LAC-TR-138

LACBWR

INITIAL

SITE CHARACTERIZATION SURVEY

FOR SAFSTOR

By:

Larry Nelson

Health and Safety/Maintenance Supervisor

October 1995

Revised: December 2009

Dairyland Power Cooperative

3200 East Avenue South

La Crosse, WI 54601

LAC-TR-138

PAGE 1

LACBWR

INITIAL SITE CHARACTERIZATION SURVEY FOR SAFSTOR

1.0 INTRODUCTION

The La Crosse Boiling Water Reactor (LACBWR) is owned and was operated by

Dairyland Power Cooperative.

The LACBWR was a nuclear power plant of nominal 50 Mw electrical output, which

utilized a forced-circulation, direct-cycle boiling-water reactor as its heat source. The

plant is located on the east bank of the Mississippi River in Vernon County, Wisconsin,

approximately 1 mile south of the village of Genoa, Wisconsin, and approximately 19

miles south of the city of La Crosse, Wisconsin.

The plant was one of a series of demonstration plants funded in part by the U.S. Atomic

Energy Commission (AEC). The nuclear steam supply system and its auxiliaries were

funded by the AEC, and the balance of the plant was funded by the Dairyland Power

Cooperative. The Allis-Chalmers Company was the original licensee; the AEC later sold

the plant to the Dairyland Power Cooperative (DPC) and provided them with a

provisional operating license.

La Crosse Boiling Water Reactor achieved initial criticality on July 11, 1967, and the low

power testing program was completed by September 1967. In November 1967, the

power testing program began. The power testing program culminated in a 28-day power

run between August 14 and September 13, 1969.

Dairyland Power Cooperative has operated the facility as a base-load plant on its system

since November 1, 1969, when the Commission accepted the facility from Allis-

Chalmers.

The La Crosse Boiling Water Reactor was permanently shut down on April 30, 1987.

Final reactor shutdown was completed at 0905 hours on April 30, 1987. The availability

factor for LACBWR in 1987 had been 96.4%.

Final reactor defueling was completed on June 11, 1987. Eleven fuel cycles over the 20

years of operation have resulted in a total of 333 irradiated fuel assemblies being stored

in the LACBWR Fuel Element Storage Well.

During this time the reactor was critical for a total of 103,287.5 hours. The 50 MW

generator was on the line for 96,274.6 hours. Total gross electrical energy generated

(MWH) was 4,046,923. The unit availability factor was 62.9%.

LAC-TR-138

PAGE 2

2.0 OPERATING EVENTS WHICH COULD AFFECT PLANT CLEANUP

(1) Failed Fuel

During refueling operations following the first few fuel cycles, several fuel elements

were observed to have failed fuel rods. These fuel failures were severe enough to have

allowed fission products to escape into the Fuel Element Storage Well and reactor

coolant. These fission product particles then entered, or had the potential to enter and

lodge in or plate out in, the following systems:

1) Forced Circulation

2) Purification

3) Decay Heat

4) Main Condenser

5) Fuel Element Storage Well

6) Overhead Storage Tank

7) Emergency Core Spray

8) Condensate system between main condenser and condensate demineralizer

resin beds

9) Reactor Vessel

10) Seal Injection

11) Waste Water

12) Reactor Coolant Post-Accident Sampling System

13) Control Rod Drive System

(2) Fuel Element Storage Well Leakage

The stainless steel liner in the Fuel Element Storage Well (FESW) has had a history of

leakage. From the date of initial service until 1980, the leakage increased from

approximately 2 gallons per hour (gph) to just over 14 gph. In 1980, epoxy was injected

behind the liner and leakage was reduced to approximately 2 gph. In 1993, the FESW

pump seals were discovered to be defective and were replaced which reduced the leak

rate to approximately 1 gph. FESW leakage has stabilized over the years to an average of

approximately 21 gallons per day. This leakage is contained within the steel shell of the

Reactor Building.

(3) Release of Contaminated Water To The Controlled Area, July 2, 1982 at 0630

The failure to close the resin inlet valve in the resin addition line to the number one full

flow demineralizer following the addition of resins subsequently caused the release of

water to the Turbine Building Floor through the gasket on a bulls-eye sight glass in that

line. Approximately 75 gallons of contaminated water could not be accounted for in the

waste water storage tanks. Approximately 20 gallons of this water is estimated to have

entered the ground in the radiological controlled area outside the west turbine hall door

and the turbine hall truck bay door. Contaminated ground was removed over a 3 sq. ft

area by the west turbine hall door and 2 sq. yard area by the truck bay door. It was placed

in waste storage barrels and later sent to burial at Barnwell, S.C.

