Systems 3-41 3.8 Refueling Operations 3-46 3.9 (Deleted) ----

-3.10 control Rods & Power Distribution Limits 3-58 13.10.1 Shutdown Margin Requirements 3-58 3.10.2 Individual Rod Worth 3-58 3.10.3 Part-Length Control Rods 3-58

~3.10.4 Misaligned or Inoperable Control Rod or Part-Length Rod 3-60 3.10.5 Regulating Group Insertion Limits 3-60 3.10.6 Shutdown Rod Limits 3-61 3.10.7 Low Power Physics Testing 3-61

-3.10.8 Center Control Rod Misalignment 3-61 3.11 Power Distribution Instruments 3-65 3.11.1 Incore Detectors 3-65 3.11.2 Excore Power Distribution Monitoring System 3-66a 3.12 Moderator Temperature Coefficient of Reactivity 3-67 3.13 Containment Building and Fuel Storage Building Cranes 3-69 PROPOSED TS0584-0002-NLO2

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~ PALISADES PLANT TECHNICAL SPECIFICATIONS

' TABLE OF CONTENTS - APPENDIX A SECTION DESCRIPTION PAGE NO

- 3.0 LIMITING CONDITIONS FOR OPERATION (Continued) 3.14 Control Room Air Temperature 3-70

-3.15 Reactor Primary Shield Cooling System 3-70

,3.16 Engineered Safety Features System Initiation

. Instrumentation Settings 3-71 3.17 Instrumentation and Control System 3-77 3.18' Secondary Water Monitoring Requirements 3-82

..3.19 Iodine Removal System 3-84

'3.20' Shock Suppressors (Snubbers) 3-88

'3.21 Movement of Shielded Cask In Fuel Handling Areas 3-92 3.22 . Fire Protection System 3-96

-3.22.1- Fire Detection Instrumentation 3-96 3.22.2 Fire Suppression Water System 3-98 3.22.3 Fire Sprinkler System 3-100 3.22.4 Fire Hose Stations 3-101 3.22.5 . Penetration Fire Barriers 3-102 3-103 13.231 Power Distribution Limits

.3.23.1 Linear Heat Rate (LHR) 3-103

.3.23.2 Radial Peaking Factors 3-111 3.23.3 Quadrant Power T.H - Tq 3-112

[3.24 Radiological Effluent Releases 3-114 3.24.1. . Radioactive Liquid Effluent Monitoring

, . Instrumentation 3-114 Basis for 3.24.1 3-128 3.24.2 Radioactive Gaseous Effluent Monitoring Instrumentation 3-117 Basis for 3.24.2 3-128 3.24.3 Liquid Effluents Concentration 3-120 Basis for 3.24.3 3-129 j 3.24.4' . Liquid Effluent Dose 3-121 Basis for 3.24.4 3-129

'3.24.5 Gaseous Effluents Dose- 3-122

! Basis for 3.24.5 3-130 3.24.6 Gaseous Waste Treatment-System 3-125

' Basis-for 3.24.6 3-132 i3.24.7 Solid Radioactive Waste. 3-126 Basis for 3.24.7 3-132 3.24.8 To tal' Dose 3-127 Basis for'3.24.8 3-133 z4.0~ _ SURVEILLANCE REQUIREMENTS 4-1 1 4.1 Instrumentation and Control 4-1

'4.2 Equipment and Sampling Tests 4-13

= 4.3i Systems Surveillance 4-16

'4.4- Primary Coolant System Integrity Testing 4-24

-4.5 Containment Tests 4-25 4.6 Safety injection and containment Spray Systems Tests 4-39 L4.7- .

Emergency Power System Periodic Tests 4-42 4.7.1- Diese1' Generators 4-42 4.7.2 Station Batteries 4-42 PROPOSED TS0584-0002-NLO2

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. PALISADES PLANT TECHNICAL SPECIFICA110NS TABLE OF CONTENTS - APPENDIX A i

'SECTION- DESCRIPTION PAGE NO 4.0- ' SURVEILLANCE REQUIREMENTS (Continued) 4 - .

14.-7.3  : Emergency Lighting' 4-43 4.8 -Main. Steam Stop Valves 4-44 4.9 Auxiliary Feed-Water System 4-45

' 4'.10 : Reactivity' Anomalies 4-46 4.11i Radiological Environmental Monitoring 4-47 4.11.1J Basis for Monitoring Program- 4-59a

_ 14.11.3? Basis for Land Usa Ccasus 4-59a 4.11.5 Basis for Interlaboratory Comparison Program 4-59a 4.12 -

Augmented Inservica ' Inspection Program For High Energy Lines Outside of Containment 4-60

~4.13 _ Reactor Intervals Vibration Monitoring 4-65

4.14 ' Augmented Inservice Inspection Program for Steam Generators. 4-68

, ;4.15' Primary. System Flow Measurement 4-70 4.16' Inservice Inspection Program for Shock Suppressors (Snubbers) 4-71 4;17 ' Fire Protection System 4-75 4.18 Power Distribution' Instrumentation 4-81 4-81 4.18.1 Incore Detectors n4.18.2 Excore Monitoring' System 4-82 4.19' Power Distribution Limits 4-83 4.19.1 Linear Heat Rate 4-83

4.19.2 Radial Peaking Factors 4-84 4.20- -(Left Intentionally Blank) 4-85 4.21 -(Left Intentionally Blank) 4-86

, _~4.22 (Left Intentionally Blank) 4-87

'4.23 (Left-Intentionally Blank) 4-88

-(Left Intentionally Blank) 4-89

.4.24 Radiological Effluent Releases 4-90 4.24.1 Radiological Liquid Effluent Monitoring Instrumentation 4-90

'4.24.2 Radiological Gaseous Effluent Monitoring Instrumentation 4-90 4.24.3- Liquid Effluent Concentration 4-90

~4.24.4 Liquid Effluent Dose 4-90

. 4.24.5 Gaseous Effluent Dose 4-90 4.24.6 Gaseous Waste System 4-91 4.24.7 Solid Radioactive Waste 4-91 4.'24.8 Total Dose' 4-91 5.0 DESIGN FEATURES 5-1

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5.1 -Site 5-1 5.2 Containment Design Features 5-1 5.2.1 Containment Structures 5-1

. 5.2.2 Penetrations 5-2 5.2.3 containment Structure Cooling Systems 5-2 PROPOSED TS0584-0002-NLO2 o

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS - APPENDIX A SECTION DESCRIPTION PAGE NO 5.0 DESIGN FEATURES (Continued) 5.3 Nuclear Steam Supply System (NSSS) 5-2 5.3.1 Primary Coolant System 5-2 5.3.2 Reactor Core and Control 5-3 5.3.3 Emergency Core Cooling System 5-3 5.4 Fuel Storage 5-3 5.4.1 New Fuel Storage 5-3 5.4.2 Spent Fuel Storage 5-4 6.0 ADMINISTRATIVE CONTROLS 6-1 6.1 Responsibility 6-1 6.2- .

Organization. 6-1 6.2.1 Offsite 6-1 6.2.2 Plant Site 6-1 6.2.3 Nuclear Activities Plant Organization (NAPO) 6-la 6.3 Plant Staff Qualification 6-lb 6.4 Training 6-lb 6.5 Review and Audit 6-5 6.5.1 Plant Review Committee (PRC) 6-5

-6.5.1.1 Function 6-5 6.5.1.2 Composition 6-5 6.5.1.3 Alternate Members 6-5 6.5.1.4 Meeting Frequency 6-5 6.5.1.5 Quorum 6-5 6.5.1.6 Responsibilities 6-5 6.5.1.7 Authority 6-6 6.5.1.8 Records- 6-6 6.5.2 Nuclear Safety Board (NSB) 6-6 6-6 6.5.2.1 Responsibilities 6.5.2.2 Function 6-6 6.5.2.3 Composition 6-7 6.5.2.4 Alternate Members 6-7 6.5.2.5 Consultants- 6-7 6.5.2~.6 Meeting Frequencys 6-7 6.5.2.7 Quorum 6-8

'6.5.2.8 Responsibilities 6-8 6.5.2.8.1 Review 6-8 6.5.2.8.2 Audits 6-8 6.5.2.8.3 6.5.2.8.3 Audits 6-9 6.5.2.9 Authority 6-9 6.5.2.10 Records 6-10 6.6 (Deleted) 6-10 6.7 Safety Limit Violation 6-10 6.8- Procedures 6-10 6.9. Reporting Requirements 6-11 6.9.1 Routine Reports 6-11 6.9.2 Reportable Events 6-12 PROPOSED ,

-TS0584-0002-NLO2 h

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!: PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS - APPEhTIX A S ECTION DESCRIPTION PAGE NO 6.0 ADMINISTRATIVE CONTROLS (Continued) 6.9.3 Other Reporting Requirements 6-18

. 6.9.3.1 Routine Reports 6-18 6.9.3.2 Non Routine Reports 6-25b 6.9.3.3' Special Reports 6-26 6.10 Record Retention 6-26 6.11 Radiation Protection Program 6-28 6.12 High Radiation Area 6-28 6.13 Fire Protection Inspection 6-33 6.14 Environmental-Qualification 6-33 6.15 Systems Integrity 6-33 6.16 Iodine Monitoring 6-33 6.'17 Post-Accident Sampling 6-34 6.18 Offsite Dose Calculation Manual (0DCM) 6-35 6.19 Process Control Program (PCP) 6-35 6.20 Major Modifications to Radioactive Liquid, Gaseous and Solid Waste Treatment Systems 6-36 f-l PROPOSED TS0584-0002-NLO2 t-

1.4 MISCELLANEOUS DEFINITIONS (Contd)

GASEOUS RADWASTE TREATMENT SYSTEM A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off gases from the primary system and providing for delay or holdup for the pur- I pose of reducing the total radioactivity prior to release to the environment.

MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

OFFSITE DOSE CALCULATION MANUAL (ODCM)

The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring alarm / trip set points.

PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71 and Federal and State regulations and other requirements governing the disposal of the radioactive waste.

SITE BOUNDARY Tlie SITE BOUNDARY shall be that line beyond which the land is neither owned nor otherwise controlled by the licensee.

-UNRESTRICTED AREA An -UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or, any area within the SITE BOUNDARY used for residential quarters or for industrial, com-mercial, institutional and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is 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 par-ticulates from the gaseous exhaust stream prior to the release to the environ-ment. Such a system is not considered to have any effect on noble gas affluents. Engineered Safety Feature (ESF) atmospheric clean-up systems are not considered to be VENTILATIO!j EXHAUST TREATMENT SYSTEM components.

1-6 PROPOSED TS0784-0012-NLO2 s

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i LIMITING CONDITIONS FOR OPERATION Ii :3.0 APPLICABILITY y LIMITING CONDITIONS FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the

. succeeding Specifications is required during the plant conditions or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 . Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the Action requirements is not required.

3.0.3 When a Limiting Condition for Operation and/or associated ACTION requirements cannot be satisfied because of cirsumstances in excess of those addressed in the specification, within one hour action shall be L

initiated to place the unit in a condition in which the Specification does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, i

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and j -3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

h 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 Limiting Condition for Operation. Exceptiens to thesc requirements are stated in the individual

!- Specifications.

! 3.0.4 Entry into a plant condition or other specified condition shall not be made unless the conditions for the Limiting Condition for Operation are

. met without reliance on provisions contained in the ACTION requirements. .

This provision shall not prevent passage through or te plant conditions as required to comply with ACTION requirements. Exceptiens to these requirements are stated in the individual Specifications.

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TS0784-0013-NLO2 1

'3.1 PRIMARY COOLANT SYSTEM Applicability Applies to the operable status of the primary coolant system.

Objective To specify certain conditions of the primary coolant system which must be met to assure safe reactor operation.

Specifications 3.1.1 Operable components

a. At least one primary coolant pump or one shutdown cooling pump shall be in operation whenever a change is being made in the boron concentration of the primary coolant.

b. Four primary coolant pumps shall be in operation whenever the reactor is operated continually above 5% of rated power (exception to this specification is permitted as described in Table 2.3.1, item 1).

c. The minimum flow for various power levels shall be as shown in Table 2.3.1.

The measured four primary coolant pumps operating reactor vessel flow (as determined by reactor coolant pump differegtial pressures and pump performance curves) shall be 126.9 x 10 lb/h or greater, when corrected to 532*F.

In the event the measured flow is less than that required above, the limits specified on Figure 2-3 shall be reduced by l'F in inlet temperature for each 1% of reactor flow deficiency.

Continuous operation at power shall be limited to four-pump operation. Following loss of a pump, thermal power shall be reduced as specified in Table 2.3.1 and appropriate corrective action implemented. With one or more pumps out of service, return the pumps to service (return to four-pump operation) or be in hot standby (or below) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Start-up (above hot standby) with less than four pumps is not permitted.

d. Both steam generators shall be capable of performing their heat transfer function whenever the average temperature of the primary coolant is above 325'F.

e. Maximum primary system pressure differentials shall not exceed the following:

(1) Maximum steam generator operating transient differential of 1530 psi.

3-lb PROPOSED TS0784-0014-NLO2

1 3.1 PRIMARY COOLANT SYSTEM (Continued) 3.1.1 Operable Components (Continued)

(2) Hydrostatic tests shall be conducted in accordance with applicable paragraphs of Section XI ASME Boiler &

Pressure Vessel Code (1974). Such tests shall be conducted with sufficient pressure on the secondary side of the steam generators to restrict primary to secondary pressure differential to a maximum of 1380 psi. Maximum hydrostatic test pressure shall not exceed 1.1 Po plus 50 psi where Po is nominal operating pressure.

(3) Primary side leak tests shall be conducted at normal operating pressure. The temperature shall be consistent with applicable fracture toughness criteria for ferritic materials and shall be selected such that the differential pressure across the steam generator tubes is not greater than 1380 psi.

(4) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia. A minimum temperature of 100*F is required. Only ten cycles are permitted.

(5) Maximum secondary leak test pressure shall not exceed 1000 psia. A minimum temperature of 100'F is required.

(6) In performing the tests identified in 3.1.1.e(4) and 3.1.1.e(5), above, the secondary pressure shall not exceed the primary pressure by more than 350 psi.

f. Nominal primary system operation pressure shall not exceed 2100 psia.

g. The reactor inlet temperature (indicated) shall not exceed the value given by the following equation at steady state 100%

power operation:

T hiet 538.0 + 0.03938 (P-2060) + 0.00004843 (P-2060) + 1.0342 (W-120.2)

Where: T g 7,g = reactor inlet temperature in F' P = nominal operating pressure in psia W = total recirculating mass flow in 106 lb/h corrected to the operating temperature

' conditions.

Note: This equation ir, shown in Figure 3-0 for a variety of mass flow rates.

h. A reactor coolant pump shall not be started with one or more ofthePCScoldlegtemperaturesf250*Funless1)the pressurizer water volume is less than 700 cubic feet or 2) the secondary water temperature of each steam generator is less than 70*F above each of the PCS cold leg temperatures.

3-Ic PROPOSED

-TS0784-0015-NLO2

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i 3.1. PRIMARY COOLANT SYSTEM (Cont'd) 3.1.1 Operable Components (Cont'd)

i. The PCS shall not be heated or maintained above 325*F unless a minimum of 375 kW of pressurizer heater capacity is available from

-both buses ID and IE. Should heater capacity from either bus ID or IE fall below 375 kW,either restore the inoperable heaters to provide at least 375 kW of heater capacity from both buses ID and

s 1E within 72. hours or be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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3-Id PROPOSED nul081-0008a142

3.9 EFFLUENT RELEASE-DELETED IN ITS ENTIRETY

- Infomation now contained in Section 3.24

" Radiological Effluent Releases"

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12'Pages 3-50 3-50s 3-51 3-51a 3-52 3-52a 3-53 3-53a 3-54 3-55 3-56 3-57 c.

PROPOSED T50784-0016-NLO2

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3.24 RADIOLOGICAL EFFLUENT RELEASES 3.24.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.24.1.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.24-1 shall be OPERABLE with their alarm / trip set points set to ensure that the limits of Specification 3.24.3 are not exceeded. The alarm / trip set points 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:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip set point less conservative than required by the above speci-fication, without delay suspend the release of radioactive liquid efflu-ents monitored by the affected channel or declare the channel inoperabic, or change the set point so it is acceptably conservative,

b. With less than the minimum number of radioactive liquid effluent monitor-ing instrumentation channels OPERABLE, take the ACTION shown in Table 3.24-1. Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not cor-rected in a timely manner.

c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS Refer to 4.24.1.1 3-114 PROPOSED TS0784-0012-NLO2

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i TABLE 3.24-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line (RIA 1049) (1) 28

b. Steam Generator Blowdown Effluent Line (RIA 0707) (1) 29

2. GROSS BETA OR GAMMA RADIOACTIVITY MONITORS PROVIDING ALARtf BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE

a. Service Water System Effluent Line (RIA 0833) (1) 30

b. Turbine Building (Floor Drains) Sumps Effluent Line (RIA 5211) (1) 30

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line (FIC 1051 or 1050) (1) 31

4. CONTINUOUS COMPOSITE SAMPLERS

a. Turbine Building Sumps Effluent Line (1) 30 3-115 PROPOSED TS0784-00128-NLO2 L

.. {

TABLE 3.24-1 (Contd)

TABLE NOTATION ACTION 28 - With.the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a telease:

a. At least two independent samples are analyzed in accordance

.with Specification 4.24.3.1, 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 29 -.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 analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 microcuries/al at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

~

ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a lower limit of detection of at least 10-7 microcurie /ml.

ACTION 31 - 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 hours during actual releases. Pump performance curves or tank levels may be used to estimate flow.

3-116 PROPOSED 750784-0012B-NLO2

l 3.24.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION  ;

LIMITING CONDITION FOR OPERATION 3.24.2.1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.24.2 shall be OPERABLE with their alarm / trip set points set to ensure that the limits of Specification 3.24.5.1 are not exceeded. The alarm / trip set points of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: As shown in Table 3.24-2.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip set point less conservative than required by the above Specifi-cation, without delay, suspend the release of radioactive gaseous efflu-ents monitored by the affected channel or declare the channel inoperable or change the set point so it is acceptably conservative,

b. With less than the minimum number of radioactive gaseous effluent monitor-ing instrumentation channels OPERABLE, take the ACTION shown in Table 3.24-2. Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not cor-rected in a timely manner.

c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS Refer to 4.24.2.1 3-117 PROPOSED TS0784-0012B-NLO2

TABLE 3.24-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. WASTE GAS HOLDUP SYSTEM a.

Noble Gas Activity Monitor (RIA 1113)

Providing Alarm and Automatic Termi-nation of Release (1) At All Times 35

b. Effluent System Flow Rate Measuring Device (FI 1121) (1) At All Times 36 2.

CONDENSER EVACUATION SYSTEM (RIA 0631)

a. Noble Gas Activity Monitor (1) Above 210*F 37

3. STACK GAS EFFLUENT SYSTEM

a. Noble Gas Activity Monitor (RIA 2326 or RIA 2318) (1) At All Times 37

b. Iodine / Particulate / Sampler / Monitor (1) At All Times 37 (RIA 2325)

c. Sampler Flow Rate Monitor (1) At All Times 36

d. Hi Range Noble Gas (RIA 2327) (1) Above 210*F 38 4.

STEAM GENERATOR BLOWDOWN VENT SYSTEM

a. Noble Gas Activity Monitor (RIA 2320) (1) Above 210*F 37 5.

MAIN STEAM SAFETY AND DUMP VALVE DISCHARGE LINE

a. Gross Gamma Activity Monitor 1 per Above 325*F 38 (RIA 2323 and 2324) Main Steam Line

6. p ENGINEERED SAFEGUARDS ROOM VENT SYSTEM

a. Noble Gas Activity Monitor 1 per Above 210*F 38 (RIA 1810 and 1811) room l B

3-118 PROPOSED TS0784-0012B-NLO2

=

TABLE 3.24-2 (Cont.)

TABLE NOTATION - ACTION STATEMENTS ACTION 35 - With the number of channels OPERABLE less than required by Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment provided that prior to initiating the release:

a.

At and least two independent samples of the tank's contents are analyzed, b.

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

ACTION 36 - 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for continuous releases or four hours for batch releases.

ACTION 37 - 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 38 - With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter (s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

1) either restore the inoperable Channel (s) to OPERABLE status within 7 days of the event, or

2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.3.3.d within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status l

8 3-119 PROPOSED TS0784-0012B-NLO2

yr

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3.24.3 LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 3.24.3.1 The concentration of radioactive material released in liquid efflu-ents to UNRESTRICTED AREAS (see Figure 5-1) shall be limited to the concen-trations.specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 micro-curies /m1_ total activity.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of radioactive material released in liquid efflu-ents to UNRESTRICTED AREAS exceeding the above limits, without delay, restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS Refer to 4.24.3.1.a Refer to 4.24.3.1.b

+

3-120 PROPOSED TS0784-0012B-NLO2

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w, w ,

3.24.4 LIQUID EFFLUENT DOSE

'LIMITINGCONDITIONFORO?5PyTION" K (,

3.24.4.1 The dose or dose commitment to'a MEMBER OF THE PUBLIC from radio-active materials in liquid effluents released from each reactor unit to

' UNRESTRICTED AREAS (see Figure 5-1) shall be limited: s

,w

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 mrems to any organ, and q,

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

APPLICABILITY: At all times. T s ,i ACTION:

• y,

a. With(.hhe' calculated dose from the release of radioactive materials in liqu'ic effluents exceeding any of the above limits, prbpare and submit to the Co - ission within 30 days a Special Report per Specification 6.9.3.3.d 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. This Special Report shall also include the results of radiological analyses of the drinking water source.

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS' i

Refer to 4.24.4.1 ,

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TS0784-00128-NLU '

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3.24.5 GASEOUS EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.24.5.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Fig-ure 5-1) shall be limited to the following:

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

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

APPLICABILITY: At all times.

ACTION:

a. With the dose rate (s) averaged over a period of one hour exceeding the above limits, without delay, restore the release rate to within the above limit (s).

SURVEILLANCE REQUIREMENTS Refer to 4.24.5.1.a Refer to 4.24.5.1.b 3-122 PROPOSED t TS0784-0012B-NLO2 L.._

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LIMITING

,r- CONDITION FOR OPERATION 3.24.5.2 The. air dose due to noble gases released in gaseous effluents to areas at and beyond the SITE BOUNDARY (see Figure 5-1) shall be. limited to the following: ,

r., ,

s ,,

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

b. During any <calerdar year: L'js e than or equal to 10 mrads for gamma radiation and Jess than or equal to 20 mrads for beta radiation.

..we,+- -

APPLICABILITY: At all times.J s

ACTION

<i.

a. With the calculated air' dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report per Specification 6.9.3.3.d th,at 1,dentifies the cause(n) 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 Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE ' REQUIFSMENTS 4 ,

Refer t7 o 4.24.5.2 1 # a r

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. . , '?-123 PROPOSED TS0784-00128-NLO2 '

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LIMITING CONDITION FOR OPERATION 3.24.5.3 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) 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 any organ.

APPLICAEILITY: At all times.

ACTION:

a. With the calculated dose from the release of iodine-131, 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 a Special Report per Specifi-cation 6.9.3.3.d that identifies.ihu cause(s) for exceeding the limit and define (s) the corrective actions that have t .n 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 Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

. Refer to 4.24.5.3 3-124 PROPOSED TS0784-0012B-NLO2

3.24.6 GASEOUS WASTE TREATMENT SYSTEM

^

LIMITING CONDITION FOR OPERATION 3.24.6.1 THE WASTE GAS DECAY TANK SYSTEM shall be used to reduce radioactive gaseous effluents by holding gaseous waste collected by the system for a mini-mum of 15 days up to 60 days.

APPLICABILITY: When gaseous waste exceeds a Xe-133 concentration of IE-05 pCi/ce.

ACTION:

a. If a waste gas decay tank is required to be released with less than 60 days holdup time, the system waste gas tank contents shall be evaluated and the waste gas decay tank with the lowest Xe-133 concentration shall he released.

b. Gaseous waste may be discharged directly from the waste gas surge tank through a high-efficiency filter or from a waste gas decay tank with less than 15 days of holdup directly to the stack for a period not to exceed 7 days if the holdup system equipment is not available and the release rates meet Specification 3.24.5.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS Not applicable.

D 3-125 PROPOSED TS0784-0012B-NLO2

7~ .

3.24.7 SOLID RADIOACTIVE WASTE LIMITING CONDITION FOR-OPERATION 3.24.7.1 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements.

APPLICABILITY: At all times.

ACTION:

a. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radio-active wastes from the site.

u- ,b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS Refer to 4.24.7.1 s

~ PROPOSED TS0784-00I2B-NLO2

i I

l l

3.24.8 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.24.8.1 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to raddation 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 mrems.

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 Specifica-tions 3.24.4.1.a 3.24.4.1.b, 3.24.5.2.a, 3.24.5.2.b, 3.24.5.3.a or 3.24.5.3.b, calculations should be made including direct radiation con-tributions from the reactor units and from outside storage tanks to determine whether the above limits of Specification 3.24.8.1 have been exceeded. If such is the case prepare and submit to the Commission within 30 days a Special Report per Specification 6.9.3.3.d 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 Part 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 1

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.

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS Refer to 4.24.8.1.a Refer to 4.24.8.1.b tr 3-127 PROPOSED TS0784-0012B-NLO2

i l

BASES 3.24 RADIOLOGICAL EFFLUENT RELEASES 3.24.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The. radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip set points for these instruments shall be calculated and adjisted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of IO CFR Part 20.

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.

3.24.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive . gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseoas effluents. The alarm / trip set points for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

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.

3-128 PROPOSED TS0784-0012B-NLO2

BASES 3.24.3 LIQUID EFFLUENTS CONCENTRATION This specification is provided to ensure that the concentration of radioactive materials released in liquid waste 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 UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits 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 Maximum Permissable Concentration (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.

This specification applies to the release of liquid effluents from all reactors at the site.

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 HASL Procedures Manual, HASL-300 (revised annually), Currie, L A, " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal Chem 40, 586-93 (1968), and Hartwell, J K, " Detection Limits'for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

.3.24.4 LIQUID EFFLUENT DOSE 1 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 Limiting Condition for Operation implements the guides set forth in Section II. A of Appendix I. The ACTION statements provide the required operating flexibility and at the same

. time implement the guides set forth in Section IV.A of Appendix I to assure j that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh- l 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 concentrat' ions 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 l requirements in Section III.A of Appendix I that conformance with the guides I 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 appropriate

pathwcys 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 l methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses )

3-129 PROPOSED

_TS0784-0012B-NLO2

BASES 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.

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

3.24.5 GASEOUS EFFLUENTS DOSE 3.24.5.1 DOSE RATE This specification 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 concen-trations 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 ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a '

MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY te less than or equal to 500 mrems/yr to the total body or to less than or equal to 3000 mrems/yr to the skin. These release rate limits also restrict, at all times, the corres-ponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/yr.

This specification applies to the release of gaseous effluents from all reactors at the site.

The required detection capabilities for radioactive materials in gaseous 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 HASL

-Procedures Manual, HASL-300 (revised annually), Currie, L A, " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal Chem 40, 586-93 (1968), and Hartwell, J K, " Detection

" Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3-130 PROPOSED TS0784-0012B-NLO2

i BASES 3.24.5.2 DOSE - NOBLE GASES This specification is provided to implement the requirements of Sections II.B, III. A and IV. A of Appendix I,10 CFR Part 50. The limiting Condition for

-Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements 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 gaseous effluents to UNRESTRICTED AREAS will be kept "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 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 radio &ctive noble gases in 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 for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCM equations provided _ for determining the air doses at and beyond the SITE B0UNDARY are based upon the historical average atmospheric conditions.

3.24.5.3 DOSE - IODINE-131, IODINE-133, TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM This' specification is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements 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 materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calcula-tional methods specified in 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 appropriate pathways is unlikely to be substantially underestimated. The ODCM calcula-tional methodology and parameters for calculating the doses due to the actual release rates of the subject materials 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 3-131 PROPOSED

~TS0784-0012B-NLO2

BASES Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111 " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specificaticas 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 beyond the SITE BOUNDARY. The pathways that were examined in the ,

(1) individual inhalation of airborne development of these calculations were:

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 milk and meat by man, and (4) deposition on the ground with subsequent exposure of man.

3.24.6 GASEOUS RADWASTE TREATMENT SYSTEM 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" by meeting the design objectives given in Section II.D of Appen-dix I to 10 CFR 50.

It is expected that releases of r'adioactive materials in effluer.ts shall be kept at small fractions of the limits specified in 20.106 of 10 CFR 20. At the same time _the licensee is permitted the flexibility of operation, compatible with considerations of health and safety, to assure that the public is provided a dependable source of power even under unusual operating

-conditions which may temporarily result in releases higher than such small

. fractions, but still within the limits specified in Specification 3.24.5.

3.24.7 SOLID RADIOACTIVE WASTE This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited t;, waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constitu-ents and mixing and curing times.

3-132 PROPOSED l TS0784-0012B-NLO2 l

PROPOSED BASES 3.24.8 TOTAL DOSE This specification 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 preparation 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 organ, 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 Appen-dix I and if direct radiation doses from the reactor units and outside storage tanks 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 limits. 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 at the same site or within a radius ~of 8 km must be considered. If the dose to any MEMBER OF

• iHE.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 Specifications 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

'TS0784-0012B-NLO2 3-133

E 1

~

TABLE 4b1I3 Minimum Frequenciesifor . Checks, Calibrations and Testing of Miscellaneous Instrumentation and Controls Surveillance Channel Description Function Frequency Surveillance Method

1. Start-Up Range Neutron. a. Check S a. ' Comparison of both channel count rate Monitors indications when in service,

b. Test P b.. Internal test signals.

2. Primary Rod Position a. Check S a. Comparison of output data with Indication. System secondary RPIS

b. Check M b. Check of power dependent insertion limits monitoring system.

c. Calibrate (g) R c. Physically measured rod drive position used to verify system. accuracy.- Check _ rod position interlocks.

3. Secondary Rod Position a. Check S a. Comparison of output data with primary Indication System RPIS.

b. Check M b. Same as 2(b) above,

c. Calibrate (1) R c. Same as 2(c) above, including out-of-sequence alarm function.

4. Area Monitors a. Check D a. Normal readings observed and internal Note: Process Monitor test signals used to verify instrument Surveillance Requirements operation.

are located in Tables b. Calibrate 'R b. Exposure to known external radiation 4.24-1 and 4.24-2 source.

c. Test M c. Detector exposed to remote operated radiation check source.

5. Emergency Plan Radiation a. Calibrate A a. Exposure to known radiation source.

Instruments b. Test M b. Battery check.

6. Environmental Monitors a. Check M a. Operational check.

b. Calibrate A b. Verify airflow indicator.

7. Pressurizer Level a. Check S a. Comparison of six independent level Instruments readings.

b. Calibrate R b. Known differential pressure applied to sensor,

c. Test M c. Signal to meter relay adjusted with test device.

4-10 PROPOSED TS0784-0023-NLO2

4.11 RADIOLOGICAL ENVIRONMENTAL MONITORING 4.11.1 The radiological environmental monitoring program shall be conducted as specified in Table 4.11-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 4.11-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specifi-cation 6.9.3.1B 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 4.11-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the cor-rective actions to be taken to reduce radioactive effluents. When more than one of the radionuclides in Table 4.11-2 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 4.11-2 are detected and are the result of plant effluents, this report shall be submitted if the potential' annual dose to A MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.24.4.1, 3.24.5.2 and 3.24.5.3. 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 Radio-logical Environmental Operating Report.

c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 4.11-1, identify locations for obtaining replacement samples and add them to the radiological environ-mental monitoring program within 30 days. The specific locations from

which samples were unavailable may then be deleted from the monitoring program. Identify the cause of the unavailability of samples and identify the new location (s) for obtaining replacement samples in the next Annual Radiological Environmental Report.

4.11.2 The radiological environmental monitoring samples shall be collected )

pursuant to Table 4.11-1 and shall be analyzed pursuant to the requirements of Table 4.11-1 and the detection capabilities required by Table 4.11-3.

4-47 PROPOSED i TS0784-0012B-NLO2 1

4.11.3 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 9 overland meteoro-logical sectors of the nearest milk animal, the nearest residence and the nearest garden

• of greater than 53 m2 (500 ft2 ) producing broad leaf vegetation.

4.11.4 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, aerial survey or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.3.1B and shall be included in a revision of the ODCM for use in the following calendar year.

4.11.5 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission.

4.11.6 A summary of the results obtained as part of the above required Inter-laboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.3.1.B.

• 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. Speci-fications for broad leaf vegetation sampling in Table 4.11-1.4c shall be followed, including analysis of control samples.

4-48 PROPOSED TS0784-0012B-NLO2

s:

, c -

=4, TABLE 4.11-1.

~

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

. Exposure Pathway Number of Representative Sampling and . Type.and Frequency and/or Sample Samples and Sample Locations, Collection Frequency of Analysis

1. DIRECT RADIATION 21 routine monitoring stations Quarterly. Gamma dose quarterly.

either with two or more dosin-eters or with one instrument for m.4suring and recording dose rate continuously, placed as follows:

- An inner ring of stations, one in each. overland meteorological sec-tor (9) in the general area of the SITE BOUNDARY.

An outer ring of stations, one in each overland meteorological sec-tor (9) within the 12 km range from the site.

The balance of the stations (3) to be placed to serve as control stations.

4-49 PROPOSED TS0784-0012-NLO2

TABLE 4.11-1-(Contd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Number of Representative Sampling and Type and Frequency Exposure Pathway of Analysis and/or Sample Samples and Sample' Locations *__ Collection Frequency

2. AIRBORNE Samples from.5 locations: Continuous sampler Radioiodine Cannister:

Radioiodine and Particulates operation with sample I-131 analysis weekly for 3 samples from within 6 km of collection weekly or each filter change.

the SITE BOUNDARY in different more frequently if re-sectors (2.4 km-SSW, 5.6 km - ESE quired by dust loading. Particulate Samplet.

Gross beta radioactivity and 1.6 km - N).

analysig following filter I sample from the vicinity of change. Gamma isotopic a community having the highest analysis" if gross beta calculated annual average ground > 1.0 pCi/m 3.

level D/Q (Covert - 5.6 km - SE).

