ML19317H189

In-Plant Source Term Measurements at Fort Calhoun Station- Unit 1.

ML19317H189
Person / Time
Site:	Fort Calhoun
Issue date:	07/31/1978
From:	Bunting R, Croney S, Dyer N EG&G, INC.
To:
References
NUREG-CR-0140, NUREG-CR-140, NUDOCS 8004210731
Download: ML19317H189 (203)

Text

NUREG/CR-0140

]N-PLANT SOURCE TERM MEASUREMENTS AT FORT CALHOUN STATION - UNIT 1 N. C. Dyer and Others EG Et G Idaho, Inc.

1 Prepared for U. S. Nuclear Regulatory Commission 8004210 7 3 /

f NOTICE This, report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Nuclear Regulatory Commission, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, norassumes any legalliability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, pro-duct or process disclosed, nor represents that its use would not infringe privately owned rights.

Available from National Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $9.25 ; Microfiche $3.00 The price of this document for requesters outside of the North 'American Continent can be obtained from the National Technical Infonnation Service.

NUREG/CR-0140 RR IN-PLANT SOURCE TERM MEASUREMENTS AT FORT CALHOUN STATION - UNIT 1 Principal Investigators N. C. Dyer

• J. H. Keller" R. L. Bunting

• B. G. Motes *

• S. T. Croney* D. W. Akers" C. V. Mcisaac* T.E.Cox**

R. L. Kynaston* S. W. Duce" D. R. Underwood* J. W. Tkachyk'*

t Manuscript Completed: April 1978 Date Published: July 1978 EG&G Idaho, Inc.*

Allied Chemical Corp.**

Idaho National Engineering Laboratory Idaho Falls, ID 83401 Division of Safeguards, Fuel Cycles and Environmental Research Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Under DOE Contract No. EY-76-C-07-1570

ABSTRACT This report presents data obtained from an in-plant source term measurement program conducted for the Office of Nuclear Regulatory Research in support of the Effluent Treatment Systems Branch of the Office of Nuclear Reactor Regulation. The objective of this orogram is to provide operational data that can be used in the generic evaluation of plant system design in the licensi ng process and for updating of the calculational models used by the (RC staff in their evaluation of radio-active waste management systems for operating pressurized water reactors.

, A data base is provided for radioisotope inventory in plant systems, radioactive waste management system performance, and source terms for both liquid and gaseous systens.

Data presented were obtained at the Fort Calhoun Statica - Unit 1 operated by the Omaha Public Power District (OPPD). located at Blair.

Nebraska. In-plant measurements wem conducted during the time period from August,1976 through February 1977. This plant is the first of a planned series of six (6) operating PWR's to be studied.

1 1

_ _ __ - _____ s

ACKNOWLEDGEMENTS 4

- The valuable assistance of OPPD personnel made possible this measure-ment program at Fort Calhoun Station, especially W. Dennyer, F. Franco, B. Nicholas, and M. Cassidy of the station personnel. Also, from the OPPD corporate offices, J. Keuchel, T. Harding and their staffs provided valuable assistance. The financial support and direction of the U. S. J Nuclear Regulatory Commission from C. Bartlett and D. Solberg of RES ,

and J. Collins, M. Bell and J. Lee.of ETSB made the measurement program 9 at Fort Calhoun possible.

a k

l

. 11-(

TABLE OF CONTENTS Paae Abstract ........................-..... 1

[

Acknowledgements ........................11 k- 1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . 1

( 1.1 In-Plant Measurement Program Objectives . . . . . . . . . 1 1.'2 Fort Calhoun Station .................. 1

2. Measurement Methods at Fort Calhoun ............. 3 2.1 Introduction ...................... 3 2.2 Liquid Streams ..................... 3 2.2.1 Reactor Coolant and CVCS Samples . . . . . . . . . 8 2.2.2 Steam Generator Samples ............. 10 2.2.3 Spent Fuel Pool and Associated Samples ..................... 10 i

2.2.4 Radwaste Samples . . . . . . . . . . . . . . . . . 10 2.3 Gaseous Streams . . . . . . . . . . . . . . . . . . . . . 16 2.3.1 Ventilation Sampling . . . . . . . , . . . . . . . 16 2.3.2 Waste Gas Processing Sampling. . . . . . . . . . . 22 2.3.3 Containment Atmnsphere Sampling ......... 23 2.4 Plant Operational Data ................. 23  !

3. Di s cus si on of Da ta . . . . . . . . . . . . . . . . . . . . . . 25 i

5 3.1 Introduction ...................... 25 3.2 Thermal Power Level . . . . . . . . . . . . . . . . . . . 25 3.3 Plant Capaci ty Factor . . . . . . . . . . . . . . . . . . 25 3.4 Radionuclide Concentrations Reactor Liquid System Conponents ............ 25 3.4.1 Predicted Concentrations in Reactor Coolant and Secondary Water ............... 25

'iii t e

/

f o-.

TABLE OF C0tiTENTS Pace 3.4.2 ' Measured Reactor Coolant Radionuclide Concentrations . . . . . . . . . . . . . . . . . 30 3.4.3 Radionuclide Concentrations in Steam Generator Fater ................ 34 3.4.4 Concentrations of Radionuclides in Spent Fuel Pool i and Refueling Associated Waters ........ 34 1 3.4.5 Radionuclide Concentrations in Reactor Coolant Drain Tank . . . . . . . . . . . . . . . . . . . 39 3.4.6 Radionuclide Concentrations in Haste Holdup Tanks ..................... 39 3.4.7 Radionuclide Concentrations in Spent Regenerant Tanks . . . . . . . . . . . . . . . . 42 3.4.8 Radionuclide Concentrations in Hotel Waste Tanks ...'.................. 42 3.4.9 Radionuclide Concentrations in Monitor Tanks . . 42 3.5 Auxiliary Building Ventilation System Source Terms .. 46 3.6 Gaseous Iodine Species ................ 51 3.6.1 Auxiliary Building Ventilation Iodine Species . 51 3.6.2 Waste Gas Processing System Chemical Species of Radioiodine .................. 51 3.7 Liquid Waste Flow Rates . . . . . . . . . . . . . . . . 51 3.8 Detergant Wastes . . . . . . . . . . . . . . . . . . . . 51 3.9 Chemical Wastes from Regeneration of Condensate Demineralizers .................... 54

(

3.10 Containment Purge Frequency . . . . . . . . . . . . . . 54 3.11 Containment Internal Cleanup System . . . . . . . . . . 54 3.12 Gaseous Leakage Rate to Containment Building ..... 54 3.13 Auxiliary Building Gaseous Leakage .......... 56

, iv N

TABLE OF CONTENTS

-Page 3.14 Particulate Releases for Gaseous Effluents . . . . . . 57 3.14.1 Containment Building Radioactive Particulate l . Releases for Gaseous Effluents . . . . . . . . 57 3.14.2 < Auxiliary Building Particulate Releases for Gasecus Effluents ' . . . . . . . . . . . . 57

{

3.14.3 Gas Decay Tank Radioactive Particulate Releases .for Gaseous Effluents . . . . . . . . 59 3.15 Tritium and Carbon-14 Releases . . . . . . . . . . . . 60 ..

3.15.1 Liquid Tritium Releases . . . . . . . . . . . . 60 3.15.2 Liquid Carbon-14 Releases . . . . . . . . . . . 60 3.15.3 Containment Building Gaseous Tritium and Carbon-14 Releases . . . . . . . . . . . . . . 60 3.15.A Auxiliary Building Gaseous Tritium and Carbon-14 Releases . . . . . . . . . . . . . . 60 3.15.5 Gas Decay Tank Tritium and Carbon-la Gaseous Releases . . . . . . . . . . . . . . . 62

~

3.16 Decontamination Factors for Demineralizers . . . . . . 62 3.17 Decontamination Factors for Liquid Stream Filters . . . 64 3.18 Decontamination Factor for Charcoal Adsorbers and HEPA Filters . . . . . . . . . . . . . . . . . . . . . 66

'3.19 Decontamination Factors for Evaporators . . . . . . . . 66

/ References ... . . . . . . . . . . . . . . . . . . . . . . . . . 72 l

-Appendix

A.1 I n t roduc ti o n . . . . . . . . . . . . . . . . . . . . . 73 lA.2 Reactor Power Level .................73

'A.3 -Liquid Samples . . . . . . . . . . . . . . . . . . . . .~73

'A.3.1- Reactor Coolant . . . . . . . . . . . . . . . . 73 i 1 l

'l

.. V '

i

i TABLE OF CONTENTS Pace

'A.3.2 CVCS Liquid . . . . . . . . . . . . . . . . . . 74 A.3.3 Secondary Liquid ............... 75 A.3.4 Spent Fuel Pool and Fuel Transfer Associated Liquids . . . ... . . . . . . . . . . . . . . . 75 A.3.5 Liquid Radwaste System ............ 76 A.3.5.1 Reactor Coolant Drain Tank . . . . . . 76 A.3.5.2 Waste Holdup Tanks . . . . . . . . . . 77 A.3.5.3 Spent Regenerant Tanks . . . . . . . . 77 A.3.5.4 Hotel Waste Tanks .......... 77 A.3.5.5 Monitor Tanks ............ 77 A.3.5.6 Waste Evaporator . . . . . . . . . . . 77 A.4 Gaseous Samples . . . . . . . . . . . . . . . . . . . . 78

\

(

vi l

LIST OF TABLES Page 1.1 Time Periods for In-Plant Measurements . . . . . . . . . . 2 2.1 Chemical and Volume Control System Components ...... 9 2.2a Component Design Data. Waste Disposal System Tanks . . . . 11

, 2.2b Component Design Data. Waste Disposal System Pumps . . . . 12

2.2c Component Desen Data. Waste Disposal System Process Equipment .................... 14 2.3 Auxiliary Building Ventilation Duct Flow Measurements .. 19 2.4 Auxiliary Building Sampling Station Feeds ........ 21 3.1 Predicted Concentrations in Reactor Coolant and Secondary Water .......................... 26 3.2 Parameter Values Used to Modify N-237 Activities for For t Cal houn . . . . . . . . . . . . . . . . . . . . . . . 28 3.3 Radionuclide Concentrations in Reactor Coolant (Reactor Power Operations) .................... 31 3.4 Radionuclide Concentrations in Reactor Coolant (Reactor R e fue l i n g ) . . . . . . . . . . . . . . . . . . . . . . . . 33 3.5 Radionuclide Concentrations in Steam Generator Water . . . 35 3.6 Radionuclide Concentrations in Spent Fuel Pool (Reactor Power Operations) .................... 37 3.7 Radionuclide Concentrations in Spent Fuel Pool (Reactor Refueling) . . . . . . . . . . . . . . . . . . . . . . . . 38 3.8 Tritium Balance During Refueling . . . . . . . . . . . . . 40 l

, 3.9 Radionuclide Concentrations in Waste Holdup Tanks .... 41 3.10 Radionuclide Concentrations in Spent Regenerant Tanks .. 43 i 3.11 Radionuclide Concentrations in Hotel Waste Tanks . . . . . 45 3.12 Radionuclide Concentrations in Nonitor Tanks . . . . . . . 47 3.13 Ventilation Airborne Activities ............. 49 3.14 Ventilation 131I Airborne Activities . . . . . . . . . . . 50 3.15 Average Tractional Percentage for 131I Species . . . . . . 52 3.16 Liquid Waste Flow Rates ................. 53 3.17 Extrapolated Annual Containment Building Particulate Releases for Gaseous Effluents . . . . . . . . . . . . . . 58 vii l

1

LIST OF TABLES Page 3.18 Exteapolated Annual Auxiliary Building Particulate Releases for Gaseous Effluents . . . . . . . . . . . . . . 58 3.19 Extrapolated Annual Waste Gas Decay Tank Particulate Releases for Gaseous Effluents . . ............. 59 3.20 Extrapolated Annual Releases of Gaseous Tritium and Carbon-14 ...................... 61 3.21 Decontamination Factors, CVCS Purification Demineralizer ...................... 63 3.22 Decontamination Factors, CVCS Purification Filters ......................... 65 3.23 Fourth Quarter 1976137Cs Auxiliary Building Release Data . . . . . . . . . . . . . . . . . . . . . . . . . . . 67

, 3.24 Evaporator Decontamination Factors, "Old" Tube Bundle .. 69 3.25 Evaporator Decontamination Factors, "New" Tube Bundle .. 71 A.1 Reactor Coolant Activities, Power Operations - Before Refueling ........................ 80 .

A.2 Beta, Only Emitting Radionuclide Activities, Power Operations - Before Refueling .............. 82 A.3 Reactor Coolant Activities, Power Operations - After Refueling ........................ 83 A.4 Reactor Coolant Activities, During Refueling . . . . . . . 85 A.5 Reactor Coolant Activities (OPPD Degassed Sam Operations - Before Refueling . . . . . . . . ple) Power

....... 86 A.6 Reactor Coolant Activities (OPPD Degassed Sample). Power Ope ra t i o r.s - Af te r Re fuel ing . . . . . . . . . . . . . . . 88 c A.7 Reactor Coolant Gaseous Activities, Power Oparations -

After Refueling ..................... 90 l A.8 CVCS Letdown Activities - 8/20/76, Power Operations -

Be fo re Refuel i ng . . . . . . . . . . . . . . . . . . . . . 91 l A.9 - CVCS Letdan Activities - 8/30/76, Power Operations -

l Be fo re Re fuel i ng . . . . . . . . . . . . . . . . . . . . . 92 viii

LIST OF TABLES Page A.10 CVCS Letdown Activities, During Refueling . . . . . . . . . . 93 A.11 Demineralizer Function: Refueling Cavity Water, During Refueling . . . . . . . . . . . . . . . . . . . . . . . . . . 94 A.12 CVCS Letdown Activities, Power Operations - After Refueling . . . . . . . .*. . . . . . . . . . . . . . . . . . 95 A.13 Steam Generator Blowdown Activitf as . . . . . . . . . . . . . 97 A.14 Spent Fuel Pool Activities, Power Operations - Before Re fu el i ng . . . . . . . . . . . . . . . . . . . . . . . . . . ' 99 A.15 Safety Injection and Refueling Water Tank Activities, During Refueling . . . . . . . . . . . . . . . . . . . . . 10 0 A.16 Fuel Transfer Canal and Refueling Cavity Activities. l 1

During Refueling . . . . . . . . . . . . . . . . . . . . . 1 01 A.17 Spent Fuel Pool Activities During Refueling . . . . . . . 102 A.18 Spent Fuel Pool Activities, Power Operations - After l Re fuel i ng . . . . . . . . . . . . . . . . . . . . . . . . . 103 l

A.19 Reactor Coolant nrain Tank Activities (Tank WD-1), Power Operations - After Refueling . . . . . . . . . . . . . . . 10 4 A.20 Spent Regenerant and Waste Holdup Tanks Power Operations -

Before Refueling .....................105 A.21 Liquid Radwaste Tank Activities, During Refueling . . . . . 107 l

A.22 Waste Holdup Tank Activities (Tank WD-4A), Power Operations -

After Refuel ing . . . . . . . . . . . . . . . . . . . . . . 108 A.23 Spent Regenerant Tank Activities, Power Operations - After Re fu el i ng . . . . . . . . . . . . . . . . . . . . . . . . . . 109 A.24 Monitor and Hotel Tanks, Power Operations - Before j Re f uel i ng . . . . . . . . . . . . . . . . . . . . . . . . . 1 10

) A.24A Monitor Tank Activities, Comparison Between INEL and OPPD Measurements . . . . . . . . . . . . . . . . . . . . . . . 112 i A.25 Hotel Tank Activities Power Operations - After Refueling . 114 A,26 Monitor Tank Activities, Power Operations - After Re fuel i ng . . . . . . . . . . . . . . . . . . . . . . . . . 115 A.27 Monitor Tank WD-223 Activities with Varying Tank Level, Power Operations - After Refueling . . . . . . . . . . . . 1 16 A.28' Evaporator Processing Waste Holdup Tank WD-4B . . . . . . . 117 A.29 Evaporator Processing Spent Regenerant Tank WD-13A . . . . 118 iX

LIST OF TABLES Page A.30 Evaporator Processing Spent Regenerant Tank WD-13B . . . . 119 A.31 Scrapings from Radwaste Evaporator Bundle. During Refueling ........................120 A.32 Evaporator Function Test, Feed Samples: Spent Regenerant I Tank WD-13A . . . . . . . . . . . . . . . . . . . . . . . 1 21 A.33 Evaporator Function Test. Evaporator Distillate Samples

, (Feed: Spent Regenerant Tank WD-13A) . . . . . . . . . . . 123 A.34 Evaporator Function Test, Evaporator Concentrate Samples . 125 A.35 Ventilation Airborne Activities. Sample Station #1. . . . 127 l A.36 Ventilation Airborne A::tivities Sample Station #2 . . . . 131 1 1

j _ A.37 Ventilation Airborne Activities, Sample Station #3 . . . . 135 A.38 Ventilation Airborne Activities Sample Station #4 . . . . 139 A.39 Ventilation Airborne Activities, Sample Station Waste Eva pora tor . . . . . . . . . . . . . . . . . . . . . . . . 143 l A.40 Ventilation Airborne Activities, Pipe Penetration Room . . 147 A.41 Ventilation Airborne Activities. Letdown Heat

~ Exchanger Room . . . . . . . . . . . . . . . . . . . . . . 148 A.42 131I Species Data Sample Station #2 . . . . . . . . . . . 149 A.43 131I Species Data Waste Evaporator Room . . . . . . . . . 151 A.44 Ventilation Iodine Species Comparisons . . . . . . . . . . 153 A.45 131I Species Measurements of Process and Cover Gas, Power Operations - Before Refueling . . . . . . . . . . . . . . 15 4 c

A.46 131i Species Measurements of Process and Cover Gas. During Refueling and Power Operations After Refueling . . . . . . 155 j A.47 Waste Gas Decay Tank "B" Activities . . . . . . . . . . . 15 6

! A.48 Waste Gas Decay Tank "A" Activities . . . . . . . . . . . 15 7 l

A.49 Process Gas 14C and 3H Measurements.

l Du r i ng Re fuel i ng . . . . . . . . . . . . . . . . . . . . . 158 A.50 Process and Cover Gas 14C and 3H Activities, Power L Operations - After Refueling . . . . . . . . . . . . . . . 159 A.51 Containment Activities . . . . . . . . . . . . . . . . . . 160 x

L

LIST OF FIGURES

, f_aat 2.0 Liquid and Solid Systems . . . . . . . . . . . . . . . . . . 4 2.1 Reactor Coolant'and CVCS Diagram . . . . . . . . . . . . . . 5 2.2 Liquid Radwaste System Collection Tanks .......... 6

2. 3 -Liquid Radwaste Process Diagram .............. 7 2,4a Gaseous Waste Disposal System ...............

17-s

, 2.4b Fort Calhoun Station Auxiliary Building Ventilation System . 18 A.1 Reactor Power Level: Fort Calhoun Station, Period: 8/15 - 10/3/76 . . . . . . . . . . . . . . . . . . . 161 A.2 Reactor Power Level: Fort Calhoun Station, Period: 12/15/76 - 2/2/77 ................. 162 A.3 Reactor Coolant System P and I Diagram . . . . . . . . . . . 164 A.4 Chemical and Volume Control System P and I Diagram . . . . . 165 A.5 CVCS Demineralizers .................... 166 A6 Letdown Flow Rate: Power Operations - Before Refueling, Period: 8/15 - 9/9/76 ................... 167 A.7 CVCS Letdown Parameters, Period: 8/9 - 9/3/76 ....... 168 A.8 Letdown Flow Rate: During Refueling, Period: 9/25 -

10/19/76 . . . . . . . . . . . . . . . . . . . . . . . . . . 169 A.9 Letdown Flow Rate: Power Operations - After Refueling.

Period: 1/ 31 - 2/24/ 77 . . . . . . . . . . . . . . . . . . . 170 A.10 CVCS Letdown Parameters, Period: 1/31 - 2/24/77 ...... 171 A.11 Steam Generator Flows and Blowdown Rates: Power Operations-

)

After Refueling, Period: 1/31 - 2/24/77 .......... 172 A.12 Spent Fuel Pool Cooling System Flow Diagram .. ..... 173 A.13 Waste Disposal System Flow Diagram . . . . . . . . . . . . . 174 A.14 Wast.e Disposal System Flow Diagram, Supplement No. 8 . . . . 175 A.15 Waste Disposal System Flow Diagram . . . . . . . . . . . . . 176

+

A.16 - Waste Disposal System Flow Diagram . . . . . . . . . . . . . 177 A.17. Level of Waste Holdup Tanks: Bafore Refueling . . . . . . . 178 A.18 Level of Waste Holdup Tanks: During Refueling . . . . . . . 179 xi L.... ..

LIST OF FIGURES Page A.19 Level of Waste Holdup Tanks: After Refueling . . . . . . . 180 A.20 Level of Radwaste-Tanks: Before Refueling . . . . . . . . 181 A.21 Level of Radwaste Tanks: During Refueling . . . . . . . . 182 A.22 Level of Radwaste Tanks: After Refueling . . . . . . . . . 183 J A.23 Radwaste Evaporator Parameters, Period: 8/9 4/2/76 . . . 184 A.24 Radwaste Evaporator Parameters, Period: 00:00, 8/24/76 through 00:00,8/28/76 . . . . . . . . . . . . . . . . . . . 1 85 A.25 Radwaste Evaporator Parameters, Period: 00:00, 8/28/76 through 00:00, 9/1/76 . . . . . . . . . . . . . . . . . . . 186 A.26 Radwaste Evaporator Parameters, Period: 1/31 - 2/21/77 . . 187 A.27 Radwaste Evaporator Paraneters, Period: 00:00,2/17/77 through 00:00,2/20/77 . . . . . . . . . . . . . . . . . . 1 88 A.28 Auxiliary Building Heating and Ventilating Flow Diagram . . 189 A.29 As Built Ventilating Flow System . . . . . . . . . . . . . 19 0 h

Xii

'"-PLANT SOURCE TERM E ASUREMENTS AT FORT CALHOUN STATION - UNIT 1 i

l CHAPTER 1 - INTRODUCTION l

\

1.1 In-Plant Measurement Program Objectives The objective of the in-plant source term measurement study is to provide the Nuclear Regulatory Commission (NRC) with experimental data that can be used by the NRC in its evaluaticq of plant system design.

The results of this sady will provide a data base for radioisotope inventory in plant systus, radioactive waste treatment system performance, and source _ tems for both liquid and gaseous systems.

The in-plant measurement study identifies and characterizes the sources of radioisotopes in operating LWR's. Specific objectives are to provide NRC with:

1. Source term infomation so that the parameters used in NRC's calculational models can be updated as necessary.

2. Data on the inventory of the radioisotopes present (i.e.,

locations, concentrations, etc.) in operating reactor  !

n1?nt systems. )

3. Radwaste equipment perfomance for use in NRC evaluations of radioactive waste management systems.

Measurements are made during the three stages of plant operation, i.e.,

(1) prior to refueling, (2) during refueling, and (3) following refueling.

1.2 Fort Calhoun Station The NRC in-plant source term measurement program at operating PWR's was initiated during the summer of 1976, From August,1976 through February,1977, measurements were made at the Fort Calhoun ' Station - Unit 1, l Omaha Public Power District (0 PPD),' Blair, Nebraska. Table 1.1 presents '

the operational status and the time span for measurement periods curing the time between the reactor's first refueling in the spring of 1975 until

~

the third refueling in the fall of 1977.

i 1

' Relevant system parameters for the Fort Calhoun Station are listed below:

1. Reactor and Steam Generator Vendor: Combustion-Engineering

2. Turbine Vendor: - General Electric

3. Power: 1420 MWt and 457 MWe, net

4. Reactor Coolant Volume: 6616 ft3

5. Initial Criticality: August,1973

6. Commercial Power Operations: June,1974 TABLE 1.1 TIME PERIODS FOR IN-PLANT MEASUREMENTS Plant Status Time Period Measurement Period Power Operation 5/75 - 10/1/76 Iiquid: 8/18 - 9/2/76 Gaseous: 9/3 - 10/7/76 Second Refueling 10/1 - 12/15/76 Liquid: 10/11 - 12/3/76 Gaseous: 10/7 - 12/15/76 Power Operation - 12/15/76 - 9/30/77 Liquid: 2/9 - 2/22/77 Gaseous: 12/15/76 - 2/17/77 i

J l

CHAPTER 2 - MEASUREMENT METHODS AT FORT CALHOUN 2.1 Introduction The measurement methods employed have been described.in the report,

" Procedures, Source Tern Measurement Program" (1). Brief descriptions are given in the following three sections which refer to the liquid process streams, gaseous process streSes and the plant operational information.

2.2 Liquid Streams

/

Liquid samples were collected using the installed planc sampling sys tem. The types of liquid samples collected were:

1. Reactor coolant

2. Chemical and Volume Control System (CVCS) samples both upstream and downstream of the letdown demineralizers and filter

3. Steam generator blowdown water

4. Spent fuel pool water

5. Radwaste system:

a. Collection tanks

b. Evaporator feed, distillate and concentrate

c. Monitor tanks Figures 2.0, 2.1, 2.2, and 2.3 present simplified schematic diagrams for the liquid and solid pathways at Fort Calhoun Station. For more detailed diagrams, the Piping and Instrument Diagrams (P & ID) have been presented in the Appendix.

All liquid samples except those passed through and concentrated on ion exchange resins were collected in plastic volumetric measuring devices and promptly transferred to glass counting containers. No

, measurements were performed to quantify potential plateout of radionuclides on the plastic volumetric measuring devices. The size of the liquid samples collected depended upon the expected radionuclide activity level.

Fif ty mi samples were collected from reactor coolant systems while

> 450 mi samples were generally taken from other systems. Each glass counting container h2d been prefilled with enough concentrated hcl to provide a solution containing two percent concentrated hcl to minimize plate-out of radionuclides on the glass. Samples from very low activity streams were concentrated on an ion exchange column by passing 20-100 liters of sample through the column (1). Again, no measurements were made to quantify potential plcteout on the aluminum resin holder or silicone tubing sampler feed lines.

e l~ 3

FIGl"E 2.0 Liquid and Soli

Deborating demineralizer c U '

sS n

regenerant ' -

Spent regenerant u " "E

-=l Gas stripperl-- E =

Auxiliary building floors- tanks (2) nks 2) $g Containment sump  :

3 g

Spent resin sluice water *. E Condensate f p o.

y Off standard ik Laundry  ;~

Hotel waste E ->l Evaporator l 1 recycle to -g Shower Hand sink  ;.

tanks (2) h Concdnt ate spent regenerant m:

t tank Radiation O Concentrate monitor tanks (2)

Waste demineralizer .

and filter 4 Spent cartridges Solid waste .--Misc solid waste drumming intake ,

Spent resins  : resin  : -

structure facility --o Shipmentoff-site tank '

~Discharg e

--4 yvet*

g structure i

- w: < .

%  % a___ -,

FIGURE 2.1 REACTOR COOLANT AND CVCS DIAGRAM Steam ' '. Steam gen. RC-7 RC-6 l gen.

8 T* Reactor I A

• • . vessel 4 RC-9 % + + J RC-8 y 1r Letd wn heat exchangers and a

. I Ig 5 I ui . pressure I bl I h s .

! Y  ! h reducers g CH-8B CH-8A CH-10 CH-98 C H-9A 9 l' h , h b RC-1

(@ 1r Purification I Cation i Deborating i demineralizers 8 demineralizers 4 demineralizer 4 RC-2 RC-3 RC-4

'i

'8 .it.

- CH-17A&B :

$  !." i

=CH-14, 3 _

• Sample points RC-5 Purification 3 filters Return to

, reactor loops Volume control tank INEL-A-8009

FIGURE 2.2 LIQUID RADWASTE SYSTEM COLLECTION TANKS Pressurizer quench tank drains, Control element drive mechanism leaksc Reactor coolant pump seat leakage, Volume control tank relief Reactor coolant loop drains, and drains, Safety injection system drains, Auxiliary building CVCS bleed for boron Refueling pool drains. quipment drain control.

=-

Reactor Auxiliary A ,

coolant building ;y Neutralization WD-7 drain sump , tank tank tank j

%  % l9  %

, RW-7 4 ,

_- v Sample points i j i WD-4A WD-4 WD-4C Auxiliary building floor drains and l *4 RW-3 other drain headers (including decon)

[

9 Steam generator blowdown t u _

Liquid and secondary side drains I Waste holdup tanks radwaste (if radioactive), I \ Spent resin 4 treatment Containment sump. \ sluice water f Shower and hand sink drains, inlet Chemical labor $ tory y , Laundry drains. header Y  % V  ;

'L J6

. . >4 RW-1 WD-13 WD-13 J

Spent regenerant tanks YYHotel waste tanks

~ n _____-

FIGURE 2.3 LIQUID RADWASTE PROCESS DIAGRAM Waste filters N To waste g--

Liquid radwaste +A WD- ,A W D-gas decay tanks WD-19 3, -

treatment inlet , 17A , 17B j Gas e WD-21 v >$ RW-5 header v l stnpper I Waste I 1r 'f T 8

evaporator 8

i j '

Q- - --.s Treatment l _______] 78 RW-6 system g j bypass 3 (usually hotel wastes) i WD- WD-I l Concentrate tanks 38A 388 g

8 i V V I

I l t To drumming g

station 3r { g i

_e I f m St Spent regenerant f _ , ~ J_^ - - "* , , ,

7

\eam generator blowdown tanks KWD- gS gg g WD-24Al 24B f a H RW.4 WD-22A l WD-22B Waste 8 Sams,le points demineralizers ( Monitor tanks

(/

t Missouri River , , e ,, ,,

discharge header INEL-A-8011

Gamma-ray spectra from the liquid samples were obtained using a Ge(Li) gamma-ray spectrometer installed in the NRC mobile laboratory (1).

Samples were counted several times over an extended period after collection to optimize detemination of both short- and long-lived radioisotopes. The detection efficiency for the gamma-ray detector was measured using NBS reference standards prior to moving the mobile laboratory to Fort Calhoun (1). During the initial set-up at Fort Calhoun, the calibration was re-checked for selected geometries using an NBS standard. Gamma-ray energy calibrations were made on a daily basis using a 22aTh source. The same reference standard was measured j in a fixed geometry to verify overall perfomance of the spectroneeter system.

The analysis of gamma-ray spectra was completed at INEL using the GAUSS VI computer program (2) on the IBM 360/75. This program searches each spectrum for gamma-ray peaks, provides isotopic identification and concentration and decay corrects to the sample collection time. A final output for the activity of each radionuclide found in the sample in uCi/mi is provided together with error estimates. Uncertainties quoted for radionuclide concentrations are errors due to counting statistics. An additional uncertainty of approximately 10 percent should be added to the quoted errors to account for calibration (-10%) and volume measurement

(-5%) errors. Indeteminate sampling errors have not been treated.

Liquid samples were collected of reactor coolant, spent fuel pool, and the radwaste monitor tank waters to detemine concentrations of pure beta-emitting radionuclides. Subsequent analyses of the samples were perfomed by the Department of Energy, Idaho Operations Office, Health Services Laboratory (D0E-HSL). Summaries of the analysis methods may be found in reference (1). Initially, samples (Tables A.2 and A.4) were analyzed for 3H,14C, 32p, 35S, 45Ca, ssFe, 6 3Ni, 893r, 90$r, 91Y, and 147Pm. The decision was made not to analyze the latter samples for 32p, 353, usCa. and 147Pm. These data are presented in Tables A.2 A.3, and A.26.

2.2.1 Reactor Coolant and CVCS Samples Figure 2.1 shows a simplified schematic diagram for the c Reactor Coolant (RC) and the CVCS. Table 2.1 presents specifications for CVCS components obtained from the Final Safety Analysis Report (FSAR) (3).

The RC samples were taken from points RC-6 and RC-7 wh' h are in the two reactor coolant locps. There was no observed difference in the measured c activity levels between the two loops. The input sample for the CVCS was taken at point RC-1. The output sample of the purification, cation, and deborating demineralizers were taken at points RC-2, RC-3, and RC-4, respectively, depending upon which C)CS demineralizer was in operation at the time of sampling. The cutput sample for the CVCS purification filters was taken at point RC-5. There was no sample point at the liquid output from the volume control tank. Results of measurements for the CVCS system are given in Tables A.8-A.12. Radioisotope data on reactor coolant (including pure beta-emitting nuclides) are presented in Tables A.1-A.7. ,

8

TABLE 2.1 CHEMICAL AND VOLUME CONTROL SYSTEM r0NPONENTS ION EXCHANGERS Item No's. CH-8A, 8B, 9A, 9B & 10 Quantity 5 Type Flushable (CH-9A, and CH-9B renewable)

Design Pressure, psig 200 Design Temperature, F 250 Normal Operating Pressure, psig 25 Normal Operating Temperature, F 120 Resin Volume, ft3 20 Resin Cross-sectional Area, ft2 6.8 Norwel Flow Rate, gpm 36 Maximum Flow Rate, gpm 116 Retention Screen 80 US Mesh Code for Vessel ASME III Class C Material Austenitic Stainless Steel Fluid 1 wt % Boric Acid, Maximum PURIFICATION FILTERS Item No's. CH-17A & 178 Quantity 2 Type of Elements Wound Cartridge Retention for 1 Micron Particle, % 95 Design Pressure, psig 200 Design Temperature, F 250 Design Flow, gpm 116 Normal Flow, gpm 36 Maximum Flow, gpm 160 Code for Vessel ASME III, Class 0 Material Austenitic Stainless Steel Fluid 1 wt % Boric Acid Maximum VOLUME CONTROL TANK Item No. CH-14 Quantity 1 Type Vertical, Cylindrical

> Design Pressure Internal, psig 75 Design Pressure. External, psig 15 Design Temperature F 250 Internal Volume Minimum, ft3 383 Operating Pressure Range, psig 0 to 65 Norwal Operating Pressure, psig 50 Normal Operating Temperature, F 120 Norwel Spray Flow, gpm 36 Blanket Gas Hydrogen Code ASME III, Class C Fluid 6-1/4 wt% Boric Acid, Maximum Material Austenitic St inless Steel o

9

2.2.2 Steam Generator Samples Steam generator water for both steam generators (sample points RC-8 and RC-9 in Figure 2.1) were sampled both before (8/26/76) and af ter refueling (2/8-10/77). Based on the 1311 activity levels observed from nominal 200 liter ion exchange resin samples (Table A.13) and a blowdown rate of 12,500 lbs/hr, the primary to secondary leak rate during the February 1977 in-plant measurement program at the Fort Calhoun Station s was less than 10-2 lbs/ day. Also, the samples taken in August 1976 indicate no meaningful primary to secondary leak. Since there was no '

meaningful primary to secondary leak during either sampiing period, the rest of the secondary system (both liquid and gaseous) was not sampled.

2.2.3 Spent Fuel Pool and Associated Samples The spent fuel pool water was sampled using a sample point on the input line to the spent fuel pool heat removal and water clean-up system (sample point #1 in Figure A.12). During refueling the refueling cavity and fuel transfer canal were sampled using a container dipped inM the waters. The safety injection and refueling water storage tank (SIRWT) water was sampled both before and after refueling using the sample point on the tank's recirculation line (sample point #3 in Figure A.12) . Data for spent fuel pool and associated samples are presented in Tables A.11 and A.14-A.18.

2.2.4 Liquid Radwaste Samples Figure 2.2 presents a diagram of the liquid radwaste collection tanks with information as to the sources of liquid radwaste for each collection tank. Tables 2.2 a, b, and c present the FSAR (3) component infomation for the ifquid radwaste system with comments determined from discussions with plant personnel. The sample point for the reactor coolant drain tank (RW-7) was a local sample point on the tank. Validity of this staple in relationship to the tank contents is not known; however, before a sample was collected, the sample line was purged. There were no sample points for the auxiliary building sump tank or neutralization tank. The sample points for the waste holdup tanks (RW-3), spent <

regenerant tank- (RW-2), and hotel waste tanks (RW-1) were in the recirculation lines for these tanks. Before samples were taken from these tanks, as in all cases where the sampic ooints were located in recircuhtion lines, the tank contents were recirculated for a minimum q of two tank volumes.

Figure 2.3 is a diagram of the liquid radwaste treatment system.

No sample point was available on the output stream of the waste filters.

The feed samples for the waste evaporator were taken from the radwaste collection tanks being processed while the condensate and concentrate samples were taken from points RW-5 and RW-6, respectively. The samples for the monitor tanks were obtained from point RW-4 in the recirculation line. During measurements at Fort Calhoun the waste demineralizers l

l 10 m

~

TABLE 2.2 a CCMPONENT DESIGN DATA, WASTE DISPOSAL SYSTDI TAN)3 s

Pressure, Tempera-psig ture, F No. Installed / Capacity / Tank Design / Design /

Tank Item No. gallons /ft3 Operating Operating Material

• Code 25/2 300/267 30k SS ASME Section III, Reactor Coolant 1/WD-1 900/120 Drain Tank Class C

' Waste Holdup 3/WD-kA, 45,800/6,100 15/2 200/120 CS AS!E Section III.

Tanks B&C Class C Deutralization 1/WD-7 900/120 25/2 200/120 304 SS ASME Section VIII Tank Caustic Dilution 1/WD-11 750/100 Atmos / Atmos 200/ 70 CS None Tank Spent Regenerant 2/WD-13A&B k,800/650 5/ Atmos 200/ 70 304 SS ASt!E Section VIII Tanks Hotel Waste Tanks 2/WD-15A&B 1,200/160 15/ Atmos 200/1h0 CS ASME Section VIII 14cnitor Tanks 2/WD-22A&B 6,000/800 5/ Atmos 200/lk0 304 SS ASME Section VIII Auxiliary Building 1/WD-25 700/95 25/2 200/120 304 SS ASME Section VIII Stunp Tank Gas Decay Tanks h /WD--29 A, 2,990/400 150/100 200/lk0 CS AS!!E Section III.

B C&D Class C Spent Resin Storage 1/WD-33 3,000/koo 25/2 250/120 304 SS ASME Section VIII Tank Concentrate Tanks 2/WD-38A&B 1,200/160 25/ Atmos 300/1h0 CS ASME Section VIII Resin 11easuring 1/WD-43 60/8 50/30 100/70 30k SS AS!T Section VIII Tank

• SS= Stainless Steel, CS= Carbon Steel

Y TABLE 2.2b C0!4PONENT DESIGN DATA, WASTE DISPOSAL SYSTEM PUfIPS Ho. Installed / Fluid Side Pump Item No. g Capacity Material

• Reactor Coolant 2/WD-2A&B

' Horizor.tal g 2A, 250 gpm d 75 ft. 316 SS Drain Tank Punps Centrifugal 2B, 50 rpm 6 75 ft. 316 SS Containment 2/WD-3A&B Vertical 20 gpm 6 34 ft. AI Sump Pumps Centrifugal ,

Waste fioldup 2/WD-5A&B Horizontal 50 gpm 8 177 ft. 316 SS Tank' Pumps Centrifugal, Canned Rotor Waste Holdup 1/WD-6 Horizontal 500 gpm G 85 ft. AI Recirculatis- Centrifugal Pump Neutralization 2/WD-8A&B Horizontal 200 gpm @ 82 ft. 316 SS Transfer Pumps Centrifugal, Canned Rotor (i Sample Pump 1/WD-10 Horizontal 5 gpm B 21 ft. 316 SS Centrifugal, Canned Rotor Caustic Pumps 2/WD-12A&B Positive 3 gpm G 160 ft. CS & SS Displacement, Diaphragm Type Hotel Waste 2/WD-16A&B Horizontal 50 gpm 0130 ft. AI Pumps Centrifugal Gas Stripper 2/WD-20AaB Horizontal 16 gpm 0 75 ft. 316 SS Pumps Centrifugal Canned Rotor

!!on! tor Tank 2/WD-23A&B Horizontal 50 gpm @ 160 ft. 304 SS Pumps Centrifugal Auxiliary Bldg. 2/WD-26A&B Horizontal 35 spm G 110 ft. 30k SS Sump Tank Pumps Centrifugal 1 Auxiliary Bldg. 6/WD-27A&B, Vertical 20 gpm 0 36 ft. AI Sump Pumps

~

40A&B, klA~ Centrifugal

, &B

• AI = All Iron. (continued)

SS .= Stainless Steel CS =.. Carbon Steel 12 - ..

F s

i TABLE 2.2 b (cont.)

No. Installed / Fluid Side M. Item Ido. g Capacity Material *

' Spent. Resin 'l/WD-34 liorizontal 30 gpm G 106 ft. 304 SS Pump Centrifugal f

Nesin Dewatering. 1/WD-35 Centrifugal 35 gpm 6 hk ft. CI bronze Pump Displacement, fitted Rotating Water.

Seal.

Concentrate 2/WD-39A&B Horizontal 30 gpm @ 105 ft. 316 SS Tank Pumps Centrifugal, & CS Canned Rotor.

• SS = Stainless Steel CS = Carbon Steel CI = Cast Iron

=

13

q, TABLE 2.2 e i

CO'iPONENT DESIGN DATA, WASTE DISPOSAL SYSTDI Pit 0 CESS EQUIP! TENT Waste Filters. Item ? o's WD-17A&B Description Humber 2 Type Expendable element pressure type Materials of Contruction 304 stainless steel vest,el with synthetic fiber element Vessel Design Pressure, psig 150 Vessel Design Temperature, F 250 Vessel Code AS!E Section III, Class C Flow Rs.te, each, gpm 30 Retention, % (particles >25 microns) 98 Gas Stripper. Item No. UD-19 Note A liumber 1 Type Packed column, heated liquid-vapor contact, non-condensable gas sepsration.

Design Flow Capacity lbs/hr 7500

, Design Pressure, psig 50 Design Temperature, F 300 Operating Pressure, psig 2 Operating Temperature, F 220 Code ASME Section III, Class C Decontamination Factor, . #

' Gaseous activity in feed 10 Gaseous activity in hotwell (continued) 14 ,

1

_L '

TABLE.2.2 c (cont.)

Waste Evaporator, Item No. WD-21 Description Number 1 Type Steam heated horizontal tube, spray film type Design Pressure (shell side), psig 15 Design Temperature, F 225 Operating Pressure, psig 2 Operating Temperature, F 217 Code ASME Section III, Class C.

Flow Rate, gpm. 17 (Note B)

Decontamination Factor, Activity in feed 10b Activity in distillate Concentration of Bottoms, max, % Na2 B407 20 Note B Waste Demineralizers Item No's. WD-PhA&B Note C Number 2 Type Non-regenerable, disposable vessel and resin Materials of Construction Carbon steel, phenolic lining Design Pressure, psig 100 Design Temperature, F 150 Operating Pressure, psig 30 Operating Temperature, Max. F 120 Vessel Code ASME Section VIII Resin Mixed bed, strong acid cation resin and intermediate base 3

anion resin Resin Volume, each, ft 3

{ low Rate, each, gpm 15 Waste Gas Compressors, Item No's. WD-28A&B Number 2 Type Centrifugal displacement with rotating water seal. 1 Design Flow Rate, each, CFM @ 60 F & 1 atm. h0 I

?

Design Discharge Pressure, psig 100 (Note D) I Design Standard American Society of Heating, Refrigerating and Air-Cenditioring Engineers (ASHRAE)

Fluid Handled Nitrogen, hydrogen, water vapor and trace s of fiscion gases.

Controls Bypass f]Ow regulator for suc-tion pressure control between 1 5 and 1.8 psig Note D Note A: Cas stripper was under repair during measurement period.

Note B Maximum about 15 gps; usual flow rate,10 gym.

Note C: Waste Demineralizers not used during measurements period.

Note Dr Normal discharge pressure; 50 to 86 psig.

15

)

1-

, or side-stream polishing domine'ralizers for the monitor tanks were l not used. According to station personnel, the waste disposal problems

! associated with the spent resins (e.g., high personnel exposures) from the waste domineralizers outweigh the usefulness of processing lower activity liquid radwaste from the monitor tank through the waste domineralizers. With the exception noted below, following . ,

activity measurement, the monitor tenks were released to the discharge -

! header without further treatment. At least once during the measurement i period the total gamma activity in the monitor tank was above release J limits due to dissolved noble gas activity (133Xe). The noble gas activity l was removed by sparging the monitor tanks. This released the noble gas . l to the station stack and reduced the total gamma activity so that the <

! monitor tank could be released to the discharge header. During the 2 Fort Calhoun study the gas stripper in the liquid radwaste treatment system was out of service. A project is underway by OPPD to make the gas stripper operational. Data obtained from analyses of radwaste samples are presented in Tables A.19-A.34.

I

2.3 Gaseous Streams 2.3.1 Ventilation Sampling

! Figure 2.4a presents a schematic diagram of the gaseous waste

{ disposal system. Figure 2.4b shows the auxiliary building ventilation system in more detail with the sampling locations noted. Prior to the installation of sampliag devices in the auxiliary building, velocity j profile measurements were made in the ducts in which sample probes t,ere

! installed. The duct flows were calculated from velocity profiles and were compared to the de=ign flows. Significant differences were i observed. Discussions with OPPD staff resulted in changes in the operating

status of the ventilation system. New velocity profiles were then i measured and duct flows calculated. The results of these-second measurements i are shown in Table 2.3 along with the design flows taken from the P & ID,

" Auxiliary Building Heating and Ventilating Flow Diagram" presented in the

Appendix (Fig. A-28). One minor difference in the flow path was found during sample probe installation and is shown in Fig. A-29 of the Appendix.

This area of the system is shown at coordinates F-4 in Fig. A-28. The new comparisons with design flows were deemed reasonable and sampling probes *

, were installed.

In order to verify the validity of the samples being withdrawn from the ducts,~ helium was injected intn the system upstream of each of the ,

sampling probes and the helium concentration was measureo in each of the sample streams. . The duct flows were calculated from the helium concen-tration measurements using the ratio of the helium injection rate to the measured helium concentration (1). " Results of these measurements are also shown in Table 2.3 The agreement between the flows measured

~ by standard pitot tube traverse methods and those calculated from the helium dilution. technique indicated that the samples taken from the ducts were representative of the gaseous species in the ducts.

L 16 L_

FIGURE 2.4a GASEOUS WASTE DISPOSAL SYSTEM

[ Turbine}

Primary C Vent coolant _3 _

( _,,ondenser

' V To roof v vent Condenser

~ -* vacuum -

Blowdown pumps flash V Charcoal tank adsorber Gac holdup tanks To waste holdup tanks HEPkEi!3 400 cu n b filter CVCS ente for reuse and other

• i b O2free waste 4 p disposal vents Waste gas compressors b From waste ~

holdup tanks Roll after 30 days '

Mtr pper ' Purge inlet HEPA Containment C 3 normally + -Normal filters closed 122,000 cfm--*

Purge Cooling and -Inlet fans outlet internal recirculating WASTE G AS

' I I cha coal f te s "

Roof vent fans (14)dOOO "

cfm (max) 90,000 cfm 2 units Fans c

Turbine building Roll filters

• TURBINE BUILDING VENTILATION CONTROL Charcoal filter Fuel handling

• PA Fans area pg,,

filters Safeguards F  ;  !

~* _

areas 'M Charcoal Other

-=

9,,

-75,000 cfm-* filter AUXILIARY BUILDING VENTILATION CONTROL

• CVCS - chemical and volume control system FORT CALHOUN Ground INEL-A-8013 17

FIGURE 2.4b FORT CALHOUN STATION AUXILIARY BUILDING VENTILATION SYSTEM Hot laboratory, counting

room, cold laboratory, men's room, laundry area Stack Fuel pool and handling qs pcw+

areas, baler, drumming, and -

spent resin storage areas Spent reg chent hold-up and waste concentrate rooms.

HEPA filters

• monitor and neutralization tank j j, 4 rooms, purification filter room g Charging and safety injection pumps rooms, fuel pool heat exchanger p [au@t e--

Waste gas compressor room W Sampling and pipe CVCS wasts hold-up --- > (3) Component cooling and

+ penetration O' 'oo==-

tank areas

• shutdown cooiino heat containment airlock area exchanger rooms, shutdown cooling valve room Volume control tank and waste evaporator h rooms Letdown hand shutdown heat exchanger rooms,  % Waste gas decay tank mechanical penetration rooms ,

Samplers: areas Circled number shows sample station number with samplers for particulate, iodine,3H and 14C.

Circled "L" shows local sample point with samplers for particulate and iodine. INE-A a012

__ _____ - .e 2

6 TABLE 2.3 AUXILIARY BUILOING VENTILATION DUCT FLOW MEASUREMENTS Location Design Flow, cfm Measured Flow, cfm Pitot Tube Helium Dilution_

Station #1 14,250 16,600 16,400 Station #2 39,400 28,900 28,400 Station #3 5,450 6,600 5,100 Station #4 19,200 24,500 23,800 Waste Evaporator Room 1,800 1,460 not measured 3

l l

19

)

Table 2.4 tabulates the va'rious areas of the auxiliary building feeding sampling stations 1 through 4. As indicated from the flow measurements (Table 2.3) station 2 sampled over half of the auxiliary building exhaust flow; consequently, sampling station 4 was installed in an attempt to more closely define specific sources of radioactivity.

For the same purpose, local samplers were deployed in individual auxiliary building rooms. Local sampler locations are shown schematically in Figure 2.4. <

The sampling systems installed in stations #1, #3, and #4 were modified commercial sampling systems which had been used in previous studies (4). The system consists of a sampling probe, particulate '

filter, iodine adsorber, water vapor adsorber, flow integrator, flow meter, and air mover. The sample inlet was a stainless steel nozzle sized so that isokinetic sampling conditions were met with a flow of approximately one cfm. The sample passed through a glass lineo probe into a compartment which contained both the particulate filter and iodine adsorber. Both the probe and compartment were capable of being heated to prevent condensation. However, heating was not necessary at Fort Calhoun because all samplers were inside the plant. The particulate filter used was a 5" diameter fiberglass HEPA filter. Dcwnstream of the filter was a 2" X 4" todine adsorber equally divided in three sections.

The first and last section were activated coconut shell charcoal impregnated with TEDA and the center section was silver zeolite. Af ter passing through the iodine adsorber, the sample stream was split with a small sidestream (100 cc/ min) passed through silica gel to remove water vapor. The two parts of the sample stream were recombined and passed through a dry gas integrating flow meter, and orifice meter to set flow rate and the air mover. The exhaust from the sampling system was vented into the sample duct. These as well as all samplers employed are described more fully in reference 1.

At station #2, an iodine species sampler was connected to the end of the prob?. The species sampler consisted of, in order of flow, a fiberglass HiPA filter; an elemental iodine adsorber, cadmium iodide adsorbed on chromosorb-P; a hypoidous acid adsorber, 4-iodophenol adsorbed on alumina; and c.n organic iodide adsorber, TEDA charcoal.

The charcoal bed was segmented to assure complete adsorption of all of the sample. The sample flow used in all measurements was nominally 2 cfm. c The local samplers employ the same sampling concept as the iodine species sampler described above. The dif ference being physical size and sample flow. The local samplers have a flow of 0.25 cfm. c A tritium and 14C sampler (5) was also installed along with the particulate and iodine sampler 'at stations 1, 2, 3 and 4. This system consisted of a silica gel adsorber to remove water vapor which might contain HTO, a molecular sieve adsorber to remove carbor. dioxide, a catalytic oxidizer to convert organic species to carbon dioxide and water plus any elemental hydrogen to water. Following the oxidizer was a second silica gel adsorber and molecular sieve adsorber. The sampling concept is to distinguish between oxidized and nonoxidized 14C and 3H species.

20 l

L

TABLE 2.4

' AUXILIARY BUILDING SAMPLING STATION FEEDS '

Station 1 - l

1. Letdown heat exchanger room

2. Mechanical penetration area ,

3. Shutdown heat exchanger room j

> 4. Valve room j

5. Pipe penetration area

6. Personnel air lock area

7. Sampilng room Station 2

1. Cask decon room

2. Fuel arrival area

3. Fuel storage area

4. Drum storage area

5. Waste baler room

6. Spent resin storage room

7. Vo6 s control tank room

8. Waste evaporator room

9. Waste holdup tank rooms

10. Spent fuel heat exchanger room

11. Safety injection pump rooms

12. Charging pump room

13. Charging pump valve room l

14. Fuel Pool area i Station 3

1. Waste decay tank rooms i

2. Shutdown cooling heat exchanger room l

3. Shutdown cooling heat exchanger valve room

, 4. Component heat exchanger room Station 4

> 1. Spent fuel heat exchanger room 2.- Safety injection pump rooms

3. Charging pump room

4. Charging pump valve room 21-

All of the venN tation sampling media were returned to INEL for analysis. The partit dates on HEPA filter media were analyzed using a computer based Ge(L1) gamma spectrometer. For a selected list of nuclides, based on activities of long tem counts of samples from other reactora, activity levels were detemined. If no positive identification could be made for a particular isotope, a "less than" (or lower limit of detection) value was calculated. The "less than" value was calculated by fitting a baseline curve to the area of interest of the spectrum, integrating the area, and assigning an up'er p limit in the same manner as for a true photopeak.

The' adsorbent media for iodine was also analyzed by a gamma spectrometer. In all cases the spectrometers were calibrated for iodine media counting employing secondary standards in the proper geometry ' (1 ). Samples analyzed using the secondary standard calibrations were inter-compared with DOE-HSL systems. The comparisor.s gave good agreement (to within 5 percent).

Tritium aasorber sliquots were treated with low background water (contains nondetecta J ; amounts of 3H) and counted in situ using liquid scintillation counting techniques.

-The carbon-14 adsorbents were degassed in a heated helium flow system and the released gas was collected at liquid nitrogen temperature.

. Af ter making manometric measurements to establish sample size, the sample was allowed to react with ethanol amine. The sample was then counted by liquid scintillation techniques.

In all but one case, a disintegration rate for the sample or aliquot was detennined. The disintegration rate was then converted to activity per unit volume of sample using the sample flow rate and sample time.

The exception was the carbon-14 air samples, in this case the counting results were reduced to d/m/cc of CO2 adsorbed. This value was used to multiply the abundance of CO2 in air, 330 ppm, to convert to d/m or pCi per cc of sample. In all cases the nuclide concentrations were converted to nuclide release rates in activity per unit time.

The reported results for gaseous samples include an error associated e with counting statistics and background fluctuations. In addition, an overall uncertainty of approximately 20 percent should be applied to account ~ for errors in calibration and indeterminates such as duct flow variations and in a few cases uncertainty in sampling duration. The c ventilation data are presented in Tables- A.35-A.44 in the Appendix.

