ML20138P436
| ML20138P436 | |
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|---|---|
| Site: | 05000187 |
| Issue date: | 12/31/1985 |
| From: | Crandall W NORTHROP CORP. |
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Text
_
O bl TRIGA REACTOR FACILITY DECOMMISSIONING PROGRAM FINAL REPORT NRC LICENSE NO. R-90 DOCKET NO. 50-187 DECEMBER 1985 Prepared by
" E C'"""
O with major contributions by J. Benveniste, G. Cozens, J. Chalmers and H. Woo l
NORTHROP 840 u Pa fa Los Angeles, Cahfornia 90067-2199 O
p2=mgmanh7
CONTENTS SECTION TITLE PAGE 1
INDEX......................................
1-1 2
TEXT.......................................
2-1 3
TABLES.....................................
3-1 4
FIGURES....................................
4-1 5
PHOTOS.....................................
5-1 DECOMMISSIONING DOCUMENTATION..............
1-2 BIBLIOGRAPHY...............................
1-3 1-1
(
DECOMMISSIONING DOCUMENTATION D-NO SUBJECT 2.1.1 DECOMMISSIONING PLAN 1
2.1.2 ENVIRONMENTAL PLAN 2.1.3 QUALITY ASSURANCE PROGRAM 1
3.0.1 NORTHROP LICENSE - NRC/90 (D-50-187) 4.2.1 ENVIRONMENTAL SURVEY REPOPT l
4.3.2 SUMP WATER DISPOSAL 5.1.1 FUEL STORAGE CRITICALITY ANALYSIS 5.1.2 RADIATION PROTECTION FOR FUEL CASK LOADING 5.1.3 FUEL CASK LOADING AND TRANSPORTATION l
j 5.1.4 TRANSPORTATION OF SPECIAL NUCLEAR MATERIAL 5.1.5 HEALTH PHYSICS HANDLING REACTOR STRUCTURE l
5.3.1 POOL WATER DISPOSAL 5.4.1 RADIATION MONITORING PROGRAM-NORTHROP 5.4.2 RADIATION MONITORING PROGRAM-CHEM-NUCLEAR 5.4.3 SAFETY PROGRAM-NOPTHROP O
5.4.4 SAFETY PROGRAM-CHEM-NUCLEAR 5.4.5 DUST ABATEMENT PROGRAM
.,i 5.4.6 BASELINE RADIOLOGICAL SURVEY i
5.4.7 CONTROLLED AREA RADIATION PROTECTION
'l 5.4.8 CONTAINMENT / VENTILATION SYSTEM-POOL AND l
EXPOSURE ROOM 5.4.9 EXPOSURE ROOM MATERIAL DISPOSAL 5.4.10 REACTOR POOL CONCRETE REMOVAL 5.4.11 BEAM PORT REMOVAL 1
5.7.1 RADIOLOGICAL SURVEY OF VENTS-WATER PIPES 5.8.1 WASTE MATERIAL l
6.2.1 RADIOLOGICAL SURVEY 4
I i
t i
1-2
(
I BIBLIOGRAPHY O
BIB. NO SUBJECT
[
f 4.2.1 The Elements of Nuclear Reactor Theory, j
Glasstone and Edlund, Van Nostrand, 1952 t
4.2.2 D.C. Kocher, Radioactive Decay Data Tables, j
Technical Information Center U.S. Department i
of Energy, 1981
{
4.2.3 Marion and Young, Nuclear Reaction Analysis, North-Holland Publishing Co.,
1968 4.2.4 U. Fano et al, Penetration and Diffusion of l
X-Rays, Volume 38/2, Handbuch der Physik, 1959 j
4.4.1 Tree simulation, Northrop Reactor Facility, l
Oct 1973 l
i 4.4.2 W. Crandall, Neutron Tomography Experiment,
[
1983 4.4.3 Handbook of Applied Engineering, CRC Press, 1973 l
4.5.1 Irvine, Nuclear Structure Theory, Pergamon Press, j
1972 j
4.5.2 Radiologieni Health Handbook, U.S. Department l
HEW (Fevised Edition Jan 1970) 4.5.3 D.C. Kocher and A.L. Sjorcen, Dose Rate convernion Factors for External Exposure to Photon Emitters in Soil, Health Physics, Volume. 48, No. 2 (Feb 1985) 5.9.1 Geologic and Hydrologic Investigation of Puente Hills Landfill Site, Leroy Crandall Associates (1981) l 5.9.2 Climatological Data, NOAA (1984) 6.2.1 C.F. Holoway et al, Monitoring for Compliance with Decommissioning Termination Survey criteria, NUREG/
l CR-2082 i
l 6.2.2 California Radiation Control Regulation, Titin 17, 1980 I
l 6.2.3 dBASE III, Version 1.1 Manual, Ashton Tate, 1984 l
6.2.4 ABSTAT, Release 4, Anderson Hell, 1984 O
1-3 t.
l l
l t
o TEXT PARA. NO SECTION PAGE 1.0 ABSTRACT....................................
2-3 l
2.0 INTRODUCTION
2-6 t
l l
2.1 ORGANIZATION................................
2-6 2.2 PLANNING....................................
2-7 2.3 EXECUTION...................................
2-8 2.4 REPORTING AND DOCUMENTATION.................
2-8 3.0 FACILITY DESCRIPTION........................
2-9 3.1 TRIGA MARK F REACTOR........................
2-9 3.2 COBALT-60 SOURCE HOT CELL...................
2-10 3.3 FLASH X-RAY.................................
2-10 1
3.4 SUPPORT AREAS...............................
2-11 4.0 RADIOACTIVE SOURCES.........................
2-11 i
4.1 SEALED NON-FUEL SOURCES.....................
2-11 4.2 REACTOR FUEL SOURCE.........................
2-12 4.3 HISTORICAL RADIOLOGICAL SURVEY DATA.........
2-13 4.3.1 POOL WATER RADIOACTIVITY....................
2-13 4.3.2 SUMP WATER DISPOSAL DATA....................
2-14 4.3.3 ARGON-41 EMISSION DATA......................
2-14 l
4.3.4 ENVIRONMENTAL SURVEY DATA...................
2-15 4.4 NEUTRON ACTIVATION OF CONCRETE..............
2-15 4.5 PRIMORDIAL RADIOACTIVITY IN CONCRETE........
2-21 5.0 DECONTAMINATION OF FACILITY.................
2-22 l
5.1 REACTOP DEFUELING...........................
2-24 5.2 REACTOR HARDWARE REMOVAL AND DISPOSITION....
2-26 5.3 POOL WATER REMOVAL..........................
2-26 5.4 EXPOSUFF. ROOM AND POOL AREA DECONTAMINATION.
2-27 4
2-1
TEXT (Continued)
\\
PARA. NO SECTION PAGE 5.5 COBALT-6 0 S OURCES HOT CELL..................
2-29 5.6 RADIOCHEMISTRY LABORATORY DECONTAMINATION...
2-29 5.7 DUCTS AND PLUMBING DECONTAMINATION..........
2-30 5.8 CONTAMINATED WASTE DISPOSAL.................
2-30 5.9 UNCONTAMINATED CONCRETE DISPOSAL............
2-31 6.0 RADIOLOGICAL SURVEY.........................
2-32
6.1 BACKGROUND
PADIOACTIVITY.,..................
2-32 6.2 SITE SURVEY CRITERIA........................
2-33 6.3 ANALYSIS METHODS............................
2-35 6.4
SUMMARY
OF SURVEY DATA......................
2-36 i
7.0 CONCLUSION
2-37 i
l
[
[
i l
I f
f t
}
l 2"2 l
NORTHROP TRIGA REACTOR PACILITY DECOMMISSIONING PROGRAM FINAL REPORT 1.0 ABSTRACT The key tasks involved in the decommissioning of the Northrop Triga Reactor Facility were :
to defuel the reactor and ship the fuel rods, under the auspices of the U.S. Department of Energy, to selected Universities ; to dispose of ancillary equipment and instrumentation by means of gifts to Universities ;
to determine the neutron activation of materials in the vicinity of the reactor and to remove, package and transport these radioactive material to an approved burial site ; and to complete a final radiological survey of the facility in order to show the facility is safe for unrestricted use.
The fission products created in the fuel were the major radiological hazard (1100 curies).
Af ter several dry runs in which our transfer procedures were practiced and perfected, the fuel elements and other components of the reactor were packaged and shipped to several Universities without incident.
The radioactive support structure of the reactor was packaged and shipped to the Hanford burial site.
The neutron fluence created less than one curie of radio-activity in the materials in the exposure room.
The radioactive 2-3
aluminum window, a twenty four inch layer of concrete in the
/
exposure room, and the steel rebar support structure were removed, packaged and shipped to the Banford burial site.
The capture neutrons created less than 0.1 curies of radioactivity in the materials in the region of'the beam ports.
The beam ports and the adjacent activated concrete were removed, packaged and shipped to the Banford burial. site.
The radioactivity in the remaining concrete structure, weighing approximately 1000 metric tons, is overwhelmingly due to natural primordial radioisotopes - a billion primordial radio-isotopes for each neutron capture radioisotope.
This concrete structure is not a radiological hazard and will be buried at 'a conventional dump site, or given to a purveyor of freeway paving materials.
l Gaseous effluents, primarily Argon-41, were monitored during the reactor operating life. The measured levels of radioactivity 3
averaged 300 pC1/m.
This is well below the permissible limits 3
of 40,000 pCi/m for emission into the atmosphere.
The stack structure showed no radioactivity at decommissioning time.
The sump water was monitored during the reactor operating period.
The levels of radioactivity were always well below acceptable levels for normal waste water.
After a final check, the water was disposed of as normal waste water.
2-4
Environmental surveys were made during the operating period of the reactor out to a distance of five miles.
No radiological effects were observed other than variations due to normal climatic background radiation.
All support areas in the Facility were surveyed and any contaminated material was removed, packaged and shipped to the Banford burial site.
With the dismantling of the reactor and the decontamination of the facility the block survey shows no radioactivity above the permissible limits.
O O
2-5 i
\\
)
210 INTRODUCTION Northrop management decided to decommission the Reactor Facility in mid-1984.
The decision was based on the belief that the ' f acility was not ersential to Northrop's primary business activity and that there was a more effective use for the build-ing.
A decommissioning organization was formed and the necessary planning initiated, as described in the following sections.