Waste

Spill

Turbine Building

Switchyard

July 2. 1982 Spill Area

LSA

LAC-TR-138 PAGE3

LAC-TR-138

PAGE 4

3.0 CLASSIFICATION OF THE GENOA SITE AREA BY CONTAMINATION

POTENTIAL

As per NUREG 5849 "Manual For Conducting Radiological Surveys in Support of

License Termination" the LACBWR site will be placed into two separate classifications

in accordance with section 4.2 of the NUREG. These classifications are defined as

follows:

Affected Areas: Areas that have potential radioactive contamination (based on plant

operating history) or known radioactive contamination. This would normally include

areas where radioactive materials were used and stored, and where records indicate spills

or other unusual occurrences that could have resulted in spread of contamination. Areas

immediately surrounding or adjacent to locations where radioactive materials were used,

stored, or spilled are included in this classification because of the potential for inadvertent

spread of contamination.

At LACBWR the affected area will be that area located within the LSE boundary.

Movement of radioactive material occurred throughout this area, therefore a potential for

the inadvertent spread of contamination in this area could have occurred. (See map).

Unaffected Areas: All areas not classified as affected. These areas are not expected to

contain residual radioactivity, based on a knowledge of site history and previous survey

information.

The area outside the LACBWR LSE fence will be classified as the unaffected area.

LACBWR AFFECTED AREA MAP

Switchyard

LAC-TR-138 PAGES

• ..

LAC-TR-138

PAGE 6

4.0 CHARACTERIZATION SURVEYS

In order to provide an initial site characterization survey for LACBWR the following

radionuclide determination tests were performed.

(1) Spent fuel radionuclide inventory

(2) Core internal /RX radionuclide inventory

(3) Plant loose surface radionuclide inventory

(4) Plant systems internal radionuclide inventory

(5) Reactor biological shield activation survey

(6) Circulating cooling water outfall area radionuclide inventory

(7) Site soil contamination determination survey

The results of these individual surveys provide the initial LACBWR site characterization

data that will be needed during the decommissioning period. This data will be used to

assist in the performance of the final Site Termination Survey which will be performed at

the end of the decommissioning period.

4.1 Spent Fuel Radionuclide Inventory

During June 1987 all fuel assemblies were removed from the reactor vessel.

Currently there are 333 spent fuel assemblies stored in the spent fuel pool.

The 72 fuel assemblies removed from the reactor in June 1987 have assembly

average exposures ranging from 4,678 to 19,259 megawatt-days per metric ton of

uranium. The exposures of the 261 fuel assemblies discharged during previous

refuelings range from 7,575 to 21,532 MWD/MTU. The oldest fuel stored was

discharged from the reactor in August 1972. Forty-nine of the A-C fuel

assemblies discharged prior to May 1982 contain one or more fuel rods with

visible cladding defects and 54 additional A-C fuel assemblies discharged prior to

December 1980 contain one or more leaking fuel rods as indicated by higher than

normal fission product activity observed during dry sipping tests.

The established radioactivity inventory in the 333 spent fuel assemblies was

performed by using the computer program Fact 1 and hand calculations

performed by Dr. S. Raffety (Nuclear Engineer) during July of 1987. Activity in

the fuel assemblies hardware is based on neutron activation on this hardware. All

activity values have been decay corrected to January 1988.

LAC-TR-138

PAGE 7

Spent Fuel Radioactivity Inventory

January 1988

Radionuclide Half Life

(Years)

Activity

(Curies)

Radionuclide Half Life

(Years)

Activity

(Curies)