-2.3.2- Waste Gas Processing System

!- - Short tem samples were taken from the pressurized storage l- tanks in the waste gas processing system. In all cases the samples l

were taken through the plant's Automatic Gas Analyzer System. The system is designed.to do hydrogen-oxygen analysis of various cover gases.

I 22

Iodine and gas samples for noble gas analysis were taken from this' system using the previously described iodine species sampler or a 250 'cc gas

cylinder (results are-reported in Tables A.45-A.48 in,the Appendix).

Sample duration ranged from 10-60 miriutes.

I Additional samples for 14C and tr'itium analysis were withdrawn from ~the pressurized storage tanks through the gas analyzer system.

, The analysis of these samples presented several unexpected problems.

First, radiation fields from the silica gel adsorbents due to noble gases were 200-400 mr/hr measured at contact. Secondly, the samplers were designed to handle air samples, i.e., oxygen is required for the proper operation of the catalytic oxidizer which provided separation -

of the various tritium and 14C species. With the exception of one sample-(evidenced by an exothermic reaction in the catalytic oxidizer),

the cover gas samples contained;little or no oxygen. Consequently, special sample collection and handling techniques had to be developed.

For samples containing an appreciable amount of oxygen, an aliquot of the silica gel adsorbent was put in low background water and refluxed in the presence of a helium purge. This procedure excluded the noble

- gas activities. The purified solution was then counted by liquid 4 scintillation techniques.. '

As an alternate sampling procedure for samples containing insuf- ,

ficient amounts of oxygen, 75 cc stainless steel gas cylinders were-employed for collecting these samples. The sample cylinders were .

evacuated and pressurized with sample gas. Afterwards, the gas cylinders were returned to INEL where the contents were mixed with low activity air'to provide oxygen and to reduce the potential of sample contamination.

The samples were further proc 6ssed by passing the mixture through the 3H14C sampler and then handled in the previously-described manner.

The waste gas processing system data are presented in Tables A.45-A.50 in the Appendix.

2.3.3 Containment Atmosphere Sampling Samples to determine iodine species and 3H 14C were obtained from the containment building atmosphere. In addition, a 250 cc gas cylinder sample of the containment atmosphere was obtained for noble j gas analysis. Samples.for the determination of iodine species and 3H 14C i were taken employing the previously described samplers. l 2.4 Plant Operational Data

-The plant operational data collected to characterize samples included the reactor log (computerized form), the auxiliary building operator's log, plus information obtained in discussions with plant 23

operating personnel. The inforhation obtained from the reactor log included:

1. Reactor power level.

2. Letdown flow rate.

3. Steam generators steam flow rates and blowdown rates.

4. Reactor coolant boron concentration.

The above information was tabulated on an hourly basis throughout the ,

day and is presented in Figures A.1, A.2, A.6, A.8, A.9, A.10, and A.11 '

in the Appendix. The information obtained from the auxiliary building l operator's log included: -  !

1. Levels in the liquid radwaste system tanks.

2. Pressure drop across the CVCS purification desineralizers and filters.

3. Evaporator operational parameters including liquid feed rate, bottoms boron concentration, and distillate l conductivity.

l

, 4. Time during the day when specific liquid radwaste tanks were j isolated, processed, filled and recirculatad.

5. Times during the day when the evaporator was processing a

, specific liquid radwaste tank and when it was on recirculation.

Items 1 through 3 were on a two-hour interval throughout the day while the specific times when events occurred were noted by the operator in the log. These data are presented in Figures A.7, A.10, A.17, A.18, A.19, A.20, A.21, A.22, A.23, A.24, A.25, A.26, and A.27 in the Appendix.

Discussions with personnel at Fort Calhoun Station verified that the radwaste tank sizes, demineralizer types, and evaporator were as specified in the FSAR and presented in Tables 2.1 and 2.2. Fort Calhoun did not recycle any of the liquid or gas streams that reached the radwaste system. ,

o d

1 24

CHAPTER 3 - DISCUSSION OF DATA 3.1 Introduction In this chapter, data obtained from sample analysis are discussed on a component or plant system basis. Plant systems and components are related to the principal parameters used in the NRC calculational models (6). Each component is addressed in a separate section.

3.2 Thermal Power Level The maximum designed and licensed power levei for the Fort Calhoun Station is 1420 MWt. During the measurement period, the reactor was operating near or at this power level, except for hot shutdowns on 8/31/76, 9/24/76 and 1/14/77 and the refueling outage.

In the Appenaix, Figure A.1 presents a plot of the power level for the period 8/15/76 through 10/3/76. Figure A.2 presents the power level for the period 12/15/76 through 2/24/77. From 10/1/76 through 12/15/76, Fcrt Calhoun was down for its second refueling.

3.3 Plant Capacity Factor Information trom OPPD (8) indicated that the capacity factors for the Fort Calhoun plant were: 60.4% in 1974, 52% in 1975 and 54.7% in 1976.

3.4 Radionuclide Concentrations in Pressurized Water Reactor Liquid System Components Radionuclide concentrations were measured in samples obtained from liquid system components at Fort Calhoun. These included: reactor coolant, steam generator blowdown, spent fuel pool, reactor coolant l drain tank, waste holdup tanks, spent regenerant tanks, hotel waste tanks and monitor tanks. Results will be discussed in detail later in l this section.

3.4.1 Predicted Concentrations in Reactor Coolant and Secondary Water Table 3.1 presents the expected radionuclide activity levels in the primary and secondary coolant water for the Fort Calhoun Station based on the N-237 standard (7). These values were derived by adjusting the parameters of the reference PWR to those of the Fort Calhoun plant.

Techniques for the adjustments are presented in the American National

' Standard.N-237/ANS-18.1 publication of the American Nuclear Society (7).

The values used for adjustment of the reference PWR values to the Fort Calhoun values are presented in Table 3.2. The parameters are for U-tube 25

i l

1 TABLE 3.1 PREDICTED CONCENTRATIONS.IN REACTOR COOLANT AND SECONDARY WATER Calculated for Fort Calhoun from N-237(7) Reference Values

- Reactor Coolant Secondary Water * ,

Nuclide _ uCi/gm)

( (uCi/cm) j Noble Gases - Class 1: I

• • s amKr 1.7 - Nil I asmKr 8.7 - Nil 85Kr 1.5 - Nil 87gp - 4.8 - Nil eeKr -1.6 - Nil

• • 89Kr 4.0 - Nil 131mXe 9.3 - Nil 133mXe 1.8 - Nil 133Xe 1.5 + Nil 135mXe 1.0 - Nil 13sXe 2.8 - Nil

• • 13 7Xe 7.1 - Nil 13 axe 3.5 - Nil Halogens - Class 2:

• • 8 3Br 4.0 - 5.1(-

• • 84Br 2.1 - 5.11,-

• • esBr .2.4 - 4.71 )

• • 130I 1.9 - 1.8I,-

131I 2.7 - 1.6l'-

1321 8.2 - 3.91 -

133I 3.5 - 5.EI'-

134I -3.8 - 1.6l'-

135I 1.6 -1) 7.3d-Cesiums, Rubiditans - Class 3: e

• • 86Rb 8.6(-5 5.1 -

8 erb 1.6 - 1.9 -

134Cs 2.6 - 1.3 - c 1 136Cs 1. 3 - 6.4 -

137CS ],8 1,1 -

Water Activation Products - Class 4:

• • -16N 4.0(+1) 2.9(-6) l-l Tritium - Class 5:

3H 1.0(0) 1.0(-3) 26 m

~ TABLE 3.1 (cont'd)

PREDICTED CONCENTRATIONS IN REACTOR COOLANT AND SECONDARY WATER Calculated for Fort Calhoun from N-237(7) Reference Values

Reactor Coolant Nuclide (uti/am) Secon'dary (uC1/qm W)ater*

Other Nuclides - Class 6:

51Cr 1.9 -3) 1.Ol' 3

2.71,-)

54Mn 3.1 -4) 55Fe 1.6 - 3) 9.41, , - )f 59Fe 1.0 - 6.41,- l l l l

seCo 1.6 - 9.01,- J 60C0 2.0 - 1 .21 -

895r 3.51 -

2.6ll 6.41,- /

// .

l 90Sr 1 .01 -

• • 9ISr 5 . 71 l 3. 31 2.11,-

[ i

• • 90Y 1 . 21 -

l , )

91Y 6.51 l 3.91

• • 91mY 2.91 l

8.11,- )l

• • 93Y 3.0f'l 2.3/-

9sZr 6.0Y'l 3.9Y- i'l 95Nb 5.0E'- 'I~ 3.9d- $'l 99Mo 8 . 21 l 3.2d#

99mTc 4.lf- 1.4/- l lo3Ru 4.5 - 2.6dh

• • 106Ru 1.0 -

=*103mRh

• • 106Rh 3.6 -

7.9 -

6.4fl0l 5 . 61,-

1.2i -

l

• • 12smTe 2.9 - 1.2 - p)

• • 12 W e 2.8 - l 1.2 -

• • 127Te 7.5 'l 1.1 /l 129mie 1.4 - l 7.7I,- h 129Te 1.3 - 1 . 71 -

131mTe 2.3 - 4.3 , J)

• • 131Te 8.8 - 1.2 I

• 8.6 132Te 2.6 - -

• • 13/m8a 140Ba 1.3 -

2.2 -

1.8 1.1

-)/

140La 1.4 - 7.7 i l

141Ce 7.0 - 3.9 -

14 3Ce - 3.8 - 8.2 - J) 144Ce 3.3 - 2.7 - )

• • 143Pr 5.0 - 2.5 - )

• • 144Pr 2.6 -

239Np 1.2 - 4.7 2.5 -- J)

• Calculation assumed all volatile treatment chemistry for secondary water.

• • N-237 (7) radionuclides not directly measured at Fort Calhoun.

27 i

TABLE 3.2 PARAMETER VALUES USED TO MODIFY N-237 (7)

ACTIVITIES FOR FORT CALHOUN Fort Calhoun Parameter Symbol Units Value Themal power P MWt 1420

• Steam flow rate (both generators) FS lbs/hr 6.2(6)*

Weight of water in reactor coolant WP lbs 2.9(5)**

system Weight of water in all steam WS lbs 1.56(5)***

generators Reactor coolant letdown flow FD lbs/hr 1.5(4)*

(purification)

Reactor coolant letdown flow FB lbs/hr 200 *

(yearly average for boron control)

Steam generator blowdown flow FBD lbs/hr 25,000 *

(total)

Fraction of radioactivity in blow- NBD ------

1.0

• down stream which is not returned to the secondary coolant system.

Flow through the purification FA lbs/hr 1600

• system cation demineralizer l

Ratio of condensate demineralizer NC ------

0.0

• flow rate to total steam flow rate "

Ratio of the total amount of ncble Y ------

0.0

• gases routed to gaseous radwaste from the purification system to d the total amount routed from the

~ primary coolant system to the l purification system (not i includin system) g the boron recovery

• Based on infomation obtained during measurement program.

• • Information from FSAR (3).

• • • 'Information from (9).

l 28

steams generators assuming all volatile treatment ( AVT) of the secondary coolant. In Tables 3.1 and 3.2, as in all other tables in this report, the number - in parentheses following a number represents the power of

' ten multiplier.

The radionuclides presented in Table 3.1 include all the radio-isotopes discussed in N-237. In the subsequent data tables, all of the

? radioisotopes discussed in N-237 are not included.

The techniques used in collecting the data would have detected and quantified all gamma-emitting radionuclides present in the collected

)

samples. However, our experience with these measurements has indicated that certain radionuclides are either not present or should not be reported due to potential errors in sample collection.

An example of these potential errors is the measurement of the noble gas activities in liquid samples. Only with specialized sample collection techniques is it possible to quantitatively determine the dissolved nob'le 9as concentrations in liquid samples. At almost all sampl'ng points used, the only method available for sample collection was to transfer the liquids into open volumetric measuring devices. This procedure results iri an unknown degasification factor for the dissolved noble gases (i.e.,

an indeterminate amount of noble gas'es is removed from the liquid sample).

Consequently, noble gases have not been included in most of the data tables. Where noble gas concentrations are reported, more specialized  ;

sample techniques (i.e., plumbing the sampler in series with the sample l line) were employed. This was done for selected reactor coolant samples and the data are presented in the reactor coolant section.

Certain radionuclides that are reported in N-237 have not been observed in data obtained at Fort Calhoun. For example, the radionuclides, 1301 and 86Rb, were not detected because of their very low fission yield (about 2 X 10-3 percent for each isotope). The radionuclides 16N and 93Y were not detected due to their short half-lives or low gamma-ray branching ratios.

Several of the radioisotopes listed in N-237 are daughters of I 1

measured parent radioisotopes. For example, 90Sr 90Y, 91Sr 91my, 3 99Mo 99mTc,103Ru 103mRh,106Ru 106mRh,137Cs 137m8a, and 144Ce 144Pr are such parent-daughter radioisotopic pairs. It is our experience in the ndionuclide measurement of samples from nuclear plant systems that these parent-daughter pairs are in radioac'tive decay equilibrium because 3 of production, half-life, and time considerations. Therefore, the parent activities are reported and the daughter activities can be easily calculated assuming radioactive decay equilibrium.

No bromine activities were-detected in liquid samples from Fort Calhoun.' Particular attention has been given to the detection of 84Br due to its decay scheme and relatively large fission yield. However, it has not been observed. A potential explanation for this is that bromine is a very reactive element and could attach to the surfaces of reactor system components.

29

k 3

The tellurium isotopes have not been observed with any consistency in.either reacgt coolant or other plant process streams. The tellurium radioisotopes NTe and 132 Te have large fission yields, relatively long half-lives and detectable gamma-ray emissions. Particular attention has been given to detection of these two tellurium isotopes, but they have not'been observed with any consistency. A potential explanation is that these radionuclides are fixed on the. reactor system components.

The fixing of radionuclide activities to structural materials can be a time varying phenomena. That is, it can be a function of the chemical )

and physical properties (pH, temperature, flow rate, ions present, chemical form of radioelement, etc.). The tellurium isotopes are presented in the ,

data tables when detected. e 3.4.2 Measured Reactor Coolant Radionuclide Concentrations Table 3.3 lists the number. of samples, mean values, and ranges of radionuclide concentrations in reactor coolant. Data summarized in Table 3.3 are presented in Tables A.1, A.2, A.3, and A.7 of the Appendix.

4 If only one of the samples had a measured value for a specific radio-nuclide, this value is presented as the mean along with its one sigma standard deviation error. Also, for the pure beta-emitting radionuclides, usually one sample was collected for analysis during a particular-measurement period. -If none of the samples collected during a-parti-cular measurement period resulted in a measurable value for a specific r.adionuclide, then the smallest lower detection limit is

. presented as the mean. The radionuclide concentrations presented in the first column are from Table 3.1. Values from Table 3.1 have been converted from pCi/gm to pCi/ml assuming a density of 1.0 ge/ml for water.

As indicated in Table 3.3, 1311, seCo, and 3H have significantly different concentrations for the pre-refueling and post-refueling periods.

Iodine-131 and ssCo were observed to decrease after refueling while the i 3H increased. There is no obvious explanation for the increase in tritium after refueling. However, weekly tritium analysis by OPPD of reactor coolant indicated the same increase. The iodine decrease is attributed to cleanup and decay. The seCo coolant level activity decreased by approximately 70 percent after refueling. Based on a forty-five day outage, the seCo should have decayed by only approximately 35 percent, e indicating cleanup by the CVCS demineralizers during the refueling outage.

- A decrease in 60C0 levels was not observed. This may be attributed to

~ the higher DF's for 58C0 than for 60C0 observed in both the CVCS purifi-cation and cation demineralizers (see Tables 3.23 and A.11). The higher c l observed DF's for 5sCo may be due to the fact that 58C0 and 60Co exist l as different chemical species in reactor liquids.

l l The rest of the radionuclides presented in Table 3.3 had similar radionuclide levels ~before and after refuel-ing or were highly variable.

l Table 3.4 presents radionuclide concentrations in reactor coolant during the refueling outage, before and after spent fuel movement.

Since there was only one umple of reactor coolant taken after fuel 30 l

r i

TABLE 3.3

'RADIONUCLIDECONCENTRATIONSINREACTORCOOLANT(REACTORPOWEROPERATIONS)

Measured: Before Refueling Measured: After Refueling Samples: 14:52; 8/18/76 Samples: 08:40; 2/9/77

, 09:17; 8/19/76 08:44; 2/17/77 09:58; 8/19/76 13:01; 2/22/77

~

11:19; 9/1/76 Calculated 11:54; 9/1/76

, Using N-237 (7) Mean Range Mean Range Nuclide (uCi/ml) (uCi/ml) (uCi/ml) (uCi/ml) (uCi/ml)

Noble Gases:

esmKr 8.7 -

• 1.7-1) 1.4-1.9 -1) asKr 1.5 -

• 1.3 -2) 0.8-1.8 -2) 87Kr 4.8 -

• 1.7-1) 1.4-1.9 -i) seKr 1.6 -

• 2.9 -1) 2.6-3.3 -1) 131mXe 9.3(-2

• 4.5-2) 3.1-5.2(-2) 13xiXe 1.8(-1 1. 3 - ) 1.2-1.4(-1) 133Xe 1.5 +

• 5.5 0 5.0-6.0 0) 135mXe 1.0 -

• 8.6 - 7.0-9.5-2) 13sXe 2.8 -

• 9.8 - 0.9-1.1 0) 138Xe 3.5 -

• 1.6 - 1.5-1.6(-1)

Iodines:

131I 2.7(-1)_ 2.7(-l 1. 5-3.9 (-1 ) 8.5(-2) 7.4-9.5(-2) 1321 8.2(-2) 9.7(-2 0.82-1.1(-1) 5.2(-2) 4.6-6.3(-2) 1331 3.5-1) 1.4(-l 0.34-2.4(- ) 1.l(-1) 1.0-1.3 -

134I 3.8 -2 3.3(-2 2.8-3.8(-2 3.5(-2) 2.8-4.2 -

135I 1.6 -1 6.1(-2 2.8-7.3(-2 7.3(-2) 6.8-8.2 -

Cs and Rb:

88Rb 1.6(-1) 4.5 - 3.8-5.3 - l 5.0-) 4.6-5.3 -

89Rb *** 5.9 - 4.5-7.1 - l 1.9 - 1.3-2.3 -

134Cs 2.6 - 1. 5 - 1.4-1.6 - l 2.2-}f 0.1-6.2 -

136Cs 1. 3 - 4.0 - 3.6-4.7 '

l 6.3 - ) 3.5-9.7 -

137Cs 1.8 - 1.4 - 1.4-1.6 - l 2.6 -2) 0.2-7.1 -

3.1-3.7-)

3 13aCs ***

3.3(-1 1.7-1) 1.5-2.0 -

139Cs *** 1.6i0.4-1) **

<4(-4)

Tritium:

3H 1.0(0) 4.77 0.02(-2)** 2.15 i 0.02(-1)

Other Nuclides:

14C ***

'4.7*0.4(-6) ** 7.010.2(-6) **

~*** 1.1(-2) 0.75-1.4(-2) 24Na -

1.9(-3) 1.3-2.8(-3)

• 32P ***

4.5t0.4(-6) -**

- 35S . *** 5.6i2.0(-6) * '

See bottom of page. Table 3.3 (cont.) for asterisk notes.

31

TABLE 3.3(cont'd)

RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT (REACTOR POWER OPERATIONS)

Measured: Before Refueling Measured: After Refueling Samples: 14:52; 8/18/76 Samples: 08:40; 2/9/77 09:17; 8/19/76 08:44; 2/17/77 09:58; 8/19/76 13:01; 2/22/77 3 11:19; 9/1/76  !

Calculated 11:54; 9/1/76 Using N-237(7) Mean Range Mean Range Nuclide (uC1/ml) (uC1/ml) (uC1/ml) (uC1/ml) (LCi/ml)

Other Nuclides: (cont'd) 41Ar ***

• 3.4(-3) 2.0-5.6(-3) 45Ca 1.08 0.01(-4)

• 51Cr 1.9(-3) <1(-2) 7.3 i 1.8(-4) **

56Mn 3.1(-4) 2.5(-4) 0.13-1.1 (-3) 5.6(-4) 0.022-1.6(-3) l 56Mn 3.6(-2) 3.3-3.7(-2) 1.4 0.4(-2) **

55Fe 1.6(-3) **

1.06 t 0.01(-3) 1.16 1 0.02(-4) **

59Fe 1.0(-3) 7.5 - 0.56-1.1(-4) 1.1 - 0.47-1.8-)

• • • 2.8 -

57Co ** 1.5 - 0.36-2.3 - )

58Ca 1.6(-2) 7.1 - 0.60-1.2(-2) 2.2 - 1.1-3.6( 3 SoCo 2.0(-3) 1.3 - 0.79-2.0(-4) 5.7 - 0.18-2.1 -)

6 3Ni *** 1.76 ** **

65Zn ***

0.02(-4) 3.710.2(-5) 2.8(-3) 1.1-5.0(-3) <2(-5) i 89Sr 1.22 c 0.05f-3 ** '

3.5q- 8.6 i 0.9(-6) **

90Sr 1.0h- 3.8 1 0.5(-6) ) **

<8(-7) 91Y 65.1 -

7.7 i .3(-6) ** 1.3 i 0.2 - **

95Zr 6.0 '- 4.4 - 0.8-8.1(- 4.3 i 1.2 - **

95Nb 5.0 - 6.2 - 3.8-8.6( 3.8 0. 3 - **

99Mo 8.2 - 1.3 - 0.25-1.5 ) 1.4(-) 0.064-1.4(-3) lo3Ru 4.5 - 6.9 - 0.27-1.1 - ) 1.7(- ) 1.7-1.7(-5) 110 mag *** 2.1 0.2-3.8(-4 <3(-5 124Sb 1.210.7(-5) **

3.6i2.4(-6) **

129mie 1.4 - <2 - 1.9 i 0.8(-4) **

129Te 1. 3 - <7 - <2 - .

131mTe 2.3 - <3 - <2 -

132Te 2.6 - <1 <2 -

• • • 6.0 -

139Ba 4.2-7.4 - 5.6(- ) 4.9-6.3(-2) 140Ba 2.2 - 5.0 - 3.2-5.9 - <1(-4 e

'140La 1.4 - 4.0 - 3.0-4.7 - 2.6(- ) 0.87-4.4(-4) 141Ce 7.0 - 3.0 i 1.3(-4) **

2.711.5(-4) **

143Ce 3.8 - 1.2(-2) 0.083-3.0(-2) <3(-3) l 144Ce 3.3 - <6(-4) 2.3 0.3(-3) **

l 147Pm 9.0 i 2.0(-7)

• 187W 1.9(-2) 0.26-3.8(-2) 7.0(-3) 5.4-8.6(-3) 239Np 1.2(-3) 1.4(-2) 0.53-2.0(-2) 7.0(-3) l 4.3-9.6(-3)

( *  : Analysis not perfomed for radionuclide.

I ** : for this radionuclide.

• • • Detected Radionuclidein one measurement, not treated in N-237 only,(7).

l I

32

TABLE 3.4 RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT (REACTOR REFUELING)

Measured: During Refueling Prior to Fuel Transfer After Fuel Transfer Calculated Sample: 12:36; 10/20/76 Sample: 13:15; 12/3/76 Using 11:00; 10/27/76 x

N-237 Mean Range Activity Nuclide (uCi/ml) ( 7) (uCi/ml) (uCf/ml) (uCi/ml)

Iodines:

131I 2.7(-1) 1.2 0.2(-2) **

<4(-4)

Cesiums:

134Cs 2.6(-2) 3.0(-2) 2.3 - 3.7(-2) 6.18 0.09(-3) 136Cs 1.3 - 1.5 - 0.9 - 2.1 - 1.9 0,3 -4 137Cs 1.8 - 3.0 - 2.3 - 3.8 - 6.3 0.1 -3 Tritium:

3H 1.0(0) 3.5(-2) 3.4 - 3.5(-2) 1.06 i 0.02(-2)

Other Nuclides:

14C 2.4(-5) 2.1 - 2.7(-5) 5.3 0.2(-7) 32P 1.1(-5) *

• • • 0.5 - 1.6

-5) 35S 1.4 0.9-6) 2.50 0.06(-5) 4sCa *** *

<6(-7 stCr 1.9 - <2( 3 2.9 i0.9(-3) 54Mn 3.1 - 3.1-4) 2.5 - 3.6 -4) 8.7 0.2(-4) 55Fe 1.6 - 1.3 -3) 0.39- 2.2 -3) 3.91 0.02(-3) 59Fe 1.0-3)

• • • 1.4(-4 1.2-1.6-4) 3.2 i0.4(-4) 57C0 3.2(-4 2.4 - 4.1 -4) 2.3 0.3(-4) 58Co 1.6(-2) 1.6(-1 1.3 - 1.8 - 7.0 10.1(-2) 60C0 2.0(-3) 6.7(-4 6.4 - 6.9 - 1.6 1 0.3(-3) 63N i ***

4.8(-3 3.9 - 5.8 - 3.94 i 0.02(-3) ssZn *** ** 5.1 1.0(-5)

, 89Sr .3.5 - 8.22.7 i 2.5()-5)

-5 0.53-4.8(-5) 5.3 i 0.3(-5) saSr 1.0 - 1.2-6) 0.4 - 2.0(-6) 1.08 0.05(-5) 91Y 6.5 - 9.0 -7) 0.3 - 1.5(-6) 2.13 0.08(-5) 95Zr ~6.0(-5 <1(-2) 1.1 0.2(-4)

> 95Nb 5.0(-5 <9(-3) 2.6 1.7(-4) 103Ru 4.5(-5 **

<3(-4) 110 mag ***

1.5 <1(-3 i 0.8()-4) 1.5 0.1(-4) 124Sb 9.6(-5) 0.61- 1.3(-4) 1.28 i 0.09(-4) 129mTe 1.4(-3) <4(-1) <2(-1) 1408a 2.2(-4) <5(-4) 5.3 2.1(-4) 141Ce 7.0(-5) <1(-4) <4(-5) 144Ce 147pm 3.3(-5) 2.6i2.3(-4) **

<2 (- 4 )

1.4(-6) 1.1 - 1.7(-6) l

Analysis not performed for radionuclide.

Detected in one measurement, only, for this radionuclide.

• • • : Radionuclide not treated in N-237.(7).

33

movement, the measured activity' level and the one sigma standard deviation based on counting statistics are presented instead of a mean and range. - The data summarized in Table 3.4 are presented in Table A.4 of the Appendix. _ Radionuclides with half-lives of lese than eight

. days are not presented in Table 3.4 because these was .. nroduction of

• radionuclides due to'the reactor not operating. Between the collection of the two data groups of Table 3.4 (before and af ter fuel transfer), c the reactor coolant was being processed through the cation demineralizer of the CVCS. This resulted in a significant reduction in the cesium concentrations as is shown in Table 3.4, The data for the cation demineralizer are presented in Table A.11 in the Appendix, (

3.4.3 Radionuclide Concentrations in Steam Generator Water Water samples were taken of the steam generator blowdown for both generators ~ before and af ter refueling. Table 3.5 presents the results for blowdown samples. For comparison, the expected radionuclide concentrations calculated using the N-237 model are 7.1s0 listed. Data summarized'in Table 3.5 are in Table A.13 in the Appendix. The blowdown activity levels were observed to be lower than predicted by N-237. At the time .the samples were taken, the total steam flow rate was 6 million pounds per hour and the total blowdown rate was 20,000 lbs/hr. Samples obtained prior to refueling were 450 ml in volume, which did not provide adequate measurement sensitivity for many nuclides. After refueling, large volume samp.les (about 200 liters) were processed through 101 exchange columns to improve sensitivity. As shown in Table 3.5, the presence of several radionuclides was confirmed using this technique. However, there was no indication of a primary to secondary leak (based on 1311, levels

' < 10-2 lbs/ day). Fort Calhoun Station released all the steam generator blowdown water to the discharge pipe during the period of the in-plant measurements. That is, none of the blowdown waters were processed with demineralizers prior to release or recycled within the plant.

3.4.4 Concentrations of Radionuclides in Spent Fuel Pool and Refueling Associated Waters ,

Mean values and observed ranges of radionuclide concentrations in the spent fuel pool samples are presented in Table 3.6. All measurements c are presented in Tables A.2, A.14, and A.18 of the Appendix. Before refueling, the spent fuel pool contained 32 expended fuel assemblies

-from the refueling in the spring of 1975. After refueling, the fuel pool contained 64 expended fuel assemblies. Table 3.7 presents the radio- c nuclide concentrations observed in fuel pool water during refueling.

The data are presented from measurements made before and after fuel movement to the pool. All data obtained are listed in Table A.17 of the Appendix. With the exception of IC, radionuclide concentrations in the spent fuel pool increased after fuel movement.

Using measured tritium concentrations in the reactor coolant,

- refueling cavity, spent fuel pool, and the safety injection and refueling

-34

. u - . -_

TABLE 3.5 RADIONUCLIDE CONCENTRATIONS IN STEAM GENERATOR WATER Steam Generator Blowdown Activities:

Measured: Before Refueling Measured: After Refuelina Calculated-Generator Generator A: 09:40; 8/26/76 Generator.A: 2/08 - 09/77 Water . Generator B: 09:40; 8/26/76 Generator B: 2/10 - 11/77 Using N-237(7). Mean of Both . Range Mean of Both Range-Nuclide (uCi/ml) (uCi/ml) (uCi/ml) (uCi/ml) (uCi/ml)-

Iodines: ,

131I 1.6 -3 <4(-8) 5.5(-10) 1.7 - 9.3(-10) 183I 5.6 -4 <3(-10)

Cesiums:

~ 134 Cs _l.3 - <1(-7) 5.3(-9) P.4 - 8.2(-9)

$ 136g3 g,4

• g,4 g g,7(_1]} **

137 Cs 1.1 - <1(-6) 8.3(-9) 0.38 - 1.3(-8)

Tritium:

3 ** t H 1.0(-3) 1.7 1 0.2(-5)

Other Nuclides:

14 ***

C <3(-8) t 24 Na *** *

<1(-10) 32P *** <4(-8) t 51

• 3.9 Cr 1.0 - 4.0(-10) :4.0(-10) 54Mn 2.7 - <1(-7) .

1.0(-9) 0.70 - 1.3(-9) ssFe 9.4 - 5.2 i 0.l(-6) t 59

• Fe 6.4 - <3(-10)

TABLE.3.5 (cont'd)

RADIONUCLIDE CONCENTRATIONS IN STEAM GENERATOR WATER Steam Generator Blowndown Activities:

Me'asured: Before Refueling Measured: After Refueling Calculated Generator Generator A: 09:40; 8/26/76 Generator A: 2/08 - 09/77' Water Generator B: 09:40; 8/26/76 Generator B: 2/10 - 11/77 UsingN-237(7) Mean of Both Range Mean of Both Range-Nuclide (uCi/ml) (uCi/ml) (uC1/ml) (uCi/ml) (uCi/ml)

Other Nuclides: (cont) 57CO *** *

<6(-11 58C0 9.0(-5) 3.0(-9 0.78-5.2(-9) 60Co .l.2(-5) <2(-7) 1.2(-9 0.64 - 1.8(-9) gg 63N i- ***

<8(-8) t 89Sr 2.6 - <4 -8) t 90Sr 6.4 - <2 -8) t 91Y 3.9 - <1 -8) t 95Zr 3.9 - *

< 3(-10) 9sNb 3.9 - *

<1(-10) **

99Mo 3.2 - *

< 6(-11) 103Ru 2.6 - * **

124Sb *** *

<3(-10))

2.1(-10 0.9 - 3.3(-10)

• : Radionuclide not observed.

• • : Detected in one measurement, only, for this radionuclide.

• • • Radionuclide not treated in N-237 (7).

t : Analysis not done for this radionuclide.

n o . m . . .

l l

l

' TABLE 3.6 RADIONUCLIDE CONCENTRATIONS IN SPENT FUEL POOL (REACTOR POWER OPERATIONS) l r Before Refueling: After Refueling Samples: 10:49; 08/23/76 Sample: 08:45; 02/22/77 10:31; 08/31/76 Mean Range Activity

)

Nuclide (pCi/ml) (pCi/ml) (pC1/ml)

Cesiums:

13Cs 9.4(-5) 9.2 - 9.6(-5) 7.6 0.2(-5) l 136Cs # 2.711.9(-7) **

<6(-5) . l l'37Cs 2.0(-4) 1.9 - 2.l(-4) 1.39 0.02(-4)  !

l Tri tium- l 3H 7.34i0.04(-3) 8.94 0.03(-3) l Other. Nuclides: )

14C 7.012.0(-8) 1.8 0.l(-7)  !

32P <2 - )

• l 3sS <2 - )

4sca 1.44 .02(-5) **

• 54Mn 1.7(-5) 1.7 - 1.7(-5) 1.38 0.04(-5) 55Fe 1.8510.02(-5) ** 1.12 0.02(-4) 59Fe <2(-6) 2.0 0.3(-6) 57C0 6.0(-6 5.6 - 6.4 -6) 7.3 1 0.3(-6) 58Co 8.7(-4 8.6 - 8.7 -4) 1.39 i 0.02(-3) 60Co 4.1(-5 4.0 - 4.2 -5) 4.77 0.07(-5) 63N i 2.37 i 0 4(-4) ** 3.42 0.02(-4) 65Zn <1(-6 <2(-6) 89Sr <4(-8 5.5 0.6(-7) 90Sr <2 - 1.7 0.2(-7)

, 91Y <1 - 4.8 i 0.2(-7) 95Zr 7*2- **

<1(-4) 95Nb <4l-7[ 1.2 1 0.4(-6) 99Mo 9 1 7L-7/ **

<4(-7) 110 mag 2.8(-6). 2.2 - 3.3(-6) <2(-6) 12"Sb 4.9(-7) 3.9 - 5.9(-7) <3(-6)

Is,7Pm <2(-8)

• : Analysis not performed for radionuclide.

• • : Detected in one measurement, only, for this radionuclide.

37

TABLE 3.7 RADIONUCLIDE CONCENTRATIONS IN SPENT FUEL POOL (REACTOR REFUELING)

Before Fuel Transfer: After. Fuel Transfer:

Samples: 10:05; 10/14/76 Samples: 14t10; 11/08/76~ '

11:05; 10/27/76 14:10; 12/02/76 Mean Range Mean Range Nuclide (pCf/ml) (uCi/ml) (pCi/ml) (uCi/ml)

Iodinest 1311 <3(-7) <6(-7) <6(-6) <8(-6)

Cesiums:

134Cs 1.0(.-4) 0.82-~1.2(-4) 2.5 -3)

• 2,5 - 2.6(-3) 136Cs 2.3 i 0.7(-6) ~ 4.1-5) 1.5-6.6(-5) 137Cs 2.2(-4) 1.9 - 2.5(-4) 2.7 -3) 2.6.- 2.7(-3)

Tritium:

3H 7.3(-3) 7.2 - 7.3(-3) 9.2(-3) 9.1 - 9.2(-3)

Other Nuclides:

1"C 9.2(-7) 0.67-1.2(-6) 4.6(-7) 0.3 - 8.8(-7) 32P <2(-7 <4(-7) 35S <2(-7 1.1(-6) 0.5 - 1.7(-6) 4sCa <6(-S <6(-8)

SICr **

<3(-6 7.6 i 2.5(-5) 54Mn 1.7(-5) 1.5 -1.9(-5) 8.3(-5) 8.2 - 8.3(-5) 55Fe -8.7(-6) 0.49 - 1.3(-5) 5.0(-5) 2.4

- 7.5(-5) 59Fe <4(-6) 9.5

• 1.3(-6) 57C0 5.2(-6) 4.5 - 5.9 -6) 4.0 -5) 3.7 - 4.2(-5) ssCo - 5.6 -4) 5.1 - 6.2 - 1.5-2) 1.4 - 1.5 -2) 60Co 4.8 -5) 4.5 - 5.0 - 1.4 -4) 1.2 - 1.5 -4) 63Ni 2.7 -4) 2.6 - 2.8 - 1.0(-3) 0.97- 1.0 -3) ,

65Zn <4(-6) 4.0 -6) 2.2 - 5.7 -6) 895r <4(-8 6.9 -6) 6.8 - 7.0 -6) 90Sr <2(-8 7.8-7) 7.2 - 8.4 -7) 91Y <8(-8 3.1(-6) 2.9 - 3.4(-6) a 95Zr <4(-5) <2(-3) ssNb <2(-5) <1(-3) 103Ru <8(-6) <1(-4) 110 mag 2.0(-6) 1.5 - 2.5(-6) 6.2(-6) 4.0 - 8.4(-6) 124Sb 4.0(-7) -2.9 - 5.1(-7) 6.1(-6) 0.11-1.1(-5) 140Ba <3 -5) 1.3 'i 0.4(-4) 141Ce -<3-7) <6(-6)

'144Ce <2 -6) <3(-5) 147Pm' 2.0(-7) 1.5 -2.5(-7) 4.4(-7) 1.9 - 6.8(-7)

• • ~ Detected in one measurement, only, for this radionuclide.

38

water storage tank (SIRWT), a tritium balance was estimated for these compartments during the refueling outage. The refueling cavity and SIRWT activity values.are listed in Tables A.16 and A.15 in the Appendix.

Table 3.8 gives these tritium balance estimates and the volumes of the

> respective compartments at the four different times during the refueling outage, i.e., before refilling the refueling cavity (10/12-10/20/76),

after filling the refueling cavity but before spent fuel movement (10/27-11/1/76), after spent fuel movement but before draining the refueling cavity (11/8/76), and af ter completion of refueling and draining of the cavity (12/2-12/3/76). Note that during the refueling outage the reactor coolant total tritium content decreased by about 3 Ci while the total tritium in both the SIRWT and spent fuel pool increased by about 2 C1.

This discrepancy may be related to the sampling technique (dip samples) used in the reactor cavity and the fuel transfer canal.- Sampling techniques and locations for spent fuel pool and refueling cavity are described and shown in Section 2.2.3 and Figure A.12, respectively.

3.4.5 Radionuclide Concentrations in Reactor Coolant Drain Tank ,

l The sample point for the reactor coolant drain tank was on the tank and within containment. Because access to contain-ment was limited, only one sample was taken. This sample was obtained after refueling. Table A.19 in the Appendix presents the radionuclides observed in this one sample. Figure 2.2 presents information on reactor component system feeding liquid radwastes to the reactor coolant j drain tank. Section 3.7 presents information on the flow of liquids through the reactor coolant drain tank.

3.4.6 Radionuclide Contentrations in Waste Holdup Tanks Table 3.9 presents a summary of the radionuclide concentrations measured in the waste holdup tanks before, during, and after refueling.

Data summarized in Table 3.9 are presented in Tables A.20, A.21, and A.22 of the Appendix. Figure 2.2 shows which plant components feed liquid radwastes to the waste holdup tanks. The liquid flows through the waste

- holdup tanks during the measurement period are presented in Section 3.7.

Samples taken from the waste holdup tanks before refueling were very different in measured radionuclide concentration, so the data from o these samples were not averaged but presented separately in Table 3.9.

The sample taken on 9/2/76 was from w a te holdup tank WD-4A. An inspection of the auxiliary building operator's log showed that Tank WD-4A had a large inflow of water between 22:00, 8/30/76 and 02:00, I

39

TABLE 3.8 TRITIUM BALANCE DURING CEFUELING Volume 3H Concentration Total 3H Compartment (gal.) (pC1/ml) (C1)

Before Filling Cavity Reactor Core 34,600 3.47 0.04 - 4 SIRWT 314,000 9.25 0.04 - .11 Fuel Pool 215,000 7.20 t 0.03 - 6 ,

Sum of All Components 563,600 ---

21 Filled Cavity, Before Fuel Movement Core plus Cavity 284,100 1.14 0.02-2) 12

. SIRWT 64,500 9.25 i 0.04 -3) 2 Fuel. Pool 215,000 7.32 0.05-3) 6 Sun of All Components 563,600 ---

20 Filled Cavity, After Fuel Movement Core, Cavity, Fuel Fool 499,100 1.1 0.2(-2) 21 SIRWT 64,500 9.25 0.04(-3) 2 i Sum of All Components 563,600 ---

23 Drained Cavity Core 34,600 1.06 0.02 - 1 SIRWT 314,000 9.91 0.04 - 12 Fuel Pool 215,000 9.12 0.04 - 7 Sum of All Components 563,600 ---

20 l

l l

I a . ., .

l

~

I TABLE 3.9 )

RADIONUCLIDE CONCENTRATIONS IN WASTE HOLDUP TANKS During Refueling Samp'es: 16:00; 10/11/76 20:30; 10/11/76 Before Refueling After Refuelina 23:55; 10/11/76 13:34; 8/24/76 11:40; 9/2/76 10:40; 2/11/77 Mean Range Nuclide (pCf/ml) (uC1/ml) (pC1/ml) (uci/ml) (uCi/ml)

[ Indines:

1311 5.06 i 0.07(-3) 1.27i0.01(0) 1.20 0.01(-2) 1.4(-2) 0.77-2.4(-2) ,

1.2 i 0.5(-5) ** '

s 133I 6.9 i 0.2(-5) c.29 i 0.01(-1) 2.9 0.3(-4) 1351 <2(-6) 2.7 0.4(-3) <4(-5) <4(-5)

Cssiums:

1 134Cs 1.78 0.08 -3) 1.56 i 0.05(-1) 1.08 0.01(-2) 7.0(-3) 0.058-1.1(-2) 136Cs 2.3410.05-4) 4.6 0.2(-2) 9.6 1.2(-3) 0.64-2.1(-3) 137Cs 1.86 i 0.06 -3) 1.47 0.05(-1) 1.23 3.8(-6) 0.01(-2 ) 2.7(-4) 0.95-5.9(-2)

Tritium: ,

l

• 5.56

• 3H 0.02(-2) j Other Nuclides:

• 6.89

• 14C 0.03(-6) 24Na 4.9 2.4(-7) <3(-5) <4(-6) <8(-6) s1Cr <7(-4) <5(-2) <1(-3) 1.3(-4) 1.1-1.4(-4) 54Mn 8.0 t *1.2(-6) 1.7 0.8(-4) 1.36 0.03(-4) 3.9(-4) 1.2-6.5(-4) ssFe 1.43

• 0.02(-4) 59Fe <6(-7) <2(-4) 1.8 i 1.5(-6) 5.3 -5) 3.7-6.9(-5) 57C0 <8(-7) 1.1 i 0.4 - 6.9 1.7(-6) 1.6 -5) 1.3-1.8(-5)

SCCo 6.220.1(-5) 1.4 1 0.1 - 9.30 0.08(-4 4.9 -3) 0.55-8.2(-3) 60C0 6.1 1*0.4(-6) 2.6 1 0.7 -

4.6 0.1(-5) ) 1.7*-4) 1.1-2.5(-4) 6 3Ni 1.51 0.02(-4) 65Zn <6(-7) <1(-4) <2(-5) <2(-5) 893p *

• 4.9

• 90Sr 91y 1.13 0.2(-5))

0.06(-6

• 1.7 0.1(-7) 95Zr <3(-6) <3(-4) <2(-6) <6(-5) 9sNb <1(-6) <1(-4) <9(-6) <4(-5)

, 99Mo 9.2 i 0.7(-6) 1.9 1.0(-3) 4.7 3.1(-5) 5.2( ) 4.5-5.8(-5) lo3Ru <6 - <2 - <1(-4) <4(-

110 mag <6 - <2 - 1.7 1.1(-6) <4(-

124Sb <4 - <6 - 1.4 0.5(-6) 5.3i1.9(-6) **

140Ba <2 - <6 - <4(-4) 5.7 x 1 7(-5) 14cLa 1.1 1 0.2(-6) 6.7 2.9(-5) 1.9 0.1(-5) 2.9( ) 2.8-3.0(-5) 141Ce <2(-6) <3 - <7(-6) <2 -

143Ce 2.6 0.8(-6) <1 - <3(-4) <4 -

144Ce <6- <1 - 7.3 2.0(-5) <6 -

187W e4 _ <4 <3(-5) 8.0 i 2.0(-5) 2:9Np <5 - 3.5i1.2(-3) 2.3 1.8(-4) <2(-3) o  : Analysis not perfonned for radionuclide.

• • : Detected in one measurement, only, for this radionuclide. l 41

4

.8/31/76 and then did not receive any additional liquids prior to sampling.

~

+

on-9/2/76. Also,1according to the auxiliary building operator's log, the spent resin from the CVCS deborating domineralizer, CH-98, was sluiced

, to the spent resin storage tank at 02:00,8/31/76. Normal procedure was i to send the resin sluice water to the spent regenerant tanks. However, the

auxiliary building operator's log showed that-the resin sluice water was .,

put in waste holdup tank WD-4A.- The high cesium and iodine concentrations 9 were representative of.the sluice water from a CVCS resin which had been'  !

in contact with reactor coolant. After refueling, only one of.the waste

. holdup tanks (WD-4A) contained arty radwaste liquid. This tank was sampled once-(see Table 3.9)~. ]

3.4.7 Radionuclide Concentrations in Spent Regenerant Tanks l ,

f Table 3.10 presents mean values and ranges of radionuclide .

concentrations for. the spent regenerant tanks for the' three measurement periods. The data summarized in Table 3.10 are presented in Tables A.20, A.21, A.23, and A.32 of the Appendix. Figure 2.2 shows which plant components -

feed liquid radwastes to the spent regenerant tanks. The liquid flows

! through the spent regenerant tanks during the in-plant measurements are presented in Section 3.7.

- The data indicate a large variance in spent regenerant tank activity I

before and after refueling and also within these operational periods, i -- Higher levels for radionuclide concentration before refueling were due to -

the influx 'of reactor coolant wastes from the deborating demineralizers being used to remove the boron shim during coastdown to refueling. Al so, resins from both of the CVCS deborating demineralizers'were sluiced to the spent resin storage tank during the measurements made prior to refueling. (The spent ~ regenerant tanks, except in the case noted in

Section 3.4.5, collected the spent resin sluice water.)

l 3.4.8 Radionuclide Concentrations in Hotel Waste Tanks Mean values and observed ranges for radionuclide concentrations -

I measured in the hotel weste tanks are presented in Table 3.11. The data '

1 summarized in Table 3.11 are presented in Tables A.24 and A.25 of"the Appendix. The hotel waste tanks receive liquid radwastes from the laundry .-

i drains plus the personnel decontamination shower and lavatory sink drains.

! Section 3.7 presents the liquid radwaste flow rates for the hotel waste tanks. No samples were collected from the hotel waste tanks during refueling.

i .

3.4.9 Radionuclide Concentrations in Monitor Tanks l As indicated in Figure'2.3, the monitor tanks at Fort Calhoun l-receive liquid radwaste from the hotel waste tanks and the radwaste ,

evaporator distillate. Water from the monitor tanks is released to the dischcrge pipe after activity measurement to. assure that release limits are not exceeded. ' Liquid flow rates through the monitor tanks are discussed in Section-3.7.

I~

42

v. .-

h.

TABLE 3.10 RADIONUCLIDE CONCENTRATIONS IN SPENT REGENERANT TANKS Before Refueling After Refueling During Refueling Samples: 10:48; 08/24/76 Samples: 10:20; 02/11/77 Sample: 12:35; 10/13/76 09:42; 08/26/76 09:11; 02/14/77 11:50; 08/31/76 07:30; 2/18/77 Mean Range Mean Range Activity Nuclide (uCi/ml) (uCi/ml) (uC1/ml) (uC1/ml) (uCi/ml)

Iodines:

1311 8.0f- 0.078- 1.7(-2) 5.5(-4) 0.046.1.1(-3) 5.46 1 0.08(-4) 133I .4.3f3) 3) 0.014- 1.3(-2) 1.8(-4) 0.024 - 3.2(-4) <4(-4)

• • <8(-2) 0 13sl 1.6(-3) 0.065- 3.0(-3) 1.6

• 0.4(-4)

Cesiums:

134Cs 1.4(-3) 0.33 - 2.2(-3) 6.6(-4 0.24 - 1.5(-3) 1.92 0.03(-3) 136Cs 2.2(-4) 0.53-4.0(-4) <3(-6) ) 9.2 0.2(-5) 137Cs 1.5(-3) 0.35 - 2.6(-3) 7.7(-4) 3.30 - 1.7(-3) 2.18

• 0.03,(-3)

Tritium:

3H

• 4.14i0.02(-3)

Other Nuclides ,

14C

• 2.5 1 0.1(-7) 24Na 2.0(-5) 0.25 - 3.4(-5) 2.1 - ) 1.6-2.6(-5) <4(-4) 51Cr 1.7i0.2(-5) ** 3.8 1 0.32 - 7.2(-5) 4.0 1 0.7A 5 s4Mn 6.7(-5) 0.17 - 1.2(-4) 3.0-) 2.7 - 3.2(-5) 3.04 0.04C-4))

ssFe

• 1.72i0.02(-4) 59Fe 2.1 - 0.31 - 3.8(-5) 3.6(-6) 1.4 - 7.4(-6) 1.40 2 0.04(-5) s7Co 7.5 - 0.12 - 1.0(-5) 2.9 (-6) 2.2 - 3.4(-6) 1.39 1 0.04(-5) seCo 3.5 - 2.7 - 4.8(-4) .4.6(-4) 3.4 - 5.8(-4) 3.35 1 0.07(-3) 4

r-----------

TABLE 3.10 (cont'd)

RADIONUCLIDE CONCENTRATIONS IN SPENT REGENERANT TANKS Before Refueling After Refueling During Refueling Samples: 10:48; 8/24/76 Samples: 10:20; 2/11/77 Sanple: 12:35; 10/13/76 09:42; 8/26/76 09:11; 2/14/77 11:50; 8/31/76 07:30; 2/18/77 Mean Range Mean . Range Activity Nuclide -(uCi/ml) (uC1/ml) (pC1/ml) (uCi/ml) (u_Ci/ml)

Other Nuclides: (cont'd) 60Co 2.7(-4) 0.23-6.7(-4) 2.5(-5) 1.2-3.5(-5) 5.34 0.09(~4) 63Ni 8.57 0.02(-5) **

• 65Zn 1.5(-5) 0.54-2.4(-5) <5(-7) 893p * **

8.5 i *0.8(-6)

- 5.0 0.7 -

. 90Sr 1.5

• 0.3 - ** *

• 91Y

• 4.0

• 2.0 - **

• 95Zr 2.810.6(-6) 4.0 i 1.0 - ** 5.3 0.8 -

ssNb 3.1 10.4(-6) **

<6(-6) 6.3 1 0.4 -

99Mo 1.2(-5) 0.62-1.9(-5) 6.5 i 4.1(-7) ** 4.9 0.9 -

10 3Ru <7(-7) <3( 6 <5(-5) 110 mag 5.3(-6) 2.0-8.6(-6) 2.4 - ) 0.19-5.2(-6) 4.4 0.8(-6) 124Sb <2(-7) 4.4 - ) 2.4-6.5(-7) 4.8i1.1(-7) 140Ba <5(-6) <1 - <2(-4) 140La 5. 7(- 7 4.6-6.8(-7) <2 - 1.80

• 0.04(-5) 141Ce <1(-6)) <4 - <4 -

143Ce 4.711.0(-6) **

7.7(-6 0.47-1.1(-5) <3 -

144Ce <6(-6)

<2(-6)) <2 -

187W 2.010.4(-4) 2.0i1.0(-5) ** <5 - '

239Np 3.611.8l-4) **

5.3 i 1.9(-6) ** <7 -

• : Analysis not perfomeu for radionuclide.

• • : Detected in one measurement, only, for this radionuclide.

. .- ^ e

=

TABLE 3.11 RADIONUCLIDE CONCENTRATIONS IN HOTEL WASTE TANKS After Refueling:

Before Refueling: Samples: 13:15; 2/15/77 Sample: 10:43; 8/24/76 12:04; 2/17/77 Activity Mean Range Nuclide (uci/ml) (uci/ml) (uci/ml)

Iodines:

131I 1.2 i 0.5(-7) 2.7*0.6(-7)

> 1331 <1(-7) <2(-7)

Casiums:

134Cs 1.1 0.1(-6) 2.2(-5) 1.0-3.5(-5) 136Cs < g(.8) <1(-7) 137Cs 1.69i0.09(-6) 2.7(-5) 1.1-4.2(-5)

Tritium:

3H 6.0(-6) 4.0-8.0(-6)

Other Nuclides:

14C 6.5(-7) 0.16-1.1(-6) 51Cr <4(-7) <5(-7) 54Mn - 3.2 1* 0.5(-7) 3.0-)

2.5 1.8-4.1(-7) 55Fe ) 2.0-3.0(-6) s9Fe <2(-7) <l(

57Co <1(-7) <5(-

5800 1.42 0.08(-6) 2.3 - 2.3-2.3 -

60Co 5.9 i *0.7(-7) 3.9-5.3 -

63N i 65Zn <2(-7) 4.6 1.1 - - ))l

<2( 7 0.2-2.0 -

893p

• 2,2 - ) 1.3-3.0(-8) 90Sr * <8-91y * <2-

" 95Zr <2- <3 -

9sNb <8- <1-99Mo <8- <5 -

103Ru <9- <2 -

3 110 mag <1- <2 -

124 Sb <1 - <2 -

140Ba <4- <8 -

140La <2 - <2 -

141Ce <2- <7 -

143Ce <1 - <1 -

144Ce

<7- <7 -

187W <3 -7) <4 -

239Np <3 -7) <3 -

• : Analysis not performed for radionucilde.

• • : Detected in one measurement, only, for this radionuclide.

45 1'

Table 3.12 presents the mean values and ranges for radionuclide concentrations measured in the monitor tanks during the three in-plant measurement periods. ~ The data for these tanks are presented in Tables A.2, A.21, A.24, A.24A, A.26, and A.27 in _the Appendix. .Several times during the measurement period, OPPD collected samples of monitor tank 1- - water at the same time our samples were obtained. Comparisons of OPPD's and our results for these samples are presented in Tables A.24A, A.26 -

and A.27. .The comparisons were good (within 10-20% in most cases).

3.5 Auxiliary Building Ventilation System Source Terms Previously (Section 2.3.1), the sampling stations and locations for the auxiliary building ventilation. system measurements are described.

The radionuclide release rates measured, at sample stations 1 through 4

-are presented in Table A.35 through A.38 of the Appendix. Local area sampler data are given in Tables A.39 through A.41 in the Appendix. The local areas sampled were the waste evaporator room, pipe penetration room, and the letdown heat exchanger room. The waste evaporator room was a component of sampling system 2, while the pipe penetration room and the letdown heat excharger room werecomponents of sampling system 1.