2.1 ORGANIZATION Figure 2.1.1 shows the organization formed to decommission the Reactor.
The organization was made up of personnel in the O '
' Corporation with knowledge and expertise considered important to the program.
The. head of this organization reported to the Corporate Vice President Research and Technology.
The individ-uals in the Decommissioning Organization met at least monthly to report their progress to the other menbers of the organization, to approve procedures for various stages of the decommissioning process, and to discuss plans for coming operations.
An exec-utive group composed of f our key personnel from the Decommis-sioning Organization devoted full time to the program and met weekly with the Decommissioning Supervisor to report on day-to-1 day progress and to coordinate activities.
l All plans and procedures for the operation related to radiological safety were reviewed and approved by the Reactor 2-6 j
l
Safeguard Subcommittee and the Corporate Radiation Committee.
2.2 PLANNING The Federal, State and local government regulations relevent to the decommissioning of the Reactor Facility were assembled for review, and are listed in Table 2.2.1.
A Decommissioning Plan, an Environmental Impact Report, and a Quality Assurance Program Report, were prepared and submitted to the Nuclear Regulatory Commission for review and approval (Doc 2.1.1).
The Decommis-sioning Schedule and the Man Power Loading are shown in Figure 2.2.1.
Potential recipients for the fuel, the reactor hardware, and the instrumentation were contacted.
Arrangements were made for the disposal of radioactive waste at approved burial sites.
The planning for the decontamination of the facility was based on the requirements given in Tables 2.2.3.
The decission was made to use a subcontractor to assist in the decommissioning operation.
Chem-Nuclear Corporation was selected to assist in the dismantling, decontamination, and radiological survey operation.
A fenced area, shown in Figure 2.2.2, was prepared adjacent to the facility for the temporary storage of decontaminated material.
Additional work areas inside the Facility were cleared for the dismantling and decontamination operations.
Of fice space was made available for the subcon-tractor's activity.
Specialized equipment, listed in Table 2.2.4, needed for the dismantling operation was brought on site.
2-7
The site preparation tasks are listed in Table 2.2.5.
n V
2.3 EXECUTION
)
The defueling operation, the decontamination operation, and the radiological survey operation are described in the main sections of the report.
2.4 REPORTING AND DOCUMENTATION In addition to the planning reports, all critical operations required written procedures approved by the Reactor Safeguard Subcommittee and the Corporate Radiation Committee.
Radiation exposure records were maintained on all personnel working in the facility.
Radiological survey records were kept for the decon-tamination and final survey.
Fuel transfer and radiological waste records were prepared.
Historical records related to the reactor operation will remain on file.
The final report summa-rizes the decommissioning activity.
O 2-8
O 3.0 FACILITY DESCRIPTION The Northrop Reactor Facility was located at the Northrop Hawthorne complex in Los Angeles County (see Fig 3.0.1 thru Fig 3.0.3).
A Utilization Facility License (R-90) was granted to the Northrop Corporation in March 1963 (Doc 3.0.1).
An exterior photograph of the Facility is shown in Photo 3.0.1, and interior photographs of the Facility are shown in Photo 3.0.2.
The Reactor Pacility included the TRIGA Mark F Reactor, a Cobalt-60 Source Hot Cell, a Flash X-ray machine, and support facilities.
An artist view of the reactor is shown in Figure 3.0.4.
Figure 3.0.5 shows a vertical section through the reactor, and Figure 3.0.6 shows a plan view of the reactor structure.
3.1 TRIGA MARK F REACTOR The Northrop Reactor was a research reactor used primarily for the investigation of the effects of nuclear radiation on electronic components, and on materials.
The reactor operated forty-five percent of the time at the exposure room position (see Fig 3.0.6).
The arrangement of the fuel elements and the control rods in the core are shown in Figure 3.1.1, and the configuration of a fuel element is shown in Figure 3.1.2.
The characteristics of the reactor are listed in Table 3.1.1.
2-9
The reactor generated a little over 4400 gigajoules (1200 megawatt-hours) from the time of startup in March 1963 to the shutdown in December 1984.
The fraction of the total energy generated per year is shown in Table 3.1.2.
3.2 COBALT-60 SOURCE BOT CELL The cobalt-60 source array contains 24 individual sources with a total source strength of 1828 curies, as of June 1984.
Twelve of these individual sources contain 335 curies and the remaining 12 sources contain 1493 curies.
In the most compact configuration of the sources the maximum deliverable dose rate was 180 kiloarads/hr.
The source was located in the hot cell adjacent to the exposure room (see Figure 3.0.6).
As expected, Os i
there was no activation of the concrete structure by this gamma source.
3.3 FLASH X-RAY The FEBETRON-705 Flash X-ray machine (see Figure 2.2.2) gen-erates an electron beam with a maximum energy of 2.3 MeV and a pulse length of approximately 30 nanoseconds.
The Flash X-ray is primarily used for transient radiation effects studies.
The machine and the control electronics were removed from the facility and stored until a new facility at another Northrop site is completed.
O 2-10
3.4 SUPPORT AREAS Located within the reactor building was a radiochemistry laboratory with radiochemical fume hoods, hot storage wells, hot hold tanks, and associated chemical laboratory equipment (see Figure 2.2.2).
There was also a machine shop, a counting room, storage vaults, a personnel decontamination area, and a pharmacy laboratory.
The decontamination and the radiological survey of these areas will be covered in the appropriate sections of the 4
report.
4.0 RADIOACTIVE SOURCES Radioactive sources located at the Northrop Reactor Facility at the start of the decommissioning operation were of two categories : (1) Sealed non-fuel sources ; (2) Sources created by the fissioning of Uranium-235 in the operation of the reactor.
4.1 SEALED NON-FUEL SOURCES Several sealed sources were stored in the reactor facility.
Descriptions of these sources and their disposition are given in Table 4.1.1.
The Cobalt-60 sources, described in Section 3.2, were removed f rom the hot cell and place] in temporary storage.
They will be installed in a new Northrop facility.
O 2-11
4.2 REACTOR PUEL SOURCE The reactor created 54 grams of fission products during the 22 years the reactor operated.
Table 4.2.2 lists the various burnup products created during the reactor operating period.
The reactor created approximately 0.57 mole of neutrons during the 22 year operating period.
Approximately 35 percent of these neutrons were used to produce new fissions and an addition-al 5 percent created U-236.
Table 4.2.3 shows that less than one percent of the neutrons were captured by U-238 to produce U-239 which decays to Pu-239.
All the plutonium was created and retained inside the welded fuel rods.
All the fission products produced by the reactor fuel are also contained inside the welded, stainless steel jacketed fuel rods.
Ninety nine percent of these fission products are now stable isotopes.
The remaining one percent of the fission products are primarily beta and gamma emitters.
The combined strength of the fission product sources at the time of dismantling, as derived in Table 4.2.4, was 360 curies of gamma rays and 720 curies of betas.
Most of the remaining sixty percent of the neutrons were captured by nuclei that created only prompt gamma radiation and short lived radioactve isotopes.
Less than 0.1 percent of the neutrons created long lived radioisotopes in the reactor structure.
These long lived radioisotope contributed less than 2 curies of radioactivity in the reactor structure and the reactor components.
2-12
The reactor created during its operation an expected small amount of gaseous radioactive products (e.g. Argon-41, Nitro-gen-16, Xenon-135, etc).
Most of these products decayed into stable isotopes in the fuel rods, in the reactor structure, in the pool water, or in the concrete of the exposure room.
The Argon-41 that escaped to the air was diluted with air from the room exhaust flow and ejected through the stack.
At all times during the reactor operation the radioactivity from Argon-41 was well below the allowable litit. The environmental sampling surveys of the surrounding area during the 22 years of operation detected no effects above the normal fluctuation in the natural background radioactivity created by seasonal and climatic conditions (e.g. see Doc 4.2.1).
O V
4.3 HISTORICAL RADIOLOGICAL SURVEY DATA During the operating life of the reactor radiological records were kept of any potential source of emission of rad-iation to the environment.
The following sections summarize the data that is relevent to the decommissioning operation.
4.3.1 POOL WATER RADIOACTIVITY A procedure was written for the removal and disposition of the pool water (Doc 4.2.1).
These procedures were approved by the Decommissioning Organization and the Corporate Radiation 2-13
Committee.
Table 4.3.1 lists the measured levels of the alpha, beta and gamma radioactivity in the pool water, after the removal of the reactor fuel and the reactor components.
The sum of the radioactivities was 3.92 pCi/ liter, which is below the measured level in natural rainwater, 5.42 pCi/ liter, shown in Table 4.3.6.
The water was pumped into the normal waste water system.
4.3.2 SUMP WATER DISPOSAL DATA Procedures were written for the survey and decontamination of the sump tanks (Doc 4.3.2).
These procedures were approved by the Decommissioning Organization and the Corporate Radiation Committee.
The waste water, with possible low level radioactiv-ity, was stored in a sump tank external to the reactor building.
This sump water was monitored throughout the operating life of the reactor and, with Los Angeles County Sanitation District approval, periodically disposed of in the sewer system.
Table 4.3.2 summarizes the sump water disposal history.
After the closure of the sump system, the sump tank was surveyed for possible radioactivity.
There was no contamination.
4.3.3 ARGON-41 EMISSION DATA Air from the facility was exhausted through the stack, shown in Figure 3.0.3 and Photo 3.0.1, to the atmosphere.
This air was
)
2-14
)
1 1
continuously monitored for Argon-41 emission during the reactor operation.
The analysis of the Argon-41 emission is given in Table 4.3.3 and the measurement data are summarized in Table 3
4.3.4.
The average concentration of the Argon-41 was 300 pCi/m,
3 with a maximum level of 840 pci/m.
These emissions were always 3
well below the maximum permissible level of 40,000 pCi/m (10CFR 20, appendix B).
4.3.4 ENVIRONMENTAL SURVEY DATA Environmental survey data obtained from samples of air, water, and dirt, taken out to a distance of five miles from the facility, were collected during the operating life of the reactor.
The sites where samples were obtained are identified in Table 4.3.5, and the locations shown in Figure 4.3.1.
A summary of the average specific activity versus year is tabulated in Table 4.3.6, and the average activity versus the month is given in Table 4.3.7.
The survey data during the 22 years of operation show no effects above the normal fluctuation in the natural background radioactivity, created by seasonal and climatic conditions.
4.4 NEUTRON ACTIVATION OF CONCRETE The neutron fluence created in the reactor structure and the surrounding regions is directly related to the total energy 2-15
generated by the reactor.