I sample from a control location in the least prevalent wind direc-tion' (Grand Rapids 89 km - NNE) .

4-50 PROPOSED TS0784-0012B-NLO2

]

TABLE 4.11-1 (Contd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Number of Representative Sampling.and Type and Frequency Exposure Pathway of Analysis and/or Sample Samples and Sample Locations

• Collection Frequency

3. WATERBORNE Gross beta (> 10 pCi/f

a. Lake (Surface) Plant lake water inlet. Composite samply over 1-month period. requires gamma) and tritium monthly.

Samples from plant, state park Monthly - grab sample. Gross beta (> 10 pCi/E

b. Well (Drinking) and Covert Township park wells. requires gamma) and tritium monthly, Gross beta (> 10 pCi/E

c. Lake (Drinking) I sample of South Haven drinking Composite sampig over water supply. 1-month period. requires gamma) and tritium monthly.

I sample from between north Semiannually. Gamma isotopic analysis *

d. Sediment From semiannually.

Shoreline boundary and Van Buren State Park beach.

4-51 PROPOSED TS0784-0012B-NLO2

TABLE 4.11-l'(Contd)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Representative Sampling and Type and Frequency and/or Sample Samples and Sample Locations

• Collection Frequency of Analysis

4. INGESTION

a. Hilk Samples from mi1 king animals Monthly. Gamma isotopic

• and 1-131 in 3 locations between 5-13 km analysis monthly.

distance.

i I sample from milking animals at a control location, 15-30 km distance,

b. Fish and Sample 2 species of commer- Sample in season or Gamma isotopic analysis

• Invertebrates cially and/or recreationally semiannually if they on edible portions, important species in vicinity of are not seasonal.

plant discharge area.

I sample of same species in areas not influenced by plant discharge.

I I

i j

j 4-52 PROPOSED TS0784-0012-NI,02

7 P

TABLE 4.11-1 (Contd).

. RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway . Number of Representative Sampling and Type and Frequency and/or Sample Samples and Satple Locations, Collection Frequency of Analysis

c. Food Products 1 sample each of two principal At time of harvest.E Gamma isotopic analyses

• f ruit crops (blueberries and on edible portion.

apples).

Samples of'3 different kinds of Monthly when Gamma isotopic" and 1-131 broad leaf vegetation grown near- available. analysis.

est each of two different offsite locations of highest predicted annual average ground level D/Q if milk sampling is not performed.

(SE or SSE sectors near site)

I sample of each of the similar Monthly when Gamma isotopic

• and I-131

, broad leaf vegetation grown available. analysis.

15-30 km distance in the least prevalent wind direction if milk sampling is not performed.

(SSW or S sectors) i 4-53 PROPOSED TS0784-0012-NLO2

l 4

TABLE 4.11-1 (Contd)

TABLE NOTATION Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous con-ditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If

, specimens are unobtainable due 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 Specification 6.9.3.1B. 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 alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program, b

One or more instruments, such as a pressurized ion chamber, for measuring 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 or phosphor readout zones in a packet are l considered as two or more dosimeters.

"The purpose of this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.

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

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

i 1

1 4-54 PROPOSED TS0784-0012-NLO2 1

l

. ' PROPOSED l

TABLE 4.11-1 (Contd)

TABLE h0TATION f

A composite sample is one in which the quantity (aliquot) of liquid samples is proportional to the quantity of liquid discharged and in which the method of sampling employed results in a specimen that is representative of the liquid raleased (continuous composites or daily grab composites which meet this criteria are acceptable).

I If harvest occurs more than once a year, sampling shall be performed during each discrete harvest.

4 TS0784-0012-NLO2 4-55 4

PROPOSE @' .

. TABLE 4.11-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Reporting Levels Water Airborne Particulate Fish Milk Food Products Analysis (pCi/1) or Gases (pCi/m 3) (pCi/kg, Wet) (pCi/1) (pCi/kg, Wet)

H-3 20,000*

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 D Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95 400 I-131 2 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300

• For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/2 may be used.

TS0784-0012-NLO2 4-56

- PROPOS $

TABLE 4.11-3 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS

• LOWER LIMIT OF DETECTION (LIS)

Water Airborne Particulate Fish , Milk Food Products Sediment Analysis (pCi/E) or Gases (pCi/m 3) (pCi/kg, Wet) (pCi/E) (pCi/kg, Wet) (pCi/kg,' Dry)

Gross Beta 4 0.01 H-3 2,000*

Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-Nb-95 15 I-131 1 0.07 1 60 Cs-134 15 0.05 130 15 80 150 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15

• If no drinking water pathway exists, a value of 3,000 pCi/E may be used.

TS0784-0012-NLO2 4-57

7_.

TABLE 4.11-3 (Contd) g I

TABLE NOTATION {

l.  !

• 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 8 also be analyzed and reported in the Annual Radiological Environmental Operat-ing Report pursuant to Specification 6.9.3.IB.

b Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4.13.

1. 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:

LLD =

E V 2.22 Y Exp (-AAt)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, s is the standard deviation of the background counting rate.or of the b

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 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and At for environmental samples is the elapsed time between sample collec-tion, or end of the sample collection period, and time of counting.

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

4-58 PROPOSED TS0784-0012A-NLO2

=- ;

TABLE 4.11-3 (Contd)

TABLE NOTATION 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. 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 uncon-trollable circumstances may render these LLDs unachievable. In such cases, .

the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specifica-tion 6.9.3.1B.

d LLD'for drinking water samples. If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.

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4.11 BASES-RADIOLOGICAL ENVIRONMENTAL MONITORING 4.11.1 MONITORING PROGRAM The radiological environmental monitoring program required by this specification 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 cperation. This monitoring program implementsSection 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 cad the modeling of the environmental exposure pathways. Guidance for this moni-toring program is provided by the Radiological Assessment Branch Technical Position en Environmental Monitoring. The initially specified monitoring program will be

,. effective for at least the first three 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 4.11-3 are considered optimum for routine environmental measurements in industrial labora-tories.

Detailed discussion of the. LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L A, " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry,"

Anal Chem 40, 586-93 (1968), and Hartwell, J K, " Detection Limits for Radioanalyti-cal Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

4.11.3 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the radiological cnvironmental monitoring program are made if required by results of this census.

The best information from the door-to-door survey, from aerial survey or from consulting with local 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 gardens of greater than 40 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (16 kg/yr) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To deter-nine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (ie, similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m 2 ,

4.11.5 INTERLABORATORY COMPARISON PROGRAM The requirement 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 performed cs part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section I,V.B.2 of Appendix I to 10 CFR Part 50.

4-59a PROPOSED TS0784-0012A-NLO2

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4-85 through 4-89 TS0784-0017-NLO2 PROPOSED

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~ 4. 24' RADIOLOGICAL EFFLUENT RELEASES 4.24.1 RADIOLOGICAL LIQUID EFFLUENT MONITORING INSTRUMENTATION 4.24.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the fre-quencies shown in Table 4.24-1.

l. 24.2 RADIOLOGICAL GASEOUS EFFLUENT MONITORING INSTRUMENTATION 4.24.2.1 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 operations at the frequencies shown in Table 4.24-2.

4_.24.3 LIQUID EFFLUENT CONCENTRATION 4.24.3.1.a Radioactive liquid wastes shall be sampled and analyzed according to the' sampling and analysis program of Table 4.24-3.

i 4.24.3.1.b 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 r.re maintained within the limits of Specification 3.24.3.1.

4.24.4 LIQUID EFFLUENT DOSE 4.24.4.1 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accord-ance with the methodology and parameters in the ODCM at least once per 31 days.

4.24.5 GASEOUS EFFLUENT DOSE 4.24.5.1.a The dose rate due to noble gases in gaseous effluents shall be deternined to be within the limits of 3.24.5.1.a in accordance with the methodology and parameters in the ODCM.

4.24.5.1.b The dose rate 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 limits of 3.24.5.1.b 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 4.24-5.

4.24.5.1 Cumulative dose contributions for the current calendar quarter and 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.

4.24.5.3 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 4-90 PROPOSED TS0784-0012A-NLO2

/

L

'adcordancewiththemethodologyandparametersintheODCMatleastonceper 4, 31 days.

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4.24.6 GASEOUS WASTE SYSTEM - NONE J 2 4 ~ 't SOLID RADIOACTIVE WASTE 4.24.7.1 The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICA-TION of at 1. east one representative test specimen from at least every tenth -

batch of each type of wet radioactive waste (eg, filter sludges, spent resins, evaporator' bottoms and boric acid solutions).

a. If any test specimen fails to verify SOLIDIFKCATION, tae SOLIDIFICATION of

the batch under test shall be suspended untilcsuch time as additional test specimens can be obtained, alternative SOLIDIFICATI0lPparameters can be determined in accordance with the PROCESS CONTROL PROGRAM and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be

,. resumed using the a E ernative SOLIDIFICATION parameters determined by the 4- PROCESS CONTROL PROGRAM. ,

/:

> <bf If the initial test specimen from a batch of, waste fails to verify SOLIDI-FICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and j ;,g ^* testing uf representative test specimens f/om each consecutive batch of the same type of wet waste until at least'3, consecutive initial test

h' q/-

specimens-demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modffini as required, as provided in Specification 6.19, to assure .

i SOLIDIFICATION of subseq'uent batches of waste.

• 4.24.8 -TOTAL DOSE f

4 '. 2 4. 8.1.~a Cumulative dose contributions from -Iiquid' and gaseous effluents shall be determirAd in accordance w{tKrSpecifications 4'.24.4.1, 4.24.5.2 and

~ A 24.5,.3 and in ac,cordance with the 3,ethodology - . .

end parameters in the ODCM.

j p j _4.24.8.1.b Cumulativedosecontrib[tionsfromElirectradiationfromthe WP reactor units-and;from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in Specification 3.24.8.1.a.

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TABLE 4.24-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION' SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FITNCTIONAL -

INSTRUMENT CHECK CHECF CALIBRATION TEST

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

4. Liquid Radwaste Effluent Line (RIA 1049) P P R(3) Q(1)(2)

b. Steam Generator Blowdown Effluent Line (RIA 0707) D M R(3) Q(1)(2)

2. GROSS BETA OR GAMMA RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMuTIC TERMINATION OF RELEASE

a. Service Water System Effluent Line (RIA 0833) D M R(3) Q(2)

b. Turbine Building (Floor Drains) Sumps Effluent Line_(RIA 5211) D M R(3) Q(2)

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line (FIC 1051 or 1050) D(4) NA R Q

4. TURBINE SUMP EFFLUENT COMPOSITER D(4) NA NA NA

.l l

TS0784-0012A-NLO2 4-92 i

p_

PROPOSED TABLE 4.24-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED

1. WASTE GAS HOL7UP SYSTEM

a. Noble Gas Activity ionitor-Providing Alarm and At.e',matic Termination of Release D(4) P R(3) Q(1)(2) *

b. WGDT Effluent Flow Rate Device D(4) NA R NA *

2. CONDENSER EVACUATION SYSTEM

a. Noble Gas Activity Monitor D M R(3) Q(2) Above 210*F

3. STACK GAS EFFLUENT SYSTEM

a. Noble Gas Activity Monitor D M R(3) Q(2) *

b. Iodine Particulate Sampler / Monitor W M** R(3)** NA *

c. Sampler Flow Rate Monitor D NA R NA *

d. Hi Range Noble Gas D M R(3) Q(2) Above 210 F

4. STEAM GENERATOR BLOWDOWN VENT SYSTEM

a. Noble Gas Activity Monitor D M R(3) Q(2) Abt,ve 210*F

• At all times other than when the line is valved out and locked.

• • Sampler not applicable TS0784-0012A-NLO2 4-93

1

. PROPOSED-TABLE 4.24-2(Cont.)

CHANNEL MODES IN WHICH CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED

5. MAIN STEAM SAFETY AND DUMP VALVE DISCHARGE LINE

a. Gross Gamma Activity Monitor D M R(3) Q(2) Above 325'F

6. ENGINEERED SAFEGUARDS ROOM VENT SYSTEM

a. Noble Gas Activity Monitor D M R(3) Q(1)(2) Above 210'F TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs if instrument indicates measured levels above the alarm / trip set point.

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that cotrol room alarm annunciation occurs if either of the following conditions exists:

a. Instrument indicates measured levels above the alarm set point.

b. Circuit failure.

a (3) a. The CHANNEL CALIBRATION shall be parformed using one or more of the reference standards traceable to the National Bureau of. Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range.

b. For tubsesquent CHANNEL CALIBRATION, sources that have been related to the fa) calibration may be u se/. .

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

TABLE FREQUENCY NOTATION D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Q At least once per 92 days M At least once per 31 days R At least once per 18 months P Prior to radioactive batch release W At least once per week TS0784-0012A-NLO2 4-94

TABLE 4.24-3 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimma of Detection Liquid Release Sampling Analysis Type of Activity (LLD)*

Type Frequency Frequency Analysis (pCi/ml)

A. Batch Waste P P Each Batch Each Batch Principal Gamma 5 x 10-7 Releage

-Tanks Emitters I-131 1 x 10-6 P M Dissolved and I x 10-6 One Batch /M Entrained Gases (Gamma Emitters)

P M d

H-3 1 x 10-6 Each Batch Composite Gross Alpha 1 x 10-7 P Q d Sr-89, Sr-90 5 x 10-8 Each Batch Composite B. Continuogs g W Principal Gamma 5 x 10-7 f

. Releases Continuous Composite Emitters I-131 1 x 10-6 Turbine Sump M M Dissolved and I x 10-6 Grab Sample Entrained Gases

' Steam Generator (Gamma Emitters)

Blowdown M- H-3 1 x 10-6 f f Continuous Composite Service Water Gross Alpha 1 x 10-7 f Q f Sr-89, Sr-90 5 x 10-8 L: Continuous Composite FREQUENCY NOTATION P Prior to batch release M Calendar month Q- Calendar quarter W Calendar week l

l TS0784-0012A-NLO2 h-95 PROPOSED I

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TABLE 4.24-3 (Contd)

TABLE NOTATION a TheLLLD is defined, in Table 4.11-3, note C.

A batch release is the discharge of liquid wastes of a discrete volume. Prior I to sampling for ana'.yses, each batch shall be isolated and then thoroughly I mixed to assure representative sampling.

c The principal gamma emitters for which the LLD specification applies

~

exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99,* Cs-134. Cs-137, Ce-141 and Ce-144.* This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report

. pursuant to Specification 6.9.3.1.a.

• (LLD - SE-06 because of low gamma yields.)

-d A composite sample is one in which the quantity of liquid sampled is propor-tional 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; eg,-from a volume of a system that has an input flow during the continuous release.

f To-be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

4-96 PROPOSED TS0784-0012A-NLO2

PROPOSED TABLE 4.24-5 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit of Sampling Analysis Type of Detection (LLD)*

Gaseous Release Type Frequency Frequency Acti>ity Analysis (pCi/ml)

.P P A. Waste Gas Storage Each Tank Each Tank Principal Gamma Emitters 1 x 10-4 Tank Grab Sample P P B. Containment PURGE Each PURGE Each PURGE Principal Gamma Emitters 1 x 10-4 Grab Sample C. Stack Gas Effluent Continuous

• W' I-131, I-133 1 x 10-12 Charcoal Sample Continuous

• W Principal Gamma Emitters 1 x 10-11 Particulate (I-131, Others)

Sample Continuous

• Q 1 x 10-11 Composite Sr-89, Sr-90 and Particulate Gross Alpha Sample Continuous

• Noble Gas Noble Cases IE-06 Monitor Gross Beta or Gamma TS0784-0012A-NLO2 h-97

r'

}

l TABLE 4.24-5 (Contd)

TABLE NOTATION

The LLD is defined, in Table 4.11-3, note C.

The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-:33m, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65,

' Mo-99*, Cs-134, Cs-137, Ce-141 and Ce-144* for particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, _ hall also be analyzed-and reported in the Semiannual Radioactive Effluent Rc' ease Report pursuant to Specification 6.9.3.1.a.

• (LLD 1E-10 because of low gamma yields)

"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 Specifications 3.24.5.1, 3.25.5.2 and 3.24.5.3.

d Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing or after removal from sampler.

'With channels operable on iodine monitor RIA 2325 less than required per Specification 3.24.2, sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, start-up or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 1 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 attivity has not increased more than a factor of 3.

1 l

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k-98 PROPOSED l TS0784-0012A-NLO2

e

' 5. 0 ' DESIGN FEATURES 5.1 SITE The Palisades reactor shall be located on 487 acres owned by Consumers Power Company on the eastern shore of Lake Michigan approximately four and one-half miles south of the southern city a limits of South Haven, Michigan. Figure 571 shows the plan of the l site. The minimum distance to the boundary of the exclusion area as t

-defined'in-10 CFR 100.3 shall be 677 meters.

5.2 CONTAINMENT DESIGN FEATURES 5.2.1  : Containment Structure

a. The containment structure completely encloses the primary coolant system to minimize release of radioactive material to the environment should_a failure of the primary coolant system occur. The prestressed, post-tensioned concrete structure provides adequate biological shielding for both normal operation and accident situations and is designed for low leakage at a design pressure of 55 psig and 283*F.

The principal design basis for the structure is that it be capable of withstanding the internal pressure resulting from a design basis loss-of-coolant accident. In this event, the total energy contained in the water of the primary coolant system is assumed to be released into the containment through a double-ended break of the largest primary coolant pipe coincident with a loss of normal and standby electrical power.

Subsequent pressure behavior is determined by the engineered safety features and the combined influence of energy sources and

,- heas sinks.

b. The external design pressure of the containment shell is 3 psig.

This value is approximately 0.5 psig greater than the maximum external pressure that could b* developed if the containment were sealed during a period of tow barometric pressure and high temperature and, subsequently, the containment atmosphere were cooled with a concurrent major rise in barometric pressure.

. Vacuum breakels are therefore not provided.

c. .The containment is designed as a seismic Class I structure.

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. ADMINISTRATIVE CONTROLS

-QUORUM '

A. quorum of NSB shall consist of_the Chairman and four (4) members. No (The Vice

16. 5l. 2. 7 more than

'Ch;irman may be a voting member when noc acting in the capacity of Chairman.) It c cinority of the quorum'shall have line responsibility for operation of the facility.

la the responsibility of the Chairman to ensure that the quorum convened for a meeting contcins appropriately qualified members or has at its disposal consultants sufficient to

.ctrry out the review functions required-by.the meeting agenda. m 6.5'.2.8 ' RESPONSIBILITIES i

REVIEW 16.5.2.8.1 -NSB shall be responsible for the review of:

a. Significant. operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

b. 'All reportable events and other violations (of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements or of internal-procedures or instructions) having nuclear safety significance.

c. -Issues of safety significance identified by the Plant General Manager, the NSB Chairman, Executive Engineer - NAPO or the PRC.

d. ~ Proposed changes in the operating license or Appendix "A" Technical Speci-fications.

e. The results of actions taken to correct deficiencies identified by the audit

' program specified in Specifications 6.5.2.8.2 and 6.5.2.8.3 at least once l every six months.

d tests if. . or Safety evaluations for changes to procedures, equipment, or systems an experiments completed under the provisions of 10 CFR 50.59, to verify that such actions did not constitute an unreviewed safety question.

g. -Maintain cognizance of PRC activities through NAPO attendance at scheduled PRC meetings or through review of PRC meeting minutes.

JAUDITS 16.5.2.8.2 Audits of operational nuclear safety-related facility activities shall be performed under the cognizance of NSB. These audits shall encompass:

a.' iThe-conformance of plant operation to provisions contained within the Technical Specifications and applicable license conditior.s at least once per

12. months.

b. The performance, training and qualifications of the entire facility staff at least once per 12 months.

6-8 PROPOSED nu0283-0452a443-63 ru==_ _

NDMINISTRATIVECONTROLS

c. The perform:nca cf getivitisc rsquired by the optrational qualif ty' assurance program (CPC-2A QAPD) to meet the criteria of Appendix "B",'10 CFR 50, at least once per 24 months.

d. The Site Emergency Plan and implementing procedures at least once per 12 months.

e. The Site Security Plan and implementing procedures (as required by the Site Security Plan) at least once per 12 months,

f. Any other area of plant operation considered approriate by NSB or the'Vice President - Nuclear Operations.

g. The plant Fire Protection Program and implementing procedures at-least once per 24 months.

h. An' independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified offsite licensee personnel or an outside fire

_ protection firm.

i. An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than 3 years.

Audit reports' encompassed by Specification 6.5.2.8.2 above shall be forwarded to the NSB Vice Chairman'and Secretary, and Management positions responsible for the areas audited within thirty (30) days af ter completion of the audit.

6.5.2.8.3 Audits of Nuclear Operations Department activities shall be performed under the cognizance of the NSE. These audits shall encompass.

.a. The radiological environmental monitoring program and the results thereof at least once per 12 mcnths.

b. The OFFSlTE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

c. The PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months.

Audit reports encompassed by Specification 6.5.2.8.3 above shall be forwarded to the NSB Vice Chairman and Secretary, and Management poistions responsible for the areas audited within thirty (30) days after completion of the audit.

AUTHORITY 6.5.2.9 The NSB Chairman shall report to and advise the Vice President -

Nuclear Operations of significant findings associated with NSB activities and of recommendations related to improving plant nuclear safety performance.

6-9 PROPOSED TS0784-0024-NLO2

d. Alphs Ridiorctivity Total curies of gross alpha-emiting material determined to be released in liquid effluents.

e. Volumes (1) Total measured volume (liters), prior to dilution, of liquid effluent released.

(2) Total determined volume, in liters, of dilurion water used during the period of the report.

4. Solid Wastes The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:

a. Container burial volume,

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

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g.,

dewatered spent resin, compacted dry waste, evaporator bottoms),

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

f. Solidification agent or absorbent (e.g., cement, asphalt).

=<

6-23 PROPOSED TS0784-0025-NLO2

=

5.. Radi:lcgien1 L;p;ct on Man The Radioactive Effluent Release Report to be submitted

_n within 60 days after January 1 of each year shall include potential doses to individuals and populations calculated using measured effluent and averaged meteorological data in accordance with the methodologies in the ODCM.

a. Total body and significant organ doses (greater than

.1 millirem) to individuals in unrestricted areas from receiving water-related exposure pathways.

b. The maximum of fsite air doses (greater than 1 millirad) due to beta and gamma radiation at locations near ground level from gaseous effluents.

c. Organ doses (greater than 1 millirem) to individuals in unrestricted areas from radioactive iodine and radioactive material in perticulate form from the major pathways of exposure,

d. Total body doses (greater than 1 manRem) to the population and average doses (greater than 1 millirem) to individuals in the population from receiving water-related pathways to a distance of 50 miles from the site.

e. Total body doses (greater than 1 manRem) to the population and average doses (greater than millirem) to individuals in the population from gaseous effluents to a distance of 50 miles from the site.

6. PCP and ODCM Changes The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 4.11.4.

6.9.3.1.B ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.3.1.B Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.

The annual radiological environmental operating reports shall include summaries, interpretation and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate) and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use census required by Specification 4.11.3.

6-24 PROPOSED TS0784-0026-NLO2

l 7

6.9.3.1.B (C:ntinund)

The annual radiological environmental operating reports shall

- s -

include summarized and tabulated results in the format of Table

~4.11-4 of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The b missing data shall be submitted as soon as possible in a supple-mentary report.

The reports shall also include the following: a summary descrip-tion of the radiological environmental monitoring program including sampling methods for each sample type, a map of all sampling locations keyed to a table giving distances and directions from one reactor and the results of land use censuses required by the Specification 4.11.3, and results of the inter-laboratory comparison program required by Specification 4.11.5.

( .-

9 6-25 PROPOSED TS0784-0027-NLO2 ,

DELETED IN ITS ENTIRETY PROPOSD TS0784-0028-NLO2

. - ":n+

16.9.3.2 Nonroutine Reports A lnq> ort. shall be _ submitted in the event that (a) the radiological 1 monitoring programs are not substantially conducted as described in Section 4.11; or (b) an unusual or important event occurs from plant operation that causes a significant environmental impact or affects a

'Je ' potential environmental impact. Reports shall be submitted within 30 days, s

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.s, 6-25b PROPOSED g.

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,A TABLE 6.9-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM

SUMMARY

Name of Facility Docket No

• Location of Facility Reporting Period (County, State)

Type and Lower Limit .

location with Highest Annual Mean

{ Medium or Pathway Total Number of ations Number of Detection, All Indicator p) ControlLocagions g Sampled of Analyses Mean(g Name Mean(fg Mean(g) REPORTABLE (Unit of Measurement) Performed (LLD) Range Distance and Direction Range Range OCCURRENCES Air Particulates (pC1/m3 ) Cross 8 416 0.003 0.08 (200/312) Middletown 0.10 (5/52) 0.08 (8/104) 1 (0.05-2.0) 5 miles 340* (0.08-2.0)- (0.05-1.40)

Y-Spec 32 Cs 0.003 0.05 (4/24) Smithville 0.08 (2/4) <LLD 4 (0.03-0.13) 2.5 miles 160* (0.03-0.13)

Ba 0.003 0.03 (2/24) Podunk O.05 (2/4) 0.02 (1/8) 1 (0.01-0.08) 4.0 miles 270* (0.01-0.08)

Sr 40 0.002 < LIE - -

<LLD 0 Sr 40 0.0003 <1.LD - -

<LLD 0 e

8 --

O M10684-0014A-RSO4 t

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TABLE 6.9-1 (Continued) , g Type and Lower Limit location with Highest Annual Mean

. Medium of Pathway Total Number of ations All Indicator p}

.?

ControlLocagiens Number of Sampled of Analyses Detection

• Name Mean (f (Unit of Measurement) Performed Mean({ Mean.({} REPORTABLE (LLD) Range Distance and Direction Range Range OCCURRENCES Fish

.pCi/kg (dry weight) y-Spec 8 l 137 Cs 80 .<LLD <LLD 90 (1/4) O Cs 80 <LLD -

<LLD <LLD 0 l m 60 .

[ Co ' 80 120 (3/4) River Mile 35 See Column 4 <LLD 0 K (90-200). Podunk River Nominal lower Limit of Detection (LLD) as defined in table notation a of Table 6.6-3 of Specification 6.6.

Mean and range based upon detectable measurements only. Fraction of detectable measurements at specified locations is indicated in parenthese:i. (f)

Note: The example data are provided for illustrative purposes only.

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MIO684-0014A-RSO4 -

?

(I 6.9.3.3 Specini R1 port

[F -

1 -a. opedias seput ta mitted covering the activitics identified below pursuant to the requirements of the applicable referenced specification:

p Specification Area Reference L Prestressing, Anchorage, 4.5.4 90 Days After Liner and Penetration 4.5.5 Completion of the Tests Test

• Primary _ System 4.3 Five Years Surveillance Evaluation and Review -

• A test is considered to be complete after all associated mechanical, chemical, etc., tests have been completed.

b. Bimonthly status reports on the program to improve the reliability of the paths to prevent post-LOCA boron precipitation shall be submitted to the Division of Operating Reactors _until completed.

c. Deleted.

6.9.3.3.d Special reports shall be submitted to the Director of the NRC F

Regional Office listed in Appendix D, 10 CFR Part 20, with a copy to the Director, Office of Inspection and Enforcement, US Nuclear Regulatory Commission, Washington, DC 20555 within the time period specified for each report.

6.10 RECORD RETENTION (Records not previously required to be retained shall be retained as required below commencing with the effective date of Technical Specification Change No 20. A system for efficient record retrieval shall be in effect not later than June 1976.)

6.10.1 The following records shall be retained for at least five years:

^

a. Records and logs of facility operation covering time interval at each power level,

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nucicar safety,

c. Reportable Occurrences.

d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

6-26 PROPOSED TS0784-0029-NLO2 '

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3 SPECIAL TECHNICAL SPECIFICATIONS PURSUANT TO AGREEMENT DELETED IN ITS ENTIRETY t

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' T80784-0030-NLO2.

Is s'N'

-.7 6.18 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.18.1 The ODCM shall be approved by the Commission prior to implementation.

6.18.2 Licensee initiated changes to the ODCM:

1. Shall be submitted to the Commission in the Semiannual Radio-active Effluent Release Report for the period in which the changes (s) was made effective. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered together with appropriate analyses or evaluations justifying the change (s);

b. A determination that the change will not reduce the accuracy

~

or reliability of dose calculations or setpoint O' determinations; and

c. Documentation of the fact that the change has been reviewed and found acceptable as defined by Specification 6.18.3.

6.18.3 Licensee initiated changes to and implementing procedures for the ODCM shall becomo effective upon review and approval by the respon-sible Nuclear Operations Department per CPC 2A (Quality Assurance Program).

h19 PROCESS CONTROL PROGRAM (PCP) 6.19.1 The PCP shall be approved by the Commission prior to implementation.

6.19.2 Licensee initiated changes to the PCP:

1. Shall be submitted to the Commission in the Semiannual Radio-active Effluent Release Report for the period in which the changes (s) was made. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

c. Documentation of the fact that the change has been reviewed and found acceptabic as defined by Specification 6.19.3.

6.19.3 Licensee initiated changes to and implementing procedures for the PCP shall become effective upon review and approval by the responsible Nucicar Operations Department per CPC 2A (Quality Assurance Program).

6-35 PROPOSED TS0784-0012-NLO2

y _ _ .

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i ADMINISTRATIVE CONTROLS

[

6.20 MAJOR MODIFICATIONS TO RADIOACTIVE LIQUID, CASE 0US AND SOLID WASTE TREATMENT SYSTEMS 6.20.1 Licensee initiated major modifications to the radioactive waste systems (liquid, gaseous and solid):

1. Shall be reported to the Commission in a special report pursuant i

to Specification 6.9.3.3d within six months of the time the safety evaluation was reviewed by the (PRC). The discussion of each modification shall contain:

a. A summary of the evaluation that led to the determination that the modification could be made in accordance with 10 CFR Part 50.59, n*- b. A description of the equipment, components and processes involved and the interfaces with other plant systems;

c. Documentation of the fact that the modification was reviewed and found acceptable by the (PRC).

2. Shall become effective upon review and acceptance by the Plant General Manager.

L TS0784-0012-NLO2

7.

l..

R,:

PROPOSED PALISADES NUCLEAR POWER PLANT OFFSITE DOSE CALCULATION MANUAL July 31, 1984 HIO583-0010A-RS03 z

i l

l TABLE OF CONTENTS ,

l I. GASEOUS EFFLUENTS A. ALARM / TRIP SETPOINT METHOD

1. Allowable Concentration

2. Monitor Response 2.1) Normal Releases 2.2) Accident Releases B. DOSE RATE CALCULATION

1. Appendix I Calculation Basis 1.1 Equations and Assumptions for Noble Gas 1.2 Equations and Assumptions for Iodines and Particulates 1.3 Design Basis Quantities (DBQ's) 1.4 Land Use Census and DBQ Changes C. DESIGN BASIS QUANTITY (DBQ) LIMITS

1. Batch Releases

2. Continuous Releases

3. Exceeding DBQ Limits

4. Releasing Radionuclides Not Listed in Table 1.9 D. OPTIONAL QUARTERLY DOSE CALCULATIONS

1. Methodology

" 1.1 Simplified, Conservative 1.2 Realistic Method E. GASEOUS RADWASTE TREATMENT SYSTEM OPERATION

1. System Description

2. Determination of Satisfactory Operation F. RELEASE RATE FOR OFFSITE MFC (500 mrem /yr)

G.- PARTICULATE AND HALOGEN SAMPLING H.- NOBLE GAS SAMPLING I. TRITIUM SAMPLING II. LIQUID EFFLUENTS A. CONCENTRATION

1. Radiological Effluent Tech Spec Requirement

2. Prerelease Analysis

3. Maximum Permissible Concentration - Sum of the Ratios MIO583-0010A-RS03 L

c- -

[ 11 B. . INSTRUMENT SETPOINTS I, - Setpoint Determination

2. Post release analysis C. DOSE

1. RETS Requirement

2. Release Analysis 2.1 Water Ingestion 2.2 Fish Ingestion

3. Annual Analysis D. OPERABILITY OF LIQUID RADWASTE EQUIPMENT E. GASEOUS RADWASTE TREATMENT SYSTEM OPERATION

'F. RELEASE RATE FOR OFFSITE MPC (500 mrem /yr)

~III. URANIUM FUEL CYCLE DOSE A. SPECIFICATION B. ASSUMPTIONS C. DOSE CALCULATION IV. RADWASTE SYSTEM MODIFICATIONS MIO583-0010A-RS03

I iip . (

(.

.4 I. GASEOUS EFFLUENTS A. ALARM / TRIP NTPOINT(tETHOD

\

(

Specification 3.24.5.1 requires that MPC'is not exceeded when averaged over a period not to exceed I hour. Based on the defin-ition of MPC, the dose rate in unrestricted areas due to gaseous effluents from the' site shall be limited at all times to the fol-( .' ,

lowing values:{

1. .500 arem/ytto the total body and 3,000 mrem /y to the skin from noble gases.

2. 1,500 mrem /ytoanyordnfromradioiodinesandparticulates,

.due to inhalation.

Specification 3 24.'2.I' requires gaseous effluent monitors to have 4

alarm / trip setpoints to ensure that offsite concentrations, when averaged over 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, will not be greater than MPC. This section of the ODCM describes the methodology that will be used to determine these setpoints.

The methodology for determining alarm / trip setpoints is divided into two major parts. The first consists of calculating an allowable concentration for the nuclide mixture to be released. The second consists-of: determining monitor response to this mixture in order to establish the physical settings on the monitors.

1.0' Allowable Concentration

-The total MPC-fraction (R ) f r each release point will be calcu-k lated by.the relationship defined by Note 1 of Appendix B, 10 CFR 20:

R g = (--) (F) g = $1.0 1 (1.1) p M10583-0010A-RS03

2

>, - )

• h ;NyhW~' '

• Ci = Concentration, at ambient temperature and pressure  !

of nuclide 1 (pCi/cc)

MPC g = The MPC of nuclide i from 10 CFR 20, Appendix B R(k) = The total MPC-fraction for release point k X/Q = Most conservative sector site boundary dispersion i (4.8E-05 sec/m3 )

'F = Release flow rate (83,000 cfm = 39.2 m /sec) 3 for stack  !

l ,

monitor- considerations; variable for other monitors.