Ce-144 7.801 E-1 2.636 E+6 Sr-90 2.770 E+1 1.147 E+6

Cs-137 3.014 E+1 1.666 E+6 Pu-241 1.440 E+1 1.138 E+6

Ru-106 1.008 E+0 1.524 E+6 Fe-55 2.700 E+0 5.254 E+5

Zr(Nb)-95 1.754E-1(9.58E-2) 3.555 E+5 Zr-95 1.750 E-1 3.52 E+2

Cs-134 2.070 E+0 3.291 E+5 Ni-59 8.000 E+4 2.87 E+2

Kr-85 1.072 E+1 1.160 E+5 Tc-99 2.120 E+5 2.76 E+2

Ag-110m 6.990 E-1 1.018 E+5 Sb-125 2.760 E+0 2.73 E+2

Sr-89 1.385 E-1 1.009 E+5 Eu-155 4.960 E+0 1.68 E+2

Te-127m 2.990 E-1 8.238 E+4 U-234 2.440 E+5 6.37 E+1

Co-60 5.270 E+0 6.395 E+4 Am-243 7.380 E+3 6.31 E+1

Ru-103 1.075 E-1 6.334 E+4 Cd-113m 1.359 E+1 1.78 E+1

Pm-147 2.620 E+0 4.129 E+4 Nb-94 2.000 E+4 1.59 E+1

Ni-63 1.000 E+2 3.540 E+4 Cs-135 3.000 E+6 1.40 E+1

Ce-141 8.890 E-2 2.638 E+4 U-238 4.470 E+9 1.22 E+1

Cm-242 4.459 E-1 1.858 E+4 Eu-156 4.160 E-2 8.63

Am-241 4.329 E+2 1.474 E+4 Pu-242 3.760 E+5 8.58

Pu-238 8.774 E+1 1.262 E+4 U-236 2.340 E+7 6.32

Pu-239 2.410 E+4 8.837 E+3 Sn-121m 7.600 E+1 4.44

Pu-240 6.550 E+3 7.165 E+3 Np-237 2.140 E+6 2.19

Eu-154 8.750 E+0 4.020 E+3 U-235 7.040 E+8 1.89

Cm-244 1.812 E+1 3.603 E+3 Sm-151 9.316 E+1 1.51

Cr-51 7.590 E-2 3.002 E+3 Sn-126 1.000 E+5 7.01 E-1

Te-129m 9.340 E-2 1.170 E+3 Se-79 6.500 E+4 5.52 E-1

H-3 1.226 E+1 5.510 E+2 I-129 1.570 E+7 3.90 E-1

Fe-59 1.220 E-1 5.120 E+2 Zr-93 1.500 E+6 1.11 E-1

Eu-152 1.360 E+1 5.110 E+2 I-131 2.200 E-2 2.00 E-3

Am-242m 1.505 E+2 4.900 E+2

Total Activity = 1.00 E7 curies

NOTE: Attachment 1 is an inventory of the Spent Fuel Radioactivity decay corrected.

LAC-TR-138

PAGE 8

4.2 Core Internals/RX Components Radionuclide Inventory

Reactor components in and near the reactor core during power operation become

radioactive due to nuclear interaction with the large neutron flux present in this

region. Most of the residual radioactivity is produced by n,γ reactions with the

atomic nuclei of the target material although n,P reactions, for example the

production of C-14 from N-14, are also significant.

The residual radioactivity in the materials of the various LACBWR reactor

components has been estimated using activation analysis theory and, where

available, actual data from laboratory analyses of irradiated metal samples. Best

estimates of neutron fluxes in and irradiation histories of specific components

were used in these calculations. Original material chemical compositions were

obtained from actual material certification records when readily available but

standard compositions for specified materials were used in some cases.

Radioactive decay of the activation product nuclides has been taken into account

to obtain the best estimate values for the residual radioactivity as of January 1,

1988.

As of January 1, 1988, the largest contribution (210,000 Ci) to the radioactivity

inventory in LACBWR activated metal components is the isotope Fe-55

(HL = 2.7y). The next largest contributor (101,000 Ci) is Co-60 (HL = 5.27y).

The activity of these two relatively short-lived isotopes will decrease very

significantly during the proposed SAFSTOR period. The major long-lived

contributor to the radioactivity inventory (10,700 Ci) is Ni-63 (HL =100y). The

activity of other activation product nuclides have been lumped together in two

categories, those with half lives less than 5 years and those with half lives greater

than 5 years. The group with HL <5y consists mostly of Zr-95 (64d) in Zircaloy

components and Cr-51 (27.7d) and Fe-59 (44.6d) in stainless steel components

along with small quantities of Sn-113 (115d), Sn-119 (293d), Sn-123 (129d), Hf-

175 (70d), Hf-181 (42.4d), W-181 (121d) and W-185 (75.1d). The group with

HL > 5y consists mostly of Ni-59 (8x104y) with small quantities of C-14 (5730y),

Zr-93 (1.5x106y), Sn-121 (50y), Cd-113 (14.6y), Nb-94 (2x104y) and Tc-99

(2.13x105y).

LAC-TR-138

PAGE 9

Core Internal/RX Component Radionuclide Inventory - January 1, 1988

Estimated Curie Content

Other Nuclides

Components

Co-60

Fe-55

Ni-63 T1/2 < 5y T1/2 > 5y

Total

In Reactor

Fuel Shrouds (72 Zr, 8 SS) 22,109 63,221 1,352 2,810 15 89,507

Control Rods (29) 4,886 4,826 817 24 15 10,568

Core Vertical Posts (52) 1,270 594 63 2,396 4 4,327

Core Lateral Support

Structure 9,108 21,477 770 105 8 31,468

Steam Separators (16) 33,439 78,851 2,826 386 30 115,532

Thermal Shield 1,443 3,402 123 17 1 4,968

Pressure Vessel 347 1,029 10 2 ~ 0 1,388

Core Support Structure 6,458 15,230 546 75 6 22,315

Horizontal Grid Bars (7) 173 408 15 2 ~ 0 598

Incore Monitor Guide Tubes 307 188 611 7 5 1,118

Total 79,540 189,226 7,133 5,824 84 281,807

In FESW

Fuel Shrouds (24 SS) 13,667 14,988 2,384 ~ 0 26 31,065

Fuel Shrouds (73 Zr) 918 1,007 95 27 3 2,050

Control Rods (10) 3,456 2,386 910 ~ 0 17 6,769

Start-up Sources (2) 3,177 2,285 156 5 3 5,626

Total 21,218 20,666 3,545 32 49 45,510

NOTE: Core Internals/RX Components were removed and disposed of in 2007.