The thermal neutron fluence created in 2
the core of the reactor was 2.2 x 1019 n/cm.
The thermal neutron fluence versus distance from the core is shown in Figure 4.4.1, and the fast neutron fluence created by the reactor is shown in Figure 4.4.2.
Table 4.4.1 shows the neutron fluence, based on the above data, generated during the operating lifetime of the reactor at specific regions of the pool structure.
The fractions of the time the reactor operated at various positions in the pool are also listed in Table 4.4.1.
The neutrons diffuse through the surrounding structures, and interact with specific isotopes in the materials of these structures.
These neutrons create a diffuse, relatively long-lived net of radioactive sources in the materials.
The produc-tion of the distributed radioactive sources in the reactor structure and the surrounding materials is determined by the neutron capture reactions :
1)
Yb+n Xa + Prompt emission products 2)
X Ze + decay radiation products a
where X is a specific radioactive isotope created by a neutron a
captured in a base isotope, Y with the probability of the b,
O 2-16
}
reaction occurring given by the cross-section, C.
The net b
production rate of the isotope is given by:
Net Production Rate = Production Rate - Decay Rate 3) dNa/dt b Cb r
Na/T f
N
=
a where f is the neutron flux, N is the number of surviving r
a active isotope atoms, X Per em3, N is the number of base a'
b isotope atoms,Y, and T is the mean life of the active isotope.
b a
The activity, A' Per gram of material containing these isotopes a
is :
Na / (T R) 4)
A
=
a a
m 3
where R is the density in gm/cm of the material.
Solving m
eq. 3 for N and substituting the result into eq. 4 gives:
a F /R ) Z ((F /F )/T )EXP(-t /T l 5)
Aa = (Nb Cb r
m n
r a
n a
for 0<tn<Ta where F is the integrated neutron fluence (n/cm2), at the r
location of the material, during the operating life of the reactor, and F is the neutron fluence for a specific year, at n
time t before the shutdown of the reactor.
Defining a
n, parameter for the summation term:
OV 2-17
l 6)
H I (F /F ) (T /T )EXP(-t /T )
for 0<tn<To
=
a n
r o
a n
a where T is the operating life of the reactor, Equation 5 o
becomes:
7)
A N
F Ha/gT a
b Cb r
=
o The operating history of the reactor is shown in Table 3.1.2.
Also shown in this table are the fractions of the neutron fluence, F /F generated at the core of the reactor.
The n
r, neutron fluence in the pool water versus the radial distance from the reactor core is plotted in Figure 4.4.1.
The summation term, H versus lifetime, T,
are plotted in Figure 4.4.3.
a, a
Equation 7 can be expressed in terms of dimensionless parameters that are more directly related to the measured parameters, or the values are known and available from handbooks:
Gb a
b Qe d L
Au/N, 8)
A
=
H P
a e
r where values of the parameters are listed in Table 4.4.1 thru Table 4.4.8, and the definition of the parameters are :
1 A
is the specific activity of isotope, a,
in a
curies /gm (see Table 4.4.7-8).
Gb = N /N, is the naturally occurring abundance of b
- isotope, b.
Values of these parameters are known and published (see Table 4.4.2).
2-18
l j
H, is a factor dependent on the operating history of the reactor and the halflife, T,
Values for H, for specific radioisotopes can be found from Figure 4.4.3 (see Table 4.4.2).
J, = R,/R, is the weight fraction of the element of which isotope b is a component in the structural material of density R.
Values of 4
m these parameters were determined by chemical j
and spectral analysis of the structural i
materials (see Table 4.4.3).
Lr= (F /Fu) (O /O ) is the ratio of the fluence generated r
r u
j in the activated material at a specific location, to the fluence generated in the core of the reactor.
Figure 4.4.1 shows the ratio of the thermal neutron fluence versus radial distance from the core.
Table 4.4.1 lists the neutron fluence at the specific locations.
I Pb = C /C is the ratio of the thermal neutron cross-b u
section for the base isotope b to the cross section for U-235.
These cross sections for all isotopes of interest appear in Table 4.4.2.
Q, = M /M is the ratio of the gram atomic masses of the e
u base element to that of U-235.
These ratios for all the elements appear in Table 4.4.3.
A
=N C
Fu/M K
T is the average fission rate of u
o u
u c
o 4
i 2-19
_~
U-235 in the core, during the operating life of the reactor, expressed in curies per gram of U-235.
N is Avogadros number, Kc = 3.7 x o
10 10 (d/s)/Ci, and the remaining terms have already been defined (Table 4.4.2).
The calculated rate is 1.28 Ci/gm.
The important long-lived radioactive isotopes created in the reactor facility, the base isotope from which they were created by neutron reactions, and the materials in which they were distributed are listed in Table 4.4.2-4.
The specific activities created in these materials, calculated from the measured and known parameters, are listed in Tables 4.4.4-7.
These data form the basis for the removal of the radioactive materials created in the reactor facility.
The measured radioactivities produced by the capture of neutrons at several depths in the concrete of the exposure room are listed in Table 4.4.8.
The neutron activation versus depth in the concrete, based on these measurements, is shown in Figure 4.4.4.
The relaxation distance, 10 cm.,
is consistent with the handbook value for fast neutron attenuation (Bib 4.4.3).
The neutron induced radioactivity is comparable to the natural background radioactivity at approximately 22 inches.
The sources of this naturally ocurring radioactivity are primordial radioisotopes.
They will be discussed in the following section.
2-20
Dd 4.5 PRIMORDIAL RADIOACTIVITY IN CONCRETE The measured radioactivities due to naturally occurring primordial radioisotopes in the concrete are listed in Table 4.5.1.
This background radiation is generated by three primor-dial parent radioisotopes, potassium-40, thorium-232 and uran-ium-238.
Eighty nine percent of the potassium-40 decays are by beta emission to calcium-40, and eleven percent of the decays are by electron capture, or positron emission, creating energetic gammas, to argon-40.
Thorium-232 generates a series of alpha and beta decays ending with the stable isotope lead-208.
Uranium-238 generates a similar series of alpha and beta decays ending with lead-210.
Tables 4.5.2-8 describe the basic reaction products, the specific activities, and estimates of the radiation emitted from the concrete.
The specific activity created by neutron capture decreases exponentially from the inside surface of the exposure room, and becomes negligible compared to the natural background beyond a depth of 22 inches (see Figure 4.4.4).
For each radioisotope created by neutron capture in the remaining concrete there are more than a billion natural background primordial radioisotopes (see Table 5.9.1).
The measured radioactivities of naturally occurring primor-dial radioisotopes in the soil are also listed in Table 4.5.1.
The natural background radioactivity in the soil is comparable to the natural radioactivity in the concrete.
2-21 J
O The analysis and measurements contained in this section and the previous section will be used in the following sections to determine the volume of radioactive concrete that must be removed
'and shipped to the Hanford burial site.
5.0 DECONTAMINATION OF FACILITY As described in the Introduction, the site was prepared for the handling and temporary storage of materials and equipment needed to accomplish the dismantling operation.
These site preparations tasks are listed in Table 2.2.5.
Materials and equipment not directly related to the dismantling operation were removed from the facility.
The dismantling operations were planned and performed in a manner which assured that the radiation exposure of the dismantling team workers would be as low as reasonably achievable (ALARA).
The following factors in accord with the ALARA prin-ciple, and in accord with the existing utilization license for the facility, were adhered to throughout the dismantling oper-ation :
1)
The dismantling operation started four months after shutdown in order to reduce the level of radioactivity t
from the fuel elements.
3 2)
Experienced and well trained personnel were used in 1
2-22
the dismantling operation.
3)
A competent and well trained health physics team monitored all dismantling operations.
4)
Approved protective clothing was worn by the workers in the dismantling operation.
5)
The dismantling plan was reviewed and approved by the Dismantling Subcommittee and the Corporate Rad-iation Committee.
6)
The exposure of the workers to the most intense radia-tion sources was minimized by loading the fuel rods under water into a transfer cask.
The transfer cask was transported to, and inserted into, the shipping cask before the fuel was transfered to the shipping cask. This was done to shield the workers from exposure during the fuel transfer operation. Dry runs of the transfer operation were carried out to reveal any planning faults and to minimize the transfer time.
7)
The movement of radioactive dust was minimized by the use of enclosures, controlled atmosphere, and filtra-tion systems.
The collective radiation dose to all the workers for the entire dismantling operation was 0.630 man-rem.
i 2-23
O 5.1 REACTOR DEFUELING Procedures were written for the transfer and shipment of the fuel (Doc 5.1.1-5).
These procedures were approved by the Reactor Safeguards Subcommittee and the Corporate Radiation Committee prior to the transfer of the fuel. A criticality analysis was made for the storage of fuel in the shipping cask (Doc 5.1.1).
The fuel was transfered from the reactor core to the storage racks in the reactor pool.
The possible radiation hazard to the workers was determined by measuring the level of radiation emitted from an unshielded fuel element.
The measured radiation levels are listed in Table 5.1.1.
The fuel rods that were in the reactor at shutdown had an average radioactivity of 1400 mR/hr/ rod at three feet from the rod.
This agrees with the calculated level of 1430 mR/hr/ rod, based on the 4.1 Ci/ rod calculated from the operating history (see Table 4.2.4).
Several dry runs were performed for the fuel transfer operation, prior to the actual transfer, to assure the efficient and smooth transfer of the fuel.
Arrangements and approvals for the transport of fuel to the recipients, with the required security support, were completed.
A small transfer cask, shown in Fig 5.1.1 and Photo 5.1.1, was used to transfer the fuel to the BMI-1 shipping cask, shown in Figure 5.1.2 and Photo 5.1.2.
This action was taken to avoid possible contamination of the pool water by the shipping cask, 2-24
and also to avoid a potential hazard in lifting the heavy cask s
into the pool.
A plastic skirt, shown in Photo 5.1.3, was placed between the pool and the shipping cask on the trailer to guide and collect any water that might drip from the transfer cask.
Three fuel elements were transfered in each transit of the transfer cask to the shipping cask.
A long-handled transfer tool, shown in Photo 5.1.4, was used to load the fuel under water into the transfer cask.
The same tool was used in the removal of the fuel elements from the transfer cask to the shipping cask.
Thirty fuel elements were transferred and loaded into the shipping cask in approximately one hour.