The waste evaporator room sampler was in operation throughout the measurements at Fort Calhoun. The pipe penetration room and letdown heat exchanger room samplers were operated only near the end of the l ~ measurement period. The latter two sampler stations (Fig. 2.4b) were installed in an attempt to better define the source tenn in system 1.

Inspection of Tables A.35-A.38 show systems 1 and 2 to be the primary ventilation system sources for 1311. System 2 was not divided into additional components other than the waste evaporator room and system 4 components (Fig. 2.4b).

Corresponding sampling interval data (Table 3.13) for the pipe pene-tration room, the letdown heat exchanger room, and sample station 1 show i

the component rooms sampled were not the main source in system 1.

Consequently, the major source in system I would have to be one of the other components in system 1 (Table 2.4). Based on discussions with OPPD personnel, it is believed the plant sampling room was the major source in system 1.

• As mentioned above, the waste evaporator room and sampling system 4 were components of sampling system 2. With the exception of two sampling intervals (9/3-9/9/76 and 9/23-9/30/76, Table 3.14) sampling ~

station 4 was the major contributor to 1311 released via system 2.

No individual components of system 4 (Table 2.4) were sampled. However, i

based on discussions with OPPD personnel, it is believed that due to the frequency of charging pump leaks., the charging pump room was the primary source in this system. .

i 46

TABLE 3.12 RADIONUCLIDE CONCENTRATIONS IN MONITOR TANKS Before Refueling: After Refueling: During Refueling:

Samples: 13:?5; 8/24/76 Samples: 09:30; 2/10/77 15:15; 8/24/76 08:14; 2/14/77 08:17; 9/1/76 08:49; 2/19/77 Sample: 07:30; 10/14/76 Mean Range Mean Range Activity Nuclide (pC1/ml) (pC1/ml) (uCi/ml) (pC1/ml) (pCi/ml)

Iodines:

1311- 1.6(-5) 1.2-2.2(-5) 4.2(-7) 2.5-5.2(-7) 6.4 0.3(-7) 133I 5.5(-7) 1.8-8.9(-7) <2(-7) <4(-7)

Cesiums:

D 134Cs 2.6 - 1.1-3.7 - 4.8(-6) 3.3-7.0(-6) 3.8 i 0.2(-6) 136Cs 3.9 - 1.0-7.1 - <2(-7) 2.5 0.6(-7) 137Cs 3.0 - 1.4-4.1 - 5.9(-6) 4.1-8,5(-6) 4.59 0.07(-6)

Tritium:

3H 3.75 1 0.02(-3) 5.36 i 0.02(- 3)

Other Nuclides:

14C 2.6

• 0.2(-7) ** 3.5i0.1(-7) 32P <4(-8) *

• 35S <2(-7) ** *

• 4sCa 1.79 i 0.02(-5) 51Cr <2(-6) <7(:7) <6(-7) 54Mn 2.7-8.5(-7) 2.4(-7) 2.0-2.7(-7) 4.2 0.3(-7) ssFe 5.2 5.6(-7) 0.1(-6 ) ** 5.0 i 1.0(-7) **

59Fe <1(-7) <2(-7) <2(-7) 57Co <2(-7) <4(-8) 8.6 1.9(-8) 58C0 3.1(- 0.88-5.0(-6) 2.2(-6) 1.5-2.6(-6) 4.7 60C0 7.2(- 5.9-8.2(-7) 3.8(-7) 3.6-3.9(-7) 1.05 0.1(-6))

0.04(-6 63Ni 3.6 i 0.1 7) ** 4.310.1(-7) **

65Zn <1(-7) <2(-7) <2(-7)

-_ - ~. .- . - -.

TABLE 3.12 (con't'd)

RADIONUCLIDE CONCENTRATIONS IN MONITOR TANKS L Before Refueling After Refueling: During Refueling:

l Samples: 13:25; 8/24/76 Samples: 09:30; 2/10/77 j 15:15; 8/24/76 08:14; 2/14/77 l 08:17; 9/1/76 08:49; 2/19/77 Sample: 07:30; 10/14/76 Mean Range Mean Range Activity Nuclide (uC1/ml) (uCi/ml) (uCi/ml) (pCi/ml) (pCi/ml)

~

- Other Nuclides: (cont'd) 895r 3.0 i 1.0(-8) **

<2-8)

• 9aSr <2 - <1 -8)

• 91y <1 <2

• 95Zr <1 - <3-9sNb <5 - <6(-7) co

<2 -

99Mo <2 - 1.7 i 1.6(-8) ***

<3l(-<3 -7) 103Ru <1 - <2 - <2I-110lMg <9 - <2 - <2d,-

12'Sb <9 -8 <2 - <21 -

140Ba <4 -7 <6 - <7 140La <8-7) 9.2 i 4.0(-7) ***

141Ce <3(-7) <6(-8)

<3idl-

<7I ,-

143Ce <4(-7) <2(-7) 144Ce 1.2 1 0.4(-6) ***

147pm

<3(-7) <2tl-*

<3l -

<2(-8)

• 187W <2(-7) <4(-7) 239 Np <1 -6

<1(-6) <5(-7) <4 -7

• : Analysis not done for this radionuclide.

• • -: One measurement, only, for this radionuclide.

• • • Detected in on9 measurement, only, for this radionuclide.

.o u

TABLE 3.13 VENTILATION AIRBORNE ACTIVITIES (uci/sec)

Sample Period One Sample Period Two Pipe Letdown Heat Pipe Letdown Heat Penetration Exchanger Sample Penetration Exchanger Sample Sample Room Room Station #1 Room Room Station #1 P riod 2/3 - 2/10/77 2/3 - 2/11/77 2/_3 - 2/10/77 2/10 - 2/11/77 2/11 - 2/17/77 2/10 - 2/17/77 131I 7.9 0.8(-6) 6.8 i 0.4(-6) 1.76 0.07(-4) 1.4 0.6(-6) 3.0 0.5(-6) 1.9 2 0.1(-4) 134Cs <1.2 -7) 1.0 0.4(-7) <2.4(-6) <1.3-6) <1.3(-7) <8.3 -

136Cs <9.9 -8) <1.4(-7) <1.6(-6) <2.6 -7) <9.9(-8) <1.5 -

137Cs <1.2 -7) 2.7 0.5(-7) 2.6 i 0.7(-6) <1.2-6) <3.2(-7) <1.7 -

3H [A] [A]

s1Cr <9.9 - <1.4 - <2.4(-5) <4.6 -6 1 <7.0 - <5.2-)

54Mn <9.9 - <1.4 - <1.6(-6) <4.2 - ;l <1.3 - <6.8 -

59Fe <9.9 - <3.5 - 5.0 t 2.0(-6) <7.5 p <3.0 - <3.0-)J s7Co <1.6 - <3.5 - <7.2(-7) <3.0 - <6.A ,-8) <1.4I 7;l 58C0 <9.9 - 2.5 0.6(-7) <1.6(-6) <3.2 - )) <2.0L-7) <2.41,6J 60Co 1.0 0.2(-6) 2.7 0.6(-7) 6.0 2.0(-6) <1. 31'- <1.11 -6l}

65Zn <7.9(- ) <2.8 - ) <8.0 - ) <5.0-7l)l

<5.2 -6 ' <2.0ll- <2.1I ;- I l 95Zr <2.0 - <2.8-;l <3.2 -6 I <3.4 <1.3 - <1.31,- l -

95Nb <9.9 - <1.4 - I <1.6-6;J <2.0 - <1.3 (I-

<7.9 I; l 103Ru <1.2 - <1.4 - h <1.6 - 1 <3.0 - <1.2I- <1.0f'l 106Ru <1.2 - <1.4-l <1.6 'l <1.1 - <1.3f- <1.3f'l lion %g <1.2 - <1.4 - )l <1.6-l } <2.8 - <2.5ll- <1.2f-ll 124Sb <1.2 - <1.4 - i <1.6 - l <7.3 - <1.4I -

<1.4f - l 12sSb <4.0 - <2.8 - <4.0 'l <5.2 - <2.8f- <2.4f -

140Ba <5.9 - <3.5 - <5.6 - <5.7 - <4.8f -

<2.6f -

ll 140La <5.9 - <5.7 - <1.6 - <5.0 - <1.3f- l 141Ce <4.0 - <6.4 - <1.6 - <8.7 - <2.5d- <1.7(f-

<7.9 - l 152Eu <4.0 - <4.2 - <7.2 - <1.5(-6 < 3. 2 t'-7) <4.9(-6 154Eu <1.2 - <4.2 - <1.6 - <8.3(-6 <3.3(-7) <3.3(-5;l )

• Not Measured

[A] SAMPLE STATION #1; 14C 3H Shutdown due to electrical circuit overload.

TABLE'3.14 VENTILATION 131I AIRBORNE ACTIVITIES (pC1/sec)

Date Waste Evaporator Room Samling Station #4 Sa mling Station #2 9/3 - 9/9/76 1.63 1 0.01 - I 6.13 0.06(-4) 9/9 - 9/15/76 4.70 0.05 - l <4.5(-6)(-4) 1.6 0.1 1.92 t 0.06(-4) 9/15 - 9/23/76 4.80 1 0.04 - l 8.4 0.7(-5) 6.0 2 0.2(-4)[A]

8.3 0.4(-5)

'9/23 - 9/30/76 1.92 1 0.03 - l <3.2(-6) 9/30 - 10/7/76 1.55 i 0.03 ,1 2.1 0.4I - l 1.4310.01(-3) .

-10/7 - 10/14/76 1.91 0.02 'l 2.3 0.1I - ll' 3.80 0.02( 4) 10/14 - 10/22/76 3.21 l 7.3 0.7I l 3.4 0.3(-4 10/22 - 10/28/76 4.37 0.06-ll[B]

0.07 - 7.6 0.6I - 1.66 0.011 10/28 - 11/4/76 [C] 7.8 0.7i ,, h) 1.57 i 0.011,-

11/4 - 11/10/76 1.51 i 0.04(-~5) 8.0 1 .01,- ) 5.27 0.04I,- ,

8.0 1.0l -5,1 6.47 1 0.04I;-

oi 11/11 - 11/23/76 1.22 0.01 5) 0.61;,-5, o 11/23 - 11/30/76 6.0 1 0.1 - 8.0 J 1.75 0.01I -

11/30 - 12/15/76 6.0 1 2.0 - 1.8 0.2-ll 1.06 1 0.091!- [D]

12/15/76 - 1/6/77 8.0 2.0 - 7.8 0.5 il 6.6 i 0.4 -

1/6 - 1/20/77 6.5 0.7 5.0 0.4 -

1/20 - 2/3/77 <1.1(-7) (-6) 2.5 i 0.4 2.3 0.1 - ')l 1.8 i 0.2 -

2/3 - 2/16/77 1.03 0.07(-6) 5.0 2.0 -6) 2.1 i 0.2 -5)

[A] sample from 9/15 - 9/18/76.

[B] Sample from 10/22 - 10/27/76.

[C] No sample due to sampler motor malfunction.

[D] Sample from 12/1 - 12/15/76.

O .b

i It should be pointed out that the total releases from the waste evaporator room and sample station 4 were at times greater than the release rate of station 2. This is attributed to the before mentioned indeter-minate errors (estimated to be 20% overall) associated with varying duct flows.

3.6 Gaseous lodine Species 3.6.1 Auxiliary Building Ventilation Iodine Species The iodine species distribution in the Auxiliary Building ventilation air was variable (Tables A.42-A.44 in Appendix). The average percent values are shown in Table 3.15 along with their ranges. The usefullness of averaging these data is questionable when the wide range of values are considered. Furthemore, there is no oovious relationship between species and the environment in which the measurements were conducted, i.e., high temperature and humidity of the waste evaporator room versus lower humidity and temperature of sampling systm 2 components.

3.6.2 Waste Gas Processing System-Chemical Species of Radiciodine In the waste gas processing systed (i.e., waste gas decay tanks and various cover gases), the predominate iodine species was found to be organic iodine. One explanation for this is that the radiation fields in these gas volumes were high and could cause the fomation of free radicals. Any organic free radical, from hydrocarbon impurities in the gas, could form organic iodine. Secondly, elemental iodine has a much greater tendency to plate out on surf aces than does organic iodine.

The remaining species in the gas phase would therefore be enriched in organic iodine. This latter phenomenon causes the ot.servation that the older a sample containing gaseous iodine is, the more likely it is to be highly organic. As shown in Table A.46, when the short-lived iodine-135 nuclide was observed, iodine was less converted to organic iodine than when 135I was not observed.

l 3.7 Liquid Waste Flow Rates At Fort Calhoun, information coverning specific sources of liquid radwaste flow was not collected by picnt personnel. However, there wt re data from the auxiliary building operator's log which presented liquid radwaste tank levels on a bi-hourly basis. From the log, average daily flow rates into the tanks were calculated for the three in-plant measure-ment periods. Table 3.16 presents the average flow rates into the liquid radwaste tanks together with the source of radwaste liquids foi each tank.

Graphs of liquid radwaste tank levels are presented in the Appendix (Figures A.17, A.18, A.19, A.20, A.21, and A.22).

3.8 Detergent Wastes At Fort Calhoun, detergent wastes are collected in the hotel waste i tanks. Table 3.16 presents the average daily flow rates for the hotel waste tanks. The radionuclide concentrations in the hotel waste tanks 51

+

I' TABLE 3.15 AVERAGE FRACTIONAL PERCENTAGE FOR 131I SPECIES i

Particulate Filter I2 Hol Organic Number Average Ranoe Average Ranoe Average Range Average Range Station of Samples (%) (%) (%) '(%) (%) (%) (%) (%)

1. 2 17 6.6 (0-25.0) 33.3 (6.6-61.3) 15.9 (1.1-41.9) 44.4 (23.3-68.2)

Waste Evaporator 15 12.1 (0-33.5) 38.7 (1.2-75.0) 17.5 (0-80.2) 31.6 (8.1-56.0)

E m

/1 _ _ fl .

. o. " "

i TABLE 3.16 LIQUID WASTE FLOW RATES.

Into Radwaste Tanks ,

~

Flow Rates (Gal / day)

Before During After Refueling Refueling Refueling Tank Source of Liquid

• 8/2/-9/3/76 10/1-16/76 1/31-2/25/77 Reactor Coolant Reactor Coolant-Leakage. ** 250. 500 Drain Tank Fuel Pool Drains

Waste Holdup- Reactor Coolant Drain Tank 2500 6800 690 Tanks Aux. Bld. Equip. Drains LVCS Bleed

. VCT Relief and Drains

. Spent Regenerant Aux. Bld. Floor Drains 3800 2400 2500 g; Tanks Secondary Side Drains Spent Resin Sluice Water Deborating Demins Containment Sump Hotel Waste Tanks Leundry Drains 1100 2500 570

! Shower and Hand Sink Drains i

Monitor Tanks Evaporator Condensate 7000 4600 3600 Hotel Tanks

• : See Liquid Radwaste Schematics, Figure 2.2 and 2.3.

• • : Information not collected.

i

are presented in Table 3.11. Liquid radwastes from equipment, drum, and '

cask decontaminations fed into the drain headers in the auxiliary building and were collected by the spent regenerant tanks.

3.9-Chemical Wastes from Regeneration of Condensate Demineralizers There are no demineralizers in the secondary system for processing c the condensate. . The steam generator blowdown waters are nomally released to the discharge pipe without treatment. Also, the nomal. practice was to dispose of all demineralizer spent resins as radioactive solid waste.

C 3.10 Containment Purge Frequency The containment purge frequency at Fort Calhoun via high volume purging av.t venting is 37 times annually; however, the approved plant procedures for containment purge releases do not differentiate between containment purging and containment venting. (8) 3.11 Containment Internal Cleanup System Due to limited access to the containment building, measurements I were not made for evaluation of the containment cleanup system.

In the Fort Calhoun containment building there are two banks of charcoal adsorbers dth associated HEPA filters. The design flow through each unit i's 90,000 cubic feet per minute which results in 5.4 containment j volume changes per hour. These units were operated prior to entry into containment of maintenance personnel and/or when high volume purges were done.

3.12 Gaseous Leakage Rate to Containment Building Two samples (Table A.51) of the containment building atmosphere were obtained. One was taken during the refueling outage (Oct.1976) approximately 10 days af ter purging of the containment atmos-phere. .The second was collected during power operations in February,1977.

The former was taken from a plant installed containment penetration used for pressure testing of the containment building. The latter .

sample.was taken inside the containment structure, consequently, it is the more representative sample. During October, the measurement utilized only an iodine species sampler, while during February an iodine species and MC and 3H samplers were used. The additional tritium data ,

l available for the February measurement allowed the leakage rate to be calculated on the basis of both iodine and tritium.

l-l The 'results were calculated using the following formulas (3):

l (1) gC xvxK=T c c Tg x D'

(2) Tc" in 2 54

(3) D' x 100% = 0 x C xKxV L L where:

C = Airborne concentration of nuclide in containment - pCi/cc g

b Vc = Free volume of containment - ft3 K = Conversion factor = 2.83 X 10 cc/ft3 T = Total activity in containment atmosphere - pCi c

T = Half-life of nuclide - days D' = Leakage - pCi/ day Cg = Concentration of nuclide in primary coolant pC1/cc Vg = Volume of primary coolant - ft3 D = Percent of nuclide inventory in primary coolant release per day.

Iodine-131 data from the October measurement indicate a daily leakage rate of the iodine inventory in the reactor coolant to the containment atmosphere of 1.6 X 10-4%/ day. Leakage rates calculated using 1311 and 1331 data from the February measurement averaged 1.35 X 10-6%/ day. The corresponding leakage rate based on tritium (HTO) was 2.5 X 10-7% / day. The higher leakage rate for the October sample j was based on data taken af ter the reactor had been down for refueling l for several days and the containment was secured and was in the process of being pressure tested. Therefore, the mechanism for release of

^

iodine into the containment atmosphere could be drastically different than during the February measurement period when the reactor was at full power.

Data from two other pressurized water reactors (6) indicate a ,

l containment leakage rate of 0.001%/ day. The best estimate of the present leakage into the containment building is en the order of work 1 X 10-fog %/ day.

55

3.13 Auxiliary Building Gaseous Leakage Leakage of primary coolant contributing to the source term of the auxiliary building was calculated for both 1311 and tritium (both oxidized and nonoxidized species). One set of gaseous 1311 and tritium values used for the calculations was taken from ventilation sampling stations 1, 2 and 3 for the sampling interval 9/3-9/9/76 (TablesA.35-A.37).

A Second set of gaseous 1311 and tritium values was taken from the same ventilation sampling stations for the period 1/6-1/20/77 (Tables A.35-A.37 in the Appendix). Ventilation stations 1, 2, and 3 ,

account for 80 percent of the total auxiliary building ventilation exhaust flow and in excess of 80 percent of the total activity exhausting the auxiliary building. Consequently, the calculated numbers could under-estimate the actual leakage by as much as 10 percent. The basis for the latter statement is that approximately one half of the balance of the flow comes from an area outside the personnel monitoring control point, and' has a lower activity. This can be seen from Figure 2.4 and Table 2.3 plus Figures A.28 and A.29 in the Appendix.

Results calculated using 1311 and 3H values from the first set of data indicate primary coolant leakage rates of 570 lbs/ day and 36 lbs/ day, respectively. The 1311 and 3H data obtained in February yield leakage rates of 85 lbs/ day and 77 lbs/ day, respectively. In the iodine calculations an iodine partition factor of 0.0075 (5% volatile iodine in primary coolant with a partition coefficient of 0.15 (6))was used. A partition factor of 1.0 was used for tritium. The formulas used in the calculations were:

sl) C xg60 sec/ min x 1440 min / day = M (2) Cg x 45 g/lb x P" '

where:

Cg = airborne nuclide activity per time pCi/sec M = activity per day pCi/ day l Cg = nuclide activity per unit mass pC1/g P = nuclide partition factor - dimensionless L = primary coolant leakage rate - lbs/ day.

The 1311 value from the first set of data is obviously inconsistent with the other three data results. An explanation of this inconsistency is the unpredictive chemical behavior of iodine, i.e., partition coefficient.

Therefore, it is believed that tritium yields a more valid estimate of the primary coolant leakage rate into the auxiliary building.

56

3.14 Particulate Releases for Gaseous Effluents 3.14.1 Containment Building Radioactive Particulate Releases for Gaseous Effluents i b

Table 3.17 shows the extrapolated annual radioactive particulate releases for gaseous effluents, from the containment structure. j The containment building calculations are based on information stated earlier, i.e., 37 high volume purges annually with ventings being treated as high volume purges. Furthermore, it was assumed that duration of a high volume purge is long enough to reduce the activity to an insignificant level. A DF of 70 (see Sect. 3.18) was used for the containment building exhaust filters. . The containment free volume is 1.05 million cubic feet.

The data (Table A.51)'were handled as shown below: ,

1 (1) Cc*YA Xhx vf=R A

! (2) C, x Pf x x 28320 cc/ft3=VA where:

C c = nuclide Particulate concentration pCi/cc V

A

= total volume of containment atmosphere released - cc/ year j (DF) = decontamination factor

.R g = nuclide released - C1/ year C, = containment free volume - ft3 l P = purging frequency - year ~1 )

f 3.14.2 Auxiliary Building Particulate Releases for Gaseous Effluents .

Table 3.18 presents the extrapolated annual radioactive particulate releases for gaseous effluents for the auxiliary building.

l The data used in the calculations were obtained from sampling stations 1, 2 and 3 (Tables A.35-A.37). For each station, only p'o'sitive values, (i.e., no lower detection _11mits) were used to obtain an average pCi/sec release rate. The average release rate was multiplied by the

- number of seconds per year (3.15 (+7) seconds) to obtain an annual release

-rate before filtration. A DF of 70 (see Section 3.18) was used to obtain the values in Table 3.18.

57

4 T58tE3.17 EXTRAPOLATED ANNUAL CONTAll#4ENT BUILDING PARTICULATE RELEASES FOR GASEOUS EFFLUENTS Nuclide C1/ year [A]

18"Cs 4.6(-8) y 137 Cs 5.8(-8) .

58 Co 8.0(-10) q 80 Co 5.4(-10) 5"Mn. 2.0(-10) 59

• Fe

- Not observed.

[A] A DF of 70 was assumed for HEPA filters.

1 TABLE 3.18 -

EXTRAPOLATED ANNUAL AUXILIARY BUILDING PARTICULATE RELEASES FOR GASEOUS EFFLUENTS Average Release Rate Extrapolated Sum of Stations 1, 2, 3 Annual Releases [A]

Nuclide . (uCi/sec) (Ci/ year) 134Cs 5.1(-5) 2.3(-5) 137Cs 5.8(-5) 2.6(-5) 58C0 6.5(-5) 2.9(-5) 60Co 8.6(-6) 3.9(-6)

S$n

• 59pe *

- Not observed.

~

[A] A DF of 70 was assumed for HEPA filters.

-58

l l

i l

3.14.3 Gas Decay Tank Radioactive Particulate Release $ for Gaseous Effluents Table 3.19 shows the extrapolated annual radioactive releases via the waste gas processing system. The releases via this pathway were calculated for 36 waste gas decay tank releases per year at a maximum achievable discharge pressure of 86 psig and a minimum initial pressure of 10 psig in the gas decay tanks (8). The decay tank volumes are 400 cubic feet. The data in Table 3.19 do not assume a DF for any

> filter system. The calculations are based on a sample collected from waste gas decay tank "A" on 10/13/76. The tank had been isolated for decay on 10/4/76. The parti were 59Fe(1, .1 ),X60Co(2.3 10-gulateX 10-radionuclidgs

), and 13'<s(7.2observed and their X 10- ) microcuries cgcentrati per cubic centimeter. The formula used in the waste gas release calculations ,

was:

o cc x 19 6 C1/pCi R

WG

=C T xT y xR x f x g where:

1 C = nuclide particulate concentration in tank pCi/cc T

Ty = tank volume - ft3 R = release frequency - year 1 i

f R yg = annual nuclide release - Ci/ year TABLE 3.19 EXTRAPOLATE > ANNUAL WAS1E GAS DECAY TANK PARTICULATE ELEASES FOR GASEOUS EFFLUENTS

.. Nuclide Ci/ year 13"Cs 1.2(-6) 137gg . ,  ;

l seCo 3.8(-6) 60

• Co 59 Fe 1.8(-6)

• - Not observed.

[A] Assumes no DF for any filters.

I 59

3.15 Tritium and Carbon-14 Rel' eases 3.15.1 Liquid Tritium Releases As indicated in Figures 2.2 and 2.3, except for the steam generator blowdown waters, all liquid radwastes including secondary side c drains at Fort Calhoun were released via the monitor tanks. Using the '

data from Tables 3.12 and 3.16, the mean liquid tritium releases from the monitor tanks were estimated to be 9.8(-2) C1/ day before refueling and 7.2(-2) C1/ day after refueling which averages to 8.5(-2) C1/ day. 6 Using the data from Table 3.5 plus Sections 3.3 and 3.4.2, the mean liquid tritium releases by the steam generator blowdowns were estimated to be 2.1(-3) Ci/ day. This sums to 8.7(-2) Ci/ day which translates to 3.2(1) C1/ year for the liquid tritium releases.

3.15.2 Liquid Carbor-1% Re1 eases

. At the Fort Calhoun Station, all liquid radwastes except for steam generator blowdown waters, were released via the monitor tanks.

Using the data from Table 3.16 and the Appendix (Tables A.2 and A.26), the mean liquid 14C releases were estimated to be 6.9(-6) C1/ day before refueling and 4.8(-6) Ci/ day af ter refueling. Tbfs averages to 5.9(-6)

C1/ day which translates to 2.2(-3) Ci/ year of liquid carbon-14 releases.

The method used for analyzing these samples measured only inorganic 1"C (CO 2 ), therefore, the results are potentially biaset low by the degree that the sample contained any organic 14C.

3.15.3 Containment Building Gaseous Tritium and Carbon-14 Releases The tritium and 14C release calculation was handled in the same manner as described for particulate releases, i.e., same volumes, purging frequencies, and pressures. Data were taken from Table A.51.

Both oxidized and nonoxidized chemical species were used in the tritium analysis. However, due to a sample analysis problem, only the oxidized 14C species was used. Consequently, the- 14C containment building

- releases could be as much as a factcr Lf 10 higher. The results of the l calculations are shown in Table 3.20. ,

3.15.4 Auxiliary Building Gaseous Tritium and Carbon-14 Releases l The annual tritium and 14C release (both oxidized and non- ,

oxidized species) via the vapor pathway from the auxiliary building ventilation system are given in Table 3.20. The data were obtained by sampling stations 1, 2, and 3 (Tables A.35-A.37). It should be noted that sampling station 2 includes the fuel pool area. Consequently, the calculated releases include tho'se related to fuel movement during the refueling outage. The average total release from sampling stations 1, 2 and 3 were multiplied by the number of seconds in a year (3.15 (+7) seconds) to obtain an annual release rate.

l 60

5 i

i a

TABLE 3.20 EXTRAPOLATED ANNU'AL RELEASES OF GASEOUS TRITIUM AND CARBON-14 9 '

Ci/ year b Auxiliary Containment Building Building Waste Gas Processing System 3H 5.9(-2) 0.9 0.65 ,

18. C 7.8(-2)^ 0.3 0.81

^

Cxidized species only 1

l

> l e

61

3.15.5 - Gas Decay Tank Tritium and Carbon-14 Gaseous Releases

-Both .Mized and nonoxidized chemical forms of tritium and 14C were included in the waste 96 decay tank annual release calculations. The annual releases presented in Table 3.20) are based on average concentrations in the decay tanks (Tables A.49 and A.50 of Appendix). The "less than" (lower detection limit) value for tritium in Table A.50 was not included. e The results were obtained in the same manner as described in Section 3.14.3 above, i.e.', the same release frequencies, volumes, and pressures.

c 3.16 Decontamination Factors for Demineralizers At Fort Calhoun the demineralizers present in the liquid process systems were the purification, cation, and deborating demineralizers of the CVCS and the waste or polishing demineralizers which were a side stream in the recirculation line for the monitor tanks. Nomal plant operating practice was not to use the waste demineralizers, which was continued during the in-plant measurements. According to station personnel, the waste disposal problems associated with the spent resins from the waste demineralizers (i.e., manual removal of the resins) override the usefulness of processing the liquid radwastes from the monitor tanks through the waste demineralizers. This was particularly true since the activity in the monitor tanks was well within the release limits for the station.

The purification demineralizers in the CVCS letdown stream are of i

the' mixed bed type. Table 3.21 presents the input activities and the decontamination factors (DF's) measured for several radionuclides during the August,1976 (before refueling) and February,1977 (af ter refueling) measurement periods. In a case where the input activity contained two components (dissolved and raspended activities), the two components were added and the one str.ndard deviation counting statistics errors were propagated. If one component was not measurable (i.e., only a lower limit of detection was obtained), the quoted input activity contains only the measurable concentration. The error was adjusted to reflect the fact that a second component (of unknown magnitude and possibly as large as the lower limit of detection) could exist by propagating the error in the measurable .

concentration together with an error equal to one-half the lower limit of

• detection (lower limit of detection values are at the two sigma level of counting statistics). The data used to calculate the DF's are presented in Tables A.8, A.9, and A.12. Also presented in Table 3.21 aie when the demineralizer was placed in service, how long it was in use, and bed volumes passed through before measurement, together with the letdown flow rate, demineralizer pressure drop, and reactor coolant boron level at the time of measurement. These parameters were extracted from Figures A.5, A.6, A.7, A.8, A.9, and A.lu.

For the purification demineralizers, the DF's, are defined as the ratios of the input to the output radionuclide concentrations. The errors 62

TABLE 3.21 DECONTAMINATION FACTORS CVCS PURIFICATION DEMINERALI2ER Demineralizer CH-8B Demineralizer CH-8B Demineralizer CH-GA 10:57; 8/20/76 10:00; 8/30/76 10:20; 2/16/77 Placed in Service: 3/24/76 Placed in Service: 3/24/7F. Placed in Service': 1/13/77 Used for:.93 days Used for: 103 days Used for: 34 days Bed Volumes Thru: 3.1(4) Bed Volumes Thru: 3.5(4) Bed Volumes Thru: 1.1(4)

Letdown Flow Rate:28.6 gpm,4.2 gpm/ft2 Letdown Flow Rate:27.4 gpm,4.0 gpm/ft2 Letdown Flow Rate:30.60pm,4.5 ppm /ft2 Demin Pressure: 1.8 rsi Demin Pressure: 2.0 psi Demin Pressure: 1.4 psi Reactor Coolant Boron: 112 ppm Reactor Coolant Boron: 58 ppm Reactor Coolant Baron: 449 ppm Input Input Input Activity Decontamination Activity Decontamination Activity Decontamination Nuclide '(vCi/ml) Factor (uCf/ml) Factor (vCi/ml) Factor Activation:

-o,24Na 3.4i0.2(-) 13. 11, 2.2i0.5(-3 6.1 2.6 1.84 0.05(-2) 59. 12.

" 54Mn 6.3 t'0.3(- ) 1.4 0.1 5.2 0.4(-4 22. 51. 1.8 0.3(- ) 14. 1 4.

58C3 3.44 0.06 -3) 8.6 0.3 2.05 0.09(3) >49. 1.87 0.04 -3 10.1 0.3 60Co 6.3 1.9(-5) 2.6 0.9 8.7 3.1(-5) 2.9 2.4 2.6 15.0( 5)) 0.9 5.1 Irdines:

131I 2.41 0.02(-1) 13. i 1. 2.40 0.02 - 20. 1. 9.0 0.1(-2) 1730. i 1460 132I 7.1 0.1(-2) 14. 3. 1.01 0.01 - 51. 11. 5.0 0.2(-2) >16.

133I 1.20 0.01(-1) 16. i 1. 2.59 0.02 - 25. 1. 1.04 0.02(-1) >170.

13sl 7.3 0.2(-2) 7.4 1.8 1.3010.04(-1) >12. 6.9 0.2(-2) >23.

Other Fission Products:

i serb 4.9 0.2(-1) 2.7 1.1 9.5 0.6(-1) 2.6 1.0 7.2 0.4(-1) 2.3 0.2 134Cs 1.54 0.01(-2) 13. 1. 1.37 0.06(-3) 3.8 0.3 1.91 0.02(-1) 61.6 0.9 136Cs 4.10 0.09(-3) 14. 1. 6.2 1.5(-4) 5.2 1.5 1.4 10. 4.

137Cs 1.56 i 0.01(-2) 13. i1. 1.46 t 0.04(-3) 3.8 0.3 2.20 0.5(-4) 0.02(-1 ) 58.0 0.6 13eCs 2.9 i 0.1(-1 29. 41. 5.6 t 0.2(-1) 11. 5. 2.2 0.1(-1) >110.

140Ba 9.0 i 1.7(-4 5.3 1.4 <3(-4) <2(-2) 140La 2.6 i 0.2(-3 1.3 0.2 1.3 0.5(-3) >8. 2.7 ?.2(-2) >140.

i on the DF's are the propagation of the one standard deviation counting statistics errors on the measured radionuclide concentrations for.the input and output. An additional propagated error of 14 percent should be added to.the stated strors in the DF's to account for other uncertainties _y (e.g.,~ uncertainties in calibration, decay correction, volume measurement, random summing correction). Sampling uncertainties are unknown and

.therefore cannot.be estimated. When a DF is expressed as a " greater than" number this means that the radionuclide was measurable in the demineralizer e input but not in the output and the output lower limit of detection and input values have been used to calculate the DF.

Except for serb, the measured decontamination factors were higher for domineralizer CH-BA in February,1977 than for demineralizer CH-8B in August,1976. The fact that domineralizer CH-8A was newer (used for only 34 days and 1.1(4) bed volumes) than demineralizer CH-8B (used for 93-103 days and 3.1(4)-3.5(4) bed volumes) may account for the higher decontamination factor.

3.17 Decontamination Factors for Liquid Stream Filters At Fort Calhoun there were liquid stream filters in the CVCS letdown stream, purification filters (CH-17A and B), and in the radwaste process stream, waste filters .(WD-17A and B). Sample points were available for investigating the purification filters but not the waste filters. (Table 2.1 presents the physical information on the purification filters.) Table 3.22 presents the radionuclide decontam-ination factors (DF's) for the purification filters measured before refueling (August,1976) and af ter refueling (February,1977). The data used to calculate the DF's in Table 3.22 are found in Tables A.8, A.9, and A.12 of the Appendix. Also presented in Table 3.22 are the letdown flow rates, filter pressure drops, and reactor coolant boron levels at the time of measurement. These parameters were extracted from Figures' A.6, A.7, A.8,' A.9, and A.10 of the Appendix. For the purification filters, the DF's are defined as the ratios of the input to the output radioauclide concentrations. The errors shown were obtained in the same ~

manner as discussed in Section 3.16.

As shown in Table 3.22, there is a large variability in the measured decontamination factors for all the radionuclides except serb which has ,

a measured DF of about 1.0. For the measurement of 2/16/77, many of the radionuclides had a'DF of less than one which means that radioactivity was being eluted from the filter by the letdown stream. . The purification filters are designed to trap loose resin beads from the demineralizers and are not intended for radioactive clean-up, so the lack of consistent DF's is not surprising.-

64

TABLE 3.22 DECONTAMINATION FACTORS CVCS PURIFICATION FILTERS Filter. ' H-17B C Filter CH-17B Filter CH-17A 10:57'; 8/20/76 10:00; 8/30/76 10:20; 2/16/77 Letdown Flow Rate: 28.6 gpm Letdown Flow Rate: 27.4 gpm Letdown-Flow Rate: 30.6 ppm Filter Pressure: 1.0 psi Filter Pressure: 0.5 psi Filter Pressure: 1.0 psi Reactor Coolant Boron: 117. ppm Reactor Coolant Boron: 58 ppm Reactor Coolant Boron: 449 ppm Input Input Input Activity Decontamination Activity Decontamination Activity Decontamination

'Nuclide ~(uC1/ml) Factor (uCi/ml) Factor (uCf/ml) Factor Activation:

24Na 2.6 2.1 4) >2.6 3.6i1.3(-4) >3.6 3.1 0.6(-4) 0.22 0.05 5Mn 4.5 i 0.1 4) 14. 8. 2.4 5.5(-5) 1.6 5.4 1.3 0.3(-5) 0.72 0.17

$ 58C0 4.0 1 0.1 4) >19. <4(-5) 1.85 0.04(-4) 3.4 0.3 60Co 2.4 0.4 -5) 11. 4. 3.0 1 1.5(-5) 0.75 0.53 2.97 0.07(-5) 7.4 0.9 Icdine:

1311 1.92 0.03(-2) 450.i270. 1.18 i 0.02(-2) 59. 6. 5.2 4.4(-5) 0.0080 0.0068 Other Fission Products:

serb 1.8 0.7(-1) 1.2 0.6 3.7 1.4(-1 1.2 1.5 3.2 0.2(-1) 0.97 0.08 134Cs 1.21 0.02(-3) 30. 8. 3.6 2 0.2 - 3.8 2.2 3.10 0.03(-3) 0.190 i 0.002 136Cs 2.9 0.2(-4) >1 ?. 1.2 0.2 - 20 40 1.4 0.2(-5) 0.58 0.13 137Cs 1.21 i 0.02(-3) 12. 5. 3.8 0.3 - 4.1 i 0.8 3.79 1 0.02(-3) 0.201 0.002 13cCs 1.0 i 1.4(-2) 0.5 0.7 5.0 1 2.2 - 1.7 6.8 <2(-3)

. . ~ . . .

l 3.18 Decontamination Factor for Charcoal Adsorbers and HEPA Filters It was not possible to install samples up and downstream of the fuel handling area charcoal adsorber because of the physical arrangement of.the ventilation ducts. For this reason, no measurements were made of- the DF of the charcoal adsorbers. A direct measurement of the DF

~

of the' auxiliary building HEPA exhaust filters was not possible because samplers could not be installed in valid sampling locations downstream 1 of these filters.

A 137Cs DF was calculated for the HEPA filters (installed new April, g l 1976) using the measurements of this work in conjunction witii plant particulate release data for. the fourth quarter of 1976 (8). The data i are presented in Table 3.23. Decontamination factors for other

, radionuclides could not be calculated due to limited corresponding. data.

i The data in Table 3.23 yield a DF, of ~ 70,. .

4 In addition, the plant numbers include releases from the containment

. and/or the waste ges releases. Consequently, a DF of 70 is somewhat conservative ds the containment and waste gas systems do not enter the i HEPA filters and are not measured as a part of the input value, but are included as a part of the output value.

3.19 Decontamination Factors for Evaporators A't Fo'rt Calhoun Station, the radwaste evaporator is used for processing all liquid radwastes except detergent wastes or wastes from 4

the hotel waste tanks. Table 2.2 presents the design information for the radwaste evaporator and Figure'2.3 presents a schematic diagram of the  !

! feed and output streams for this evaporator. Operating practice i was to completely process a full radwaste collection tank with a constant

feed rate (varying from 5 to 13 gpm, nominally 10 gpm). ' The concentrate bottoms were removed in a batch process mode when the bottoms boron ,

concentration reached 8 to 10 percent. An anti-foaming agent (General

Electric Co., Type AF-72) and sodium thiosulfate were added to the waste r

holdup and spent regenerant tanks just prior to processing the tank liquids with the radwaste evaporator. The sodium thiosulfate was added l to keep the radioiodines in solution. For the waste holdup tanks, the anti-foam agent was added to a level of 110 ppm while the sodium thiosulfate

• was added to a level of 80 ppm. For the spent regenerant tanks, the anti-l -foam agent was added to a level of 84 ppm while the sodium thiosulfate

was added to a level of 22 ppm.

. In October,1976, the radwaste evaporator tube bundle was replaced j with a new bundle of the same design as the original one. The old tube bundle had developed leaks such that the distillate water quality could not be properly controlled. 'That is, the conductivity of the distillate became high and uncontrollable (see Figure A,23). The conductivity i .of the distillate is a measure of the carry-over of ions from the

! ' feed and/or bottoms of an evaporator. If an evaporator is working properly,'only water vapor is-transferred from the feed and/or bottoms

( to the distillate. A malfunction of the evaporator occurs when moisture droplets are carried over. from the feed and/or bottoms to the l'

l- -66

..:.--. :-- _ ..-. i_. . . , . . . - -n - , - , - , < , , - . . , - .-. --

/

TABLE 3.23 FOURTH QUARTER 1976 137Cs AUXILIARY BUILDING RELEASE DATA  !

Total of Sampling Stations EAl

?

Date - (1976) 1. 2.~ and 3~ (uC1/sec) 9/30 - 10/07 3.1(-5) 3 10/07 - 10/14 9.8(-5) 10/14 - 10/22 4.2(-5)

.10/22;- 10/28 5.4(-5) 10/20 - 11/04 6.0(-5) 11/04 - 11/10 5.7(-5)

'11/10 - 11/23 1.4(-5) 11/23 - 11/30 6.0(-5) 12/01 - 12/15 2.6(-5) 12/15 - 1/06 1.3(-4)

AVERAGE: 5.72(-5)

Plant Data (8):

137Cs: 6.29(-6) Curies per quarter or: 6.29(-6)' Ci/7.88(+6) seconds [B] x 1(+6) pC1/Ci -

= 7.98(-7) pCi/sec.

[A] Tables: A.35, A.36 and A.37 in Appendix.

[8]~ 365 days / year x 0.25 years x 1440 min / day x 60 sec/ min =

7.88(+6) sec.

67.

.. i

=

distillate. These moisture drt$plets contain ions or radicions and transport these ions to the distillate. This results in a high conductivity of the distillate and reduces the radionuclide clean-up properties of the evaporator. As can be seen by comparing the distillate conductivity data presented in Figures A.23 and A.26, the new tube bundle installed in'0ctober,1976 caused the radwaste evaporator at Fort Calhoun Station to operate in a better manner as far as the distillate conductivity. c Measurements were made of the decontamination factor (DF) for the radwaste evaporator with both the old and new tube bundle for several  !

different radionuclides. The evaporator DF was defined as the ratio #

of the feed radionuclide concentration to the radionuclide concen-tration of the distillate. The evaporator feed and distillate samples along with a sample of the evaporator bottoms were taken at the same time to provide a sample set for measurement of the DF. It is realized that there was a finite time delay between when a particular water molecule of the feed stream reached the distillate stream because of the evaporator operation. However, this time delay was not assessible and the measurements made at Fort Calhoun on the radwaste evaporator do provide the data on how well a radwaste evaporator operating in a nuclear plant enviror. ment will remove radionuclides from liquid radwastes.

Table 3.24 presents the radionuclide feed concentrations and decontamination factors (DF's) measured for three sample series taken i on the "old" tube bundle in August,1976. The data for Table 3.24 l can be found in Tables A.28, A.29, and A.30. Also presented in l Table 3.24 are the identity of the radwaste tank feeding the evaporator, i radwaste tank level, the evaporator feed rate, the evaporator bottom's

boron concentration, and the distillate conductivity at the time of j the measure %.nts. These parameters were obtained from Figures A.17, A.20, A.24 and A.25. No pH measurements were made on the feed liquid

! to the radwaste evaporator. The errors shown were calculated in the

same way as discussed in Section 3.16.

I From the measured DF's presented in Table 3.24, only the radio-l nuclides 54Mn, 60Co, and 1311 have any indications of an increased DF

.with increased feed concentration. The other radionuclides do not show l

this. The observed DF for 58C0 was the highest for the sample series .

j taken on 8/26/76. For this sample series, the spent regenerant tank WD-13A L

was at a fairly low level (800 gal of a capacity of 6000 gal or about 1/8 full).

At Fort Calhoun, the activation or " crud" radionuclide (51Cr, 54Mn, 55Fe, 59Fe, 58C0, and 60Co) with the highest activity level was ssCo. The larger DF for .

58C0 from the 8/26/76 sample series was probably due to a higher fraction of suspended particulate 58Co material in the feed because of a lower tank level than during the sample series.done on 8/25/76 and 8/31/76. The uistillate conductivity for the sample series taken on 8/31/76 was higher, indicating higher moisture and/or ion carry-over to the evaporator distillate. The measured DF's for seCo,134Cs, and 137Cs were lower on 8/31/76 and reflect this higher moisture carry-over. However, the DF's for 54Mn, OCo, and 1311 were not lower on 8/31/76. These apparent discrepancies in relating the DF's with distillate conductivity l

-68

, , # w _ _

TABLE 3.24 EVAPORATOR DECONTAMINATION FACTORS "Old" Tube Bundle - August,1976 Feed: Waste Holdup WD-4B Feed: Spent Regenerant WD-13A Feed: Spent Regenerant WD-138 -

13:43; 8/25/76 09:28; 8/26/76 13:A5; 8/31/76 Tank Level: 34.400 gal Tank Level: 800 gal . Tank Level: 3330 gal Feed Rate: 7 gpm Feed Rate: 5 gpm Feed Rate: 13 gpn Bottoms Boron Conc: 4.6% Bottoms Boron Conc: 5.5% Bottoms B0ron Conc: 6.5%

Distillate Cond: 5.8 umho Distillate Cond: 3.4 umho Distillate Cond: 8.2 umho Feed Feed Feed Activity Decontamination Activity Decontamination Activity Decontamination Nuclide (uC1/ml) Factor (uci/ml) Factor (uC1/ml) Factor 54Mn- 8.0 1.2(-6) 26 i 6 1.65 0.07(-5) 80 50 1.20 0.06(-4) 67 4 S SPCo 6.2i0.1(-5) 60 i 4 2.% i 0.01(-4) 330 60 2.74 0.01(-4) 39 1 60Co 6.1 0.4(-6) 12 1 2.3 0.1(-5) 40 10 6.74 0.06(-4) 630 i 40 131I 5.06 0.07(-3) 179 2 4 6.80 0.07(-3) 320 i 10 1.65 0.01(-2) 880 10 13Cs 1.78 0.08(-3) 1350 t 90 2.23 0.01(-3) 830 120 1.49 0.01(-3) 368 i 5 137Cs 1.86 0.06(-3) 1110 50 2.57 0.02(-3) 950 180 1.49 i 0.01(-3) 313 3

M may be explained by the feed concentration levels, because higher feed concentrations result in higher DF's. That is, for 54Mn, 60Co.

and 1311, .the feed concentrations on 8/31/76 were up by a factor of 5 to 1D times over the levels on 8/25/76 and 8/26/76. This increased

, feed activity level for 54Mn, 60Co, and 1811 caused the measured DF's to be higher even though the distillate conductivity was higher. The

• i feed concentrations for SOCo,13'+Cs, and 137Cs were essentially the same c for all three sample series. Therefore, for seCo,134Cs,' and 13ts, the variance in measured DF's with distillate ccnductivity were not j complicated by a simulataneous variance of feed concentration.

Table 3.25 presents the radionuclide feed concentrations and decontamination factors (DF's) measured for three sample series taken l on the "new" tube bundle in February,1977. The data for Table

3.25 can be found in Tables A.32, A.33, and A.34. Also presented l in Table 3.25 are the identity of the radwaste tank feeding the l evaporator, the radweste tank level, the evaporator feed rate, the evaporator botton's boron concentration, and the distillate conductivity at the time of the measurements. These parameters were extracted from Figures A.22 and A.27. No pH measurements on the feed to the radwaste evaporator were made. The errors on the DF's are the propagation of the one standard deviation errors on the measured radionuclide concentrations for the feed and distillate.

) For the measurements on the "new" tube bundle, the distillate conductivities were significantly lower than for the August,1976 data on the' "old" tube bundle (1.2-1.3 paho for "new" and 3.4-8.2 paho for "old"). This lower distillate conductivity, which indicates less

! moisture and/or ion carry-over for the "new" tube bundle, is why the DF's 1

were much higher for all radionuclides in February,1977 than in August,1976.

4 For the evaporator DF measurement series taken in February,1977, a spent regenerant tank (WD-13A) was followed from a 75% level down to a 55%

level .while its contents were being processed through the radwaste evaporator 1

at a constant feed rate of 5 gpm. Since the evaporator feed for this measurement series was from the same tank (WD-13A), the feed radionuclide

concentrations for the six isotopes (54Mn, seCo, soCo, 1311, 134Cs, and 137Cs)

! did not vary significantly for the three sample sets (see Table A.32).

i

. However, as is also shown in Table 3.25, the measured DF's for the cobalt

• and cesium radionuclides increased with time that the tank WD-13A had i been feeding the evaporator. This indicates that the longer a constant feed is fed to the radwaste evaporator, then an equilibrium condition is established r between evaporator feed, bottoms, and distillate, resulting in higher radio- '

! ' nuclide decontamination factors. Therefore, the evaporator DF measurements

! taken at Fort Calhoun Station in August,1976 and February,1977 indicate that radwaste evaporator DF is a function of the radionuclide j feed concentration, distillate conductivity (i.e., moisture and/or ion carry-over), and the length of time the evaporator processes a given and constant: feed stream. . That is, the higher the feed radionuclide concentration, the higher is the radionuclioe DF. . Also, the higher the

, distillate conductivity, the lower is the DF while the longer a constant j, feed stream,is processed, the higher the DF becomes.

70 L

,-- , m .o _

TABLE 3.25 EVAPORATOR DECONTAMINATION FACTORS "New" Tube Bundle - February, 1977 Feed: Spent Regenerant Tank WD-13A -

10:25; 2/18/77 13:45; 2/18/77 15:55; 2/18/77 Tank Level: 4530 gal Tank Level: 3800 gal Tank Level: 3330 gal Feed Rate: 5 gpm Feed Rate: 5 gpm Feed Rate: 5 gpm Bottoms Boron Conc: 6.4% Bottoms Boron Conc: 6.8% Bottoms Boron Conc: 6.9%

Distillate Cond: 1.3 pmho

~

Distillate Cond: 1.2 umho Distillate Cond: 1.2 umho Feed Feed Feed Activity Decontamination Activity Decontamination Activity. Decontamination Nuclide (pCi/ml) Factor (pCi/ml) Factor (pC1/ml) Factor 5'+Mn - 3.3710.07(-5) 1080 170 3.00 1 0.06(-5) 2260 440. 2.97 0.07(-5) 1800 290 3 ssCo 5.92 1 0.05(-4) 750 1 10 -5.56 0.u7(-4) 1030 20 5.81 1 0.09(-4) 1310 1 30 60C0 2.8610.03(-5) 330

• 20 2.5i0.2(-5) 460 40 2.53 t 0.08(-5)- 490 i 30 131I 9.0 1 0.2(-4) 6770 340 1.13 0.01(-3) 5890 160 1.09 0.02(-3) 6470 330 134Cs 1.22 1 0.01(-3) 2980 100 1.31 i 0.02(-3) 3190 470 1.38 0.02(-3)- 4440 140 137Cs 1.47 1 0.02(-3) 2870 180 1.54 0.02(-3) 2750 40 1.64 0.02(-3) 4210 180

-l

REFERENCES

1. NUREG-0384, " Procedures, Source Tern Measurement Program,"

N. C. Dyer, E. B. Nieschmidt, J. H. Keller, and B. G. Motes.

Idaho National Engineering Laboratory, EG&G Idaho Allied Chemical Company, National Technical Infomation Service, Springfield, Virginia 22161, December,1977. c

2. ANCR-1113, " Gauss VI A Computer Program for the Automatic Batch Analysis of Gamma-Ray Spectra from Ge(Li) Spectrometers," J. E. Cline, M. H. Putnam, and R. G. Helmer, Aerojet Nuclear Cos.pany, National '

Reactor Testing Station, Idaho Falls, Idaho 83401, June,1973.

3. Final Safety Analysis Report and Supplements Omaha Public Power District. Fort Calhoun Station, Unit No.1, AEC Docket No. 50-285, 1970-71.

4. "Results of Independent Measurements of Radioactivity in Process Systems and Effluents at BWRs," Dir. of Reg. Op., USAEC, May,1973.

5. NUREG/CR-0068, "An Atmospheric 3H and 14C Monitoring System,"

J. L. Thompson, S. W. Duce, and J. H. Keller, Idaho National Engineering Laboratory, Allied Chemical Co., May,1978.

6. NUREG-0017. " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)," Office of Standards Development, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, April, 1976.

7. N-237, "American National Standard, Source Tem Specification" ANS-18.1, ANSI N-237-1976, American Nuclear Soceity, 244 East Ogden Av>nue, Hit.sdale, Illinois 60521, May,1976.

8. Letters from T. P. Harding (0 PPD) to N. C. Dyer (EG&G Idaho, Inc.)

dated November 2 and 3,1977.

9. Fort Calhoun Unit #1. Evaluation of Fort Calhoun Station #1 in '

Accordance with 10CFR Part 50 Appendix I, June 1976.

l. 72

APPENDIX A.1 Introduction The data for the in-plant measurement program at the Fort Calhoun Station are presented in this Appendix. The data were collected over a period of about 6.5 months from 8/15/76 through 2/24/77. During this period Fort Calhoun was down for its second refueling from 10/1/76 through mid-December,1976.

A.2 Reactor Power Level The reactor power level for the period' 8/15/76 through 10/3/76 is presented in Figure A.1. Figure A.2 presents the reactor power level for the period 12/15/76 through 2/24/77. Noted on both Figures A.1 and A.2 are the times when reactor coolant samples were taken at Fort Calhoun Station.

A.3 Liquid Samples Liquid measurements were made from 8/18/76 through 9/2/76 for the end of an operating cycle prior to refueling, from 10/11/76 through 12/3/76 for the refueling period, and 2/9-22/77 for the beginning of an operating cycle following refueling.

A.3.1 Reactor Coolant Figure A.3 is a reproduction of the FSAR diagram of the reactor coolant system. Samples were taken at sample points #6 and #7 shown on Figure A.3. Tables A.1-A.7 show the radionuclide concentrations measured in reactor coolant. The reactor loop number and time of collection are noted on the tables. Reactor coolant radionuclide concentrations for the periods of power operation prior to refueling and after refueling are shown in Tables A.1 and A.3, respe,:tively. Concentrations of pure beta emitting radionuclides in reactor coolant, spent fuel pool water and a monitor tank are shown in Table A.2.

Three reactor coolant samples were taken during the refueling outage. The results for these samples are shown in Table A.4. The reactor vessel cover was removed on 10/22/76 and the refueling cavity and fuel transfer canal were filled on 10/26/76. The spent fuel was transferred from the reactor to the spent fuel pool between 11/2/76 and 11/7/76. The refueling cavity was drained and the reactor vessel cover replaced by 12/2/76 for testing prior to start-up.