NOTE: If a batch release is made while a continuous release or another batch release is in progress, the sum of all

. values of R aus e ess an .0.

k 12.0 Monitor Response '

Normal radioactivity releases consist mainly of well-decayed fission gasses. Therefore, monitor response calibrations are performed to fission gas typical of normal releases (mainly Xe-133). Response of monitors used to define fission product release rates under accident conditions may vary from that of Xe-133, however. Monitor response

- for the two categories of monitor is determined as follows:

1)~ Normal Releases (aged fission gasses)

Total gas concentration (pCi/cc) at the monitor is calculated.

The calibration curve or constant for cpm /(pci/ce) is applied to determine cpm expected. The setting for monitor alarms is established at some factor (b) greater than 1 but less than 1/Rk (Equation 1.1) times the allowed concentration (c):

s=bxc (2.1)

MIO583-0010A-RS03

l l

3 1

2) Accident Releases '

Monitors sre preset to alarm at or before precalculated offsite I dose rates would be achieved under hypothetical accident conditions. These setpoints are established in accordance with Emergency Plan requirements for defining Emergency Action Levels and associated actions. Emergency Implementing Proce-dures contain monitor-specific curves or calibration constants for conversion between epm and pCi/cc (or R/hr and pCi/ce),

depending on monitor type, for fission product mixtures as a function of mixture decay time.

When these monitors are utilized for other than accident condi-

^ '

- r tions, either an appropriately decayed " accident" conversion

. curve may be used, or a decayed fission gas calibration factor may be applied. In these cases, setpoints are established as in 1) above.

Setpoints of accident monitors (if set to monitor normal releases) are reset to the accident alarm settings at the end of normal' release. .Setpoints of other release monitors are maintained at the level used at the latest release (well below ir o'g' s

the level which would allow MPC to be exceeded at the site e s

boundary), or are reset to. approximately three times background in order to detect leakage or inadvertent releases of low level gases.

-, e B. DOSE RATE CALCULATION

, u-1 <1.0' The first step involves calculating a dose rate based on the design

( ' objective' source term mix used in Appendix I licensing calculations.

Historical meteorological data used in licensing are used in this calculation. Doses are determined for (1) noble gases and (2) iodines and particulates. . Dose rates as defined in this section are MIO583-0010A-RS03 V . _- . ... . - - _. - . - . -

7- .-

4 in terms of 10 CFR 50 Appendix I limits of mrem per quarter and millirem per year. All dose pathways of major importance in the Palisades environs are considered.

1.1 Equations and assumptions for calculating doses from noble gases are as follows:

1.1.1 Assumptions

1. Doses to be calculated are the maximum offsite point in air, total body and skin.

2. Exposure pathway is submersion within a cloud of noble gases.

3. Noble gas radionuclide mix is based on the historically observed source term given in Table 1.1, plus additional I nuclides.

4. Basic radionuclide data are given in Table 1.2.

5. All releases are treated as ground-level.

6. Meteorological data expressed as joint-frequency distribution of wind speed, wind direction, and atmospheric stability for the period rt.ulting in X/Q's and D/Q's shown in Table 1.3.

7. Raw meteorological data consist of wind speed and direction measurements at 10m and temperature measurements at 10m and i- 60m.

8. Dose is to be evaluated at the offsite exposure points where maximum concentrations are expected to exist (overland sector site boundaries), and nearest residents.

9. Potential maximum population (resident) exposure points are identified in Table 1.4. ,

MIO583-0010A-RS03

5

10. A semi-infinite cloud model is used.

1 1.' For person exposures, credit is taken for shielding by residence (factor of 0.7).

12. Radioactive decay is considered for the plume.

13. Building wake effects on effluent dispersion are considered.

14. A sector-average dispersion equation is used.

15. The wind speed classes that are used are as follows:

Wind Speed Class Number Range (m/s) Midpoint (m/s) 1 0.0-0.4 0.2 2 0.4-1.5 0.95 3 1.5-3.0 2.25 4 3.0-5.0 4.0 5 5.0-7.5 6.25

, 6 7.5-10.0 8.75 7 >10.0 --

-16. The stability classes that will be used are the standard A through G classifiestions. The stability classes 1-7 will correspond to A=1, B=2, . . ., G=7

17. Terrain effects are not considered.

'1.1.2 Equations To calculate the dose for any one of the exposure points, the following equations are used.

For determining the air concentration of any radionuclide:

9 7 i 2 1/2 f jk 01 P expf-A g x I (1.1)

R Izk "j ( "*!") "j

- j=1 K=1 ( l

. M10583-0010A-RS03 i

1

L:

6 4

where:

xg = Air. concentration of radionuclide i, pCi/m8 .

fg = Joint relative frequency of occurrence of winds in wind speed class j , stability class k, blowing toward this exposure point, expressed as a fraction.

Q = Average release rate of radionuclide i, pCi/s.

p = Fraction of radionuclide remaining in plume.

I = Vertical dispersion coefficient for stability class k (m).

u = Midpoint valt.e of wind speed class interval j , m/s.

x = Downwind distance, m.

n = Number of sectors, 16.

Ag = Radioactive decay coefficient of radionuclide i, s.1, 2nx/n = Sector width at point of interest, m.

For determining the total body dose rate:

D (1.2)

TB i i i

where:

D TB

• Y ** #* *' "#'"!Y' a Xg = Air concentration of radionuclide 1, pCi/m .3 DFB f = Total body dose factor due to gamma radiation, mrem /y per pCi/m3 (Table 1.5).

Fdr determining the skin = dose rate:

D, = Xg (DFS + 1.11 DFYg ) (1.3) i MIO583-0010A-RS03

7 where:

D, = Skin dose rate, mrem /y.

Xg = Air concentration of radionuclide i, pCi/m 8.

DFS = Skin dose factor due to beta radiation, mrem /y per pCi/m 8(Table . 5) .

1.11 = The average ratio of tissue to air energy absorption coefficients, mrem / mrad.

7 DFY g = Gamma-to-air dose factor for radionuclide i, mrad /y per pCi/m 3 (Table 1.5).

For determining dose rate to a point in air:

D, = 3.17E-02 Xg (DFYg or DFBg ) (1.4) i s where:

D, = Air dose rate mrem /yr

-DFB - Air dose. factor for beta radiation (Table 1.5) 3.17E-02 = Conversion'from release /yr-(Ci) to pCi/m3 divided by seconds /yr.

1.2 Equations.and assumptions for. calculating doses from radioiodines

'and particulates are.as follows:

1.2.1  : Assumptions".

1. -Dose is to be calculated for the critical organ, thyroid, and

, the critical age groups, infant (milk) and child (green, leafy vegetables).

2. Exposure pathways from iodines and particulates are milk

' ingestion, ground contamination, green leafy vegetables from home gardens, and inhalation.

3. The radioiodine and particulate nix is based on the histori-cally observed source term given in Tab ~' 1.1.

, MIO583-0010A-RS03 e

hid

8

4. Basic radionuclide data are given in Table 1.2.

'5. All releases are treated as ground-level.

6. Mean annual average X/Q's for the period January 1,1978 -

December 31, 1982 are given in Table 1.3.

7. Raw meteorological data for ground-level releases consist of wind speed and direction measurements at 10m and temperature measurements at 10m and 60m.

'8. Dose is to be evaluated at the potential offsite exposure points where maximum doses to man are expected to exist.

9. Real cow, goat and garden locations are considered.

10. Potential maximum exposura points (Table 1.4) considered are the nearest cow, goat and home garden locations in each sector.

11. Terrain effects and open terrain recirculation factors are not considered.

12. Building wake effects on effluent dispersian are considered.

13. Plume depletion r~l radioactive decay are considered for air-concentration calculations.

14. Radioactive decay is considered for ground-concentration calculations.

15. Deposition is calculated based on the curves given in Figure 1.1.

16. Milk cows and goats obtain 100% of their food from pasture grass May through October of each year.

17. Credit is taken for shielding by residence (factor of 0.7).

1.2.2 Equations To calculate the. dose for any one of the potential maximum-exposure points, the following equations are used.

1.2.2.1 Inhalation Equation for calculating air concentration, X, is the same as in the Noble Gas Section 1.1.2 (Equation 1.1).

MIO583-0010A-RS03 1

• - +6 v'w-e

9 For determining the organ dose rate:

D7 = 1.t 108 X DFI BR (1.4) i where:

Dg = Organ dose rate due to inhalation,strem/y.

Xg = Air concentration of radionuclide i, pCi/m 3.

DFI f

= Inhalation dose factor, mrem /pci (Table 1.7).

3 BR = B.eathing rate 1400 m /y, infant; 3700 m 3 y, child; or 8000 m 8/y adult.

1 x 106 = pCi/pCi conversion factor.

1.2.2.2 Ground Contamination For determining the ground concentration of any nuclide:

7 Gg = 3.15x107 fg Q1 DR

[1-exp -(A tb))

7 (1.5) k=1 where: ,

G g = Ground concentration of radionuclide i, pCi/m 3.

2% k = Stability class.

E f = Joint relative frequency of occurrence of winds in stability class k blowing toward this exposure point, expressed as a fraction.

L' A Q.1 = Average release rate of radionuclide i, pCi/s.

DR = Relative deposition rate, m~1 (Eigure 1.2 for DR/2nX).

r x = Downwind distance, m.

n = Number of sectors, 16.

2nx/n = Sector width at point of~ interest, m.

Ag = Radicactive decay coefficient of radionuclide i, y"1 t = Time'for buildup of radionuclides on the ground, 35 y.

e- .3.15x107 = s/y conversion factor y

-MIO583-0010A-RS03

b ,

'~

. y 10

'For. determining the total body or organ dose rate from ground contamination:

D G

=(, x 09 Gg DFG g (1 4

. i where:

.D = Dose rate due to ground contamination, mrem /y.

G G = Ground concentration of radionuclide i, pCi/m 2, f

DFG .= Dose factor for standing on contaminated ground, mrem /h per pCi/m2 (Table 1.8).

8,760 .= Occupation time, F/y.

1x108' = pCi/pCi conversion factor.

1.2.2.3' Milk Ingestion For determining the concentration of any nuclide (except C-14 and H-3) in and~on vegetation:

7 f

'CV f

~

= 3,600 k 0 1 DR r[1-exp ~ki e ]+ (1.7)

(2n/n) Y g y

Bg [1-exp (-Afbt )l PA i j where:

Cg = Concentration of radionuclide i in and on vegetation, pCi/kg.

k. = Stability class.

f = Frequency of this stability class and wind direction k

combination, expressed as a fraction.

Q = Average release rate of radionuclide i, pCi/s.

DR = Relative deposition rate, m 2 (Figure 1.1).

L l

MIO583-0010A-RS03

l?

(;

11 x = Downwind distance, m.

n = Number of sectors, 16.

'~

2nx/n = Sector width at point of interest, m.

L r = Fraction of deposited activity retained on vegetation (1.0 for iodines, 0.2 for particulates).

A Ei *Ai + A w

, where Ei

= Effective removal rate constant, A A is the radioactive decay coefficient, h"1, and A is a measure of physical loss by weathering (Aw = .0021 h~1).

t = Period over which deposition occurs, 720 h.

Y = Agricultural yield, 0.7 kg/m 2, B = Transfer factor from soil to vegetation of radioruclide h

i (Table 1.9).

A = Radioactive decay coefficient of radionuclide i, h~1 t = Time for buildup of radionuclides on the ground, b

3.07x105 h (35y).

P = Effective surface density of soil, 240 kg/m 2, 3,600 = s/h conversion factor.

For determining the concentration of C-14 in vegetation:

3 CVg = 1x10 Xg (0.11/0.16) (1.8) where:

CV g. = Concentration of C-14 in vegetation, pCi/kg.

Xg = Air concentration of C-14, pCi/m 3.

0.11- = Fraction of total plant mass that is natural carbon.

0.16

~

= Concentration of natural carbon in the atmosphere, g/m 3.

lx103~ = g/kg conversion factor.

For determining the concentration of H-3 in vegetation:

CVT = 1x108 XT (0.75)(0.5/H) (1.9)

MIO583-0010A-RS03 L

O.

V- - -

q 12 where:

CV T = Concentration of H-3 in vegetation, pCi/kg.

X = Air concentration of H-3, Ci/m 3.

T 0.75 = Fraction of total plant mass that is water.

0.5 = Ratio of tritium concentration in plant water to tritium concentration in atmospheric water.

H- = Absolute humidity of the atmosphere, g/m8 .

1x108 = g/kg conversion factor.

For determining the concentration of any .uclide in cow's or goat's milk:

CMg = CV FMg Q gexp (-A t) g (1.10) where:

CM g = Concentration of radionuclide i (including s:-14 and H-3) in milk, pCi/1.

CV = Concentration of radionuclide i in and on vegetation, pCi/kg.

FM = Transfer factor from feed to milk for radionuclide i, d/1 (Table 1.6).

Qg = Amount of feed consumed by the milk animal per day, kg/d.

Ag = Radioactive decay coefficient of radionuclide i, d'1 tg = Transport time of activity from feed to milk to receptor, 2 days.

For determining the organ dose rate from -ingestion of green leafy vegetables and milk:

D = 1x10s CM DF UM (1.11) 1

-MIO583-0010A-RS03

13 where:

D = Organ dose rate due to ingestion, mrem /y.

CN g = Concentration of radionuclide i in vegetables or milk, pCi/Kg (or liters).

DFG = Ingestion dose factor, mrem /pCi (Table 1.7).

f UM = Ingestion rate for milk, 3301/y; for vegetables 26 Kg/yr (child), no ingestion by infant.

-1x108 = pCi/pCi conversion factor.

c 1.2.2.4 Organ Dose Rates For determining the total thyroid dose rate from iodines and particulates:

O

.D=D7+Dg+Dg+Dy (1.12) ,

where: -

D = Total organ dose rate, mrem /y.

D = Dose rate due to inhalation, mrem /y.

7 Dg = Dose rate due to ground contamination, mrem /y.

'D = Dose rate due to milk ingestion, mrem /y.

3 Dy. = Dose rate due to vegetable ingestion, mrem /y.

1.2.3 The maxim'm u organ dose rate, maximum total body dose rate, and maximum skin dose rate calculated in the previous section (Sec 1.2.2).are used to calculate design basis quantities as described in Section 1.3.

1.3 Design Basis Quantities The design basis quantity of a radionuclide emitted to the atmos-phere is the: amount of that nuclide, when released in one year, which would result in a dose not exceeding any of the following:

MIO583-0010A-RS03

14

a. 15 millirem to any organ of an individual from iodines and ps particulates.

Q%;.

\ \ .

.j 't . b. 15 millirem to skin of an individual from noble gas.

c. 5 millirem to the total body of an individual from noble gas.

Design basis. quantity (Ci) is the smallest value for each nuclide, calculated by dividing the dose limits (a through e, above) by the appropriate dose. calculated in step 1; the result then is multiplied by the amount of radionuclide (Ci) used to conservatively estimate the doses of Section D, as listed in Table 1.1 (or assumed a hypothetical 1 Ci/ year for nuclides not actually present):

DBQ = AI (C ) (1.13)

D c

where:

D =' Appendix I dose limit (mrem or mrad).

f7

-D = -Calculated dose from step 1 (mrem or mrad).

C. = Quantity of nuclide resulting in dose D (Ci).

DBQ = Design Basis Quantity (Ci).

The limiting values for-Design Basis Quantities for radionaclides released to the atmosphere are given in Table 1.9.

MIO583-0010A-RS03 l

15 1

.The inverse of the ratio C /D in the above equation (i.e., D /C )

is a useful value, since it represents the most limiting dose per unit quantity of each nuclide released. Use of the D /C ratio in quarterly evaluation of offsite dose is discussed in section D.

Values of D /C are given in Table 1.9.

1. 4 - Land Use Census and DBQ Changes Specifications 4.11.3 and 4.11.4 describe the requirements for an annual land use census and revision of the ODCM for use in the following calendar year. Areas of the ODCM which will be reviewed, and-changed if appropriate, are Table 1.4 (Land Use Census data by Sector), and Table 1.9 (Gaseous Design Basis Objective Annual Quantities). Changes will be effective on January 1 of the year following the year of the survey C. DESIGN OBJECTIVE QUANTITY (DBQ) LIMITS ON BATCH AND CONTINUOUS RELEASES 1.0 . Batch Releases

~

I Prior to each batch release (waste decay tank release or containment Purge), the quantity of each nuclide identified is summed with the quantity of that nuclide released since the first of the current calendar year. The cumulative total for each nuclide then is divided by the design objective quantity for each nuclide (from Table 1.9), and the resultant fractions are summed in order to assure that the sum fraction of all nuclides does not exceed 1.0:

Ai

< 1.0 (I*

(DBQ)

The amount in any calendar quarter should not exceed 0.5. This is checked by subtracting the value obtained at the end of the previous quarter from.the value obtained from the cumulative total to date, including the batch to be released.

MIOS83-0010A-RS03 2

NI '

i 16

^

2.0 Continuous Releases Low level continuous releases from the vent gas collection header and other low level sources are totaled on a weekly basis and summed with any batch releases for the week'in order to establish the

-cumulative DBQ fraction from. batch plus continuous releases for the year to date. Calculations are performed in the same manner as for batch releases described in C.l.0.

= 3. 0' Exceeding DBQ Limits

~

As discussed under B.l.3, the DBQ is a very conservative estimate of activity which could give doses at Appendix I limits. Because different organs are summed together and doses to different people are summed, the DBQ typically overestimates dose by about a factor of five. Thus, if calculations of DBQ fraction exceed 1.0 for year-to-date or 0.5 for the quarter, technical specifications probably still would not be exceeded. However, further discre-tionary releases should be deferred until an accurate assessment of dose is made by use of GASPAR computer code or by analysis of

appropriate _ release data via the segmented gaussian dose model used

.in emergency planning (inhalation dose, total body external dose, and, boundary dose in air). See also Section D.1.2.

EIt should be noted that Palisades Plant to date (based on review of semiannual effluent data) has never exceeded the annual or quarterly DBQ fraction, despite its conservatism. Thus, it is not expected that_an alternate to the DBQ method will be required unless the plant is in a significantly off-normal condition.

4.0 Releasing Radionuclides Not' Listed in Table 1.9

, Table 1.9 contains all nuclides identified to date as routine constituents of gaseous releases at Palisades Plant, plus those MIO583-0010A-RS03

17 common to PWRs in general, even if not previously detected at Palisades. From time to time, however, other nuclides may be

' detected.

If the unlisted nuclide constitutes less than 10% of the

-MPC-fraction for the release, and all unlisted nuclides total less than 25% of the MPC-fraction, the nuclide may be considered not present.

If the unlisted nuclide constitutes greater than 10% of the MPC-fraction, or all unlisted nuclides together constitute greater than

'25%, then each nuclide should be assigned a DBQ equal to the most conservative value listed for the physical form of the nuclide involved (noble gas, halogen or particulate).

~

Should a nuclide not-listed in Table 1.9 begin to appear in signifi-cant quantities on a routine basis, revision to this ODCM should be made in order to include a design basis quantity specific to that nuclide.

D. OPTIONAL QUARTERLY DOSE CALCULATIONS 1.0 Methodology for Optional Quarterly Dose Calculations

, -This option may be used in place.of, or in addition to, the design basis quantity (DBQ) fraction calculation described by Equation 1.4.

This optional conservative calculation relates the DBQ fraction to the doses from which it was originally derived. Use-of this method may assist in identification of the critical dose pathway or char-

.acteristics of the assumed critical individual (infant, child, adult), since Table 1.9 indicates these parameters.

.MIO583-0010A-RS03

18

,dk 1.1 Simplified Conservative Approach This method utilizes a limiting dose concept such that the limiting dose for each nuclide is summed with the limiting dose for each other nuclide, regardless if cuch sum is physically possible. It also assumes critical pathways, such as milk and vegetables, are in effect even in winter when the pathway is absent.

As such, the method is highly conservative and significantly over-estimates dose. If limits appear to be exceeded by this method, Section D.I.2 (a concise method, but requiring computer support) will be utilized.

1.1.1' Assumptions-

1. All assumptions of Section 1.1 are utilized.

2. Limiting doses for each gaseous nuclide are summed, regardless of limiting decay mode (gamma or beta).

3. Limiting doses for each particulate and iodine nuclide are summed, regardless of dose point location, exposure pathway or organ affected.

4. -Doses are summed for detected nuclides such that all nuclides which contribute greater than 10% individu' ally or 25% in aggregate, to the MPC of released radioactivity, are included in the dose calculation.

1.1.2 Equations For determining gaseous effluent dose:

g Dg = A$g(Dc I c LG < ' l4" * ' "' lY'

-MIO583-0010A-RS03

)

19 l where:

Dg = Dose from gaseous effluents (mrad).

A = Quantity of gaseous nuclide i released (Ci).

iG

= Dose per Ci factor for gaseous nuclide i (mrad /Ci).

(D /C )iG The-limit for this mixture is conservatively taken as that for gamma exposure (5 mrem / quarter, 10 mrem / year) although as indicated in Table 1.9, a majority of the gaseous effluents are beta-limiting and on an individual basis have the higher limit of 10 millirem / quarter and 20 millirem / year.

For determining iodine and particulate dose to organs:

i D py = Apy (D /C )pyg - ( 7.5 mrem /q, 15 mrem /y 0

where:

D py. = Dose from particulates and iodines (mrem).

Apyg = Quantity of particulate or iodine nuclide i released (Ci).

(D /C )py = Dose per Ci factor for particulate or iodine nuclide 1 (mrad /Ci).

1.2 Realistic Calculations This methodology'is to be used if the h'ighly conservative calcula-

~

tions described in C.I.1, C.I.2 or D.l.0 yield values that appear to exceed applicable limits.

Doses for_ released particulates, iodines and noble gases will be 1 determined by use of the NRC GASPAR computer code. The computer run will-utilize the: annual average joint frequency meteorological data

, 4 based on not less than 3 years of meteorological measurement, and will reflect demographic and land use information from the land use MIO583-0010A-ks03

-m

'i 20 survey generated in the uost recent prior year. Where appropriate,

' seasonal adjustments will be applied to obtain realistic dose estimates since both recreational and agricultural activities can vary greatly in relation to season of the year.

An alternative to GASPAR for offsite dose calculation is the use of the Palisades Segmented Gaussian Plume Emergency offsite dose calculation program. This dose model allows evaluation of dose under the actual meteorological conditions present at the time of release. It is anticipated that the system may be used if major short-term releases such as containment purges are to be made under conditions which depart significantly from mean annual conditions.

E. GASEOUS RADWASTE TREATMENT SYSTEM OPERATION

The gaseous radwaste treatment system (GRTS) described below shall be maintained and operated to keep releases ALARA.

1. System Description A flow diagram for the GRTS is given_ in Figure 1-1. The system consists of three waste gas compressor packages, six gas decay tanks, and the associated piping, valves, and instrumentation.

Gaseous wastes are received from the following: degassing of the reactor coolant and purging of the volume control tank prior to a cold shutdown, displacing of cover gases caused by liquid accumulation in the tanks coanected to the vent header, and boron recycle process operation.

Design of the system precludes hydrogen explosion by means of ignition source elimination (diaphragm valves, low flow diaphragm compressors and system electrical grounding), and minimization of leakage outside the system. Explosive mixtures of hydrogen and oxygen have been demonstrated compatible with the system by operational experience over the past 13 years.

MIO583-0010A-RS03

G:

21

2. Determination of Satisfactory Operation Design basis quantity fraction will be calculated for batch and continuous releases as described in Section C. These calcula-tions will be used to ensure that the GRTS is operating as designed. Because the plant was designed to collect and hold

-for decay a vast majority of the high level gasses generated within the primary system, and because the 13-year operating

-history (to date of writing the initial ODCM) of the plant has demonstrated the system's consistent performance well below Appendix I limits, no additional operability requirements are specified.

F. RELEASE RATE.FOR OFFSI'Io MPC (500 mrem /yr) 4 10 CFR 20.106 requires radioactive effluent releases to unrestricted areas be in concentrations-less than the limits specified in Appen-dix B, Table II when averaged over a period not to exceed one year.

(Note

There are no unrestricted areas anywhere within the. site boundary as defined-by Figure 2-1.) Concentrations at this level if present for one year will' result in a dose of -500 mrem due to external exposure or inhslation depending on the auclide(s) released. 101CFR 50.36a requires that the release of' radioactive

i

. materials be kept as low as reasonably achievable. However, the

_ section.further states that the licensee is permitted the flex-ibility of operation,' to assure a dependable source of power, to JE release quantities of material higher than-a small percentage of 10:CFR 20-limits but not exceeding those limits under unusual operating conditions. --Appendix.I to 10 CFR 50 provides the numer-

~

- ical~ guidelines on limiting conditions for operations to meet the as low as reasonably achievable reqtirement.

4 The GASPAR code has been run to determine the dose due to external

- radiation and inhalation. - The source term used is listed in Table

' 1.1. The meteorology data is given in Table 1.3. Dose using annual MIO583-0010A-Rh03 q ,

j _

gar 22 average meteorology, to the most limiting organ of the person assumed to be residing at the site boundary with highest X/Q, is 2.15E-02 mrem (for one year). The release rate which would result in a dose rate equivalent to 500 mrem / year (using the more conserva-tive total body limit) is the Curies / year given in Table 1.1 multi-plied by 500/2.15E-02 or 1.11 Ci/sec.

G. PARTICULATE AND IODINE SAMPLING Particulate and iodine samples are obtained from the continuous sample stream pulled from the plant stack. Samples typically are obtained to represent an integrated release from a gas batch (waste gas decay tank or containment purge, for example), or a series of samples are obtained to follow the course of a release. In any event, sample intervals are weekly, at minimum.

Because HEPA filters are present between most source inputs to the stack and the sample point, releases of particulates normally are

~significantly less than pre-release calculations indicate. This provides for conservatism in establishing setpoints and in esti-mation of pre-release design basis quantity fraction. However, for the sake of maintaining accurate release totals, monitor results I

(for gasses) and sample results (for particulates and iodines) are utilized rather than the pre-release estimates, for cumulative records.

Gamma analytical results for particulate and halogen filters are combined for determination of total activity of particulates and halogens released. Beta and alpha counting also is performed on the particulate filters. Beta yields of the gamma isotopes detected on particulate filters are applied to determine " identified" beta, and the " identified" count rate is subtracted from the observed count rate to give " unidentified" beta. The " unidentified" beta is assumed to be Sr-90 until results on actual Sr-90 (chemically MIO583-0010A-RS03 m

23 separated from a quarterly composite of filters) are obtained.

Similarly, alpha activity not identified as natural radium or thorium or their daughters is assumed as Pu-239 until results of detailed analyses are obtained from quarterly composites.

H. NOBLE GAS SAMPLING Noble gasses will be sampled from Waste Gas Decay Tanks prior to release and the Containment prior to purging. Analysis of these samples will be used for accountability of noble gasses. Off gas will be sampled at least weekly and used to calculate monthly noble gas releases. Nonroutine releases will be quantified from the stack

-' noble gas monitor (RE 2326) which has a LLD of IE-06 pCi/cc (if RE1815 is used because RE 2326 is out of service, the LLD will be SE-05 pCi/cc).

' I. TRITIUM SAMPLING Tritium has a low dose consequence to the public because of low production rates. The major contributors to tritium effluents are evaporation from the fuel pool and reactor cavity (when flooded).

Because of the low doce impact, gaseous tritium sampling will not be

required. Tritium effluents will be estimated using conservative evaporation rate calculations from the fuel pool and reactor cavity.

7 MIO583-0010A-RS03 a

FIGtlRF I-1

HiGH -

ACTIVITY 1 r PALISADES GASEOUS EFFLUENTS GAS pg T-67 -68A

[

T-68B

[

T-68C TIO1A

[ T101C T101B STACK V y hFLOW CONTROL

- 1 I I I IODINE /PART

' Y RE ~~

WASTE GAS Y r7 l 2325 2326 SURGE TANK '~ ~ i J-- 8 NOBLE RE 1113 GAS RE 2320 RE t

' p ATMOSPHERE 12327

__ u E

HEPA

! T-29 T-29B V V LOW LEVEL = 0 i

I GAS COLLECTION HEADER C ~

STEAM GENERATOR HEPA '

l BLOWDOWN TANK VENT RE mD1

' CONDENSER

- - , n OFF GAS

EAST 8 RE 1810 '

AUX l HEPA/ CHARCOAL GUARD SAFE d/ BLDG e HEPA

, . , , , SUPPLY F RA C:

l AREA RE 1809*

WE I HEPA p

qf  ;

GUARD RE 1811 HEPA (NORMAL)

CONTAIN-SUPPLY -o -- HANDLG l FUEL U r  ;

MENT BLDG

• AREA O HEPA/ CHARCOAL (REFUELING) p_________,

,_ -___ _, HEPA FUEL i RAD-  ! HEPA  : AREA -l/'  : SUPPLY

-l SUPPLY

, a WASTE 6 M -

RE 5712 ADDITN

i F AREA ta aJ -

ADDIT N RE 5711'

'RE 1809. 5711 AND 5712 TRIP SUPPLY AND ONE OF TWO EXHAUSTERS.

FLOW IS NOT TERMINATED PALISADES EFFLUENT FLOW PATHS i GASEOUS

I 25 l

TABLE 1.1 PALISADES GASEOUS AND LIQUID SOURCE TERMS, CURIES / YEAR (

Nuclide Gaseous ( } Liquid (

H-3 5.5 159 Kr-85 4.1 NA Kr-85m 0.12 NA Kr-87 8.4E-02 NA Kr-88 2.1E-01 NA Ar-41 3.1E-02 NA Xe-131m 2.2 NA Xe-133 1493 NA Xe-133m 0.43 NA Xe-135 1.11 NA Xe-135m. 0.3 NA I-131 0.025 3.21E-03 I-132 2.91E-03 NA I-133 6.5E-03 4.7E-05 I-134 4.8E-04 NA I-135 1.84E-02 NA

- Na-24 1.5E-06 NA Cr-51 2.5E-04 3.9E-03 Mn-54 4.1E-04 7.8E-03 Co-57 2.1E-06 3.2E-05 Co-58 .8.6E-04 2.9E-02 Fe-59 6.6E-06 4.1E-04 Co-60 1.1E-03 1.24E-02 Se 3.7E-06 NA Nb-95 2.4E-05 4.53E-04 Zr-95 4.7E-06 1.79E-04 Mo-99 -1.5E-07 NA Ru-103 .3E-07 .1E-05 Sb-127 NA 3.5E-05 Cs-134- 4.5E-05 0.7 Cs-136 NA 1.8E-06 Cs-137 2.6E-04 1.36E-02

.Ba-140 2.8E-07 NA

1. 'La-140 7.5E-07 1.1E-04 Uniientified beta 3.9E-04 3.3E-03 (1) Data derived from taking the effluents released during July-December

.- 1978 through January-June 1982 and dividing by 4.

(2) Nuclide values listed as NA have not been observed at detectable levels

9. in these-waste streams.