The following personnel were used in the transfer operation:
AT POOL Health Physicist - measurements, and data handling Technician - fuel transfer Technician - assist in fuel transfer Crane operator AT BMI-l SHIPPING CASK Health Physicist - measurements, and data handling Technician - fuel transfer Technician - assist in fuel transfer Photographer - Video and photo documentation O
2-25
The total radiation exposure of the personnel during the fuel transfer operation, was less than 0.27 man-rem.
The maximum exposure to an indvidual was 0.08 man-rem.
Photographs of the fuel transfer operation are shown in Photos 5.1.3 thru 5.1.4 The 88 fuel elements at the Facility were shipped to three recipients.
The University of Texas received 59 fuel elements in two shipments of 30 elements and 29 elements.
The University of Kansas received 18 fuel elements, and the University of Illinois received 11 elements.
The data for the transfer of fuel to these recipients is summarized in Tables 5.1.1-6.
5.2 REACTOR HARDWARE REMOVAL AND DISPOSITION The reactor hardware, consisting of radioactive components located in the region of high neutron fluence in the lower shroud structure, and non radioactive components, are shown in Photos 5.2.1.
The various components from the reactor, listed in Table 5.2.1, were given to Universities for use in their training programs.
5.3 POOL WATER PEMOVAL A procedure was written for the removal and disposition of the pool water (Doc 5.3.1).
These procedures were approved by the Decommissioning Organization and the Corporate Radiation Committee.
The measured level of the radioactivity of the pool 2-26
water, after the removal of the reactor fuel and the reactor components, was 3.92 pCi/ liter (see Table 4.3.1), which is below the measured level of natural rainwater, 5.42 pCi/ liter (see Table 4.3.6).
The water was pumped into the normal waste water system.
5.4 EXPOSURE ROOM AND POOL AREA DECONTAMINATION Procedures were written for removal of activated materials in the exposure room, and the decontamination of the room and the pool area (Doc 5.4.1 thru 5.4.10).
These procedures were I
approved by the Decommissioning Organization and Corporate Radiation Committee.
The exposure room and, to a limited extent the beam port area, were the regions most susceptible to neutron activation.
Photos 5.3.1 show the enclosures constructed to control the movement of dust from these areas.
The dust in the air in the enclosure was removed prior to exhausting the air into the facility area by passing it through HEPA filters.
The exposure room window, the wooden liner in the exposure room, and the concrete wall adjacent to the window were first removed, to lower the radiation level in the enclosure, and to allow free movement of air and personnel between the exposure room and the pool area.
These operations are shown in Photo 5.3.2 thru Photo 5.3.5.
The measured radioactivity in the concrete in the exposure room versus depth f rom the surface is summarized in Tables 4.4.8.
The activated concrete in the exposure room, removed to a depth of approximately twenty four inches, weighed 2-27
130 metric tons.
Table 5.4.1 analyzes the effect of geometry of p
the concrete on the measured radiation level.
This analysis shows the measured gamma radiation level in the exposure room cavity, af ter the removal of the contaminated material, was 21.3 uR/hr for a 4 pi geometry exposure.
The radiation level in the cavity for a 2 pi geometry exposure was determined to be 10.6 uR/hr (Table 5.4.1).
The radiation level in the exposure room agrees with the natural background gamma radiation level, 11.6 UR/hr, measured in areas remote from the reactor facility (Table 6.1.1),
and also with the value calculated f rom the measured natural primordial radioactivity in the concrete (Table 4.5.6).
The measured gamma radiation level in the exposure room for 4 pi geometry (see Final Survey Analysis, Table 6.4.1 and Table 6.4.4) slightly exceed the levels specified in NUREG 2082, Monitoring for Compliance with Decommissioning Survey Criteria".
Therefore, in order to demonstrate achievement of the 5 uR/hr criterion without resorting to a geometrical correction extra wall and ceiling surfaces of the exposure room had to be re-moved.
To this end, it was necessary to cut a swath through the concrete structure from the hot cell at the north end through the exposure room and the south wall of the biological shield ( Figure 5.4.1 and Photos 5.4.1 ).
The measured gamma radiation level with the 4 pi geometry effect removed is very close to the background level.
O 2-28
The beam ports and the concrete removed from the west wall O
l in the region of the beam ports weighed approximately 4 metric tons.
The concrete was activated by the diffusion of neutron through the beam port walls during the period the beam ports were unplugged for experiments.
No other sources of contamination were found in these areas.
5.5 COBALT-60 SOURCES HOT CELL A new Northrop Hot Cell Facility will be built for the Cobalt-60 source array.
For this reason the sources were trans-fered to a shipping cask and transported to an approved interim storage facility.
There was neither activation of the concrete, O
nor contamination in the existing Hot Cell.
The nonradioactive O
hot cell hardware are shown in Photos 5.5.1.
5.6 RADIOCHEMISTRY LABORATORY DECONTAMINATION Minor low level contamination of the exhaust hood, bench tops, storage wells, and the drain pipes were found in the Radiochemistry Laboratory.
Figure 5.6.1 shows where the measure-ments were made and Table 5.6.1 lists the measurements.
Photos 5.6.1 show the waste storage wells.
The contaminated materials were either decontaminated or, where more expedient, packaged and shipped to the Hanford burial site.
2-29
5.7 DUCTS AND PLUMBING DECONTAMINATION m
Procedures were written for the decontamination of the ducts and plumbing system (Doc 5.7.1).
The procedures for the radia-tion survey, decontamination, and disposition of the ducts and water pipes were approved by the Decommissioning Organization and
- t. h e Corporate Radiation Committee.
Figure 5.7.1 shows the locations of the radiological measurement on the ventilation ducts and Table 5.7.1 lists the measurements.
There was no significant contamination of the ventilation ducts.
Figure 5.7.2 shows the location of the measurements on the plumbing system and Table 5.7.2 lists the measurements.
The A
location of the drain lines that were removed in the decontam-
, V ination of the plumbing are shown in Photo 5.7.1.
5.8 CONTAMINATED WASTE DISPOSAL Materials showing levels of radioactivity above the accept-able limits given in Table 2.2.2, were decontaminated, or were packaged and shipped to the Hanford site for burial.
The major portion of the shipped material was relatively low activity concrete removed from the exposure room walls and from the beam port region of the reactor pool.
Table 5.8.1 lists the shipments of waste materials to the Hanford site.
The total mass was approximately 138 metric tons.
O 2-30
5.9 UNCONTAMINATED CONCRETE DISPOSAL O
A large f raction of these primordial radioisotopes will survive the disintegration of the concrete on burial.
In contrast, the majority of the radioisotopes created by neutron activation will decay to stable isotopes while still chemically bonded in the concrete, due to their relatively short lifetimes.
An analysis of the radioactivity of the eroded fraction of the radioisotopes versus time after burial is presented in Table 5.9.1 and the data are tabulated in Table 5.9.2 and plotted in Figure 5.9.1.
The plot shows the radioactivity created by neutron capture are always less than one percent of the eroded fraction of the natural background activity, and become com-pletely negligible af ter 100 years.
The eroded concrete rubble buried at the landfill site will be exposed to water infiltrating into the soil f rom natural rainfall.
The radioisotopes from the eroded concrete will be dissolved in the water and transported into the soil and rock structure.
An analysis of the specific activity of the ground-water, based on the report " Geologic and Hydrologic investigation of Puente Hills Landfill Site" (Bib 5.9.1) and the climatological report (Bib 5.9.2), is given in Table 5.9.3 and the dilution of the radioactivity versus time is tabulated in Table 5.9.4.
The specific activity of the radioisotopes versus time is plotted in Figure 5.9.2.
The neutron induced radioactivity introduced into the groundwater at the site is negligible compared to the natural 2-31 1
primordial radioactivity, and, in any event, will be completely gU contained within the site due to : the low permeability of the sandstone formation (75 ft/yr) and the siltstone formation (0.05 ft/yr) ; the construction of leachate barriers in the canyons ;
and the large area of the landfill site (1365 acres) that precludes any of the induced radioactivity leaving the site during the relatively short lifetime of these neutron induced radioisotopes.
6.0 RADIOLOGICAL SURVEY The final radiological survey of the f acility is to deter-mine the suitability of the facility for unrestricted use.
The final release survey plan conformed to the NUREG-2082 " Monitor-ing for Compliance with Decommissioning Survey Criteria" (Bib 6.2.1).
The final release decontamination limits were in conformance with NRC Regulatory Guide 1.86 and State of Califor-nia Release Limits (Bib 6.2.2).
A background level of radiation was measured at sites sufficiently removed from the facility, the various areas with similar categories of potential radiation hazard determined, the areas gridded to determine block loca-tions, and the measurements made and analyzed.
The following sections describe these operations.
6.1 BACKGROUND
RADIOACTIVITY Background data were obtained from locations remote from 2-32
the reactor f acility, on concrete surfaces laid down during the same time period as the reactor facility construction.
The background survey data are summarized in Table 6.1.1 and the location of these areas relative to the reactor building shown in Fig u re 6.1.1.
The average gamma radiation at 1 meter from the surface is 11.5 uR/hr, with a maximum likelihood value of 12.7 uR/hr.
The measured background radiation level is consistent with the measured primordial radioactivity in the concrete as shown in Table 4.5.7 and Table 5.4.1.
6.2 SITE SURVEY CRITERIA 1
Procedures were written for the final survey (Doc 6. 2.1).
l These procedures were approved by the Decommissioning Organiza-tion and the Corporate Radiation Committee.
The Facilty was l
divided into three categories of areas:
o Areas of low contamination potential o
Areas of medium contamination potential o
Areas of high contamination potential Areas where previous surveys indicated contamination was below the release limits and the function of the area was not conducive to creating a contamination problem were designated areas of low contamination potential.
These areas were section-ed into 3 meter by 3 meter blocks.
Thirty percent of the blocks l
were surveyed on a random basis to give a statistically signif-2-33
5 s
i}
icant determination of the radiation level.
I
/
Areas where:l previous survey results indicated contamination near or above release limits and/or the function of that area constituted a potential for contamination were designated areas of medium contamination potential.
Subsequent to decontamina-tion, these areas were sectioned into 2 meter by 2 metet blocks.
Fifty percent of these blocks were surveyed on a random basis to
,give a statistically sign f cant determination of the radiation ii level.
Areas of high contamination potential were sectioned into 1 meter by 1 meter blocks.
Seventy five percent of these blocks were surveyed on a random basis to give a statistically significant determination of the radiation level.