LAC-TR-138

PAGE 10

4.3 Plant Loose Surface Radionuclide Inventory

A plant smear survey was performed of all accessible interior building surfaces in

an attempt to determine the amount of loose surface contamination in the plant.

The specific isotopic identification of the contamination was also determined.

Each smear was gamma scanned to determine not only a correlation factor in

µCi per DPM/100 cm2 but also the percentile of each radioisotope present in the

mixture. From previous, part 61, analysis of plant smears, it has been determined

that Fe-55 is the major beta emitter in the plant and is in approximately the same

percentage as Co-60. Fe-55 will be the only beta emitting isotope listed as a

contaminant. Alpha activity on the surfaces has been checked by the use of an

Internal Proportional Counter and has been found to be negligible and so will not

be considered. The survey did indicate that the major isotopes present in the

plant's loose surface contamination are the following isotopes:

Isotope 1/2 Life Co-60 5.27 years Cs-137 30.1 years Mn-54 312.2 days Fe-55 2.7 years It must be realized when reviewing the results of this survey that a 100% smearing of plant surfaces was not performed and therefore the following data is subject to significant error.

The majority of the loose surface contamination throughout the plant is found in the plant contaminated areas. The plant areas that were classified as contaminated areas at plant shutdown in 1987 are listed below.

a) Waste Treatment Building (1) decon area

(2) basement

(3) resin liner room (4) high level storage pit

b) Turbine Building (1) stop valve area

(2) full flow room

(3) condensate bay

(4) area under main condenser

(5) feed pump bed plates (6) tunnel - includes 4500 WT cubicle and crawlway to stack

c) Containment Building (1) 701 south by FESW

(2) mezz around core spray pumps

(3) west NI platform

(4) purification platform

(5) FESW IX cubicle

(6) purification pump area

(7) basement - includes UCRD platform, purification IX cubicle, retention tank cubicle, subbasement, and FCP cubicles.

LAC-TR-138 PAGE 11

PLANT LOOSE SURFACE CONTAMINATION - JANUARY 1988

Isotopes Present, in µCi Location Co-60 Cs-137 Mn-54 Ce-144 Co-57 Cs-134 Fe-55

Total Area µCi

Content

Turbine Building (TB)

a) Main Floor 0.83 0.07 -- -- -- -- 0.83 1.73

b) Mezzanine - including stop valve area 0.49 0.14 0.04 -- -- -- 0.49 1.16

c) Grade Floor - includes feedwater heater area 0.42 0.06 0.02 0.42 0.92

d) Tunnel 0.81 0.18 0.06 -- -- -- 0.81 1.86

Containment Building (CB)

a) Above grade 3.16 0.20 0.39 -- -- -- 3.16 6.91

b) Below grade 31.44 7.40 2.36 0.04 0.04 0.08 31.44 72.80

Waste Treatment Building

7.57 0.48 0.66 -- -- -- 7.57 16.28

Totals

44.72 8.53 3.53 0.04 0.04 0.08 44.72 101.66

LAC-TR-138 PAGE 12

4.4 Plant System Internal Radionuclide Inventory

The internal surfaces of many inplant systems were exposed to radionuclide contaminants during plant operation. To obtain the most accurate analysis of these systems, piping/component destruction would be necessary. This was determined to be an undesirable method of analysis at this time. A method of nondestructive sampling was developed. Each system that would be sampled was looked at and a nondestructive entry point was found. Once the system was open, the piping was dried and a 1 cm2 area was scraped to bare metal. As the scraping was being done, the corrosion layer that was being removed was vacuumed into a glass fiber filter 47 mm in diameter. This filter was then gamma scanned for radionuclide identification using the computer-based gamma spectroscopy unit connected to a GeLi detector. A sample of the crud layer from the bottom of the FESW was obtained using a vacuum system. This crud was mixed with the water and an aliquot was taken and gamma scanned. This was used to determine the activity in the FESW crud layer. Because of the inaccessibility of the bottom of the reactor vessel, the gamma spectrum obtained from the FESW sample was used as the gamma spectrum for any crud layer in the vessel bottom. Conversation with representatives from other facilities who have looked at their reactor vessel bottom indicates a very small crud layer in the vessel. Each sample was also alpha counted to determine the total alpha content. The individual alpha isotopic mixture was not determined. Alpha analysis was performed with a Canberra Internal Proportional counter. In systems where the piping/component radiation levels varied significantly, the radiation level of the sample area was found. The remainder of the system was surveyed and the piping/component area was classified by radiation level. This technique allowed the sampled area uCi/cm2 value to correspond with a radiation level. As the system radiation level varied, a uCi/cm2 radiation level value was used to proportion the system areas to that of the sample area, thus allowing a total system uCi content to be determined. After sampling, each system was returned to an as-found condition. Fe-55, will be the only beta emitting isotope to be considered in the plant isotopic content. This will be in the same percentage as that of the Co-60 found in the systems. The isotopic composition occurring in the piping systems consists of the major nuclides listed as follows: Isotope 1/2 life Co-60 5.27 years Cs-137 30.1 years Mn-54 312.2 days Fe-55 2.7 years This testing of piping systems is not an absolute value. Significant error can be associated with this sampling technique due to the inability to actually analyze a system component or pipe. The values listed are subject to error due to the required non-destructive sampling techniques used and the inability to sample all components and areas.