Y 9

M10583-0010A-RS03

26 TABLE 1.2 BASIC RADIONUCLIDE DATA NUCLIDE HALF-LIFE LAMBDA BETA GAMMA (days) (1/s) (MEV/ DIS) (MEV/ DIS) 1 Tritium 4.49E 03 1.79E-09 5.68E-03 0.0 2 C-14 2.09E 06 3.84E-12 5.17E-02 0.0 3 N-13 6.94E-03 1.16E-03 4.91E-01 1.02E 00 4 0-19 3.36E-04 2.39E-02 1.02E 00 1.05E 00

5. F-18 7.62E-02 1.05E-04 2.41E-01 9.88E-01 6 NA-24 6.33E-01 1.27E-05 5.55E-01 4.12E 00 7 P-32 1.43E 01 5.61E-07 6.95E-01 0.0 8 AB-41 7.63E-02. 1.05E-04 3.63E-01 1.28E 00 9 CR-51 2.78E 01 2.89E-07 3.75E-03 3.28E-02 10 MN-54 .3.03E 02 2.65E-08 4.17E 8.36E-01 11 MN-56 1.07E-01 7.50E-05 7.93E-01 1.76E 00 12 FE-59 4.50E 01 1.78E-07 1.18E-01 1.19E 00 13 CO 7.13E 01 1.12E-07 2.05E-01 9.78E-01 14 CO-60 1.92E 03 4.18E-09 9.68E-02 2.50E 00 15 ZN-69m 5.75E-01 1.39E-05 0.0 4.15E 00 16 ZN-69 3.96E-02 2.03E-04 3.19E-01 0.0 17 BR-84 2.21E-02 3.63E-04 1.28E 00 1.68E 00 18 BR-85 2.08E-03 3.86E-03 1.04E 00 8.40E-01

'19 KR-85m 1.83E-01 4.38E-05 2.53E-01 1.59E-01 12 0 KR-85 3.93E 03 2.04E-09 2.51E-01 2.21E-03 21 KR-87 '5.28E-02 1.52E-04 1.32E 00 7.93E-01 22 KR-88 1.17E-01 6.86E-05 3.75E-01 1.96E 00 23 KR-89 2.21E-03 3.63E-03 1.23E 00 2.08E 00 24 RB-88 1.24E-02 6.47E-04 2.06E 00 6.86E-01 25 RB-89 1.07E-02 7.50E-04 0.00 2.40E 00 26 SR-89 5.20E 01 1.54E-07 5.73E-01 1.36E-04 27 -SR-90 1.03E 04 7.79E-10 1.96E-01 0.0 28 SR-91 4.03E-01 1.99E-05 6.50E-01 6.95E-01 29- SR-92 1.13E-01 7.10E-05 1.95E-01 1.34E 00 30 SR-93 5.56E-03 1.44E-03 1.61E 00 6.28E-01 31 Y-90 2.67E 00 3.00E-06 9.36E-01 0.0 32 Y-91m. 3.47E-02 2.31E-04 0.0 5.56E-01 33 Y-91 5.88E 01 1.36E-07 6.06E-01 3.61E-03 34 Y-92 1.47E-01 5.46E-05 1.44E 00 2.50E-01 35 Y-93 4.29E-01 1.87E-05 1.17E 00 8.94E-02

-36 -ZB-95 6.50E 01 1.23E-07 1.20E-01 7.35E-01

-37 NB-95m 3.75E 00 2.14E-06 2.85E-01 5.87E-02 38 NB-95 3.50E 01 2.29E-07 4.50E-02 7.64E-01 39 'MO-99

- 2.79E 00 2.87E-06 3.96E-01 1.62E-01 40 TC-99m 2.50E-01 3.21E-05 4.85E-03 1.43E-01 41 TC-99 7.74E 07 1.04E-13 8.38E-02 0.0 42 TC-104 1.25E-02 6.42E-04 0.0 0.0 MIO583-0010A-RS03

27 TABLE 1.2 (CON'T)

^

BASIC RADIONUCLIDE DATA NUCLIDE HALF-LIFE LAMBDA BETA GAMMA (days) (1/s) (MEV/ DIS) (MEV/ DIS) i, 43- RU-106 3.67E 02 2.19E-08 1.01E-02 0.0 441 TE-132- ~3.24E 00 2.48E-06 1.00E-01 2.05E-01 45 I-129 6.21E 09 1.29E-15 4.02E-02 3.77E-03 46 I-131 8.05E 00 9.96E-07 1.94E-01 3.81E-01 47 MI-131 8.05E 00 9.96E-07 1.94E-01 3.81E-01 48 I-132- 9.58E-02 8.37E-05 5.14E-01 2.33E 00 49 'HI-132' 9.58E-02 8.37E-05 5.14E-01 2.33E 00 50 I-133 8.75E-01 9.17E-06 4.08E-01 6.10E-01 51 MI-133 8.75E-01 9.17E-06 4.08E-01 6.10E-01 152 I-134 3.61E-02 2.22E-04 6.10E-01 2.59E 00 53 MI-134 3.61E-02 2.22E-04 6.10E-01 2.59E 00 54 I-135 2.79E-01 2.87E-05 3.68E-01 1.58E 00

'55 MI-135. 2.79E-01 2.87E-05 3.68E-01 1.58E 00 56 XE-131m 1.18E 01 6.80E-07 1.43E-01 2.01E-02 57 -XE-133m 2.26E 00 3.55E-06 1.90E-01 4.16E-02 58 XE-133' 5.27E 00 1.52E-06 1.35E-01 4.54E-02 59 .XE-135m 1.08E-02 7.43E-04 9.50E-02 4.32E-01 i., 60 'XE-135- 3.83E-01 2.09E-05 3.17E-01 2.47E-01 61 XE-137 2.71E-03 2.96E-03 1.64E 00 1.94E-01 62 XE-138 1.18E-02 6.80E-04 6.06E-01 1.18E 00 63 'CS-134- 7.48E 02 .1.07E-08 1.57E-01 1.04E 00 64 CS-135 1.10E 09 7.29E-15 5.74E-02 0.0

-65 CS-136 1.30E 01 6.17E-07 1.01E-01 2.20E 00 66 CS-137 1.10E 04 7.29E-10 2.52E-01 5.97E-01 67 CS-138 '2.24E-02 3.58E-04 1.23E 00 2.30E 00

68 BA-139 5.76E-02 1.39E-04 6.54E-01 5.50E-02

'69 'BA-140 'I.28E 01 6.27E-07 3.15E-01 1.95E-01

'70L LA-140 1.68E 00 4.77E-06 5.40E-01 2.31E 00 71 CE-144 2.84E 02 2.82E-08 9.13E-02 3.24E-09 72 PB-143 1.36E 01 .5.90E-07 3.14E-01 0.0 73 PR-144 1.20E-02 6.68E-09 1.21E 00 2.18E 00 MIO583-0010A-RS03 s;

28 Table 1.3

.GICUND LEVEL RELEASE - TOP OF CONTAINMENT SUILDING I 00 DECAY, UNDEPLETED ame-. I j

ANNUAL AVERAGE CN!/4 (SEC/ METER CUSED) DISTANCE IN MILES SECTOR 0.250 0.500 . 0.750 1.000 1.500 2.000 2.500 3.000 3.500 )

4.000 4.500 j SSW 3. 00 9E-06 1. 00 3E-06 5.514 E-07 3.6 33E-07 2.013E-0 7 1.327E-07 9. 619E-08 7.397 E-08 5. 927 E-08 4.8 95 E-08 4.136 '

SW 2.4 57E-06 8.043E-0 7 4.376 E-37 2.8 9 0 E-07 1.610E-07 2. 0 68E-0 7 7.7 74 E-08 6. 0 02 E-08 4.826 E-08 3. 9 96 E-08 3.l WSW 2.942E-06 9.490E-07 5.050E-07 3.32fE-07 1.853E-07 1.236E-07 9.050E-08 7.019E-08 5.665E-08 4.707E-08 3.999E-08 W 2.830E-06 9.005E-07 4.735E-07 3.109E-07 1.731E-07 1.161E-07 8.548E-08 6.658E-04 5.393E-08 4.495E-08 3.830E-08 j WNW ~ 3.7 06E-06 1.175E-06 6.161E-07 4. 037E-0 7 2.245E-07 1.504E-0 7 1.111 E-0 7 8.66 3E-08 7. 024 E-08 5.85 9E-08 4. 9 95 E-08 NW 4.623E-06 1.463E-06 7.674E-0 7 5. 018E-0 7 2.784 E-07 1.870 E-0 7 1.378E-07 1. 0 75E-0 7 8.717E-08 7.27 3 E-08 6.2 02E-08 l MNW 7.102E-06 2.229E-06 1.165E-06 7.615E-07 4.230E-07 2.847E-07 2.202E-07 1.642E-07 1.333E-07 1.114E-07 9.505E-08 ,

M 9.645E-06 2.971E-06 1.536E-06 1.001E-06 5.560E-07 3.765E-07 2.795E-07 2.193E-07 1.787E-07 1.497E-07 1.282E-07 i NME 8.47 3E-96 2.627 E-06 1.374E-06 8. 976 E-07 4.989 E-07 3.365E-0 7 2.4 9 0 E-07 1. 948 E-0 7 1.584 E-07 1.325 E-0 7 1.132E-0 NE 4.4 02E-06 1.386 E-06 7. 335E-0 7 4.8 08 E-07 2.675E-0 7 1.7 96 E-0 7 1. 323E-07 1. 0 32E-0 7 8.362E-08 6. 975 E-08 5.946 E-0 ENE 4.261E-06 1.398 E-06 7.5 56 E-07 4.958E-0 7 2.740E-07 1.8 06 E-07 1. 30 9E-07 1. 0 0 7 E-07 8.0 72 E-08 6. 66 9E-08 5.6 37E-0 E 3.472E-06 1.200E-06 3.590E-06 1.139E-06 6,486E-07 6.120E-07 4.228E-07 4.001E-07 2.316E-072.203E-07 1.515E-07 1.446E-07 1.092E-07 1.045E-07 8.356E-08 4.014E-08 6.671E-08 6.409E-08 5.491E-08 5.285E-08 4.627E-0F 4.460E-08 ESE SE 3.375E-96 1.130E-06 4.250E-06 1.423E-06 6. 075E-07 3. 948 E-0 7 2.155E-0 7 1.4 08E-07 1. 013E-07 7.752E-08 6.185E-08 5. 088 E-08 4.286 E-08 i

$$E 7.669E-07 4.976E-07 2.710E-07 1.766E-c7 1.268E-07 9.683E-08 7.713E-08 6.338E-08 5.332E-08 '

4.306E-06 1.433E-06 7.784E-07 5.078E-07 2.782E-07 1.820E-07 1.311E-07 1.004E-07 8.013E-08 6.596E-08 5.558E-08 i ANNUAL AVERAGE CN!/4 (SEC/ METER CUSED) DISTANCE IN MILES GEARING 5.000 7.500 10.000 15.000 20.000 25.000 30.000 35.000 40.000 45.000 50.000 SSW 3.559E-08 2.0 03E-08 1.334 E-08 7.58 9E-0 9 5. 0 99E-0 9 3.752E-0 9 2. 924E-0 9 2. 36 9E-0 9 1. 976 E-0 9 1.68 5E-0 9 1.461E-0 9 SW 2.919E-08 1.656E-08 1.111E-04 6.364E-09 4.300E-09 3.178E-09 2.485E-09 2.020E-09 1.690E-09 1.444E-09 1.255E-09 W5W 3.4 58 E-08 1. 98 3 E-08 1.341 E-08 7.76 5 E-0 9 5.28 9E-0 9 3. 934 E-0 9 3. 0 92E-0 9 2. 525 E-0 9 2.12 0 E-0 9 1.817E-0 9 1.58 4E-0 9 W 3.320 E-08 1. 921 E-08 1.308 E-08 7.64 0 E-0 9 5.236 E-0 9 3.914 E-0 9 3.08 9E-0 9 2.531E-09 2.131 E-09 1.831E-0 9 1.6 0 0E-0 9 WNW 4.333E-08 2.514E-08 1.714E-08 1.004E-04 6.894E-09 5.160E-09 4.077E-09 3.343E-09 2.817E-09 2.423E-09 2 119E-09 NW 5.381E-08 3.125E-08 2.133E-08 1.251E-08 8.594E-09 6.439E-09 5.091E-09 4.177E-09 3.521E-09 3.030E-09 2.650E-09 MMW 8.254E-04 4.810E-OS 3.249E-08 1.935E-08 1.335E-08 9.997E-09 7.914E-09 6.500E-09 5.485E-09 4.724E-09 4.134E-09 N 1.115E-07 6.560E-08 4.515E-08 2.679E-08 1.456E-08 1.399E-04 1.111E-08 9.153E-09 7.743E-09 6.683E-09 5.861E-09 NME 9.8 37 E-08 5.753E-08 3. 944 E-08 2.328E-08 1.6 0 7E-08 1.208E-08 9.574 E-0 9 7.873E-0 9 6.64 9E-0 9 5.731E-0 9 5.02 0 E-09 NE 5.157E-08 2.993E-08 2.041E-08 1.195E-08 8.206E-09 6.141E-09 4.852E-09 3.979E-09 3.352E-09 2.883E-09 2.520E-09 ENE 4.854E-08 2.740E-08 1.833E-08 1.046E-04 7.056E-09 5.211E-09 4.072E-09 3.309E-09 2.766E-09 2.363E-09 2.053E-09 E 3.835E-08 2.155E-08 1.437E-08 8.171E-09 5.504E-0 9 4. 06 0 E-0 9 3.170E-09 2.574 E-0 9 2.15 0E-09 1.8 36 E-0 9 1.595E-09 ESE 3.973E-08 2.22 0 E-08 1.474E-08 8. 330 E-0 9 5.543E-0 9 4.10 4 E-0 9 3.195 E-0 9 2.588 E-0 9 2.158 E-0 9 1.8 3 9E-0 9 1.5 95E-09 SE 3.67 9E-08 2.055 E-04 1.364E-08 7.704E-0 9 5.16 5E-0 9 3.7 96 E-0 9 2.955 E-0 9 2.394 E-0 9 1. 996 E-0 9 1.7 01 E-0 9 1_ alae-0 9 SSE 4.573E-08 2.544E-04 1.684E-08 9.481E-09 6.347E-09 4.659E-09 3.624E-09 2.933E-09 2.444E-09 2.082E-09 1.805E-09 4.773E-08 2.668E-08 1.773 E-08 1.0 02E-08 6.722E-0 9 4. 941E-0 9 3.848E-0 9 3.117E-09 2.599E-0 9 2.215E-0 9 1. 921E-0 9 CH!/4 (SEC/ METER CUBED) FOR EACH SEGMENT SEGMENT B0UNDARIES IN MILES DIRECTION '.5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 FROM SITE 40-50 S 5.681E-07 2.068E-07 9.705E-08 5.954E-08 4.147E-08 2.052E-08 7.765E-09 3.780E-09 2.378E-09 SSW 4.539E-07 1.653E-07 7.839E-08 4.846E-08 1.688E-3.394E-08 1.695E-08 6.502E-09 3.200E-09 2.027E-09 1.447E-SW 5.270E-07 1.906E-07 9.121E-08 5.687E-08 4.009E-08 2.026E-08 7.919E-09 3.959E-09 WSW 4.961E-07 1.784E-07 8.609E-08 5.412E-08 2.533E-09 1.821E-3.838E-08 1.959E-08 7.780E-09 3.936E-09 2.538E-09 1.434E-W 6.459E-07 2.315E-07 1.119E-07 7.048E-08 5.006E-08 2.$63E-08 1.022E-08 5.189E-09 3.352E-09 W'NW 4.038E-07 2.874E-07 1.388E-07 8.747E-08 2.427E-6.215E-08 3.185E-08 1.773E-08 6.475E-09 4.188E-09 3.035E-NW 1.222E-06 4.368E-07 2.117E-07 1.338E-07 9.525E-08 4.899E-08 1.968E-08 1.005E-08 6.5172-09 4.731E-NNW 3.617E*66 5.751E-07 2.813E-07 1.793E-07 1.284E-07 6.672E-08 2.721E-08 1.406E-08 9.175E-09 6.693E-M 1.441E-06 5.153E-07 2.507E-07 1.589E-07 1.134E-07 5.857E-08 2.367E-08 1.214E-08 7.892E-09 5.740E-NME 7.663E-07 2.758E-07 1.333E-07 8.392E-08 5.939E-08 3.051E-08 1.217E-08 6.176E-09 3.989E-09 2.488E-NE 7.828E-07 2.818E-07 1.321E-07 8.198E-08 5.652E-08 2.807E-08 1.069E-08 5.247E-09 3.320E-09 2.368E-ENE 6.349E-07 2.266E-07 1.054E-07 6.439E-08 4.473E-08 2.209E-08 8.364E-09 4.089E-09 2.553E-09 1.840E-0 E 6.70FE-07 2.385E-07 1.102E-07 6.703E-08 4.641E-08 2.278E-08 8.535E-09 4.136E-09 2.598E-09 1.843E-S ESE 6.290E*07 2.221E-07 1.023E-07 6.215E-08 4.299E-08 2.109E-08 7.8 95 E-0 9 3.825 E-0 9 2.403E-09 1.705E-SE 7.929E-07 2.794E-07 1.281E-07 7.752E-01 5.349E-08 2.613E-08 9.724E-09 4.695E-09 2.944E-09 2.087E-R SSE 8.036E-07 2.465E-07 1.324E-07 8.051E-08 5.575E-08 2.738E-08 1.027E-08 4.978E-09 3.128E-09 2.220E-VENT AND BUILDING PARAMETER $8

.. RELEASE NE2GHT (METEks) 58.10 REP. WIND HEIGMT (METERS) 10.0 DIAMETER (METERS) 0.0 SUILDING HEIGHT (METERS) 58.1 EXIT VELDCITY (METER 5) 0.0 BLDG. MIN.CRS.5EC AREA (5G.NETERS) 2000,0 NEAT EMIS$10N RATE (CAL /5EC) 0.0 AT THE RELEASE ME!OHT AT THE MEASURED WIND HEIGHT C 10.0 METERS):

VENT RELEASE MODE WIND SPEED (METERS /SEC) VENT RELEASE MODE WIND SPEED (METERS /SEC) WIND SPEED (METERS /SEC)

STABLE CCNDITION5 UNSTABLE / NEUTRAL CONDIT!

ELEVATED LESS THAN 0.0 ELEVATED LESS THAN 0.0 LESS THAN 0.0 MIXED SETWEEN 9.0 AND 0.0 MIXED .0 0.0 GROUND LEVEL A80VE SETWEEN AND 8ETWEEN 0.0 AND 0.

0.0 GRCUND LEVEL ABDVE 0.0 ABOVE 0.0 l

29 Table 1 3 (Contd)

GROUND LEVEL RELEASE - TCP OF CONTAINMENT 8UILDING 2.260 DAY DECAY, UNDEPLETED (NNUAt AVERAGE CN1/0 (SEC/ METER CUBED) DISTANCE IN MILES SECTOR 0.250 0.500 0.750 1.000 1.500 2.000 2.500 3.000 3.500 4.000 4.500 5 3.008E-06 1.002E-06 5.506E-07 3.626E-07 2.007E-07 1.322E-07 9.571E-08 7.353E-08 5.886E-08 4.856E-08 4.099E-08 SSW 2.455E-06 8.074E-07 4.369E-07 2.883E-07 1.604E-07 1.063E-07 7.731E-08 5.963E-08 4.788E-08 3.961E-08 3.352E-08 SW 2.940E-06 9.479E-07 5.042E-07 3.318E-07 1.847E-07 1.231E-07 9.001E-08 6.973E-08 5.622E-08 4.666E-08 3.960E-08 WSW 2.828E-06 8.995E-07 4.727E-07 3.101E-07 1.725E-07 1.156E-07 8.499E-08 6.612E-08 5.349E-08 4.454E-08 3.790E-08 W 3.704E-06 1.174E-06 6.151E-07 4.028E-07 2.237E-07 1.501E-07 1.105E-07 8.606E-08 6.969E-08 5.807E-08 4.946E-08 WNW 4.620E-06.1.461E-06 7.661E-07 5.007E-07 2.774E-07 1.861E-07 1.370E-07 1. 06 BE-0 7 8.6 48 E-08 7.208 E-08 6.13 9 E-08 NW 7.098E-06 2.226E-06 1.163E-06 7.598E-07 4.215E-07 2.834E-07 2.090E-07 1.631E-07 1.323E-07 1.103E-07 9.407E-08 NNW 9.640 E- 06 2. 96 7 E-0 6 1.5 33 E-06 9. 98 9E-0 7 5. 541E-07 3.747 E-07 2. 77 9E-0 7 2.178 E-0 7 1.773E-07 1.484E-07 1.268E-07 N 8.469E-06 2.624E-06 1.371E-06 8.957E-07 4.973E-07 3.351E-07 2.477E-07 1.936r-07 1.5720-07 1.313E-07 1.121E-07 NME 4.400E-06 1.385E-06 7.323E-07 4.798E-07 2.667E-07 1.789E-07 1.316E-07 1.025E-07 8.302E-08 6.917E-08 5.891E-08 ,

NE 4.259E-06 1.396E-06 7.546E-07 4.948E-07 2.732E-07 1.799E-07 1.303E-07 1.001E-07 8.017E-08 6.616E-08 5.547E-08 I ENE 3.470E-06 1.137E-06 6.108E-07 3.991E-07 2.194E-07 1.438E-07 1.038E-07 7.949E-08 6.349E-08 5.228E-08 4.406E-08 E 3.588E-06 1.199E-06 6.476E-07 4.219E-07 2.308E-07 1.508E-07 1.086E-07 8.301E-08 6.620E-08 5.443E-08 4.581E-08 ESE .3.373E-06 1.129E-06 6.065E-07 3.940E-07 2.148E-07 1.402E-07 1.008E-07 7.700E-08 6.137E-08 5.043E-08 4.243E-08 SE '4.247E-06 1.421E-06 7.655E-07 4.965E-07 2.700E-07 1.757E-07 1.260E-07 9.613E-08 7.648E-08 6.277E-08 5.274E-08

$5E 4.303E-06 1.431E-06 7.772E-07 5.067E-07 2.773E-07 1.813E-07 1.304E-07 9.973E-08 7.9531-08 6.539E-08 5.504E-08 AMNUAL AVERAGE CNf/4 (SEC/ METER CU8ED) DISTANCE IN MILES BEARING 5.000 7.500 18.000 15.000 20.000 25.000 30.000 35.000 40.000 45.000 50.000

$ 3.524E-08 1.973E-08 1.310E-08 7.363E-09 4.897E-09 3.567E-09 2.751E-09 2.207E-09 1.822E-09 1.538E-09 1.320E-09

$5W 2.887 E-08 1.62 9 E-08 1. 087 E-08 6.153E-0 9 4.111 E-0 9 3. 0 04 E-0 9 2. 32 3 E-0 9 1.868 E-0 9 1.544 E-0 9 1. 3 0 5 E-0 9 1.122 E-0 9 SW 3.421E-08 1.951E-08 1.312E-08 7.514E-09 5.062E-09 3.724E-09 2.896E-09 2.339E-09 1.942E-09 1.647E-09 1.421E-09 WSW 3.282E-08 1.889E-08 1.278E-08 7.381E-09 5.001E-09 3.696E-09 2.884E-09 2.326E-09 1.944E-09 1.653E-09 1.428E-09 W 4.285E-08 2.473E-08 1.677E-08 9.711E-09 6.595E-09 4.882E-09 3.815E-09 3.094E-09 2.579E-09 2.194E-09 1.898E-09 WNW 5.320E-08 3.073E-08 2.085E-08 1.2092-08 8.220E-09 6.090E-09 4.763E-09 3.866E-09 3.225E-09 2.746E-09 2.376E-09 MW 8.159 E-08 4.727 E-08 3.214 E-08 1.86 #E-0 8 1.27 3 E-08 9.447 E-0 9 7.397 E-0 9 6. 011E-0 9 5. 018 E-0 9 4.276 E-0 9 3.7 0 3 E-0 9 MNW 1.10 3 E-0 7 6.448E-08 4.412E-08 2.569E-08 1.774 E-08 1.322E-08 1. 0 3 9 E-08 8.47 3E-0 9 7.0 94 E-0 9 6. 061E-0 9 5.261E-0 9 ,

N 9.734E-08 5.663E-08 3.462E-0J 2.*56E-08 1.541E-08 1.147E-08 8.998E-09 7.325E-09 6.125E-09 5.227E-09 4.532E-09 i NNE 5.7 0 4 E-08 2. 947 E-08 1. 9 99 E-0 8 1 !$ 9E-08 7.87 3 E-0 9 5. 8 31 E-0 9 4. 56 0 E-0 9 3.7 01E-0 9 3. 08 6 E-0 9 2.6 27 E-0 9 2.2 7 3E-0 9 NE 4.8 0 6 E-08 2.7 0 0 E-08 1.7 97 E-08 1. 015E-0 8 6.777 E-0 9 4. 954 E-0 9 3.832E-0 9 3. 082 E-0 9 2.550 E-0 9 2.157 E-0 9 1.855E-09 ENE 3.783E-08 2.112E-08 1.399E-08 7.843E-09 5.214E-09 3.796E-09 2.926E-09 2.346E-09 1.936E-09 1.632E-09 1.401E-09 E 3.930E-08 2.184E-08 1.442E-08 8.058E-09 5.344E-09 3.884E-09 2.991E-09 2.396E-09 1.976E-09 1.666E-09 1.429E-09 '

ESE 3.634E-08 2.020E-08 1.3341-08 7.448E-0 9 4.9 38 E-0 9 3.588 E-0 9 2. 76 3 E-0 9 2.214 E-0 9 1.826 E-0 9 1. 5 3 9 E-0 9 1. 320 E-0 9 SE 4. 518 E-08 2.4 98 E-08 1'.644 E-06 9.141 E-0 9 6. 046 E-0 9 4. 386 E-0 9 3. 372E-0 9 2.6 98 E-0 9 2.22 3 E-0 9 1.872E-0 9 1. 6 04 E-0 9 SSE 4.722E-06 2.626E-08 1.735E-08 9.701E-09 6.438E-09 4.682E-09 3.608E-09 2.892E-09 2.387E-09 2.014E-09 1.728E-09 CN1/4 (SEC/ METER CUSED) FOR EACM SEGMENT SEGMENT 80VNDARIES IN MILES CIRECTION .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 FROM SITE 5 5.673E-07 2.062E-07 9.657E-08 5.913E-08 4.110E-05 2.023E-08 7.541E-09 3.595E-09 2.216E-09 1.541E-t

$$W 4.532E-07 1.648E-07 7.797E-08 4.801E-05 3.360E-08 1.668E-08 6.293E-09 3.027E-09 1.875E-09 1.308E-t SW 5.262E-07 1.900E-07 9.071E-08 5.644E-08 3.970E-08 1.994E-08 7.670E-09 3.750E-09 2.347E-09 1.651E-t WSW 4.953E-07 1.778E-07 8.560E-08 5.369E-08 3.798E-08 1.927E-08 7.523E-09 3.719E-09 2.343E-09 1.656E-t W 6.449E-07 2.308E-07 1.113E-07 6.994E-08 4.9f*E-08 2.522E-08 9.894E-09 4.912E-09 3.104E-09 2.199E-t WNW 8.025E-07 2.865E-07 1.380E-07 8.679E-08 6.15aE-08 3.133E-08 1.232E-08 6.127E-09 3.878E-09 2.751E-t NW 1.220E-06 4.353E-07 2.105E-07 1.327E-07 9.427E-08 4.817E-08 1.903E-08 9.503E-09 6.029E-09 4.284E-t NNW 1.614E-06 5.732E-07 2.797E-07 1.778E-07 1.271E-07 6.561E-08 2.632E-08 1.330E-08 8.496E-09 6.071E-t N 1.438E-06 5.137E-07 2.494E-07 1.578E-07 1.123E-07 5.767F-08 2.295E-08 1.153E-08 7.346E-09 5.23&E-t NNE 7.652E-07 2.750E-07 1.326E-07 .8.332E-08 5.904E-08 3.005E-08 1.180E-08 5.867E-09 3.712E-09 2.632E-t FE 7.818E-07 2.810E-07 1.114E-07 8.053E-08 5.603E-08 2.766E-08 1.039E-08 4.991E-09 3.094E-09 2.162E-t ENE 6.337E-07 2.258E-07 1.047E-07 6.379E-08 4.419E-08 2.166E-08 8.039E-09 3.826E-09 2.355E-09 1.636E-t E 6.698E-07 2.377E-07 1.096E-07 6.652E-08 4.595E-f8 2.242E-98 8.266E-09 3.916E-09 2.406E-09 1.670E-t ESE 6.281E-07 2.214E-07 1.018E-07 6.167E-08 4.256E-08 2.075E-08 7.641E-09 3.618E-09 2.223E-09 1.543E-t SE 7.916E-07 2.744E-07 1.273E-07 7.687E-08 5.291E-08 2.567E-08 9.387E-09 4.423E-09 2.710E-09 1.877E-t SSE 8.024E-07 2.456E-07 1.317E-07 7.992E-08 5.521E-08 2.695E-08 9.950E-09 4.721E-09 2.904E-09 2.018E-t VENT AND BUILDING PARAMETERS RELEASE NEIGHT (METER 5) 58.10 REP. WIND NEIGNT (METERS) 10.0 DIAMETER (METERS) 0.0 BUILDING NEIGHT (METERS) 58.1 EX1T VELCCITY (METERS) s.O BLDG. MIN.CRS.5EC. AREA (50. METERS) 2000.0 NEAT EMISSION RATE (CAL /5EC) 0.0 AT THE RELEASE NE!GMT* AT THE MEASURED WIND NEIGNT C 10.0 METERS):

VENT RELEASE MODE WIND SPEED (METERS /5EC) VENT RELEASE MODE WIND $ PEED (METdRS/5EC) WIND SPEED (METER $/$EC)

STABLE CONDITIONS UN5 TABLE / NEUTRAL CONDIT!!

ELEVATED LESS TNAN 0.0 ELEVATED LESS TNAN 0.0 LESS TNAN 0.d MIXED BETWEEN 0.0 AND 0.0 MIXED BETWEEN 0.0 AND 0.0 BETWEEN 0.0 AND 0.t GROUND LEVEL ABOVE 0.0 GROUND LEVEL ABOVE 0.0 ABOVE 0.0

30 Tcble 1 3 (Contd)

GROUND LEVEL RELEASE - 7CP DF CONTAINMENT BUILDING 8.000 DAY DECAY. DEPLETED ANNUAL AVERAGE CN1/4 (SEC/ METER CUBED) DISTANCE IN MILES SECTOR 0.250 0.500 0.750 1.000 1.500 2.000 2.500 3.000 3.500 4.000 4.500 '

~

S 2.848E-06 9.154E-0? 4.912E-07 3.178E-07 1.708E-07 1.098E-07 7.781E-08 5.866E-08 4.616E-08 3.749E-08 3.119E-08 SSW 2.325E-06 7.379E-07 3.898E-07 2.528E-07 1.366E-07 8.829E-08 6.288E-08 4.759E-08 3.757E-08 3.060E-08 2.552E-08 SW 2.784E-06 8.663E-07 4.498E-07 2.909E-07 1.572E-07 1.022E-07 7.320E-08 5.565E-08 4.411E-08 3.605E-08 3.015E-08 WSW 2.678E-06 8.221E-07 4.217E-07 2.719E-07 1.469E-07 9.604E-08 6.913E-08 5.278E-08 4.198E-08 3.442E-08 2.887E-08 W. 3.507E-06 1.073E-06 5.488E-07 3.532E-07 1.905E-07 1.247E-07 8.986E-08 6.868E-08 5.469E-08 4.486E-08 3.766E-08 WNW 4.374E-06 1.335E-06 6.835E-07 4.390E-07 2.362E-07 1.546E-07 1.115E-07 8.521E-08 6.786E-08 5.569E-08 4.676E-08 NW 6.720E-06 2.035E-06 1.037E-06 6.662E-07 3.589E-07 2.354E-07 1.700E-07 1.302E-07 1.038E-07 8.526E-08 7.166E-08 NNW 9.127E-06 2.712E-06 1.368E-06 8.754E-07 4.718E-07 3.115E-07 2.260E-07 1.738E-07 1.391E-07 1.146E-07 9.661E-08 N 8.018E-06 2.398E-06 1.224E-06 7.852E-07 4.233E-07 2.783E-07 2.014E-07 1.545E-07 1.233E-07 1.014E-07 8.534E-08 NME 4.165E-06 1.266E-06 6.533E-07 4.207E-07 2.270E-07 1.486E-07 1.070E-07 8.180E-C8 6.512E-08 5.342E-08 4.484E-08 NE 4.032E-06 1.276E-06 6.731E-07 4.337E-07 2.325E-07 1.494E-07 1.059E-07 7.986E-08 6.287E-08 5.198E-08 4.251E-08 ENE 3.286E-06 1.039E-06 5.451E-07 3.500E-07 1.869E-07 1.196E-07 8.445E-08 6.351E-08 4.988E-08 4.044E-08 3.360E-08 E 3.397E-06 1.096E-06 5.777E-07 3.699E-07 1.965E-07 1.253E-07 8.828E-08 6.625E-08 5.194E-08 4.20$E-08 3.489E-08 ESE 3.194E-06 1.031E-06 5.411E-07 3.454E-07 1.828E-07 1.164E-07 8.196E-08 6.146E-08 4.815E-08 3.896E-08 3.231E-88 SE 4.021E-06 1.299E-06 6.830E-07 4.353E-07 2.299E-07 1.460E-07 1.025E-07 7.675E-08 6.004E-08 4.852E-08 4.019E-08 SSE 4.074E-06 1.308E-06 6.934E-07 4.442E-07 2.361E-07 1.505E-07 1.061E-07 7.958E-08 6.239E-08 5.051E-08 4.191E-Oc ANNUAL AVERAGE CN!/4 (SEC/ METER CUBED) DISTANCE IN MILES BLARING 5.000 7.500 10.000 15.000 20.000 25.000 30.000 35.000 40.000 45.000 50.000 5 2.6 45 E-08 1.406 E-0 8 8. 926 E-0 9 4.67 9E-0 9 2. 944 E-0 9 2.047 E-0 9 1.516 E-0 9 1.172 E-0 9 9. 36 5E-10 7.6 6 5 E-10 6.3 96 E-10 SSW 2.16 9 E-08 1.16 2E-08 7.42 0 E-0 9 3. 920 E-0 9 2.47 9E-0 9 1.7 31E-0 9 1.286 E-0 9 9. 9 7 6 E-10 7. 98 7 E-10 6. 551 E-10 5.476 E-10 SW 2.57 0 E-0 8 1.3 92E-08 8.955E* 0 9 4.784 E-0 9 3.0 51E-0 9 2.14 3E-0 9 1.6 01 E-0 9 1.247 E-4 9 1. 0 0 3E-0 9 8.25 3 E-10 6. 921 E-10 W5W 2.4 6 6 E-08 1.348E-0 8 8.72 9E-0 9 4.7 0 5E-0 9 3. 019 E-0 9 2.131E-0 9 1.5 98 E-0 9 1.24 9 E-0 9 1. 0 07 E-0 9 8.306 E-10 6. 98 0 E-10 W 3.220E-08 1.764E-08 1.144E-08 6.185E-09 3.976E-09 2.811E-09 2.110E-09 1.651E-09 1.332E-09 1.100E-39 9.253E-10 WNW 3.998E-08 2.193E-08 1.4 24 E-08 7.7P2 E-0 9 4. 956 E-5 9 3.506 E-0 9 2.6 34 E-0 9 2. 06 2E-0 9 1.6 6 5E-0 9 1.375E-0 9 1.157E-0 9 MW 6.132E-08 3.374E-08 2.195E-08 1.191E-08 7.680E-09 5.442E-09 4.093E-09 3.208E-09 2.591E-09 2.142E-09 1.803E-09 NNW 8.288 E-08 4.6 02 E-08 3. 013E-08 1.64 9 E-08 1. 0 70 E-08 7.612E-09 5.746 L-0 9 4. 517 E-0 9 3.6 58E-0 9 3. 031E-0 9 2.556 E-0 9 N 7.311E-08 4.038E-08 2.634E-08 1.435E-08 9.274E-09 6.585E-09 4.961E-09 3.894E-09 3.149E-09 2.607E-09 2.197E-09 NME 3.833E-08 2.101E-08 1.363E-08 7.367E-09 4.737E-09 3.349E-09 2.514E-09 1.968E-09 1.588E-09 1.311E-09 1.103E-09 NE 3.608E-08 1.924E-08 1.224E-08 6.447E-09 4.074E-09 2.842E-09 2.111E-09 1.637E-09 1.311E-09 1.075E-09 8.987E-10 ENE 2.847E-08 1.511E-08 9.520E-09 5.021E-09 3.164E-09 2.202E-09 1.632E-09 1.264E-09 1.010E-09 8.274E-10 6.907E-10 E 2.952 E-08 1.5 58 E-08 9.8 4 32-0 9 5.131E-0 9 3.221 E-0 9 2.235E-0 9 1.654 E-0 9 1.278 E-0 9 1.020 E-0 9 8.34 5 E-10 6. 96 0 E-10 ESE 2.7 34 E-08 1.442E-08 9 @5 E-0 9 4.7 4 5 E- 0 9 2. 978 E-0 9 2. 06 7 E-0 9 1.52 9 E-0 9 1.182 E-0 9 9.4 33E-10 7.718 E-10 6.4 38 E-10 SE 3.397E-08 1.784E-08 1.124E-08 5.834E-09 3.655E-09 2.533E-09 1.871E-09 1.445E-09 1.153E-09 9.422E-10 7.854E-10

$$E 3.547 E-08 1,873E-08 1.184 E-08 6.173E-0 9 3.8778-0 9 2.6 92E-0 9 1.9 92 E-0 9 1.54 0 E-0 9 1.229E-0 9 1. 0 06 E-0 9 8.391E-10 CMI/O (SEC/ METER CUBED) FOR EACH SEGMENT SEGMENT 80UNDARIES IN MILES DIRECTION .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 FROM SITE S 5.084E-07 1.764E-07 7.867E-08 4.643E-08 3.130E-08 1.453E-08 4.852E-09 2.074E-09 1.181E-09 7.699E-SSW 4.063E-07 1.409E-07 6.354E-08 3.778E-08 2.561E-08 1.199E-08 4.057E-09 1.752E-09 1.004E-09 6.578E-SW 4.717E-07 1.625E-07 7.392E-08 4.433E-08 3.025E-08 1.433E-08 4.940E-09 2.168E-09 1.255E-09 8.286E-WSW 4.441E-07 1.521E-07 6.977E-08 4.219E-08 2.896E-08 1.385E-08 4.85DE-09 2.154E-09 1.256E-09 8.337E-W 5.783E-07 1.974E-07 9.068E-08 5.494E-08 3.777E-08 1.812E-08 6.372E-09 2.841E-09 1.661E-09 1.104E-WNW 7.196E-07 2.450E-07 1.125E-07 6.818E-08 4.689E-08 2.252E-08 7.934E-09 3.544E-09 2.074E-09 1.380E-NW 1.094E-06 3.723E-07 1.715E-07 1.043E-07 7.186E-08 3.463E-08 1.226E-08 5.499E-09 3.226E-09 2.150E-NNW 1.440E-06 4.902E-07 2.279E-07 1.397E-07 9.686E-08 4.715E-08 1.695E-08 7.688E-09 4.541E-09 3.041E-N 1.290E-06 4.393E-07 2.031E-07 1.239E-07 8.558E-08 4.141E-08 1.476E-08 6.652E-09 3.915E-09 2.616E-NME 6.860E-07 2.352E-07 1.080E-07 6.543E-08 4.497E-08 2.158E-08 7.590E-09 3.385E-09 1.979E-09 1.316E-NE 7.007E-07 2.403E-07 1.071E-07 6.323E-08 4.267E-08 1.987E-08 6.681E-99 2.878E-09 1.648E-09 1.080E-EME 5.682E-07 1.932E-07 8.544E-08 5.018E-08 3.373E-LB 1.562E-08 5.209E-09 2.231E-09 1.273E-09 8.309E-E 6. 0 0'5'E- 0 7 2.034E-07 8.934E-08 5.226E-08 3.502E-08 1.613E-08 5.329E-09 2.265E-09 1.287E-09 8.382E-ESE 5.631E-07 1.895E-07 8.295E-08 4.845E-08 3.244E-08 1.493E-08 4.928E-09 2.094E-09 1.190E-09 7.752E-SE 7.097E-07 2.383E-07 1.038E-07 6.043E-08 4.035E-08 1.849E-08 6.066E-09 2.568t-09 1.455E-09 9.464E-SSE 7.193E-07 2.443E-07 1.073E-07 6.278E-08 4.207E-08 1.938E-08 6.411E-09 2.728E-09 1.551E-09 1.010E-VENT AND SUILDING PARAMETERS:

RELE8SE NEIGMT iMETERS) 58.10 REP. WIND HEIGHT (METER $) 10.0 DIAMETER (METER $) 0.0 BUILDING MEIGHT (METERS) 58.1 EXIT VELOCITY (METERS) 0.0 BLDG. MIN.CRS.SEC. AREA (50. METERS) 2000.0 NEAT EMISSION RATE (CAL /SEC) 0.0 AT THE RELEASE NEIGHT: AT THE MEASURED WIND HEIGHT ( 10.0 METERS):

VENT RELEASE MODE WIND SPEED (METERS /SEC) VENT RELEASE MODE WINJ SPEED EMETERS/SEC) WIND SPEED (METERS /SEC)

STABLE CONDITIONS UNSTA9LE/ NEUTRAL CONDIT!