I Reactor coolant samples were degassed by Fort Calhoun operating personnel in the following manner.1.A 300 mi sample was collected in an in-line~ steel container on the same sample line as for the non-degassed samples. The reactor coolant gas was transferred to an j evacuated 250 mi glass container by bubbling air through the steel container. The 133Xe activities in both the liquid and gas fractions were used to determine the degassing efficiency.

i 73 -

1 g

For each of the before and after refueling measurement periods, reactor coolant sample.s were collected and degassed by Fort Calhoun personnel prior to our counting. Tables A.5 and A.6 present the analysis results for these degassed samples. By comparing the' data in Tables A.1 and A.3 with Tables A.5 and A.6 it is apparent.that the degassing did not significantly increase the sensitivity for the non-detected -

radionuclides. Also, the degassing procedure delayed the first count c on these samples to 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after collection. This precluded

measurement of the short-lived radionuclides.

For each of the liquid reactor coolant samples presented in Table c

- A.6, a sample of the reactor coolant gases was obtained from OPPD. . The

- radionuclide analysis results of these samples are presented in

-Table A.7.

- . - Shown in Tables A.1, A.5, A.6, and. A.7 are the results from arialysis of similar reactor coolant samples by OPPD.

A.3.2 .CVCS Liquid A diagram of the chemical and volume control system (CVCS) is

presented in Figure A.4. The indicated local sample point on the volume control tank _ was not installed during construction. The infonnation on when the mixed bed purification ion exchangers (CH-8A and CH-88) were in service and recharged is presented in Figure A.5 for the period 3

July,1973 'through February,1977. For these demi.ieralizers, the resin was replaced and the used resin disposed of as solid wastes.

'Ouring the before refueling in-plant measureeent period at Fort

[ Calhoun, the purification ion exchanger (CH-88) was in service and the cation ion exchanger (CH-10) was not used. The samples for tne CVCS i were obtained from the points number 1, 2, and 5,_ in Figure A.4. The i three samples (50 m1, each) were collected at essentially the same time l and counted within 20 minutes of. collection, with the input (#1) l counted first.

Tables A.8 and A.9 present the radionuclide analysis data for the

,- samples from the CVCS on 8/20/76 and 8/30/76, respectively. Figure A.6 .

! presents a graph of the letdown flow rate for the CVCS during the period 8/15/76 through 9/9/76._ Figure A.7 presents the gra'phs of the reactor coolant pH and conductivity plus the pressure drops across the purification demineralizers and filters for the period 8/9/76 through 9/2/76. Sample times - -

are noted on both figures.

During the refueling measurement period, the purification ion exchangers were cperated in _ series for clean-up of reactor coolant prior to reactor. vessel head removal.- Samples were taken on 10/11/76

when CH-8A and CH-88 were in series 'and Table A.10 presents the results.

"These samples were taken from points numbered 1, 2, and 5 on Figure A.4. Figure A.8 presents the letdown flow rate for the period 9/25/76 through 10/19/76.

74

After movement of spent fuel from the core to the fuel pool, the cation ion exchanger (CH-10) of the CVCS was used to clean up the refueling cavity water prior to placement in the safety injection and refueling water tank (SIRWT). On 11/11/76, samples were taken of the input and output of CH-10 (points #1 and #3, respectively in Figure A.4) during the processing of refueling cavity water. Table A.11 presents 5 the analysis results.

During the post-refueling measurement period in February,1977, a sample series was taken of the letdown flow being processed through the

,2 purification ion exchanger CH-8A. These samples were taken from points 1, 2, and 5 on Figure A.4. A filtration procedure (described in the procedures report (1)) was used on these samples to determine the suspended radioactive solids with sizes larger than 0.5 micron. The data from these samples is presented in Table A.12. In this and all subsequent tables where dissolved and suspended activities are reported, the suspended activities represent radionuclides observed on the filter.

Dissolved activities represent activities found in the filtrate.

The letdown flow rate during the period from 1/31/77 through 2/24/77 is presented in Figure A.9. Figure A.10 presents graphs of reactor ,

coolant boron concentration and pressure drop across the purification demineralizers and filters for the same period.

A.3.3 Secondary Liquid Steam generator blowdown samples were collected from both generators before refueling on 8/26/76 and af ter refueling on 2/8/77 and 2/10/77. The February samples were over 200 liters in size and were collected with an ion exchange column technique described in

- the procedures report (1). The radionuclide analysis results are presented in Table A.13. Because of other possible pathways for radioactivity to enter secondary liquid, the radionuclide concentrations measured in the steam generator blowdown were not necessarily indicative of a primary to secondary steam generator leak. Figure A.ll presents the steam generator blowdown and flow rates for the period 1/31/77 through 2/24/77.

• A.3.4 Spent Fuel Pool and Fuel Transfer Associated Liquids Figure A.12 presents a diagram of the spent fuel pool cooling and clean-up system. The local sample points are indicated. During the

- measurement period, the valve for sample point #2 was not operable.

Therefore, only spent fuel pool water or the input to the fuel pool clean-up system could be sampled (#1). The volume of the spent fpel pool is 215,000 gallons. According to the auxiliary building operator s log, the spent fuel pool was full duririg the in-plant measurement periods. The temperature of the spent fuel pool water is presented on the tables with the radionuclide data. Table A.14 presents the radionuclide concentrations for two sampTes taken fron the spent fuel pool during the before refueling. Table A.2 presents the activities from a beta analysis of a spent fuel pool sample taken before refueling.

At this time, the pool contained spent fuel rods from one-third of ~

the core removed during the first refueling in March,1975.

75

During refueling, wa'ter is' taken from the safety injection and refueling water tank (SIRWT) to fill the refueling cavity and fuel transfer canal. The volume of the SIRWT and refueling cavity are 314,000 and 249,500 gallons, respectively. The refueling cavity and fuel transfer canal were filled from the SIRWT on 10/26/76. The spent fuel movement occurred between 11/2/76 and 11/7/76. Water was transferred from the refueling cavity and fuel transfer canal aftar clean-up with the CVCS  ;

by 12/3/76.

' Table A.15 presents the radionuclide concentrations in samples taken from the SIRWT before and after the reactor refueling (from d

1. 3 on Figure A.12). Table A.16 presents the radionuclide concentrations for samples taken from the fuel transfer canal and refueling cavity before and af ter movement of the spent fuel (dip samples). _ Table A.17 presents radionuclide concentrations for four spent fuel pool samples taken at different times during the refueling process.

During the post-refueling measurement period, a sample was taken from the spent fuel pool. This sample was filtered using the technique described in the procedures report (1). Table A.18 presents the radio-nuclide concentrations in the spent fuel pool sample taken in February, 1977.

i A.3.5 Liquid Radwaste System Figures A.13 through A.16 present uf agrams of the liquid radwaste system. The radwaste tanks which were sampleable were the reactor coolant drain tank (WD-1), the spent regenerant tanks (WD-13A and WD-138), the waste holdup tanks (WD-4A and WD-4B), the hotel waste tanks (WD-15A and WD-158) and the monitor tanks (WD-22A and WD-228). The sample point for the reactor coolant drain tank was a local valve on the tank,

• within containment. Sample points for the other tanks are noted by numbers one through four on Figures A.14 through A.16. Liquid radwaste samples obtained after refueling were filtered using the technique described in the procedures report (1).

Figures A.17, A.18, and A.19 present the tank levels as a function of time for the waste holdup tanks for the three measurement periods.

Figures A.20, A.21 and A.22 present the tank level information for the

• reactor coolant drain tank, the spent regenerant tanks, the hotel tanks, and the monitor tanks for the three measurement periods.

A.3.5.1 Reactor Coolant Drain Tank ~

fable A.19 pre,sents the radionuclide concentrations observed in the reactor coolant drain tank for a sample taken during the

, after refueling measurement period.

s l

l 76

A.3.5.2 Waste' Holdup Tanks Table A.20 presents the radionuclide concentrations detennined on two samples taken from the waste holdup tanks during the before refueling measurement period. A duplicate of the sample '

taken 11:40, 9/2/76 was analyzed by DOE-HSL and their results are also shown. Table A.21 presents the radionuclide concentrations in three waste holdup tank samples taken during refteling. Table A.22 presents t'ie

> radionuclide concentrations measured in a sample from waste holdup tank WD-4A taken after-refueling. A duplicate sample was also analyzed by DOE-HSL and their results are shown.

D A.3.5.3 Spent Regenerant Tanks Table A.20 presents radionuclide concentrations measured on three samples of the spent regenerant tanks taken during the oefore refueling measurement period. Table A.21 presents data on one sample taken during refueling. Radionuclide concentrations of two samples taken from the spent regenerant tanks during the after refueling measurement period are shown in Table A.23.

A.3.5.4 Hotel Waste Tanks Table A.24 presents radionuclide concentrations observed in a sample taken from hotel tank WD-15A before refueling.

Radionuclide data measured on two hotel tank samples taken af ter refueling are shown in Table A.25.

A.3.5.5 Monitor Tanks Tables A.2 and A.24 present radionuclide data for three samples taken from the monitor tanks before refueling. One monitor tank sample was taken during refueling and the results are shown in Table A.21. Table A.26 presents data from two monitor tank samples taken after refueling. Af ter refueling, a series of four samples were taken at different times in the release of monitor tank WD-22B to the discharge pipe. These results are presented in Table A.27. Several times during the measurement period, OPPD collected samples of monitor tank water at the same time our samples were obtained. Comparisons of OPPD's and our

- results for these samples are presented in Tables A.24A, A.26 and A.27.

In order to make a valid comparison, our results reported in these tables are for unfiltered samples (i.e., same procedures used by 0 PPD).

- A.3.5.6 Waste Evaporator Evaporator feed samples were taken from the tanks being processed and the distillate and concentrate samples were taken from points 5 and 6 onTigure A'.16~, respectively. Tables A.28 through A.30 present the radionuclide data for the samples taken while the evaporator was processing foed from waste holdup tank WD-4B, spent .

regenerant tank WD-13A and spent regenerant tank WD-13B, respectively.

77

i Figure A.23 presents the radwaste evaporator parameters, evaporator feed, bottoms boron concentration, and distillate conductivity for the period 8/9/76 through '9/2/76. Figures A.24 and A.25 present the same evaporator parameters with an expanded time scale to better show the variation of these parameters during the period when the samples of Tables A.28 through A.30 were taken.

During October,1976 the evaporator tube bundle was replaced with ,,

a new bundle. Scraping samples were obtained from each end of the evaporator bundle on 10/13/76. Concentrations shown in Table A.31 are in pCi per sample and provide a relative indication of the radionuclide buildup on the evaporator tube bundle.

4 Figure A.26 presents the evaporator operating parameters, feed, bottoms boron concentrations and distillate conductivity for the period 1/31/77 through 2/21/77. The bottoms were removed in a batch mode. A series of measurements were made in February,1977 to evaluate the waste evaporator performance with a constant feed. The operational parameters are shown on an expanded time scale in Figure A.27. The radionuclide concentrations measured for the feed, distillate, and concentrated bottoms are shown in Tables A.32, A.33, and A.34, respectively.

A.4 Gaseous Samples In the text of this report (Section 2.3.1), the sampling stations and locations for the gaseous streams are described. Figure A.28 is the P. & I.D. for the ventilation system at Fort Calhoun Station with sample stations #1 through #4 noted on the figure. Figure A.29 presents an as-built modification to the P. & I.D. shown in Figure A.28.

The radionuclide activities measured at sample. stations #1 through #4 are presented in Tables A.35 through A.38, respectively. Additional samplers were placed in rooms to determine the airborne activities reaching the ventilation system from these areas. These rooms were the waste evaporator room, the pipe penetration room, and the letdown heat exchanger room. Tables A.39 through A.41 present the results of these measurements. Iodine-131 species data were collected at ventilation sample station #2 and the waste evaporator room during most of the in-plant measurement period at Fort Calhoun. Tables.A.42 and A.43 present these data. During the post-refueling measurement period,1311 species data '

were collected using the ventilation sample stations for station #1, letdown heat exchanger rotmi and pipe penetration room. These data are presented in Table A.44.

In addition to the ventilation samples, grab samples were taken from portions of the pro' < gas system. In all cases, the samples were taken through the plant's ..dtomatic Gas Analyzer System. Iodine specie; samplers were used to obtain samples of several of the subsystems. Two samples of gas from different waste gas decay tanks were also taken for gamma analysis. The results of these masurements are shown in Tables A.45 through A.48.

78

Carbon-14 and tritium analyses were perfori.ed c : the same process and cover gas subsystems as were the iodine species determinations. The sampler developed for air monitoring was not used because of the absence of oxygen in these systems. As an alternate sampling procedure, 75 cc stainless steel gas bombs were used to obtain these gas samples. The gas bombs were returned to INEL where the contents were mixed with low activity air and processed through the normal sampling system. The results of these measurements are shown in Tables A.49 and A.50.

d Two samples of the containment atmosphere were obtained. The first sample, taken in October,1976, was an iodine species sample taken through a plant installed containment penetration. The second sample, taken in January 1977, consisted of iodine species and 14C and tritium measurements.

These latter samplers were set up inside the containment structure.and were operated for approximately two days. The results of these measurements are shown in Table A.51.

P 4

l l

l 79 l

TABLE A.1 REACTOR COOLANT ACTIVITIES Power Operations - Before Refueling.

Loop #1 Loop #1 OPPD Loop #2 Loop #1 OPPD Loop #2 12:58 10:05 14:52; 8/16/76 ~ 09:17; 8/19/76 8/19/76 09:58; 8/19/76 11:54; 9/1/76 9/1/76 11:19; 9/1/76 Nuclide (uCi/ml) (vC1/ml) (uCi/ml) (uC1/ml) (uCi/ml) (uC1/ml) (uC1/ml)

Activation Nuclides:

24Na 2.2 0.4(-3) 1.9 0.2(-3) 2.8't 0.3(-3) 1.3*0.2(-3) 1.4 i 0.1(-3) s1Cr <1(-2) <2(-2) <7(-3) <2(-2) <3(-2) <3(-2)

SMn 1.3 0.1(-4) 2.1 1 0.2 - 1.9(-2) 3.0 0.3 -4) 1.1 1 0.2(-3) <6(-3))

1.2(-2 3.5i1.3('4).

56Mn <1(-3) 3.3 1 0.3 - 3.7 0.4-2) 3.6 i 0.3(-2) 3.6 i 0.7(-2) 59Fe 1.1 0.2(-4) 5.6 i 1.6 - <1(- 5.8 1.6 -5) <

2(-4) <9(- <8(-4)-

57Co <l(-4) 2.8 i 1.5 - <5(- <8(-5) <2(-4) <5(- <4(-4)

"o 58C0 9.0 ;t 0.!(-3) 6.03 0.05 3) 2.4 1) 6.13 0.05(-3) 7.1 0.3(-3) 3.2 2) 1.24*0.08(-2) 60C0 1.6 0.7(-4) 7.9 1.0(-5) 2.8 -3 9.6 0.9(-5) 2.010.9(-4) 2.4(-3 <4(-4) 65Zn <1(-4) <2(-5) <2(-3) ) 5.0 3.8(-3) 1.1 0.6(-3) <1(-3)) 2.2 1 0.8(-3) 187W 2.6 0.8(-3) 8.3 i 3.4(-3) 1.4 0.2(-2) 3.0 1 0.6(-2) 3.8 0.7(-2) 239Np <6s -3) 1.8 0.6(-2) 2.0 1.5(-2) 5.3 1.9(-3) <2(-2)

Iodine Nuclides:

131I 1.48 0.02 -1) 2.03 1 0.02(-1) 2.4(-1 2.05

• 0.04 -1) 3.92 0.01 -1 4.3 - 3.93 1 0.02 -

, 1321 1.13 0.01 -1) 8.9 - 8.49 i 0.09 - 1.05 0.01 -1 6.8 - 1.01 1 0.01 133I 134I 3.36 8.2 1 0.05(-2 0.01 -2) 8.03 0.1(-2) )9.5 - 8.57 0.07 - 2.40 1 0.01 -1 2.4 - 2.40 1 0.01 3.8 0.9(-2) 3.2 0.3(-2) 2.2 - 3.55 0.67 - 2.8i0.3-2) 1.6 - 3.0 1 0.3(-2 135I 2.8 i 0.2(-2) 6.5 1 0.2(-2)- 6.1 - 6.7 0.2(-2) 7.2 1 0.2 -2) 6.1 - 7.3 1 0.2(-2

_ _ _ _ _ _ _ _ _ _ _ m o

7 3 . ~.. .

TABLE A.1 (cInt.)

REACTOR COOLANT ACTIVITIES Power Operations - Before Refueling '

Loop #1- Loop #1 .OPPD Loop #2 Loop #1' -OPPD Loop.#2

12:58 10:05 14:52; 8/18/76 09:17; 8/19/76 . 8/19/76 09:58; 8/19/76 11:54; 9/1/16 9/1/76 11:19; 9/1/76 Nuclide - (uC1/ml) (uC1/ml) (uC1/ml) (uCi/ml) (uCi/ml) (uC1/ml) (uC1/ml)

Fission Nuclides:-

esRb' 5.3 i 0.4 - 5.210.2(-1) 4.4 0.6(-1) 3.8 i 0.2 - 4.0 0.2(-1) 89Rb 6. 2 i 0. 3 - 4.8 i 0.4(-2) 4.5 i 0.4(-2). 7.0 i 0.4 - 7.1 i 0.3(-2) 95Zr 8.0 1 2.7 - <3(-5) <1(-3) <1(-4) 8.1

• 3.2 - <8(-4) <7(-4) 9sNb <9(-5) 8.6 3.0(-5) <7(-4) 3.8 i 1.5(-5) <3(-4) <5(-4) <3(-4) 99Mo IL4 0.2(-2) 1.5 0.2 - 8.4(-3) 1.32 i 0.07(-2) 9.5 0.6(-3) .1.3(-3) 2.5 0.4(-3) 101Mo j. 4

. 0.6(-2) 1.1 0.3 - 9.5 i 6.2(-3) 2.1 0.5 - ' <2(-2) 104Tc <6(-3) 3.3 i 1.3 - <2(-3) 3.7 i 1.1 - .

2.4 i 1.6 -

. 103Ru <2(-4) <2(-4) <8(-4) <3(-4) 2.7 1.1 - <6(-4) 1.1 0.6 -

110AAg 2.7 2.1(-4) 3.5 4.5(-4) 3.8 i 5.0(-4) 0.2 4.7 - -

02 4.7 -

<5(-4) < 1(-3) <1-4) 124Sb 1L2 0.7(-5) <1(-4) <1(-3) <2(-4)'

1263h  : <1(-4) <1(-4) <3(-4) <5(-4) <1 -4) 127Sb 1.1 0.3(-3) <2 - <2(-4) <7(-4)- s9 -4 129Te. <1(-2) <9 - <l(-2) <7(-3) <8 -3 129mTe <3 - <2 - <2 -3) <8(-3) <9 -

131mie <6 - <3 - <3 -4) <6(-4) <6 -

132Te <1 - <2 - <2 -3) <8(-3) <4 -

134Cs- 1.56 i 0.01(-2) 1.36 0.01(-2) 1.2(-2) 1.45 i 0.01(-2) 1.47 0.04(-2) 5.6(-3) 1.44 0.01(-2) 136Cs 3.8 0.1(-3) 3.74 1 0.06(-3) 3.2(-3 137Cs 1.56 1 0.01(-2) 3.6 1.38 1 i 0.01(-2 0.1(-3) < 9(-4)

)1.7( -2 ) 1.38 0.01(-2) 4.7 1.40 i 0.3(-3) 0.02(-2 ) 1.6(-2 4.1 1.42

• 0.2(--2 ) )

1 0.02 13sCs 3.7 0.1(-1) 3.1 0.1(-1) 2.4(-1) 3.4 0.1(-1) 3.22 0.07(-1) 2.3(-1 3.13 1 0.10 1)

139Cs <4(-4) <7(-4) <8(-4) <3(-4) 1.6 1 0.4(-

139Ba 7.4 0.7(-2) 6.410.4(-2) 5.8 0.6(-2) 6.4 i 0.8(-2) 4.2: 1.1(-

140Ba 3.2 i 1.9(-4) 5.8 i 1.3(-4) <3(-3) 5.9 1.5(-4) <2(-3) <3(-3) <6(-4)

! 140La . 4.2 i 1.0(-4) 3.0 2 0.7 - <7(-4) 4.7 i 0.8(-4) <4(-4) <5(-4) <2 -

l 141Ce <3(-4) 3.0 2 1.3 - <1(-3) <1(-4 <3(-4) <1(-3) <2 -

8.317.5(-4) 3.0 1 0.6 - <8 -

14 3Ce <4(-3 4.2 1 1.7(-3) 144Ce . <1(-3) <2(-3) <6(-4 *<1(-3) <3 -

4

__-__m___..___.________.___ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ m __ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - e

TABLE A.2 '

BETA. ONLY, EMITTING RADIONUCLIDE ACTIVITIES-Power Operations - Before Refueling

. Loop #1 Spent Fuel .

Reactor Coolant Pool .

Monitor Tank 22B

.12:00; 9/1/76 '10:25; 8/31/76 -08:17; 9/1/76 Nuclide (uC1/ml) (uC1/ml) _ (uC1/ml) 3H .4.77 0.02(-2) 7.34 0.04(-3) 3.75 i 0.02(-3) 14C A 4.7 0.4(-6)- 7 1 2(-8) 2.6 'O.2(-7) 32P 4.5 0.4(-6) <2(-7) <4(-8) 35S 5.6

• 2.0(-6) <2(-7) <2(-7) 4 5Ca . 1.08 0.01(-4) 1 44 0.02(-5) 1.79 0.02(-5) g ssFe 1.06 0.01(-3) 1.84 0.02(-5). 5.2 i 0.1 (-6) 6 3Hf 1.76 0.02(-4) 2.37 1 0.04(-4) 3.6 0.1(-7)-

89Sr 1.22 2 0.05(-3) <4(-8) 3 1(-8)'

saSr 3.8 0.5(-6) <2(-8) <2(-8) 91Y 7.7 i 1.3(-6) <l(-7) <1(-8) 147pm 9 2(-7) <2(-8) <2(-8)

A Analysis results are for inorganic 14C only (CO2) o_ n.

TABLE A.3 REACTOR COOLANT ACTIVITIES Power Operations - After Refueling Loop #1 Loop #2 Loop #1 08:40; 2/9/77 08:44; 2/17/77 13:01; 2/22/77 3

Nuclide (uCi/ml) (uCi/ml) ,(uci/ml)

Activation Nuclides:

's 24Na 7.5 i 1.2 'l 1.41 i 0.03(-2) 1.2 0.2(-2) 51Cr 7.3 i 1.8 - l l <2(-2) <1(-2) 54Mn 4.8 0.3 - p 1.6 0.1(-3) 2.2 0.8(-5) ssMn 1.4 i 0.4 - I <4(-3) <6(-3) 59Fe 4.7 i 0.6 1.8 0.2(-4 <3(-5) 57Co 3.6 1 0.8 - )l 2.3 0. 4(-3 1.9 0.2(-3) seCo 2.00 0.02-3) 3.57 1 0.04( 3) 1.1310.05(-3) 60Co 3.2 i.0.4(-5) 1.2 1 0.2(-4 1.8 0.7(-5) 6sZn <2(-5) <6 - <3(-5) 187W 8.610.6-3) <5 - 5.4 1 0.6 -3 239Np 4.3 i 1.8 -3) <1 - 9.6 5.7 -3 Iodine Nuclides:

131I 9.54 i 0.08 - 8.62 0.03 - 7.3710.05-1321- 6.30 1 0.08 - 4.56 i 0.09 - 4.87 0.06 -

133I 1.27 i 0.01 - 1.04 1 0.01 -1) 1.02 1 0.01 -

1 34I 4.2 0.3 - 2.8 0.2 - 3.6 1 0.3 -

135I 8.2 1 0.3 - 7.0 0.2 - 6.8 i 0.2 -

Beta Only, Nuclides:

3H

• 2.15 0.02(-1) I

• 7.0 14C A 55Fe

• 1.16 0.2(-6) 0.02(-4 )

63Ni *

• 3.7 0.2(-5)

• 8.6 0.9(-6) 895r 90Sr <8(-7)

• 1.3 0.2(-6) 91Y l

• Not analyzed for' beta emitters.

A Analysis results are for inorganic 14C only (CO2 ).

83 I

TABLEA.3(cont.'d)

REACTOR COOLANT ACTIVITIES Power Operations - After Refueling a

Loop #1 Loop #2 Loop #1 08:40; 2/9/77 08:44; 2/17/77 13:01; 2/22/77 Nuclide (uCi/ml) (uci/ml) (uC1/ml)

Fission Products:

. . serb 4.6 1 0.2 - 5.3 0.3(-1) 5.0 0.3(-1) 89Rb 2.1 i 0.2 - 1.3 0.3(-2) 2.3 0.2(-2) 95Zr 4.3 1.2 - <3 - <2(-4) 95Nb 3.8 1 0.3 - <2 - <1(-4) 99Mo 1.4 1 0.9 - <8 - 6.4 6.3(-5) 103Ru <3(-5) 1.7 0.5(-5) 1.7 0.6(-5) 110 mag <3(-5) <2 - <8 -

124Sb 3.6

• 2.4(-6) <2 - <8 -

129Te <5 - <2 - <6 -

129mTe <3 - <5 - 1.9 1 0.8(-4) 131mTe <2 - <3 - <3(-4) 132Te <2 - <2-3) <2(-3) 134Cs 1.09 0.02(-3) 6.23 0.03(-2) 3.9 0.1(-3) 136Cs 3.5 0.4(-5) 9.7 1.3(- ) 5.6 1.9(-5) 137Cs 1.53i0.02(-3) 7.1310.03-2) 4.48 0.06(-3) 13sCs 1.73 0.09(-1) 1.47 0.07 -1) 2.0 0.l(-1) 139Cs <2(-3) <4(-4) <3(-3) 139Ba 6.3 11.4(-2) 4.9 0.2(-2) 5.5 0.3(-2) 140Ba <1(-4) <5(-3) <4 -

140La 8.7 2.3(-5) 4.4 1.4(-4) <6 -

141Ce 2.7 1 1.5(-4) <2 - <5 -

143Ce <3(-3) <4 - <3 -

144Ce <2(-3) <7 - 2.3 0.3(-3) i

• Not analyzed for beta emitters.

l A : Analysis results are for inorganic 14C only (CO2 ).

i 84

TABLE A.4 REACTOR COOLANT ACTIVITIES During Refueling Prior to Head After Head After Head Replacement Removal Removal Prior to Start-up

12:36; 10/20/76 11:00; 10/27/76 13:15; 12/3/76 Nuclide (uci/ml) (uC1/ml) (uCi/ml)

Activation Nuclides:

9 51Cr <5 (- 3) <2(-3) 2.9 0.9(-3 54Mn 3.60 0.06(-4) 2.5 0.1 - 8.7 0.2 -

59Fe 1.6 0.1(-4) 1.2 1 0.3 - 3.2 0.4 -

5700 2.35 0.08(-4) 4.1 0.8 - 2.3 0.3 -

58C0 -

1.34 1 0.03(-1) 1.78 0.05 1) 7.0 0.1 -

60Co 6.4 0.1(-4) 6.9 0.3(-4) 1.6 0.3(-3) 652n <7(-5) 8.2 1 2.5(-5) 5.1 i 1.0(-5)

Iodine Nuclides:

1311 1.2 t 0.2(-2) <2(-3) <4(-4)

Fission Products:

9sZr <1 - <3(-2) 1.1 0.2(-4) 9sNb <9 - <2(-2) 2.6 1.7(-4) 10 3Ru <1 - 1.5 0.8(-4) <3(-4) 110 mag <1(-3 <2(-3) 1.5 1 0.1(-4) 124Sb 6.1 i 0.5(-5) 1.3 0.1(-4) 1.28 0.09(-4) 129mpe <4(-l) <5(-1 ) <2(-1) 134Cs 2.38 0.02(-2 3.65 0.03(-2) 6.18 0.09(-3) 136Cs 2.12 0.08(-3 9.1 1.0(-4) 1.9 0.3-4) 137Cs 2.26 1 0.05(-2 3.83 0.07(-2) 6.3 0.1 -3) 140Ba <5-4) <9(-3) 5.3 2.1 -4) 141Ce <4-4) <1(-4) <4(-5) 144Ce <5 -1) 2.6 2.3(-4) <2(-4)

Beta Only Nuclides:

3H 3.49i0.04(-2) 3.44 0.02(-2) 1.06 1 0.02(-2) 14C 2.74 0.04(-5) 2.10 0.02(-5) 5.3 0.2(-7) 1.6 5.

• 32P 0.7(-5) 35S <4(-6) 1.4 3.(-6))

0.9(-6 2.50 i 0.06(-5) 45Ca <2(-6) <6(-7) ssFe 2.17 0.04(-3) 3.89+0.03(-4) 3.91 0.02(-3) 63Ni 5.75.i.0.04(-3). 3.86 0.02(-3) 3.94 i 0.02(-3) 89Sr 4.8 1 0.4 - 5.3 0.7(-6) 5.3 0.3(- )

90Sr 2.0 i 0.4 - 4. 2.(-7) 1.08 0.05-5) 91Y 1.5 0.4 - 3. 2.(-7) 2.13 0.08 -5) 147Pm -1.1 0.2 - 1.7 0.1(-6) *

• Sagle not analyzed for these, radionuclides.

85

q TABLE A.5_

REACTOR COOLANT ACTIVITIES (OPPD Degassed Sample)

Power Operations - Before Refuelino

. Loop #1 OPPD Loop #2 OPPD ' Loop #1-09:04 08:33-14:39; 8/23/76 -8/26/76 09:04; 8/26/74 8/27/76 08:33;.8/27/76 Nuclide (pC1/ml) (uCi/ml) (uCi/ml) _

(pCi/ml)_ (vC1/ml)

Activation Nuclides:

24Na 4.2 0.2(-3) 2.8 0.1(-3) 2.5 i 0.2(-3) s1Cr. <2(-2) <7(-3) <2(-2) <7(-3) <1(-2) 54Mn <1(-4) 3.2(-2) 1.1 0.5(-2) 2.7(-2) 2.64 0.06(-4) 56Mn 2.7 2.9(-3). 3.6 3.1(-3) .

<4(-3) 59Fe <5(-5) <2(-3) <2(-3) <2(-)- 6.7 1 2.7(-5) 57C0 1.7 0.4(-3) <5(-4) <1(-4) <5(- ) <7(-5)'

g 58C0 1.30 0.05(-2) 2.4(-2) 1.22 i 0.06(-2) 1.1 -2) 2.86 0.04(-3 60C0 65Zn 1.0 1 0.2(-4) <1(-3) <1(-4) <1(-3) 3.2 0.8(-5))

<4(-5) <2(-3) <2(-3) <2(-3) <2(-5) 187W <4(-4) <3(-3)

<2(-4) 239Np 9.7 1 4.9(-3) <1(-2) <1(-2)

Iodine Nuclides: i 131I 2.18 0.02 - 2.4(-l 1.99 1 0.02 - 1.91;-1} 1.71 1 0.01( 1) 132I 1.51 1 0.01 - 8.7(-2 1.06 0.02 - 7.91,-2. 7.8 0.8(-2 asal 3.20 0.02 - 3.2(-l 2.97 0.06 -~ 2.5 2.41 1 0.06( 1) 134I lasI 1.08 0.07 -

2.11 0.02(-1 5.0(-2 1.3(-1 5.8 1 0.3(-2) 1.55 0.03(-1) 2.2 ((-)

9.71,- )

) <5(-1) 1.24 1 0.03(-1) 1

• n.  ?

. _ -m _ . _ . . . . . - ._

._ _._ _ 3 ._..7_.__.___.. f ,

b

- TABLE A.5(cont.).

t REACTOR COOLANT ACTIVITIES (OPPD Degassed Samle)'

Power Operations - Before Refueling Loop #1 OPPD Loop #2 OPPD.  : Loop #1 09:04 .

08:33 .

14:39; 8/23/76 8/26/76- 09:04; 8/26/76 8/27/76 08:33; 8/27/76'

. Nuclide (uC1/ml) (uC1/ml) (uci/ml)- - (uCf /ml) '

(uC1/m11 Fission Products:

ssZr 6.3*1.5(-4) <1(- <2(-3) <1(- <1(-4) ssNb <1(-4) <8(- <8(-4)- <8(- <6(-5) 99Mo 2.45 1 0.06(-3) 1.5 3) 1.0 i 0.1(-3) 1.2 3) 3.1 1 1.1(-4) lo3Ru <1(-4) <8(-4) <9(-5) .

<7(-4) '

< 2(-5) .

110 mag 0.3.i2.6(-4) 1.5 0.7(-4) .

O.53 0.54(-4) 124Sb 2.2

• 0.3(-4) <2 -3 <8(-5)- <2(-3) <2 -

ca 12ssb- <1(-4) <8(-5) '2-

." <2 -

127Sb -2.5 i 1.8 - <2(-3) .

~34Cs 4.5 1 0.1 - <9 - 1.8 0.5 - <9 - 1.2.i0.2(-3) 136Cs 1.E 0.1 - <1 - 4.5 i 1.3 - <1 - 3.06 i 0.08 -

2 137Cs 4 64 0.07 3) <9 - 1.6 i 0.1 - -<9 - 1.19 1 0.01 - r 140Pa <5 -4) <3 - <3(-4 <3 - '1.40'i 0.05 -

'140La <4 - <9 - <1 - <9 - 2.8 i 1.6(-3) 141Ce <1 - <1 - <2 - <1 - <1 -

143Ce <4 - e4 - <3 -

144Ce <6(-4 <1 -3) <5 - .,

4 4

TABLE A.6 REACTORCOOLANTACTIVITIES(OPPDDegassedSample)

Power Operations - After Refueling Loop #1 Loop #2 08:16; 2/9/77 08:21; 2/17/77 INEL OPPD INEL OPPD Nuclide (uCi/ml) (uci/ml) (uci/ml) (uC1/ml) '

Activation Nuclides:

24Na 5.9 2 0.1(-3 1.49 1 0.02(-2 ci s1Cr 1.0 i 0.1(-3 2.2 8.9i2.2(-4)) < 5(-3) 54Mn 6.6 i 0.3(-5 2.18 0.5(-2))

0.07(-2 2.3 i 1.0(-4) 2.03 0.07(-2) ssMn <3(-3) <7(-3) 59Fe 8.5 0.6(-5 2.7 i 1.1(-3) 1.0 0.3(-4) 5.5 1.1(-3) 57C0 '2.3 0.6(-5 <4(-4) 9.7 2.0(-4) <4(-4) seCo ~5.69 1 0.07( 3) 9.7 t 0.4(-3) 3.48 0.04( 3) 6.3 0.6(-3) 60C0 7.5 1 0.6(-5) <6(-4) 9.3 3.0 - <6(-4) ssZn <2(-5) <1(-3) 1.4 0.5 - <1(-3) 187W 9.9 i 0.4(-3) 8.7 1 0.4 -

239Np 4.1 0.7(-3) 1.8 1.3 -

Iodine Nuclides:

1311 8.29 i 0.03 - 1.02 0.01(-1) 8.85 i 0.08( 2) 9.25 0.09(-2) 132I 5.46 0.07 4.5 0.4(-2) 5.5 1 0.1(-2 4.5 1 0.4 -

133I 1.10 0.01 1.23 0.01(-1) 1.11 0.02(1) 9.9 0.1 -

134I 3.7 0.3 - 2.2 0.2 - 3.8 0.4 - 1.6 0.2 -

1351 7.2 i 0.1 - 5.7 0.2 - 7.6 i 0.2 - 5.3 0.2 -

Fission Products:

95Zr 6.6 i 0.5(-5) <8(-4) <3(-4) <8(-4) 95Nb 4.6 1 0.3 - <5(-4) <2(-4) <5(-4 99Mo 3.3 1. 6 - 4.3 i 1.0(-4) 3.9

• 1.2(-4) <9(-5 103Ru 1.9 i 0.6 - <5(-4) <2(-3) <5(-4 110 mag 1.3 0.3 - <2(-3) 124Sb 4.7 1 1.2 - 4.4 1.4(-3) 3.9 i 3.9(-5) <1(-3) .

129Te <3 - <2 -

129mie <4 - <5-131mTe <5 - <3 -

132Te <2 - <2 - -

134Cs 8.2 1 0.1(-4 <5 - 6.51 0.05(-2) 5.08 0.09(-2) 13sCs 5.6 i 1.3(-5 <6 - 1.610.3(-4) <6(-4) 137Cs 1.18 0.02 3) <5 - 7.44 0.04(-2) 7.14 13sCs 2.30 0.08 -1 0.09(-2) 1.1210.03(-1) 2.9 i 0.1(-1) 1.39 i 0.04(-1) 139Ba 3.2 1 0.1(- ) ) 4.3 0.5(-2) 140Ba <9(-5) <2 - <6 - <2 -

140La 1.4 0.8(-3) <6 - <7 - <5 -

141Ce <2 - <7 - <8 - <7 -

143Ce <3 - 1.310.3(-3) 144Ce <1 - <4(-3) 88

l I

TABLE A.6 (cont'd)

REACTOR COOLANT ACTIVITIES (OPPD Degassed Sample)  !

Power Operations - After Refueling l Loop #1 13:38; 2/22/77 INEL OPPD e Nuclide (uCi/ml) (pC1/ml)

Activation Nuclides:

b 24Na 1.310.3-51Cr 2.0 0.4 - <5(-3) 54Mn 9.2 i 1.2 - 2.08 0.06(-2) 56Mn <4(-3) 59Fe 1.4 0.4(-4) 2.3 1.1(-3) 57Co <1(-3) <4(-4) l 58Co 1.04i0.05(-2) 1.07 i 0.06(-2) 60Co 1.5 i 0.2(-4) 3.3 0.7(-3) 65Zn <8(-5) 3.8 1.3(-3) 187W 7.5i0.9(-3) 239Np 3.3 1.9(-3)'

Iodine Nuclides:

131I 7.3i0.1(-2) 7.58 0.08(-2) 132I 4.7210.08(-2) 4.7 0.4(-2) 133I 9.60 i 0.05(-2) 8.84 0.09(-2) 134I 3.4 i 0.3 - 1.2 i 0.2 -

135I 6.3 0.1 - 4.6 0.2 -

Fission Products:

95Zr- 2.4 0.2 -4 <7 -

ssNb 1.8 2 0.4 -4 <4- '

99Mo <1(-3 <9 - l 10 3Ru 9.5 2.4(-5) 4.2 0.7(-3) 110 mag <2 -

. 124Sb ,<2 - <1(-3) 129Te <6 -

129mTe <6 -

131mie <4 -

. 132Te <2 -  !

134Cs 3.9i0.1(-3) <5(-4) 136Cs 1.1 0.2(-4) <6(-4) 137Cs 4.5310.07(-3) 3.3

-13sCs 2.64 1 0.08(-1) 1.16 0.6(-3))

0.03(-1 139Ba 4.3 0.3(-2) l 140Ba <7(-4) <2 -  ;

<5 -

140La 1.1 1 0.2(-4) 141Ce - <2 - <6 -

143Ce- <3 -

-144Ce- <9 -

89 i

-y TABLE A.7 REACTOR COOLANT RASE 00S ACTIVITIES.

Power Operations - After Refueling -

Loop #1 .

Loop #2 Loop #1 08:16; 2/9/77 '

08:21; 2/17/77 13:38; 2/22/77 INEL OPPD INEL- OPPD INEL .0 PPD Nuclide (vC1/ml) (uCi/ml) (vC1/ml) .(pCi/ml) (pC1/ml) (vCi/ml) 4 1Ar 5.6i0.7(-3) 5.02(-3) 2.6 i 0.5(-3) 5.05(-3) 2.0 0.3(-3) 3.15(-3) esKr 8.4 1.2(-3) 1.8 0.4(-2) <8(-1) esmKr 1.40 0.05(-1) 1.95(-1 1.88 0.02(-1) 2.30 - 1.7 i 0.1(-1) 1.88 -

87Kr 1.43 1 0.02(-1) 1.67(-1 1.92 0.08(-1) 2.05 - 1.87 0.06(-1) 1 .51 -

saKr 2.6 i 0.1(-1) 0.1(-2, 2.54(-1 3.310.2-) 3.00 - 2.93 0.05(-1) 2.42 -

gg 131mXe 3.1 5.2 i 0.3 - ) 5.2- 0.3(-2)'

133Xe 5.0 0.1(0) 1.32 +1) 6.1 0.1 0 1.43(+1) 5.54 0.03 0 1.11 +

133mXe 1.16 0.01(-1) 1.27 -1) 1.38 0.02(-1) 1.20(-1) 1.29 0.01 - ) 8.91 -

13sXe 8.5 1 0.3 - 1.06 0 1.05 0.01(0) 1.24 0 1.04 i 0.02 0 9.83 -

135mXe 7.0 1 0.7 - 6.67 - ) 9.5 0.4(-2) 9.32-) 9.2 0.2(-2) 6.88 13 axe 1.5 i 0.2 - 3.22 - ) 1.58 0.07(-1) 5.72 - ) 1.56 i 0.05(-1)- 4.28 -2 '

TABLE A.8 CVCS LETDOWN ACTIVITIES - 8/20/76 Power Operations - Before Refueling

-CH-88: Placed in service, 3/24/76; use, 93 days by 8/20/76 Input to CVCS Output From Output From 6 Ion Exch CH-88 Ion Exch CH-8B Filters 10:56; 8/20/76 10:57; 8/20/76 10:58; 8/20/76 Nuclide (uti/ml) (uC1/ml) (uCi/ml)

Activation Nuclides:

24Na- 3.4 1 0.2(-3) 2.6 2.1(-4) <1(-4) s4Mn 6.3 0.3(-4) 4.5 0.1(-4) 3.2 i 1.9(-5) 56Mn 3.5 i 0.2(-2) <2(-3) seCo 3.44 0.06(-3) 4.0 0.1(-4) <2(-5) 60C0 6.3 1.9(-5) 2.4 0.4(-5) 2.1 0.7(-6) 187W <4(-3)

Iodine Nuclides:

131I 2.41 1 0.02(-1) 1.92 1 0.03( 2) 4.2 i 2.6(-5) 132I 7.1 1 0.1(-2) 5.0 0.9(-3 <8(-4) 133I 1.20 0.01 1) 7.6 1 0.3(-3 <5(-4) 134I 3.1 1 0.3(- <2(-3) 1351 7.3 0.2(- 9.9 2.4(-3) <5(-3)

Fission Products:

88Rb 4.9i0.2-;l 1.8 0.7(-1) 1.5 0.4(-1) 89Rb 5.0 0.4 '> <1 -

95Zr 3.9 1.7 'l <2 -

95Nb 7.5 1.6 - l <8 -

99Mo- 4.6 0.2-ll 3.1 i 1.0(-5) <2(-5)

• 101Mo 1.8 0.4(-21 <1(-2) 110 mag 1.6 0.8(-5l} <1(-5) 3.2 i 3.7(-4) 134Cs 1.54 0.01 - 1.21 0.02(-3 4.0 1.0(-5)

-0.2(-4))

136Cs 4.10 0.09 - 2.9 <2(-5) 137Cs 1.56 0.01 - 1.21 0.02(-3) 9.9 4.0(-5) 138Cs 2.9 1 0.1 - ) 1.0 i 1.4 -2) 1.9 0.5(-2)

. 140Ba 9.0 1 1.7 -4 1.7 i 0.3 -4) <1(-4) 140La 2.6 0.2 -3 2.0 i 0.3 -3) <1(-4) 91'

TABLE A.9 CVCS LETDOWN ACTIVITIES - 8/30/76

. Power Operations - Before Refueling CH-88: Placed in service, 3/24/76; Use,103 days by 8/30/76 Input to CVCS Output From- Output From Ion Exch CH Ion Exch CH-8B Filters 3 09:58; 8/30/76 10:02; 8/30/76 10:00; 8/30/76 Nuclide._ (uCi/ml) (uC1/ml) (uC1/ml)

Activation Nuclides: o 24Na 2.2 1 0.5(-3) 3.6 1.3(-4)- <1(-4) 54Mn 5.2 1 0.4(-4) 2.4 i 5.5(-5) 1.5 i 3.7(-5) 56Mn <1(-3) ssCo 2.05i0.09(-3) <4(-5) 9.2i1.5(-5) 60Co 8.713.1(-5) 3.011.5(-5) 4.0 2.0(-5) 187W 3.2 0.4(-2) <2(-4)

Iodine Nuclides:

131I 2.40 i 0.02 - 1.18 0.02(2) 2.0 0.2(-4) 1321- 1.01 i 0.01 - 2.0 0.4(-3 <8(-4) 133I 2.59 1 0.02 - 1.03 0.01( 2) <6(-4) 134I 5.3 1 0.7(-2) <2(-3) 135I 1.30 i 0.04(-1) <1(-2)

Fission Products:

88Rb 9.510.6(-1) 3.7 1 1.4(-1) 3.2 3.9(-1) 89Rb 8.5 i 0.4(-2) <1(-3) 95Zr <2(-4) ssNb <8(-5) 99Mo 3.3 0.4 - <6(-5) lo1Mo 4.2 1.0 - <2(-2) 110 mag 2.6 i 1.9 - 0.6 1.3(-4 2.1 5.5(-3) 134Cs 1.37 0.06( 3) 3.6 1 0.2(-4 9.4 5.5(-5) 136Cs 6.221.5(-4) 1.2 1 0.2 - 6 12(-6) ~

137Cs 1.46 1 0.04(-3) 3.8 0.3 - 9.3 1.6(-5) 138Cs 5.6i0.2(-1) 5.0 1 2.2 - 0.3 1.2(-1) 140Ba <3(-4) 140La 1.310.5(-3) <1(-4) ,

i 92 l

, , 7 7 TABLi ' 10 CVCS LETDOWN ACTIVITIES During Refueling (CH-8A and CH-8B in Series) f' . Input to CVCS Output From Output From Ion Exchangers. Ion Exchangers - Filters 16:45; 10/11/76 17:30; 10/11/76 .16:45; 10/11/76-Nuclide~ (uC1/ml) (uCf/ml) (u Ci/ml)

Activation Nuclides:

stCr <3(-3) <3(-4) 5.0 1.4(-5) 54Mn 3.4 0.1(-4) 3.5 0.2(-5) <5(-7) 59Fe 2.19 0.09(-4 2.6 1 0.5(-5 3.8 i 0.2(-5) 57Co 2.0 t 0.2(-4) ) 1.6 i 0.4(-5 <3(-6) seCo 1.24 0.02(-1) 9.1710.09(3) 5.4 i 0.2(-5) 60Co 5.2 0.2(-4) 4.5 0.4(-5) 6.1 i 0.6(-6)

Iodine Nuclides:

131I 1.64 0.02(-2) 1.50 i 0.02(-3) 3.0 1 0.3(-5) 1331 5.4 2.3(-5) <7(-5) <4(-5)

Fission Products:

95Zr <5(-3) <4 -4 <8(-6) 9sNb <3(-3) <2 -4 <5(-6) 99Mo 9.1 1.3(-4) 5.4 & 0.3(-5) 1.8 i 1.0(-5) 110mg <2(-4) <6(-5) <3(-5) 124Sb 5.4 0.6(-5) 9.0 2.0(-6) <3(-5) 134Cs 6.82 i 0.05(-3) 1.55 i 0.02(-3) 7.24i0.06(-4) 136Cs 8.5 0.2(-4) 7.3 0.3(-5) 1.0 i 0.1(-5) 137Cs 6.07 i 0.08 3) 1.53 0.02(-3)

• 7.8 i 0.1(-4) 140Ba 2.2 i 0.8 - <3-4) <2(-4) 140La 2.7 0.5 - <2 -4) <2(-5).

4 TABLE A.11

~DEMINERALIZER FUNCTION: REFUELING CAVITY WATER During Refueling.

Input: Cation Output: Cation

• Demineralizer CH-10 Demineralizer, CH-10 10:10; 11/8/76 10:15; 11/8/76 Nuclide (uCi/ml) (uC1/ml)

Activation Nuclides:

51Cr 4.4 1 0.3(-4 2.0 0.3 -

54Mn 1.3 0.2 - 1.1 0.2 -

59Fe 5.0 1.3 - 2.2 0.3 -

57Co 4.0 t 0.1 - <4(-7) seCo 1.8010.05 2) 3.2 2 0.6 -

60C0 1.7 0.4 - 2.2 0.5 -

65Zn 4.2 1 0.9 - 1.1 0.2 - ,

Iodine Nuclides:

131I 6.8 1.3(-4) 5.29 0.06(-4)

Fission Products:

95Zr 1.2 0. 3 - 4.2 t 0.3 -

9sNb 3.4 1.0 - 7.0 0.6 -

103Ru 1.8 0.3 - 5.3 0.3 -

110 mag 2.710.2(-5 4.2 0.2-6) 124Sb 1.4 0.1(-4) 1.31 0.08(-4) 134Cs 1.85 1 0.02(-3) 1.47 0.02(-5) 136Cs 8.S 0.4(- <3(-5) 137Cs 1.96 0.04 3) 1.70 0.02(-5) 14aBa 5.0 3.9(- <3(-6) 141Ce <7(-6) 1.9 0.3(-6) 144Ce <3(-5) 2.0 1 0.6(-6) -

I L

94

TABLE A.12 CVCS LETDOWN ACTIVITIES - Power Operations - After Refueling (2/16/77)

Input to CVCS Ion Exch CH-8A Output from CVCS Ion Exch CH-8A Dissolved Suspended Dissolved Suspended Activity Solids Activity Solids Ne:11de -(uCi/ml) (uCf/ml) (uCi/ml) (uti/ml)

Activation Nuclides:

24Na 1.83 0.05(-2) 5.0 3.1 i 0.6(-4)[A] <1(-6) s1Cr <2(-2) 3.78 0.5(-5))

0.01(-4 <8(-4) 3.58i0.06(-5) 54Mn 1.6 0.3(-4) 2.08 0.09(-5) <6(-6) . 1.29 0.02(-5) 56Mn <8 - 2.8 0.2 - 3.7 2.6(-3)[A] 6.5 1 0.4 -

59Fo <6 - 3.3 0.3 - <3(-6) 3.9 0.1 -

57Co <2 - 9.4 4.1 - <4(-5) 5.9 0.4 -

58Co 1.47 0.04(-3) 3.98 0.07( 4) 9.0 0.3(-5) 9.5 0.2 -

60C0 <3(-4) 2.620.1(-5 2.8 0.6(-6) 2.69 0.03-5) 65Zn <5(-4) <5(-6) <3(-6) 1.6 0.2(-6) lo7W 7.2 1.0(-3) 7.5 1.0(-5) <8(-5) 1.4 0.3(-6) 239Np <1(-2) 1.1 0.2(-4) <5(-4) 1.02 0.06(-5)

Iodine Nuclides:

131I 8.5 1 0.1 - 5.46 0.04(-3) 5.0 4.4(-5) 2.1 0.1(-6) 1321 4.7 0.2 - 2.52 0.03(-3) A' 1.8 0.6(-6) 133I 9.8 0.2 - 6.2 1 0.3(-3) <3(-3

<6 - ['A' 1.10 0.08(-6) 1 34I 3.4 1 2.0 - <2(-3) <2 - lA'. <6(-5) 135I 6.5 0.2 - 4.35 1 0.09(-3) <3 - 6.2 2.6(-7)

Fission Products:

7.2 *1 0.4(-1)[A]

• 3.2

• esRb 0.2(-1)[A]

89Rb <2 -

• ssZr <2(-4) 2.2 0.1(-5) <1 - 2.45 0.07 -

• 95Nb <9(-5) 1.65 0.07(-5) <6 - 3.08 0.08 -

99Mo 5.7 i 3.3(-3) 2.5 0.1(-5) <4 - 1.13 0.02 10 3Ru <4 - 6.7 i 0.8(-6) <8 - 4.8 0.5(-7 11ongg <4 - <2(-5) 2.9 1.2(-6) 1.8 0.6(-6 124Sb e4 - 2.4 0.5(-6) <7(-5) <3(-7) 134Cs 1.90 0.02(-1) 5.5 i 0.1(-4) 3.10 t 0.03(-3) 2.5 0.2(-6) 13sCs 1.410.5(-4)[A] <2(-5) 1.4 0.1(-5) <4(-6) 137Cs 2.19 0.02(1) 7.08 0.07(-4) 3.79 0.02(-3) 3.13 0.09(-6) 138Cs' 2.2 t 0.1(-1 [A] <2(-3)[A]

139Ba 3.6 1 0.6(-2 <2(-4) 7.2 2 0.7(-2)[A] <6(-6) 140Ba <2(-2) <5(-5) <3(-4 <9(-7) 140La 2.71 2.2(-2)[A] '4.4 1.2(-6) < (-4 2.4 0.7(-7

'141Ce <4 - <2(-6) <8 - 4.3 t 0.3 -

143Ce <4 - <3(-4) <2 - 4.0 1.2 - l 144Ce <2 - 1.1 1 0.5(-5) <4 - 8.4 t 1.7 -

I 95

TABLEA.12(cont'd)

CVCS LETDOWN ACTIVITIES - Power Operations - After Refueling (2/16/77)

Output from CVCS Filters Dissolved Activity Suspended Solids Nuclide (uC1/ml) (uCf/ml)

Activation Nuclides: "

24Na 1.4 i 0.1(-3) [A] <1(-7)

SICr <1(-3) 1.1 1 0.2(-6) o 54Mn 1.8 i 0.1(-5) [A] 8.3 1*0.8(-7) 56Mn <3 - [A]

59Fe <2 - <3(-7) 57Co <2 - 6.312.1(-8)

Seco 5.0 0.4(-5) 4.1 0.1(-6) 60Co 3.0 0.5(-6) 1.02 i 0.03(-6) 652n <2 - <2(-7) 187W <3 - <1(-6) 239Np <8 - 2.3 0.2(-6)

Iodine Nuclides:

131I 6.44 0.07 7.4 1 0.6 -

132I 3.8 1 0.4 - 3)A' 2.9 0.2 -

133I 7.7 0.7 - 'A

. 8.2 0.2 -

134I <4(-3) [

135I 7.2 0.6(-3) [A] 5.6 0.1(-5)

Fission Products:

serb 3.3

• 0.2(-1) 'A]

• 89Rb 5.2 + 1.4(-3) lA]

95Zr <2 - 3.4 1 0.4 -

95Nb <7 - 3.3 i D.3 -

99Mo <2 - 1.f 0'. 3 -

103Ru <4 - 7.4 3.2 -

110 mag <4 - 1.5 1.9 -

124Sb <4 - 3.6 3.3-8) -

134Cs 1.63 0.01(-2) 7.1 0.2 -6) 136Cs 2.4 <3(-7) 137Cs 1.89 0.4(-5))

0.02(-2 9.711.3(-6) 13eCs 2.7 -

139Ba 7.1 0.3(-2)A]((A]

0.4(-2)

• 140Ba <2 - <2(-6) thola <6 - 1.3 0.3(-7) 141Ce <3 - 9.1 2.2(-8) 143Ce <3 - <6(-6) 144Ce <2(-4) <4(-7)

• : Sample not counted soon enough for these radionuclides.