LAC-TR-138 PAGE 13

PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JANUARY 1988

Nuclide Activity, in µCi Plant System Fe-55 Alpha Co-60 Cs-137 Mn-54 Cs-134 Nb-95 Co-57 Zn-65

System Total µCi Content

CB Ventilation 1.6 E3 -- 1.6 E3 1.7 E2 1.3 E2 1.40 -- -- -- 3.5 E3

Offgas - upstream of filters

6.0 E2

--

6.0 E2

4.4 E4

1.0 E2

--

--

--

--

4.5 E4

Offgas - downstream of filters

6.5 E2

--

6.5 E2

8.3 E2

--

--

--

--

--

2.1 E3

TB drains

1.7 E4

4.0 E1

1.7 E4

5.0 E3

7.8 E2

--

--

--

--

4.0 E4

CB drains

3.8 E4

3.2

3.8 E4

2.4 E3

2.6 E3

--

--

--

--

8.1 E4

TB Waste Water

3.6 E3

6.8

3.6 E3

1.2 E2

1.5 E2

--

5.6 E1

--

--

7.5 E3

CB Waste Water

2.1 E5

7.9 E1

2.1 E5

2.3 E3

1.7 E4

--

--

1.3 E2

1.4 E3

4.4 E5

Main Steam

2.6 E5

2.9 E2

2.6 E5

--

2.0 E4

--

1.0 E3

--

--

5.4 E5

Turbine

9.3 E2

1.8

9.3 E2

2.0 E2

1.8 E2

--

--

--

--

2.2 E3

Primary Purification

8.9 E4

1.2 E1

8.9 E4

--

8.8 E3

--

--

--

2.1 E3

1.9 E5

Emergency Core Spray

1.8 E3

5.0

1.8 E3

1.1 E2

1.4 E2

--

--

--

--

3.9 E3

Overhead Storage Tank

1.3 E4

3.4 E1

1.3 E4

7.8 E2

9.5 E2

--

--

--

--

2.8 E4

Seal Inject

1.6 E3

3.8

1.6 E3

5.5 E1

1.5 E2

--

--

7.5

--

3.4 E3

LAC-TR-138 PAGE 14

PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JANUARY 1988 - (cont'd)

Nuclide Activity, in µCi Plant System Fe-55 Alpha Co-60 Mn-54 Co-57 Co-58 Zn-65 Other

System Total µCi Content

Decay Heat

1.0 E5 4.9 E2 1.0 E5 3.1 E4 1.6 E2 3.2 E3 3.5 E3 2.4 E5

Boron Inject

1.4 E5 6.6 E2 1.4 E5 4.2 E4 2.1 E2 4.3 E3 4.7 E3 3.3 E5

Reactor Coolant PASS

9.9 E3 4.6 E1 9.9 E3 2.9 E3 1.5 E1 3.0 E2 3.3 E2 2.3 E4

Alternate Core Spray

2.0 E4 9.4 E1 2.0 E4 5.9 E3 3.0 E1 6.1 E2 6.7 E2 4.7 E4

Shutdown Condenser

2.3 E5 1.1 E3 2.3 E5 6.9 E4 3.5 E2 7.1 E3 7.8 E3 5.5 E5

Control Rod Drive Effluent

1.5 E5

7.2 E2

1.5 E5

4.6 E4

2.3 E2

4.7 E3

5.1 E3

3.6 E5

Forced Circulation

1.5 E6 7.0 E3 1.5 E6 4.4 E5 2.3 E3 4.5 E4 5.0 E4 3.5 E6

Reactor Vessel and Internals

2.5 E6

1.2 E4

2.5 E6

7.6 E5

3.9 E3

7.8 E4

8.6 E4

5.9 E6

Condensate after beds & Feedwater

2.1 E5

2.8 E2

2.1 E5

3.2 E4

--

1.6 E3

3.1 E3

4.6 E5

Condensate to beds 3.9 E4 3.1 E1 3.9 E4 1.3 E4 1.5 E1 6.3 E2 6.7 E2 Fe-59 = 5.2 E2 9.3 E4

Nb -95 = 1.1 E2

Ru-103 = 4.9 E1

Ce-144 = 1.2 E2

LAC-TR-138 PAGE 15

PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - JANUARY 1988 - (cont'd)

Nuclide Activity, in µCi Plant System Fe-55 Alpha Co-60 Mn-54 Cs-137 Ce-144 Zn-65 Other