ELEVATED LESS THAN 0.0 ELEVATED LESS THAN 0.0 LESS THAN 0.0 MIXED BETWEEN 0.0 AND 3.0 MIXED BETWEEN 0.0 AND 0.0 BETWEEN O.0 AND 0.

GROUND LEVEL ABOVE 0.0 GROUND LEVEL ABOVE 0.0 ABOVE 0.0

31 Tsble 1.3 (Contd)

GROUND LEVEL RE' LEASE - TOP OF CONTAINMENT SUILDING eneseesseesssammannan RELATIVE DEP051 TION PER UNIT AREA (Mas-2) AT FIXED FOINTS SY DOWNWIND SECTORS emmessessummaassa DIRECT!0M CISTANCES IM MILE 5 Ft0M SITE 0.25 0.50 0.75 1.00 1.50 2.00 2.50 3.00 3.50 4.00 4.50 5 2.696E-08 9.115E-09 4.680E-09 2.874E-09 1.433E-09 8.690E-10 5.875E-10 4.257E-10 3.237E-10 2.550E-10 2.065E-10

$$W 3.809E-08 6.119E-09 3.142E-09 1.929E-09 9.618E-10 5.833E-10 3.944E-10 2.858E-10 2.173E-10 1. 12E-10 1.386E-10 SW 1.921E-98 6.496E-89 3.336E-89 2.048E-09 1.021E-09 6.193E-10 4.187E-10 3.034E-10 2.307E-10 1.818E-10 1.471E-10 WSW 1.621E-08 5.480E-09 2.814E-09 1.728E-09 8.614E-10 5.224E-10 3.532E-10 2.560E-10 1.946E-10 1.533E-10 1.241E-10 '

W 2.248 E-0 8 7.6 01E-0 9 3.90 3E-0 9 2. 3 96 E-0 9 1.195 E-0 9 7.246 E-10 4.8 9 9E-10 3.5 50 E-10 2.6 9 9E-10 2.127 E-10 1.722E-10 WNW 3.084E-08 1.04 3E-08 5.35 5E-0 9 3.288E-0 9 1.6 3 9E-0 9 9. 942E-10 6.722E-10 4.871E-10 3.784 E-10 2. 918 E-10 2.362E-10 MW 4.86 6 E-08 1.6 46 E-04 8.44 9 E-0 9 5.188E-0 9 2.584 E-0 9 1.56 9 E-0 9 1.061E-0 9 7.685E-10 5.8 44 E-10 4.6 0 4 E-10 3.727 E-10 ,

NMW 6.104E-08 2.064E-08 1.060E-08 6.507E-09 3.244E-09 1.96&E-09 1.330E-09 9.640E-10 7.330E-10 5.775E-10 %.675E-10 '

N 5.189E-04 1.755E-08 9.009E-09 5.532E-09 2.758E-09 1.673E-09 1.131E-J9 8.195E-10 6.231E-10 4.909E*10 3.974E-10 MME 2.813E-0 6 9.512E-0 9 ( .884 E-0 9 2.99 9 E-0 9 1.4 95 E-0 9 9. 067 E-10 6.130 E-10 4.44 2 E-10 3. 378 E-10 2.6 61E-10 2.154 E-10 NE 4.17 3E-0 8 1.411E-08 7.24 5E-0 9 4.44 9E-0 9 2.218 E-0 9 1. 34 5 E-0 9 9. 0 95E-10 6.591E-10 5. 011E-10 3. 948 E-10 3.196 E-10 I EME 3.5 0 0 E-08 1.184 E-08 6. 078 E-0 9 3.7 32E-0 9 1.861 E- 0 9 1.128 E-0 9 7.62 9 E-10 5.528 E-10 4.2 04 E-10 3.312 E-10 2.6 81 E-10 E 3.684 E-88 1.314 E-08 6.748 E-0 9 4.14 3E-0 9 2.0 6 6 E-0 9 1.253E-0 9 8.47 0 E-10 6.138 E-10 4.6 67 E-10 3.677 E-10 2.9 77 E-10 I ESE 3.857 E-08 1.30 4 E-08 6. 6 97 E-0 9 4.112E-0 9 2. 0 50 E-0 9 1.24 3E-0 9 8.40 7E-10 6. 0 92E-10 4.6 32 E-10 3.6 5 0 E-10 2. 955 E-10 SE 5 156E-08 1.743E-08 3.951E-09 5.497E-09 2.740E-09 1.662E-09 1.124E-09 8.143E-10 6.191E-10 4.878E-10 3.949E-10 SSE 4.978E-08 1.683E-80 s.643E-09 5.307E-09 2.846E-09 1.605E-09 1.085E-09 7.862E-10 5.978E-10 4.710E-10 3.813E-10

!!RECTION CI5TANCES IN MILES 7RCM $1TE 5.00 7.50 10.00 15.00 20.00 25.00 30.00 35.00 40.00 45.00 50.00 5 1.708E-10 8.370E-11 5.251E-11 2.654*-11 1.607E-11 1.077E-11 7.718E-12 5.796E-12 4.506E-12 3.600E-12 2.938E-12

$$W 1.146 E-10 5.618 E-11 3.225E-11 1.782E-11 1. 0 78 E-11 7.231E-12 5.181E-12 3.8 9 0E-12 !.025 E-12 2.416 E-12 1.972E-12 SW 3.217 E-10 5. 96 5E-11 3.74 3E-11 3.8 92E-11 1.14 5 E-11 7.67 7E-12 5.501E-12 4.13 0 E-12 3.212E-12 2.56 5 E-12 2. 0 94 E-12 WSW 1.027E-10 5.032E-11 3.157E-11 1.596E-11 9.659E 12 6.476E-12 4.640E-12 3.485E-12 2.709E-12 2.164E-12 1.766E-12 W 1.424E-10 6.979E-11 4.379E-11 2.213E-11 1.34 0 E-11 8. 982E-12 6.4 36 E-12 4.8 3 3E-12 3.758E-12 3.0 0 2 E-12 2.4 5 0 E-12 WNW 1.954E-10 9.576E-11 6.009E-11 3.037E-11 1.838E-11 1.232E-11 8.831E-12 6.631E-12 5.?i6E-12 4.119E-12 3.362E-12

'N 3.08 3E-10 1.511E-10 9.48 0 E-11 4.7 92E-11 2. 9 0 0 E-11 1.!44 E-11 1. 3 9 3 E-11 1. 0 4 6 E-11 8. 4 35E-12 6.4 98 E-12 5.3 0 4 E-12 s

e- 3.8 6 7 E-10 1. 8 9 5 E-10 1.16 9 E-10 6. 010 E-11 3. 6 38 E-11 2.4 3 9E-11 1. 7 4a E-11 1. 312 E-11 1. 0 2 0 E-11 8.151 E-12 6. 6 5 3 E-12 a 3.287E-10 1.611E-10 1.011E-10 5.109E-1 3.092E-11 2.073E-11 1.486E-11 1.116E-11 8.674E-12 6.929E-12 5.655E-12 NNE 1.782E-10 8.733E-11 5.460E-11 2.770E-1)1 1.676 E-11 1.124 E-11 8. 054 E-12 6. 048E-12 4.7 02 E-12 3.756 E-12 3. 0 6 6 E-12 NE 2.644E-10 1.296E-10 8.130E-11 4.199E-11 2.487E-11 1.667E-11 1.195E-11 8.972E-12 6.976E-12 5.572E-12 4.548E-12 ENE 2.218 E-10 1.087 E-10 6.819E-11 3.4 47 E-11 2. 086 E-11 1.39 9E-11 1. 0 02E-11 7. 526 E-12 5.8 52 E-12 4.674 E-12 3.815 E-12 E 2.462E-10 1.207E-10 7.571E-11 3.827E-11 2.316E-11 1.553E-11 1.113E-11 8.356E-12 6.497E-12 5.190E-12 4.236E-12 ESE 2.444E-10 1.198E-10 7 5131 11 3.798E-11 2.299E-11 1.541E-11 1.104E-11 8.293E-12 6.448E-12 5.151E-12 4.204E-12 SE 3.266E-10 1.601E-10 f.004E-10 5.377E-11 3.073E-11 2.060E-11 1.476E-11 1.108E-11 8.619E-12 6.885E-12 5.619E-12 SSE 3.154E-10 1.546E-10 9.698E-11 4.902E-11 2.967E-11 1.989E-11 1.425E-11 1.070E-11 8.321E-12 6.647E-12 5.426E-12 useassamaammunassunsemuk RELATIVE DEPOSITION PER UNIT AREA (Mem-2) SY DOWNW1MD SECTORS maanssessmensmesanessmen SEGMENT BOUNDARIES IN MILES DIRECTION .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 FROM SITE 4.863E-09 1.502E-09 5.979E-10 3.267E-10 2.076E-10 8.919E-11 2.766E-11 1.096E-11 5.854E-12 3.623E-1 3.264E-09 1.009E-09 4.013E-10 2.193E-10 1.394E-10 5.987E-11 1.857E-11 7.358E-12 3.929E-12 2.432E-1 3.466E-09 1.071E-09 4.261E-10 2.328E-10 1.480E-10 6.357E-11 1.971E-11 7. 812 F.-12 4.172E-12 2.582E-1 U5W 2.924E-09 9.033E-10 3.594E-10 1.964E-10 1.248E-10 5.362E-11 1.663E-11 6.591E-12 3.519E-12 2.178E-1 W 4.055E-09 1.253E-09 4.985E-10 2.724E-10 1.731E-10 7.437E-11 2.306E-11 9.141E-12 4.881E-12 3.021E-1 WNW -3.564E-09' 3.719E-09 6.841E-10 3.738E-10 2.376E-10 1.021E-10 3.165E-11 1.254E-11 6.698E-12 4.146E-1 MW 8.779E-09 2.712E-09 1.079E-09 5.898E-10 3.748E-10 1.610E-10 4.993E-11 1.979E-11 1.057E-11 6.541E-3 NNW 1.101E-08 3.422E-09 1.354E-09 7.398E-10 4.702E-10 2.020E-10 6.263E-11 2.482E-11 1.325E-11 8.204E-1 M 9.361E-09 2.892E-09 1.151E-09 6.289E-10 3.997E-10 1.717E-10 5.324E-11 2.110E-11 1.127E-11 6.974E-1 UN 5.074E-09 1.568E-09 6.238E-10 3.409E-10 2.167E-10 9. 3 0 7 E -11 2.886t-11 1.144E-11 6.108E-12 3.781E-1 M 7.528E 09 2,326E-09 9.255E-10 5.058E-10 3.214E-10 1.381E-10 4.282E-11 1.697E-11 9.062E-12 5.609E-1 ENE 6.315E-09 1.953E-09 7.764E-10 4.242E-10 2.696E-10 1.158E-19 3.592E-11 1.423E-11 7.602E-12 m.705E-1 3 7.011E-09 2.146E-09 8.619E-10 4.710E-10 2.994E-10 1.286E-10 3.987E-11 1.580E-11 8.440E-12  !.224E-3

[1I 6.959E-t1 2.?50E-09 8.555E-10 4.675E-10 2.971E-10 1.276E-10 3.958E-11 1.569E-11 8.377E-12 5.185E-1 9.301E-09 2.874E-09 1.143E-09 6.248E-10 3.171E-10 1.706E-10 5.290E-11 2.*17E-11 1.120E-11 6.930E-1 f$gE 8.980E-09 2.774!-09 1.104E-09 6.033E-10 3.834E-10_ 1.647E-10 5.197E-11 2.024E-11 1.081E-11 6.691E-1 s

~

VENT AMD SUILDING PAR AMETERS RELEASE MEIGHT (METERS) 58.10 REP. WIND MEIGHT (METERS) 10.0 DIAMETER (METERS) 0.0 BUILDING HEIGHT (METER 5) 58.1 EXIT VELOCITY (METER $1 0.0 BLDG. MIN.CRS.SEC. AREA (50. METERS) 2000.0 MEAT EMISSION RATE (CAL /5EC) 0.0

'AT TME RELE ASE HEIGHT: AT THE MEASURED WIND ME!GMT ( 10.0 METERS):

VENT RELEASE MODE WIND SPf ?' .MF' lRS/SEC1 VENT RELEASE MCDE WIND SPEED (METER +/SEC) WIND SPEED (METERS /SEC)

STABLE CONDITIONS UMSTABLE/ NEUTRAL COND171 ELEVATED LESS TMAM 0.0 ELEVATED LES5 TMAM 0.0 LESS TMAM 0.0

. MIXED 'BETWEEN 0.0 AMD 0.0 MIXED BETWEEN 0.0 AND 0.0 BETWEEN C.0 AMD D.

GROUND LEVEL ABOVE 0.0 GROUND LEVEL ABOVE 0.0 ABOVE 0.0

,..-m- - - - . .w , .r-- r -.- - _ _ ,-%_, +. , - - . - - . ,

m .:

L,; cf

.il,+

32 o ,

TABLE 1.4 7'

PALISADES, 1983 SURVEY

Distance to the nearest residence, garden, beef cow, dairy cow and goat in each~

sector.

SECTOR RESIEENCE GARDEN BEEF COW DAIRY COW GOAT 1N >5 mi >5 mi >5 mi >5 mi >5 mi

NNE -1.0 mi 1.5 m'i >5 mi >5 mi >5 mi NE 1.5 mi .1.5 mi 3.9 mi 3.0 mi >5 mi ENE~ 1.3 mi. 1.6 mi 4 mi 4.0 mi >5 mi E 1.0 mi 2.0 mi 3.5 mi 2.5 mi >5 mi ESE .9 mi .9 mi 3.1 mi >5 mi 3.0 mi

'. SE .9'mi 1.2 mi >5 mi >5 mi >5 mi SSE .75 mi. 1.5 mi - >5 mi >5 mi >5 mi S .5 mi 1.5 mi >5 mi >5 mi >5 mi

.SSW~ .75 mi 1.5 mi >5 mi >5 mi >5 mi 4

Y

.w

$ Y a

a

'MIO583-0010A-RS03 0

g.- , e- 4 -. - ,..-.s - , .,,w-. _ , _ . , . , ,,.s. - v--, . - , , - - - cvw., .. e- - -. -

r i 33

[. .

TABLE 1.5 DOSE FACTORS FOR SUBMERSION IN NOBLE GASES DFBI DFY2 DFS1 DFB2 Kr-85m 1.17(+3)8 1.21(+3) 1.46(+3) 3.86(+3)

Kr-85 1.61(+1) 1.69(+1) 1.34(+3) 3.83(+3)

Kr-87 5.92(+3) 6.05(+3) 9.73(+3) 2.01(+4)

Kr-88 1.47(+4) 1.50(+4) 2.37(+3) 5.72(+3)

Kr-89 1.66(+4) 1.59(+4) 1.01(+4) 1.88(+4)

Xe-131m 9.15(+1) 1.53(+2) 4.76(+2) 2.18(+3)

Xe-133m 2.51(+2) 3.17(+2) 9.94(+2) 2.90(+3)

Xe-133. :2.94(+2) 3.46(+2) 3.06(+2) 2.06(+3)

Xe-135m 3.12(+3) 3.30(+3) 7.11(+2) 1.45(+3)

Xe-135 1.81(+3) 1.88(+3) 1.86(+3) 4.84(+3)

.Xe-137 1.42(+3) 1.48(+3) 1.22(+4) 2.50(+4)

Xe-138 8.83(+3) 9.00(+3) 4.13(+3) 9.25(+3)

Ar-41 8.84(+3) 9.76(+3) 2.69(+3) 5.54(+3)

1. mrem /y per pCi/m a

2. m' rad /y.per pCi/m3

3. 1.17(+3) = 1.17x108

\

MIO583-0010A-RS03 h

YTa - > , , , ,. --

w i

34

t TABLE 1.6 STABLE ELEMENT TRANSFER DATA gt ELEMENT F - MILK F - MILK Ff - MEAT m m (COW) (GOAT)

'H. 1.0E-02 1.7E-01 1.2E-02 C 1.2E-02 1.0E-01 3.1E-02 Na 4.0E-02 4.0E-02 3.0E-02 P 2.5E-02 2.5E-01 4.6E-02 Cr 2.2E-03 2.2E-03 2.4E-03 Mn 2.5E-04 2.5E-04 8.0E-04 Fe 1.2E-03 1.3E-04 4.0E-02

-Co 1.0E-03 1.0E-03 1.3E-02 Ni 6.7E-03 6.7E-03 5.3E-02 1Cu 1.4E-02 1.3E-02 8.0E-03 Zn 3.9E-02 3.9E-02 3.0E-02 Rb 3.0E-02 3.0E-02 3.1E-02 Sr 8.0E-04 1.4E-02 6.0E-04 Y 1.0E-05 1.0E-05 4.6E-03 Zr 5.0E-06 5.0E-06 3.4E-02 Nb .2.5E-03 2.5E-03 2.8E-01 Mo 7.5E-03 7.5E-03 8.0E-03 4

Tc 2.5E-02 2.5E-02 4.0E-01 Ru 1.0E-06 1.0E 4^.0E-01 Rh 1.0E-02 1.0E-02 1.5E-03

- Ag- 5.0E-02 5.0E-02 1.7E-02 Te~ 1.0E 1.0E-03 7.7E-02

-I 6.0E-03 6.0E-02 2.9E-03 Cs' 1.2E-02 3.0E-02 4.0E-03 Ba. 4.0E-04 4.0E-04 3.2E-03 La~ 5.0E-06 5.0E-06 2.0E-04 Ce 1.0E-04 1.0E-04 1.2E-03

-Pr 5.0E-06 5.0E-06 4.7E-03 Nd 5.0E-06 5.0E-06 3.3E-03 W 5.0E-04 5.0E-04 1.3E-03

~

Np 5.0E-06 5.]E-06 2.0E-04 A

MIO583-0010A-RS03 J

b-':

35 I

i t

TABLE 1.7 INHALATION DOSE FACTORS FOR INFANT (MREM PER PCI INHALED)

Page 1 of 3 NUCLIDE' BONE' LIVER T. BODY THYROID KIDNEY LUNG GI-LLI

'H ! 3 - NO DATA 4.62E-07 4.62E-07 4.62E-07 4.62E-07 4.62E-07 4.62E-07 C .14 '1.89E-05 3.79E-06 3.79E-06 3.79E-06 3.79E-06 3.79E-06 3.79E-06 NA 24i 7.54E-06 7.54E-06 7.54E-06 7.34E-06 7.54E-06 7.54E-06 7.54E-06 P. 32 1.45E-03 8.03E-05 5.53E-05 NO DATA NO DATA NO DATA 1.15E-05 iCR 51 NO DATA NO DATA 6.39E-08 4.11E-08 9.45E-09 9.17E-06 2.55E-07

-MN 54 NO DATA ~ 1.81E-05 3.56E-06 NO DATA 3.56E-06 7.14E-04 5.04E-06

-MN 56' NO DATA 1.10E-09, 1.58E-10 NO DATA 7.86E-10 8.95E-06 5.12E-05 0  : FE 55. . 1.41E-05~ 8.39E-06 2.38E-06  : NO DATA NO DATA 6.21E-05 7.82E-07 FE 59- 9.69E-06 1.68E-05 6.77E-06 NO DATA .NO DATA 7.25E-04 1.77E-05

' ~

C0'58 NO DATA 8.71E-07 1.30E-06 NO DATA NO DATA 5.55E-04 7.95E-06

. CO 60- . NO DATA 5.73E-06 8.41E-06 NO DATA NO DATA 3.22E-03 2.28E-05 WI 63'- 2.42E-04 1.46E-05 8.29E-06 NO DATA NO DATA 1.49E-04 1.73E-06 NI 65- 1.71E-09 2.03E-10 8.79E-11 NO DATA NO DATA 5.80E-06 3.58E-05

' CU 64 NO DATA ~ 1.34E-09 5.53E-10 NO DATA 2.84E-09 6.64E-06 1.07E-05

. 2N 65- '1.38E-05: 4.47E-05 2.22E-05 NO DATA 2.32E-05 '4.62E-04 3.67E-05

~

4

~

1 f2N 69~ .3.85E-11' 6'.91E-11 5.13E-12 NO DATA 2.87E-11 1.05E-06 9.44E-06

, JBR 81 .

'NO DATA lNODATA 2.72E-07 NO DATA NO DATA NO DATA LT E-24

.BR 84 NO DATA. :NO DATA 2.86E-07 NO DATA NO DATA NO DATA LT E-24

, .BR 85- NO DATA NO DATA 1.46E-08 NO DATA NO DATA NO DATA LT E-24 lRB'86 NO DATA. 1.36E 6.30E-05 NO DATA - NO DATA NO DATA 2.17E g. .RB 88i ;NO DATA ~ 3.98E-07 2.05E-07 NO DATA NO DATA NO DATA 2.42E-07 m ,..................................................................................................

RB 89 - NO DATA -2.29E-07. 1.47E-07. - NO DATA NO DATA NO DATA 4.07E-08

SR 89 2.84E-04 NO DATA 8.15E-06 NO DATA NO DATA 1.45E-03 4.57E-05

. SR 90 : - 2.92E-02; .NO DATA 1.85E-03 ~NO DATA NO DATA 8.03E-03 9.36E-05

....L........~.....................................................................................

1

--SR 91 6.83E-08 NO DATA ;2.47E-09 NO DATA NO DATA 3.76E-05 5.24E-05 A<s^

~

SR 92 7.50E-09 , NO DATA 2.79E-10 NO DATA NO DATA 1.70E-05 1.00E-04 1 '3 Yl90 l2.35E-06: NO DATA 6.30E-08 NO DATA 'NO DATA 1.92E-04 7.43E-05 4

iY: 91m '- 2.91E-10: NO DATA- ~9.90E-12 NO DATA NO DATA 1.99E-06 1.68F-06 LY 91 '4.20E-045 NO DATA' 1.12E-05. NO DATA- NO DATA 1.75E-03' 5.02E-05

" Y . 92 - .-1.17E-08 _ "NO DATA- 3.29E-10 NO DATA NO DATA 1.75E-05 9.04E-05 1

MIO583-00105-R50 C y

[_ b <

e_

p- -

4 4

36 TABLE 1.7 (CONT'D)

INHALATION DOSE FACTORS FOR INFANT (MREM PER PCI INHALED)

Pase 2 of 3 b- NUCLIDE BONE LIVER T. BODY THYROID' KIDNEY ' LUNG CI-LLI

Y -95 '1.07E-07' NO DATA 2.91E-09 NO DATA NO DATA 5.46E-05 1.19E-04

-ZR 95 18.24E 1.99E-05 1.45E-05 NO DAIA 2.22E-05 1.25E-03 1.55E-05 ER 97 1.07E-07 1.83E-08 8.36E-09 NO DATA 1.85E-08 7.88E-05 1.00E-04 NB 95 1.12E-05 4.59E-06 2.70E-06 NO DATA 3.37E-06 3.42E-04 9.05E-06 MO 99 NO DAIA

~

.1.18E-07 2.31E-08 NO DAIA 1.89E-07 9.63E-05 3.48E-05 TC 99m 9.98E 2.06E-12' 2.66E-11 NO DATA 2.22E-11 '5.79E-07 1.45E-06 TC101 4.65E-14 5.88E-14J 5.80E-13 NO DAIA 6.99E-13 4.17E-07 6.03E-07 RU103 1.44E-06 NO DATA 4.85E-07 NO DATA 3.03E-% 3.94E-04 1.15E-05

. RU105 8.74E-10 NO DATA 2.93E-10 NO DATA 6.42E-10 1.12E-05 3.46E-05 RU106: 6.20E-05 NO DAIA 7.77E-06 NO DATA .7.61E-05 8.26E-03 1.17E-04

-. AC110m '7.13E-06 5.16E-06 3.57E-06 NO DATA 7.80E-06 2.62E 2.36E-05 f"t125m 3.40E 1.42E-06 4.70E-07 1.16E-06 NO DATA 3.19E-04 9.22E-06 TE127m .1.39E-05 -4.93E-06 1.48E-06 3.48E-06 2.68E-05 9.37E-04 1.95E-05 IE127: 1.59E-09 6.81E 3.49E-10 1.32E-09 3.47E-09 7.39E-06 1.74E-05 TE129m' 1.01E-05 4.35E-06 1.59E-06 3.91E-06 2.27E-05 1.a0E-03 4.93E-05

-TE129 5.63E-11. 2.48E-11 1.34E 4.82E-11 1.25E-10 2.14E-06 1.88E-05

. TE131m . 7.62E-08 3.93E-08 2.59E-08 6.38E-08 1.89E-07 1.42E-04 8.51E-05

. TE131 1.24E 5.87E-12 3.57E-12 1.13E-11 2.85E-11 1.47E-06 5.87E-06

..................~................................................................................

TE132 2.66E 1.69E-07 1.26E-07 1.99E-07 7.39E-07 2.43E-04 3.15E-05 I 130 4.54E-06 9.91E-06 3.98E-06 1.119-03 1.09E-05 NO DAIA 1.42E-06 i~

cI 131- 2.71E-05 -3.17E-05 1.40E-05 1- sE-02 3.70E-05 NO DATA 7.56E-07

.................................................................................................. e

-I 132' 1.21E-06 2.53E-06 8.99E-07 1.21E-04 2.82E-06 NO DATA 1.36E-06

~ I 133 -- 9.46E-06 1.37E-05 4. 00E-06 2.54E-03 1.60E-05 NO DAIA 1.54E-06 1 11 134 6.58E-07 1.34E-06 4.75E-07 3.18E-05 1.49E-06 NO D 9.21E-07 il?ft ~ .i...............................................................................AIA .................

I 135 2.76E 5.43E-06 ~ .1.98E 4.97E-04 6.05E-06 NO DATA 1.31E-06

, CS134 . 2.81E-04: 5.02E-04 -5.32E-05 NO DATA 1.36E-04 5.69E-05 9.53E-07

~ CS136 - 3.45E-05 9.61E-05 3.78E-05 NO DATA 4.03E-05 8.40E-06 1.02E-06 m ..................................................................................................

CS137- -3.92E-04. 4.37E-04 3.25E-05 NO DATA 1.23E-04 5.09E-05 9.53E-07 CS138 3.61E-07 5.58E 2.84E-07 NO DATA 2.93E-07 4.67E-08 6.26E-07 BA139 1.06E-09 7.03E-13 3.07E-11 NO DATA 4.23E-13 4.25E-06 3.64E-05 1

. l l

MIO583-00105-RS03 l

c _ - , .- -. --. .- . . . . - - . - . - - . - . - _ _ . -

1 i

1 I

4 37

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I l

)

IABLE 1.7 (CONT'D) l l

i INHALATION DOS 2 FACTORS FOR INFANT l (MREM PER PCI INHALED)

Page 3 of 3

'.NUCLIDE BONE' LIVER T. BODY THYROID -KIDNEY LUNG GI-LLI BA140 4.00E-05 4.00F-08 2.07E-06 NO DAIA 9.59E-09 1.14E-03 2.74E-05 RA141 1.12E-10 7.70E-14 3.55E-12 NO DATA 4.64E-14 2.12E-06 3.39E-06 1BA142 2.84E-11 2.36E-14 1.40E-12 N3 DAIA 1.36E-14 1.11E-06 4.95E-07 LA140 3.61E-07 1.43E-07 3.68E-08 NO DATA NO DATA 1.20E-04 6.06E-05

, LA142 7. 36E-10 2.69E-10 6.46E-11 NO DAIA NO DATA 5.87E-06 4.25E-05 CE141 1.98E-05 1.19E-05 1.42E-06 NO DAIA 3.75E-06 3.69E-04 1.54E-05 CE143 2.09E-07 1.38E-07 1.58E-08 NO DATA 4.03E-08 8.30E-05 3.55E-05 CE144 2.28E-03 8.65E-04 1.26E-04 NO DAIA 3.84E-04 7.03E-03 1.06E-04 PR143 1.00E-05 3.74E-06 4.99E-07 NO DAIA 1.41E-06 3.09E-04 2.66E-05 PR144 3.42E-11 1.32E-11 1.72E-12 NO DATA 4.80E-12 1.15E-06 3.06E-06

' MD147 - 5.67E-06 5.81E-06 3.57E-07 NO DAIA 2.25E-06 2.30E-04 2.23E-05 W 187 9.26E-09' 6.44E-09 2.23E-09 NO DATA NO DAIA 2.83E-05 2.54E-05 NP239 2.65E-07 2.37E-08 1.34E-08 NO DAIA 4.73E-08 4.25E-05 1.78E-05

? *

(- g a

M10583-00105-RS03 m

i Q:i ;a I i

E 38 TABLE 1.7 (CON'T)

INHA1.ATION DOSE FACTORS FOR CHILD (MREM PER PCI INHALED)

Page 1 of 3

.NUCLIDE BONE -LIVER T. BODY THYROID KIDNEY UfNG GI-LLI H~3' NO DAIA: ~3.04E-07 3.04E-07 3.04E-07 3.04E-07 3.04E-07 3.04E-07 C :14- 9.70E-06 1.82E-06 1.82E-06 1.82E-06 1.82E-06 1.82E-06 1.82E-06 NA 24 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.352-06 P 7,04E-04 3.09E-05 2.67E-05 NO DATA NO DATA NO DAIA 1.14E-05 CR 51 NO DAIA- NO DATA 4.17E-08 2.31E-08 6.57E-09 4.59E-06 2.93E-07 MN 54: ^ NO DATA 1.16E-05 2.57E-06 NO DAIA 2.71E-06 4.26E-04 6.19E-06 NN $6 NO DAIA 4.48E-10 8.43E-11 NO DATA 4.52E-10 3.55E-06 3.33E-05

- FE 55 1.28E-05 6.80E-06 2.10E-06 NO DAIA NO DAIA 3.00E-05 7.75E-07 FE 59 5.59E-06 9.04E-06 4.51E-06 NO DAIA NO DATA 3.43E-04 1.91E-05 C0 $8 NO DATA 4.79E-07 8.55E-Os NO DATA NO DAIA 2.99E-04 9.29E-06 00 60 NO DAIA 3.55E-06 6.12E-06 NO DAIA NO DAIA 1.91E-03 2.60E-05 NI 63 2.22E-04 1.25E-05 7.56E-06 NO DAIA NO DATA 7.43E-05 1.71E-06 NI 65 8.08E-10 7.99E-11 4.44E-11 NO DATA NO DATA 2.21E-06 2.27E-05 CU 64 NO DAIA 5.39E-10 2. 90E-10 NO DATA 1.63E-09 2.59E-06 9.92E-06 ZN 65 1.15E-05 3.06E-05 1.90E-05 NO DATA 1.93E-05 2.69E-04 4.41E-06 ZN 69 1.81E-11 2.61E-11 2.42E-12 NO DATA 1.58E-11 3.84E-07 2.75E-06 BR 83' NO DATA' NO DATA 1.28E-07 NO DATA NO DATA NO DATA LT E-24 BR 84 .NO DATA NO DATA ~1.48E-07 NO DATA NO DATA NO DATA LT E-24 BR 85 NO DATA NO DAIA 6.84E-09 NO DATA NO DATA NO DATA LT E-24 RB 86- NO DAIA 5.36E-05 3.09E-05 NO DAIA NO DATA NO DAIA 2.16E-06 RB 88 'NO DATA 1.52E-07 9.90E-08 NO DATA NO DATA NO DATA 4.66E-09

-RB 89 NO DATA 9.33E-08 7.85E-08 NO DAIA NO DATA NO DAIA 5.11E-10 SR 89' 1.62E-04 NO DAIA: 4.66E-06 NO DAIA NO DAIA 5.83E-04 4.520-05 SR 90 2.73E-02 NO DATA 1.74E-03 NO DATA NO DAIA 3.99E-03 9.28E-05

- SR 91 3.28E-08 NO DATA' 1.24E-09 NO DATA NO DAIA 1.44E-05 4.70E-05 SR 92 3.54E NO DAIA 1.42E-10 NO DAIA NO DAIA 6.49E-06 6.55E-05 Y 90 1.11E-06 NO DAIA 2.99E-08 NO DATA NO DATA 7.07E-05 7.24E-05 Y 91m 1.37E-10 NO DATA 4.98E-12 NO DATA NO DATA 7.60E-07 4.642-07 Y 91 - 2.47E-04 NO DAIA 6.59E-05 NO DAIA NO DAIA 7.10E-04 4.97E-05 Y ~ 92 5.5C:-09 NO DATA 1.57E-10 NO DATA NO DATA 6.46E-06 6.46E-05 MIO583-0010B-R503

b b L

+]

g

, 39

11' IABLE 1.7 (CONT'D)

INHALATION DOSE FACTORS FOR CHIID

-(MREM PER PCI INHALED)

Page 2 of 3 NUCLIDE BONE LIVER 'T. BODY THYROID KIDNEY LUNG GI-LLI l Y - 93 ' - 5.04E-08 ~NO DATA 1.38E-09 NO DATA NO DATA 2.01E-05 1.05E-04

- 23t 95 : T5.13E-05 1.13E-05 1.00E-05 NO DAIA 1.61E-05 6.03E-04 1.65E-05 2R 97 s 5.07E-08 7.34E-09 4.32E-09 NO DAIA 1.05E-OS 3.06E-05 9.49E-05 MB 95 '6.35E-06 2.48E-06 1.77E-06 NO DAIA 2.33E-06 1.66E-04 1.00E-05

- MD 99 - NO DATA 4.66E-08 - 1.15E-08 NO DAIA 1.06E-07 3.66E-05 3.42E-05

TC 99m - L4.81E-13' 9.41E-13 1.56E NO DAIA 1.37E-11 2.57E-07 1.30E-06

. TC101' 2.19E-14 2.30E-14 2 ' t i-13 NO DAIA 3.92E-13 1.58E-07 4.41E-09 RU103: - 7.55E-07 NO DATA 2.90E-07 NO EATA 1.90E-06 1.79E-04 1.21E-05

.RU105- '4.13E-10 NO DAIA 1.50E-10 NO DAIA- 3.63E-10 4.30E-06 2.69E-05

< s ....................................................................................................