[A] : Represents total activity (dissolved and suspended) in an unfiltered sample. Filtrate counted too late to observe nuclide.  ;

96

u

~ ~( , .-

TABLE A.13 STEAM GENERATOR BLOWDOWN ACTIVITIES Before Refueling: . After Refueling: , t Generator A Generator B Generator A Generator B Start Sagle: 14:30; 2/8/77 Start Sample: 14:20; 2/10/77 09:40; 8/26/76 09:40; 8/26/76 Time: 17.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Volume: 211 Liters Volume: 450 ml Time: 17.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Volume: 212 Liters Volume: 450 ml Dissolved Suspended Dissolved Suspended Total Activity Total Activity Activity Solids Activity- Solids

.Nuclide (uC1/ml) (pCi/ml) (uci/ml) (uci/ml) (uci/ml) (uC1/ml)

Activation Nuclides:

24Na * *

• * <9(-11) <4(-11) <6(-11) <6(-11) 51Cr

<8(-10) 4.0

• 0.5(-10) <6(-10) 3.9 t 0.8(-10) 54Mn <1(-7) <1(-7) 1.0 9.5(-9) 3.0

.*

• 0.2(-10) 6.2 0.6(-10) 7.5i1.9(-11) 59Fe 7.7 e * *

<3(-10) 1.7(-11) <2(-10) <2(-10)-

57Co <2(-10) 6.0 seCo * '*

4.2(-12) <5(-11) <2(-11) 60Co 3.57i0.09(-9) 1.63 i 0.05(-9) 4.4 0.3(-10) 3.4 0.3(-10)

<2 (-7) <2(-7) 1.09 0.04(-9) 6.9 ' O.2(-10) 3.3 0.3(-10) 3.1 = 0.3(-10)

Irdine Nuclides:

131I <4(-8) <2(-7) 8.9 0.6(-10) 4.1 1331 *

• 0.6(-11) 1.' O.3(-10) <2(-11)

<4(-10) <4(-11) ^ -10) <8(-11)

Fission Nuclides:

95Zr * *

<4 - <2(-10) <2 - <1 - 1' 95Nb *

<3 - 1.1 0.5(-11) <7 - <6 -

99Mo <2 -

10 3Ru * * <3-

<9(-12) 6.2 0.8(-11)

<5 -

<2 -

<2 -

<8 - J

)h 124Sb

• 2.8 1.1 - ) 4.5i1.2(-11) 9.0 10.1(-11) <7 -11) 134Cs <1(-7)* <1(-7)* 7.8 0.1 - 3.5 0.2(-10) 2.32 t 0.08(-9) 6.3 i 0.9(-11) 136Cs 8.4 t 4.4 - ) <1(-10) <7(-11) <6(-11) 137Cs <1(-6) <1(-6) 1.23 0.02-8) 5.1 i 0.2(-10) 3.61 0.08(-9) 1.4 0.4(-10)

l l

TABL E A.13 (cont'd)

STEAM GENERATOR BLOWDOWN ACTIVITIES Before Refueling: After Refueling:

Generator A Generator B i Generator A Generator B Start Sample: 14:30; 2/8/77 Start Sample: 14:20; 2/10/77 09:40; 8/26/76 09:40; 8/26/76 Time: 17.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Volume: 211 Liters Time: 17.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Volume: 212 Liters Volume: 450 ml Volume: 450 ml Dissolved Suspended Dissolved. Suspended Total Activity Total Activity Activity Solids Activity Solids Nuclide (uCi/ml) (uC1/ml) (vC1/ml) (uC1/ml) (vCi/ml) (uC1/ml) 53 eta Only, Nuclides:

3H 1.7 i 0.2(-5) * *

• 14C <3(-8) * *

• 32P <4(-8)

• m.ssFe 5.2 *

• 0.1 (-6) * *

• 63Ni <8 -

893p * * *

<4 90Sr <2 - * * *

.91y <1 * *

• O : Sample not analyzed for these radionuclides.

e. D

TABLE A.14 SPENT FUEL POOL ACTIVITIES Power Operations - Before Refueling Spent Fuel Pool Spent Fuel Pool Temp: 78 F Temp: 75*F

.10:49; 8/23/76 10:30; 8/31/76

'P Nuclide (uCi/ml) (uC1/ml) 56Mn 1.67 i 0.03(-5) 1.72 0.06(-5) 57Co 6.4 0.4(-6) 5.6 i 0.2(-6) 58C0 8.7 0.l(-4) 8.61 0.04(-4)

\ 4.03 60Co 0.04(-5) 4.18 0.06(-5) 4 65Zn <l(-6) 1.1 0.2(-6)

, 95Zr 7.4 2.1(-7) <9(-7) 9sNb <4(-7) <4(-7) 99Mo 9.2 i 7.1(-7) <3(-7) 110 mag 2.2 0.l(-6) 3.3 0.6(-6) 124Sb 3.911.0(-7) 5.9 0.9(-7) 134Cs 9.210.l(-5) 9.6 0.l(-5) 13sCs <5(-7) 2.7 1 1.9(-7) 137Cs 1.93 0.02(-4) 2.06 0.04(-4) l l

I l

l 99 e m-

TABLE A.15_

SAFETY INJECTION AND REFUELING WATER TANK ACTIVITIES' Durina Refueling .

Before Refueling After Refueling 10:04; 10/12/76 14:25; 12/2/76 Mclide (uCi/ml) (uCi/ml) .

Activation Nuclides:

51Cr 2.4 0.5(-5) 6.4 .

s% 1.18 1 0.01( 4) 2.19 1.7(-4))

0.05(-4 59Fe - 7.0 0.4(-6 4.3 0.8(-5) 57Co 2.9 0.1(-5 - 7. 3 0.8(-5) seCo 5.6310.06(-3) 2.52 1 0.07(-2) soCo 1.41 t 0.01(-4) 3.09 0.05(-4) 65Zn 2.5 0.4(-6) <4(-5)

Iodine Nuclides:

1311 7.5 0.2(-4) <8(-5)

Fission Products:

95Zr <2 - 3.611.4(-5) 95Nb <2 - 5.6 1.5(-5) 10 3Ru <4 - <2(-4) 110 mag 1.3 0.1(-5) 2.6 0.5(-5) 124Sb 2.4 0.2(-6) 8.4 0.5(-5) 134Cs 1.44 0.02(-3) 3.97 1 0.05(-3) 136Cs 8.7 1 0.2(-5) 4.4 1.0(-5) 137Cs 2.28 0.02(-3) 4.1010.08(-3) 140Ba 9.0 2.1(-6) <6(-4) 141Ce <2 - <1 -

144Ce <9 - <5 -

Beta Only, Nuclides:

3H 9.25 i 0.04(-3) 9.91 0.04(-3) 14C 8.1 0.2(-7) 3.

1.(-8) 32P 6. 1 5.(-7) 35S 4.9 1.2(-7) 1.42 0.04(-5) '

( 45Ca <6(-8) ssFe 2.23 0.02(-4) 1.27 0.02(-3) 63Ni 1.09 0.02(-3 3.46 0.02(-3) 89Sr 7.1 0.7(-7)) 1.55 0.08(-5)

-90Sr 8. 2.(-8) 2.7 0.1 -

91Y 1.9 0.2(-7) 6.5 0.2 -

l 147Pm 3.1 0.2(-7) 4.4 0.1 -

• Sa g le not analyzed for these radionuclides.

100 l

l l

~

TABLE A.16 FUEL TRANSFER CANAL AND REFUELING CAVITY ACTIVITIES

.During Refueling Before Fuel Before Fuel After Fuel Movement Movement Movement

-Fuel Transfer Refueling Refueling Cavity Cavity.

f Canal 18:30; 11/1/76 18:00; 11/1/76 09:45; 11/8/76 Nuclide (uCi/ml) (uCi/ml) (uci/ml)

Activation Nuclides:

51Cr 1.5 0.7(-4 6.0 1.2(-4) 6.4 0.4 -

54Mn 6.5 2 0.7(-5 1.62 0.03(-4) 1.6 0.3 -

59Fe 1.1 1 0.1(- 3.9 0.4(-5) 7.3 2.5 -

57Co 3.54 1 0.07 6.3 0.4(-5) 3.94 0.07-5) seco 1.44 0.04 - 3.46 0.08(-2) 1.84 0.05 2) 60Co 1.12 0.02 - 2.00 0.07(-4) 1.7 0.4(-

65Zn 1.8 0.7(-6) <3(-5) 6.6 0.9(-

Iodine Nuclides:

131I 2.2 0.9(-4) 9.7 1.7(-4) 6.2 1.3(-4)

Fission Products:

95Zr <2(-3) <4 - <3(-3) 95Nb <2(-3) <2 - 1.5 0.3 -

10 3Ru 1.1 0.5(-5) <1 - 1.4 2 0.3 -

110 mag 2.66 0.09 - 1.30 0.06( 4) 5.7 0.2 -

124Sb 1.19 0.09 - 1.410.1(-4) 1.6 0.3 -

134Cs 2.06 0.02 - 2.59 0.03 - 1.85 0.03(-3) 136Cs 1.04 0.06 - 1.29 0.08 - 6.2 1.1(-5) 137Cs 2.37 0.04 - 2.76 0.05 - 1.96 0.04(-3)=

140Ba <4(-4) <4 -4) <5(-4) 141Ce <4(-6) <2-5) 9.2 3.1(-6) 144Ce <2(-5) <7-5) 2.2 0.8(-5) r.

- Beta. Only, Nuclides:

3H 9.29 0.05(-3) 1.14 0.02(-2) 1.19 0.02(-2) 14C 1.02 1 0.02(-6) 1.02 1 0.02(-6) 4.83 0.04(-6) 32P 5. 2.(-7) 2.0 0.5(-5) 1.3 0.3(-5) 355 7.. 2 (-7) ' 02 1 0.03(-5)

. 1.12 0.03(-5) 4sCa <6(-8) 55Fe 2.9710.02(-4) 3.99 0.02(-4 8.03 0.04 -4) 63Ni 9.71 '0.02(-4) 1.47 0.02(-3 9.25 0.03 -

89Sr 2.2 0.2 - 1.15 0.67 - 1.04 0.05 -

90Sr 2.2 0.3 - 1.40 t 0.07 - 1.07 0.06 -

91Y 2.4 i 1.0 - 1.73 0.07 - 5.2 0.1(-6) 147Pm *

• l.87 0.05(-6)

• Sanple not analyzed for these radionuclides.

101

TABLE A.17 SPENT FUEL P0OL ACTIVITIES

~

During Refueling Refueling Cavity Refueling Cavity

.. . Filled- Filled

_ Before Filling: Before Fuel After Fuel After Draining:

Refueling ' Cavity Transfer Transfer -Refuelino Cavity Teg: 67 F Temp: 58'F- Teg: 59'F Temp: 57'F _

10:05; 10/14/76- 11:05; 10/27/76 14:10;11/8/76 14:10; 12/2/76 ,

(

Nuclide ~(uCi/ml) (uC1/ml) (uC1/ml) (uC1/ml)

Activation Nuc1kdes: a 51Cr <5(-6) 7.6 2.5 - <7(-5) 54Mn <3(-6) 1.53 i ' 0.02(-5)1.87 0.07(-5)' 8.3 0.3 - 8.2 i 0.2(-5) 59Fe- <4(-6) <5(-6) 9.5 i 1.3 - <2(-5)

.57C0 4.5 0.1(- )- 5.9 0.4 - 3.7 t 0.1 - 4.2 0.2(- )

seCo '5.07i0.06-4) 6.2 0.2 - 1.54i0.04(2) 1.38 i 0.04 -2) 60Co 4.53 i 0.05 -5) 5.0 0.1 - 1.22 t 0.02(-4 1.48 0.03 -4) 65Zn <4(-6) <4(-6) 2.220.9(-6)) 5.7 1.9(-6)

Iodine Nuclides:

131I <3(-7)- <6(-7) <6(-6) <8(-6)

Fission Products:

95Zr <4 - <7 - <3- <2 -

95Nb <2 - <4 - <2 - <1 -

103Rh <8(-6) <1 - <1(-4) <1(-4) 4 110 mag 1.53 i 0.09(-6) 2.5 0.3(-6) 4.0 0.6(-6) 8.4 2.7(-6) 2.9

+

124Sb 0.3(-7) 5.1 1.5(-7) 1.1 1.1 i 0.1(-5) 134Cs 8.2 i 0.1(-5) 1.21 1 0.02(-4) 2.52 0.3(-6) 0.02(-3 ) 2.5510.02(-3) 13sCs <2(-5) 2.3 0.7(-6) 6.6 0.6(-5)'

137Cs 1.87 0.02(-4) 2.53 0.05(-4) 2.71 0.05(-3) 1.5 2.63 i 0.3(-5) 0.04(-3 )

140Ba <3- <4-5? 1.3 0.4(-4) <4 -

141Ce <3 - <2 -6J <6 -6 <8-144Ce - <2 - <6 -6) <3 -5 <4 -

. Beta Only. Nuclides: -

3H 7.20 1 0.03(-3) 7.32 0.05(-3) 9.21 0.05(-3) 9.12 0.04(-3) 14C 6.7 0.2(-7) 1.16 0.02(-6) 8.8.+0.2(-7) 3. 1.(-8) 32P <2 - <1 - <4(-7) * ~

ass. <2 - <2 - 5. 2. 1.7i0.2(-6) 45Ca. <6 - .6-

< <6(-8)(-7)

• 55Fe 4.9' O.2(-6) 1.25 0.04(-5) 7.53 0.05(-5) 2.41 0.03(-5) 63Ni - 2.55 0.02(-4). 2.79 0.04(-4) 9.74 0.03(-4) 1.02 0.02(-3) 89Sr. <4 - <4-8) 7.0 '0.4(-6) 6.8 0.3(-6) 90Sr <2 - <2 -8) 7.2 0.5(-7) 8.4 0.5(-7) 91Y' .1-

< <8-8) 2.92 0.08(-6) 3.37 0.09(-6) 147Pm 2.5i.0.2(-7) .1.46 i 0.08(-7) 1.9 1 0.2(-7) 6.8 0.2(-7)

• J: Sagle not analyzed for these radionuclides.

'102

I TABLE A.18 SPENT FUEL POOL ACTIVITIES Power Operations - After Refueling -

Temp: 55*F P 08:45; 2/22/77 Dissolved Activity Suspended Solids Nuclide (uti/ml) (uti/ml)

- Activation Nuclides:

54Mn 1.25 0.04( 5) 1.26 i 0.03(-6) 59Fe. 1.7t0.3(-6 3.4 0.2 -

57Co 7.3 i 0.3(- 4.5 i 0.4 -

seCo 1.38 0.02 3) 7.0 i 0.2 -

60C0 4.54 0.07 -5) 2.26 0.04(-6) 65Zn <2(-6) <2(-7)

Iodine Nuclides:

131I 3.31 0.05(-5) 1.7 i 0.1(-7)

Fission Products:

95Zr <1(-4) 7.5 1.5(-8) 95Nb- 1.1 0.4(-6) 8.2 0.9(-8) 99Mo <4 - <8(-9) llomAg <2 - 8.2 3.7 -

124Sb <3 - 1.8 0.6 -

134Cs 7.6

• 0.2(-5) 4.5 0.3 -

136Cs- <6(-5) 4.9 2.1(-8) 137Cs 1.38 0.02(-4) 6.1 0.2(-7)

)

Beta Only, Nuclides:

3H 8.94 0.03(-3)[A]

14C 1.8 55Fe 1.12 0.1(-))

0.02 -4 63Ni 3.42 0.02 -4) 895r 5.5 1 0.6 -

90Sr 1.7 0.2 -

91Y 4.8 0.2 -

[A]: Total activity, suspended solids activity not determined in beta analysis.

' 103

l TABLE A.19 ~

. REACTOR C0OLANT DRAIN TANK ACTIVITIES (TANK WD-1)

Power Goerations - After Refueling 09:15; 2/10/77

. Dissolved Activity Nuclide (uC1/ml) Susp(ended uC1/ml) Solids Activation Nuclides:

24Na ' 1.18 t 0.03(-3) 1.1 1 0.2(-5 51Cr' <6(-3) 7.0 i 0.7(-5 5% 2.27 2 0.05( 5)

' 1.5620.06(-4) 56Mn <3(-4) <1(-6) e-59Fe. -<6(-6) 1.12 t 0.06 5)

. 57Co 3.66 0.34 - 3.1 2.9(-

58Co 3.46 1 0.04 - 2.19 0.02 4)

SOCo 8.77 0.29 - .3.54 1 0.05 -5) 65Zn 1.78 0.60 - <5(-6) 187W 1.17 i 0.11 - 2.7 0.5(-5) 239NP <2(-3)' <1(-3)

Iodine Nuclides:

131I 3.49 0.06 - 8.3110.08-3 9.60 2 0.67 -

1321

'133I 134I 1.99

<7(-5) 0.03 - 9.110.8(-}i) 4.7 2*0.2(-

13sl 3.09 0.09(-3) 8.9 i 0.2(-4)

Fis' .aucts:

ssZr <1(-4) 4.2 0.4(-6) 95Nb <7(-5) 1.32 i 0.05(-5) 99Mo 5.11 1 0.57(-5) 2.2 i 1.0(-6) lo3Ru <6(-5) 1.5 1 0.4(-6) 110 mag 1.67 1 0.39f- 1 <2(-6) i 124Sb 8.34 i 1.05f- h 3.9 i 2.3(-7) 134Cs 2.33 1 0.02f' h 2.2 0.1(-5) t 136Cs 2.70 <5(-6) 137Cs 2.98 0.48ll-)

0.04l -

) 3.16 0.05(-5) 139Ba <3 - '

140Ba <2 - <5 -

140La <3 - <7 -

141Ce <1 - <2 -

-143Ce <1 - <4 - ~

144Ce' '6-

< . <6 -

Beta .Only Nuclides:

-3H 1.13 0.01(-1) 14C - <1(-7) ssFe 1.5420.02(-4)

.63Ni 5.39 1 0.02(-4) 89Sr 1.33

• 0.07(-5)

-90Sr 1.38 0.09(-6 91Y "8.310.9(-7) )

• ~:' Sample not counted soon enough for these radionuclides.

• • : . Suspended solids activity not determination in beta analy' sis.

104

_____.__-________._.____l___.__ _ _ _ . -.

• _ _ _ _ _ _ w - . - - .-- M

o .

y 9 TABLE A.70 SPENT REGENERANT AND WASTE HOLDUP TANKS Pnwer Onorations - Before Refuelino Spent Regenerant Spent Regenerant Spent Regenerant Waste Holdup HSL Analysis: Waste Holdup Tank WD-138 Tank WD-13A Tank WD-138 Tank WD-48 Duplicate' Sample Tank WD-4A

~10:48; 8/24/76 09:42; 8/26//6 11:50; 8/31/76 13:34; 8/24/76 11:40; 9/2/76 11:40; 9/2/76 Nuclide (uC'/ml) (uCi/ml) (uCi/ml) (uCi/ml) (uCi/ml) (uC1/ml)

Activation Nuclides:

24Na 2.510.9(-6) 2.9 0.3(-6) 3.4 1 0.3(-5) 4.9 2.4(-7) <3(-5) siCr IL7 0.2(-5) <6(-4) <5(-4) <7(-4) <5(-2) 54Mn 6;53 1 0.07 5) 1.65 0.07(-5) 1.20 0.06(-4) 8.0 1.2(-6) 4 2(-4) 1.7 1 0.8(-4) 59Fe 3.1 1 0.4(- 3.8 1 2.6(-5)

<1(-6) <6(-7) <2(-4) 57C0 3.7 0.3(- 1.0 0.1(-5) <8(-7) 1.1 1 0.4 -

ssCo 4.81 1.2 1i 0.01(-4 0.05(-4) 2.96 0.6(-6) ) 2.74 0.01(-4) 6.2 0.1(-5) 2.600.08(-3) 1.4 1 0.1 -

60Co 1 08 i 0.01(-4) 2.3 0.1(-5) 6.74 0.06(-4) 6.1 0.4(-6) 1.8 0.7(-4) 2.6 0.7 -

5 sn ssZn 5).411.9(-6) <1(-6) 2.4 0.4 - <6 - <1(-4) 187W <2(-5) <7(-6) 2.0 0.4 - <4 - <4(-4) 239Np <2(-5) <4(-4) 3.6 i 1.8 - <5 - 3.5 i 1.2(-3)

Iodine Nuclides:

131I 7.77 0.07(-4) 6.80 1 0.07(-3) 1.65 i 0.01 - 5.06 0.07(-3 1.21 0.04(0) 1.27 i 0.01(0) 133I 1.43 0.03(-4) 2.72 1 0.02(-4) 1.26 i 0.01 - 6.9 1 0.2(-5) ) 2.29 1 0.01(-1) 135I 6.5 i 0.4(-5) <2(-6) 3.04 0.02 - <2(-6) 2.7 0.4(-3)

TABLE A.20 (cont.)

SPENT REGENERANT AND WASTE HOLDUP TANKS Power Oneratinnt - Before Refuelina Waste Holdup Spent Regenerant Spent Regenerant Spent Regenerant Waste Holdup - HSL Analysis:

Tank WD-138 Tank WD-13A Tank WD-13B Tank WD-4B Duplicate Sample Tank WD-4A 10:48; 8/24/76 09:42; 8/26/76 11:50; 8/31/76 13:34; 8/24/76 11:40; 9/2/76 11:40; 9/2/76 Nuclide (pC1/ml) (pCi/ml) (vC1/ml) (uti/ml) (pCi/ml) (uC1/ml)

Fission Products:

ssZr 2.8 0.6 - <3(-6) <4(-6)- <3(-6) <3(-4) 95Nb 3.120.4- <1(-6) <2(-6) <1(-6) <1(-4) 99Mo 6.2 1.5 - 11 1 0.08(-5) 1.9 0.2(-5) 9 2 1 0.7(-6) 1.9 i 1.0(-3) 103Ru <7(-7) <2 - <2(-5) <6 - <2I;-

11amA9 2.0 0.3(-6) <2 - 8.6 1 1.5(-6) <6- <2I -

124Sb <3(-7) <2 - <4 - <4 - <6il-126Sb 5.7 2.7(-7) <2 - <5 - <5 - <2i -

g 127Sb <1(-5) <4 - <1 - 1.8 1 0.6(-6) ' <3d-134Cs 3.32 0.02-4) 2.2310.01-3) 1.49 0.01 -3) 1.7810.08(-3) 1.5310.05 - 1.56 2 0.05(-1) 136Cs 5.32 1 0.08 -5\ 2.00 1 0.04 -4) J.95 1 0.06 -4) 2.34 0.05 4) 4.5 1 0.2 - 4.6 i 0.2(-2) 137Cs 3.48i0.01-4) 2.57 0.02 -3) 1.49 i 0.01 -3) 1.8610.06j-3) s 1.62 0.05 - 1.47 1 0.05(-1) 140Ba <5(-6) <7(-5) <2(-5 <2(-4) <6(-3) .

14cLa 6.8 1.5(-7) 4 6 i 1.2(-7) <2(-7 1.1 1 0.2(-6) 6 7 t 2.9(-5) 141Ce <1(-6) <1 -6) <5(-6 <2(-6) <3 -

143Ce 4.711.0(-6) <1-4) <1(-4) 2.6 0.8(-6) <1 -

144Ce <6(-6) <6 -6) <2(-5) <6(-6) <1 -

e t

9 E

o .

o u TABLE A.21 LIQUID RADWASTE TANK ACTIVITIES During Refueling Waste Holdup Waste Holdup Waste Holdup Spent Regenerant Monitor Tank Tank WD-4A Tank WD-4B Tank WD-4C . Tank WD-138 - WD-228 20:30; 10/11/76 23:55; 10/11/76 16:00; 10/11/76 12:35; 10/13/76 07:30; 10/14/76 Nuclide (uCi/ml) (uCi/ml) (uC1/ml) (uci/ml) fuCf/ml)

' Activation Nuclides:

24Na <3(-5) <2s-5) <8(-6) <4(-4) <4(-7) 51Cr <8(-3) 1.4 0.3(-4) 1.1 4.0 <6(-7) 54Mn 1.18 0.07(-4) 4.04 0.04(-4) 6.45 0.5(-4) 0.08(-4 ) 3.04 0.7(-5) 0.04(-4 ) 4.2 0.3(-7) 59Fe <3(-5) 3.7 0.3-5) 6.9 0.3(-5) 1.40 0.04(-5 <2(-7) s7Co <4(-5) 1.3 0.4 -5) 1.8 0.2(-5) 1.39 0.04(-5 8.6 1.9(-8) ssCo 5.5 0.1(-4) 8.2 0.1 -3) 5.82 0.07(-3) 3.35 0.07(-3 4.7 60Co 1.1 C.4(-4) 2.5 0.2 -4) 1.4 0.1(-4) 5.34 0.09(-4 1.05 0.1(-6) 0.04(-6 )

ssZn <3-5) <2-5) <2(-5) 8.5 0.8(-6) <2 -7) 187W <2 -4) <2 -3) 8.0 2.0(-5) <5(-4) <1 -6) 239Np <6 -3) <2 -3) <2(-3) <7(-5) <4-7)

Iodine Nuclides:

g 131I 2.37 0.03(-2) 7.69 0.05(-3) 9.83 2 0.07(-3) 5.46 0.08(-4) 6.4 0.3(-7) u 133I <3 -3 <5 -4 1.2 i 0.5(-5) <4(-4) <4(-7) 135I <2 -4 <4 -5 <6(-5) <8(-2) <1(-5)

Fission Products:

95Zr <6(-5 <6(-4) <4(-4) 5.3 2 0.8 - <6-7) 95Nb <4(-5 <4(-4) <3(-4) 6.3 0.4 - <3 -7 99Mo <4 - 5.8 0.4(-5)- 4.5 0.4(-5) 4.9 0.9 - <3-8ll l 103Ru <3 - <5(-4) <4 - <5(-5) <2 -

lion %g <3 - <5(-4) <4 - 4.4 0.8(-6) <2 )1 124Sb <3 - 5.3 i 1.9(-6) <4 - 4.8 1.1(-7) <2 -J 126Sb <3- <5(-4) 2.3 1.1(-5) <4(-5) <2 (I-7) 1275b <9(-5 2.6 1.2(-5) <7(-4) <4(-4) <8L-7) 134Cs 5.84 0.05 - 1.12 0.01 - 9.31 0.08(-3) 1.92 0.03(e?) 3.8 0.2(-6) 136Cs 2.1310.03- 6.43 0.08 - 9.48 0.08(-4) 9.2 0.2(-5) 137Cs 5.92 t 0.08 - 1.21 1 0.02 - 9.5 0.1(-3) 2.18 0.03(-3) 2.5 4.59 t 0.6(-7) 0.07(-6 )

140Ba <9 -3) 5.7 1.7(-5) <2(-3) <2(-4) <7(-7) 140La <4 -3) 2.8 0.2(-5) 3.0t0.2(-5) 1.80 0.04(-5) <3(-7) 141Ce <7-5) <2(-5) <2(-5) <4(-6) <7(-8) 143Ce <2 -3) <6(-4) <4(-4) <3(-5) <2(-7) 144Ce <3(-4) <6(-5) <8(-5) <2(-5) <3(-7) 1

TABLE A.22 WASTE HOLDUP TANK ACTIVITIES (Tank WD-4A)

Power Operations - After Refueling w

Analyzed by HSL 10:40; 2/11/77 10:40; 2/11/77 Dissolved Activity Suspended Solids Total Activity

Nuclide (uC1/ml)- (uci/ml) (uC1/ml)

Activation Nuclides: 'C 24Na <1(-6) <3(-6) 51Cr <8(-4) <5(-4) j 54Mn 6.6 t 0.1 - "

1.29 i 0.03(-4) 1.7 0.2(-4) 59Fe <3(-6) 1.8 0.2 -

57Co 6.42 i 1.62 - 4.9 3.9 -

58C0 8.67 1 0.08 - 6.25 0.07 5) 1.17 0.03(-3 j

i 60Co 3.26 0.12 - 1.36 0.02 -5 4.4 t 0.4(-5) )

65Zn 187W

<2(-5) 3.1 0.8(-7))

<3(-5) <5(-6) 239Np 2.33i1.06(-4) <3(-4)

Iodine Nuclides:

131I 9.44 0.05(-3) 2.5410.09(-3) 1.4 133I 0.2(-2) las!

2.20 i 0.29(-4) 6.9 i 0.2(-5)

<5(-6) <4(-5)

Fission Products

95Zr <2(-6) 4.7 1 0.7(-7 ssNb <9(-6) 7.2 i 0.7 -

99Mo 4.42 3.13(-5) 2.7 0.5 -

103Ru <1(-4) 2.8 0.7 -

4 11049 1.13 i 1.11 - 5.6 1.0 -

124Sb- 1.34 1 0.53 - 4.7 t 3.4 -

134Cs 1.08 1 0.01 - 2.11 0.03-5) 1.18 0.02(-2) 136Cs 9.63 i 3.68 - <2(-6) ,

137Cs 1.23 0.01 - '2.81 0.04(-5) 1.57 0.03(-2) 140Ba <4(-4)

<4(-6) l 140La 1.86 i 0.11(-5) 3.4 i 1.1(-7) -

141Ce - <7(-6) <3 -

143Ce <2(-4) <1 -

144Ce .7.31i2.02(-5) <2 -

Beta Only, Nuclides:

' 3H 5.56 i 0.02 - **

14C 6.89 i 0.03 -

ssFe 1.43 0.02 -

! 63Ni 1.51 0.02 -

i 89Sr 90Sr-4.9i0.2(-5) 1.13 i 0.06(-6) 91Y 1.7 i 0.1(-7) l

• • : Suspen'ded solids activities not determined on these sanples.

' 108 t

- , , , - m. - - , - -

7- -er - . . .~, _ _ . _ _ a -

TABLE A.23 SPENT' REGENERANT TANK ACTIVITIES Power Operations - After Refueling Tank WD-13A Tank WD-13B 10:20; 2/11/77 09:11; 2/14/77 Dissolved Suspended Dissolved Suspended

. Activity Solids Activity Solids Nuclide (uci/ml) (uCi/ml) (uC1/ml) (uci/ml) p Activation Nuclides:

2i.Na <6(-81 <4(-6) 2.61 0.09(-5) <8(-6) 51Cr <5(-6} 3.2 t 0.4 - <4(-5) 6 56Mn 2.11 0.04(-5) 5.7 i 0.1 - 1.96 0.04(-5) 5.9 1.09 2 0.02(-5 0.4(-6) )

59Fe 6.7i1.3(- 7.4 i 0.5 - 5.8

• 1.3(- 1.3 0.3(-6) 57C0 2.8 i 0.2(- 2.3 0.3 - 2.6 1.0.2(- 8.0 0.3(-7) 58Co 3.07 i 0.07 4) 3.09 t 0.08 -5) 3.36 0.06 4) 1.12 t 0.03(-4) 60Co <6-ll 1.16 0.02 -5) 1.29 0.01 -5) 2.22 i 0.04'(-5) 65Zn <1 ,1 <5 - <1(-6) 2.2 0.7(-7) 187W <2 - <7 - 2.0 1 0.3(-5) <2(-5) 2 39Np <4 - ) <4 - 5.3i1.9(-6) <9 (-7 )

Iodine Nuclides:

1311 3.07 i 0.06(-5) 1.56 1 0.06(-5) 4.89 0.06 - 1.09 0.08(-6) 133I <3(-6) 2.4 0.6(-6) 3.22 0.06 - <4{-6 135I <2(-7)

• 1.59 i 0.06 - <1s 4 Fission Products:

95Zr <1 - l 2.0 0.4 - <2(-5 5.8 0.8 -

95Nb <6 - l 4.0 1 0.5 - <8(-6 1.0 0.1 -

99Mo <2 - l 3.2 i 2.0 - 5.6

• 4.1(-7) 8.9 2.0 -

103Ru <3 - <2 - <5(-6) <5(-7) 110mA9 1.9i1.1(-7) <2 - 6.821.2(-7) 1.1 0.1(-6) 124Sb <3(-6) <2 - <5(-7) 134Cs 2.52 0.02(-4) 1.5 0.1(-6) 2.4 2.35 i 0.8(-7) 0.02(-4 ) 6.2 0.2(-6) 136Cs <1(-6) <2(-6) 3.1 1.1(-7) <5(-6) 137Cs 3.05 t 0.04(-4) 2.14 0.07(-6) 2.87 0.03(-4) 8.5 0.2(-6) 140Ba <1(-5 l <5(-7) <2 - l

. 140La <2(-6 <3(-6) <2 - <2-6'h

<6 -7 141Ce <4 - 5.0 1.8(-8) <5 - l 1.111.1(-5) <2-7;j

<3 -7 14 3Ce <1 - 4.7 i 4.7(-6) 144Ce <2 - <2(-7) <2(-6) 2.6 1.0(-7)

Beta Only Nuclides:

Sli 4.14 0.02( 3) 14C '2.5 i 0.1(-7 ssFe 1.72 i 0.02( 4) 63Ni' -8.57 0.02( 5) 89Sr 5.0,1 0.7 -7 90Sr 1.5'i 0.3 7 91Y 4. 2 2.(-

• : Sanple not counted soon enough for these radionuclides.

• • : Sample not analyzed for suspended solids.

• • • : Beta analysis not done on this sample.

109 .

TABLE A.24 MONITOR AND HOTEL TANKS Power Operations - Before Refuelino Monitor Tank Monitor Tank Monitor Tank - Hotel Tank.

WD-22A WD-22B' WD-228 WD-15A 15:15; 8/24/76 13:25; 8/24/76 08:17; 9/1/76 10:43; 8/24/76 Nuclide (uCi/ml) (uC1/ml) (uC1/ml) (uC1/ml)

Activation Nuclides: '

I SICr <2(-6) <2(-6) <2(-6) <4(-7) 54Mn 5.5 0.6(-7) 8.5 1 0.7(-7) 59Fe <2(-7) 2.7 i 0.3(-7) 3.2 0.5(-7) s7Co

<2(-7) <1(-7) <2(-7)

<2(-7) <2(-7) <9(-5) <1(-7) seco 3.4 0.l(-6) 60Co 8.2 2.6(-7) 5.0 i 0.1(-6) 8.8 i 0.4(-7) 1.42 0.08(-6 ssZn 7.4 i 1.7(-7) 5.9 0.4(-7) 5.9' O.7(-7))

<2(-7) <2(-7 <1(-7) <2(-7) 187W <4(-7) g 2 39Np

<4(-7 <2(-7)- <3(-7)

O

<2(-6) <1(-6 <l(-6)' <3(-7)

Iodine Nuclides:

1311 1.42 133I 0.02(-5) 1.20 0.04(-5) 2.2 i 0.2(-5) 1.2 t 0.5(-7) 5.9 1.9(-7) 1.8 1.2(-7) 8.9 0.7(-7) <1(-7)

~

. .. u. 9 TABLE A.24 (cont.)

MONITOR AND HOTEL TANKS Power Operations - Before Refuelin9 Monitor Tank Monitor Tank Monitor Tank Hotel Tank WD-22A WD-22B WD-22B WD-15A 15:15; 8/24/76 13:25; 8/24/76 08:17; 9/1/76 10:a3; 8/24/76 Nuclide (t.Ci /ml) (uCi/ml) (uC1/ml) (uCi/mi)

Fission Products:

952r <2(-7) <3(-7) <1 -7) <2(-7)

"a :ll:H :ll:H  :::0

<1 -

"f:0

<9(-8 to3Ru <2-l l <2 -

lion %g <2 - <2 - <9 - <1(-7 3

124Sb 126Sb

<2 - )) <2 -

<2 -

<9 -

<9 -

<1(-7

<9(-8) 1275b <2-))

<3 - <4-7) <1(-7 <2(-7) 134Cs 1.11 0.05(-6) 3. 7 0.2 -6) 2.87 0.08(-6). 1.1 0.1(-6) 136Cs <1(-7) 7.1 0.6 -7) 3.5 0.3(-7) <8(-8) 137Cs 1.4 0.2(-6) 4.1 0.2 -6) 3.37 0.07(-6) 1.69 0.09(-6) 140Ba <8(-7) <9 - <4(-7 <4 -7) 140La <1(-6 <8 - <9 - <2 -7 )

141Ce <4 - <3- <2 - <2 -7 l 143Ce <6 - <4 - <4 - <1(-7 l 144Ce <2 - 1.2 0.4(-6) <7 - <7(-7I

4 i

TA'BLE A.24A MONITOR TANK ACTIVITIES Comparison Between INEL and OPPD Measurements Tank WD-22A Tank WD-22B INEL OPPD INEL OPPD 15:15; 8/24/76 15:10; 8/24/76 13:25; 8/24/76 13:20; 8/24/76 Nuclide (uCi/ml) (uC1/ml) (uci/ml) (uC1/ml) c Activation Nuclides:

51Cr <2(-6) 1.7i0.5(-6) <2(-6) <1(-6) 54Mn 5.5 0.6(-7) 8.3 1 0.5(-7) 8.5 i 0.7(-7) 7.9 0.6(-7) 59Fe <2(-7) <8(-8) <2(-7) <7(-8) 57Co <2(-7) <9 (-8) <2(-7) <1(-7) 58Co 3.420.1(-6) 4.6i0.1(-6) 5.0 t 0.1(-6) 5.6 0.1(-6) 60Co 8.2 i 2.6(-7) 6.0 0.5(-7) 7.4 1.7(-7) 6.6 0.6(-7) 65Zn <2(-7) ' <8(-8) <2(-7) <7(-8)

Iodine Nuclides:

131I 1.42 1 0.02(-5) 1.57 0.02(-5) 1.20 0.04(-5) 1.25 i 0.02(-5) 133I 5.9 1.9(-7) <9(-8) 1.8 t 1.2(-7) <2(-7)

Fission Products:

95Zr <2 -7) <8-8ll <3 - <1(-7 .

95Nb <1 - <4 -8J <1 - <2 -

99Mo <2 - <3 - ll <? - <3 -

10 3Ru <2 - <2 - <1 -

124Sb <2 - <8-)J

<6 - <2 -7) <6 -

134Cs 1.11 1 0.05(-6) 1.0320.07(-6) 3.7 0.2 - 3.2 0.1(-6 136Cs <1(.7) <6(-8) 7.1 0.6 - 5.1 i 0.6(-7 137Cs 1.4 i 0.2(-6) 1.56 0.07(-6) 4.1 0.2 - 4.4 0.1(-6 140Ba <8-7) c2 -7) <9(-7) <3(-7) 140La <1 -6) <3-8) <9(-7) <3(-8) 'l 14tCe <4 -7) <2-7) <3(-7) <2(-7) i

?

I l

l l

112

k 4

~

TABLE A.24A (cont'd)

MONITOR TANK ACTIVITIES Comparison Between INEL and OPPD Measurements , ,

, Tank WD-228' Tank WD-22B INEL OPPD INEL .0 PPD-08:17; 9/1/76 08:17; 9/1/76 07:30; 10/14/76 07:30; 10/14/76 Nuclide (uC1/ml) (uC1/ml) (uCi/ml) (LCi/ml)

Activation Nuclides:

SICr <2(-6) <2(-6) <6(-7) <8(-7) 54Mn 2.710.3(-7) 2.7 1 0.4(-7) 4.2i0.3(-7) 4.7 0.5(-7) 59Fe <1(-7) <6(-8) <2(-7) <8(-8) 57Co <9(-5) 1.4 1 0.2 - 8.6 1.9(-8) 3.E a 0.5 -

seCo a.8 0.4(-7) 9.7 i 0.6 - 4.7 4.9 1 0.1 -

60Co 5.9 i 0.4(-7) 2.9 0.3 - 1.05 0.1(-6))

0.04(-6 8.5 t 0.6 -

6 5Zn <1(-7) <6(-8) <2(-7) <9(-8)

Iodine Nuclides:

131I 2.2i0.2(-5) 2.34 0.02(-5 6.4 0.3(-7) 7.3 i o.7(-7) 133I 8.9 0.7(-7) 7.8 0.6(-7)) <4(-7) <1(-7)

Fission Products:

95Zr <1 - <S - 1 <6 - <9-8) ssNb <5 - <4 - h~ <3 - <5 -8) 99Mo <8 - <7 - <3 - <4 -

lo3Ru <1 - <8 - <2 - <1 - '

124Sb <9 - <6 - <2 - <6 - 1 134Cs 2.87 0.08( 6) 2.60 1 0.09(-6) 3.8 0.2(-) 3.8 0.1 -

136Cs 3.5 0.3(-7 2.8 0.4(-7) 2.5 t 0.6(- ) 2.1 0.6 -

137Cs 3.37 0.07( 6) 2.9 0.1(-6) 4.59 i 0.07 -6) 4.6 0.1 -

140Ba <4 - <3 - <7 - <3 -

, 140La <9 - <3 - <3 - <3 -

141Ce <2 - <5 - <7 - <3 -

l 113

< )

TABLE A.25 HOTEL TANK ACTIVITIES Power Operations - After Refueling Tank WD-15A Tank WD-15B 12:04; 2/17/77 13:15; 2/15/77 Dissolved Activity Suspended Solids Dissolved Activity Suspended Solids Nuclide (uC1/ml) (uti/ml) (uCi/ml) (uci/ml)

Activation Nuclides:

51Cr <5(-7) <8(-8) <1(-6) <1(-7) 54Mn 7.2 i 6.1(-8) 1.07 i 0.09(-7) 3.110.4(-7) 1.0 0.1 (-7)

R: :li:71  :!f:?!  :!f:ll  :!f:*l seCo 1.76 0.08(-6 5.3 3.2(-7) 1.7 2 0.1 -( 6) 5.9 0.2(-7) 60Co 2.110.4(-7)) 1.7 0.1 (-7) 3.110.5(-7) 2.2 i 0.1(-7) 6sZn <1 <g. <g . <g .

187W <3 - <2 - <5 - <2 -

239Np <3 - <6 - <8 - <8 -

Iodine Nuclides:

131I 2.5i0.6(-7) 1.7 0.5(-8) <1(-7) 3.5 0.7(-8) 133I <2(-7) <6(-8) <4(-7) <6(-8)

Fission Products:

95Zr <2 - <8 - <3 - <7-8) 95Nb <9 - <5- <1 - <4 -8) 99Mo <3- <3 - <1 - 44 2.4(-8) 10 3Ru - <2 - <3- <4 - <5 -

110 mag <2 - <3 - <4 - <5 -

124Sb <2 - <3 - <4 - <5 -

9. 0.2(-6) 1.7 0.1(-7) 134Cs 3.41 0.06(-5) 3.5i0.1(-7) 136Cs <9(-8) <5(-8) <1(-7) <4(-7) 137Cs 1.12 0.02(-5) 2.2 0.1(-7) 4.16 0.05(-5) 5.1 0.3(-7) 140Ba <8 - <1 - <2 - <2 -

14oLa <2 - <4 - <2 - <5 -

141Ce <2 - <6 - <2 - <9 - -

143Ce <1 - <2 - <3- <9 -

144Ce <6 - <3 - <9 - 7.6 3.8(-7)

Beta Only Nuclides: -

3H 4. i 1.(-6) ** 8. 1.(-6) **

14C 1.6 i 0.1 - 1.13 0.02( 6) '

95Fe 2.0 i 0.1 - 3.0 0.1(-6 63N i 2.0 0.2 - 2.0 0.1(-6 89Sr 3. 1.(- ) 1.3 t 0.9

P :ll:]  :!{::l~(-8

• • : Sanple not analyzed.for suspended solids.

114 i .. _ . _ _ _ . _ _ _ _ _ _ _ _ __.

TABLE A.26 MONITOR TANK ACTIVITIES Power Operations - After Refueling Tank WD-22A Tank WD-22B INEL OPPD INEL OPPD 08:49; 2/19/77 08:50; 2/19/77 09:30; 2/10/77 09:25; 2/10/77 Nuclide (vCi/ml) (vCi/ml) (uci/ml) (uCf/ml)

Activation Nuclides:

. 51Cr <7 - <7(-7) <1(-6) <2(-6) 5'+Mn <2 - 1.0 1 0.2(-7) 2.0 0.2(-7) 2.9 0.3(-7) 59Fe <2 - <4(-8) <2(-7) <6(-8) 57Co <5 - <8(-8) <6(-8) <2(-7) 58Co 1.45 0.06(-6) 1.61 i 0.07(-6) 2.44 0.07(-6 3.3 0.1(-6) 60C0 3.9 0. '(-7) 1.2 i 0.2(-7) 3.6 0.4(-7) ) 2.2 2 0.3(-7) 65Zn <2(-7) <5(-8) <2 ~,) <7(-8)*

187W <4(-7) <5 -7) 239Np <5(-7) <7 -7)

Iodine Nuclides:

1311 5.2 0.5(-7) 5.0 0.6(-7) 2.5 i 0.6(-7) 4.1 0.8(-7) 133I <7(-7) <8(-8) <2(-6) <? -7)

Fission Products:

95Zr <3 - <6 - <3 - '1-95Nb <2 - <3- <2 - <5 -

99Mo <4 - <2 - <6 - <3-to 3Ru <2 - <8 - <5 - <2 -* l 110 mag <2 - * <5 -

124Sb <2 - <6(-8) <5 - <6(-8) 134Cs 3.25 i 0.07(-6) 3.4 0.1(-6) '

7.0 0.1(-6) 8.710.2(-6) 136Cs <2(-7) <5(-8) <2(-7) <5(-8) 137Cs. 4. 0.1(-6) 4.3 i 0.1(-6) 8.5i0.3(-6) 1.04 0.02(-5) 140Ba <6 - <2 - <2(-6) <3 -

140La <4 - <3 - 9.2 4.0(-7) <3 -

141Ce <8 - <2 - <1 - <2 -

• <3 -

• 14 3Ce <2 -

144Ce <4 - * <5 -

Beta Only, Nuclides:

3H *** *** 5.36 i 0.02(-3) 14C 3.5 0.1(-7) 55Fe 5. i 1.(-7) 63Ni 4.3 0.1(-7) 89Sr <2 -

90;r <l -

91Y <2 -

a

• : Radionuclide not detennined by OPPD.

• • • Beta analysis not done on sample.

115

TABLE A.27 M)NITOR TANK WD-22B ACTIVITIES WITH VARYING TANK LEVEL Power Operations - After Refueling '

Prior to Release Start of Release Middle of Release End of Release Tank Level: 89 inches (5930 gal) Tank Level: 77 inches Tank Level: 37 inches INEL Tank Level: 21-inches OPPD (5130 gal) (2470 gal) (1400 gal) 08:41; 2/14/77 08:10; 2/14/77 11:15; 2/14/77 12:40; 2/14/77

~Nuclide (uCi/ml) 13:15; 2/14/77 (vCi/ml) (uCf/ml) (uCi/ml) (uCi/ml)

Activation Nuclides:

51Cr <7(-7) <2(-6) <1(-6) <2(-6) 5.5 2.3-7) 54Mn 2.7 0.3(-7) 2.3 i 0.3(-7) 1.6 0.4(-7) 59Fe ' <2(-7) <5(-8) 2.7 1 0.6(-7) 1.9 0.3-7)

<3(-7) <3(-7) 1.3 t 0.5 -7) 57Co~ <4(-8) <2(-7)

<9(-8) <6(-8) 58C0 2.62 0.08(-6) 3.6 <3(-6) 3.9 0.1(-6) 2.5t0.2(-6) 2.2 0.1(-6) 1.94 i 0.06(-6 SoCo 0.5(-7) 3.5 0.4(-7) 4.5 0.4(-7) 65Zn <2(-7) <7(-8) <3 - 3.9 i 0.9(-7)

<3(-7

'3.3

<2 -

• 0.5(-7) )'

187W <6(-7) *

<l -

2 39Np <5(-7) * <8(-7 <6 -

<7 - <1(-6 <8 -

M Iodine Nuclides:

131I 5.0 2 0.7(-7) <2(-7) 6.7 0.6(-7) 133I <2(-7) <2(-7) 4.7 i 0.8(-7) 5.8 0.7(-7)

<2(-6) <7(-7) <7(-7)

, Fission Products-95Zr <3(-7) <9 - 1.9 t 0.6(-7) <41 -7) <3-7) 95Nb <2(-7) <6 - <3 -7) <31,;-

99Mo 1.711.6(-8) <4 - <8-8) <2 -

l 10 3Ru <2-7) <2 - <7 -7)

<61 ,-

<31,-

<2-lJ 110 mag <2 -7) *

<7 -7) <3(-

<3 - )

124Sb <2 -7) <6(-8) <7-7) <3(- <3 - )

134Cs 4.1 0.1(-6) 8.6 0.2(-6) 4.5 0.2(-6) <3 - J 136Cs 4.5 0.2(-6) 4.3 i 0.1(-6)

<2(-7) <4(-8) <3(-7) 137Cs 5.2 0.1(-6) <3(-7) <2(-7) 1.10 i 0.02(-5)' 5.2 i 0.4(-6) 5.3 0.2(-6) 140Ba <8(-7) <4(-7) <3(-6) 5.4i0.2(-6) 140La <3 - <2{;-6) <2(-6)

<3(-8) <5 - <7i,-7 <51;-7) 141Ce <6 - <3(-7) <2 -

143Ce <2 - *

<3- <11l- <4{l-

• <41 -

<3 144Ce <3(-7) <7 - <5d- <2(1-0: Radionuclide not determined by OPPD.

.s * . - - - _ - _ -

)

TABLE A.28 EVAPORATOR PROCESSING WASTE HOLDUP TANK WD-4B Evaporator Evaporator Waste Holdup WD-4B Distillate Concentrate 13:34; 8/24/76 13:43; 8/25/76 13:43; 8/25/76 Nuclide (uC1/ml) (uC1/ml) (uCi/ml)

Activation Nuclides:

24Na 4.9 2.4(-7) <6(-8) 5.9.t1.2(-5) 51Cr <7(-4) <2(-6) <3(-2) 54Mn 8.0 i 1.2(-6) 3.1 0.5(-7) 7.7 0.5(-4) 59Fe <6(-7) <1(-7) <1(-4) 57Co <8(-7) <3(-4) <1(-4) seCo 6.2 1 0.1(-5) 1.04 1 0.04(-6) 1.68 i 0.01(-2) 60Co 6.1 2 0.4(-6) 5.2 0.3(-7) 1.57 0.07(-3) 6sZn <6 - <l(-7) <9(-5) 187W <4 - <3(-7) <7(-4) 2 39Np <5 - 5.8 1 2.6(-7) 7.9 5.6(-4)

Iodine Nuclides:

131I 5.06 1 0.07(-3) 2.8210.04(-5) 8.07 1 0.03(-1) 133I 6.9 i 0.2(-5) <9(-8) 8.6 z 0.5(-3) las! <2(-6) <2(-7) 4.6  ;.2(-4)

Fission Products:

95Zr <3(-6) <2-7) <4(-4) 95Nb <1(-6) <8 - <2(.4) 99Mo 9.210.7(-6) <3 - 9.1 1.1(-4) 103nu <6 - <8 - <4-3) 11oIPAg <6 - <8 - <4 -3) 124Sb <4 - <8(-8 <2 -5 126Sb <5 - 2.0 1 0.5(-7) <3 -3 127Sb 1.8 0.6(-6) <2(-7) <5(-4 134Cs 1.78 0.08(-3 1.32 1 0.06(-6) 3.37 i 0.01(-1) 13cCs 2.34 1 0.05(-4 <8(-8) 2.90 0.02(-2) 137CS l.86 1 0.06(-3 1.68 0.06(-6) 3.81 0.02(-3) 140Ba <2(-4) <3-) <1 -

<1 -

140La 1.1 1 0.2(-6) 143Ce <2(-6) <1

<5 -- )I <2 -

143Ce 2.610.8(-6) <6 - h 1.5 0.4(-3) 144Ce <6(-6) <5 - ll <9 (-4) l 117

. TABLE'A.29 EVAPORATOR. PROCESSING SPENT REGENERANT TANK WD-13A

. . Evaporator Evaporator Spent Regen WD-13A Distillate Concentrate 09:42; 8/26/76 09:28; 8/26/76 09:28; 8/26/76

• Nuclide .