A typical example of radiological survey data used to assign the category of the areas is shown in Table 6.2.1.
The various I
assigned survey areas are listed in Table 6.2.2 and the location of these' areas shown on Figure 6.2.1.
The survey procedure was in conformance with NUREG-2082.
Each designated block was surveyed in the following manner:
p o
A gamma reading (uR/hr) was taken at 1 meter above the centerpoint of the block.
Beta gamma, gamma and alpha readings were taken at five o
2-34
equally spaced points on a one acter square grid within a designated block.
o A
beta-gamma survey scan of the block was conducted and the maximum beta gamma point identified.
o A smear sample and beta-gamma measurements were taken at the maximum beta gamma point, o
The five point survey was recorded and the average of the five points recorded as an unbiased measurement.
o The measurements at the beta gamma maximum point were recorded as a biased measurement.
1 NUREG-2082 suggests 30 uniformly spaced samples for each area aoove 2 meters from the floor level and for overhead horizontal surfaces for a statistically significant measurement meeting release criteria.
Thirty percent of these blocks (greater than 30 uniformly spaced points) were measured on vertical surfaces above 2 meters and overhead horizontal sur-faces.
6.3 ANALYSIS METHODS i
The survey blocks were selected for measurement in a random j-manner.
Block numbers were assigned to all blocks and a random number generator selected the specific blocks to be measured.
A typical example of a survey block layout is shown in Figure 6.3.1 i
2-35
O The method used in processing the very large volume of data ( approximately 20,000 measurements) is outlined in Table 6.3.1.
The data sheet created to record the block measurements is shown in Table 6.3.2, with the symbols that were used in the computer program in order' to process the data inserted in the table.
An example of one of the completed data sheets is shown in Table 6.3.3.
The recorded data for all the blocks in a specific area were transfered to a database file and processed.
An example of the computer printout of the data base file is shown in Table 6.3.4.
An example of the output of the statis-tical analysis of the data from a specific area is shown in Tables 6.3.5, and a histograms of the data distribution for the specific area are given in Table 6.3.6.
6.4
SUMMARY
OF SURVEY DATA The final survey data for the areas of highest potential hazard are summarized in Tables 6.4.1, for medium potential hazard in Table 6.4.2, and for low potential hazard in Table 6.4.3.
An overall summary of the analysis of areas of each category of potential radiation hazard, the net levels of the radiation above background, and the permissible levels are shown in Table 6.4.4.
The marking of the blocks on the walls and the ceiling in the holdup room and the measurement of the gamma radiation level at one meter from the surface with the Ludlum 19 Micro R meter are shown in Photo 6.4.1.
The marking of the 2-36
blocks in the counting room are shown in Photo 6.4.2.
The Tennelec LB 5100 Automatic System for counting surface wipes is shown in the forground of this photograph.
The measurement of overhead surfaces and the ventilation ducts, utilizing a person-nel bucket extension arm is shown in Photo 6.4.3.
7.0 CONCLUSION
These data show that the residual radiation levels in the Facility will meet criterion set down for unrestricted use of the Facility.
Following termination of our operating licence the remaining concrete structure will be demolished and the rubble will be buried at a local landfill site.
O O
2-37
1 J
TABLES TABLE SUBJECT PAGE
2.0 INTRODUCTION
2.2.1 DECOMMISSIONING REGULATIONS...................
3-4 2.2.2 EXCERPT FROM REGULATORY GUIDE 1.86............
3-5 2.2.3 RADIOLOGICAL CONTROL CRITERIA.................
3-6 2.2.4 DECOMMISSIONING EQUIPMENT.....................
3-7 2.2.5 SITE PREPARATION FOR DISMANTLING..............
3-8 i
f 3.0 FACILITY DESCRIPTION 3.1.1 NORTHROP REACTOR CHARACTERISTICS..............
3-9 3.1.2 REACTOR OPERATING HISTORY.....................
3-10 4.0 RADIOACTIVE SOURCES 4.1.1 SEALED NON-FUEL RADIOACTIVE SOUFCES...........
3-11 4.2.1 URANIUM-235 NEUTRON CAPTURE REACTIONS.........
3-12 4.2.2 NORTHROP REACTOR FUEL BURNUP PRODUCTS.........
3-13 4.2.3 PLUTONIUM-239 BUILD-UP........................
3-14 4.2.4 FISSION PRODUCT RADIOACTIVITY ANALYSIS........
3-15 4.3 HISTORICAL RADIOLOGICAL SURVEY DATA 4.3.1 POOL WATER RADIOACTIVITY, NORTHROP TENNELEC DATA...............................
3-16 4.3.2 SUMP WATER DISPOSAL HISTORY...................
3-17 4.3.3 ARGON-41 FMISSION ANALYSIS....................
3-18 4.3.4 ARGON-41 EMISSION HISTORY.....................
3-19 4.3.5 ENVIRONMENTAL SURVEY, LOCATIONS OF i
SAMPLING SITES..............................
3-20 4.3.6 ENVIRONMENTAL SURVEY
SUMMARY
, AVERAGE ANNUAL DATA.................................
3-21 4.3.7 ENVIRONMENTAL SURVEY
SUMMARY
, AVERAGE MONTHLI DATA................................
3-22 4.4 NEUTRON ACTIVATION ANALYSIS OF CONCRETE i
4.4.1 NEUTRON FLUENCE AT SPECIFIC LOCATIONS.........
3-23 4.4.2 RADIOISOTOPE PARAMETERS.......................
3-24 4.4.3 MATERIAL COMPOSITION PARAMETERS...............
3-25 j
4.4.4 NO DILUTION CASE PARAMETERS...................
3-26 4.4.5 SPECIFIC ACTIVITY IN MATERIALS IN CORE........
3-27 4.4.6 SPECIFIC ACTIVITY IN MATERIALS NEAR WINDOW....
3-28 4.4.7 SPECIFIC ACTIVITY IN CONCRETE.................
3-29 4.4.8 SPECIFIC ACTIVITY OF CONCRETE VEFSUS DEPTH....
3-30 4
i I
O 3-1
(
TABLES (Cont)
D)
TABLE SUBJECT PAGE 4.5 PRIMORDIAL RADIOACTIVITY IN CONCRETE 4.5.1 MEASURED RADIOISOTOPE SOURCES IN CONCRETE AND SOIL....................................
3-31 4.5.2 POTASSIUM-40 SOURCE...........................
3-32 4.5.3 URANIUM SERIES SOURCES........................
3-33 4.5.4 THORIUM SERIES SOURCES........................
3-34 4.5.5
SUMMARY
OF BACKGROUND RADIOACTIVITY...........
3-35 4.5.6 EXTERNAL EMISSION OF RADIATION................
3-36 4.5.7 EXTERNAL EMISSION OF RADIATION AT SURFACE.....
3-37 4.5.8 RADON COMPONENT...............................
3-38 5.0 DECONTAMINATION OF FACILITY 5.1 TRANSFER OF NORTFROP FUEL ELEMENTS 5.1.1 FUEL ROD RADIATION LEVELS.....................
3-39 5.1.2 FIRST SHIPMENT TO UNIVERSITY OF TEXAS.........
3-40 5.1.3 SECOND SHIPMENT TO UNIVERSITY OF TEXAS........
3-41 5.1.4 SHIPMENT TO UNIVEPSITY OF KANSAS..............
3-42 5.1.5 SHIPMENT TO UNIVERSITY OF ILLINOIS............
3-43
(
5.1.6
SUMMARY
OF SHIPMENTS..........................
3-44 5.2.1 REACTOP HARDWARE RADIOACTIVITY AND DISPOSITION.................................
3-45 5.4.1 EXPOSURE POOM CAVITY DOSE RATE................
3-47 5.6.1 RADIOCHEMISTRY LABORATORY DECONTAMINATION DATA........................................
3-48 5.7.1 VENTILATION DUCTS DECONTAMINATION DATA (OVERHEAD AND WALL LOCATIONS)...............
3-49 5.7.2 PLUMBING DECONTAMINATION DATA.................
3-51 5.8.1 RADIOACTIVE WASTE SHIPMENTS...................
3-52 5.9 RESIDUAL CONCRETE RADIOACTIVITY 5.9.1 RADIOACTIVITY AT BURIAL TIME..................
3-53 5.9.2 LANDFILL RADIOACTIVITY LIMITS.................
3-54 5.9.3 EROSION RADIOACTIVITY ANALYSIS................
3-55 5.9.4 EROSION RADIOACTIVITY VERSUS TIME BURIED......
3-56 5.9.5 GROUNDWATER ANALYSIS..........................
3-57 5.9.6 GROUNDWATER RADIOACTIVITY VERSUS TIME BURIED......................................
3-58 O
3-2
~ _.. _ _ _.
TABLES (Cont)
TABLE SUBJECT PAGE 6.O RADIOLOGICAL SURVEY 6.
1.1 BACKGROUND
SURVEY DATA........................
3-59 6.2.1 EXAMPLE OF HISTORICAL SURVEY RECORD
( Ju ne 19 8 4 ).................................
3-60 6.2.2
SUMMARY
OF SURVEY AREAS.......................
3-61 6.3 BLOCK SURVEY DATA PROCESSING 6.3.1 DATA PROCESSING OUTLINE.......................
3-64 6.3.2 DATA SHEETS WITH SYMBOLS......................
3-66 6.3.3 WALKWAY DATA SHEET EXAMPLE....................
3-67 6.3.4 WALKWAY COMPUTER PRINT-OUT EXAMPLE............
3-68 6.3.5 WALKWAY STATISTICAL ANALYSIS EXAMPLE.........
3-70 6.3.6 WALKWAY HISTOGRAM PLOT EXAMPLE................
3-71 6.4 FINAL SURVEY OF NOPTHROP REACTOR FACILITY 6.4.1 HIGH CONTAMINATION POTENTIAL AREA DATA........
3-74 6.4.2 MEDIUM CONTAMINATION POTENTIAL AREA DATA......
3-76 6.4.3 LOW CONTAMINATION POTENTIAL AREA DATA.........
3-78 6.4.4 ALL CATEGORIES, BACKGPOUND AND PEPMISSIBLE LEVELS......................................