• RU106: -3.68E-05 NO DATA 4.57E-06 NO DATA 4.97E-05 3.87E-03 1.16E-04

. AC110m 4.56E 3.08E-06 2.47E-06 NO DATA 5.74E-06. 1.48E-03 2.71E-05 EE125m - 1.82E-06 6.29E-07 2.47E-07 5.20E-07 NO DATA. 1.29E-04 9.13E-06

- nTE127m- 6.72E-06 2.31E-06 8.16E 1.64E-06 1.72E-05 4.00E-04 1.93E-05 lDE127 j ' 7.49E-10 2.57E-10 '1.65E-10 5.30E-10 1.91E-09 2.71E-06 1.52E-05 TE129m '5.19E-06 1.85E-06 8.22E-07 1.71E-06 1.36E-05 4.76E-04 4.91E-05

~e- ..................................................................................................

TE129-  : 2.64E-11' 9.45E-12 6.44E-12 1.93E-11 6.94E-11 7.93E-07 6.89E-06

, TE131m 3.63E-08 1.60E-08 - 1.37E-08 2. 64E-08 1.08E-07 5.56E-05 8.32E-05

.T: :TE131. 2.28E-12 1.78E-12 15.59E-12 1.59E-11 5.5SE-07 3.60E-07

,o

,...........'5.87E-12' TE132: ~1.30E-07 7.36E-08 7.12E-08 8.53E-08 4.792-07 1.02E-04 3.72E-05' iI 130 .- 2.21E 4.43E-06 2.28E-06 4.99E-04 6.61E-06 NO DATA 1.38E-06

.I 131  : 1.30E 1.30E-05 7.37E-06 4.39E-03 2.13E-05 NO DATA 7.68E-07 I 132 5.72E-07. 1.10E ' 5.07E-07 5.23E-05 1.69E-06 NO DATA 8.65E-07

~

I 133' 4'.48E-06 T.49E-06 2.08E-06 1.04E-03 9.13E-06 NO DATA 1.48E-06 f ;I 134 3.17E-07' '5.84E-07 2.69E-07 1.37E-05 '8.92E-07 NO DAIA- 2.58E-07 ss II 1352 1.33E-06 2 2.36E-06 1.12E-06 2.14E-04 3.62E-06 NO DATA 1.20E-06

' CS134 : =1.76E-04' 2.74E-04 6.07E-05 NO DAIA 8.93E-05 3.27E-05 1.04E-06 (CS136L ~ 1.76E :4.62E-05 -3.14E-05 NO DATA 2.58E-05 3.93E 1.13E-06 n - CS137, 7.45E-04, 2.23E-04 3.47E-05 NO DATA 7.63E 2.81E-05 9.78E-07

. CS138 i U1'.71E-07 2.27E-07 1.50E-07 NO DATA' 1.68E-07 1.84E-08 7.29E-08

'RA139 4.98E 2.66E-13: 2 1.45E-11. NO DAIA 2.33E-13 1.56E-06 '1.56E-05 o s' ..................................................................................................

n

'M10583-00103-RS03

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lL5' e . TABLE 1.7 (CONT'D)

Wu ;

% INHALxIION DOSE FACTCPS FOR CHIII (MREM PER PCI INHALEW Page 3 of 3 NUCLIDE- BONE LIVER T. BODY THYROID KIDNEY- LUNG CI-LLI h' BA140 ' .2.00E-05 1.75E-08 1.17E-06 NO DATA 5.71E-09 4.71E-04 2.75E-05 BA141. '5.29E-11 2.95E-14 1.72E-12 - NO DATA 2.56E-14 7.89E-07 7.44E-08 BA142~ :1.35E 9.73E-15 7.54E-13 NO DATA 7.87E-15 4.44E-07 7.41E-10 LA140 - 1.74E 6.08E-08 '2.04E-08 NO DATA NO DATA 4.94E 6.10E-05 1 LA142 3.50E-10 1.11E 3.49E-11 .NO DATA NO DATA 2.35E-06 2.05E 05 C'

_ }CE141.. 1.06E-05 5.28E-06

7.83E-07 NO DATA 2.31E-06 1.47E 1.53E-05

' CE1431 -9.89E-08 .5.37E-08 7.77E-09 NO DATA 2.26E-08 3.12E-05 3.44E-05

CE144 '1.83E-03 5.72E-04 9.77E-05 NO DATA 3.17E-04 3.23E-03 1.05E-04

.PR143 4.99E _1.5CE 2.47E-07 NO DATA 8.11E-07 1.17E-04 2.63E-05

. PR144 - 1.61E 4.99E-12 8.10E-13 NO DATA 2.64E-12 - 4.23E-07 5.32E-08 KD147 2.92E-06 2.36E-06 1.84E-07 NO h,. 1.30E-06 8.87E-05 2.22E-05

. 4 187 '

4.41E-09 ~2.61E-09 1.17E-09 NO DATA NO DATA 1.11E-05 . 2.46E-05

+

,NP239 .1.26E-07 9.04E-09 .6.35E-09 NO DATA 2.63E-08 1.57E-05 1.73E-0; s

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-t-4 y-+------t--+as----*w-wm-r-a-rw----e,-et-ww-we-&w++--

1 i i

1 41 l

i i

TABLE 1.7 (CON'T)

INHALATION DOSE FACTORS FOR ADULTS (MREM PER PCI INHALED)

Page 1 of 3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG CI-LLI H' 3 NO DATA. 1.58E 1.58E-07 1.58E-07 1.58E-07 1.58E-07 1.58E-07 C 14 2.27E 4.26E-07 4.26E-07 4.26E-07 4.26E-07 4.26E-07 4.26E-07 NA 24- 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 P 32- 1.65E-04 9.64E-06 6.26E-06 NO DATA NO DATA NO DATA 1.08E-05 CR 51- NO DATA NO DATA 1.25E-08 7.44E-09 2.85E-09 1.80E-06 4.15E-07

'MN 54 NO DATA 4.95E-06 7.87E-07 NO DATA 1.23E-06 1.75E-04 9.67E-06 NN 56 NO DATA 1.55E-10 2.29E-11 NO DATA 1.63E-10 1.18E-C6 2.53E-06 FE 55 3.07E-06 2.12E-06 4.93E-07 NO DATA NO DATA 9.01E-06 7.54E-07 FE 59 1.47E-06 3.47E-06 1.32E-06 NO DATA. NO DATA 1.27E-04 2.35E-05 CO 58 NO DATA 1.98E-07 2.59E-07 NO DATA NO DATA 1.16E-04 1.33E-05

- CO 60 - NO DATA 1.44E-06 1.85E-06 NO DATA NO DATA 7.46E-04 3.56E-05 NI 63 5.40E-05 3.93E-06 1.81E-06 NO DATA NO DATA 2.23E-05 1.67E-06 NI 65 1.92E-10 2.62E-11 1.14E-11 NO DATA NO DATA 7.00E-07 1.54E-06 CU 64 NO DATA- 1.83E-10 7.69E-11 NO DATA 5.78E-10 8.48E-07 6.12E-06

'ZN 65 4.05E-00 1.29E-05 5.82E-06 NO DATA 8.62E-06 1.08E-04 6.68E-06 Z3 69 4.23E-12 8.14E-12 5.65E-13 NO DATA 5.27E-12 1.15E-07 2.04E-09

.BR 83 NO DATA NO DATA 3.01E-08' NO DATA NO DATA NO DATA 2.90E-08 BR 84 ~ NO DATA NO DATA 3.91E NO DATA NO DATA NO DATA 2.05E-13 1 BR 85 NO DATA - NO DATA 1.60E-09 NO DATA NO DATA NO DATA LT E-24 RB 86~ -NO DATA 1.69E-05 7.37E-06 NO DATA NO DATA NO DATA 2.08E-06

~RB 88 NO DATA . 4.84E-08 2.41E-08 NO DAIA NO DATA NO DATA 4.18E-19 RB 89 NO DATA 3.20E-08 2.12E-08 NO DATA NO DATA NO DATA 1.16E-21 SR 89 3.80E-05 NO DATA 1.09E-06 NO DATA NO DATA 1.75E-04 4.37E-05

'SR 90' 1.24E-02 NO DATA 7.62E-04 NO DATA NO DATA 1.20E-03 9.02E-05 SR 91 7.74E-09 NO DATA 3.13E-10 NO DATA NO DATA 4.56E-06 2.39E-05 SR 92 8.43E-10 NO DATA 3.64E-11 NO DATA NO DATA 2.06E-06 5.38E-06

^Y 90' 2.61E-07 NO DATA 7.01E-09 NO DATA NO DATA 2.12E-05 6.32E-05 O ' Y ; 91m . 3.26E-11 NO DATA 1.27E-12 NO DATA - NO DATA 2.40E-07 1.66E-10 Y. 91 < 5.78E-05 . NO DATA 1.55E-06 NO DATA NO DATA 2.13E-04 4.81E-05 Y. 92 1.29E-09 NO DATA 3.77E NO DATA NO DATA 1.96E-06 9.19E-06 4

MIO583-00108-RS03 s

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42 TABLE 1.7 (CONT'D)

INHALATION DOSE FACTORS FOR ADULTS (MREM PER PCI INHALED)

Page 2 of 3 NUCLIDE BONE LIVER T.BCDY THYROID KIDNEY LUNG GI-LLI Y 93~ 1.18E-OS NO DATA 3.26E-10 NO DATA NO DATA 6.06E-06 5.27E-05 ZR 95' 1.34E-05 4. 30E-06 2.91E-06 NO DAIA 6.77E-06 2.21E-04 1.88E-05 ZR 97 1.21E-08 2.45E-09 1.13E-09 NO DATA 3.71E-09 9.84E-06 6.54E-05 NB 95 1.76E-06 9.77E-07 5.26E-07 NO DATA 9.67E-07 6.31E-05 1.30E-05 MO 99 NO DATA 1.51E-06 2.87E-09 NO DATA -3.64E-08 1.14E-05 3.10E-05 TC 99m 1.29E-13 3.64E-13 4.63E-12 NO DATA 5.52E-12 9.55E-08 5.20E-07

- TC101-- 5.22E-15 7.52E-15 7.38E-14 NO DATA 1.35E-13 4.99E-08 1.36E-21 RU103 1.91E-07 NO DATA 8.23E-08 NO DAIA 7.29E-07 6.31E-05 1.38E-05 RU105 9.88E-11 NO DATA 3.89E-11 NO DAIA 1.27E-10 1.37E-06 6.02E-06 RU106' 8.64E-06 NO DATA 1.09E-06 NO DATA 1.67E-05 1.17E-03 1.14E-04 AG110m 1.35E-06 1.25E-06 7.43E-07 NO DATA 2.46E-06 5.79E-04 3.78E-05 TE125m 4.27E-07 1.98E-07 5.84E-08 1.31E-07 1.55E-06 3.92E-05 8.83E-06 TE127m 1.58E-06 7.21E-07 1.96E-07 4.11E-07 5.72E-06 1.20E-04 1.87E 05 IE127 1.75E-10 8.03E-11 3.87E-11 1.32E-10 6.37E-10 8.14E-07 7.17E-06 TE129m 1.22E-06 5.84E-07 1.98E-07 4.30E-07 4.57E-06 1.45E-04 4.79E-05 TE129 6.22E-12 2.99E-12 1.55E-li 4.81E-12 2.34E-11 2.42E-07 1.96E-08 TE131m 8.74E-09 5.45E-09 3.63E-09 6.88E-09 3.86E-08 1.82E-05 6.95E-05 TE131- 1.39E-12 7.44E-13 4.49E-13 1.17E-12 5.46E-12 1.74E-07 2.30E-09

-TE132 3.25E-08 2.69E-08 2.02E-08 2.37E-08 1.82E-07 3.60E-05 6.37E-05 I 130 '5.72r-07 1.68E-06 6.60E-07 1.42E-04 2.61E-06 NO DAIA 9.61E-07 I 131 3.15e 66 4.47E-06 2.56E-06 1.49E-03 7.66E-06 NO DAIA 7.85E-07 I 132 1.45E-07 4.07E-07 1.45E-07 1.43E-05 6.48E-07 NO DATA 5.08E-08 1 133 1.08E-06 1.85E-06 5.65E-07 2.69E-04 3.23E-06 NO DAIA 1.11E-06 I 1 34 8.05E-08 2.16E-07 7.69E-08 3.73E-06 3.44E-07 NO DAIA 1.26E-10 I 135 3.35E-07 8.73E-07 3.21E-07 5.60E-05 1.39E-06 NO DATA 6.56E-07 CS134 4.66E-05 1.06E-04 9.10E-05 NO DAIA 3.59E-05 1.22E-05 1.30E-06 CS136 4.88E-06 1.83E-05 1.38E-05 NO DAIA 1.07E-05 1.50E-06 1.46E-06 CS137 5.98E 7.76E-05 5.35E-05 NO DATA 2.78E-05 9.40E-06 1.0$E-06

_ CS138 4.14E-08 7.76E-08 4.05E-08 NO DATA 6.00E-08 6.07E-09 2.33E-13 BA139 1.17E-10 8.32E-14 3.42E-12 NO DATA 7.78E-14 4.70E-07 1.12E-07 MIO583-0010B-RS03

43 TABLE 1.7 (CONT'D)

INHALATION DOSE FACTORS FOR ADULTS (MREM PER PCI INHALED)

Page 3 of 3 NUCLIDE' BONE LIVER T. BODY THYROID KIDNEY LUNC GI-LLI

, 'EA140 4.88E-06 6.13E-09 3.21E-07 NO DAIA 2.09E-09 1.59E-04 2.73E-05 BA141 .1.25E-11 9.41E-15 4.20E-13 NO DATA 8.75E-15 2.42E-07 1.45E-17

, RA142 3.29E-12 3.38E-15 2.07E-13 NO DATA 2.86E-15 1.49E-07 1.96E-26

.LA140 4. 30E-08 . 2.17E-08 5.73E-09 NO DATA NO DAIA 1.70E-05 5.73E-05

. LA142 8.54E-11 3.88E 9.65E-12 NO DAIA NO DAIA 7.91E-07 2.64E-07 CE141 2.49E 1.69E-06 1.91E-07 NO D/.IA 7.83E-07 4.52E-05 1.50E-05 CE143 2.33E-08 1.72E-08 1.91E-09 NO DAIA 7.60E-09 9.97E-06 2.83E-05

. CE144 -4.29E-04 1.7?E-04 2.30E-05 NO DATA 1.06E-04 9.72E-04 1.02E-04 Tk143 1.17E-06 4.69E-07 5.80E-08 NO DAIA 2.70E-07 3.51E-05 2.50E-05 PR144 3.76E-12 1.56E-12 1.91E-13 NO DAIA 8.81E-13 1.27E-07 2.69E-18 ND147 6.59E-07 7.62E-07 4.56E-08 NO DAIA 4.45E-07 2.76E-05 2.16E-05 W 187..

1.06E-09 8.85E-10 3.10E-10 NO DAIA NO DAIA 3.63E-06 1.94E-05 NP239 2.87E-08 2.82E-09 1.55E-09 NO DAIA 8.75E-09 4.70E-06 1.49E-05 i

i t- .

r Mid583-00108-R$03 >

. , s.

7:4 W

44 TABLE 1.8 EXTERNAL DOSE FACTORS FOR STANDING ON CONTAMINATED GROUND (mrem /hr per pei/m 2)

ELEMENT TOTAL BODY SKIN H-3 0.0 0.0 C-14 0.0 0.0 Na-24 2.50E-08 2.90E-08 P-32 0.0 0.0 Cr-51 2.20E-10 2.60E-10 Mn-54 5.80E-09 6.80E-09 Mn-56 1.10E-08 1.30E-08 Fe-55 0.0 0.0 Fe-59 8.00E-09 9.40E-09 Co-58 7.00E-09 8.20E-09 Co-60 1.70E-08 2.00E-08 Ni-63 0.0 0.0 Ni-65 3.70E-09 4.30E-09 Cu-64 1.50E-09 1.70E-09 Zn-65 4.00E-09 4.60E-09 Zn-69 0.0 0.0 Br-83 6.40E-11 9.30E-11

, Br-84 1.20E-08 1.40E-08 Br-85 0.0 0.0 Rb-86 6.30E-10 7.20E-10 Rb-88 3.50E-09 4.00E-09 Rb-89 1.50E-08 1.80E-08 Sr-89 5.60E-13 6.50E-13 Sr-91 7.10E-09 8.30E-09 i -Sr-92 9.00E-09 1.00E-08 Y-90 2.20E-12 2.60E-12 L Y-91m 3.80E-09 4.40E-09 Y-91 2.40E-11 2.70E-11 Y-92 1.60E-09 1.90E-09 Y-93 5.70E-10 7.80E-10 Zr-95 5.00E-09 5.80E-09 Zr-97 5.50E-09 6.40E-09 Nb-95 S.10E-09 6.00E-09

.Mo-99 1.90E-09 2,20E-09 Tc-99m 9.60E-10 1.10E-09 Tc-101 2.70E-09 3.00E-09 Ru-103 3.60E 09 4.20E-09 Ru-105 4.50E-09 5.10E-09 Ru-106 1.50E-09 1.80E-09 Ag-110m 1.80E-08 2.10E-08 Te-125m 3.50E-11 4.80E-11 Te-127m 1.10E-12 1.30E-12 Te-127 1.00E-11 1.10E-11 k ;h10583-0010A-RS03 i

I

/

45 TABLE 1.8 (CON'T)

ELEMENT- TOTAL BODY SKIN

'Te-129m 7.70E-10 9.00E-10

.Te-129 7.10E-10 8.40E-10 Te-131m 8.40E-09 9.90E-09 Te-131 2.20E-09 2.60E-06 Te-132 1.70E-09 2.00E-09 I-130 1.40E-08 1.70E-08 I-131' 2.80E-09 3.40E-09 I-132 1.70E-08 2.00E-08 I-133 3.70E-09 4.50E-09 I-134 1.60E-08 1.90E-08 I-135 1.20E-08 1.40E-08 Cs-134 1.20E-08 1.40E-08 Cs-136 1.50E-08 1.70E-08

-Cs-137 4.20E-09 4.90E-09 Cs-138 2.10E-08 2.40E-08 Ba-139 2.40E-09 2.70E-09 Ba-140 2.10E-09 2.40E-09 Ba-141 4.30E-09 4.90E-09 Ba-142' 7.90E-09 9.00E-09 La-140 1.50E-08 1.70E-08 La-142 1.50E-08 '1.80E-08 Ce-141 5.50E-10 6.20E-10 Ce-143 2.20E-09 2.50E-09 Ce-144 3.20E-10 3.70E-10 Pr-143- 0.0 0.0 Pr-144 .2.00E-10 2.30E-10 Nd-147 1.00E 1.20E-09 W-187 . 3.10E-09 3.60E-09 Np-239 9.50E-10 1.10E-09 q

r w

I

, MIO583-0010A-RS03

I 46 l

1 TABLE 1.9 '

PALISADES PLANT 1984 GASEOUS DESIGN OBJECTIVE ANNUAL QUANTITIES Pathway- Dc/Cc Design Objective Site - Dose Factor Annual Quantity Nuclide Ase* Organ mrem /Ci (Ci)

H-3 V-1-C Total Body 1.23E-04 4.07E+04 C-14 V-1-C Bone 1.07E-01 1.40E+02 Ar-41 P-1-y Total Body 1.96E-04 2.55E+04 Cr-51 V-1-A ,

GI Tract 2.40E-03 6.25E+03 Mn-54 V-1-T GI Tract 2.48E-01 6.05E+01 Fe-55 V-1-C Bone 2.11E-01 7.11E+01 Mn-56 V-1-C. GI Iract 2.51E-07 5.98E+07 Co-57 V-1-T GI Tract 8.69E-02 1.73E+02 Co-58 V-1-C Total Body 5.16E-02 9.69E+01 Fe-59. V-1-T GI Tract 2.46E-01 6.10E+01 TCo-60 V-1-C Total Body 3.07E-01 1.63E+01 Ni-65 V-1-C GI Tract 1.12E-07 1.34E+08

-Zn-65 V-1-C Total Body 4.67E-01 1.07E+01 Kr-83m P-1-$- Skin 4.75E-07 3.16E+07 Kr-85 P-1-p Skin 4.29E-05 3.50E+05 Kr-85m P-1-y Total Body 2.60E-05 1.92E+05

'Kr-87 P-1-p Skin 4.60E-04 3.26E+04 Kr-88 P-1-y Total Body 3.26E-04 1.53E+04 Rb-88 V-1-C Total Body 2.25E-32 2.22E+32 Kr-89 P-1-y Total Body 3.68E-04 1.36E+04

-Sr V-1-C Bone 9.28E+00 1.62E+00 Sr-90 V-1-C Bone 3.82E+02 3.93E-02 Nb-95 V-1-T GI Tract 1.10E-01 1.36E+02 Zr-95 V-1-T GI Tract 3.25E-01 4.62E+01 Mo-99 M-4-I Kidney 4.54E-03 3.30E+03 Tc-99m V-1-C GI Tract 4.98E-07 3.01E+07 Ru-103- V-1-T GI Tract 1.41E-01 1.06E+02 Sb-124 V-1-T GI Tract 7.91E-01 1.90E+01 Sb-125 V-1-T iGI Tract 4.46E-01 3.36E+01 Te-127 V-1-T GI Tract 4.45E-05 3.37E+05 I-131 M-4-I Thyroid 7.55E+00 1.99E+00 Xe-131m P-1-p Skin 1.89E-05 7.94E+05 I-132 V-1-C Thyroid 3.62E-07 4.14E+07 I-133 M-4-I Thyroid 7.00E-02 2.14E+02 Xe-133 P-1-y Total Body 6.52E-06 7.67E+05 Xe-133m P-1-p Skin 3.96E-05 3.79E+05 Cs-134 V-1-C Liver 7.07E+00 2.12E+00 I-134 V-1-C Thyroid 2.79E-13 5.38E+13 1-135 V-1-C Thyroid 4.71E-04 3.18E+04 Xe-135 P-1-y Total Body 4.02E-05 1.24E+05 Xe-135m P-1-y Total Body 6.92E-05 7.23E+04 IC1282-0012B-RS01 .

47 Pathway- Design Objective Site - Dose Factor Annual Quantity Nuclide Age

• Organ mrem /Ci (Ci)

Cs-136 M-3-I Total Body 7.50E-02 6.67E+01 Cs-137 V-1-C Bone 6.91E+00 2.17E+00 Xe-137 P-1-p Skin 4.24E-04 3.54E+04 Xe-138 P-1-y Total Body 1.96E-04 2.55E+04 Ba-140 V-1-C Bone 4.68E-02 3.21E+02 La-140 V-1-A GI Tract 6.95E-03 2.16E+03 Ce-141 V-1-T GI Tract 1.29E-01 1.16E+02 Ce-144 V-1-T GI Tract 3.44E+00 4.36E+00 Codes are as follew: ,

Pathways V - Green leafy vegetable ingestion P - Plume submersion M - Milk ingestion

- Site locations 1 - Residence with garden, 0.5 mi, South sector

'3 - Milk (goat), 3.0 miles, ESE sector 4 - Milk (cow), 2.5 miles, East sector -

Aae Groups A - Adult T - Teen C - Child I - Infant S - All ages, beta skin exposure y.- All ages, gamma total body exposure 4

-IC1282-0012B-RS01 1

48 II. LIQUID' EFFLUENTS-A. CONCENTRATION

1. 'RETS Requirement Specification 3.24.3.1 of the Radiological Effluent Technical Speci-fications (RETS) requires that the concentration of radioactive ma-terial. released at any time from the site to unrestricted areas shall be limited to the Maximum Permissible Concentration (MPC) specified in-10 CFR 20, Appendix B, Table II, Column 2 for nuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2x10~4 pCi/ml total activity. To ensure compliance, the following approach will be used for each release.

2. Prerelease Analysis Most tanks will be recirculated through two volume changes prior to sampling for release to the environment to ensure that a repreeer tative sample is obtained. The appropriate recirculation time for 7; those tanks too large to provide two volume changes will be the time that the suspended particulate concentration reaches steady state.

Either a one-time test, or prior sampling data, may be used to deter-mine appropriate recirculation time.

V Prior I.o_ release, a grab sample will be analyzed for each release, and the concentration of each radionuclide determined.

n C= Cg (2.1) i=1 MIO583-0010A-RS03

49 where:

C = Total concentration in the liquid effluent at the release point, pCi/ml.

C f

= Concentration of a single radionuclide i, pCi/ml.

3. Total MPC-Fraction The Total MPC-Fraction (R ) for each release point will be calculated by the relationship defined by Note 1 of Appendix B,10 CFR 20:

R= g (2.2) i where:

C = Effluent concentration of radionuclide i, pC/ml.

-HPC f

= The MPC of radionuclide i, as specified in Section 2.1.1, pCi/ml.

R = The Total MPC-Fraction for the release point.

The sum of the ratios at the discharge to the lake must oc $1 due to p ,the releases from any or all concurrent releases. The following relationship will assure this criterion is met:

f 3(R g-1) + f (R 2 2~1) + f 3 3

~

(

where:

f ,f 'f = The effluent flow rate (gallons / minute) for the respec-3 2 3 tive releases, determined by plant personnel.

R ,R ,R , = The Total MPC-Fractions for the respective releases as 3 2 3 determined by Equation 2.2.

F = Minimum required dilution flow rate. Normally, a conservatively high dilution flow rate is used, that

'is, flow rate used = (bg )(F) where bg is a conserva-tive factor greater than 1.0.

MIO583-0010A-RS03 i = __

TL 50

~

B. INSTRUMENT SETPOINTS

.1. Setpoint Determination

.The setpoint for each liquid effluent monitor will be established using plant instructions. Concentration, flow rate, dilution,

. principal gamma emitter, geometry and detector efficiency are com-bined to give an equivalent setpoint in counts per minute (cpm). The physical and technical description, location and identification number for each liquid effluent radiation detector is contained in Figure 2-2.

.The- respective alarm / trip setpoints at each release point will be sct such that the sum of the ratios at each point, as calculated by Equation 2.2, will not be exceeded. The value of R is directly related to the total concentration calculated by Equation 2.1. An increase in the concentration would indicate an increase in the value of R. A large increase would cause the limits specified in Section 2.1.1 to be exceeded. The minimum alarm / trip setpoint value is equal to the release concentration, but for ease of operation it may be desired that the setpoint (S) be set above the effluent concentration (C) by the same factor.(b) utilized in setting dilution flow. That is:

S=bxC (2.4)

Liquid effluent flow paths and release points are indicated in Figure 2.1.

2. Post-Release Analysis A post-release analysis will be done using actual release data to ensure that the limits specified in Section 1 were not exceeded.

MIO583-0010A-RS03

51 A composite list of concentrations (C ), by isotope, will be used with the actual liquid radwaste (f) and dilution (F) flow rates (or volumes) during the release. The data will be substituted into Equation 2.3 to demonstrate. compliance with the limits in Section 1.

This data and setpoints will be recorded in auditable records by

' plant personnel.

C. DOSE t:

1. RETS Requirement s

i Specification 3.24.4.1 the Radiological Effluent Technical Specifi-cation (RETS) requires that the quantity of radionuclides released be limited such that the dose or dose commitment to an individual from radioactive materials in liquid effluents released to unrestri:ted areas from each reactor (see Figure 2-2.) will not exceed:

a. During any calendar quarter, 1.5 mrem to the total body and 5 mrem to any organ, and

b. -During any calendar year, 3 mrem to the total body and 10 mrem to any organ.

To ensure compliance, quantities of activity of each radionuclide released will be summed for each release and accumulated for each

, quarter as follows in Section 2.

re

~

MIO583-0010A-RS03 i

G

~

J1 m

52 2.' Rel' ease Analysis Calculations shall be performed for each batch release, and weekly for continuous releases according to the formula:

A g /C $ 0.5 (2.5) where:

Ag = Cumulative quarterly activity of nuclide i identified in liquid release (Ci).

C = Design objective annual quantity of radionuclide i from f

Table 2.2.

' Radionuclides may be omitted from the summation if they fall under the criteria of allowed omission specified by Note 5 to Appendix B, 10 CFR 20.

The design basis quantities are derived in such a conservative manner that doses may be greatly overestimated by this technique. As a consequence of this conservatism, and in light of historically consistent operations with releases well below annual design basis quantities, the Palisades Plant technical specifications do not require monthly dose projections. Instead, if at any time, calcula-tions by Equation ',2.5) results in values greater than 0.5 for a given quarter or 1.0 for year-to-date, the NRC LADTAP code will be run to ensure that Specification 3.24.4.1 has been met.

Values for the design basis quantities (Ci), and the dose per Curie (D /C )g for each nuclide i shown in Table 2.2, were calculated as follows in Sections 2.1 and 2.2 HIO583-0010A-RS03

53 2.1 Water Ingestion The dose to an individual from ingestion of radioactivity from any source is described by the following equation:

i D = (DCF) xI grem (2.11) i=1 where:

th D = Dose for the j organ from radionuclides released rem.

J = The organ of interest.

(DCF) = Adult ingestion dose commitment factor for the j th organ from the i radionuclide rem /pC1, see attached as Table 2.1.

h I = Activity ingested of the i radionuclide, pCi.

f I is described by:

I f= ( i)( )(365) Ci' (2.12)

(1000)(d) where: ,

365 = Days per year.

Ag = Annual activity released of i radionuclide pCi.

V = Average rate of water consumption (730 ml/d ICRP 23,

p. 358).

d = Dilution water flow for year.

1000 = Dispersion factor from discharge to nearest drinking water supply.

MIO583-0010A-RS03 l

k.

iyr: a p

1. t

, 54 )

l

'The dose equation'then becomes:

i D= ' ( ij * ^i "#*"

(2.13) i=1-

- J2.2 Fish Insestion

? The dose to an individual from the consumption of fish is described by th Equation 2.13. In this case the activity ingested of the i radio-nuclide (I ) is described by:

f I = ^i i pCi-f 15d (2.14) where:

h A-g = Annual released of i radionuclide, pC1.

-B = Fish concentration factor of i radionuclide pCi/gm I'

see Table 2.0. pCi/ml, F = Amount of fish eaten per year (21 kg).

115 =. Dispersion factor from discharge to fish exposure point.

d = Dilution water flow for year.

-Substitution of Equation 2.14 into Equation 2.11 gives:

i D = ^i

• i

• i

"#** (*

, i=1 MIO583-0010A-RS03

F r

55

3. ' Annual Analysis A complete analysis utilizing the NRC computer code LADTAP with the total source release will be done annually in conjunction with the annual environmental report. 'This analysis will provide estimates of dose to the total body and various organs in addition to the dose limiting organs considered in the method of Section 2. The following th approach is utilized in LADTAP. The dose to the j f ,

radionuclides, D , is described by:

m D = D rem (2.16) i=1 m

= (DCF) xI rem (2.17) i=1 where:

th D '= Dose to the j -organ from the i radionuclide, rem.

q J = The organ of interest (bone, GI tract, thyroid, liver, kidney, lung or total body).