(uC1/ml) (uCi/ml) (uCi/ml)

- Activation Nuclides: ,

, _ 24Na 2.9i0.'3(-6) <2(-7) 9.9 i 3.3(-5) stCr <6(-4) <4(-6) <6(-2) 1 54Mn 1.65i0.07(-5) 2.0 1.3(-7) 2.1 1.5(-4) 59Fe <l(-6) <3(-7) <2(-4) 57Co 1.2

• 0.06( 6) <4(-5)

, seCo 2.9f 0.01 -4 8.9 1.7(-7) 1.6 1.45 1i 1.6(-

0.02 -2) ) '

60Co 2.3 2 0.1(- ) ) 5.8 1.5(-7) 1.36 i 0.06 -3)

652n '<l - <3 - '2-

, 187W <7 <7 <7 l' 239Np ~ <4 - <3 - <4 -

Iodine Nuclides:

1311 6.8010.07(-3) 2.13 i 0.08(-5) 8.32 0.04(1) 1 133I 2.72 i 0.02(-4) <5(-7) 8.1 0.5(-3 135I <2(-6) 2.7 4.0(-6) 1.2 i 0.2(-3 Fission Products:

ssZr <3(-6) <4 - <5(-4) ssNb. <1(-6) <2 - <2(-4) 99Mo 1. 1 0.08(-5) <5 - 7.3 1.5(-4)

, 10 3Ru <2 - <5 - <3 -

110 mag <2 - <5 - <3 -

124Sb <2 - <5 - <7 -

126Sb <2 - <5 - <3 - i 1275b .4-

< <5 - <6 - -

134Cs 2.231.0.01-3) 2.7 0.4(-6) 3.48 0.02(-1) 136Cs 2.00 1 0.04 -4) <2(-7) 2.96 0.06(-2)

, 137Cs 2.57 .0.02-3) 2.7 i 0.5(-6) 3.86 0.03(-1) i

- 149Ba - <7(-5) <2 - <1 - -

140La 4.6 i 1.2(-7) <2 - <3 -

i 141Ce <1 - <8 - <3 -

l 143Ce <1 - <1 - <1 -

. 144Ce <6 - <1 - <1 -

118

TABLE A.30 ,

t

,EV APORATOR PR OCESSING SPENT REGENERANT TANK WD-138 1-Evaporator Evaporator Distillate Spent Regen WD-138 Concentrate-11:50; 8/31/76 ' 13:45; 8/31/76 13:40; 8/31/76

. Nuclide (uCi/ml) (uC1/ml) (vC1/ml)'

- Activation Nuclides:

', 24Na 3.4'i0.3(-5) <1(-7) .7(-5)

51Cr .<5(-4) <5(-6) <2(.-2) i 5Mn 1.20 i 0.06( 4)- 1.79 0.03(-6) 6.3 0.3(-4)-

59Fe 3.812.6(- <3(-7) <2(-4) 57Co 1.0 1 0.1(- <1(-4) 1.4 seCo 2.74 1 0.01 4) 6.99i0.07(-6) 3.45 0.5(-))

0.04 -3 60Co 6.74 0.06 -4) 1.07 i 0.06(-6) 2.39 i 0.03 -3) 1 - 65Zn 2.410.4I;- <2(-7) <1(-4) 187W 2.0 t 0.4I -

2.320.9(-6) 1.1 0.3(-2) j 2 39Np 3.6 1.8d- 5.7 1.4(-6) <2(-2)

Iodine Nuclides:-

l. 1311 1.65 0.01 - 1.87 0.02(-5) 2.98 i 0.03(-1) 133I' 1.26 1 0.01 - 1.1 i 0.2(-6) 1.62 0.02(-2) i 1851 3.04 1 0.02 - <4(-7) 3.3

• 0.2(-3)

Fission Products

1 95Zr <4(-6) <6 - <2(-4) 95Nb <2(-6) <3 - <7(-5) J l 99Mo 1.9 i 0.2(-5) <1 - 1.6 0.6(-3) 'l 10 3Ru <2(-5) <1(-7 <2(-3) i

110mg 8.611.5(-6). 2.6 1.8(-7) 7.5*5.0(-5) 124Sb <4 - <1 - <2 -

126Sb. <5 - <1 - <2 -

4 127sb <1 - <8 - <2 -

, 134Cs- 1.49 2 0.01 - 4.05t0.05(-6) 1.87 2 0.01 -

!- 136Cs 3.95 i 0.06 - 4.5 1.0(-7) 1.35 t 0.01 -

137C$. 1.49 i 0.01 - 4.7620.04(-6) 2.09 0.01 140Ba <2 - <5 - 1 3.0 0.5(-3 14oLa <2 - <2-h 3.9 1.5(-3

- 141Ce <5 - <2 h <2(-4) 143Ce

<1 - 5.2 0.6(-2)

!~ 144Ce <2 - <1-)J

<9 - <8(-4) i 4

119 4 .

r --r- --

e -- -- - - - - - - - --- -- - -- + * - - , - - = - - - -

I TABLE A.31 SCRAPINGS FROM RADWASTE EVAPORATOR BUNDLE Dut#tej Refueling

. Inlet - Outlet End U-Tube End 11:30; 10/13/76 11:30; 10/13/76 Nuclide (uC1) (uCi) ,

Activation Nuclides:

24Na <6(-2) <3(-2) 51Cr 5.5 1.3(-2) 1.4 i 0.7(-2) 5"Mn 1.3310.02(0) 1.64 0.03()

59Fe 1.4 2 0.1(-2) 3.0 i 0.6(-2

$7Co 3.86 i 0.06 -2) 3.98 0.08(2) saco 4.93 0.06 ) 5.2 i 0.1(0) 60Co 1.36 1 0.01 ) 1.21 1 0.02(0) 65Zn 1.7 i 0.2(-2 2.3 1 0.7(-2) 187W <4(.] ) <7(.]

239Np <2(-2) <6(-3 Iodine Nuclides:

131I 2.3 0.5(-2) 3.1 0.5(-3) 133I <2(-1) 3.3 i 0.9(-3) 135I <2(-1) <9(-2)

Fission Products:

95Zr 5.6 1.7(-3) 7.2 2.2(-3) 9sNb 1.0 0.2(-2) 8.7 1.6(-3) 99Mo <2(-3) <2(-3) 103Ru <2(-1) <4(-2) 110 mag 1.7 0.3(-2) 2.3 0.3(-2) 124Sb 2.5 0.6(-3) 3.5 0.8(-3) -

126Sb <2(-1) <4(-2) 127Sb <3(-1) <6(-1) 134Cs 3.04 0.09(0) 1.50 i 0.03(0) 136Cs 2.6 i 0.2(-2) 7.0 1.9(-3) -

137Cs 3.71 i 0.05(0) 1.87 0.02(0) 140Ba <6(-1) <2(-1) 140La 1.8 0.4(-3) 2.1 0.5(-3) 141Ce <3(-3 <3(-3) 143Ce <7(-3 6.8 4.2(-3) 144Ce <2(-2 <2(-2) 120

i

.l TABLE A.32 EVAPORATOR FUNCTION TEST

~

Feed Samp1'es: Spent Regenerant Tank WD-13A s

Sample 144** First Pro:ess Sample Prior to Test Tank I.evel: 70" (4670 gal)

Tank Level: 80" (5330 gal) 10:3G; 2/18/77 07:30; 2/18/77 Dissolved Suspended Total Activity Activity Solids-(uti/ml)

Nuclide (uci/ml) (uci/ml)

Activation Nuclides

24Na 1.6 1 0.2(- 1.3

• 0.1(-5)

SICr 7.2 i 0.9(- <2(-4) 7.68 0.09(5) 54Mn 3.19 0.07 5) 2.08 0.06(-5) 1.29 0.03( 5) 59Fe 7.4 0.7(- <2(-6) 9.2

• 0.2(-

57C0 2.2 0.7(- 2.7 6.9 0.3(-

seCo - 5.8220.07 4 3.67 1.1(-6) 0.03(-4 ) 2.25 0.04 4) i, 60Co 2.8 0.2(-5)) 9.5i0.8(-6) 1.91 0.02 -5 i 65Zn <3 - <2(-6) 3.8 i 0.7(-7) )

f 18 % <9 - <3(-5) <3(-5) 239Np <2 - 4.1 i 1.5(-5) <3(-6)

Iodine Nuclides:

131I 1.11 0.02(-3) 8.89 0.15(-4) 1.05 0.03(-5) 133I 2.19 0.07(-4) 1.73 i 0.16(-4)

• 1351- .<5(-5) <3(-6)

Fission Products:

95Zr 4.0*1.0(-6) <2 - 3.3 0.4 -

95Nb <3f- <8 - 4.6 0.5 -

1.0 -

99Mo <2E- <2 - 2.9 108Ru <7d- .

<6 - 9.6' 1.0 -

,.t-110'%g 5.2 0.7(-6 <6- 3.7 0.1 -

124Sb 6.5 i 4.2(-7 <5 - 2.0 0.3 -

134Cs ~1.47 0.02(3) 1.20t0.01(-3) 2.09 0.06(-5 136Cs <3(-5) 3.6 0.4(-7) )

137Cs 1.72 i 0.02(-3) 2.1 1.44

• 0.02(-3 1.6(-6) ) 2.78 0.05 5) 140Ba <3 - <2(-4 1.9 1 2.1 -

140La <9 - - 1.6 1 0.3 -

141Ce- .3-< <7(J

<4 6 5 5.1 0.3 -

143Ce <5 - <4 -5 <9(-7)

.144Ce - <1 - (1

<2,-5 8.2 2.4(-7) f1

.. 121

- , . -- .  : ., a, -- -- . . . . - - -- .-

TABLEA.32(cont'd)

EVAPORATOR FUNCTION TEST Feed Samples: Spent Regenerant Tank WD-13A Third Process Sample Second Process Samp(le60" 4000 gal)

Tank Level: Tank Levci: 50" (3330 gal) 12:50; 2/18/77 16:00; 2/18/77 Dissolved Activity Suspended Solids Dissolved Acti.vity Suspended Solids

• Nuclide (uC1/ml) (uCf/ml) (uCi/ml) (uci/ml)

Activation Nuclides: ,

1.3 0.1(-5)

• 24Na <3(-6) <7(-6)

_51Cr <2(-4) 6.01 i 0.07 -5) <2(-4) 6.3 54Mn 2.03t0.05(-5) 9.7 1.96 0.07(-5) 1.01 0.2(-5))

0.02(-5 59Fe <3(-6) 7.47 i 0.3(-))

0.08 6 <4(-6) 6.8 0.4(-6 57Co 2.7 0.7(-6) 4.9 0.2(- <2 (-6) 5.6 i 0.4(-7 58Co 3.90 2 0.06(-4) 1.66 2 0.03 4) 4.06 0.08(-4) 1.75 i 0.04( 4) 60C0 1.1t0.2(-5) 1.43

• 0.02 -5) 1.06 0.07(-5) 1.47 0.03(-5) 65Zn <2(-6) 1.7 0.6(-7) <3(-6) 4.0 i 1.3(-7) 187W 1.8 2 0.4 -5 <6(-6) <6(-5) <3(-5) 239Np 5.0 i 4.3 -6 1.4i0.7(-6) 6.7 i 3.3(-6) <2(-6)

I:: dine Nuclides:

131I 1.12 0.01(-3) 8.7 i 0.1(-6) 1.08 0.02(-3) 1.00 i 0.02(-6) 133I 1.9i0.1(-4) <8(-7) 1.67 t 0.04(-4) 1.9 0.4(-6) 135I <1(-5) <8(-5)

Fission Products:

95Zr <3 - 2.3 0.3 - <3 - 2.5 i 0.2(-6 95Nb <2 - 2.7

• 0.3 - <2 - 2.8 0.3(-6 99Mo <1 - 2.3 0.7 - <2 - 2.6 0.4 -

103Ru <3 - 7.6 1 0.8 - <3 - 7.6 i 1.5 -

110 mag 2.010.4(-6) 3.1 0.2 - 6.1 i 1.9(-6) 3.0 0.1 -

124Sb <3(-5) 1.6 1 0.2 - <3(-5) 1.6 t 0.4 -

134Cs 1.30

• 0.02(-3 1.39 1 0.03 5) 1.36 0.02(-3) 1.90 0.05 5) 13sCs 4.2i4.7(-7)) <2(-5) <7(-6) -

137Cs 1.52 2 0.02(-3) 4.0 1.82 11 0.02(-5 0.9(-7) ) 1.62 0.02(-3)' 2.47 0.03(-5) 140Ba <1 - <2(-6) <2 - <2(-6) 140La <9 - <6(-6) <9 - <8(-7) 141Ce <2 - 5.4.11.7(-8) <4 - 3.6 1 0.5(-7) -

143Ce <2 - <4(-7) <5 - <4(-7) 144Ce <9 - 4.9 1.4(-7) <2 - 9.0 1 2.6(-7)

• Sample not counted soon enough for these radionuclides.

• • Sample not filtered.

I p 122 L

TABLE A.33 EVAPORATOR FUNCTION TEST Evaporator Distillate Samples (Feed: Spent Regenerant Tank WD-13A)

First Process Sample 10:20; 2/18/77 Dissolved Activity Suspended Solids Nuclide (uCf/ml) (uci/ml)

Activation Nuclides:

24Na <8(-9) <1(-8) 51Cr <7(-8) <2(-8) 54Mn' 2.2 0.5(-8) 9.0 1.1(-9) 59Fe <3(-8) <1(-8) s7Co 1.0 i 0.2 'l <1(-9) 58C0 7.1 10.1 - l,l l 7.7 0.3(-8) 60Co 5.3 0.4 3.4 2 0.2(-8) 65Zn <3(-8) <1 -

187W <](.7) <2 239Np 1.3t1.1(-8) <1 -

Iodine Nuclides:

131I 1.28 0.06(-7) 5.3 0.9(-9) 133I 1.7i1.0(-8) <6(-9) 135I <7(-8) <7(-8)

Fission Products:

95Zr <5 - <2-;l 9sNb <3-99Mo <5 - <6

<1 - /l 103Ru <3 - l lion %g 124Sb

<3 -

<3 -

<6-l

<6

<6

- l

,1 134Cs 3.710.1(-7) 4.3 0.3(-8) 136Cs <3(-8) <6(-9) ,

137Cs 4.8 0.1(-7) 3.1 i 3.0(-8) i 140Ba

<8 -} 1.1 0.4(-7) l 140La <4 - <7 -  !

141Ce <8 - <2 - l 143Ce <3 - <4 -

144Ce <4 - <9 -

I 123

TABLE A.33 (cont'd)

EVAPORATOR FUNCTION TEST Evaporator Distillate Samples (Feed: SpentRegenerantTankWD-13A) i Second Process Sample Third Process Sample 13:40; 2/18/77 15:50; 2/18/77 Dissolved Activity Suspended Solids Dissolved Activity Suspended Solids -

Nuclide (uct/ml) (uC1/ml) (uC1/ml) (uci/ml)

' Activation Nuclides:

2*Na * * *

• 51Cr <1(-7) <2 - <1(-7) <2(-8) 54Mn 1.3i0.2(-8) <4 - 1.4*0.3(-8) 2.4 0.6(-9) 59Fe <2(-8) <6 - <2(-8) <6(-9) s7Co 6.5 i 1.2 - 1 <9-10) 3.2i1.7(-9) <5(-10) seCo 6.6

• 0.2(-8) 3.89 t 0.06(-7) 5.4 0.3(-8) 60Co 4.7*0.1-lh 3.9

• 0. 2 - l 1.6 1 0.1(-8) 3.6 i 0.3(-8) 1.54 0.09(-8) 652n <2 - 1 <6(-8) <1 - <5 -

187W <8 - h <2(-8) <8 - <3 -

239Np <9-ll 1.0 1.0(-8) <9 - <8 -

Iodine Nuclides:

]1 1.8410.05(-7) 7.8 i 0.9(-9) 1.65 1 0.08(-7) 2.2 i ,0.7(-9) 135I * * *

• Fission Products:

95Zr <4 <gl'.9) <4 - 2.2 0.7(-9) 95Nb <2 - )

<4Il-9) <2 - <7 -9) 99Mo 103Ru

<3 -

<1 -

l <91 ,- ) <5 - <6 - )

l <41 ,- <2 - <3 -

110 mag <1 - i <41 -

<2 - <2 -

124Sb <1 - ) <41!- <2 - <2 -

13kCs 3.70 i 0.06(-7) 4.1 i 0.2(-8) 2.93 0.09(-7) 1.76

• 0.09(-8) 138Cs <2(-8) <4(-9) <2(-8) <7(-9) -

137Cs 5.00 1 0.07(-7) 6.1 i 0.2(-8) 3.8 0.1(-7) 1.1 1.1(-8) 140Ba <61'- m2(-8) <1(-7 <9 -

! 140La - <6f <7(-9) <6(-8 <6 -

141Ce l-

<61 - 4.8 i 5.4(-10) <9 - <1 - -'

143Ce <31 ,- <4(-9) <3 - <3 -

144Ce <3L- <7(-9) <5 - <5 -

• : Sample not counted soon enough for these radionuclides.

124

TABLE A.34 EVAPORATOR FUNCTION TEST Evaporator Concentrate Samples 7

Before Test Start First Pro ass Sample 07:30; 2/18/77 10:20; 2/18/77 Dissolved Activity Suspended Solids Dissolved Activity suspended Solids Nuclide (uC1/ml) (uct/ml) (uC1/ml) (uCi/ml)

Activation Nuclides:

24Na 3.7 1.5(-5) <7(-5) 8.0 12.2(-5) <5(-5) 51Cr <5(-3) <1(-4) <5(-3) 2.1 1 0.6(-4) 54Mn 1.410.2(-5) 1.08 0.01(-3) 1.4 ' i 0.2 (-4) 1.61 1 0.02( 3) 59Fe <8(-5) 4.6 1.3(-5 <1(-4) 7.3 1.3(-5 57Co <2(-4) 6.9 i 0.5(-5 <2(-4) 9.2*i0.6(-5 seCo 3.6 0.1 (-3) 1.05 0.02(2) 3.87 0.07(-3) 1.62i0.04(-2) 60Co 2.0 1 0.4 (-4) 8.4 2 0.2(-4) 1.5 0.1 (-4) 1.27 i 0.02(-3) 652n <7 - 5.1 1.4(-5) <1(-4) <1 -

187W <8 - <8(-4) <5( 4) <4 -

239Np <3 - <7(-5) 2.5 2.6 (-4) <2 -

Icdine Nuclides:

131I 2.6220.03(-2) 2.9 i 0.6(-6) 2.54i0.05(-2) 6.1 1.5(-5) 133I 2.6. t 0.1 (-3) <7(-5) 2.5 0.2 (-3) <1(-4) last <c(-4) <2(-4) <3(-4) <3(-3)

Fission Products:

95Zr <4I;- 2.1 0.6(-5) i'-4) 4.4 1.0(-5) 95Nb < 31,- 3.5 0.4(-5) d -4 <1 -

99Mo <21'- <6(-6) <2-) I <5 -

103Ru <31 <7(-5) <3 - h <1 -

9.5 i 0.7(-5) 1.3 0.1(-4) 110 mag l-

<3l -

<3-)J

<3 - <1(-4) 124Sb <31,- <6(-5) 134Cs 6.6 i 0.1 (-2) 1.32 0.02(-5) 6.50

• 0.05(-2) 1.8 1 0.3(-3) 136Cs <3(-4) <2(-4)- <2(-4) <2(-4) 137Cs 8.0 0.2 (-2) 1.66 i 0.03(-5) 7.7 0.2(-2) 2.45 i 0.04(-3) 140Ba <2(- <4 - 7.4 i 2.3 (-4) <4 -

14oLa <5 - <2f- <2 -

<2I'-- <3f- <1 -

- 141Ce <41 <2 -

143Ce <10,- 1.4 1 0.8(-5) <1ll- <1 -

144Ce <21,-} <7(-5) <2 I,- <5 -

r 125

- _~ - - .- - . .

l

-TABLEA.34(cont'd)

EVAPORATOR FUNCTION TEST Evaporator Concentrate Samples Second Process Sample Third Process Sagle ,

13:36; 2/18/77 15:50; 2/18/77 '

Dissolved Activity Suspended Solids Dissolved Activity Suspended Solids Nuclide- '(uCi/ml) (uC1/ml) (uCi/ml) (uCi/ml)

~

Activation Nuclides:

124Na <1(-4) <5(-5) -

<2(-4) <2(-4) 51Cr <5(-3) 2.8 t 0.3(- .

<1(-2) 3.2 0.4(-)

54Mn 1.4- 0.2(-4) 1.25 t 0.03 3) 2.6 1 0.2 (-4) 1.6110.03-3)-

59Fe <1(-4) .

8.7 1 0.9(- <2(-4) 1.20 i 0.08 4)

57Co 2.1 5,7 . -

7.3 1 0.3(- <3(-3) 9.8 0.3(-

58Co 4.C' i 0 2 -

1.25 t 0.03 2) 4.7 2 0.1 (-3) 1.59 0.04(2)  ;

! 60Co 1.6 t 0. 4 -

1.02 1 0.02 -3) 2.3 i 0.7 (-4) 1.25 i 0.03(-3) '

. 65Zn -<1

<4(-5) - <1 - <2 -

l 187w <2 - 4.6 i 2.2(-4) <1 - <3 -

239Np <4 - <1(-4) <1 - <1 -

Iodine Nuclides

, 131I 2.8020.06(-2) 4.9 t 0.5(-5) 2.9710.08(-2) 7.5 0.5(-5) 133I 2.9 i 0.1 (-3) <1(-4) 2.3 0.2 (-3) <1(-4)

13sI <1(-3) <3(-3)

• Fission Products:

95Zr <1 -

5.4 1 0.9(-5) <5 - 3.5 0.6(-5).

l 95Nb <4(,- )[

1 6.8 t 0.7(-5) '3-

< 6.2 1 0.9(-5) 99Mo. - l- <5(-6) <2 -

~ <4I <4(-6) 10 3Ru <4l- h <4(-5) <2 - <5(-5) i 110 mag <3(- 8.7 0.6(-5) <2 - 1.1 0.1(-4

'124Sb <3(-)h <3(-5) 2.8 1.4(-5) 1.7 0.3(-5 134Cs- 7.05 1 0.05(-2) 1.51i0.02(-3) 7.2 0.1 (-2) 1.94 0.06( 3) ' ,

13sCs <4(-4) <5(-4) <3(-4) <1(-3) 4

'137Cs 8.2 *0.2~(-2) 5.6 7.0(-4) 8.4 i 0.1 (-2) 6.8i10.9(-4) 140Ba <2 - .'2(-4)

< <1 - <2(-4) 140La '<?- 2.5 1.2(-5) <5 - 1.5 0.3(-5)

~'

' - 141Ce <1 - . <1 - <4 - 8.0 1 3.0(-6) 1143Ce <E - <3 - <3 - <3(-5) 144Ce -<4. <4 - <2 - <4(-5)

• . Sample not counted'soon enough for these radionuclides. -

i-f

.126

a . ' o TABLE A.35 VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #1 (u Ci/sec)

Sample Period 9/3 - 9/9/76 9/9 - 9/15/76 9/15 - 9/23/76 9/23 - 9/30/76[A] 9/30 - 10/7/76 1311 8.06 0.07(-3) 8.1 0.2(-3) 2.19 0.09(-4) 1.22 0.08(-4) 1.72 0.04(-3)-

134Cs <2.2(-6) <9.2(-6) <8.6(-6) 8.6 i0.8(-7) <7.9(-6) 136Cs <5.4(-6). <2.6(-5) <1.8(-6) <3.4(-7) <1.1(-5)

-137Cs 3.2 0.7(-5) 2.6 1 0.7(-5) 2.7 0.6(-5) 1.1 0.2(-6) 1.4 0.5(-5) 3H - 3.0 i 1.0(-3)- 7.2t0.8(-3) 2.0 1.0(-3) 5.0 i 1.0(-3)~ '

9.510.9(-3)[8]

14C 6.5 0.8(-4) 2.4 0.1(-3) 7.Ri0.d(-A)fc]

51Cr <1.1 -4) <1.4(-5) <4.1 -5) <2.4(-6) 5.3 1 2.9(-5) 54Mn <9.7-6) <3(-6) <2.7-6) <7.5(-7 <6.6(-6) 59Fe <3.9 -6) <1.2(-5) <4.7 -6) <1.2(-4 <7.9(-6) g 57Co 3.7 i 2.3(-5) 1.2 0.7(-5) 1.2 1 0.7(-6) <3.9(-7 <2.2(-5) w ssCo <3.4(-6)' <3.4(-6) <6.0(-6) 2.0 .

• 1.0(-6) <4.9 -6) 60C0 8.3 3.0(-6) <1.4(-5) <4.2(-6) 9.0- 2.0(-7) <1.1-5) ssZn <8.0(-6) <8.9(-6) <3.0(-6) <1.9(-6) <5.6 -6) 95Zr <7.1(-6) <3.5(-6) <5.6(-6) <2.3 - <7.3(-6 95Nb <2.2 -6) <1.5 - <6.3(-6) <2.2 - <6.2(-6 lo3Ru <1.8 -6) <6.3 - <1.8(-6) <8.7 - <4.1(-6 106Ru <1.5-5) <4.6 - <1.6-5) <1.3(-5) <3.0(-5 110 mag <6.4 -6) <2.6 - <3.9 -6) <1.4(-6) <3.9(-6) 124Sb <7.4(-6) <6.2(-6) <7.3-6) 3.0 1 0.3(-5) 6.3 i 3.5(-6) 12sSb <9.0f-6) <1.8(-5) <9.4 -6) 4.0 1 2.0(-7) <1.4(-5) 160Ba. <9.6(-6) 6.2 2.1(-5) <9.4(-6) <1.2 -6) <1.4(-5) 140La 4.6 i 0.3(-4) <1.6 -4) <3.0(-5) <4.0 -7) <1.3(-3) 141Ce 2.8 i 1.7(-5) <2.4-5) 1.0 0.7(-6) <3.4-7) <1.5(-5) 152Eu <1.9(-5) <1.4 -5) <1.2(-5) <1.1 -6) <2.0(-5) 154Eu <6.1-f) <8.4(-6) <1.2(-6) <1.3(-5)

IssEu <1(-5)

<7.5(-4 ) <1.0(-4) 1.1 1 0.4(-4) <3.1(-4; <1.4(-4)

[A] 60,000 second count

[B] Sample from 9/9 - 9/23/76

[C] -Sample from 9/23 - 10/7/76 l

- aW TABLEA.35(Coni;.)

VENTILATION AIRBORNE ACTIVITIES ,

, SAMPLE STATION #1 (uCi/sec)

Sample PGriod 10/7 10/14/76 10/14 - 10/22/76 10/22 - 10/28/76 10/28 - 11/4/76 11/4 - 11/10/76 131I- 5.6 0.2(-4) 3.0 1 0.1(-4) 4.6 0.6(-4) 7.6 0,7(-6) 3.2

• 0.6(-5) 134Cs <9.1(-6) <1.7(-5) <1.1(-5) 9.2 1 3.9(-6) <6.6(-6) 136CS <7.8(-6) <7.8(-6) <1.0(-5) <1.4(-6) <9.2(-6) 137Cs 5.4 1 1. 8(-5) 2.0 .1 0.5(-5) 5.1 t 1.3(-5) 6.0 1 1.0(-5) 5.4 2 1.5(-5) 3H 5.0 i 1.0(-3) 9.8 1 0.9(-3) 3.8 i 0.3(-2) 7.2 1 0 8(-3) 6.4 0.1(-2) 14C 1.7 0.1(-3)[D] 2.97 0.02(-2)[E] 1.11

• 0.04(-2) 51C r 3.0 2.0(-5) <3.5(-5) <1.5(-5) <7.6 - <1.0 -5) 54Mn <3.9(-6 <9.6I -7) <2.5(-6) <2.3 - <5.4 -6) g 59Fe <1.1(-5 <6.2i <2.0 - <1.1 - <1.1 -5) co 57Co' 58C0

<4.1 -

<5.6 -

<1.

<5.8(-

7([- <5.3 -

<3.7 -

<4.6 -

<1.1 -

5.0 2.0(-5)

<9.2(-6) 60C0 <5.3 - <1.2 - <7.3 - <2.5(-5) 652n <2.8 - <9.4 - i )l

<5.5 <2.0(-5) <1.0 -5 95Zr <2.2 - <2.0 <4.0(-6) tl.6 -

<9.0{-i

<8.61 - S

'l 95Nb <1.2(-6 <3.4 - l <4.5(-6) <2.2 - <5.2fil 103Ru <3.1 -

<1.7 -

<1.4 'l 9.1 1 4.6(-6) <3.7 - <1.2fl 106Ru <4.0 - 1 <3.6(-5) <1.6 -

110 mag 124Sb

<3.5 -

<7.8(-6)

<2.3-!) <9.5(-6) <8.9 - <2.8Yl

<1.31 - ll l

<6.4(-6) <9.8(-6) <1.2-5) <2.0 -

125Sb <6.9(-6) <4.9(-6) <2.5(-5) 1.9 1 0.8(-5) <4.2 -

140Ba <2.0 '5 <4.3(-6) <1.6(-5) <1.6(-5) <4.1 -

l'0La <9.9 -5 <8.5(-4) <6.6(-5) <1.9(-4) <2.5(-3) 141Ce <3.8 -5) <1.2(-5) <4.6(-5) <6.3(-5) <8.1(-5) 152Eu <1.0(-5) <2.1(-5) 2.8 1.4(-5) 1.3

• 0.8(-5) 2.6 i 1.7(-5) 154Eu <1.1(-5) <6.1(-6) <2.0(-5) <9.7(-6) <2.9(-5)

IssEu <3.2(-4) <2.1(-4) <6.9(-4) <1.6(-4) <1.2(-3)

[D] Sagle from 10/7 - 10/22/76

[E] Sagle from 10/22 - 11/4/76

s

~

TABLE A.35 (Cont.)

-VENTILATION AIRBORNE ACTIVITIES' .

SAMPLE-STATION #1 -(pCi/sec) w iSample

Period- 11/10'- 11/23/76 :11/23 - 11/30/76 12/1 - 12/15/76[g 12/15 - 1/6/77 1/6 - 1/20/77 131I- 2.8 0.5(-5) 1.1 . 0.4(-5) <2.8(-5) 1.33 2 0.05(-6) 1.96 t 0.07(-4) 13"Cs <6.3(-6) <1.8(-5)- <1.3-5) 9.0 0.6(-6) 9.0 -

-13sCs-0.8(-6)

<7.0(-6) <1.3(-5) <4.7-5) <2.3(-7) <3.9(-6)

.137Cs l6.0 : 4.0(-6) 5 0 t 1.0(-5) -

<4.2-5) 1. 0 - 0.1(-5) '1.1 1 0.1(-5) 3H 9.3 . 0.2(-3) 1.8 0.1(-3) 4.4 0.1(-3) 1.6 0.2(-2) 4.7. t 0.2(-2).

1"C 2.0 0.1(-2) 3.6 m 0.3(-3) 6.8

• 0.3(-3) 2.4 i 0.1(-3). .7.1 1 0.2(-3)

+ .

51Cr' s4

<6.2 - <6.1 - <4.8 -6). <5.2(-6)

Mn '<1.6 - <7.91-)i

<4.41- <9.5 - <4.4 - <2.0(-7 59Fe~

57 C0

<8.1 - <3.2I- l} <2.6 - "

<3.6 - .<6.7'~-6 2.0 i 1.0 (-5) . <3.11,-5ll 5.0 i 3.0(-5)' <2.9 - <1.1k-7 ssCo 80

.5.0 i .0(-5) <2.1l -51 5.0 2.0(-5) 2.3 1 0.5(-6) 4.0 1 3.0(e7)

Co <3.4 - <1.41 l-5l l

<9.7(-6) <6.3(-7) <2.2(-7 65Zn <7.4 - <2.1 d- 1 <9.4I;-6J <2.5(-7) 95Zr <2.6 <2.3 d55!) <5.9ll-7

<4.1(-51 <4.5(-7) <5.6(-7 "95Nb <1. 61 ;-

<3.3f-6) ~ <8.9f-5I <1.8I;-6) to3Ru ~ <6.51 <2.9I'-7)

<1. 0l!-5'1 <1.4d-Sh <8.4(-7): <1.2l -6 106Ru '

110 mag.

<3.61,- <4.8f-5 l <2.2f-5 l <5.8f-6 l ' <3.41 l-)

<3.2ll- <9.8d-6ll <4.4ll-5ll <1.1ll-7h <2.8d6l 7h

124Sb < <4.21 -5

<1.8f-5) <7.5l -6 1.1

• 0.1(-5) 1.2f- - )i <1.51 ;-5))

-12sSb <2.5 <4.4f-5) <2.3 Ll-6)) <1.6(-6) 1408a <1.8 - 9 <1.6f <2.3-6)

<3.0(-6)

<1.2{1'-4) <1.5d,-4))

1"0La <1.2 - h <4.8 -3) -5 <2.3t-6) <1.2 -5)

. 1" Ice <5.1 - 1 <6.9( -

<5.6ll- <6.2(-7) <3.1 -7)

- 152Eu <2.1 - I <3.9I <1.11 5'.0 1 4.0(-7) <2.4 -6) 15"Eu <8.6-) <2.8d- <2.3d-- <6.2(-7) <6.0-7)

[F] Total Iodine Sampler was only operative between.

12/1 -'12/3/76. 1"C and 3H sampler was operative

'12/1 - 12/15/76..

~

TABLEA.35(Cont.)

VENTILATION AIRBORNE ACTIVITIES; 4

SAMPLE STATION #1 (uCi/sec).

' Sample Period 1 1/20 - 2/3/77- 2/3 - 2/10/77 -2/10-2/17/7L t

131I 2.75 i O.08(-4) _1.76 0.07(-4) 1.9 0.1(-4) _

'13"Cs <1.4(-6). <2.4(-6) <8.3(-6)

-13sCs <<1.6(-7) <1.6(-6) <1.5(-6)

, 137Cs. ~< 1.4(-6). 2.6 0.7(-6) <1.7(-6) 1.8 i 0.1(-2) 3H 'G' 14C 4.3 1 0.2(-3) 'G.'

, [G' G' s1Cr- - < 1. 71'- <2.4(-5) <5.2f-6) 54Mn <4.0f- <1.6(-6) <6.81,-7)

SSFe <6. 5 ll- 5.0 2.0(-6) <3.01 5)

<1. 4I ,-

57Co . 1.71

< <7.2(-7)

.y 4 _seCo <8.4d-7) . <1.6(-6) <2.4 I- ll 60Co

<2.6ll-7) 6.0 1 2.0(-6) <1.11!- lll 65Zn- <4.5I,-6) <8.0-6) <2.11'-5 I .

ssZr <6.6 -7)~ <3.2 -6) <1.3I-5ll ssNb <1.8-6). <1.6 - <7.9I'l

, 103Ru <1.3 -7) <1.6 - l i

losRu <2.4-6) <1.6 - <1. 0(f

llomAg 43.0 -7) <1.6 - <1.2(-3

<1. - ll l

124Sb 9.0 i 4.0(-7) <1.6(-6 <1.4 -

12sSb <7.4 -7) <4.0(-6 <2.4 -

140Ba <6.6-7) <5.6 - <2.6 -

140La <6.2 -6) <1.6 - <1.7 -

141Ce k2.4(-7) <1.6 - <7.9 -

t 152Eu <5.7(-7) <7.2 - <4.9 -

154Eu <7.1(-7) <1.6 -5) <3.3 -

[G] . SAMPLE STATION #1 14C 3H Shutdown due to electrical circuit overload.

7- ,.

TABLE A.36 VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #2 (uCi/sec)

Sample Period 9/3 - 9/9/76 9/9 - 9/15/76 9/15 - 9/18/76 9/23 - 9/30/76 9/30 - 10/7/76 131I 6.~13 1 0.06(-4)- 1.92 0.06(-4) 6.0 L 0.2(-4) 8.3. 0.4(-5) 1.43 0.01(-3) 134Cs <2.7(-6) ' <1.5(-5) <2.6(-5) <2.1(-6) <7.1(-6) 136Cs <4.4(-4) <2.8(-6) <1.5(-5) <2.7(-6) <3.1(-6) 137Cs 8.0 3.0(-6) 4.8 2.4(-6) 2.0 i 1.0(-5) <9.0(-6) 6.8 3.9(-6) 3H <1.0(-3) <1.0(-3) 2.0 0.4(-3) 14C 6.0 1.41 1 0.01(-1 1.0(-3) ) 1.9 0.1(-3)[A] 2.0 1.4a i 0.4(-3) 0.09(-3 )[B]

stCr <6.6(-5) <1.3' ") <4.6(-5 2.3 1.1(-5) <2.8 -5)

L 54Mn <1.1(-5) <2.8 - <1.0 - <7.7(-7) <2.1 -

E! '59Fe <3.7(-5) <6.9 - <7.5 - <4.4 -6) <3.6 -

57C0 1.5 0.7(-6) <1.2 - <2.0 - <9.1-7) <4.6 -

58Co <1.4 -5) <4.0 - <5.5 - <3.0 -6) <3.3 -

60Co <6.6 -6) <5.3 - <2.0 - <2.9-6) <5.0 -

65Zn <1.2 - <2.5 - <4.6(-6 <4.0(-6) <1.9 -

95Zr <1.0 - 6.0 2.0(-6) <3.7(-6) <4.6(-6 <7.2 -

95Nb <2.4 - <2.0(-6 <1.4 - <3.7 -6 <2.6 -6) 103Ru <1.7 - <4.7 - <4.0 - <2.5 -6 <3.3 -6) lo6Ru <8.4 - <2.1 - <3.8 - <2.7 -5) <9.2 -6) 110 mag <5.8 - <7.2 - <1.2 - <3.8 -6) <1.7 -6)

, 124Sb <1.2 - <2.2(-6) <1.4(-5) <2.3(-6) <6.2-6) 12sSb <2.5 - <1.6(-5) <1.4(-5) <5.0(-6) <1.6 -6) 140Ba <6.2(-3) <1. 0(-5 ) <2.3(-5) <6.5(-6) <1.7 -5) 140La <3.3(-6) <4.4(-5) <4.7(-3) <4.3(-5) <1.1 -3) ,

141Ce <9.0(-6) <4.7(-6) 4'. 5 2.6(-6) 2.8 1.6(-6) <2.6(-6) 152Eu <1.4(-5) <1.3(-5) <6.2(-6)- <8.7(-6) <5.8(-6) 154Eu <7.2(-6) <4.5(-6) <2.3(-5) <1.2(-5) <1.1(-5)

IssEu 4.0 1 2.0(-6) <4.8(-6) <8.5(-6) <3.4(-6) <7.5(-7)

[A] Sample from 9/9 - 9/23/76

[B] Sample from 9/23 - 10/7/76

..j TA8LE A.36 (Cont,)

VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #2 (uci/sec) l Sample Period 10/7 - 10/14/76 10/14 - 10/22/76 10/22 - 10/28/76 10/28 - 11/4/76 11/4 - 11/10 76 1311 3.80 0.02(-4) 3.4 0.3(-4) 1.66 0.01(-4) 1.57 0.01(-4)' 5'.72 i 0.04(-5) 13"Cs <5.7(-6) <1.7(-6) <2.2(-6) <8.6(-7) <3.5(-7) 136Cs <1.7(-6) <3.1(-6) <9.6(-7) <8.1(-7) <1.1(-6) '

137Cs 3.1 1.1(-5) <1.5 0.4(-5) 2.1 1.0(-6) 1.3 0.6(-6) 1.5 1 0.8(-6)'

3H ' 2.0 1 1.0(-3) 2.0 2 1.0(-3) 4.7 0.5(-3) 4.7 i 0.5(-3) 2.0 1.0(-3) 14C 2.5 0.1(-3)[C] 9.4 0.3(-3)[D]

SI Cr 1.16 2 0.03(-2)

<1.3(-5) <2.9(-5) 5.2 12.6(-6) <3.8(-6) <6.7(-6

'54Mn <2.5(-6) <7.3(-6) 2.9 N$ 59Fe <8.0(-6) 0.8(-6) <1.1(-6) <4. 7(-7llJ

<1.2(-5) <2.8(-6) <1.9-6) <1.3(-6) s7Co <l.5(-6) <2. 0(-6 ) 6.5 2.6(-7) <4.3 -7) <3.2(-7) 58C0 5.4 1.7(-6) 4.6 2.7(-6) 8.2 1.6(-6) <1.5-6) 60Co <1.1(-5) <4.8(-6) 9.8 2.1 i 0.8(-6) 652n -<3.1(-6) 1.5(-6) 2.4 i 1.0(-6) <1.6(-6)

<8. 5(-6) <5.3(-6) <8.8(?) <1.1-6)

-ssZr <7.0(-6) <5.1(-6) <2.2(-6) 9sNb <1.1(-6) <7.7 -7)

<2.8(-6) <3.9(-6) <7.9(-7) <6.7(-7) 103Ru 2.4 1.4(-6) <1.9(-6) <2.0(-6) 106Ru- 2.0 0.9(-5) <1.4(-5) <9.2(-6)

<1.3 -6)

<4.7 - <E.0l-7)

<5.7I -7ll

<4.81 -6;l 110 mag 42:6(-6) <4.2(-6) <1.3(-6) <4.7 - <8.7d-8) 124Sb w.0(-6) 3.6 1 2.2(-6) 1.7 0.7(-6) <1.1 - <1.7d-6j 12sSb <1.9 - <1.6(-6) <4.2(-6) <2.2(-6) <2.8d,-6 'l 140Ba <9.7 - <1.2(-5) <5.1(-6) 2.9 1.6(-6) <2.0I -6!l 140La <1.2 - <3.4(-6) <5.8(-6) <2.0(-5) <1.3d-51 141Ce <2.8(-6 2.2 1.3(-6) 5.2 2.6(-6) 4.2 1 2.2(-6) <9.2d-7!!

152Eu <1.7 - <6.7 - 2.0 i 0.7(-6) <2.2 is"Eu <1.2 - <5.0 -

<1.8q-ll 1ssEu <3.2 - <2.9 -

<3.0(-6)

<1.2(-6)

<1.1([- ll i <1.3 l

<1.1(- ll <1.4(1 -- ll

[L) Sample from 10/7 - 10/22/76

[D] Sample from 10/22 - 11/4/76 t

e- e 4

TABLEA.36(Cont.)

- VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #2 -(pC1/sec)

Sample .'

1/6 - 1/20/77 EE]-

Period 11/10 -'11/23/76 11/23 - 11/30/76 12/1 - 12/15/76 .12/15f76-1/6/77 131I 6.47L 0.04(-5) '1.75 0.01(-3) 1.06 0.09(-4) 6.6 0.4(-5)' '5.0 . 'O.4(-5) 134Cs 2.5 - 0.5(-6) <1.2(-6) '2.0 - 0.4(-6) 9.8 0.2(-5) 4.96 i 0.03(-5) 136Cs <7.6(-7) <5.3(,7) <2.1(-6) .

<9.3(-7) <7.6(-8)

.137Cs 3.1 i 0.7(-6) 3.0 1 0.9(-6) 4.6 0.8(-6) 1.20 0.03(-4) 5.85 0.03(-5) 3H ~2.2 1 0.1(-2) 1.9 i 0.9(-2) 5.5 i 0.1(-2) 1.0 0.1(-2) 6.3 0.1(-3)

I r.C - 1.88 ,0. 04 (-2 ) ' 4.3 0.2(-3) 1.8 0.2(-2) 3.2 'i 0.2(-3) 1.12 0.04(-2)

U 51Cr <8.2(-6) <5.5 -6) 4.0 3.0(-6) <1.5 -5) <3.8(-4) s4Mn 9.0 1 4.0(-7) <2.5 -6) 2.0 0.5(-6) <7.3 -7) ~6.3- t 0.8(-7) 59Fe <1.0(-6) <6.5 -7) <1.1(-6) <5.6 -7) <1.9(-6)

-57C0 <3.0(-7) 1.4 0.9(-7 5.0 12.0(-7) <6.1(-7) <1.4(-7) seCo 9.0 1.0(-6) 2.6 0.8(-6 3.5 0.2(-5) 2.0 1 0.4(-6) 1.0 0.1(-5) 60Co 1.0 2 0.3(-6) 1.9 1 0.7(-6 2.5 0.5(-6) <5.0(-7) 1.05 0.04(-6) 65Zn <4.4(-7) <7.3(-7) <9.8(-7) <1.4(-6) <3.0(-7) 952r <3.4(-7) <1.2(-6) <2.2(-6) <1.4(-6) <1.6(-6) 9sNb <1.2(-6) <1.1(-6) <1.8(-6) 3.0 2.0(-7) <7.4(-6) 103Ru <8.4(-7) 5.0 1 3.0(-7) 6.0 4.0(-7) 5.0 4.0(-7) (8.6(-6) 106Ru <9.9(-6) <4.0(-6) <3.5(-5) <9.2(-6) <2.4(-6) 110 mag <2.8(-7) <1.2(-6) <5.2(-7) <1.3(-6) <3.1(-7) 124Sb' 3.3 1 0.7(-6) <1.0(-6) 3.0 1.0(-6) <8.2(-5) <2.6(-4) 12sSb ~<2.2(-6) <2.5(-6) 1.3 0.7(-6) <5.7(-6) <8.7(-7) 140Ba 1.7 1 0.0(-6) <4.6(-6) 3.0 2.0(-6) <7.9(-6) <5.5(-7) 140La <1.6(-5) <1.9(-5) <3.9(-7) <5.1(-7) <1.3(-4) 141Ce 2.0 i 1.0(-7) 6.0 ' 3.0(-7) <1.3(-6) <1.4(-6) <1.5(-5) 152Eu 1.0 0.5(-6) <2.3 -6 <4.1 -6 <5.0 -6 <6.0 -7 154Eu <1.1(-6) <9.6 -7 <4.8 -7 <1.D -7 <1.9 -7

[E] 60,000 second count

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

4 N-TABLEA.36(Cont.)

VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #2 (uct/sec) e 1/20 --2/3/77 2/3 - 2/16/77 131I. 1.8 ' t 0.2(-4) 2.1 0.2(-5) 134Cs- <4.0(-6) 3.0 i1 136c5 <9.3(-8)' <4. 2 (-7) . 0(-5) 137Cs 4.5 - ~- 0. 6 (-6 ) 3.4 1 0.2(-5) 3H [F] '6.0 t.1.0(-3) 14C [F] 5.0 . 0.3(-3) y 51Cr <2.9-6) <1.0 -5)

'34Mn <1.9-7) <7.2 -7) 59Fe <6.1 -7) <3.8 -6) 57Co 38

<2.3(-7) 9.0 7.0(-8)

SO 0o '2.0 1 1.0(-7) 2.5 1 0.5(-6)

Co <1.7 -7) <9.8(-7)

, 6sZn- <7.6 -7) <4.9(-7 ssZr <4.9 -7) <6.0(-7 9sNb <2.6(-7)~ <5.0(-7 103Ru <4.0-7) <1.5 -6) 106Ru- <2.0 -5) <6.1 -6) 110 mag <3.0 -7) <6.7 -7) 124Sb 5.2 1 0.3(-6) 3.0 2.0(-5) 12sSb L<5.5(-7 <3.9(-6 140Ba <2. 5 (-6 <5.9(-6 140La <5.9(-6 <2.2(-5

<4.4(-7 1.1 1 0.9(-6) 141Ce 152Eu 4.0 3.0(-7) <4.4(-6) 154Eu <3.2(-7) <4.2(-6)

[F] Sample station #2.- 14C3H Sampler inoperable due to motor malfunction.

e t

TABLE A.37 "t' VENTILATION AIRBORNE ACTIVITIES

- - SAMPLE STATION #3 (pCf/sec) ,

' Sample.

Period 9/3 - 9/s/76 9/9 - 9/15/76 9/15 - 9/23/76 9/23 - 9/30/76 9/30 - 10/7/76--

-131I' 1. 7 . 'O.1(-5) 2.9 0.6(-6) 2.3: 0.6(-6) 4.5

• 0.7(-6) 6.5 .i 0.4(-5)

'134Cs <3.1(-6) <1.5(-6) <1.1(-6) <9.9(-7) <2.9(-6)

. <1.8(-6)-

~

'136Cs '7.4(-7)

< <1.7(-6) <1.7(-6) <1.5(-6) 137gg- 9.4- 2.3(-6) 8.3 3.1(-6) 4.6 i 1.3(-6) 5.2 i.1.6(-6) 1.0 .i 0.2(-5)'

I cH 3

<3.0(-4) <3.0(-4) <3.0(-4) <3.0(-4) <3.0(-4).

14C 1.6 i 0.7(-4) 1.9 0.2(-4)[A] A.310.3(-4)[B]

s1 Cr

<1.8(-6) <1.0 -5) <1.8(-5) < 1.'1(- 5 ) <2.2(15)

U- 5"Mn <1.7(-6) <1.1 -6) <1.1(-6) <1.0(-6) <4.1(-7)

M <1.5(-6) 59Fe <3.0(-6) <4.1-6) <9.9(-7) <1.2(-6) 57C0 9.2 - .4.1(-6) <6.0 -6) 3.3 2.3(-6) 3.5 i 1.6(-6) <6.9(-6) ssCo ' <4.5(-7) <1.3(-6) <7.3(-7) <8.8(-7) <5.2f-6):

60Co <2.2(-6) <4.8(-6) <1.4(-6) <1.2(-6) <1.1(-6) ssZn <1.9(-6) <1.7(-6) <1.1(-6) <5.4(-7) 95Zr <2.1(-6) <7.1(-7) <8.4(-7) <9.6(-7) -<1.1

<1.3(1-6)-

,-6) 9sNb <1.3(-6) <6.2(-6) <2.7(-7) <7.9(-7)

<2.3(1-6) 103Ru .2.2(-6)-

< <1.4(-6) <1.3(-6) <2.0(-6) <1.5 ,-6) lo6Ru < 1. 0 (~-5 ) - <1.3(-5 <9.6(-6) <4.3-6)- < 1.1ll-5).

110 mag <8.8( 7) <1.3(-6 <6.4(-6)~ <2.4 -6) <7.2ll-6) 124Sb ' <2.4(-6) <2.0(-6

<2.1 - <1.3-6) <1.7 12sSb <3.0(-6) <9.8(-7 <2.9 - <8.3 -7) <6.7 (I-6)

,-6) 140Ba <6.8(-6) <5.7(-6) <5.0 - <1.7-6) < 5. 31;-6)

<1.9(-5) <1.8 -5) <7.6 -6) <2.51,-4) 140La <1.2-5) .

6.1 141Ce <1.~2 -5) 5.3 2.4(-6) 5.9 i 3.1(-6) <8.8 -6) 3.8(-6) 152Eu <5.3 -6 <1.8-6) <4.7(-6) <9.8 - 3.3 1 1.7(-6) 154&.1 <1.2 -6 <2.0 -6) <2.5(-6) <2.2 - -

<4.1(-6)

Issta <3.6(-5 <1.8 -4) <9.2(-5) <2.2 - <5.2(-5) i

[A] Sample from 9/9 - 9/23/76

[B] Sample from 9/23 - 10/7/76

d TABLEA.37(Cont')

VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #3 -(vCi/sec)

Sample -

Period. 10/7 - 10/14/76' 10/14 - 10/22/76 10/22 - 10/28/76 -10/28 - 11/4/76 11/4 - 11/10/76

1311. 1.6 0.2(-5) 1.3 10.2(-5) 5.0

• 2.0(-6) 2.0 i O.'4(-6) 4.0 0.9(-6)

"e 137Cs

1:!{:'l 1.2 2 0.2(-5)

6.8:'{:!l 2.7(-6)

'{:'l  :!:ll:ll :1:!{:R 1.1 1 0.5(-6) 7.0 2.0(-6) 2.4 1.3(-6) 3H- 1.0 2 0.3(-3) 1.0 1 0.3(-3) <3.0(-4) 1.3 0 2(-3) 8.0 1 2.0(-4) 14C 2.0 i 0 2(-4)[C] 1.A 0 1(-4)[D] 1.45 i 0.06(-3).

51Cr ~< 3.5I;-6 ~Z.0 - <2.21 <3.1 -

g

<1.9-5)

- 54Mn <1. 31,-6 <1.3 - <1.81 -

<2.2 - <1.4 -6) 59Fe <2.01 -6 <3.6 - <3.4I -

<2.4 - <3.4 -6 57Co <2.61 -6 <1.1 - <2.21 - .

<6.7 - <1.0 -5 s8C0 < 1. 51 -6) <1.2(-5 <3.31 -6) <4.0(-6) <8.4(-7 60C0 <2.01 -6) <5.0(-6 <3.91 -6) <2.8(-6) <3.0(-6 65Zn <3.9 -7) <2.6I -6 <7.0ll-7

<3.0 - <2.6 -

95Zr <6.2 -6) <1.01 -5 <2.5(-6 <8.5 - <2.6 -

9sNb <1.8 - <1. 61 -6) <1.0(-6 <4.9 - <1.2 -

103Ru <1.7 - <1.41 -6) <5.1(-6 <5.0 - 1.7 0.9(-7) losRu <2.4 - <7.3I;-6 <2.8(-5) <6.1 -6 <9.8(-6) 110 mag <6.6 - <1.61,-6 <9.3(-7) <9.1 -6 <1.9(-6) 124Sb <1.3(-6) <1.4 -6 <6.6 -6) <1.4 - <9.1 -

12sSb <4.4 - <1.8 - <7.2 - <2.8 - <5.8 -

140Ba <4.7 - <8.0 - '<1.7 - <7.2 - <2.6 -

140La <4.2 - <1.1 - <3.2 - <1.4 - <2.0 -

141Ce- 4.5 1 2.6(-6) 5.9 3.2(-6) <4.0 - <9.9 - 1.0 0.5(-6) is2Eu <3.5(-6) <8.3(-7) <4.9 -6) <4.9-6) is4Eu

.<1.2 -6)

<3.0(-6). <2.2(-5) <5.7 -6) <1.6 -6) <3.8 -6)

IssEu' <4.1(-5) <1.4(-4) <5.5 -5) <2.6-6) 5.4 i 2.6(-5)

[.' ; Sample from 10/7 - 10/22/76

[D] Sample from.10/22 - 11/4/76

, TABLE'A.37(Cont.) .

VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #3 (uC1/sec)

~11/10 --11/23/76 11/23-11/30/76[E] 11/30 - 12/15/76 12/15 - 1/6/77 1/6-'1/20/77[F]

1311. 3. 9 , 0.1(-5) 4.3 0.9(-6) 3.8 0.6(-5) 7.0 12,0(-7) 2.0 1 0.6(-6) 13"Cs.

13s <3.2(-6) <2.1(-7) <6.6(-7) <3.6(-7) 4.2 i 0.1(-7).

137 Cs <1.3(-7) <1.7(-6) <1.4(-7) <1.1(-8)

Cs <1.2(-6) 5.0 i1 0(-6) <2.7(-7) 1.8 i 0.7(-6) <3.0(-7) 4.6

• 0.2(-7) 3 14 H 7.0 1.0(-4) <3.0(-4) 9.3 i0.1(-3) 2.3 10.1(-3) 3.6 10.2(-2)

C 1.56 i.0.06(-3) 2.8 2 0.2(-4) 4.3 0.4(-3) 6.2 0.4(-4) 1.62

• 0.08(-3)

O stCr 6.0 1 4.0(-6) <8.6-4) 9.0 i 4.0(-6) <7.9 - <2.8(-5).