System Total µCi Content

Fuel Element Storage Well System

8.5 E5

3.9 E2

8.5 E5

1.4 E4

--

1.0 E4

--

1.7 E6

Fuel Element Storage Well - all but floor

1.3 E3

4.9

1.3 E3

6.0 E2

4.6 E3

--

4.5 E2

8.3 E3

Fuel Element Storage Well floor

2.6 E7

7.6 E3

2.6 E7

5.0 E5

4.1 E4

--

1.1 E5

Cs-134 = 1.3 E2 Co-58 = 1.3 E2

5.3 E7

Resin Lines 1.3 E5 1.0 E2 1.3 E5 4.2 E4 -- 4.0 E2 2.2 E3 Fe-59 = 1.7 E3 Co-57 = 4.8 E1 Co-58 = 2.1 E3 Nb-95 = 3.5 E2 Ru-103 = 1.6 E2

3.1 E5

Main Condenser 1.1 E7 8.5 E3 1.1 E7 3.6 E6 -- 3.4 E4 1.9 E5 Fe-59 = 1.4 E5 Co-57 = 4.1 E3 Co-58 = 1.7 E5 Nb-95 = 3.0 E4 Ru-103 = 1.4 E4

2.6 E7

NOTE: Attachment 3 is an inventory of the plant system - Internal Radionuclide Inventory decay corrected.

LAC-TR-138 PAGE 16

4.5 Reactor Biological Shield Activation Survey

During 1993 LACBWR revised the decommissioning cost study. It was felt necessary at this time that a determination be made as to the extent of activation of the biological shield. The amount of biological shield activated will directly effect the cost of removal and disposal. On October 20, 1993, an outside contractor completed boring of the biological shield. The initial boring site had to be abandoned due to the presence of shield cooling system piping. This hampered the drilling and at one point caused the bit to become stuck. The boring site was moved approximately two feet below the original site. The boring through the biological shield was surveyed and analyzed on October 21, 1993. The borings were removed in several sections and are designated as below. These are listed as from the outside of the shield wall to the inside. These boring sections are being kept on the West TB Grade for further analysis. 1A-2A - 9.5 inches long 2A-3A - 14 inches long 3A-4A - 27 inches long 4A-5A - 15 inches long 5A-6A - 2 inches long 6A-7A - 39 inches long Section 6A-7A, being closest to the reactor, was surveyed using an RO-3 portable radiation detector to determine activities present. The following dose rates were obtained. These are listed as distances from the inside shield wall. 0 inches - 130 mrem/hour 6 inches - 130 mrem/hour 12 inches - 55 mrem/hour 18 inches - 16 mrem/hour 24 inches - 6 mrem/hour 30 inches - 2.5 mrem/hour 36 inches - 1.2 mrem/hour 39 inches - <1 mrem/hour A smear of this core section was taken with the following results: Co-60 8.42 E-3 µCi/smear Eu-152 1.27 E-2 µCi/smear No other surveys of this core were taken due to the fact that the dose reading indicated complete activation of this section.

LAC-TR-138 PAGE 17

One inch sections of the remaining cores were cut and gamma-scanned to determine the depth of the concrete activation.

Distance from inside shield wall

Co-60 (µCi/sample)

Eu-152 (µCi/sample)

39 inches 1.79 E-3 9.57 E-4

42 inches 4.49 E-4 3.82 E-4

48 inches 1.77 E-4 3.67 E-4

49 inches 8.13 E-5 2.36 E-4

52 inches 2.01 E-5 9.08 E-5

56 inches NP NP

Surveys of these sections with a frisker indicated <100 CPM increase over background. Smears indicated <MDA loose surface contamination. Analysis indicates that the biological shield is activated at a distance of 56" from the inside of the shield wall.

LAC-TR-138 PAGE1 8

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50 40 30 20 10 0

BIOLOGICAL SHIELD CORE RADIATION SURVEY

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LAC-TR-138 PAGE 19

4.6 Circulating Cooling Water Outfall Area Radionuclide Inventory As part of LACBWR's dismantlement cost study it was determined that the extent of contamination in the circulating cooling water outfall area be determined. During February 1994 EMS (Environmental Marine Services) was contracted to send divers to LACBWR to perform a sampling survey in the outfall area. This sampling survey would be used to estimate the radionuclide inventory that exists in the plant outfall. The divers report that the area from the Mississippi River's edge to approximately 60' from the edge is covered with large rip rap and no sample could be obtained. From the 60' area a silt sample was taken approximately every 10' out and 110' from the rivers edge. The divers then turned downstream with the river current. They continued to obtain silt samples every 10' out to 150' from the outfall. The following table lists the activities found in these samples.

ρCi/kg as of February 1994

Sample Distances

# From Outfall (ft) Co-60 Cs-137

1 60 1,070 1,980

2 70 1,260 2,200

3 80 1,330 2,230

4 90 813 1,870

5 100 216 362

6 110 82.3 127

7 120 8.96 21.4

8 130 NF 7.22

9 140 NF 20.11

10 150 9.33 22.6

(NF = None Found)

An attempt to obtain samples at various depths was tried. The divers didn't have the necessary equipment to perform this sampling so cross contamination occurred. Because of this no determination of the depth of the contamination can be made. The data we presently have would indicate a clean up area 20' wide by 95' long.