(DCF)gJ = Adult ingestion dose commitment factor for the j th organ from the i radionuclide, rem /pCi, see Table 2.1.

I

g. = Activity' ingested of the i radionuclide, pCi.

Ig for water ingestion is described by:

A Ig = ^i pCi. (2.18)

~

vd and for fish ingestion'I is described by:

f I

Ig = i i pCi (2.19) vd^

.MIO583-0010A-RS03 1

1 JT~ -

56 i

l 1

F where:

th

-A g = Activity released of j radionuclide during the year, pCi.

V' = Average rate of water consumption (730 ml/d).

T = Number of days during the year'(365 d).

v ;= Dispersion factor from point of discharge to point of

. exposure, d = Dilution water volume (ml).

Eg = Fish concentration factor of the i radionuclide, fj/gm F - = Amount of- fish eaten per day (57.5 gm).

D. . OPERABILITY OF LIQUID RADWASTE EQUIPMENT The Palisades liquid _radwaste system is designed to reduce the radioactive materials in liquid wastes prior to their discharge (by recycle or shipment for disposal) so that radioactivity in liquid effluent releases to unrestricted areas-(see Figure.2-1) will not exceed Specification 3.24.4.1. Maintaining the cumulative fraction of allowable release for each batch release and weekly for continuous releases assures compliance with this requirement. In addition, 13 years of operating experien;e (to'the date this ODCM was first adopted) has shown that design basis quantities never have been exceeded.

E. RELEASE RATE FOR OFFSITE MPC (500 mrem /yr) 10 CFR 20.106 requires radioactive effluent releases to unrestricted areas be less than the limits specified in Appendix B, Table II when averaged over a period not to exceed one year. Concentrations at this level, if ingested for_ one year, will result in a dose of 500 millirem to the total body or its equivalent to internal organs. In addition, 10 CFR 50.36a requires that the release of radioactive materials be kept as low as is reasonably achievable. However, the section further states that the licensee is permitted the flexibility MIO583-0010A-RS03

n 57 of operation, to assure a dependable source of power, to release quantities of material higher than a small percentage of 1 CFR 20 limits but not exceeding those limits under unusual operating condi-tions. Appendix I to 10 CFR 50 provides the numerical guidelines on limiting conditions for operations to meet the as low as is reason-ably achievable requirement.

The.LADTAP code has been run to determine the dose due to drinking water at plant discharge concentration (1,000 x nearest drinking water intake concentration). The source term used is given in Table 1.1. Dose to the most limiting organ of the person F .athetically

. drinking this water is (TBD) mrem. The release rate which would result in a dose rate equivalent to 500 mrem / year (using the more conservative total body limit) is the Curies / year given in Table 1.1 (TBD) times 500(TBD) or (TBD) Ci/yr = (TBD) Ci/sec.

III. URANIUM FUEL CYCLE DOSE A. SPECIFICATION In accordance with Specification 3.24.8.1, if either liquid or (fx gaseous quarterly releases exceed the quantity which would cause ffsite doses more than twice the limit of Specifications 3.24.4.1, 3.24.5.2.or 3.24.5.3, then the cumulative dose contributions from combined release plus' direct radiation sources (from the reactor unit and radwaste storage tanks) shall be calculated. The dose is to be determined for the member of the public projected to be the most 5C1 o highly exposed to these combined sources.

B. ASSUMPTIONS

1. The full time resident determined to be the maximally exposed individual (excluding infant) is assumed also to be a fisherman.

This individual is assumed to drink water and ingest local fish at the rates specified in Sections II C.2.1 and .I C.2.1.

HIO583-0010A-RS03

58 i l

l l

l l

2. Amount of shoreline fishing (at accessible shoreline adjacent to site security fence) is conservatively assumed as 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per i quarter (average of approximately hour per day each day of the quarter) for the second and third quarters of the year, 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for the fourth quarter and 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> for the first quarter.

' C ~. DOSE CALCULATION Maximum doses to the total body and internal organs of an individual shall be determined by use of LADTAP and GASPAR computer codes, and doses to like organs and total body summed. Added to this sum will be a mean dose rate, calculated or measured for the shoreline due to plant presence during the quarter in question, times the assumed fishing time:

Do=Dg+DL 4 + (RT )( ) ( . 0) where:

D4 o = 40 CFR 190 Dose (mrem)

Dg = Limiting dose tc an individual from gaseous source term (mrem)

D -= Limiting dose to an individual from liquid source term (mrem)

R = Mean dose rate calculated to be applicable to T

Lake Michigan shoreline adjacent to plant site (mrem /hr)

T = Assumed shoreline fishing time for the quarter in question MIO583-0010A-RS03

59 IV. RADWASTE SYSTEM MODIFICATIONS 1.0 Definition of Major Radwaste System Modification 1.1 Purpose The purpose of this definition is to assure that Technical Specifi-cation 6.20 will be satisfied under clearly identifiable circum-stances, and with the objective that current radwaste system capabilities are not jeopardized.

1.2 Definition A major radwaste system modification is a modification which would remove (either by bypassing or physical removal) or replace with less efficient equipment, any components of the radwaste system:

1) Letdown filters or demineralizers

2) Vacuum degassifier

3) Miscellaneous or clean waste evaporators

4) The present waste gas compressor / decay tank system 5)' Fuel pool filters /demineralizers

6) Radwaste polishing demineralizers 7)- Radwaste solidification system Improvements or additions to improve efficiency will not be considered major modifications unless a complete substitution of equipment or systems is made with equipment of unrelated design. Examples would be 1) replacement of mechanical degassifier with steam; jet degassifier 2) replacement of waste gas system with cryogenic system 3) replacement of ashphalt solidification with cement system 4) change from deep bed resins to Powdex 5) etc.

MIO5g3-0010A-RS03

FIGURE'2-1 PALISADES RADWASTE EFFLUENTS - LIQUID S

CLEAN _ f ll WASTE ~ '

4 SYSTEM STEAM GEN f

BLOWDOWN PRIMARY ,/ h TANK i SYSTEM T29B I STORAGE TANK N / RE 0707

/g /g f/-

^ T66A T66B T-90 ST I

TANKS (2) \/ \/ Ny TRIP & FLOW CONTROL r- - - ,

/

)

RE 1049 RE 0833 /

SERVICE C_

WATER '

l MISCELLANEOUS WASTE SYSTEM - 1; j (DIRTY WASTE) 3, J LAKE LAUNDRY UTILITY RE 1323 l ,

SYSTEM WATEH STORAGE Q 8

\ TANK -

a--

! T-63 MIXING :1

! FILTERED V T-91 TURBINE BLDG - -

BASIN j$ -

WASTE DRAINS -

l } I TANK  : TURBINE SUMP RE 5211 DlLUTION FLOW \

m i I - SERVICE WATER  :  ;

LAKE INLET g 4

\

a e

j - . . . *b

^

\

/ / '/ '

.I A >

/

l k / 13 f

/

j , t g \ + 1 / 8

-/

/

/

/ '

\ / ,/. l1

\

\ \

g

} - . . - .

4 / / -

\ 1* ./

' lr',

\ / /

{

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\

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\ o I / i

\ l l ,

l A l '\

l \ NNW f NNE I I N

\ NW

1. \

\

/ ,' / 1S g '/

% \ ;I.

/ NE /

/

/'

'"""~"" j

, a.

• sg s

/, x,p ,

c W

WNW f/ - 5. ,.+ / M

\ ,, .

GASEOUS EFFLUENTS

'- / '

I

• '.',,% N

% \ / ,,2

/ STACK 782 elev. #

/

7

~ #

./c- ~

3, l' l4 /7, -

E LIQUID DISCH ARGE ,# f, ,/, '( _ %g [- g-- qd.,, , .

',./,,, .. ,

'g'~

SU8 FACE 586 elev. p gg -

9 L WsW c

/:1.a s N 's.\ ,

. I.

l 'I g Est ~

f

,( N .jy  %, . , ~~~

I.

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/

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16 A y,

/ / , f  ; y ': - , - -

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/ sw f  ; I

\ sE s // N t -- .:  :

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, \ \ N *

,  ? " - .,,,

4 ssW * \ '

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\ sse \

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V s g JI \ \ ,

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\ \ N

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l -

l \ \

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[

7 li { 1"= 215 METERS A

g \ --s 70

• \ UNRESTRICTED AREA ~

i / gggg SITE BOUNDARY \

l;l \ 18 g N g TLD STATIONS ,

I A20 \ - \ '

N "

PALISADES NUCLEAR PLANT j g.- SITE DOUNDARY

/ 6 -

FIGURE 2-1 2

il

e, : v -

62

=

TABLE 2.0 BIOACCUMULATION FACTORS (pCi/kg per pCi/ liter)

FRESHWATER ELEMENT FISH H 9.0E-01

.C 4.6E 03 NA 1.0E 02 P 1.0E 05 CR 2.0E 02 MN 4.0E 02 FE 1.0E 02 CO 5.0E 01 NI 1.0E 02 CU 5.0E 01 ZN 2.0E 03 BR 4.2E 02 RB 2.0E 03

-SR 3.0E 01 Y 2.5E 01 ZR 3.3E 00 NB 3.0E 04 NO 1.0E 01 TC 1.5E 01 RU 1.0E 01 RH 1.0E 01 TE 4.0E 02 I 1.5E 01 CS 2.0E 03 BA 4.0E 00 LA 2.5E 01 CE 1.0E 00 PR 2.5E 01 ND 2.5E 01 W 1.2E 03 NP 1.0E 01 HIO583-0010A-RS03

r1

, 63 IABLE 2.1 ADULT INCESTION DOSE FACTORS (MREM /PCI INCESTED)

Pane 1 of 3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNC CI-LLI

~

H -3' 'NO DAIA 1.05E-07 1.05E-07 1.05E-07 1.0$E-07 1.05E-07 1.05E-07 C 14 2. 84E-06 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E 07 5.68E-07 NA 24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 P 32 1.93E-04 1.20E-05 7.46E-06 NO DAIA NO DAIA NO DATA 2.17E-05 CR 51' NO DATA' NO DAIA 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 NN 54 E NO DATA 4.57E-06 8.72E-07 NO DATA 1.36E-06 NO DATA 1.40E-05 NN 56 NO DATA 1.15E-07 2.04E-08 NO DAIA 1.46E-07 NO DAIA 3.67E-06 FE 55 2.75E-06 1. 90E-06 4.43E-07 NO DAIA NO DAIA 1.06E 06 1.09E-06 FE 59 4.34E-06 1.02E-05 3.91E-06 NO DATA NO DATA 2.85E 06 3.40E-05

-00 58 NO DATA 7.45E-07 1.67E-06 NO DATA NO DATA NO DATA 1.51E-05 CO 60 NO DATA 2.14E-06 4.72E-06 NO DATA NO DATA NO DATA 4.02E-05 MI 63 1. 30E-04 _9.01E-06 4.36E-06 NO DATA NO DATA NO DATA 1.88E-06 NI 45 '5.28E-07 6.86E-08 3.13E-08 NO DATA NO DATA NO DATA 1.74E-06

.CU 64 NO DATA 8.33E-08 3.91E-08 NO DATA 2.10E-07 NO DATA 7.10E-06

.ZN 65 4.84E-06 1.54E-05 6.96E-06 NO DATA 1.03E-05 NO DAIA 9.70E-06 ZN 69 1.03E-08 1.97E-08 1.37E-09 NO DATA 1.28E 08 NO DATA 2.96E-04 BR 83. NO DATA NO DATA 4.02E-08 NO DATA NO DATA NO DATA 5.79E-08 BR 84 ' NO DATA- NO DATA 5.21E 08 NO DATA NO DATA NO DATA 4.09E-13

'3R 85 - NO DATA NO DATA 2.14E-09 NO DATA NO DATA NO DATA LT E-24 RB 86 NO DATA 2.11E-05 -9.83E-06 NO DATA NO DATA NO DATA 4.16E-06 RB 88 NO DATA 6.0$E-08 3.21E-06 NO DATA NO DATA NO DATA 8.36E-19 RB 89- NO DATA 4.01E-08 2.82E-08 NO DATA NO DATA NO DATA 2.33E-21 SR 89 3. 08E-Of. NO DATA 8.84E-06 NO DATA NO DATA NO DATA 4.94E-05 SR 90. 7.58E-03 NO DATA 1.86E-03 NO DATA NO DATA NO DATA 2.19E-04 SR 91 5.67E-06 NO DATA 2.29E-07 No DATA NO DATA NO DATA 2.70E-05

,SR 92 2.15E-06 NO DATA 9.30E-08 NO DATA NO DATA NO DATA 4.26E-05 Y '90 9.62E-09 NO DATA 2.58E-10 NO DAIA NO DATA NO DATA 1.02E-04

Y 91m  ? 09E-11 'NG DATA 3.52E-12 NO DATA NO DAIA NO DATA 2.67E-10 Y 91' 1.41E-07 NO DATA- 3.77E-09 NO DATA NO DATA NO DATA 7.76E-05 Y 92 '8.45E-10 NO DATA 2.47E-11 NO DAIA NO DATA NO DATA 1.48E-05

. . . e e

M10583-00108-RS03 k

y; 4

64 IABLE 2.1 (CONT'D)

ADULT INCESTION DOSE FACTORS (MREM /PCI INGESTED)

Pase 2 of 3 NUCLIDE BONE LIVER' T. BODY THYROID KIDNEY LUNG CI-LLI Y '93 2.68E-09 NO DAIA 7.40E-11 NO DATA NO DAIA NO DATA 8.50E-05 ZR 95 3.04E-08 9.75E-09 6.60E-09 NO DATA 1.53E-08 NO DAIA 3.09E-05 ZR 97 1.68E-09 3.39E-10 1.55E-10 NO DAIA 5.12E-10 NO DATA 1.0$E-04 NB 95 6.22E-09 3.46E-09 1.86E-09 NO DATA 3.42E-09 NO DATA 2.10E-05

.MO 99 NO DATA 4.31E-06 8.20E-07 NO DAIA 9.76E-06 NO DATA 9.99E-06 TC 99m 2.47E 6.98E 10 8.89E-09 NO DATA 1.06E-08 3.42E-10 4.13E-07 TC101. 2.54E-10 3.66E-10 3.59E-O'9 NO DATA 6.59E-09 1.87E-10 1.10E-21 RU103' 1.85E-07 NO DAIA 7.97E-06 NO DATA 7.06E-07 N1 DATA 2.16E-05 RU105 1.54E-08 NO DATA 6.08E-09 NO DAIA 1.99E-07 NO DATA 9.42E-06 RU106 2.75E-06 NO DAIA 3.48E-07 NO DATA 5.31E-06 NO DAIA 1.78E-04 AC110m' 1.60E-07 1.48E-07 8.79E-09 NO DATA 2.91E-07 NO DATA 6. 04E-05 TE125m 2.68E-06 9.71E-07 3.59E-07 8.06E-07 1.09E-05 NO DATA 1.07E-05 TE127m 6.77E-06 2.42E-06 8.2SE-07 1.73E-06 2.75E-05 NO DATA 2.27E-05 TE127 1.10E-07 3.95E-08 2.38E-08 8.15E-08 4.48E-07 NO DATA 8.68E-06 TE129m 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 NO. DATA 5.79E-05 TE129 3.14E-06 1.18E-08 7.65E-09 2.41E 08 1.32E-07 NO DAIA 2.37E-08 TE131a 1.73E-06 8.46E-07 7.05E-07 1. 34E-06 8.57E-06 NO DAIA 8.40E-05 TE131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 NO DATA 2.79E-09 TE132 2.52E-06 1.63E 06 1.53E-06 1.80E-06 1.57E-05 NO DAIA 7.71E-05

-I 130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 NO DATA 1.92E-06 I 131 4.16E-06 5.95E-06 3.41E-06 1.95E-03 1.02E-05 NO DATA 1.57E-06 I 132 2.03E-07 5.43E 07 1.90E-07 1.90E-05 8.65E-07 NO DAIA 1.02E-07 1 133 -1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31E-06 NO DAIA 2.22E 06

'I 134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 NO DAIA 2.51E-10 I 135 4.43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 NO DATA 1.31E-06 CS134 6.22E-05 1.48E 04 1.21E-04 NO DAIA 4.79E-05 1.59E-05 2.59E 06 CS136 6.51E-06 2.57E-05 1.85E-05 NO DATA 1.43E 05 1.96E-06 2.92E-06 CS137 7.97E-05 1.09E-04 7.14E 05 NO DATA 3.70E 05 1.23E-05 2.11E-06 CS138 5.52E-08 1.09E-07 5.40E-08 NO DAIA 8.01E-08 7.91E-09 4.65E-13 BA139 9.70E-08 6.91E-11 2.84E-09 NO DATA 6.46E-11 3.92E-11 1.72E-07 MID583-00108 R$03

x. .

65 TABLE 2.1 (CONT'D)

ADULT INCESTION DOSE FACTORS (MREM /PCI INCESTED)

Page 3 of 3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNC CI-LLT BA140 2.03E-05 2.55E-08 1.33E-06 NO DATA 8.67E-09 1.46E-08 4.18E-05 RA141 4.71E-08 3.56E-11 1.59E-09 NO DATA 3.3tE-11 2.02E-11 2.22E-17 BA142 2.13E-08 2.19E-11 1. 34E-09 NO DATA 1.85E-11 1.24E-11 3.00E-26 LA140 2.50E-09 1.26E 09 3.33E-10 NO DATA NO DATA NO DATA 9.25E-05 LA142 1.28E-10 5.82E-11 1.45E-11 NO DATA NO DAIA NO DATA 4.25E-07 CE141 9.36E ai 6.33E-09 7.18E-10 NO DATA 2.94E-09 NO DATA 2.42E-05

'CE143 1.65E-09 1.22E-06 1.35E-10 NO DATA 5.37E-10 NO DATA 4.56E-05 CE144 4.88E-07 2.04E-07 2.62E*06 NO DAIA 1.2LE-07 NO DATA 1.65E-04 PR143 9.20E-09 3.69E-09 4.56E-10 NO DATA 2.13E-09 NO DATA 4.03E-05 PR144 3.01E-11 1.25E-11 1.53E-12 NO DAIA 7.05E-12 NO DATA 4.33E-18 c.

ND147 6.29E-09 7.27E-09 4.35E-10 NO DAIA 4.25E*09 NO DATA 3.49E-05 W 187 1.03E-07 8.61E-08 3.01E 08 NO DATA NO DATA NO DATA 2.82E-05 NP239- 1.19E-09 1.17E-10 6.45E-11 NO DAIA 3.65E-10 NO DAIA 2.40E-05 I

> l M10583 00108 RS03 y _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ _ _ _ _ _

hII b

66 F

TABLE 2.2 y

LIQUID EFFLUENT DESIGN OBJECTIVE ANNUAL QUANTITIES Design Objective Dose Factor Annual Quantity Nuclide Limitina Dose (mrem / Curie) (Curies)

H-3 Total Body 1.31E-06 2.29E+06 Cr-51 GI Tract 1.49E-03 6,730

, Ma-54 GI Tract 6.22E-02 161 Fe-55 GI Tract 1.21E-02 827 Fe-59 GI Tract 3.77E-02 265 Co-58 GI Tract 8.38E-03 1,190 Co-60 Total Body 4.50E-01 6.67 Rb-86 Total Body 2.28E-01 13.2 Br-89 Bone 1.07E-02 935 Sr-90 Bone 1.64E-01 61.0 No-99 GI Tract 1.11E-03 9,020 Te-99m -GI Tract 6.88E-05 145,000 Te-127m GI Tract 1.01E-01 99.2 Te-127 GI Tract 2.85E-02 259 Te-129m GI Tract 2.57E-01 38.9 Te-131m GI Tract 3.73E-01 26.8 Te-132 GI Tract 3.42E-01 29.2 I-130 Thyroid 5.28E-02 190 1-131 Thyroid 3.24E-01 30.9 I-132 Thyroid 1.32E-02 755 I-133 Thyroid 4.05E-02 254 I-135 Thyroid 2.83E-02 353 Co-134 Total Body 2.83E+00 1.06 Co-136 Total Body 3.94E-01 7.61 Cs-137 Total Body 1.67E+00 1.80 Ba-140 GI Tract 1.86E-03 5,390

-La-140 GI Tract 2.57E-02 390 Np-239 GI Tract 2.66E-03 3,750 MIO583-0010A-RS03

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

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PROPOSED e .

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, Consumers Power Company Palisades Plant - Docket 50-255 f:

[ PROCESS CONTROL PROGRAM (PCP).

f ', July 31, 1984 F ,

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T80784-0031A-MLO2

~

1 PALISADES PROCESS CONTROL PROGRAM (PCP) 1.0 ASPHALT VOLUME REDUCTION SYSTEM The Palisades Plant utilizes a Waste Chem volume reduction and solidi-fication system (VRS) to process various radioactive liquid waste

, streams.

The process utilizes thermal energy (heat) to evaporate water from the radioactive wastes thus reducing waste volume to anhydrous waste residue.

This residue is then er capsulated in a thermoplastic matrix (asphalt).

i The end product is a monolithic, freestanding solid with no free liquid.

Fifty-five gallon drums are used to contain the encapsulated waste for temporary storage, transport and burial.

1.1 VARIABLES INFLUENCING SOLIDIFICATION The purpose of this section is to identify ar.d define those process variables which have a direct effect on the ability of the final product to form a-free. standing monolith with no free liquid.

The following variables influence the properties and consistency of the final product:

a. Asphalt type

b. Waste chemical species used as feed

c. Ratio of waste-to-asphalt, and

d. . Process temperature 1.2 ASPHALT TYPE Asphalt utilized in the system shall conform to ASTM-D-312-71. Type III.

This is an oxidized petroleum-based asphalt, such as Witco Chemical Company's Pioneer 221. The specifications for this asphalt are provided in Appendix A.

This grade of asphalt has a low residual volatile content and a high molecular weight. At room temperature, and at all normal ambient temperature conditions, this material is a freestanding, monolithic, solid.

Utilization of an asphalt complying with ASTH-D-312-71. Type III, is the

• means by which process control of this variable is achieved, rp0684-0677a-42-63

r 2

1.3 WASTE CHEMICAL SPECIES The type and relative quantity (waste-to-asphalt ratio) of waste chemicals being incorporated into the asphalt matrix has a direct influence on the properties of the final product. Encapsulation of inorganic salts and solids typically " stiffen" and harden the end product, whereas organic liquids have the opposite tendency. When the specified ratio of waste-to-asphalt is maintained, final product properties for typical power plant wastes are independent of the waste type.

However, certain chemical specifications are required as an outer bounds to limit end product tendencies to soften at lower temperatures.

A maximum limit of 1% oil by weight will be applied to the waste feed streams. Most oils found in power plants are low viscosity fluids, which are liquid at room temperature. Based on calculations for a typical waste stream with 20% solids by weight and 1*. oil by weight, Waste Chem has found the total concentration of oil in the end product would be approximately 2.5%. This would then lower the end product softening point by approximately 5'F, or approximately 2*F lower per percent of oil. This is within an acceptable range and, therefore, is the basis for the limit of 1% oil in the feed stream.

Uther chemical specifications on feed streams are specified below. These are required primarily for calculating waste-to-asphalt ratio which is important to end product, and equipment protection (which will have no discernable effect on the end product).

REQUIRED ANALYSIS Concentr)tes Resin /Powdex pH(Equip;1limit). pH (Equip Limit)

% Solids  % Slurry Sp Gravity .

Oil %

1.4 WASTE-TO-ash! ALT RATI,0 The ratio of waste-to'-asphatt contained in the end product has the most significant effect on the viscosity and physical consistency of that product. Process control is achieved by placing limitations on the range of wcste-to-asphalt ratios allowable for each waste type.

Vaste-to-asphalt ratics (mass) and evaporative rates should not exceed the verification test values specified for the waste feeds as follows:

rp0684-0677a-42-63 L

3

. Ratio of Vaste-to-Asphalt Feed in the End Product

1. Boric Acid Concentrates at 120 L/hr Evaporative Rate 5 1.0/1.0

2. Spent Resins at 80 L/hr Evaporative Rate 5 .67/1.0

3. Powdex at 80 L/hr Evaporative Rate 5 .67/1.0 Should the ratio of waste-to-asphalt be increased above the range specified in the foregoing table, the end product viscosity will increase and may exhibit a grainy texture. This could lead to " pyramiding" of the product in the container, thereby decreasing the container filling efficiency. In all cases, the product will cool to form a freestanding monolith. If lower than specified waste loadings are realized, the end product properties will approach that of pure asphalt. Again, solidification is assured; however, toward this end of the spectrum, economical volume reduction may not be realized.

Maximum concentrate feed rate can be determined by the following formula:

Conc Feed Rate = 0.528 GPM (1.0 - Solids Fraction) (Sp Gravity)

NOTE: 0.528 gpm = 120 L/hr evaporative rate.

The corresponding asphalt feed is calculated by:

Asphalt Feed (GPM) = (Cone Feed Rate GPM) (Solids Traction) (Sp Gravity)

(Waste-to-Asphalt Ratio) where the recommended waste-to-asphalt ratio is 1.0.

NOTES: 1. The density of Type III asphalt is 1.0 so a density correction is not needed.

2. The minimum asphalt flow is 0.065 spm bncause of lubrication requirements of the twin screws.

3. If either the concentrate or asphalt flows cannot be met, the calculated flows can be ratioed to new values to maintalu the 1.0/1.0 waste-to-asphalt ratio as long as the maximum concentrate flow or the minimum asphalt flows is not exceeded, rp0684 0677a-42-63

4 Maximum bead resin or Powdex can be determined by the following formula:

Resin Feed = 0.35 GPM (1.0 - Solid Fraction)

NOTES: 1. Solid fraction = slurry fraction & 2. Example is a 50%

slurry = 25 weight %.

2. 0.35 gpm = 80 L/hr evaporative rate.

The corresponding asphalt feed is calculated by:

Asphalt Feed (GPM) = (Resin Feed) (Solid Fraction)

(Waste-to-Asphalt Ratio)

Where the recommended waste-to-asphalt ratio is 0.67, the notes on the preceding asphalt calculation apply.

The operator can also visually confirm that the quality of the end product is approximately being maintained. A CCTV camera " views" the discharge from the extruder-evaporator, and a TV monitor located in the Solid Radwaste Building Control Room allows the operator to observe the physical consistency of the product as it is discharged into the container. At evaporative rates higher than specified, there will be excessive steaming at discharge nozzle. At higher waste-to-asphalt loading the discharge will appear grainy and stringy.

1.5 PROCESS TEMPERATURES A proper temperature profile along the length of the extruder-evaporator is required to provide adequate evaporative (process) capacity, and to assure that free water is not discharged from the machine.

Process temperature profiles for waste feeds should be maintained as recommended below:

Waste Type Process Temperature (*F)

Zones: 1 2 3 4 5/6 7 Boric Acid Concentrates LWS Concentrates 300* 280' 280' 300* 300*

• Chemical / Laundry Waste Spent Resins /Powdex 300* ** ** ** **

• NOTE: No zone shall tu maintained below 240*F.

rp0684-0677a-42-63 t

i 5 p- Low temperature alarms are provided to alert the operator to a low

)'

R. '

temperature out-of-specification condition which could potentially lead to the discharge of free water. If the out-of-specification condition persists for two (2) minutes, the extruder-evaporator is automatically tripped to prevent free water from being discharged into the container.

Free water cannot be discharged in the interim, since the residual heat of the extruder-evaporator itself is sufficient to effect evaporation.

Verification of the absence of free water and product solidification will be made~on a minimum of one container from every tenth batch or run.

Verification is recommended to be done on every tenth drum shipped. The required container shall be examined through a removable lid bung or equivalent means for solidification by checking penetration with a solid tool and inverted for a minimum of eight hours to check for free water.

Evidence of free water other than a few drops of condensation shall be cause for rejection and evaluation system product.

2.0 DEWATERING SOLIDS IN HIGH INTEGRITY CONTAINERS (HIC) 2.1 Solids such as bead resin, filter cartridges and powdered resin (Powdex) may be dewatered and shipped in HICs per approved vendor procedures and the HIC certificate of compliance.

2.2 1,ligh~ integrity containers are approved by the individual burial ground agreement states as meeting 10 CFR 61 waste form stability requirements.

2.3 Free water determination shall be verified by the successful completion and documentation of the vendors approved dewatering procedure.

3.0 DELAWARE CUSTOM MATERIAL (DCM) - SILICATE CEMENT Liquid wastes can be solidified by the DCM method. The silicate solidifies and the cement gives structural strength.

3.1 For solidification, acquire a representative sample of waste. Before following the guidelines outlined below, determine the type of waste to be solidified, example, lab waste, laundry waste, decon solutions, boric acid, oil, etc. Sample for pH, boric acid, visible organics and radioactivity.

Use analysis to determine the proper laboratory procedure to test.

All batches shall be lab tested prior to solidification in a larger container unless sample analysis (pH 0.2 and Boron i 20%) matches the analysis of a waste type which has previously passed lab test criteria.

For all oil waste, do not exceed 50% by volume. 011 must be emulsified with a detergent or boric acid in some type of neutral aqu-ous waste or tap water.

NOTE: 011 cannot be shipped to Barnwell, South Carolina.

rp0684-0677a-42-13

6 For spent resins, liquid absorbent, or other earthen-like material, dilute with an equal volume of concentrate or tap water to solidify.

All results must be recorded initially, at approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after testing. Grade observations to evaluate sample mixes. The 48-hour test can be omitted if the 24-hour test is good.

The quantity of chemicals added to solidify radwaste shall be within 20*.

of the quantity'as determined by the laboratory test of Step G.

3.2 SOLIDIFICATION AND FREE WATER DETERMINATION Solidification shall be. considered successful if, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after completion of Appendix A solidification, there is not standing water on the waste surface and the surface is not penntrated more than 2" with a 1" diameter rod. If deeper penetration is possible, then the drum can still be considered solid if the penetration hole remains open after the rod is withdrawn.

Silicate cement shall cure for a minimum of 28 days prior to shipment for disposal. For silicate ceme.,t drums, the following shall be done:

Each drum shall be inspected for absence of detectable freestanding

' quid after curing at least 28 days. With the drum lid installed, li invert each drum and allow drum to remain upside down for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, inspect each drum by placing upright and removing the lid. The RMC Supervisor or deaignate and a QC Inspector shall inspect each drum for presence of liquid. Drums which failed the 48-hour solidi-fication evaluation should be capable of passing at this point. If no detectable freestanding liquid is present, the drum can be prepared for shipment. RMC Supervisor and QC Inspector document if no detectabla freestanding liquid is present.

In the event liquid is observed, those drums with liquid shall be drained of all liquid. When no further liquid can be drained from the drum in a 24-hour period, the drum shall be core-bored or overpacked with two bags of approved absorbent and inspected by QC and RMC to verify that the drum is dry. After this verification (and documentation) the drum may be prepared for shipment.

Inspect the drum lid and gasket for defects prior to lid installation.

Install lid. Use a different lid if defects are found which prevent a tight seal between drum and lid.

4.0 10 CFR 61 REQUIREMENTS 4.1 10 CFR 61 classification requirements will be met using Wastetrak computer software program using the scaling factor methodology of rp0684-0677a-42-13

>Q

7 AIF/NESP-027, Methodologies for Classification of Low-Level Radioactive

> Wastes From Nuclear Power Plants, 1983.

The scaling factors will be updated by an ongoing analysis program of actual waste streams. The program will initiate with semiannual samples of availabic waste streams and may be modified to longer intervals if the data base warrants. Waste streams should include, if available; bead resin, concentrates, reactor coolant, clean waste, filter crud and compacted trash.

4.2 10 CFR 61 waste form stability requirements will be met by generic testing of the asphalt / waste stream product. The generic waste streams will be boric acid, bead resin and chemical regenerative wastes.

4.3 Documentation of the waste stream analysis, waste form stability and computer software scaling factor security shall be maintained by the Radiological Services Department.

9 rp0684-0677a-42 13

APPENDIX A Consumers Power Company Palisades Plant -

PROCESS CONTROL PROGRAM (PCP) t T80784-0031A-NLO2

c. - -.-

h L

E ASPHALT TECHNICAL DATA

SUMMARY

WITCO CHEMICAL - PIONEER 221

1. Basic Constituent Pioneer 221 is an oxidized petroleum base asphalt. Oxidation is accom-plished by' air bl7 wing at temperatures ranging from 200'C (392*F) to 300*C (572*F). Air blowing results in a product with minimum volatile content (0.1%), greater inertness and higher molecular weight.

2. Flash Point .

The Flash Point of Pioneer 221 is in excess of 288'C (549'F). The Flash Point is determined by the Cleveland Open Cup (ASIN D92-71) method. It is the lowest temperature at which surface vapors will momentarily ignite when a test flame is passed over the surface.

3. Fire Point The Fire Point of Pioneer 221 is in excess of 300*C (572*F). The Fire Point, like the Flash Point, is determined by the Cleveland Open Cup

.(ASTM D92-72) method. It is the lowest temperature at which the surface vapots will burn for at least 5 seconds before going out, the vapors being ignited as in the test for Flash Point.

4. Innition Point The Ignition Point of Pioneer 221 is approximately 400'C (752'F).

The Ignition Point is the lowest temperature at which the heat loss from the combustible mixture is exceeded by the heat projuced in the chemical reaction. It is thus the lowest temperature at which combustion begins and continues in an air environment.

5. Softenina Point The Sof tening Point of Pioneer 221 is in the temperature range of 88-94*C (190-201*F).

The Softening Point is determined by the Ring and Ball method (ASTM D-36-70).

6. Viscosity The Viscosity of Pioneer 221 in the temperature range from 250*F to 400*F is presented in attached graph.

A-1 u

T:

The graph is based on the following data from Witco Chemical:

Saybolt Furol Viscosity at 205'C 54 see at 177'C 161 sec

7. Penetration The Penetration of Pioneer 221 by ASTM Method D-5-73 for various temperatures is given below:

25'C (77'F) 22-30 dem 46*C (115'F) 40-60 den 0*C (32*F) 13-18 dem The abbreviation " den" means one-tenth of a millimeter. The number of dom's represents needle penetration under standard conditions of loading and time for a given temperature.