54Mn <1.8(-6) <2.6 -7) <6.1-7) <1.3 - - 4.4 0.7(-8) 59Fe <2.3(-6) <2.3 <1.4-6) <2.3 - 1.0

• 0.2(-6) 57Co 2. 0 1 1.0(-6) <1.4 - <2.4 -6) <7.6 - <4.0(-8) seCo 1.2 1 0.2 <1.8 - 6.0 1.0(-6) 2.0 i 1.0(-7) 1.9 0.3(-7) soCo <8.6(-7) 1.0 1 0.4(-7) <1.9(-6) <8.6(-8) 5.5 0.5(-8) ssZn <1.4(-6) <4.9(-7) <3.5(-7) <2.0(-6) <4.6(-8) 95Zr <1.8(-6) <4.9(-6) <2.1-6) <1.1(-6) <1.7(-7)  !

95Nb <9.0(-7) <2.8(-5) <4.0 -6) <2.0(-7) <1.2(-6) lo3Ru <2.4 -7) <1.5(-5) <2.8 -7) <1.0(-7) <1.8(-7) lo6Ru <1.4-5) <3.5(-6) <1.1 -5) <1.3(-6) <2.8(-7) 110 mag <7.5 -7) <4.0(-7) <3.1(-7) <1.2(-7) <1.5(-8) 124Sb 1.8 1 0.9(-6) <3.1(-6 <1.0(-6) <l.6(-7) <2.0(-6) 125Sb 3.0 1 1.0(-6) <4.5(-7 <2.0(-6) <3.2(-7) 2.4 0.7(-8) 140Ba <2.7(-6) <3.9(-7 <2.2(-5) <2.0(-7) <7.8(-8) 140La <2.3(-5) <1.0(-7) <3.5(-7) <1.3(-6) <1.3(-8) '

141Ce <8.7(-6) <3.9(-5) <1.4(-6) <5.8(-8) <2.1(-6)

Is2Eu <1.1(-6) <4.5(-7) <1.8(-6) <7.8(-8) 8.0 1 0.8(-8)

IS4Eu <2.1(-6) <3.5(-7) <9.7(-7) <2.6(-6) <6.1(-8)

' E" 24,201.9' second count 229,501.1 second count lF

~

TA3 LEA.37(Cont.)

VENTILATION AIRBORNE ACTIVITIES- '

. SAMPLE STATION #3 (uci/sec)'

. Sample '

• Period. 1/20 - 2/3/77 2/3 - 2/16/77 131I 1.6. 0.5(-6) 7.0 2.0(-7) 134Cs <1.3 -7)' <1.5(-7 136Cs '<2.4 -7) <3.6(-8 137c5" <4,4 _7} <4,9(.8

~

3H 2.0 t 0.1(-3) 1.9 0.3(-3) 14C 1.38' O.06(-3) 1.26 0.08(-3) '

., s1Cr- <2.0 -6) <2.4 -

M' '54Mn <1.6 -7) <9.9 -

59Fe <2.9-6)' <2.9 - -

57C0 <1.6 -7) <3.3 -

58Co -<4.3(-7) <2.6(-7) 60Co .<4.2(-7) <1.7(-7) 6sZn <2.0(-6) <2.9(-7) 95Zr <2.2(-7) <1.2(-7) 9sNb <2.2(-7) <1.6(-7)

-~103Ru '1.8(-7)

< <8.4(-8) .

106Ru <2.2(-6 <1.2(-6) 110 mag <1.8(-7 <1.6(-7) 124Sb <3.3(-7 <1.8(-7 12sSb <4;1(-7 <1.3(-7 140Ba <1.2(-6) <2.8(-6 140La <2.8(-6) <5.6(-6 141Ce- <2.7-7) <6.4(-8 152Eu <4.5 -7) <3.6(-7 .

154Eu <3.1-6) <3.2(-6 uA *W m

-_._______._____._.-.J -

TABLE A.38 VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #4 (uti/sec)

Sample Period 9/3 - 9/9/76 -9/9 - 9/15/76 9/15 9/23/76 9/23 - 9/30/76 9/30 - 10/7/76-1311 <4.5(-6) 1.6 2 0.1(-4) 8.4 0.7(-5) <3.2(-6) 2.1 i 0.4(-3) 134Cs <9.7(-6) <1.8(-5) <8.8(-6) <1.0(-5) <5.7(-6) 136CS < 8.4(-6) <4.4(c5) <6.8(-6) <2.8(-6) <4.9(-5) 137Cs 4.6 1.1(-5) 4.7 t-2.0(-5) 4.0 1.0(-5) 2.9 t 0.8(-6) 3.6 '0.9(-5) 3H <1.0(-3) <1.0(-3) <1.0(-3) <1.0(-3) <1.0(-3) l 14C 3.0 .6(-4) 2.1 t 0 4(-4)[A] 3. ?. 0.6(-4)[B]

s2Cr <4.3 - <1.7 - <4.5(-5 <3.4(-5) ,

54Mn <2.7 - <2.1 - <9.3(-6 <3.6(-6) <7.0ll-5)J

<3.0t -6 U '59Fe <2.1 - <2.1 - <4.9(-6 <1.0 -5) <1.1ll-51

• <2.8 -5) <2.3I -5) 57Co <3.7 - <1.7 - 2.6 1.2(-5) 58C0 <5.0 - <8.0 - <4.3(-6) <6.7 -6) <3. 5 ( -61 6000 <1.5 - <1.4 - <9.2(-6) <1.1-5) <1.61 -Sh 6sZn <1.0(-5) <1.8(-5) <6.7(-6) <7.4(-6) <8.2(-6) 95Zr <2.1(-5) <1.1(-5) <6.3(-6) <4.4(-6) <4.1(-6) 95Nb <7.1(-6) <1.3(-5) <1.5(-6) <8.1(-6) 103Ru <1.7(-5) <4.4(-6) <1.0(-5) <6.4(-6) <7.5(l-6)

<3.8i -6) 106Ru <7.5(-5) <1.6(-5) <2.2(-5) <2.0(-5) <2.6(-5)

<6.6(-6) <9.5(-6) <3.5(-6) <3.71 -b) 110 mag <9.6(-6) 124Sb <2.1(-5) <4.5-6) <4.2(-6) <1.5(-6) <1.2Il-5) 12sSb <1.4(-5) <3.4 -5) <1.5(-5) <1.6(-5) <1.8L-5) 140Ba <2.7(-5) <1.5 -5) <2.0 -5) <3.0(-5) <1.6 -

1401a <1.4(-4) <2.0 -4) <8.3-5) <8.8(-5) <1.0 -

141Ce 4.2 12.2(-5) <5.7(-6) <8.1 -6) <4.6(-5) <5.0 -

152Eu 2.0 1.0(-5) <3.3 - <2.0 -5) <2.8-5) <1.2 -

154Eu <1.9(-5) <5.9 - <3.6 -5) <9.6 -5) <9.2 -

155Eu <5.4(-4)~ <3.2 - <3.2 -4) <2.8 -4) <6.3 -

[A] Sample from 9/9 - 9/23/76

[B] Sample from 9/23 - 10/7/76

L TABLEA.38(Cont.)

C VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #4 ( uCf /sec) i Sample Period 10/7 - 10/14/76 10/14 - 10/22/76 10/22 - 10/28/76 10/28 - 11/4/76 11/4 - 11/10/76 131I 2. 3 - 0.1(-4) 7.3 0.7(-5) 7.6 -1 0.6(-5) 7.8 1 0.7(-5) 8.0 i 140(a5) 134C's 7.1 t 4.0(-6) <6.5(-6) <2.3(-5) <5.6(-6) <9.8(-6) lasts <5.2(-6) <5.3(-6) <5.3(-6) 6.5 <1.4(-5) <5.2(-6) 137Cs 2.3(-5) 2.5 1.0(-5)' 5.4 2.4(-5) 3.2 1 0.9(-5) 1.2 1 0.7(-5) 3H <1.0(-3) <1.0(-3) 2.3 0.2(-2) 1.5 i 0.3(-3) <1.0(-3) 14C 2.74 i 0.07(-3)[C] 1.00 0.07(-3)[D] 2.7 i 0.1(-3).

- 51Cr 8 <1. 0(-5) 5.1 1 2.8(-5) <6.2 -5) <2.6 -5) 4.8 i 3.5(-5) 54Mn <8.7(-6) <2.6(-6) <8.9-6) <3.7-6) <2.9(-5)

SSFe <8.8(-5)

<1.1(-5) <1.0 -4) <5.7 -6) <1.1(-5) 57Co 3.2 1.3(-5) <2.1(-5) <4.2 -5 <7.2-6) <4.2(-5 seCo <1.3 -5) <4.5(-6) <2.4-6) J <1.0 -5) <2.0(-5 saCo <1.7 - <1.2 - <5.0 'l <9.5 - <1.0 -

55Zn <6.1 - <9.1 - <6.7 - <3.4 -

ssZr 95Nb

<1.5 -

<2.1 -

<4.0 - <9.2-)I

<8.2 - <3.0 - <7.0 -

<2.2 - <6.3-h <4.3 - <7.9 -

103Ru <4.5 - <3.1(-6) <3.9 - <7.1 - <2.5 -

10sRu 110 mag 3.4 i1.6(-5) <5.2(-5) <4.3 - <5.0 - <9.7 -

<7.5(-6) <1.2(-5) <4.3 - <4.6 - <5.4 -

124Sb <1.6(-5) <3.1(-6) <1.3 -

12sSb

<8.3 - <1.1 -

<2.3(-5) <1.2(-5) <4.3(-6) <1.6(-5) 140Ba <4.2(-5) <1.2(-5) <5.4(-6) 140La

<1.4(-5) <1.8(-5) <4.5(-5)

<1.6(-4) <1.2(-3) <5.8 - <1.7 - <2.6(-4) 141Ce <1.7(-5) 2.4 i 1.2(-5) <7.0 - <2.1 -

is2Eu 1.5 0.9(-5) <6.2(-5)

<2.0(-5) <1.1 - <9.8 - 3.6 i 1.8(-5) 154Eu <1.0(-5) <1.2(-5) <1.1 -

IssEu 9.6

<1.3 -5) <2.2 -5)

<5.9(-4) 6.8(-5) <8.7(-4) <2.5 -4) <3.3-4)

[C] Sample from 10/7 - 10/22/76

[D] Sagle from 10/22 - 11/4/76

f

' TABLE A.38 (Cont.)

4-VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION #4 (uCf/sec)

Sample Period 11/10 - 11/23/76 11/23 - 11/30/76 11/30 - 12/15/76 12/15 - 1/6/77 1/6 - 1/20/77 131!. 8.0 i 1.0(-5) 8.0 0.6(-5) l'. 8 0.2(-4) 7.8 t 0.5(-5) 6.5 1 0.7(-5) 134Cs ~<5.0(-6) '<5.6(-6) <6.5(-6) 1.55 1 0.04(-4) <6.5(-7) 136cg <7,4(;6) <3.1(-6) <1.0(-5) <2.1(-6) <4.0(-7) <

137CS 1.4 i 0.4(-5) 2.0 1 0.6(-5) 9.0 3.0(-6) 1.82 0.06(-4) <1.1( i) 3H 7.4 2 0.1(-2) 2.2 *0.1(-2) 1.8 i 0.1(-2) 1.3 1 0.1(-21 9.3 1 0.4 -

14C 1.00 t 0.04(-2) 4.6 0.3(-3) 1.73 1 0.04(-t) 3.6 i0.2(-3} 4.8 i 0.2 -

g 51Cr <8.516) <2.2(-5) <3.5(-5) <3.7(-5) <2.6(-5) 54Mn <3.7 - <5.3(-6) <2.9(-6) 1.4 -6) <3.5(-7)-

59Fe <1.3 - <1.9(-5). <1.9(-5) <6.7 -7) <1.1(-5) 57Co .< 2.5 - <2.6(-5) 1.0 1 0.0(-5) <1.6 -6) <5.3(-7) seCo 6 i 3(-6) <4.6(-6) 2.4 0.5(-5) 1.7 0.6(-6) <2.1(-6) 60Co <4.2(-7 <1.2-5) <6.2(-6) <2.5 - <3.7(-7) 65Zn <1.5(-6 <9.1 -6) <1.4(-6) <1.7 - <7.4 -6) 95Zr <2.5(-5 <4.3 - <3.9(- <1.7 - <9.6 -7) 9sNb <2.5 -6 <1.8 - <3.5 - <1.9 - <4.5 -

103Ru <6.4 - <8.2 - <3.8 - <4.8 - <4.5 -

106Ru <3.8 - <3.4 - <4.8 - <1 9 - <2.5 -

110 mag <1.6 - <3.4 - <7.8 - <4.3 - <3.4 -

124Sb <5.3-6) <1.3 - <5.3 - <1.3 - <6.2-7) 12sSb 7.0 i40(-6) <2.0 - <5.7 -

<4.3 -

<1.3 -5) <9.4 -7) 140Ba <1.0 5) <5.9 - <1.6(-5) <7.2(-7) 140La <4.9 -5) <3.1 - <1.9 - <2.7(-6) <5.1(-4) 1*1Ce <1.3 - 2.0 1.0(-5) <5.2(-5) <3.6-6) <5.5-7) 152Eu <9.0 - <3.2(-5) <7.7(-6) <1.4 -5) <6.8 -7) 154Eu <5.3 - <1.5(-5) <2.4(-6) <1.9 -6) <1.7-6)

(

TABLEA.38-(Cont.)

~

VENTILATION AIRBORNE ACTIVITIES SAMPLESTATION#4-(uC1/sec)- Q

' Sample:

Period. -1/20 - 2/3/77 2/3 - 2/16/77 131I 2.3 1 0.1(-4) 5.0'.12,0(.-6)'

134Cs . 9.0 1.0(-6) <1.6 -

136Cs <4.4(-7) <5.2 -

137Cs 1.1 1 0.2(-5) <9.7 -

3H~ 2.7 *.~ 0. I(-2) - 1.1 0.2(-2) 14C- 4.8 ~ 0.3(-3) 5.2 1 0.6(-3) siCr- <4.8(-6) <4.7 -6 54Mn <7.1(-7 <4.1-7) 59Fe <l.7(-6 <1.2 -6) s700 <6.0 -7 <2.4 -

-58C0 <1.6 -6 <6.4 - '

60C0 <6.2.-7) <7.2 -

652n <1.6(-6) <8.6 -

95Zr <5.5(-6) <5.2(-7 95Nb -<4.8(-7) <5.3(-7 103Ru <7.2 - <9.4(-8 106Ru <5.2 - <4.1(-6 110 mag <4.7 - <6.9 -7 124Sb 1.2 0.2(-5) <1.1-6) 12sSb .<3.3-6) <5.6 -7) 140Ba <5.8-6) <7.9(-7)

• 140La 141

<1.2 -4) < 2.2(-5)

Ce <7.3-7) <1.4(-7) 152Eu <5.4 -6) <5.3(-6) 154Eu <1.3-5) <1.4(-5)

• ~' ' '

,4

' T' ABLE A. 39

• VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION WASTE EVAPORATOR- (uC1/sec)

. Sample. .

Period 9/3 - 9/9/76 9/9"- 9/15/76 9/15 - 9/23/76 9/23 - 9/30/76 . 9/30 - 10/7/76-1311- 1.63 2 0.01( 4) 4.70- 0.05(-5) ..4.80 i 0.04(-5) 1.92 1 0.03(-5) 1.55 '0.03(-5)-

134Cs. <2.4(-7)- <6.4(-7) <5.0(-7) <6.6(-7): . <3.7 -

136Cs .<2.4(-7) .<1.1(-6) <1.0(-7)~ <1.4(-7) . <3.2 -

137C$- .<6.0(-7) 7.0 - . 4.0(-6) 1.0 1 0.3(-7) '3.81 1 1.9(-7). <5.9 -

't 51Cr- 20'210(-6) <2.4 - <1.1(-6) .< 2.1 -6).

<1.9 -6) 54Mn ~ <4.0(-7) <3.4 - <2.6(-7) <9.8-8) <1 3 -7)-

59Fe_ <6.7(-7) <1.1 - <2.8(-7). <4.4 -6) . <5.7 -7)~

57Co - : 3.0 - 1.0(-7) . <5.7 -8) 5.0 3.0(-8)

<1.9 -7) ' <8.2 -8)' -

58C0' 46.0(J7) <6.8 -7) <5.4 -7) .<2.4 -7) <2.6 -7)"

y , 60Co ~< 4.2(-7) <5.5-7) <1.5(-7) <3.0(-7)-

10 ' 65Zn <8.8(-8) .<1.8 -7) <4.5-8))

<1.7 -7 <1.4(-7) .<7.2(-7 ssZr. <7.7(-7) <1.4 -7) <1.5(-7) <5.2(-7 '

95Nb <1.8(-7)- <2.1 -7)' <3(-7)

<3.8(-7 )- <3.7(-7) <1.4(-7 1103Ru <2.7 - <2.2(-7 <1.7(-7) <3.6-7) <3.1(-7) 106R u' . <2.5 - <1.8(-6 <1.9(-6) <4.9 -7) <1.9(-6) 110 mag <3.0 - <3.0(-7 .<3.1(-7) <4.6 -7)- <5.2(-7) 1124Sb <9.6-7) <2.1(-7) <5.4(-7) <3.2 -7) <4.4(-7) 12sSb '4.0((-7)

< <1.3(-6) <1.8(-7) <2.5(-7) <4.6(-7) 140Ba <5.5 -7) <4.9(-7) <1.0(-6) <1.4(-6) <1.0 -

140La <4 0((-6) <4.9 - <6.5(-6) <5.0(-6) <4.1 -

141Ce :3.8.'1.7(-7) <3.1 - . 2.0 11.0(-7) 1.4 i0.9(-8) <1.4 -

152Eu <1. 9(-7) <3.8 - <5.5(-7) 4.9 3.7(-7) ~ <7.5 -

154Eu <3.0(-7) <6.8 - <2.6(-7) <4.8(-6) <9.9 - 4 1ssEu 3.0 _ 2.0(-7) <5.1 - <5.0(-7) <2.3(-7) <3.2 -7)-

eI._._____.__________.______._______ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . - - - --m 'a .

+

' . I q.

i$

9 TABLE A.39 (Cont.)

VENTILATION AIRBORNE ACTIVITIES-- ,

, ~ SAMPLE STATION WASTE EVAPORATOR' (uC1/sec) ~

1 haple Period. E10/7'- 10/14/76' '10/14 10/22/76 -10/22 - 10/27/76~ '

10/27 - 11/4/76EA3 ! 11/5~- 11/10/76'

~

1.91

~

131I. 0.02(-5) 3 21 i 0.06(-5) .4.37 i 0.07(-5).' 1.51 ' O.04'(-5)

.134Csi <9.1(-7)- ' <8.2(-7) 5.1 2.7(-7)' '. <9. 3(-7) 13sCs- <4.9(-7). <4.1(-7) <2.7(-7) ,

<4.5(-7)'. . -

(137CS- : 3.4?i.0.6(-6)~ 1.6-1.0.5(-6) .  ; 3.2 0.7(-6) 8.4; .3.2(-7)l 51Cr 1.7 i 1.1(-6) <6.9 - <4.6 - .<5.6 -

54Mn- 1.0. 1 0.3(-6) <8.8 - <3.1 - <1.6 -

59Fe <9.4(-7) <9.1 - <6.2 - <7.1 -

57Co. 9.6 5.2(-8) 1.2 i 0.7(-7) <1.3 - J .<9.3(-8

seCo 1.5 0.4(-6)' 1.7 0.6(-7) <8.6 'l '< 3.7 -

• ~ .60Co 6.3 1 2.2(-7) ' 4.2 -7)

< <2.9 'l <3.8 -

65Zn <g,g _g} - <5.1-7) <4.8 'l <9.3 -

95Zr <3.5 -7)' <5.0 -7) * <6.6-!l <2.9 -

95Nb' <2.4 -7) <1.0 - <4.3 'l <1.6 -

103Ru <8.4 -7) , <2.1 -

<6.3-h 106Ru 3.1 - 1.5(-6) <7.5 - <1.0(-7)

<1.8 - J <1.0 110mA9 <1.5(-8) <3.6 - <6.9 -7) <7.0(-6) 124Sb- 9.1 3.2(-7) <9.7(-7) <5.6 -7) <2.5f--

12 sSb'. <8.5-7) <1.4(-6) <1.2-6) <5.3I 140Ba ~ <1.2 -6) <3.3(-6) <3.1 -6) <1.lf- -

140La <8.4 -6) <3.5 - <1.3-4) <1.2d-141Ce <3.5 -7) <2.1 - <2.5 -7) 1.4 1.0(-7) 152Eu <7.0(-7) <8.5 - 7.7 4.4(-7) <9.7(-7)'

154Eu <4.7(-6) <8.0(-7) <1.2(-6) <1.7(-6) issEu <3.0(-7) <2.2(-7)

<2.0(-7) <6.5(-7)

'[A] No sample due. to sampler motor malfunction

-TABLE A. 39 (Cont.)

VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION WASTE EVAPORATOR (u ci/sec)

Sample Period 11/11 - 11/23/76 -11/23 - 11/30/76 11/30 - 12/15/76 12/15.- 1/6/77 1/6f--1/20/77: ,

131I 1.22't'0.01(-5) . Si0 0.1(-6) 6.0 i 2.0(-6) 8.0 ' 2.0(-7) <1.1(-7) 134Cs < 3.7(-7) .<4.7(-7) 5.0 1 2.0(-7) 2.9 0.1(-6) <1.6(-7) 136Cs <9.9(-8) . <2.4(-7) <2.8(-7) <9.9(-8) <5.1(-7) .

137CS 5.0 3.0(-7) 6.0 . 3.0(-7) 1.1 0.3(-6) 3.5 i 0.3(-6) 1.0 0.5(-7) s1Cr . <2.0(-6) <6.7 -7) <2.8(-6) - 8.0 . 4. 0(-7) <4.7(-7 54Mn <2.6(-7) <2.3-7) <3.7(-7) 1.5 0.6(-7) <6.2 -8 59Fe < 1. 9 (-.7 ) <9.2 -7) <2.8(-7). <8.5(-7) <1.1 -7 57Co 8.0 4.0(-8) 1.0 0.5(-8) 4.0 2.0(-8) <6.5(-8) <4.8 -7) .:

seco ' 4.0 '2,0(-7)

2.7(-7) 1.0 1 0.3(-6) 6.0 t 1.0(47) 2.1 1 0.7(-7) g

.m 60Co <5.8(-7) <4.0(-7) <4.2(-7) <1.1(-7) <9.9 -8) 6sZn -2.1(-7) <2.2(-7) <2.0(-7) <8,5(-8) <6.7-8) 95Zr <3.2(-7) <1.7(-7) <8.5(-8) <6.8(-8) <6.2-7) 9sNb <1.3(-7) <1.7(-7) <1.4(-7) <8.5(-8) <3.7 -7) 103ku -<4.7(-8) <1.9(-7) <2.5(-7) <2.7(-?) ' <5. .' -8 lo6Ru <6.2(-7) <1.4(-6) <3.2(-6) <1.3(-6) <7.1 -7 110 mag <1.1(-6) <1.6 -7) <1.8(-7) <7.8(-8). <7.1 -8 124Sb ' <3.2(-7) <2.3 -7) 1.5 i 0.4(-6) <2.3(-6) <1.8 -7) 12sSb <1.8(-7) <8.5 - <7.1 -7) 1.4 0.7(-7) <8.5 -8) 140Ba <4.4(-7) <1.2 - <4.0 -7) <7.1(-7) <2.8 -

140La <5.2(-6) <1.4 - <7.1 -8) <7.8(-7) <1.8 -

141Ce <80 4 0(-8) <2.1(-7) <2.6(-7) 60 1 3.0(-8) <7.8 -

152Eu <6.2(-7) <7.1(-7) <4.4(-7) <5.4(-7) <5.5 -

154Eu <4.7(-7) <5.3(-7) <2.1(-7) <9.2(-8) <1.3 -7)

TABLE A.39 (Cont.)

VENTILATION AIRBORNE ACTIVITIES SAMPLE STATION WASTE EVAPORATOR (uci/sec) ,

o$ 1/20 - 2/3/77 2/3 - 2/15/77 131I 2.5 0.4(-6) 1.03 1 0.07(-6) 134Cs 3.5 0.7(-7) 2.1 0.4(-7) 136Cs <2.0(-8)- <5.7(-8) 137Cs 4.0 1.0(-7) 4.3 0.4(-7) 51Cr- <5.2 - <3.5(-7)

, 54Mn <4.0 - 7.0 1 2.0(-8)

.59Fe <1.3 - <1.4(-7)

g. 57Co <1.1 - <1.4(-8) ssCc 4.0 i 1.0(-7) 2.8'i0.3(-7)

- 60C0 <5.4(-7) 1.6 0.3(-7) ssZn <7.8(-8) <1.4 -7) 95Zr <1.3(-7) <1.4 -7) 95Nb <2.3(-8) <5.7 -7 lo3Ru <9.2(-8) <7.1 -

IC6Ru <1.2(-6) <6.4 -

110 mag <3.2(-8) <6.4 -

124Sb <4.0(-7) <6.4(-8) 12sSb <8.5(-8) 7.0 2.0(-8) -

140Ba <1.9(-7) <2.8(-7) 140La <1.0(-6) <2.1(-7) 141Ce <2.9(-8) <2.8(-8) 152Eu <9.2(-8) <1.4(-7) 154Eu <1.5(-6) 4.0 i 3.0(-8)

M ' g $

s'

' TABLE A.40 LVENTILATION AIRBORNE ACTIVITIES (uC1/sec)

PIPE PENETRATION ROOM Sample Period. 2/3 - 2/10/77 2/10 - 2/11/77 ,

o- .1311 7.9 ' 0.8(-6) 1.4 1 0.6(-6) 134Cs - <1.2(-7) <1.3(-6) 136Cs  :<9.9(-8) <2.6(-7) 137Cs <1.2(-7) <l.2(-6) 3H [A] [A]

j s1Cr '<9.9(-7) <4.6(-6)  ;

5 Mn - <9.9(-8) <4.2(-7) 59Fe <9.9(-7) <7.5 (-6 ) j 57Co <1.6(-7) <3.0(- 7 l 58C0 <9.9(-8) <3.2 -

60C0 1.0 0.2(-6) <5.0 - ,

65Zn <7.9 -7) <5.2 - -

95Zr ' <2.01 -7) <3.4(-7) 95Nb <9.91 -8) <2.0(-7 103Ru- <1.21 -

<3.0(-7 106RU <1.21 - <1.1 -6 110 mag ' <1. 21 -

<2.8-6) I 124Sb' <1.21 -7) <7.3 -7J 12sSb <4.01,-7) <5.2(-7;l  !

14oBa ' <5.9I -7 <5.7 'l 140La- <5.9(-7 <5.0 - l 141Ce' <4'0 -

. <8.7 - )

152Eu .< 4.0 - <1.5-6) 154Eu <1.2 - <8. 3 -6)

[A] '.Not Measured l

--147

~

a L

TABLE A.41

. VENTILATION AIRBORNE ACTIVITIES (uC1/sec)

LETDOWN HEAT EXCHANGER ROOM-

~ Sample . .

Period 2/3 - 2/11/77 2/11 - 2/17/77 131I 6.8 0.4(-6) 3.0 0.5(-6)

• 134Cs 1.0 1 0.4(-7) <1.3(-7) 136Cs <1.4(-7) <9.9(-8) 137Cs 2.7 1 0.5(-7) <3.2(-7) 3g' [A] [A]

51Cr <1.4(-6) <7. 0(-7 )

54Mn <1.4 -7) <1.3(-7) 59Fe <3.5-7) <3.0(-7) 57C0 <3.5 -8) <6.4(-8) 58C0- 2.5-i0.6(-7) <2.0(-7) 60C0 2.7 0.6(-7) <1.3 -7) 65Zn <2.8 - <2.0 -

95Zr <2.8 -

95Nb 103Ru-

<1.4 -

< 1. 4 (- 7

-<1.3-ll

<1.3 -

<1.2(-7

~106Ru <1.4 - <1.3 -

. 110 mag <1.4 - <2.5 -

124Sb <1.4 - <1.4 -

12sSb <2.8-7) <2. 8(-7 )

+

140Ba <3.5 -7) <4.8(-7) 140La . <5.7 - <1.3(-5) 141Ce .<6.4 - - <2. 5 -7) is2Eu <4.2(- <3.2 -7) 154Eu <4.2(- <3.3-7)

[A] Not Measured i

o

, 148 l ;. _- -

/

TABLE A.42-131I SPECIES DATA Sample Station #2 Fractional 131I Distribution (%)

Particulate Filter I2 H0I drganic Sample Dates: ,

9/3 - 9/9/76 ~. 2.3 30.1 ,

29.5 38.0 9/9 - 9/15/76 0.8 44.6 13.1 41.5 9/15 - 9/18/76 25.4 42.0 6.4 26.2 9/23 - 9/30/76 8.2 44.1 3.4 44.3 9/30 - 10/7/76 1.6 24.4 41.9 32.2 10/7 - 10/14/76 8.9 32.4 11.9 40.8 10/14 - 10/15/76 .7.5 48.6 20.6 23.3 10/22 -.10/28/76 9.2 29.2 15.6 46.0 10/28 - 11/4/76 20.2 25.5 12.1 42.2  !

11/4 - 11/10/76 7.2 30.2 22.2 40.3 A*<

E a

l 4

i J149

TABLE A.4 2 (cont.)

-131I SPECIES DATA Sample Station #2 Fractional 13II Distribution (%)

Particulate I Sample Dates Filter ,

2 H0I Oraanic-i

- 11/10- --

11/23/76 11.3 32.0 -23.9 32.9 11/23 -

11/30/76 0.7 41.3 1.1 56.9 12/1 -- 12/15/76 4.4 6.6 14.0 75.0

, 12/15 /76 - 1/6/77 4.5 61.3 8.6 25.6 1/6 -

1/20f77 ND 18.1 13.6 68.2 1/20 -

2/3/77 ND 33.5 66.5[A] ____

2/3- -

2/16/77 ND 23.1 ND 76.8

[A] represents HOI + organic ND - Iodine activity not detected on this sample component.

4 4

6 4

1 l

l' e

150

TABLE A.43 1311' SPECIES DATA Waste Evaporator Room Fractional 131I Distribution (%)

Particulate Filter 12 HOI Organic Sample Dates: __

4 8/27 - 9/3/76 5.3 75.0 11.5 8.1 9/3 - 9/9/76 7.0 69.7 10.2 13.1 9/9 9/15/76 2.4 1.2 80.2 16.1 9/15 19/23/76 20.1 45.6- 18.9 15.3 9/23 - 9/30/76 23.4 52.4 12.4 11.7

, 9/30 - 10/7/76 6.3 40.5 11.3 41.9 10/7 - 10/14/76 33.5 20.7 14.0 31.8 10/14'- 10/22/76 10.3 44.0 18.9 26.8 10/22 - 10/27/76 21.0- 23.7 15.8 39.5 11/4 - 11/10/76 21.5 11.6 10.8 56.0 e

I 151 l

i~

- TABLE A.43 (cont.)

131I SPECIES DATA Waste Evaporator Room Fractional 131I Distribution (%)

Particulate y2 Sample Dates Filter HOI Organic 11/11 -

11/23/76 8.4 22.7 19.8 49.1 11/23 -

11/30/76 11.5 22.2 22.2 44.0 11/30 -

12/15/76 ---- ---- ----

100 IA3 12/15 /76 - 1/6/77 ND 46.6 ND 53.4 1/6 -

1/20/77 ND ND ND ND 1/20 -

2/3/77 ND 59.6 ND 40.4 2/3 -

2/15/77 11.7 45.4 16.2 26.6

[A] Iodine activity only visible as organic due to excessive decay period prior to analysis.

ND - Iodine activity not detected on this sample component.

3 e

G 152 1

- . . - ~ - -. - =. . - - - - = . . . . - = .

TABLE A.44 1, VENTILATION. IODINE-SPECIES COMPARISONS x Fractional 1311 Distribution (%)

Sanole Date '131I conc.(uci/sec) part. filter 12 HOI ~ Orgar.ic Station #1 2/3 - 2/10/77 1.76 i 0.07(-4) '8.7 69.2- 13.8 14.5' 2/10 - 2/17/77 1. 9 ' O.1(-4)- 6.0 26.0 4.8~ 63.1 Letdown Heat 2/3 - 2/11/77 6.8 1 0.4(-6) 4.2 19.4 5.0 71.4 Exchanger Room gjjj - 2/17/77 3.0 0.5(-6). ND fG ND' 100 h Pipe Penetration -2/3 - 2/10/77 7.9 1 0.8(-6) ND 54.5 35.6 1.0 -

Room 2/10 - 2/17/77 1.4 1 0.6(-6) ND 100 ND~ ND j.

j ND - Iodine activity not detected on this sanple component.

/

-l I

TABLE A.4 5 131I SPECIES MEASUREMENTS OF PROCESS AND COVER GAS Power Operations - Before Refueling 4 Fractional 131I Species:

Total- Particulate 1311 Filter I HOI anic Decay Corrected To: (uct/cc) (%) h (%) Org(%)

Waste Gas' Decay. Tank "B" Isolated: 01:30; 8/19/76 Sampled: 11:05; 8/26/76 911:00; 8/26/76 3.0 (-7) 0 1.2 15.2 83.5 l

Sampled: 10:52; 8/31/76 '

011:00; 8/26/76 3.7 (-7) 0.5 1.2 12.3 86.0 Cover Gases: l Waste Holdup "B" Tank:

I Sampled: 15:17; 8/26/76  :

016:00; 8/26/76 4.0 (-6) 0 0.7 9.3 90.0 -l Neutralization Tank:

Sampled: 15:54; 8/26/76 917:00; 8/26/76 1.5 (-5) 0 1.4 17.2 81.4 Spent Resin Storage Tank:

Sampled: 09:05; 8727/76 -

911:00; 8/27/76 2.4 (-7) 0.5 0 43.5 56.0

Aux. Bldg. Sump Tank: .

-Sampled: .09:25; 9/1/76

~

911:00; 9/1/76- 9.0 (-6) 0 0.2 11.0 88.8 System Header:

Sampled: 09:16;-8/27/76

. 911:00; 8/27/76- '4.6 (-6) 1.1 - - 0.6 30.0 68.2 ,

154

_ - , - , .-_ ..m - r.. -

-r, - ,

~ . ._ _ ._

TABLE A.46 131I SPECIES MEASUREMENTS OF PROCESS AND COVER GAS Durina Refueling and Power Operations After Refueling Fractional Distribution (%) .

Part. I2 Sample location Date ._ Corrected to Filter HOI Organic ~ Total (uti/cc) i j Waste. Gas Decay Tank "A" 10/13/76 1457 10/13/76' 131I 0.2 0.9 18.6 80.3 2.14.t 0.03(-7)

! ' isolated 1240 10/2/76

. Waste Gas Decay Tank "D" 10/12/76 1600 10/12/76 131I ND 0.3 17.6 82.1 1.01-t0.01(-6) isolated 2104.10/11/76 134I 73.9 25.8 ND ND 1.0 .0.5 -

13sI ND ND ND 100 1.3 0.9 - -

Waste Holdup Tank A 2/10/77 1115 2/10/77 131I ND 0.1' 13.6 86.2 1.84.1'0.02(-7) 88.3 5.6 0.5(-9)'

y 30% full' 133I ND ND 11.7 Neutralization Tank 2/10/77 1245 2/10/77 131I 0.07 ND 8.5 91.7 1.72 0.03(-6) '

50% full 133I ND ND 2.9 97.1 8 1(-8)

Spent Resin Storage Tank 2/10/77 1400 2/10/77 1311 ND 24.1 ND 75.9 3.8 i 0.5(-9) 80% fu11 133I ND 100 ND ND 3 1- .

las! ND 100 ND ND 7 5-

. Vent Collection Header 2/11/77 0850 2/11/77 131I ND 2.5 25.0 '72.5 2.4 0.1 -

133I ND ND ND 100 2.5 0.6 -

Auxiliary Bld. Sump Tank 2/11/77 1000 2/11/77 1311 ND 0.5 8.9 90.'7 7.1 0.1(-7) 133I ND 0.7 11.9' 87.4 4.0 0.1(-8; Reactor Coolant Drain 2/11/77 1057 2/11/77 131I 0.1 2.5 28.3 69.0 8.11 0.08(-7)

Tank 133I 0.1 3.0 27.7. 69.1 1.78 0.04(-7) las! ND NO 39.6 60.3' 1.2 1 0.4(-8)

ND - iodine activit'< not detected on this sample component.

TABLE A.47 WASTE GAS DECAY TANK "B" ACTIVITIES Tank Isolated 15il5 10/4/76 - Results decay corrected to 17:00 10/13/76 Iodine Species Sample Taken 16:12 10/13/76 Fractional 131I Distribution (%) -

Particulate I2 Filter _ __

HOI Organic Total (uCi/cc) 131I 0.08 1.9 26.3 71.7 3.86 i 0.06(-7) 250 cc gas bomb taken 17:15 10/13/76 131mXe 3.8 1 0.03(-2)

I?3Xe 1.04 i 0.05(0) 133mXe 1.57 0.05(-3)

, asKr 1.38 0.03(-1) l 9

156

TABLE A.48 WASTE GAS DECAY TANK "A" ACTIVITIES Tank isolated 17:15 1/20/77 - Results decay corrected to 09:26 2/10/77 Iodine Species Sample Taken 08:41 2/10/77 I* Fractional 131I Distribution (%)

Particulate Filter I2 HOI Organic Total fuci/cc) 1311 ND ND 1.8 98.2 3.98 0.08(-8) 250 cc gas bmib taken 09:04 2/10/77 131mXe 1.34 0.07(-3) 133Xe 5.49 0.05(-3) esKr 1.34 0.03(-2)

ND - Iodine activity not detected on this sample co@onent.

9 9

h 157.

2

TABLE A.49 PROCESS GAS 14C AND 3H TRITIlM MEASUREMENTS During Refueling Waste Gas Decay Tank _"A" Isolated: 12:46; 10/2/76 Total 14C Fraction Not Sample Time (uC1/cc) OxidirM (%) HTO (uCf/cc) HT or RT (uCi/cc)

-1506; 10/17/76. 7.29 i 0.01(-4) 89 9.63 0.02(-6) 3.86 0.01(-4) ,

t j

t i

i e

l l

! 158

\ +k 4> ) IMAGE EVALUATION

// q < #

NNNN TEST TARGET (MT-3) 1.0 ilenua

@ IE iu L* llllE I.8 1.25 1.4 1.6 l

e

MICROCOPY RESOLUTION TEST CHART

1. 4 4%

+$fAf #3+h4 4g h,/// 9

//

W*R>

\

IMAGE EVALUATION

$<#A '%

TEST TARGET (MT-3) 1.0 l9E4ILM 582 EE I.l $l,EN&

1.8 1.25 IA j i.6

= 6" - - - - -

=

MICROCOPY RESOLUlit N "E , :: HART

1. 4 ' # 4'4%

• %fh/

b,i 43<:Q

/

TABLE A.50_

PROCESS AND COVER GAS 14C AND 3H ACTIVITIES Power Operations - After Refueling Fraction Not Sample 1ccation - Date Total 14C (uCi/cc) Oxidized (%) HTO (uCi/cc). HT or RT (uCi/cc)

Waste Holdup Tank A 2/10/77 7.40 0.05(-6) 90 3 1(-8) <2(-8)

Neutralization Tank 2/10/77 2.68 0.02(-5) 97 1.1 0.2(-7) <2(-8)

Spent Resin Storage 2/10/77 8.22 0.08(-6) 65 <2(-8) <2(-8)

Vent Collection He,Tder 2/11/77 8.7 0.4(-7) 89 <2(-8) <2(-8)

E 2.65 <2(-8) <2(-8)

Reactor Coolant Lrain Tank 2/11/77 0.06(-6) 91 Waste Gas Decsy Tank A 2/11/77 2.54 0.01(-4) 92 <2(-8) <2(-8)

TABLE A.51 CONTAINMENT ACTIVITIES 1

Fraction Distribution (%)

Date Particulate Filter I2 HOI Organic Total (uci/ce) 10/12/76 1311- [A] 23.5 22.4 54.1 4.68 0.07(-8)[A]

2/16/77- 131I 13.9 41.0 4.8 40.3 5.46 0.05(-11) 2/18/77 133I 35.6 56.0 8.4 ND 1.4 0.5(-11) 58C0 4 1(-14) 60Co 2.7 0.6(-14)

$ 134Cs 2.29 0.06(-12) 137Cs 2.94 0.09(-12) 54Mn 1.0 0.5(-14) 3H as HTO 2.00 0.01(-8) 3H as HT or RT 2.24 0.01(-8) 14C oxidized 5.60 t 0.01(-8) 14C unoxidized [B]

[A] filter not analyzed; assumes no activity on filter.

[B] sample lost during analysis ND - iodine activity not detected on this sample component.

. . o .

Figure A.1 Reactor Power Level: Fort Calhoun Station Period: 8/15 - 10/3/76 I I I I I I j : Pnmary Coolant Samples Taken 51500 - -

1 a., , _ ,,ye-e --- : e - e- e--ee o

a j k

e e T **

7* r-e r '* * ~ * ~ ***'e _ e _-

e a e f E \

e soo _ ,

/ -

- .e  ;

m l S

.\- . - .l 8/15 8/17 8/19 8/23 8/27 8/31 9/4 9/8 1500 - -

g -

" '*-e-e-e-6-o-o-o.e.e ,.,

• a f -e.*

- e 32000 s __

a E

• e

• %e- - o -o.e -o e - e - ..e-e, 500 -

I -

o-e - e - o

!  ! I J' I  ! I 0

9/9 9/13 9/17 9/21 9/25 9/29 10/3

. Date in 1976 ML-A-ts2o

(

N 1:

v l .,

Figure A.2 Reactor Power Level: ' Fort Calhoun Station i :L;

, Period: 12/15/76-2/2/77

-i 1500 , , , , . ., ,

-2 w -

.l l1000 - -

4 g' 500, -

E .0 ' ' ' ' f '

12/15 12/19 12/23 12/27 12/31 1/4 1/8

,.g 2000 , , , , , ,

I E1500 -

i -g1000 - -

8 Q 500 -

O ' ' I I ' 8 1/9 1/13 - 1/17 1/21 1/25 1/29 2/2 2/6 Date in 1976 or 1977 INEL-A-2529 8.

9 h

i' e ..

• 4 e

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

• ' ,- S .. ..

M

, Figure A.2 (cont.) t Reactor Power Level: Fort Calhoun Station Period: 2/3 - 24/77 2000 , , , , , ,

+

O

. j1500 -

. , _ , i_ _f s a y '

O $1000 -

IO O: Reactor Coolant Sample Taken S ym K

' I I I i 0 -

2/3 2/7 2/11 2/15 2/19 2/23 2/27 Date in 1977 INEL-A-2525 s

D E

t 6

i b

l '

I I

Figure A 3

--0--4i h .

<<.-- ~ . - -

?

. q . q' ~,. g .4' e=a e "A ye , ,

~% w ,-l ;54 >- @ * - n . __dz ' r. ere*

t yy n g

- ~< ~ ----

n,.y---

,a.,,.

3 ...........-

_' .*..~*

<* -e a..

, .x y ......... .....,

,AD..

g :4.Q

>,r. ,,n

.m.9 i , . w ,.

--e-..

l "' ~ . . . . .

r*ro ,,.a. ~~ , .~- . s.

I _ .

. . . . . ....J  ! ##* "h , p ,

wm: ., , ...,x v I PE-i L

~61 l

.. g PH" " L .. $

4e g. g.. . ". J s 1 ', .-

. . ,.r ,.

_-'_*,m w -

@t l@-  : ,. ' & + 4 . .. .

' ,, ; . 7 s i .-

u ' '~~

%7W~m Q-- l J., e. .bQ;zr hw . +.,  :

~,., .7 e '

n, , ,

,, c:

i c.-~~.- .... ----- . . .... _ _._:___,, ,:

I e , e.. 6 e s;

[, , L... ,0 .....

y ,

,.... , -,i.i .:

o n- , , ;, '

' ; [. . . . .! . ', . l. y.' *

..L i r

3. 's

..n , u

.1s *. e

h.

@ a @ @.

' a me e*w s

( /asy f ow? 4.e * , g g , [es W 8'8P'WE S I O Sa

,F,,.,,.nent y , .ee,np . gee ', 8, ,m(.rra*F,F ,

ff'.% er F .M TF,) g ' '

qe. 3 B

.....- ....... ..s..n~.. ,m g . .$

• e J

. ,.. , . ,, , e i ' ' ~ * * ' '

l 8a= 5."2si am 4 %g cam'

• i<

y n

, n,, ..._.. ,

'"N' 1"me- -l ;;o=,... . ... :'~-. .

. . os a

e 0.E 9

II

.E.. . ..,

,'O'.....

3

?

i i ,,u

@@@@G' ....J l

. i , ,

i u ~, , _ , . - ,,

( ) ) )

w l l g

.,,i 7 .-

7 isA W '-. **. anae -A = $  %.'T. . ,a qq l

a v M D 1 7,[.", Z,,,,

. ,s f AEM7<.2e

,es m <

< L:t f

e a e T" ILoJP ta

_.m..,

l

• I ' =* * * * > ppqM@**4 & st pro a

,. ,e , i.u n s~ gjo w e

( ,m. .a.d . .,. _ ,4 ,;g,,2

, . ~ ~ . . ,

, i. e's l  ; .

. . . 2: . . - f[ . . O I. iM-N- Ni! - E. 2.-If-s84-h'-i . _ .Y-Y I '-[i.':Y T L.. .. . .. 9. . . t E.:. ..:.: 5. . .i.?.!. ./.; i' .=/. . . . : .:.75. . . .:.;=. ..:. .: : .t.:5 5_3.;. .: .:n. . .:. N.,;-i.', .... . . . .' ..r,. r J

. e499,99 999ttete

,97.,,.a

. . . ... IT,T 1 s.q.- [-

. . . . .. n n .a ,

v-ilse d. .486*4* i }

D 400 h..- th y I.

I

• f.

b i

p . s ,, .

.> s-

~

h~ --L DL- , . .

LA-l i Q U E. M 9 i,j hu C t

. e- - >

~

%HMh* g

^

0M <M_7.~' ,

CruuL- 'l a'  ? '.-. - q j DUM" 3

I

.g' f. . . g-r-~

9-@

. v<r,,  ! ,

4ep'. - -,= g

-t% ~ rzn

-~ -- he' n21 EQ

^

A." y, .- ,

% Y

(' s. v.r?

1 ., . 1 a c. a, c. . .

vs

-s_-_~w w r, r, r r, n n,e, -

a P9f5 5aMgfd orvr-aw , -

..<m...

v y ' .' ~s <~ ~ e,'km ,,. ...., .

s -@ . ~ - - ~ . .

~,;;<;~ ~

,,,.,- .n a

- & ---;, *4 415<?;.~

m- x-q

• 3 n,,, [" ,...-,,,,,.a.

3, ;,.,,;~-

c - s-v"

. , w~ a..,...,.-...

72%, ne (spact r- s, s rw

• , ys, g3 ot****** o .ow see. s,o.s. s ser Q 1.1.n2

• 2 roc.

• s-r

,4, . ms l y s.-a.,.,f , ,- . ,,. n .<.n ..u.

e as w.- . *= .~.ar

.- . o. . .

t 4* ..ch ; ' .p g

-.....e

, es p

V . t' *,,,

t .

g,f .' .r=~,.u~,ap n, .

, , 'r------ W r%W '

to ' ,e n o. 9 % $ _ f c'

() r ~** , ra: rw e ysepy ,,. ,

V'an ccm Ed n^ &oN3

, :o: :

..D mm

<m ~

}A g: &

a*=< ,

~:,r 6,.,_',,.,,,.

. , ,, jef,

(* '% , j . .

d @f

' ****'"s, gY ,~ ; <p; I 5.-~".*A "CX, 1- I

{

.M - ls l,h b b b N

, ,,,.- t i Pfi A.1".2-j .,, GEstearor t , . ,._) , e ,

og e- ---.

fbi ;I

,f g.

I  ;;y,'  ;. i (  ;

l 3

@g. 6> d

.i; i w. .. ..

@ $_  : . 1 , .A

%n _s .,~~-='=~wT- e V /*

' I '

h ,.4 M h) h _- ...... . ,..

]

a i T M 4.A/ .

j j j gge t g g , g

• TA(({A

'h*. w ,.y

' ' ~ ~

. t.  ; .

I

~~_-W.

• z ^-i s:';L ~-l_5 :.:

• . _a..__

l lI T "' i. :. . 5_ -. :_: ....

.. ' -l-:&. .f. : ~ . -.~.-. - .*:. 3 ~ . .. i. _i . E :: ,= = , .=_J._ _: '__ . ._)_ _ . . .

May 1971

@ ,k Numbers to Sample system:

Supplement No,. 8 Reactor Coolant System Omaha Public Power District Figure Pand I Diagram fort Calhoun Station Unit No.1 4.31 l-164 k

l' Finure A g.. .......___-- .

......_7 . ...._......... ____. . -Q1

! c4 ___1._q________.__._____....__...*4 i !__  : :

a .=.#r _, t i d,

$- ~ 4 .

2

, i! : 0p e

+f-, L . . .c e 7 3M _ g.._d.

NM7fxc m f4 3 .._ s 7~~~~~ r------ --

s@ r'<- --

p._ .. .. J , <p

<5; ts--s

/_ l c.

,' 1

_a -..--,

, ~ . - .

...y aI

&} /* h - ----- &

r9 e a an , . . .

ww.m~ ~ ~ ~ .

- ~~

c.3

.s.,n ,_. __.__........

6.

.... .a 3_

-3 $ - ,,,.,_, -_ ~

~. --_, f:

T-.-C,D ss. w. g._i__3 92@  ;!

R {&-e .. _4 9D

-> ._ ~'.- }m.'4 ll o

i

h. h , ,' , y- 1 th, ,

-p-- -* b~""_

\ Adi- 6.-++' 4

. 4-

-- h 4--- z.= g J

.:f(Y-

. -. *sgw t'nA'm't .

N."=12 J ~ &

es;g=n .=nu

,g ,1' _=-*

g g!@i:

,r-~ - t

• :%> \- wu eux_>,: , ~~~ Q:,

m

{ -.

x

-) 4

~<-.,e w- *, .,

m.

gI [ kr4Fi@feev6ft ., I s

~

I I- , ,

: H -- ) I ."

• 6 {di < :

y + tt =

2h V" . .m  :~ $, - d t-lh

+, _

~y #, ; ,81 ei pm

- ~g' _ -

, .' <--e h

"{r.g h:.

a (h, y

~

l ,- t,"A em?

@-c )' m J.Cl W S 3s l

• n 3 kgiFL. t" ~._

. gg A**' /\

l .- - g</~ , _k \ , y, s.

1 s--m  !

1 l

+w >

g-. ) *- /*mes 1#

4,2;; ~ 2 ~~#

15 #

1 J ,

4 4E. ee w

~6 " l W. s . .w , . .,' '*?'

sm-x/*

tm, y-

" Q: -

, _r-sm.1_3"*.'ehrm/" . -

- jTt i, 4 !'-9 41 g,D h-i .

<- e u .3 r

,q

& "e

.'. l A' . _

Y

,7,

, i Q.4 *

• g* ' *~r.- .. ., y, t~~- .. ,

t, - . _ ......~

. ..-s .,0._

i._~me-. /i i-.mo - m._ mi i e r -- !d,

~~

.W lo;- J f. W ~,F*'**

h, e

9 A s1.= r- .

)e ~ +*.

s  : r~t _ _ _ . _ _ _ _

vm -

,.s

-s I

,[, ,

f

,! , _, 7'}'* r h{ ..r iJ b f f,,_ f,Nb f pp g

- t m 3

v. ,i. r .,

%-w.m 3 g-v,t. , .p u en.-)  :, 4. gw 3 g 'g s

p d' 't

'q

.a  %{-.tw4-Ch-M >

t.u %r.< , - . < , -

.4 y . 3y,, ,, < .

,f _ , < . . . g j'p-.t^g g

y. , 1 f \' - * '.o.

g .

W.,

},' f*/ .se  ! ;N  : c

,< i e<4. *c t- T

• - . I p-d * .,.

6 4- ..

k .* (  ! g t ,' 3 j'. 4,, ,q

}(

v

.z P

P 'u_ Qj' t

• ""g.p y .. U. . . ps '

}(pp  %. (,/g

• m, ~ =.** ~-

\

1 r~ $-%y i j .j-G**~ ~ ZTJ I~4- .

>>-~

! 41 r,,. . . ~!;-H@@ 7hy y w$p,, T; h. _.h ,jhNy e l , ..s f,,-

3 r- ,

l <" ~- -- I$4 4

$.iN % MOM 4 I dl..;

e a .- a .

.g

2. x r

, I s ,. -

1. 3 a""" i # 2

• M, gr. ,'.,C'**" .. ,,,

--: [-

Y' I$

l.*,'.,

n!'M C" ,<-~~

h~ .-7 f.=p}1974F.T..'i$tr~

j" ~~~

^

- ?"A *U-@ -

t

.- Q ,

+*-=** 8"9 (*'

"D ,

s....,.,

~

< V L 'ff) @ /.* '#' g K',

_ _ . . . . ' - },0 d, I b .. . e, ' **

m

~l(

t g )< rg [

$ $4 Q ~' &=..=I'

^ ~ ~ ~

ey,'* . "~9(G u m f ~yf'.7 ' ' S Q ,

.f?R erre Q .Q n

p . ..=w-~ ' , .:,. ,,;-

?> a. -

(tps) g s

E CW-f.- ia6g ,,,cm m - , ,

g g"-g,k. l3.s

1. +-

ps,ede sal i 9 e pg r#

h..-

ao ga..a. -

.,g----

.- -2a.- , Q' 4t- A* P 4 e ~6=en u r +  ;

l'6 cl M.~~_ -

, s.am e,,

rr . .

.t e -

Q r .. ~ '

%'W

?.'" .' : % "3*N.*5*X I e  % {$

, .hp e " " ~

i... y _ v. D < r# ! cwat ,a

('4r9806'8 1 g e M y/

3.

, +.ie v g .. , l dat <-

i,)3 r h.,t, D I E" e

e J v .. '. s

'e T *** *W [**

'ete#e 9 j] k i ; 4 0--- O

• q

-O,a.e--==.a.= === a * *(N'

• T , (=irap i( h "e

p] i**$6'*.f fW

<.E. . . mi E 7f"' m. '

mD_ '*'

,.r.-. ,,s.,. * : Indicates Points to Sample System.

'* ~.;s.~ :'7, : Ti: l.' ~ ~ ' * * : Local Sample Points

*:.;'L*": :.'.~.:T'X::".::=~.,*.