4.7 Initial Site Soil Contamination Determination Survey

In accordance with NUREG/CR-5849 "Manual for Conducting Radiological Surveys in Support of License Termination" an initial site soil survey was performed during August - September 1995. The LACBWR affected area (area within the LSE fence) was marked out into 10 m2 areas for sample location identification. No samples were taken under surface coverings such as blacktop. After gridding of the area was complete one surface soil sample was collected from the top 15 cm of the soil in each grid location. Each sample collected was approximately 1 - 2 kg.

LAC-TR-138 PAGE 20

Grass, rocks, sticks and foreign objects were removed from each sample to the degree practical. Each sample was then dried and placed in a disposable marinelli container for gamma analysis. All samples were allowed to sit for at least 72 hours to reduce the affect of any short lived natural decay chains. All samples were then counted for 3 hours on the ENV HPGE. The following is a listing of activity found in these samples. No natural occurring isotope is listed. The actual sample gamma scans are being kept for further review if needed. All sample locations can be found using their corresponding number using the enclosed site grid map.

NOTE: All activity values are in ρCi/kg

Sample Grid #

Co-60

Cs-137

Sample Grid #

Co-60

Cs-137

Eu-155

1 10.5 106 27 13.9 58.4

2 34.8 81.6 28 <MDA 78.7

3 2.9 126 29 <MDA 74.1

4 16.7 78.5 30 <MDA 68.3

5 4.8 92.1 31 <MDA 64.2

6 8.8 80.2 32 <MDA 27.7

7 <MDA 51 33 31.2 129

8 <MDA 12.2 34 <MDA 171

9 <MDA 23.2 35 <MDA 91.7 53.5

10 <MDA 38.9 36 <MDA 61 20.3

11 <MDA 71 37 40.9 124 40.4

12 <MDA 93.3 38 <MDA 44.6

13 767 983 39 <MDA 130

14 100 312 40 <MDA 49

15 5.8 58.9 41 <MDA 74.9

16 <MDA 22.5 42 <MDA 55.6

17 <MDA 15 43 20.7 211

18 <MDA 14.4 44 9.1 145

19 <MDA 43.5 45 36.5 310

20 <MDA 44.2 46 19.5 197 30.8

21 <MDA 72.6 47 85.1 177 27.7

22 <MDA 55.6 48 12 158

23 <MDA 45.5 49 56.4 329

24 <MDA 67.9 50 23.8 99

25 7.4 31 51 19.3 192 33.4

26 10.7 28.7 52 17.4 353 11.9

LAC-TR-138 PAGE 21

Sample Grid # Co-60 Cs-137 Eu-155

53 6.2 291 23

54 8.6 150

55 19.1 307 11.4

56 11.3 213 40

57 28.9 367 56.9

58 27 109

59 12.8 84.7 32.8

60 <MDA 30.7

61 48.5 137

62 76.9 206

63 28.3 226

64 <MDA 200

65 <MDA 122

66 26.7 66.8

67 19.2 185 27.3

68 276 1,300 52.2

69 44.8 252 26

SITE SOIL SAMPLING GRID MAP

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LAC-TR-138 PAGE 22

LAC-TR-138

PAGE 23

The unaffected area around LACBWR but within the owner controlled area was also checked. These samples were taken and analyzed as the samples taken from the affected area. The following results were obtained.

Location Co-60 Cs-137

Area W of #2 warehouse <MDA 77.3

Area S of parking lot <MDA 25.5

Area N of LACBWR admin building <MDA 149

Area by G-3 ash silo <MDA 38.2

Area outside G-3 offices <MDA 76.3

Area by G-3 outfall <MDA 17.2

All results are in ρCi/kg

To determine potential background Cs-137 levels for the area surrounding LACBWR several soil samples were taken at various locations outside the owner control area. These samples were collected and analyzed as per NUREG/CR-5849 as all the other samples. The following results were obtained.

Location Co-60 Cs-137

3.3 miles S @ boat landing along the Bad Axe River

22.7

381

Pedretti Farm Substation E of plant

<MDA

227

Radio Tower NE of plant <MDA 35.00

Junction of Cty Hwy O and K <MDA 188.00

E of Stoddard @ Junction of Cty Hwy O and 162

<MDA

66.5

NOTE: All activities are in ρCi/kg

LAC-TR-138

PAGE 24

ATTACHMENT 1

SPENT FUEL RADIOACTIVITY INVENTORY

Decay-Corrected to October 2009

Radionuclide

Half Life (Years)

Activity (Curies)

Radionuclide

Half Life (Years)

(Curies)