8. Specific Gravity The Specific Gravity of Pioneer 221 is approximately 1.0 gram per cc.

Specific Gravity is determined by ASTM Method D-70-72, which employs a pyenometer. A pyenometer is a container of known volume which is weighed empty and filled with sample.

9. Solubility

. Pioneer 221 may be considered to be entirely waterproof and insoluble in water. Pioneer 221 is soluble in petroleum solvents such as naphtha, mineral spirits and kerosene, in addition to carbon tetrachloride, carbon disulfide and trichlorethylene.

A-2

t i

PIONEER 221 LAMINATING & INDUSTRIAL ASPHALT PIONEER E-7465 FOR SALT CARTON MANUFACTURERS I

Pioneer 221 is an all-purpose, tough, medium sof tening point asphalt for use in laminating paper, foil-to-paper, as a base pigment for paints and var-mishes, or in the manufacture of sealers and adhesives.

Pioneer 221 complies with federal specifications set forth by the Food & Drug Administration for use in packaging and sealing food products and will not stain, or impart an oder or taste when used properly in connection with packaging products.

• PHYSICAL CHARACTERISTICS Softening Point 190-210*F Penetration 9 77'F 20-30 dem Ductility 9 77'F 2.5 ces +

Solubility CCL 99.0% + +

Flash Point (C.O.C.) 550*F +

Weight Per Gallon 8.3 lbs Use Temperature 400*F 1 25' Viscosity 9 400*F .94 secs Viscosity 9 375'T .174 secs Viscosity 9 350*F .360 secs Packasing: Bulk - Tankwagon ( 5000 gal), tank car ( 10,000 gal) i Packages - 100 lb cartons e 1 4

I i

A-3

TABLE 1 - COMPARIS0N OF ELEMENTAL ANALYSES AND OTHER CHEMICAL PROPERTIES OF ROAD TAR, COAL-TAR PITCH AND PETROLEUN ASPilALTS ASPRALT ROAD ROOFING R00FERS MATERIAL CEMENT TAR ASPilALT PITCil CRADE AC-10 RT-12 TYPE III TYPE A ELEMENTAL ANALYSIS, perconte C 85.8 92.2 86.0 92.8 H 9.7 5.2 9.9 5.1 N 0.6 1.5 0.5 1.5 0 0.5 1.0 0.7 ---

S 2.8 0.6 2.9 1.53 C/H atomic ratio 0.74 1.49 0.73 1.53 MOLF.CUI.AR WETCHT (Numbern Ava) 1030 420 1160 497 CARBON ATOM DISTRIBUTION:

(percent of total carbon)

AROMATIC CARBON 34 80 37 79 NAPitTHENE CARBON 23 15 23 18 PARAFFIN CARBON 43 5 40 3 1

A-4 750784-0031-MLO2 1-

ATi ACHMeur A eq

.  % Con 30mtf5 Power y company 1

..,,,, o,,,.. m . ,. .,,,.. ...~ .. m . o,e. w ,.is, u .. .... S w> r U June 7,1982 Dennis H Crutchfield, Chief Operating Reactors Branch No 5 Nuclear Reactor Regulation US Nuclear Regulatory Commission k'ashington, DC 20555 DOCKET 50 155 - LICENSE DPR 6 -

BIG ROCK PO!h7 PLANT AND DOCKET $0 255 - LICENSE DPR PALISADES PLANT ADDITIONAL INFOR!!ATION PERTAINING TO RADIOLOGICAL EFTLUENTS TECHNICAL SPECIFICATIONS (RETS)

Consumers Power Company agreed in a meeting with the NRC, held in Grand Rapids on January 4 7, 1982 to supply Franklin Research Center, which was contracted by the NRC to review and evaluate Consumers Power Company's RETS submittals, additional information to amplify and corroborate Consumers Power Company's previous RETS submittals. Consumers Power Company letter dated February 3, 1982 provided a portion of the requested information. It is the intent of this correspondence to furnish the NRC/FRC with additional information related to the following technical areas:

1. For Big Rock Point, provide rationale to demonstrate that 5 mR/hr cor-responds to a reasonable curie limit in each tank (ie, provide basis for number of curies per 5,000-ga11on tank). Please aee the enclosed Appen-dix (1) for Consumers Power Company's response.

2. For Palisades, provide justification for omitting Fn-55 analysis. Please see the enclosed Appendix (2) for Consumers Power Company's response.

3. For Big Rock Point and Palisades, perform 100/E calculations to prove that curie limits in 11guld and gas tanks are not needed. Please see the enclosed Appendix (3) for Consumers Power Company's response.

nu0682 0131a 43-153

T 2

To briefly summarise the information provided, the following synopsis is given:

Appendix (1) j Exposure rates of 5 mR/hr and above correspond to reasonable curie limits for tanks considered at Big Rock Point (T15A and 715B which are waste hold tanks and T-40 which is the condensate storage tank). Exposure rates of 5 mR/hr above background for each of these tanks corresponds to a curie content which is a small fraction of the design basis quantity limits (for liquid releases).

If one of these tanks ruptured with a 5 mR/hr reading and released its con-tents to the environment, the offsite doses would be a small fraction of 10 CTR 50, Appendix ! limits. Big Rock Point liquid tanks may be controlled -

as described in this Appendix.

Appendix (2)

Effluent calculations must include any radionuclide which contributes greater than 20*. of the HPC fraction in a release (10 CFR 20, Note 5 to Appendix B).

It is shown that Te 55 is below 0.251, of total MPC fraction and, therefore, may be excluded from effluent reporting requirements.

Appendix (3) i The last E determination at Palisades was used to calculate the gaseous radio-nuclide ratio present in primary coolant. The activities corresponding to these ratios were increased to the Technical Specification limit and compared to design basis quantity limits (for gaseous releases). If one gas tank contained the entire primary coolant gaseous activity based on operating at the Technical Specification limit, it would not exceed the design basis quantity curie limit. Curie limits are not needed on gas tanks at Palisades (as shown in Appendix 3) since this represents a conservative upper limit.

Big Rock Point has no gas tanks.

The liquid tanks of concern at Palisades ($1Rh', T 90, T 91 and T 2) are all largeenoughtoholdtheentireprimarycoolantvolume. This volume (10,900 ft ) was used in conjunction with the last I calculation to derive the total activity in the coolant at the maximum average concentration of 0.1 wC1/se dose equivalent 1-131. If this total activity were in one tank, it would be roughly 1.23 times the design basis quantity curie limits (for liquid releases). This amount, however, does not take credit for the domineralizers and filters that reduce the activity by a factor of at least ten, prior to entry to the tanks. It is extremely unlikely that any tank would contain more than the design basis quantity curie limit and total activity 1 faits for these tanks should not be imposed.

'the enclosed Appendix (4) provides the Design Calculation Checklist as required in Corporate Procedure HPS 09.

As agreed upon with the NRC, one additional commitment remains to be fulfilled by Consumers Power Company. A draft Offsite Dose Calculation Manual (ODCM) will be transmitted to the NRC/TRC when completed by October 1982. This ODCM nuo642 0131a 43 153

[ ,

3 i

will be commensurate with the sequoyah model ODCM, except for modifications for plant s'ecific p applications.

Please contact us if you have any questions not addressed in this '

serrespondence, t

David J VandeWalle (Signed)

David J VandeWalle Nuclear Licensing Administrator i CC Administrator, Region !!!, L'SNRC NRC Resident Inspector Eis Rock Point '

NRC Resident Inspector Palisades i

I i

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l TISA and 715P. WASTE N01,D TAW S Tanks T15A and T15B are 5000 gallen vaste h'old tanks. If one of these tanks  !

ruptures, the spillage vould be released directly to the environment. If we assume that the tank contains the design basis quantity curie limit and cal-eulate the exposure rate at the surface of the tank, we can limit the curie ,

content of the tank by simply measuring the exposure rate and comparing to i the exposure rate obtained at the design basis quantity limit.

[

Table 1 lists the radionuelides of interest, design basis quantities, and '

parameters needed in subsequent calculations. Also attached is a data sheet ,

for the vaste hold tank.

To determine the photon flux density at the surface of the tank (at the mid- '

point) the method described in Section 5.8 of " Concepts of Radiation Deci-metry" by Kenneth R. Kase and Walter R. Nelson was used for the no shield case. ,

The flux density (4) for a right-circular cylinder source is given by the following formula:

8

, , s yn F(e.ta ) '

2(a + 2.)

Where g a photon flux density (photons /cm2sec)

By a source strength (photons /cm8sec)

Ro a radius of cylinder (cm) a = distance from cylinder to dose point (cm) l F(4,b) = funetton to approximate the flux from the cylinder as it relates to a line source E = self absorption distance (cm) b2

• Wat where ys is the total scatter coefficient for  !

vetor (cm *)

orsphieally, we haye T Where Ro = 122 cm (k feet) a=0

~T ' h = k88 em (i

, 9a=0.074ca"ffeet) from p. 138 j l

6', of Rad. Health Handbook '

h ,

for 0 905 MeV gammas f

i

, b2 = Wa t = 2.88 from Figures ,

, A.21 and A.22 of " concepts

,***, e of Radiation Dosimetry" l

.. - _ _ _ _ _ _ _F _-

• _ . _ _ - -

r.

2 F(0,b2) = LE-2 from Figure A.1h of " Concepts of Radia-tion Dosimetry" (using 0 = 81 )

Sy = 6.26E + 5 photons /c=3see (from Table 1)

The photon flux density at the surface of the tank is then:

1. " (6.26E + 5 phetons/cm3sec )'122cm)2 (LE-2 )

2(0 + 38.9e=)

= h.79E + 6 photons /cm 2see Page 132 of the Bad Health Handbook gives a graph of particle fluence to give

1 R/hr. - The graph lists 5.8E + 5 photons /cm2 see to give 1 R/hr for 0.905 MeV photons. . The exposure rate at the surface of the tank is then:

R/hr = h.79E 5 8E ++ 56 = 8.26 R/hr The only thing not considered thus far is the shielding p ovid'ed by the tank vall. This is corrected,.by the,fdlloUing relationship:s

, D=De'-p[to

. N

.gWliere: D = Exposure rate - shielded

\ >

.s 5 s Do = Exposure rate - unshielded s

~

,t m 4-

' ~

p = Atte'nuation coefficie'nt, (c/ iron and 0 905 MeV photons D

J( = 0.063h~ em2 fg. c 3..s '

g,

.p = DEniN y efsiron = 7.86 g/cq3 5 "

4' s (,* *

k . ,

., t = thie5tnessiof'Dall = k." = 0.63YeIn ,

i s , ~ -

\ ..D = 8.26 e -(0.0631)(7.86)(0 635) a 6.02 It/hr

- < Therefore, if one vaste hold tank cond,ains the design basis Avantity curie limit,

~

y thelc5ntact exposure rate at the cank mid N int is 6102 R/hi.3

[Efi jsc 's dould certainly correspond to a resonable curie. limit in the tank.A, reading 1/100of 5=R i

~

In fact, 71'i. T Cof the limit would read'60 mR/hr.

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T h0, COIDENSATE STORAGE TANK The same =ethodology is used as for the vaste hold tanks. Table 1 data was used with a volume of 25,000 gallons (9.h5E + 7 cc) instead of 5000 gallons.

To find the photon flux density ve use the following parameters:

Ro = 305 cm (10 feet) a=0 h = 366 cm (12 feet) ps = 0.07h cm-1 b2

• UsZ = 3.86

-F(0,b2 ) = 1.3E - 2 (using e = Th.1 )

sv= 6.26E + 5 (5000/25,000) = 1.25E + 5 The photon flux ' density at the surface of the tank is then:

, (1.25E + 5)(305)2(1.3E - 2) = 1.h5E + 6 photons /em23,e 2(0 + 52.2)

The exposure rate at the. surface of the tank is:

R/hr = 1 ,

= 2 50 R/hr The vall thickness of the tank is assumed to be one inch of aluminum. It is probably not this thick and assu=ing one' inch is conservative. The exposure rate after passing through the vall is:

R/w - 2 5 e-( * '* )( '" ')( = 1.6 R/w Where: 25 = unshielded exposure rate (R/hr) 0.06h8 = attenuation coefficient (cm2 /g) 2.699 = density of alu=inum (g/cm3) 2 5h -= vall thickness (cm)

A reading of 5 mR/hr at the midpoint of the condensate storage tank (6 feet above grade) corresponds.to a reasonable curie limit in the tank.

_ , , - . - - . . - - . ,, --,, ,-w- - , , ,----,,--------m , - - - ~ - - - -

I H . :3 v .o J

k TABLE 1 l.

DESIGN BASIS NUCLIDE  : ENERGIES (MeV) YTELD (%) QUANTITY (Ci) .Pir0 TONS /cc nec cr-51 0.320 90 6320.0 1.11E +.6 Mn-Sh O.'835 100.0 129.0 2.53E + 5 co-58 0.511 30.0 '887.0 '2.27E + 6 O.810 99 0 0.865 1.h 1.67 0.6 Fe-59 'O.1h3 0.8 709 0 1.hhE + 6

~0 .192 2.8-1.095 56.0 1.292 hh.0 co-60' 1.173- 100.0 10.0 3 92E + h 1.332 100.0 Zn-65 0 511 3.h 366.0 3.75E + 5 1.115 49.0 Sr-89 0 91 0.009 588.0 1. 0h E + 2 Sr-90 -

61.0 Ag-110m 0.658 96.0 63.03' 3.66E + 5 0.68 16.0 0.706 19 0 0.76h 23.0 0.818 8.0 0.885 71.0 0 937 32.0 1.38h 21.0 1 505 11.0 Sb-12h 0.603 97.0 321.2* 1.19E + 6 0.6hh 70 0.72 1h.0 0.967 2.4 1.0h8 2.h

n. '

, j..

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. 2.: .

TABLE'1 (continued)

DESIGN BASIS NUCLTDE ENEF:GIES (MeV) YIELD (%) . QUANTITY (Cl) .PIf0 TONS /cc see sb-124 :(cont) 1.31 3.0 1 37 '5 0 1.h5 2.0 1.692 50.0 2.088 7.0 1-131 0.08 2.6 23 9- h.60E + h 0.28h 5.h 0.364 82.0 0.637 6.8 0 723 1.6 Cs-13h 0 57 - 23.0 0.829 ~ 3.67E + 3 O.605 98.0 0.796 99 0 1.038 1.0 1.168 19 1 365 3.4 Cs-137 0.662 -85 0 1.39 2.31E + 3 La-lho 0 329 20.0 h55 0 1.67E + 6 0.h87 h0.0 0.815 19 0 0.923 10.0 1 596 96.0 2.53 3.0 I Yield = 1673.7 8.77E + 6 g , E(E g)(Yield) _151h.67 = 0 905

.I Yield 1673.7 Limit =8.7y+6 = 6.26E + 5 Notes

1. Nucliden were obtained from the 1981 Seminnnual Effluent Reportn for Big Rock Point. Thene were all the nucliden identified in liquid ef.']uentn.

7 =. -

D.

3 TABLE 1 (concluded) 2.- -Volume of vaste hold tank is 5000 gallons = 1.89E + 7 cc.

3. Energi.es and yields are from the Radiological Health Handbook (1970).

h. Design Basis Quantities are from our-submittal to the'NRC dated November 13, 1978. Those with an

, asterisk are from nemo SBB 82-03 5 Photons /ce'sec =-(Design. Basis QuantityiC1)(E% yieldg)(106pC1/C1)(3 7E + h, photons /sec 901)

(100)(1.89E + 7ec)

Where: Design Basis Quantityi = Design Basis Quantity for Radionuclide i I % Yieldg = Summation of % Photon yicids for Radionuclide 1

6. The design basis quantity listed for each radionuclide vould give 10CFR50, Appendix I, done limits as if only that radionuclide were present. In other words, if the sum of all'1h radionuclides were used, the total dose would be lh times 10CFR50, Appendix I limits. Therefore, the tank would contain the design basis quantity curie limit, if one fourteenth of each radionuclide were in the tank. This is why the total photons /cc see is divided by 14

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, s BASIS FOR OMISSION OF Fe-55 A? ALYSIS REQUIREENTS

. Iron-55 production can.be chovn to give rise to activity levels not exceeding a factor of 38 times.that of-Iron-59 in reactor vessel ec.ponents or primary coolant of an operating reactor:

-Activity (Fe-55) , IN o4N4(1-eo ) , c 5N " = 38 at t = =

Activity (Fe-59) NVo c8U 8(1-e' # ) 09Na Where: INo = thermal neutron fluence rate (n/sec) og = themal cross section Fe-Sh (2.3 barns) to produce Fe-55 og = themal cross section Fe-58 (1.16 barns) to produce Fe-59 N4 = Atoms Fe-Sh (5.8% total Fe)

Ng = Ato=s Fe-58 (0.3% total Fe) 1-e-At = equilibrium fraction (1.0 at t = = )

Iron-59 data for radioactive liquid discharges over the past 6 to 9 years frem the Palisades and Big Rock' Point plants have been reviewed (see Table I attached) and the percentage contribution of activity due to Iron-59 has been found to aver-age 0.k2% for Palisades and 0.62% for Big Rock Point. These percentages agree well with primary coolant analyses which indicate that Fe-59 is in the range of 0 3%.to 0 9% of activity from coolant, based upon radionuclides with half-lives greater than approximately 10, days (those typically found in effluent releases).

-In all cases,-tritium and dissolved gasses have been excluded from the calcula-tions. Inclusion of tritium in the total activity released reduces the percen-tage radioactivity contribution from Fe-59 to well belov 0.1% and Fe-55 to belov 3.8% in all cases.

On a basis of Maximum Pemissible Concentration (MPC per 10CFR20, Appendix B, TableII),Fe-55/Fe-5g/5-x105=ho.(soluble)

(insoluble) = 2 x 10- Thus,=Iron-55 8 x 10-4/6 x 1013.3 is from 5 = to 13 LO 3 and Fe-55/Fe-59 times

'less limiting than Iron-59 from an MPC (or dose) standpoint. Iron-59 has been observed only in insolub,1_e form where separation by solubility has been per-formed (primary coolant E analyses). This indicates that the maximum quantity of Iron-55 (38 times the quantity of Iron-59) provides an MPC-fraction essen-tially equal to that of Iron-59 in effluent releases 38 x Fe-59/(ho)(MPC Fe-50)=

0 95 x Fe-59 MPC fraction.

During the periods when Fe-59 was the highest percentage of released activity at Big Rock Point and Palisades (approximately 1% - see Table I), the fraction of total MPC accounted for by Fe-59 was 0.25%. Since we have shown above that in the vorst case (infinite activation time), the MPC-fraction of Fe-55 approaches that of Fe-59, the maximum contribution of Fe-55 to the total MPC-fraction re-leased also is 0.25%, calculated in accordance with Note 5 to Appendix B of 10CFR20. This note states: " ... a radionuclide may be considered as not present

m 2

~

in a mixture if ,(a) the ratio of the concentration of that radionuclide in the mixture (Cg ) to'the concentration limit for that radionuclide specified in Table II of. AppendixlB (MPC A ) does not exceed 1/10 (i.e. C A/MPCA 5 1/10) and (b) the sum'of such ratios for all radionuclides considered as not present in the mix-ture.does not. exceed 1/k i e. (C A/!GC A + C3 /!GC B . . . .+ 5 1/h ) . " Since it is shown that Fe-55 is less than approximately 0.2:% of the total !GC-fraction of

the releases, there should be.no further analysis or reportin6 required fer Fe-55

(E o

TABLE I I.' '

C: Fe-59 AND TOTAL ACTIVITY FROM SEMIANNUAL EFFLUENT REPORTS y

E- PAL PAL PAL BRP BRP 2RP

[ Fe-59 Total  % Fe-59 Total  %

1973 0.055 23.3 0.2h 0.068 27.8 0.2h 197h. .0.00059 3.08 0.02 0.00077 5 27 0.015 1975 0.0027 3.28 0.082

-1976 'O.0016 0.ho 0.ho 0.0011h 0 566 0.20 0.0018- 0.h3 0.k2 -

0.205 -

1977 ND :0.088 - -

0.103 -

ND 0.00h8- - -

0.392 -

1978 - ND _ 0.059 -

0.00036 0.038 0 95 0.00029 0.1h9 0.20 1979 0.00022 0.028 0.79 0.0050 0 573 0.87 0.0011 'O.10 1.10 -

0.33 -

1980 ND 0.00k1 - -

0.695 -

ND 0.00h6 -

0.00206 0.225 0.92 1981 ND 0.00L6 -

0.00237 0.26h 0.90 0.k26

• 0.39 0.618

• 0.38 LN_

AFFE'.JIX 3 e

4 5

e

.A _ - - ~ _ . - . _ . _ . . _ _ . . . _ . _ . . _ . . _ _ _ . .,,

7 I

1 -

ACTIVITY IN PALISADES' GAS TANKS The last Palisad'es E calculation (dated June 6, 1981) was used to obtain the maximum primary coolant gaseous activity at the technical specification limit u~

of 1.0 pCi/gm of dose equivalent I-131. The iodine activities June 6, 1981

' vere _obtained from Chemistry Form F6.2.2 (PCS Activity Analysis) for the week of June 1, 1981 to June 7, 1981. These activities were converted to dose equivalent I-131 and compared to the 1.0 pCi/gm limit. The results are pro-

~

vided in Table 1. Since the actual dose equivalent I-131 is 11 7 times below

. the technical specification limit, all the other nuclides were multiplied by this a= cunt to get the activities _at the limit (1.0 pCi/gm dose equivalent I-131 vill be reached before 100/E).

Table 2 is a. tabulation of the total gaseous activity in the pri=ary coolant

- at _the technical specification limit and the fraction of the design basis quan-tity for each radionuclide. The total fraction of the design basis quantities is from Table 2, 25.8%. Curie limits on the gas tanks should, therefore, not be. imposed.

ACTIVITY IN PALISADES' LIQUID TANKS The same type of methodology is applied to the liquid tanks as the gas tanks.

During the past 10 years of operation of the Palisades facility, the dose equi-valent I-131 concentration in the primary coolant has rarely been above 0.1 pCi/gm (except.during occasional iodine spikes). Only during the past year have ve seen activities approaching 0.1 pCi/gn.

The actual activity concentrations in the June 6, 1981 5 calculation were mul-tiplied by 1.17 to reach the maximum average dose equivalent I-131 concentra-tion of 0.1 pCi/gm. The filtrate and crud activities were compared to the de-sign basis quantity limits and added to give a total fraction of design basis quantity (see Table 3). If the entire primary coolant volume at the maximum average activity were contained in one tank, it would represent 128% of the design basis quantity curie limit.

It is extremely unlikely that one tank would contain this much activity, espe-

, cially considering the removal of activity by filters and demineralizers prior to entry to the tank. This conservative assumption alone vould account for at least a factor of 10 reduction.

In view of the above', curie limits on Palisades' liquid tanks should not be imposed.

,---- . . - , - - - - , , . . _ . , - - , - , ~a,-~.-r -, , , . , . - . , - , , - - , - , ., , - , . . -

.g .

,- ;E-1 i >'

~

TABLE 1 Radionuclide ' Actual Conc. (pCi/ml) 1 ' X DCP (rad /pC1) = DEF (rad /ml)

I-131 0.0338 1.h8 0.'05002 I-132 0.1397 0.0535 0.007h7 I-133 0.1150 0.h' O.oh60 I-134 0.1965- 0.025 0.00h913 I-135 0.1h67 0.12h 0.01819 Dose equivalent I = 0.127 rad /ml Dose Equivalent-I-131 = 0 = 0.0855 pC1/mi Factor to achieve Tech. Spec. limit * *I 0 0 55 Noteg From Chemistry Work Sheet F6.2.2 for June 6, 1981 Dose conversion factors from TID-lh8hh

3) Dose equivalency factors

]

1 i

i i

TABT.E 2 Half Total Radionuclide Actual Conc. (pCi/m 1} I,ife (hr) Activity (Cl) DBQ (Cl)(3) Ci/DBQ ')

Ar h1 2.087E - 2 1.83 1.278E + 0 2,830 h.52E h Kr-85. 5.687E 4 9.343E + h h.60E + 0(5) 19,800 2 33E - h Kr-85m 3.151E - 2 h.h 2 75E + 1 16,600 1.66E - 3 Kr-87 3.8h0E - 2 1.27 3.18E - 1 2,560 1.2hE 14 Kr-88 h.597E - 2 2.8 1 363E + 1 Kr-89 - - -

1,700 8.02E - 3 Xe-131m ' 9 01hE I4 283.2 149,700 5.058E + 9 1.02E h Xe-133 3.638E - 1 126.5 2.76E + 315) 70,900 3.89E - 2 Xe-133m 8.729E - 3 54.2 1

!s.327E + 1 21,800 4

1.ThE - 3 i

Xe-135 1.1h0E - 1 9 1h 2.652E + 2 12,800 2.07E - 2 Xe-135m 1.291sE - 2 0.26 9.65E - 7 27,778(6) _

Xe-137 - - - - -

Xe-138 - - - - -

I-131 1.0E + 0 192 2 957E - 1(7) 1 59- 1.86E - 1 l

i I 2 58E - 1 or 25.8%

i j Notes:

Actual concentrations are from June 6,1981 Palisades E, where the filtrate and crud were added

to get the total.

Total gaseous activity is calculated by the following:

1 Total Activity (ci) = (Actual Conc.,pCi/ml)(10,900 ft )(2.832E 3

+ h ml/ft )(1.6)(11.7)(e-At)(10-6C1/pCi) 3 Where: 13,900 ft 3 = volume of primary ecolant, FSAR, Table 1.1 4 1.6 = correction in total gan activity in primary coolant (see JI.B 82-28, Palinaden-Calculation of Steam Generator Trak Rate),

i i

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'1 2

. TABLE 2 ' (concluded )

Notes:

11 7 = increase to achieve Tech. Spec. limit concentrations (see Table 1).

e-At = decay factor in the Degas process. A = 0.693/Th, t = 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The degassing of the primary system can he done in as short a time as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(3)DBQ = Annual Design Basis Quantities.

(b)Ci/DBQ - Fraction of design basis quantity for each radionuclide.

In addition to the calculation of Note (2) above, the activity in the pressurizer is also added for

, Kr-85 and Xe-133. Only these radionuclides were considered since their abundance and longer half-liven allows the accumulation in the pressurizer. The additional activity was calculated as follows:

Ci = ( Actual Conc.,pci/ml)(10)(700 ft3 )(2.832E + h ml/ft3 )(ll.7)(e -At)(10 6C1/uCi) 1 Where: 10 = gaseous activity in the pressurizer is conservatively assumed to be a j factor of ten higher than in the primary coo 19nt.

i j 700 ft3 = gas volume of the pressurizer FSAR, Table h.8.

l (6)Taken from memo SBB 81-19 f (7)I-131 activity was calculated as follows: I 1

j Ci = (1/1000 pCi/ml maximum)(10,900 ft3 )(2.832E + h ml/ft 3)(e- At)(10-6Ci/pC1) j Where: 1/1000 = the fraction of iodine that evolves from the primary coolant in the de-gassing proccas (partition coefficient).

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w iTABLE 3 FILTRATE (DEGASSED)

Actual ~. Total. -

Activity (Cl)III'

+

Radionuelide - - Cone'. (pCi/ml) DBQ'(Cl)(2) Ci/DBq(3)

F-18 2.909E'- 2 c6.h5E 1 (5) -

Na-2h 2.30TE~- 3: 5.11E - 2 551 5(6) - 9.27E - 5 Co-58 7.725E - 5 1.76E - 3 1190 1.h7E - 61

.Rb-88 - 9.Sh3E - 2 2.11E + 0 1:335 3(6) h.87E Is Rb-89 h.371sE - 2 9.698 - 1~- (5) .-

6.526E - 3 Mo-99 1.lshE - 1 9020 1.60E --5 Tc-99m 2.968E - 3 6.58E - 2 1h5,000 14 53E - T-I-131 1. 0E - 1 1.89E + OII') 30 9 6.12E - 2.

I-132 - -

755 -

I-133 - -

25h -

I- 1 31: .-- --

123h.6(6) _

I-135- .

353 -

Cs-131s 2.01:2E Is b.56E - 3 1.06 1.30E - 3 8

Cs-136 6.505E - 5 1.h0E - 3 7.61- 1.h0E - 2 Cs-137 5 522E 1 1.23E - 2 1.80 6.83E - 3 Cs-138 2.282E - 1 ~ 5.05E + 0 1298.7(6) 3.89E - 3 La-lho 2.829E - 5 '5.85E h 390 1.50E - 6 I 7.82E - 2 CRUD (PARTICULATE)

Cr-51 8.672E h 1 92E - 2 6730 2.85E - 6 Mn-Sh 2.313E - 5 5.12E - h 161 3.18E - 6 Mn-56 2.097E h h.68E - 3 1612 9(6) 2.90E - 6 Fe-59 2.61:0E - 5 5.85E Is 265 2.21E - 6 Co-57 1.122E - 6 2.188E - 5 683.h(6) 3.63E - 8 1.067E - 3 Co-58 2.36E - 2 1190 1.99E - 5 Co-60 2.h81E - 5 5 50E - h 6.67 8.25E - 5

7 g .

2 TABLE 3 (continued)

CRUD (PARTICULATE)

Actual Total

~Radionuclide Conc. (uci/ml) Activity (C1)fl) DBQ (C1)(2) Ci/DBq(3)

Ni-56 2.186E - 6 h.81sE - 5 (5) -

Rb-88 h.81:0E h 1.08E - 2 1335.3(6) 2.h8E - 6 Zr-95 '2.573E - 5 5.70E - h 581 9.80E - 7 Zr-97 2 986E - 5 6.61E - 4 3232.8(6) 2.05E - 7 Nb-95 3 900E - 5 8.63E - h 855 1.01E - 6 Nb-97 4.018E '5 8.89E 14 h316.6(6) 2.06E - 6 Mo-99 h.lsilsE .5 9.77E 14 9020 1.08E Tc-99m h.061E - 5 -9 00E 14 1h5,000 6.20E - 9 Tc-101 - -

(5) -

Ru-103 1.1:20E - 5 3.15E - 4 1091s.9(6)

Sb-122 2.88E - 7 3.777E - 6 8.37E - 5 (5) -

Te-132 2.861E - 5 6.33E - h 29 2 cs-137 2.16E - 5 2.23hE - 6 h.95E - 5 1.8 2.75E - 5 cs-138 6.713E 1: 1.149E - 2 1298.7(6) 1.1hg - 5 Ba-139 14.031sE h 8.898 - 3 614,377.7(6) 1.38E - 7 Io-lho 9.338E - 6 2.07E - h 390 5.31E - 7 ce-lhi 6.858E - 6 1 52E - h 8108.1(6) 1.87E - 8 Ce-lh3 1.887E - 5 h.18E - h (5) -

ce-lbh h.778E - 6 1.06E la 1923.1(6) 5.50E - 8 W-187 h.510E - 5 9.99E h 50 5(6) 1.98E - 5 Re-188 2 5h3E - 5 5.63E - h (5) -

Np-239 6.311E - 5 1.h0E - 3 3750 3.71sE - 7 Ni-65 9.763E - 5 2.11E - 3 51518.6(6) h.08E - 7 3r-85 9.801E - 8 2.16E - 6 (5) -

I 2.03E Is

[aI$ n i e + crud) = (7.82E - 2) + (2.03E h) = 7.8hE - 2 ror 5000 gallonn Fraction of Design Basis Quantity" 3.09E + 8 for entire primary coolant volume * -

1.8926E + 7,

= 1. + 0 or 12M

3

'3 1

TABLE 3 (concluded)

Notes:

1 Total activity is calculated as follows:

Ci = (Actual Cone. pCi/ml)(1.8926E + 7 ml)(1.17)(10-6C1/pC1)

Where: 1.8926E + 7 = m1 in 5000 gallons 1.17 = increase in activities to reach the maximum average of 0.1 pCi/gm dose equivalent I-131 DBQ = Design Basis Quantity Ci/DBQ = Fraction of design basis for a 5000 gallon volume. The total primary coolant volume is 3.09E + 8 ml and the total fraction of design basis quantity for the primary coolant in 1.28, as listed above.

This quantity is for dose equivalent I-131 which considers the other iodine radionuclides. The maximum average of 0.1 pCi/gm is assumed.

Design Basis Quantities were not found for these radionuclides, however, their contribution is ex-pected to be insignificant.

Design Basis Quantities are from memo SBB 81-15, 1

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APPENDIX 4 b.

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L-E y

1 TS0784-0031A-NLO2 rn

I I i

DES!G:i CALCULATIOI CHECKLIST EIF DCCCT T CC!a EITS The design calculation input shall include ite=s such as:

A. Basic functions of each structure, h//p/70s) syste= and ce=ponent anc,/or purpose g of calculation referencing same. /

/)soso+ n ts B. Perfor=ance require =ents for:

EYM &s ity iod

1. system, or /14 Tf D (,[4/f/

2. calculation C. Docu=ents 1.

/dfid Le o gg go, $~gw axsaa44 codes '

2. standards g ,, g Esc rie 4 e7 /TLPonts fcA' 3 regulatory require =ents 73,p /7,g y /3,, a 7, ggg D. Design or hypothetical conditions: Arp 1

1. FSAR description /T40s,4,#ient "UN # **

2. O!JiMiY/tf,,,

gpn gg 4 dy Npri Regulatory Guide //tsecid //soci7t< y,19 70

3. Standarc Reviev Plan yyy,g

k. Other 5d/*1, fBBlot -03 wD A+3e . .s asymi-nnt- 7o E. Mitigating factors fi-/4 o d PnJ/ W llNb OWYdD ////f/7f 84 tis &ne'7/7//S W sksstrW /Mf/S

1. tackup systems gues-re r/(s.

2. operator action 3 procedural requirements 7t<# Js'st ) PIM /*'#4'd4 /t'M M

k. Tech Spec require =ents po p rs end ~

5 ce==it=ents //tstn3 /s/ ~7 / 2 y,yge4n y ge es.xwy g,, ,p .yg F. Environmental factors if/7/V/Yf W M1~

?4 stsdalf

1. pressure, te=n, humidity

2. corrosion

3. meteorology

k. dilution 5 concentration factors

6. duration of exposure

__ _ _ ___ _ - . . . _ _ _ . _ . - . - _ _