' *""ll*f.f *r;"" "- ~ ~ April 1972 Supplement No. 16 i 165 Chemical and Volume Control System Omaha Public Power District Figure ,

P and I Diagram - Sheet 1 fort Calhoun Station-Unit No.1 93 j s l

r

<,! t ,)l

.! c !

' 'i'

'.'a2 $

a5 o< g5

- - ~

~ nixo51

- m _

u~ ~

• A 8

_ e H ic C v

r e

S in r

e iz L

9 la r

0 e _

in _

m e .

D d -

e .

B 5 d E -

. .M mM E -

e . +

x i 8 M 8 H

C

- ~ - -

s -9 5 " $ - t N4a> $~

_Ae I,e~

Oh- -

~

11 1l llllll lI1l1llll1ll- l 111 l l l1l ll,llllllllll llll

'~

• p. .

i Figure A.6 1

' Letdown Flow Rate: Power Operations- Before Refueling Period: 8/15 - 9/9/76 i

1 ..

ue-140 Il: Samples Taken no -

100 -

T qy .

4 .,

e a

.~ a s* - -

I }, M, 8

.o - .

s J

.o -

20

o. A._ _._,--.-.-m e

.d a m._._ q : e <%g a

1 I I e , ,

0 1 1 8/15 8/17 8/19 8/23 8/27 8/31 g/4 9/8 Date in 1976 mn-usa i

4-i i

4

Fi9ure A.7 CVCS LETDOWN PARAMETERS

- Period: 8/9 - 9/3/76 15 , , , , , , , , , i i i Reactor coolant boron levels not available, but less than 100 ppm for the period. '>

10 - _

p I Reactor coolant pH meter stopped working

. 5 - -

o 8 0 't i e i e i t t i e i f30 E

1 , , , , , , , , , , , ,

Ey!EO5 3 Reactor coolant conductivity meter stopped working 8 0"' ' I i 8 I 8 I ' I 8 I 5 M DCH'8A and B in series'

' CH-8A dnd B in s'eriesd

.$ _ 10 eDeminer*Iizer CH-8A in servicer;: Demineralizer CH-8B in service  :

3 G g Arrow indicates g

j {g 8 -

$ samples taken -

Eo

@D 6 - -

c2 0

ma 8 4 - -

4 -$2 -

F R U -

' I I I f ' ' '

0' I 8 2.5 i , , i i i ,

Filter CH-178 in service U k 2.0 _

E c g 1.5 - -

p Ea *2 1.0 -

N5 Y E $ 05 - -

1 I I I I I I I 8/9 8/10 8/12 8/14 8/16 8/18 8/20 8/22 8/24 8/26 8/28 8/30 9/1 9/3 Date in 1976 INEL-A-6445 e e *a s

, .. .. .,. 7 ,

'h' 7:

3 . ..

i^ .

1 h.

1

~ ,

i 4' ,

t '

Figure A.8 i'

Letdown Flow Mate: During Refueling i Period 9/25 - 10/19/76

}. 100 i i i i i i g O: Sample Taken .

] g g 80 - -

, e. -

2 -

+

y60 -

E o .

E 40 -

l c ,,,; - ; c.

B

. 020 -

4

I 8 i ~

0 ' ' '" ' - ' - ' -

9/25 9/29 10/3 10/7 10/11' 10/15- 10/19 10/23 I' Date in 1976 INEL-A-2528 9

• I.

• r i

i j i

~

1 t-f

.-.o__

, i i i!

1.

il g 67 2 n 2 5 i _ - _ -

2/ 2-l A-e

_. u L E

. f N e I R _ b r 4

, 2 e _

' /

2 t

f  : ,

. A s

n 0 o ,

_ ' 2 i / _

t 2 a

r e

p

,O _

7 r7 6 e/ ,

' 1 9 ' w4 2 4 /

2 _

o/ -

A P2 _

e r  : 1 u e3 t / -

g a1 2 i

F Rd ,

- ' 1

/

wio r 2

lo e -

FP n

w o -

.d

.t ,

8

/

e 2 L n e -

k a -

T e -

lp

, m .

4

/

a 2 S .

O 1

- - - - 3

/

0 0 0 0 01 0

1 8 M 4 2 O In8 Oj aoE j$*2 d

aM l

, . o ,

a Figure A.10 .

CVCS LETDOWN PARAMETERS Period: 1/31 - 2/24/77 k -

1 I .I I I I I I i i i i 3 Reactor coolant pH and k8 550 conductivity meters not functionat-o3- #_ during this period -

8 y 500 -

gg i Arrow indicates y ,

g g 450 t samples taken -

28 e co400 -

I I I I I I I I I I I I g $0 MO 2 10 i i i i i i i i i i i i e To Demineralizer CH-8A in service during this period d

gE 8 -

CH-8A and -

i@ CH-10 in series em 6 - i ou i 5$4 EN yi 2 ., __ ,_

- -d -

S o e i i i i i i e i i i i 2.5 . i -i i i i a i i i i i ey Filter CH-17A in service during this period 2 g 2.0 - -

=a E o 1.5 -

o , 1.0 w . ,

5$

g { 0.5 - -

i i i i i i i i e i i o

1/31 2/2 2/4 2/6 2/8 2/10 2/12 2/14 2/16  ?>18 2/20 2/22 2/24 m a Date in 1977 w.---___-________-- . _ - _ - _ _ _ _ _ . - _ - - - _ _- - . - _ _ _ _ _ _ _ _ _

j O

' O g

n i

l e

u f

e R

r e 8 2 3 1

t - - - - - - / 5 f 2 2 -

A A-

- L E

s _

N I

n _

4 o ,

i ' /2 t 1 2 a _

r -

e7 =

p7 / -

O4 2

'r/

e2

, ~

' /

0 2

w-01 2

- P /3 1

sd eo I. 6 tar i ,

' 1 1

ReP -

/ 7 27 1 n 9

- w 1 A o n d i e w . e r lo 2 a t

u B

' 1 D

g t - /

i d 2

- F n a

C s

w lo C 6 F

o t

a r

e n

r d

o ir P

e M -

' 8 2

/

/

g e

G e lp m

a

,m ' 4

/

2 e a t

S S

N 1

- - - - - - 3 0 /

0 o0 0 0 0 0 0 0 O1 4 2' 0 0 2 0 0 0 0 0 0 4 2 4 2

~ -

- [ B y U c E 'i ; U a 7 r b E N 0 0 E g 6 N O 5

g9 E Sm ySg gsw c$ E@5 $gEEG e$ g5 e$u$

• _ ~

_ N 4

I _I a I a i e i o 't M.

i Figure A.1:

1 4' M . o 9e M

e .. "M==----- 4 ci) gm.a rip i.e

" at be us Safety injection and s m y .w q

< ; ""* e.a 9,, .

f 3 Refuehng WSter Tank **'a'M *W t *i S Y 2 '~' 'f <* + }p 1

q -- -.h ~ s>

u l

. . a. _- _. .

of4 jaj <! =. .:

s,. 'l ==:3.=T4'*"*

C.*r  %

seewT evet root ,..  !, . . u. .si

• • .;4*A. 3

. L *, L,p ~ 'a "" ** ,. A

_ q , .-..., -

5; y ..

J 5: .w. -

• "' W a.. . . . .

J) :: .6 Ec

.a 2 -

_ss -

3.

I

\.. .,.

%j } J = a= A[f ==. .

. YEs*e*

u ,

h

,j P

b* 3.....

• • • 8**

2~ ,f

, oN_'u

. .. o _ . s 96sweeu.a poet I

CIRC. PW M P S en.enM 1

e.

• I a

.u....

ess N -WP gn.et 64 TO .F.asT F956 Peek 4

=w..vT= ,

asewT ea Asg Tegne %st go t a m n..s.m-. l

.g . awgst .

• < =

MN g .

nm y  ;

+ m...'

"3 E . , . .

AEs Se8 ',1 l AS*564

.) I e

"4 ,,, . . .;,, , ,,

,e ..; n w,o * *-' s, at++ a + -

~..

I u.n .r,. _ ._ _ _ _ - m m, v. <

9 tp L POOL.

l 4

I ~

l a i a l c l 0 4 v

)

l l

e l ,

l a  ; , ;_ )

P l

3 ar- 79t

\iFE "n..'.

Y 9 (" " Qi &

a. ... n _. :

s,e 2 1 s- .-

c .

es e e As 'spo 1

. L.1 - m,. -

n,.

eulIr d ee- a*

~

az. .e n as-ass 4

][ q

- -\

I h.h2 k?*;CJ

. _" '.I edmetat etas*g-1 reg 1te sfase e A4r6 H..se s w

-fEriC 9 fJ b

• • As 9 M

f. $. .=- .*

b

. .... ,l; .

l'4 asia 9e 3 as .

OTOGAd.9 POOL y

6d

,A,T. t s(t4Ame.49st i

4 i

[

i i

L.

y 173 Fom SYMeest see Fig. g g.g OesAMA Pusuc PowtR DisTatCT 0 FORT CAU60UN STATIDet upsT Isa 3 wonaenwnnoiaos A u s .s... , c..s. ,

SPtut FutL PCOL CoouNG T l' I si. w

• c.-*mT~

(De. 53 ta - i SYSTE b4 Flow DI AGEAM

• W ar . h ,

,2BEN 3 W.ma?)M_

• T M.

peta=w 's'"a"c*T +n  %

.ty .,

.- =

s-.__--_.-_ -_e.mn .

- m. m.-

.,u_. _ - e_u ... -- - . _

,,..u.,,

e e l .o l M FIG.9.6-1

I a

I -1 i e i e I e t a

Figure A.13 n e.o .. f , , ..o *g y . ... r.. *'

l /

,.v.e.

7 . .~. . . .

q..i..

O..

3 e

W.f**Y U*Y-k ..o,...

w w 4..

w

.,o w

4 , o, _

w 1

STRt-S%Ik14 TUNNEL SUMP Pus 4PS cro sams we 3

..m e.e,m ,s E6 o.S o i- - <,. ,,o ,,o . . . .. . o ...y. -

y y - 'rf W V' **7*;'.****"'*

L  ?

.=.4a se e .. . - ..- n-.aW .

• • .e.s ,.

~. i) '

t A

m..e. ..

ea. m.6...6.amas .e.c.a.o . . .. 4 ;g

h. hj.r.~ ..I,s

.._-_3 w-

~~ '

( g en...wei..e l % * '

enacte.

6oo* De. c.oosa.sv

=s (es) *"8 il

• Q"'f ,, Il

• ew....ia game ees H r, tg r.v. = .me o c.... , .... H. WJm e, l i . 8.ea -ee- .l / .*w...... g)

L ,Y

• r.......

. ^

...- .E4.tM 4 e ,o.e sa r w.. s e . = a 6

," ,,s es wo. v

< # e. .' r ed

&a.E u s.t.as, et . . . . ,.. n .e -

w '"w w F W . . , ** !'* 1 l

.... e ..  :.

.. . u. J ,

'k ., . .

mervat

.o eese- epe. a.a.s.

888 C Oeu '4eBetfE88, e s i a

--,c D ~D -

9 E k .%'4'7 _ .* . o Y_ . T'

,n...,....,

& S

--.....o..

CONTheseWENT.v.w se an.o k

i l i a i . I c l o 4

(. --

e I e i e i M l_

. a ,y,,y ty ann,em,4 ,

s. 37. + .....v.-

W.a 3 ..* f_*- r1 n *=v*.me-T

" '"' .. .a i.ut'.'.

,ag 1rgz.,..

l .e. -

p -

a .e - ass. .w. .._r*.@..,,

w.s I

g.[,~~f 7 i **"**- I , @ Qrr%e.Ie*."*. , v.y

• J" y Dw

• li

@* . e - *....* ,r asing . * **. / .

n- e c G-*++-

40.*:';4".>

..V#R,,. .. p;- . \ .>; /.*h l~ g p .:Lg s rig in-6 **'

c j *. - j l( '* M  ; _

.qgg c a s.c .C' _

. .. ' 3 1 $8 ... Ft.~ .g't-.- ~{'

r . 8 **" aaaa ==* i ** ""

h g t36,tr it ,g 9

W r--#F *'gG'n ,_2 ' '

• 4 -i

..M)e. j l A*m. ,

r..i ,- J.*4.'.,,3 1

.. J 4' gp p een

,****t e ;4 *- {,tg i .-- ,, , .. a gn,: ;.

w .. . "t - .t . . ,- , , , . -

.D e..

yg=p==  % -

/  ;,

g

.- ,\ p j

.,a

.a .... [,. ~ c...7..c: _

+

• e ..*S'..****<

h

,M N

{ u/q w ' ' * , "

.p;;;~

M_. * $.~e 1'

. +

,,,.. (.. ) , ,. . .

, .- + . . , _ , , , , ,

o @- g, c. + r-. ... .- ,

c t r. I ... .~ .

_.-. r-.... r t. T c .. ..,,.m.

. .. . ..a

-):$

+g '

, d . t i

g1 4i "'"+ *"**

< , . . . . *.J ~* ~ '~' *")

• p1 . ~ - p- *

., -- mgy w . . ~ .

I i

g r .e_t .-,,..,,,,,,,r .,.,. A.

v ,---. J'

, l l I_

.~.4+- ,r.r.t ,'"* ---.M. a.we m. . FIG H l-.

. .-J (q r . o. .

.r e,._ x. ......,'

v...., +I...,

-m -- -- -

="%* .

-I l =2 . , ,, .. , . , ,. . .

r- .w ...i

$a . . . . . . w.pg.::lllry;~~.

~d- -

• ".r..

-T

-w.A...**a ....

,M, C

• s FIG.ll 1 2 i v.. -

i \'gr . . 73 ,3,; +..-=.

I '

f a 4.ma

_s . ..

.. }... .. .. . . . _ . .. .. -.

g ..D-

- # vi

.I a. hr \.4% WAST E DtSPOSAL SYSTW NDS) n w.- / "*"' **

,,. ,,.....w-w...

r4 n wn J.Y"..

n s. u t.*.'.v.w . . ..o a ..o4 o.o. . %

P4 q

=, rig u l-6

.u....a".....-=,=~".a..

a

~

o...n...... .........~..

'g'"g':;: o

• * ";"n  :.c' = " " ' '

g::b* .. .. . 8)

.. ="~....k,.,,,...,

? ... ."..i Q Sr - n . . ;;;,yt.;.*,:.;'t,;;;c.L'.."

h. .

.......x...~ .s i

e .. ..,.u ,... . .. .. o c..=.v -n .=...e

...v.

+ oo r m -o

. .u ..u.....

,rg. .

e. . ,,,.,,,,,,,o.,,,..,.oo,,

.. ...o.......

. . . . . .. . . .. .,. t .e.o , r-i,ro)

.. ... . - .. u.. .....

1

o. . ..~ e v -.. ~ . a we M. .. .*.*.n. . M. .a*J:7.f 't '7 . . .. . .. . ... . , .,,, . ,',*.T^'. . . .'INa.'. ~ + .i .a. .u.e +- e 4 J,i -

...t...a ~.~.~~. a.-

- e .. . . ... . ...... ..

. - ....,...... .. .... e... .....

... r

+-b ~ . . . . . . . .. . . . . . . . . .

. . . .. - noo. ..... ..... -

4 f

g, y fa

.gg p.--_

o ..~.. . .. n .

• * - . -, u. <=~. . .a aw 4 .

-a;;; .. ..v..u..

m . . . . . . ...r..v......

v..m . . .

,,, -ur_r. - . ..,. .aT-

. t .

,,,,,,,..,,,,,,, .. e . . ~ ., . . . ..u e e.-

-a7. , . . . , , . ...,.

.-~ a o

~a. .e v -tu.tv. ,,,. .. ...

v - .. v.s.v. ... m ,.., ,

L' *a. ...v..,......

e i . . r . .. . . 6 <.u U ."#s' 174 F. E .T=.. 66 . .. P.g. 9. i .1 3

I . 000AMA FOIrf CAUPU.SUC POWERI MTillCT 0Ulf STATIDIIipsiIEE O l

wass cessa na iaal

.o - "i a i..'u n. m . f . *f'** WASTE DISPOSAL SYSTEM

=

yyhk i .

lgg ;*T.',"$

--w w FLOW DB AGRAM a I

.t m. I==.e,um..aie.'.n _ .

_ _ __.m . _

a.as. u . .

. i e I a i M FIG.ll.H l l i <

I 4 g

.I I e i e i e t

Figure A

, r....

E - ~*

e.

-.4 . . . . . . , .

@ c w@

• o .r

)

r-A

.m

.. . n es,

,,wu

._ r ,i ea

....g .

~ '

-:@ l 4 , t r-i

,, J i

r

.;MS *1.. .}.. . ..

-=* a -i*  : . . . . - -~

m. y. ..

} 'q, -i -

.~.g

.. . g _

-.- .,E , =. .. R. x.re .. . y . ,,....... y y, g.; ,-a rw...... t t < .,,

r r,G61:-5

. :. .~. W= = G ..= .s  :=a;~; .*;;*** l

.... 1 1 t e

. w . ..#.

- . . .. . . . . w '*' O l .

.., \

ccu,w.asur - . .!

aa evemas ,, -

rio n.i-4

~ y , .n .

y FIG 11 l- =*

... ... ..q.g > m..=w~ gp .

.m .

ma.'**** ~ . , _ ..s 4 ric n i.s .-

..  : . we....= 1 I...

• .+

.. r i o. n. ...- , . - . . + . )

4

~. .

. .~-/(*"a

~

. 1

, _.L w

< E Ig... -.... 7

, ... _Vm.3.

.t. ..

e

_) -i; - l m s- +

g_". *. . .*.. .".. g e. . . . . ,

. .g g. ...

g

, ... /c

• 1 y pi .. .. g, 7 - . .. . . , . ai 7, , . .

,A . < a . <

j

+ 5 ..

.......a  ;  ;  ;  ; ,

ep Se g Yi g;" .i. @@

,,, ,.. .,.\.\\\i i @@

64%. ,. . ,. . .

T

}  ; -) e i

3 I l L%- . -

- w. ... .

1 a.vuW .D 6..s.v

. m.oo.d

1. Av. 6..e. . .15.0... . Auu tsamade.sk.o..e M,,C P . b ak M. D,u.ti .@% .4 S.

ETE. i.,.y$.M W .h

. 2. .gseg .f .dJ. *.oO.L egA .4 C .S, 6. o .b.. a h c.a.s a. . s.e e.

l e

i 1

/

j ~

\

4 l . I c l

• I

e

, t' t e i e i o I

• I~~

\

b

.14 i

3gr. .. . .., y f.....

g re N / .f i

5

' I. .i It

. m. .r . - -:: : .: :: } ..

, a J,J!! >

f r... .. .. p .~

an. tl 4

s... . . 5g 9

hE L'**

~

b f.r1 -

. . . , . t =g am . ... --. *-

,. Fifa ll,t - 4 f

Y I. ,,, ,,,.t.s.vu .u..t.av, ..e va.st p N T . ....

, ......; ,., ,,,t . . . . .

k * ' * . <4 4 * **

3

-. , . . . w. . . e - i

* .....--.6-86*.*... 1

. . . . . . . . 1 g; gg L  ?

3 gy*,,,  !.

1 , . . . . . . . . . . =

n. r.s 4.

,, p -

• w4. , i .

.v

.. .s , y < . . . . .,,

, *! e$, ;,n..T **'"1;,-- v. *

.. .. .-. ...f._ .

=a

.L se .

g. g a

$U25Q p E 7__[ ~j ". ' . *"*f [ ,/q' 5

, o A3 e./4$

'*,,~.% _.

/~

"*. liy.".M 6 j P v [.. .. -w

? --tl$-- V

. /r3====w.,, 4@4 - $","' / '* ";r "'

.. / ,a

, 2I - J ' ' " * *

• 2 .r_ 6 - 1.' . .

6 ".g ... , i a >

z 1

a. M1' $"r%-t '.'a .* .u....._

i p*

• T. .l'O. . s " #

1. / I V

C(3:!! '

\'EU.

J-( ' *[f cI FIG 11.1-5 s

" .. A rl lb => ..".'.' j ER M.7'w.

m

. y,; ;s.,,rg ea a . . . - ..w y..-,,

,7;-*,,, Q7 P

W,, I O'**,;** WaN. , a gg w =.. ,,;^..

"" ; M

• +'._ 2, - == -..... . .

M. , j- ' - * =

i- . , . . . . . .

= " " ,/'; ~ = '

". t t g

?# * -

h y* y.'..-=.v..e.

2 -

. . . . . r..

.. s 3'.u  : .u f .. t @3

@* r - += O!" T ,I f I 17'*"ti ,-,. "* "*

,, "' y~ '* 4' * *

• J h, .- tiI1

- .~....-- 1.n .-

1 -

~= 1_ -7.l U".M*:" J

-.+,. .a

,v :: :~

. ..a..

--- t. =. -:.em .

= -s . ,, F, e,ma :r..:

v. . .e ,,

' . w.-

• 4 =,:.

p g g g:.,

g.$

. , , , . *> - s

.~ erd w,......,

. L. (.. . .-

v. m .. _4u =

... ro . . . . . . . -.. s..,

OS A b. 6VI A..b.. =%

. _ por'o-m a = * **N

. May 1971 Supplecient No. 8 ' " " " " " ' ' ' ' ' ' ' ' ' '

ca,sAMA rusuc powin.oestmct O ORT-oCAusoum.or.f.luft,U 57A .o oo r.f fu2. I p;i un W pyy tg-$ PE.-df e .

• • t WASTE DISPOSAL SYSTW FLOW DI AGIEAM 1h

. 8 ,

_w- _ ._ .- - ...

, _. __. _ _Ww ~n_-n._~m uos.u . ,

6

' l ' I

• I " FIG. ll.1-2

h I o

-I 4 i e i e i 1

a Figure A.1%

ut,N  : -

i

. . . . " . .~ " .

.m..ryj Fm. . -.- = r i

IIO II I'O FIG. I t.1 - t h i

, _ N'ile..s. 4 V FIG 18 l 6 f

I r

"#Ndf T. w- Mw.'..'.".* *M* a ".**

.e .d*d e.%. . .. e s v.....a

~

8 4. w...

$ . fi . .. i. e . y IGli

.= x -. . .. .. .. - - - .

' <--= .. . .. . ,j,

. ...LDsy Lo %

i a......

u *a%...u. .

w.. . i 2 e. nee s i.es emote s. . a ae e. .

FIG. tl.1- 6 i=-' ,

FIG il.1-6 .g i FIG il.1-6 w 5

s i,

, .. e_ih~.. .--

=*3 ~;"] - x,  : m ~3-. -L

'"7 3

- , r r e r y

w% -.

@~._ c *=* 6,,, , =* @

r '-'

• 3 -~~ -- -

g

' . . .* y",T$ .

< , . . , .s i

,,. r r .... , .: =*, ;' e p? . & M M M " '{ } *& .I

• x ' Y 'T$)-rb

=wzj "c

- gg ,,,,,,

'- . m.---,

=.

y N'.:~.:..j -

, -l -

.%....t.

u

.t ..

. .. a.s..w - .. .%......

. m. ...

g .FC ULAflO.f ***

Pew. .. g, . ' ' * -

e. 4e. Ava.m FG il.t 3 w....

. s ss n amn.as sam.g

• • • • • : wo -i.ia

.c.v

_ae se . s.a .u **-n*

a 7i, ,,...., .11.1 -1 *n.m u _ ... .

=;==.,g -.. tw . s - c, .nr . . . . . . _ . e e s.o.e e s a e.e.e-.o . Y. . = = - \

M '" * * ' * -

M * )

• f.2..,'*.*.*.L%%.. -

),.W.; . *** , m g.

.-p e

4 m. . .

s. o 4 e .. o.-

.r.....-

w..wa- . o au.

~--

3

..- 8

- 1\ 4 ' . ..

. . _f- T.t-..=. e f.

r "s:s:t

• s .. ..

, g

.c...,

y 7'" / W d i I

..n...

n

. an--

m aa . ...

g ,,,.. h w w.

y-~ m.gg,J

. i;; g . -

[ g.,p ..

. . . . ..m .- .. .x . m . . .. . .=m, ,,.,o,,.,.,.,,

_ ~~ y ( .

-@ 7 _,v.....

g.....

n v . .. . .% _ .. . . .. - xt.. . , .

• TY **'"' I h,1 .

e242. .*M. .. .'._'.n,'* .

~

4t k.

g rio ii i-2 g .. < - .A.. x...

h.

_ g[~

_w.:

g.o.'."*Jt. "^*

r .

1

• s i c o

{ l 4 l 1 m

i t ,

t ,

I o I __

l w_u- .,

j A-e,._........m ..... _ m.. .... m - ~, ,

C -

• 6. m ' *'. <a-.crea. ( a ....'* '6. .) to 6. * * * *.*** .

o -=v:t. :' ware =',*." " ~~ -" '" --" "

I c - t: = m ' = .- ; 1 ' m t :~.t- - ' ~ ~ ~~ o ~"">

w,::.m::. =m

. .,. .n..y ,

1 1.. no n i.3 .

_ .,. * - ~ ~ r. w.w.::u.s.,y .s b

i

. .u u. - -a ".-'^ -

M-~ ~'

no iu.3  ; .... -

t

_3-. m. . .

) )er n ( :7 <

- a 3' j 7

p i x-. in.I ,

Me

• q ,

. . . . . . . .... . . -%.. Mo.' . '" - -- . ,. . . . .

g- --g.. 5 .

9.

E an...r  : qp ~ ..

N dD

. . . . , , . .w. ... )

(

p.i "G"+2 .' :.. .& d[ . dff@.~"

~~f

~

$ ?i'hi lY ~n li i@

e ii i  :-

7. @ .ii @ l:i i

(

_':wm ,t

@ i

)  !

r, )  ? ) *

'* a m-.a + L A, < l .g.

( , < l x x 1, . -e..x 7 g . s.-h+ w::;:t:.. T. ... ,,, ..

~ I ~___.--_a .

)

. .. . n o. u -6 . .~. e

. . . . ... _ . . . ~_4__  :

ts L

.. ... .** L'. i"~ (

..  % ." ..g n ' ' '

-.._h-~ rw,-. ~,gM -d .

- e 2- m ..;) ~-

..- - .. 4.N;Tw.., -

M ,

(

n o.;i,.i.o e.1ud.m.s / h--ute.-

A n . .. ,!

4 " - ~ . -

- - _f -} ...... . ... J ,

p

..,,,,g,, _.

g _ . . . ,

)g ..-

j

@ (

...... a 1 g. e. - .,a%

,,0..u-3. .

. . . . .g,; . .

y

,g... ,,

_Sp.g  ;, ,.

m. ;.g u,. ~

, . . n 3

v..n., ,. . a

_WA A A a A A*F * .s . . d

1. f ~ ~ ~

s!

h'-a-@k[ - \- . .r,i gL 7'-::. .c *.i .wno s

a

. --,n- un *

.l.

t. I s

@ @ >'- g 6) @ $ #) $5 @ @l 'T'u'-'~. ." q . -

,V = ,

~~ w

& &, y, s y

t- - -

y& $1 5)> f .. -

,7*n.'.w

- ..._ _~.-

\

I P. =. m:'. Mro :fr a

.u . c=u nllF.'.'02 r 0.flJsll"X" r- - .  :=; :- s-: ;.'..c-- _.,

.. ., . v s

~~~ kb".h".$*5. i > 176

."l 'E. A. *'.".{ 4 .g

"- Nh- ....e...r..............,,.

.a .v.t..s. S. E F. (.. p g.3

p. ,,.,,,,,,

A l Ot,o.MA PU.3UC t j oun s P. OWL 1 D.i, STRICT nai .

- 1cau.u wo o.cr,.tr. w no icoe e c -- c= * - . WASTE DISPOSAL SYSTEM

h. e.mD I

ge.

}.,qs-r H .- ,pRm. .

-w. _

FLOW DIAGRAM p

1 ,.,

un . ca'* n

,1 _ m

• • .=a ****"

=

, _ t

' _ _--_-r P

__ -..a n.0$- H

, , i ,

I o I a F I G.Il.1 4 a a

. I I

3 a i e i c l o i

Figure A.16 i

~*:,- : =E FIG tl.1 4 * * * * * -

• J~ v .. - . .

.7 lL'"

j r ..-

g,, H:. th,#.

j r ..m x yg7ll;', =g e.... f

~

g rig H i 2 . r. . u .- J

_ ,g' 8 '

~

",,m c

]I" ' * "

]... .

-'~ n-- rcio:

4 use. .M.

.72"*!N W g **1..

1  :. i i p. '

m.M fig H.1-2

~~- ..~

I iI  ?

T s

g.,_

3 .

, . . , C H:

. J.,.-

.A , w-

. ..s

g. . ...

@ =;.'~4c.;g="= x **; U v .......

mg -.

r,G. ;,- i:4 r. .T. .A.

,r......-

.=- <

x ~.....

. c....

! fl,..-- M g r A p r d...  ;  ; i

}% *h

• " " '

• Y*****

$ b l j "".'".7l ; M) 4p ,. _ _ _i._i*~____j rl l

• ' it l 4$ 'a=~ e,.".M. . .,. ,P" t

i Illlll lIlll i i.* .u

',*i.O..", ]= = -

.=-m' i - i ...

s ," - ,

~ ~ . - .

i i gwY. ..."c*l*t).,.,,,.,,

8 l

' FIG.H.i 2 * -

3 w_n.

l l

l l

.11NL-O eghns 69ft.tns ' . r*J. ---; "

g

.a,e g

.e-.e. F * *!n -

'(,

! M. l f 8' t___________________;

i

( w m e . . u-, ,

. i

,A=-+/

w-

=

\ s.

g

.. a <,

4

= a*

r I-,, .e .y., Q ..

-6*.

.mm.. a

<. ,- 1 tg ,,, et --

=r. _m , ..".-- T. . . . . . .

.mes.,k

~

gu t.......

b "3"-

-f .

s t+ mt* 1

• 9 -

} ..9_ & ~. n V:**

j ge \

9 46 '

I I l A I e i e I o J

e

o j

i e 1 e o l N FIG 11 i 6 2 FIG 18. l f, n= % -.qp 4 = ..1 .- 6.+ .  ;

i 9 e 9 ,.. j

-4 s t

- Q 4e.s. ..h .O i

g 3 g 4 . . . ..... .. . ,

lt NN I

• • • '"." i

,i 7, ~a m.o r.t . . ... . .... ..

~

r. I

e -, ,, ,,,

l 2, m..".::'.T..'3."; 'rt.'.?

% ,- ^ .

. .um . . .n ...

r > .,

,m- +

Q:

p Y+,.

--- z l

t--,

l' I a ' I qj

• i_____ _______, u. 1 c ..... ,.ru. ,

-.ve ..,,u..

q =.G. is.".

ri

. -3 ,

gm x x t [i,,,'-_.._s_...........,

r G. s i. e 4 4

MQh l 4

==

ia 1 1

i . :Ic

m. I

-. I

• [.., bs3*3 c,

Le; M Qg. .v..... L---- -----

h---- y, ,-,ad,,,,,,,,,,.,,,,

l - . .. t.. = .a b  :

h.... C c . . .m r a G. il . e - 6

" a - .g e

.v..... 4 A m . . . . . . .. . .

@ ,. i -

.n.e s .

..a -

i

. e >

~

L W4'. V:

r m

FM 6~ 4.--

=b b p. " " * * * '

ft@?.  ;;~;,..; .

- 0} h. . .. . ., ...

h i i

"'~~~l .4 T.ra. .

W ..

)

~'

g.,~~4 .. I

Jiin ...

. . .,. F 9T a ,4

.= I T~ c .c s 3

.J.,.c.,, . M '

4__ . . .

..?'a

- y

/ ... e \'Q ,**,'

%s a=%

Fw.. .se g

-e6*.g 3.(.36'i. .m

• a

. w r.u v a ..

d' a n. .. . a 1 W r e tG it 8 . .4 ni .

,e t

eee .- r;. _ #6 l

4-. aj... Rirrev. er.J* .v,.... .r. ........ r - - ,, 177 v...

g ,,,,g... ,,,

.4.m..n.* FIG.l l. e - 4

.. . .. 6 . . . . . . . ,

gaa.a

~ '+ m

. lll ll ossana

=r emru.euc= ==. rowe.i.c.e.s.Ta.ic.t

• M M Rin e I .1 I '

I I NESTE D5eOSALSYST3.A irgy . . . . . .

rG

> g

'"* "* d' 6.mur syr A a,,,.A _"'"",,',,,,,,,

= - -

1 i

- r.-.i---m _ .- _ mas.u.e e ' l' ' I a I a FIG. l l.1-S -

,l

, . a .

Figure A.17 Level of Waste Holdup Tanks: Before Refueling Tank Level in Inches of H,0 8 I I I l 5 I I I I I I For Waste Holdup Tanks 440 inches a 45.800 gals

, f Samples Taken _

t o 3m _

3:

'~

h  : 1 1M - -

o on-e-e-e-e-e-e-e-e-e-e-e-e-e-e-e-e-e-: :: : e-e-e-e-e-e-e-e-e- e- e*e-e*** *** * *

• 400 _ ,8*- *- *- o _

to si  !-

/ 4-9-e-4-0-e*T*~*e pe.es .- _

'E200 _

e '*-e-e-e-e-e-e *'%- ,

100 -

,i e-e-e-.- e .-e-e e e.e-e.e-..

4w -

y300 _

I a 2mu -o-e-e-e _

g ,,,,_ e-e-e e, geo-e-e-e-ee,,,

o I i e i i . .-e-e-e-... .e-. e e e e-e-o.e e a. e e e e e e e#

8/ 9 8/10 8/12 8/14 8/16 8/18 8/20 8/22 8/24 8/26 8/28 8/30 9/1 Date m 1976 mrt-A-isas Y __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ - - _ _ _ . - _ . - - _ - - - - _ ----- - _-__- __ _

a Figure A.18 ye ~ Level of Waste Holdup Tanks : During Refueling 600 -

Tank Levelin inches of H2O

.- 500 - For Waste Holdup Tanks: 440 inches = 45,800 gals _

M 400 - - -

Q 3 300 -

I y.200 -

- O: Sample Taken -

0 500 -

g 400 -

O

~

C I

A 2% - '

100 -

0 500 -

0 400 -

a 3

.x 300 -

h200 - -

?

d -

100 -

0- 1 I I I I I I 10/1 10/3 10/5 10/7 10/9 10/11 10/13 10/15 10/18 Date in 1976 ' INEL-A-2533

/

.4 g 179 L

/!

WM.

66 2 2 5

- ~ - ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ - /

7 2-A-

4 L g i e '

2 E

/ NI 2

O n -

i l 2 e . 2 u i

/

2 f

e R 0

. 2

- r i /

2 t

e f 8 A i 1

/

2 s ls k a 6 n g' aO i 0 .

1

/

2 T2 0 pH 8 7

uf 5 47 9 ld o so i 4 ' - 1 9

/

1 = 2 1 A

Heh s e ~

i n

e c n 2 e e t si i h c -

1

/

t a

r a n n C 2 D

g u Wi i l

0 ovee 0 f

i 4 F i 4 . 1

/

leL v

s - _

2 ek k Ln a n -

8 T i Tap n

/

2 u e k

d a 6 i lo

/

T 2 H e _

t e lp 4

i s m  : /

a .a 2 W.S r

o 2 i F - O. .

_  : /

2 7

1

• . - - - - - - - - - ' - - - - - 3

/ .

0 0 0 0 0 0 0 0 0 0 M 0 '0 0 0 0 1 0 0 0 0 0 0 6 5 4 3 2 1 0

5 M30 0 2 1 0 0 5 4 0

3 m0 1

$Q3 ;s 3 jF s fF f

• = s .

Figure A.20 Level of Radwaste Tanks: Befom Refueling Tank Level in Inches of Hon I i I I I I i I l l 1 3 Tank Levels. Spent Regen 90 snches

• 6000 gals Hotel 6s enches = 1200 gats Monstor. 90 inches = 6000 gals j : Samples Taken 3im -

t 4o . *

" . , ' . _ .* *. g ..' 1 3

" f.l .,N.1 *. . .. p... .# ".* h/ \ ....../.... '*

l c 0

}$s $

• .x. . .. . . . /,..\ '. -*

V

[t*/ '/ .i....* ,.i#e A, A .*.......-..r*

1. • .4 '. .

• /

/*,A.f* ' .' . ,

O

~

_ 1,w -

w -

~ *

• I

i' ' *\_4 ' l./\ *fs..II

~

O _. '.. f**i*/ \ " d\ m r_\.Ts #e* $ . m . . e**? \

• u-O .

\.....I \.e ! \. =..I\ ln .=.=.=.e_ne_._E.o T.

..Y\.e.I\ ..=.a.11I.\ O i h ~N.'

0 -

>.s, . _ . .

-* w

.t l* ... .

<1. -

g .. s .-. . ... ,

J *\ ,., -

, ss. ,. ... a .i  ;

.\ .,' .\

\ i ~ . i i

a c

O _ .i... i i

. a

~

.i i\

, .I .

t 7".

s e- '

v. ,\. r \ .y .\ iva. . .. ?. \ i y. -. .\ p ' \ ,_r \.;_\. pi, < r-

- 0 .-,

-?

\. e,

.'\ L , . _

8'28 6'30 9-

.,1

\.1 8/12 8'14 8/16 8.18 8/20 8/22 8 24 8 26 8/9 8/10 lNE L- A-1522 Date en 1976

Figure A.21 Level of Radwaste Tanks : During Refueling Tank Levelin inches of H2O p 150 , , , , , i i 3m Tank Levels: Reactor Coolant Drain: 42 inches = 900 gals o E - 100 -

Spent Regenerant: 90 inches = 6000 gals-O$y O: Sample Taken Hotel: 65 inches = 1200 gals o3 50 -

Monitor: 90 inches = 6000 gals- ,

g 0 _n _ ___ -

E $100 -

50 -

5hm 0 O

f }

, $100 -

I aS 50 - /

0 u

~

t 4 100 - -

j 3 0 -

50 -

20 3 0 ' ' ' " ^

10/1

• 10/3 10/5 10/7 10/9 10/11 10/13 10/t5 10/17 Date in 1976 INEL-A-2532 182

. . v Finure A.22 Levet of Radwaste Tanks : After Refueling Tank Levelin inches of H2O Tank Levels: Reactor Coolant Drain: 42 inches = 900 gals Spent Regenerant: 90 inches = 6000 gals Hotel: 65 inches = 1200 gals x Monitor 90 inches = 6000 gals jj 150 , , , , , , i i i e i i

[ $o

~

O: Sample Taken O5 V

< 9' c

U 100 -

m 0 - - - - - -

s fis -

rvM MM #-

e b100 -

gm 0 -

M _ _ - , M -

I N100 -

ll':21 . A l V.A H L E 1/31 2/2 2/4 2/6 2/8 2/10 2/12 2/14 2/16 2/18 2/20 2/22 2'24 2/26 Date in 1977 INEL-A-2530

3 2

/ 4

- - - - [ - - - - - - 9 46 s A-en u L t

ae ck 1 E

N I

idt a s i i

' 9

/

I ins {

e  ;

wlp o

r m 0

i Asr a , i i 3

/

' 8 =

4I 8

i i i ' 2/ a 8

l- 6 i , i i 2

8 ' /8 4

I , i , e ' 2

/

S 8 R

E T 1 tE t 2 A i , i I i

' 2

/

R 8 3

A P 6 2 7 R /

A O 2 - 0 T / , i r i 2

o r

A R9 ' /

8 u O -

g P i A 9 F V /

E 8 i I i #

t e 8

1 E /

T  : 8 S d A o W i D r A e R P t L i i 6

1

/

8 4

i , i i 1

/

8

• i 2

1

/

- 8 e

L 6

7 0 9 1

i i 1 8/ in

- - - - - - - - - 9 e

- t 5

/ a 0 5 0 l 0 E 2 2 1 1 2 5

1 0

1 5

2 0

2 5

1 0

1 5 0 8 D Ea2 _#70EE # grE 3 320g8E5 ES b j S poo $ 5 b ll

Figure A.24 RADWASTE EVAPORATOR PARAMETERS Period: 00:00,8/24/76 through 00:00,8/28/76 Railge f'or ev'aporator parar'neters during lhis period:

30 Evaporator press: 15 - 16 psia Evaporator sump temp: 220 - 226 'F

~

Recirc pump disc press: 35 - 37 psig Evaporator level: 39 - 40 inches f Arrow indicates samples taken 25 _

g Distillate pump disc press: 33 - 36 psig Vapor condenser press: 15 - 16 psia 3 Evaporator bundle press: Vapor condenser level: 12 inches _

20 Bg Operating 16 psig Conc cooler out 1,emp: 204 - 207

• F WD4B Ea 15 Re circ: 8 psig WD13B WD4B WD13A to WD13A _

gS WD4B to to to VD22B to

io Wo22A WD22B WD A WD22B WD22B _.

5 g 5 b 8 5 -

p 5 E E ' E O I ' J 38 8 8 3 ' 3 3

10.0 i i 2 i 5 5 5 g y gi C

i g

C i i C

g

- E 7.5 - E _

O b E 5 50 g

_ g6 p

~ - a b _ _ - - -

m a c 2.5 -

81 e ' I '

0 s e i e t i 3

15.0 i  ; i i , , , , ,

i

.{ 12.5 -

E 10'.0 -

eEg SOE7.5 -

W S i i 5.0

=

( _

2.5 -

0 ' ' I I i e i t i OV;00 04:00 08:00 10.00 00:00 17.00 00.00 12:00 00:00 12:00 00.00 8/24/76 8/24/76 8/25/76 8/25/76 8/26/76 8/26/76 8/27/76 8/27/76 8/28/76 INEL-A-6443

p' -y Figure A.25 RADWASTE EVAPORATOR PARAMETERS Period: 00:00,8/28/76 through n0:00, 9/1/76 35 Range for evaporator parameters during this period: I I I I 30 Evaporator press: 15 - 16 psia Evaporator sump temp: 220 - 226 'F Arrow indicates -

2 25 Recire pump disc press: 35 - 37 psig Evaporator level: 38 - 41 inches samples taken S Distillate pump disc press: 2'3 - 36 psig Vapor condenser level: 12 inches -

$E E c1 20 - Evaporator bundle press:

Operating: 16 psig Conc cooler out temp: 204 - 207 'F _

WD4B to

&3 15 - Recirc: 8 psig WD13A WD22A WD138 WD13B W B to WD228 W 2A W 2B L W 2A 5 \

8 g

.o =

' ' ' ' =E ' ' '

03 i I

~

8 100'y,'.- , i i s.h s, ,

@ "3 z

i e

o 7.5 E s C $

3 5.0 -" O CL n o* ~

$ue 2.5 -

9 E O E t e i t l I , e p 12.5 i i , i i ,

a pg 10.0 -

eu c o 8 E 7.5 gu :L _.,

$ $ ~ 5.0 -

5 2.5

$ A -

0 I I I I Y I I I e 00.00 04:00 08:00 12:00 00:00 12.00 00 00 12:00 00.00 12:00 00:00 8/28/76 8/28/76 8/29/76 8/29/76 8/30/76 8/30/76 8/31/76 8/31/76 9/1/76 INEL-A-6444 e

• a =<

i lI ll

. 27 2 4

/ 4

_ - - - - - - - - 2 6-

, A-s en t E L

ae 0I N ick a ' i ' i I 2 idt /

2 in se -

o wlp t

r m 8 a

i As ' i

~

' e 7

1

/

2 I(

6 1

' i '

i i /

- 2 S 4 R 1 E i ' i I i /

T 2 E 7 M 7 A /

R 1

- A 2 2 1

P / e i I e /

6 2 - 2 2 R O - ^

A T A 1 e R 3 r O / 0 u P 1 i i ' i 1

/

i g A V ^ 2 F E :

d E o -

T i 8 S r A e WP i

r I i ' i '

2

/

D A '

R ,

6 I i ' i I /

i 2

m

^

g 4 i I i ' e

' /2 o-N i

A i

g- ' i 2 7

/ 7 2 91 1

i t

n e

- - - - - - - - - - 3

/

a 0

1 D 5 0 5 0 5 0 0 5 0 5 0 0 5 0 M 2 2 1 1 2 1 1 2 1 1 ke 7[ $a ho" $n.

eoIjsg ng D C

xg*a 8s a

t

g Figure A.27 RADWASTE EVAPORATOR PARAMETERS Period: 00:00,2/17/77 through 00:00,2/20/77 35 , , , , , , ,

Range for evaporator parameters during this period:

~

=

Evaporator press: 16 - 16.5 psia Evaporator sump temp: 227 - 228

• F ~

u 25 Recirc pump disc. press: 38 psig Evaporator level: 34 - 34.5 inches -

g Distillate pump disc oress: 34.5 - 35 psig Vapor condenser press: 17 - 17.2 psia E E 20 Evaporator bundle press: Vapor condenser level: 12 inches -

Sa Operating: 8 psig Conc. cooter out temp: 190 - 192

• F f8 15 - Recire. 6 psig ~

& WD13A WD13B WD138 E 10 to to to _

Arrow indicates WD22A WD22A WD22A 5 - samples taken i ' ' t t' t t '

-e 70 8 g c 10.0 i i i c i i h 7.5 -

~-h > ]g 8,j

-t

= m " ,

a h85.0 -

ls g 3 -

e8o 2.5 E 8 'g e

g E 8 -

m j 0 8 8 8 ' '  ! '

12.5 i , , , , , ,

.x.

,3 10.0 -

I $7.5 -

ao E j g35.0 -

kg 25 -

o y p -

0 ' ' ' ' ' '

00:00 04:00 08:00 12:00 00:00 12:00 00.00 12:00 00.00 2/17/77 2/17/77 2/18/77 2/18/77 2/19/77 2/19/77 2/20/77 INEL-A-6446 .

h h

I h

'~ A 6 C l l l l e ~.e+ ^

. v!' . . f,

~ - - - -

+g .O 6..', ,#

, *; t aaf t e s. **w A, I a I

l i I

.ur e6 .c.. .

j g L4 [. '.. d b

.i .g_ t d: i v av .  ; , . g-.1,,_ [

.. , ,a~.* it s rc

- u s ..

i.L' n e ' s

~"

c  :

n~

r.

n. ...y ;-

nf w .x*)--

1,'

3

.c-

~

2 y...=,,

• *

• T

'..*.T' l 7 ,

,, f ,n o. c,.

'e]t g _k,'* so '

,e w

v.E

{

}  :.: 1 1 c '-

9. - g- :,PM. an, ..,5 --9... *.:.). ~l ...yy.' . ,

o*ii. - - 3e 6.n t 6e ,* e b - L* .

- n. ,.-

[.

_ p._._r- 4.u '  ;;;s f} 7**Rl;

,I

1. (-

' I i _ . . _

I l .1 *-l - . .m l I '

. .. s

+,. ! ,e.

1 .. l

! .l it y,

.t <

g I ': 1y a -il .. .. l .~; ,

I('0

~l l I . kl .!5-l .

l l

I l llll l >

g lt

. . __1_ . . 5I l ..

' [.

'~

.. ' l I~l 3 m1 . .c:Fl*

co e

I i M4

-- :ly";2 is ,) t *B, p y $, +2-w I

t, i

l ll .

&t . . h W.T

• l  ; t

~- p - : ..=-  ! -

=.

, ,a

' );, f

. . . . . . a

.l ha

! I >

.-. l --  ; 3

.1 .. i. .

gm.: . !!

?

+

(-i l ,..: . s 5 ,,,,,

i I "" *

. . . 3 n -

Er..-m sr =

[,..

8 C iE- ~[ Th4 ;d" 4 ll [},ft'dId'~f'Yh ". <q kb Tl '

n

_4 4

c

,i, k,

i L. } % v f I

_gcr Tlq'

. ._.3_....~, 4 re u w I i 1, ;I :  !  :

l l p=

w r, i <,. 2 ' .,  :;:

2 .

l:ll""~. p.

s

'- '. 3 * ;r =>

3 .

. . . ::. :.r.u---

t o l n l 3

. 7

,T. 7.,,

3 ' - -

, e . 3 ,

~

j'; -

l; 4

t fphJ1L l '*$+

i. m; n i .e_ i _

h< V 4

t e

pg 7  ;

T; r....,m......

>= ~ ~ ~~s - ~~ --

i.....--. .~ m.

g- .__._ .

ri.. .. . .

j' f.:ix - fg f$.:j=y2_

g g i g, I ll

, d" 'sl- -riinig; =W' ul l 1:

t, t, t.

/ s

^

a I e I

,I s s .

t ' - - .

i

3. i I l F l 0 l M l

v MU.iF c .e :m.---

Firiure A.78 5

'm'.':aM:g;w:

i t V.'rs. pw L. . .; i

! y9 u .7,3f !

l y

r- .. .  %.,f L

, c..,. 7. p, .,7

- i * , . ;.; ; , ,, ,

,, g s ep. ... 9 ,

yb 1.3 h.

)h * %* I .G...,A..S. . ~ ~ N. M.5'E

<m

• % . q15 .!. b,,r.-M 1.j&g.]j .

q_; e q . . -

-l

~ --. .n .- w el ,l M' l : W)]~ -.T'-a, - -,--~*'pk .%'

I 7 .-s.' ;; -_y ,

- ' jf. - y. '.- :t- y - ":.-: .? ,

l .;

""t" "'

,L T_ _;', ',*,, ', f.l, I 4.w,, iL -@

._~.a g 4 '.

c 6 ) p.p,.,f , l . . . . , .

_ . _ (g.A.y .N) .. ,.] . . , ..

n .;. x .um_ .. .. .. t M =e .u o ACC(% gg A _ L_~ d L' -@,-l ,l l 133 -

g

. . . h

'77I u 111rit a% 12*8.L <r a

f""'"* *""

.t - _c r 3 = f -~ a ,& Hl 1 -

Q.. l.

s . w l l l

f . . . . .~ f3 2- - -.w--.w. -

4 -o--.g c?.

l Z- .s W ,, s-. -* .

.q

~~

a. e I y 4.a a c.-- (.;.-.* ,, . .

e s,- - _ - - 'l _i t- -'

i 1. - - ,..._- m ,. % i

, ;. P.ca ... ., .., .

A c. il ' ' ' - ..'

l y m n .-- 3 '",, gl , . . _ . . _

-~ -

..u c~.-

q l 4.w

~ " -

- .=

a l 1:

8 y' .. .~

I' ] Il4,b-.I ]y r.w -~

T

a( m ' , .eg m

, g,%. =...::. t TTT. '

i T r , M.r,, .T Tg ri l 11 l il lg Lg m...... ~.. ..... ..

3 i

F 'T g

$ I'

'il 1l f

l , ,,l t I it =--

h h . . . ~

I e I "r *1,-

I.El I'I- l I_l -,i l ' "I l f fli i llAl

~,1 r* I' 1 .- J~  !

l, i " .

'[. A* . .{ h"4 , -

' .- ~' -

it

• 4 l

• * ' ' ~ " ' ~

a ri"

. .g

'l '

r. ,

'r:^  :

3 I-M e 1MHbb-Il l n I

[ ii~F (f I

" * * ~ "

y[I , j.

"~

i i N, h 0 j!  ! 1:

sp , ,

si d .. , . , .

1 ftg. Mj r1 i f,i , 4% ~ .,

z.. . . . . . -

l:

c.. L

. t ,

, 3 o-, r. ti "

1 t5 g

rp I. .

,.3 1

g

, . '. ,': .- A' e, a.. . .

~ w -<

., 9.- ^

l i.Ii M$ ,l h'Y l-' ll et l.i y l:

! 2{:;,,l*2-l1 . ' -

-j j i ! % i i=. _,,s r .- - t.f:c+l U..

.. )

t.

=-

l _ ,;~ .%

. . 'z: :--vr- ,  :-= -+,.s=:.9. . .x.-

g@+E,g.-m tj ,p,u

%:: L-.- c"- veg+ +n ' .

l MI y- r-.I c-- ,5*, @'M il @ @ @ rrl L '+--".4J

(" sk #O i e$ i ed$[ M. l 42 :. ' ' ~

l A

-8 C"h.' til i I t ._

9

'}

f

. e. . - <Fe l t

, '" . 0-l -G4[7ID3

~M h'Y' ,Tl

......e ....-s I'

8 - I .

l*4. - ~ *lC

' M, l W.*') _]t m, , - M ,b" . 3 l i }' l , ' l. l 4.-4. . ~_ u , . . --

~

l-u

..es lw- L... J, s

.5

..r va. . . w< ~ *.* 3 aa .a =.

1

- ., cow m.  : ,.c., =e- l  !

sm.e...., v .,

.. ytt.,,. n:::g a , , . , com.e

.ne e ., e .. e.w . - *- s )i i

{

I earo ne a =*** *saa .ame.e. - . ' ' ,

{

j """*'

3 l l t 73 37 T f,

! k l' f ili .ljl l}

l l' 'l l? ll 15

ll .

f ' ' ~ "

• i ses ins .e 2.svwoot s er an.a . '

f . , ,

, , ~ . .u.o .... , 189 4' % .56. .

. r .,.1- ,~k m.iw. . . . . . c a.*u< r *v -1 maan . ft .e . a g

. * =**~ rjt  ? m n ** " '*J . OssAMA PusuC POWER DISTittCT

• o.' *l

,-, {pf.- .,..m '-((Mm,ur . gi Fosr1 Causoun statiose useT seo. s

b. I:'-m' .m g,, -o-< = . oi=i AdKILIARY SUILDiegG

- A>< r-.

'*s M%f.o.

,a 1.m  : a HEATHIG S VEssTILATileG i . ..

_ ., ., M

. s

. - -? # n u dam. e- c _^rf Flow oinstaas

~ ~ ~ ' ~ ^^

n# * $

QM;I(-N .$22d~.N .J, ""l',:" . - hm

• . " ' U*F --,,,,,_, us-ne e i c e l F l 0 l *e Fc. h '-i l g 4

• a f

7 2

6 w 1

- A ys -

a cm eA GL N

I D k sn aa

• GT e r o

s s

se ar m GpR 8 , m 2 ,

o C

A

. t m g n eg g e

e i nnt n t F olia poehR a m s

y f moHc oC x S o C E w 4 o -

l F 9

F g o n ,n

?. n i A i t t c n e r

e a e w ot anm g

r l S d ea q

u i t m ,

tuHhc R h

i n o 3 i S E x

F e r lee pn V F mo ati ev St a

S m t laR V

l i l u d 8

B o1 TA s V A f

• s; a4 S

i eo k