Ce-144 7.801 E-1 9.94 E-3 Sr-90 2.770 E + 1 6.65E+5

Cs-137 3.014 E+1 1.01 E+6 Pu-241 1.429 E+1 3.98E+5

Ru-106 1.008 E+0 0.46 Fe-55 2.700 E+0 1.94E+3

Cs-134 2.070 E+0 220 Ni-59 8.000 E+4 287

Kr-85 1.072 E+1 2.83E+4 Tc-99 2.120 E+5 276

Co-60 5.270 E+0 3.62E+3 Sb-125 2.760 E+0 1.14

Pm-147 2.620 E+0 128 Eu-155 4.960 E+0 7.95

Ni-63 1.000 E+2 3.04E+4 U-234 2.440 E+5 63.7

Am-241 4.329 E+2 1.42E+4 Am-243 7.380 E+3 61

Pu-238 8.774 E+1 1.06 E+4 Cd-113m 1.359 E+1 5.85

Pu-239 2.410 E+4 8.83E+3 Nb-94 2.000 E+4 15.9

Pu-240 6.550 E+3 7.15E+3 Cs-135 3.000 E+6 14.0

Eu-154 8.750 E+0 713 U-238 4.470 E+9 12.2

Cm-244 1.812 E+1 1.56E+3 Pu-242 3.760 E+5 8.58

H-3 1.226 E+1 160 U-236 2.340 E+7 6.32

Eu-152 1.360 E+1 168 Sn-121m 7.600 E+1 3.64

Am-242m 1.505 E+2 443 Np-237 2.140 E+6 2.19

U-235 7.040 E+8 1.89

Sm-151 9.316 E+1 1.28

Sn-126 1.000 E+5 0.7

Se-79 6.500 E+4 0.552

I-129 1.570 E+7 0.39

Zr-93 1.500 E+6 0.111

Total Activity = 2.18 E6 Curies

LAC-TR-138

PAGE 25

ATTACHMENT 2

CORE INTERNAL/RX COMPONENT RADIONUCLIDE INVENTORY

Estimated Curie Content

Other Nuclides

Components

Co-60

Fe-55

Ni-63 T1/2 > 5y

Total

In Reactor

Fuel Shrouds (72 Zr, 8 SS)

Control Rods (29)

Core Vertical Posts (52)

Core Lateral Support Structure

Steam Separators (16)

Thermal Shield

Pressure Vessel

Core Support Structure

Horizontal Grid Bars (7)

Incore Monitor Guide Tubes

Total

In FESW

Fuel Shrouds (24 SS)

Fuel Shrouds (73 Zr)

Control Rods (10)

Start-up Sources (2)

REACTOR VESSEL WAS

PROCESSED, PACKAGED AND

DISPOSED OF IN 2007

“IN FESW” COMPONENTS LISTED

WERE PROCESSED, PACKAGED

AND DISPOSED OF IN 2006

LAC-TR-138

PAGE 26

ATTACHMENT 3

PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY - OCTOBER 2009

Nuclide Activity, in µCi Plant System Fe-55 Alpha Co-60 Cs-137

System Total µCi Content

CB Ventilation 8 -- 104 105 217 Offgas - upstream of filter

SYSTEM

REMOVED Offgas - downstream of filters

SYSTEM

REMOVED

TB drains 82 40 1,103 3,097 4,322

CB drains 182 3 2,466 1,487 4,138

TB Waste Water 17 7 234 74 332

CB Waste Water 1,008 79 13,625 1,425 16,137

Main Steam 1,249 290 16,869 18,408

Turbine 4 2 60 124 190

Primary Purification 427 12 5,775 6,214

Emergency Core Spray SYSTEM REMOVED

Overhead Storage Tank 62 34 843 483 1,422

Seal Inject 8 4 104 34 150

LAC-TR-138 PAGE 27

ATTACHMENT 3

PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY – OCTOBER 2009

- (cont'd)

Nuclide Activity, in µCi System Total µCi Content

Plant System

Fe-55 Alpha Co-60 Decay Heat

480

490

6,488

Boron Inject

SYSTEM REMOVED

Reactor Coolant PASS

SYSTEM REMOVED

Alternate Core Spray

96 94 1,298

Shutdown Condenser

SYSTEM REMOVED

Control Rod Drive Effluent

720 720 9,732

Forced Circulation

7,203 7,000 97,323

Reactor Vessel and Internals

SYSTEM REMOVED

Condensate after beds & Feedwater

SYSTEM REMOVED

Condensate to beds

SYSTEM REMOVED

7,458

1,488

11,172

111,526

LAC-TR-138 PAGE 28

ATTACHMENT 3

PLANT SYSTEMS INTERNAL RADIONUCLIDE INVENTORY – OCTOBER 2009

- (cont'd)

Nuclide Activity, in µCi Plant System Fe-55 Alpha Co-60 Cs-137

System Total µCi Content

Fuel Element Storage Well System

4,082

390

55,150

59,622

Fuel Element Storage Well

- all but floor

6

5

84

2,854

2,949

Fuel Element Storage Well floor

124,859 7,600 1,686,940 25,398 1,844,797

Resin lines

624 100 8,435 9,159

Main Condenser

52,825 8,500 713,705 775,030

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