3-80 l
I I
l i
l
\\
3-3
/~
TABLE 2.2.1 b]
DECOMMISSIONING REGULATIONS SUBJECT AGENCY DOCUMENT COMMENT OPERATIONS FEDERAL NRC/R90 (D-850-187)
NORTHROP LICENSE AND CALIP 0006-59 NORTHROP LICENSE SECURITY FEDERAL R-90 APPENDIX (A)
NORTHROP LICENSE FEDERAL 10 CFR 50 LICENSE, USE FEDERAL NUREG 1756 DISMANT./DECOMM.
FEDERAL ANSI /ANS-15.10 DECOM./ SECURITY RADIATION FEDERAL NRC/10 CPR 20 PROTECTION / STANDARDS PROTECTION FEDERAL REG GUIDE 1.86 DECOM. LIMITS AND CALIF.
DHS TITLE 17 CAC PROTECTION /TERMIN.
SURVEYS CALIF.
DHS DECON-1 DECOM. LIMITS FEDERAL NUREG 2082 TERMINATION SURVEY FEDERAL ANSI /ANS-N-13.12 DECOM. LIMITS FEDERAL NUREG GUIDE 8.29 RA RISKS NONRADIATION FEDERAL OSHA 29 CFR WORK HEALTH / SAFETY
(/}
PROTECTION CALIF.
OSHA TITLE 8-CAC WORK HEALTH / SAFETY s_
ENVIRONMENTAL FEDERAL EPA NEPA PREPARATION EIS IMPACT CALIF.
OPR CEQA PREPARATION EIS FEDERAL NUREG-0586 PREPARATION EIS SHIPPING FEDERAL NRC/10 CFR 71 SHIPPING RA MATER.
OF FEDERAL NRC/10 CFR 73 SHIPPING FUEL RADIOACTIVE FEDERAL DOT /49 CFR SHIPPING RA MATER.
i MATERIAL CALIF.
VEH. CODE-33002 TRANSPORTATION QUALITY FEDERAL NRC/10 CFR 71.12 QA PROGRAM ASSURANCE FEDERAL NUREG GUIDE 7.10 PACK /TRANSP/RA MAT.
O v
3-4
[-
[,
[,)\\
\\
Table I-1.
Acceptable surface contamination levels
- [
Removable
'#'d Nuclides#
Average
- Maximum
- U-nat, U-235, 0 238, and 5,000 dpm a/100 cm 15,000 dpm a/IDO cm' I,000 dpm o/100 co' 3
associated decay products Transuranics, Ra-226 Ra-228, 100 dpm/100 ca 300 dpe/100 ca' 20 dpm/100 cm' r
g Th-230, Th-228, pa-231, gis g
Ac-227, 1-125, I-129 Th-nat, Th-232, Sr-90 1,000 dpm/100 ca 3,000 dpm/100 cm' 200 dpm/100 cm'
>B r
Ra-223, Ra-224 U-232, 1-126, q
I-131, 1-133 g
Beta-gamma emitters (nuclides 5,000 dpm fh/100 cm' 15,000 dpm h/100 cm' I,000 dpm h /IDO cm' 3
4 with decay modes other than alpha emission or spontaneous l
fission) except Sr-90 and O
t*
ta other noted above.
C I
I tn
,ha Y
M ahhere surface contamination by both alpha-and beta. gamma-emitting nuclides exists, the limits established for alpha-O e
and beta-gamma-emitting nuclides should apply independent ly.
M bAs used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive m1terial as O
determined by correcting the counts per minute obserted by an appropriate detector for background, efficiency, and geometric c
factors associated with the instrumentation, g
Sleasurements of average contaminant should not be averaged over more than I square meter. for objects of less surface 84 area, the average should be derived for each such object.
pa be maximum contamination level applies to an area of not more than 100 cm' h
3
- The amount of removable radioactive material per 100 cm of ss.. ace area should he determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.
IThe average and maximum radiation levels associated with surface contamination resulting from beta-gamma emitters should not exceed 0.2 mrad /hr at I cm and 1.0 mrad /hr at I cm, respectively, measured through not more than 7 milligrams per squa e centimeter of total absorber.
TABLE 2.2.3 O
RADIOLOGICAL CONTROL CRITERIA UNRESTRICTED USE TYPE ACTIVITY TOTAL REMOVABLE Alphas 100 dpm/100 cm2 20 dpm/100 cm2 Beta-Gammas 0.1 mrad /hr at 100 dpm/100 cm2 0.1 cm thru 7 mg/cm2 absorber LIMITED TO POSTED CONTAMINATED AREAS Alphas 2500 dpm/100 cm2 500 dpm/100 cm2 Beta-Gammas 1 mrad /hr at 2500 dpm/100cm2 1 cm thru 7 mg/cm2 absorber Equipment and material with activity between these limits shall be evaluated and reviewed in each specific case O
3-6
TABLE 2.2.4
\\
DECOMMISSIONING EOUIPMENT SPECIALIZED EOUIPMENT FOR DISMANTLING AND DECONTAMINATION Long-handled Fuel Handling Tool Transfer Cask Shipping Cask Radiological Vacuum Cleaner Diaphragm Pump H.E.P.A.
Filter Exhaust System Self Contained Respirators Alph, Beta, Gamma Measuring Instruments Automatic Counter System for Wipe Counting Continuous and Periodic Air Samplers HEAVY EQUIPMENT Bridge Crane Fork lift Scaffolding Aerial Manlift Hydraulic Hammer and Scoop Arc Welder MOBILE EOUIPMENT 18 Wheel Truck and Flatbed Trailer Personnel Van Storage Trailer Vacuum Pump Truck HAND TOOLS ETC Hydraulic Cutter Shears Portable Hand-held Band Saw Circular Saw Reciprocating Saw Tubing Cutter Heavy Duty Bolt Cutter Oxyacetylene Cutting Tool Electric Jack Hammer 50 Ten Jacks Dynamometer Safety Belts and Harnesses O
3-7
TABLE 2.2.5 SITE PREPARATION FOR DISMANTLING TASE PREPARED BY Remove Flash Xray from Facility NORTHROP Make a Storage Area Outside the Reactor NORTHROP Dispose of Nonessential Equipment NORTHROP Install an Overhead Crane Extension NORTHROP Arrange for a Fuel Shipping Cask CHEM-NUCLEAR Arrange for Fuel Transfer to Recipients NORTHROP Arrange for Fuel Transport NORTHROP Obtain the Fuel Handling Equipment NORTHROP/C-N Make Office Space for Subcontractor NORTHROP Make Preliminary Radiological Survey NORTHROP/C-N Layout Blocks for Final Radiological Survey NORTHROP/C-N O
J O
3-8
TABLE 3.1.1 NORTHROP REACTOR CHARACTERISTICS FUEL ELEMENTS Number 88 (86 in core + 2 spares)
Fuel-Moderator Uranium-Zirconium Hydride Uranium Fraction 8 percent by weight Hydrogen / Zirconium Ratio 1.7 atom ratio Uranium Enrichment 20 percent U-235 Overall Dimensions 1.47 in. diam x 28.3 in. long Cladding 0.20 in. Stainless Steel Dimensions Core 17 in. diam. x 15 in, high Poisons Burnable Poisons REFLECTOR Material:
Radial Water Axial Water and Graphite Thickness:
Radial Variable (2 in, at window)
Axial Water + 3.5 in. Graphite NUCLEAR CHARACTERISTICS 3396graggU-23g/sec.
()
Fuel Inventory 2.0 x 10 n/cm
(_,/
Thermal Flux at Core Excess Reactivity-initial 3.0 percent ($4.30)
Transient Max. Reactivity 2.1 percent dk/k ($3.00)
Control System Reactivity 6.3 percent dk/k O
4 Temp, Coef. at 300 C
-(1.3 + 0.1) x 10 5 dk/k/0C Energy Shutdown Coef.
+(3.3 + 0.2) x 10-1/ joule Prompt Neutron Lifetime 39 microseconds Effective Delay Neut. Fract.
0.0070 THERMAL CHARACTERISTICS Steady-state Pulsed Power 1 megawatts 1800 megawatts 10 milliseconds Pulse Time 18 megajoules Pulsed Energy O
Temperature
< 600 C
< 600 C
Cooling Natural Convection Natural Convection CONTROL 1 Pulsed B/C Rod 2.1 percent dk/k net 3 Standard B/C Control Rods 1.4 percent dk/k each Drives:
Pulsed Rod Air Piston + Mechanical Adjust Standard Control Rods Rack and Pinion (max. rate for Stand.)
(0.04 percent dk/k per sec.)
O 3-9
4 f
TABLE 3.1.2 REACTOR OPERATING HISTORY l
l-TIME PERIOD-l l-------
ANNUAL --------l CALENDAR DELAY ENERGY FRACTION ACTIVITY (YR)
(DT)
(DE)
(PC)
(DK)
(YEAR)
(YEARS)
(GJ) (PERCENT) (PERCENT) 1963 22.00 32 0.7 0.02 1964 21.00 30 0.7 0.02 l
1965 20.00 111 2.5 0.08 1966 19.00 56 1.3 0.04 1967 18.00 219 5.0 0.18 1968 17.00 160 3.6 0.14 1969 16.00 218 4.9 0.20 1970 15.00 1135 25.7 1.14 1971 14.00 884 20.0 0.97 i
1972 13.00 241 5.5 0.29 1973 12.00 270 6.1 0.35 1974 11.00 194 4.4 0.28 1975 10.00 164 3.7 0.27 1976 9.00 95 2.2 0.18 1977 8.00 89 2.0 0.19 1978 7.00 80 1.8 0.20 1979 6.00 85 1.9 0.25 1980 5.00 66 1.5 0.25 1981 4.00 69 1.6 0.35 1982 3.00 85 1.9 0.58 l
1983 2.00 84 1.9 0.95 L
1 1984 1.00 47 1.1 1.26 SUM 4414 100.0 8.19 j
DT = (1985-YR)
DE = Energy produced during calendar year.
l' DK = 0.0281 PC (TO/DT) pngrgy produced during calendar year.
PC = Praction of total
, Fraction of fission product activity surviving at dismantling time (see Table 4.2.4).
TO = 22 yrs, Operating life of reactor.
i O
1 3-10
~.
TABLE 4.1.1 O
SEAT.PD NON-FUEL RADIOACTIVE SOURCES SDRRCE SOURCE SOURCE SOURCE TX2E LSE DISPOSITION STRENGTH (CURIES) 12 CO-60 (SEALED)
NEW HOT CELL NORTHROP 1492 12 CO-60 (SEALED)
NEW HOT CELL NORTHROP 335 1 CS-137(SEALED)
GAMMA CALIPRATION NORTHROP 122 1 AM-BE (SEALED)
NEUTRON CALIBRATION NORTHROP 4.54 1 AM-BE (SEALED)
NEUTRON CALIBRATION PENN STATE 3
O O
3-11
i TABLE 4.2.1 O
URANTUM-235 NEUTRON CAPTURE REACTIONS REACTANTS U-235 NUCLEUS + NEUTRON PRODUCTS FISSION PRODUCTS (874)
U-236 (13 % )
COMPONENTS NEUTRONS FISSION FRAGMENTS U-236 PROMPT (MeV) 6(K.E.)
162(K.E.)
6(GAM) 8(GAM)
DECAY (MeV) 12(GAM) 5 (B) 5(G) 11(n'O)
DEP.EN.(MeV) 16 15 1
EARLY ACT.
VARIED VARIED U-236 LATE ACT.
Fe Co Ni Sr Cs U
(EXAMPLES) 55 60 63 90 137 236 L' TIME (YRS) 3.9 7.6 132 40.5 43.3 40 MY BETAS (MeV)
--.32
.07 0.55 1.2,0.5 4.5 O-(ALPHAS)
GAMMAS (MeV)
.23 2.5 ---
0 0.66 HQIE: Fission product fraction is increased by the build-up of plutonium-239 through neutron capture by uranium-238.
O 3-12
TABLE 4.2.2 v
NORTHROP REACTOR FUEL BURNUP PRODUCTS BURNUP OF U-235 DURING OPERATION l
El
= 173 MeV
= 2.7 x 10-11 J per U-235 burnup Ebu = 1226 MW-hrs
= 4400 GJ during operation
= 1.6 x g 23 Nbu = Ebu/E1 U-235 reactions No
= 6 x 10 atoms per mole mbu = Nbu/N
= 0.26 moles U-235 reactions o
Mbu = mbu x 235
= 62 grams U-235 burnup M
= 3396 grams U-235 at Facility r
f
- M /Mr = 0.018 = 1.8 percent fuel burn-up r
f FISSIONS DURING OPERATION fg
= 0.87 fissions per burn-up Mg =ffxMbu
= 54 grams fission products NEUTRON PRODUCTION DURING OPERATION n
= 2.5 neutrons per fission m
=nxff x mbu
= 0.57 moles neutrons n
PRODUCTS FROM NEUTRON REACTIONS (ESTIMATES) mg
= (1/n) xm
= 0.23 moles fissions
= ((n-1)/n) nx m = 0.34 moles non-fission captures ma n
m26 = (mbu -mg)
= 0.03 moles U-236 produced m39
= 0.01 moles Pu-239 produced (see Table 4.2.3) 2xmg
= 0.46 moles fission products mfp== m137 =.06 x mf= 0.014 moles m90 Sr-90 and Cs-137 a
=.03 x mg
= 0.007 moles burned Boron m
m
=.005 x m
= 0.002 moles non-fission products x
a O
3-13
3 TABLE 4.2.3 O
PLUTONIUM-239 BUILD-UP 49/dt
= (s28 N28 F) - (s49 N49 F)
REACTION:
dN 49/dt
=0 STEADY STATE:
N 49 (s28 / s49) N28 N
=
N28
= (80% / 20%) N25 (4) (s28 /E49) N25 N
=
49 BUILD-UP STATE:
N49 (t)
=N49 (1-EXP(-t / T)
T
= 1 / (s49 F)
ADDED DURNUP ENERGY:E49
= ((N49 s49)/(N25 s25)) E25 (N49/N25) M25 STEADY STATE PU-239:M49
=
ps REACTOR PARAMETERS Q
s28 = 10612.74 barns (fission + capture) s49 =
barns (fission + capture) s49f=
746 barns (fission) s25f=
582 barns (fission)
Feb =
1722 x 10 geV (U23g nucleus burnup)
Eg 1
neut/cm /sec,1 MW power
=
N25 = 3319 grams of U-235 in core N49 =
37 grams Pu-239, steady state 3.1 MeV (U235 nucleus burnup)
E 49 = 18600 hours (at 1 MW power level)
T
=
PU-239 PRODUCTION (Reactor was in build-up state) t
= 1226 hours0.0142 days <br />0.341 hours <br />0.00203 weeks <br />4.66493e-4 months <br /> at 1 MW power level t/T
= 0.066 M49
= 2.4 grams Pu-239 at shutdown 49
= 0.20 MeV at shutdown E
O 3-14
TABLE 4 2md iV FISSION PRODUCT RADIOACTIVITY ANALYSIS GAMMA ACTIVITY
- Average gamma energy is Gamma Rate = (1.9 x 10-6)
- x t-[.2gammas per sec per fission for t in days BETA ACTIVITY
- Average beta energy is 0.4 MeV Beta Rate = 2 x Gamma Rate INTEGRATED ACTIVITY Specific Activity of gammas in U235 is given by:
AAG= K AU / 100 K
= SUM ( DK ) = 8.19 pegcent - see Table 3.1.2 DK = 0.028 PC (TO/DT)
AU = 1.28 Ci/gm of U235 - see Table 4.4.4 g
AAG= 0.11 Ci/gm of U235 V)
Total activity of gammas in 88 fuel rods is:
AG = AAG
- MU MU = 3396 gms of U235 - see Table 4.2.2 ACTIVITY AT DISMANTLING TIME AG = 360 Ci of gammas at dismantling time AB = 720 Ci of betas at dismantling time ACTIVITY PER ELEMENT NR = 88 rods AR = AG/NR = 4.1 Ci/ rod Equations from The Elements of Nuclear Reactor Theory, Glasston. and Edlund (Bib 4.2.1).
O 3-15
i TABLE 4.3.1 O
POOL WATER RADIOACTIVITY NORTHROP TENNELEC DATA 5
ZZEE SPECIFIC ACTIVITY (pCi/ liter) i Al ha 0.14 P
Beta 3.78 Gamma O
Total 3.92 HQTE The activity of the pool water is lower than the rainwater activity, measured off site, of 5.42 pCi/ liter (Table 4.3.6).
O 4
l j
I l
!O l
i 3-16 i
TABLE 4.3.2 O
Yj SUMP WATER DISPOSAL HISTORY XEAB ANNUAL SPECIPIC SPECIFIC SPECIPIC ANNUAL ANNUAL ANNUAL WATER BETA-GAM ALPHAS TRITIUM BET-GAM ALPE TalT VOLUME
&CTIVITY ACTIVITY ACTIVITY ACTIV. ACTIV.
ACTIV.
(1000 1)
(pCi/1)
(pCi/1) (nCi/1)
(uCi)
(uCi)
(mci)
(Note 1)
(Note 2) 1966 28.1 459 12.0 0
12.91 0.35 0
1967 17.5 1700 16.0 0
29.93 1.39 0
1968 49.0 20 1.0 0
1.05 0.02 0
1969 50.0 15 0.0 0
0.76 0.00 0
1970 79.3 42 0.0 7364 3.30 0.00 584 1971 62.0 50 0.0 4322 3.09 0.00 268 1972 26.0 54 0.0 1192 1.41 0.00 31 1973 39.0 29 0.0 0
1.13 0.00 0
1974 20.0 24 0.0 0
0.47 0.00 0
1975 23.0 14 1.0 0
0.32 0.02 0
1976 10.3 16 0.0 0
0.17 0.00 0
1977 9.5 24 0.0 0
0.23 0.00 0
1978 0
0 0.0 0
0.00 0.00 0
1979 15.4 10 0.0 0
0.15 0.00 0
1980 27.0 1
2.0 0
0.22 0.05 0
f 'g 1981 0
0 0.0 0
0.00 0.00 0
e
(_y 1982 0
0 0.0 0
0.00 0.00 0
1983 15.1 1
0.4 0
0.13 0.01 0
1984 17.0 26 1.8 0
0.44 0.03 0
TOT 488.2 55.71 1.85 883 AVER. 25.7 2.90 0.10 47 SPEC.AV. -
130.8 1.8 678 HQ.TE 1)
Permissible limits depend on isotopic constituents.
Levels were always below permissible limits.
2)
Permissible limit is 3000 nci/1 for disposal in waste water (See 10 CFR 20, Appendix B, Table II).
i O
3-17
TABLE 4.3.3 ARGON-41 EMISSION ANALYSIS ANNUAL STACK VOLUME FLOW RATE 3
V3=vxHxN (m /yr)
- Annual stack flow rate 3
3 v
= 20000 ft /m = 34000 m /hr: Operating stack flow rate H
= 9 hrs / day
- Daily operating hours N = 4 days /wk
= 209 days /yr Annual operating days 3
V3 = 0.064 Gm /yr ANNUAL STACK ARGON-14 RATE A
(Ci/yr)
- Data in Reactor log book (see Table 4.3.4)
ANNUAL DILUTION VOLUME 3
XY (Gm /yr)
- Annual dilution volume VD=FTxFD S
(Gas Eff. Log Book) i PT= (24/9)x(365/209) = 4.7
- Operating time factor
(
FD = 165
- Diffusion area factor 3
VD = 50 Gm /yr AVERAGE CONCENTRATION OVER YEAR 3
(A / V ) x 1000 (pC1/m ) : Average concentration C
=
D (see Table 4.3.4)
O 3-18
,g TABLE 4.3.4 ARCON-41 EMISSION HISTORY CALENDAR ANNUAL ANNUAL ANNUAL 1EAB DILUTION ARGON-41 ARCON-41 VOLT}ME EMISSION CONCENTRATION (Gm )
(Ci)
(pCi/m )
3 (Note 1)
(Note 2)
(Note 3 and 4) 1963 50 5.13 100 1964 50 18.05 370 1965 50 101.21 2020 1966 50 41.04 840 1967 50 21.70 440 l
1968 50 16.21 330 1969 50 21.00 430 1970 50 12.80 260 1971 50 8.07 160 1972 50 8.86 180 1973 50 8.07 160 1974 50 7.34 150 1975 50 12.30 250 1976 50 9.15 190 1977 50 7.48 150
(~Ns) 1979 50 3.40 70 1978 50 6.56 130 1980 50 2.50 50 1981 50 3.68 80 1982 50 3.81 80 1983 50 7.05 140 1984 50 3.87 80 AVERAGE 50 14.97 390 TOTAL (22) 1078 329.28 HQIE
- 1) DILUTION VOLUME
- V, See Table 4.3.3 for calculation D
- 2) ARGON-41 ACTIVITY:
A, Recorded data in Gas Eff. Log Bk.
C = ( A/V ), See Table 4.3.3
- 3) CONCENTRATION D
3
- 4) PERPASSIBLE
- C < 40000 pCi/m averaged over one year (10 CFR,20, Append.B).
O 3-19
l l
TABLE 4.3.5 O
(m) j ENVIRONMENTAL SURVEY LOCATIONS OF SAMPLING SITES SIZE LOCATION SAMPLE TAKEN S-1 Reactor Soil, vegetation, drinking water S-2 Imperial Highway and soil, vegetation, drinking water Inglewood Blvd.
1 S-3 Imperial Highway and Soil, vegetation, drinking water Sepulveda Blvd.
S-4 Prairie Avenue and Soil, vegetation, and pond water Redondo Beach Blvd.
(Alondra Park)
S-5 Hawthorne Blvd. and Soil, vegetation, drinking water Redondo Beach Blvd.
S-6 Hawthorne Blvd. and Soil, vegetation, drinking water 190th St.
s S-7 Normandie and Soil, vegetation, drinking water y,)
El Segundo Blvd.
S-8 Rosecrans and Soil, vegetation, drinking water Central Avenue S-9 Hawthorne Blvd. and Drinking water Century Avenue S-10 La Brea Avenue and Soil, vegetation, drinking water Slauson Avenue S-ll Roof Engineering Center Air (particulate), and rainwater (900 yards west of reactor)
S-12 Roof Plant III Air (particulate), and rainwater (200 yarda east of reactor)
HQIE Summary of data - Table 4.3.6 and 4.3.7 Map showing locations - Figure 4.3.1 i
{
O 3-20
g-~
TAnLE 4.3.6 ENVIRONMENTAL SURVEY
SUMMARY
AVERAGE ANNUAL DATA (Report year, Sept thru Aug)
REPORT VEGE-SOIL WATER AIR RAIN YEAR TATION WATER (PCi/g) (pCi/g) (pCi/1) (pCi/km3) (pci/1) 75 94.10 12.80 2.06 55.90 5.41 76 67.60 12.30 1.97 20.50 3.64 77 96.70 13.60 3.06 84.40 8.64 78 99.10 14.10 3.65 129.70 14.70 79 71.00 14.60 2.85 31.40 6.30 80 63.50 13.10 2.68 65.50 2.52 81 62.20 11.10 2.06 106.00 8.69 82 55.80 18.50 1.86 15.70 1.26 83 60.20 11.30 2.44 11.80 4.15 84 62.40 9.50 2.25 13.00 2.25 85 78.70 14.50 3.09 17.20 2.06
()
AVER 73.80 13.20 2.54 50.10 5.42 l
l l
l O
l I
3-21 i
I l
TABLE 4.3.7 ENVIRONMENTAL SURVEY
SUMMARY
AVERAGE MONTHLY DATA l
REPORT VEGE-SOIL WATER AIR RAIN MONTH TATION WATER l
(pCi/g) (pCi/g) (pC1/1) (pC1/km3) (pCi/1)
SEPT 72.00 11.60 2.58 46.70 1.68 OCT 74.60 12.70 2.65 50.50 7.22 NOV 71.10 13.20 2.63 51.00 7.21 DEC 77.70 12.90 2.43 50.30 4.40 JAN 82.00 13.40 2.70 61.10 7.10 FEB 84.00 12.10 2.66 69.10 3.83 MAR 75.80 12.40 2.35 68.70 10.82 APR 63.20 13.20 2.39 60.70 9.01 MAY 72.70 12.70 2.47 48.20 3.38 JUNE 67.30 12.50 2.54 36.20 0.56 JULY 74.70 12.20 2.46 29.30 0.00 AUG 70.00 19.90 2.66 29.20 9.82
(
AVER 73.80 13.20 2.54 50.10 5.42 O
3-22
TABLE 4.4.1 NEUTRON PLUENCE AT SPECIPIC LOCATIONS LOCATION THERMAL OPERATING OPERATING QEERATING NEUTRON FRACTION FLUENCE HET FLUEN{E FRACTION FRACTION (n/cm )
(percent)
(percent) (percent)
==================================================
REACTOR CORE 2.20 x 1019 100 100 100 EXPOSURE ROOM WINDOW 1.5 x 1017 45 0.67 0.3 15 WINDOW WALL 4.9 x 10 45 0.022 0.01 15 45 0.022 0.01 CONCRETE SURFACE
- 4. 9 x 10 12 24 INCH DEPTH 9.8 x 10 45 0.000044 0.00002 POOL BEAM PORTS 13 5
0.0004 0.00002
/~
WEST WALL 9.8 x 10 b]
POOL CENTRAL POSITION 11 WALLS AND FLOOR 9.8 x 10 50 0.000004 0.000002 0
3-23
Ob TABLE 4.4.2 NEUTRON ACTIVATION ANALYSIS RADIOISOTOPE PARAMETERS l ACTIVE ISOTOPE PARAMETERS ll----BASE ISOTOPE PARAMETERS-------l SYMBOL LIFETIME OP. FACTOR SYMBOL ABUNDANCE CROSSSECT. RATIO (XA)
(TA)
(HA)
(YB)
(GB)
(CB)
(PB)
(YEARS)
(PERCENT)
(BARNS) (PERCENT)
NEUTRON ACTIVATION PARAMETERS IN STRUCTURAL MATERIALS FESS 3.90 0.50 FE54 5.82 2.50 0.43 CO60 7.60 0.67 C059 100.00 20.00 3.45 NI63 144.44 0.15 NI62 3.66 15.00 2.59
()
ZN65 0.97 0.28 ZN6 48.89 0.44 0.08 (m /
EU15 19.60 0.60 EU151 47.82 5900.00 1020.00 EU15 12.70 0.68 EU153 52.18 400.00 69.00 1
XA Radioactive isotope created by thermal neutron capture by base isotope,YB.
TA Mean lifetime of XA (years).
(TO/TA) SUM (PC EXP(-DT/TA))
Percentage of isotope, HA
=
XA, surviving at shutdown time - see Figure 4.4.3.
TO 22 years Operating life of the reactor.
=
PC Percentage of total energy produced during the l
year,YR.
1985-(YR+.5)
DT
=
YB Parent of radioisotope, XA.
GB Natural abundance of base isotope (percent).
CB Thermal neutron cross section of base isotope (barns).
PB Ratio of CB to U235 fission cross section (percent).
3-24
TABLE 4.4.3 NEUTRON ACTIVATION ANALYSIS MATERIAL COMPOSITION PARAMETERS l--BASE MATERIAL -----l l-------ELEMENTAL FRACTION ------------l ISOTOPE AT. MASS RATIO FIS. PROD ALUMINUM REBAR CONCRETE (XA)
(ME)
(QE)
(JEFP)
(JEAL)
(JEREB)
(JECON)
(GRAMS)
(--------- WEIGHT PERCENT ------------)
NEUTRON ACTIVATION PARAMETERS IN STRUCTURAL MATERIALS FE55 54 23.0 0.0 0.25000 99.300 10.2000 CO60 59 25.1 0.0 0.00010 0.024 0.0070 NI63 62 26.4 0.0 0.00100 0.010 0.0015 ZN65 64 27.2 0.0 0.07000 0.025 0.0020 EU152 151 64.3 0.0 0.0005 EU154 153 65.0 0.0 0.0005 ME Base element atomic mass (grams).
()
QE Ratio ME to MU235 (percent).
JEXXX:
Ratio base element weight to the material mass (percent).
O 3-25
TABLE 4.4.4 NEUTRON ACTIVATION ANALYSIS NO DILUTION CASE PARAMETERi ACTIVE ISOTOPE OPER. CR' SECT. AT. MASS ELEMdNT SPECIFIC ISOTOPE ABUND. FACTOR RATIO RATIO FACTOR ACTIVATION (XA)
(GB)
(BA)
(PB)
(QE)
(KA)
(AAA)
(PERC.)
(PERC.) (PERC.) (PERC.)
(uCi/gm)
NEUTRON ACTIVATION PARAMETERS IN STRUCTURAL MATERIALS FESS 5.80 0.50 0.43 23.0 0.054 691 CO60 100.00 0.67 3.45 25.1 9.200 118000 NI63 3.66 0.15 2.59 26.4 0.054 680 ZN65 48.90 0.28 0.08 27.2 0.040 510 EU152 47.80 0.60 1020.00 64.3 454.000 5810000 EU154 52.20 0.68 69.00 65.0 37.600 482000 JEXXX Dilution of Base element material (percent) in structural material (Table 4.4.3).
(GB HA PB) / QE (percent)2 KA
=
7.'
(s)
=
(NO CU FU)/(MU KC TO)/ = 1.28 LCi/gm)
=
NO
.6023 Avogadros number (1/ barns)
=
CU 580 (barns)
=
235 (grms)
MU
=
10 3.7 x 10 (d/s/Ci)
KC
=
8 22 yrs = pS 4 x 10 (sec)
TO 9
=
FU 2.2 x 10 Total neutron fluence (n/cm2)
=
O 3-26
TABLE 4.4.5 O
NEUTRON ACTIVATION ANALYSIS SPECIPIC ACTIVITY IN MATERIALS IN CORE ACTIVE BASE l-SPECIFIC ACTIVITY MATERIAL IN CORE--l ISOTOPE ISOTOPE FIS. PROD.
ALUMINUM REBAR CONCRETE (XA)
(YB)
(AAFP)
(AAAL)
(AAREB)
(AACON) l ----- ------- -----pC i/g m------ -- -- - --- - ----- - l NEUTRON ACTIVATION PARAMETERS IN STRUCTURAL MATERIALS FESS FE54 0
1750000 690000000 70000000 CO60 C059 0
120000 28000000 8200000 NI63 NI62 0
6900 69000 9200 ZN65 ZN64 0
360000 128000 10200 29000000 EU152 EU151 2400000 EU154 EU153
/~'
AAXXX AAA JEXXX 10000 (pCi/gm)
=
JEXXX See Table 4.4.3 (percent)
O 3-27
TABLE 4.4.6 NEUTRON ACTIVATION ANALYSIS SPECIPIC ACTIVITY IN MATERIALS NEAR WINDOW ACTIVE BASE l-SPECIFIC ACTIVITY-l ISOTOPE ISOTOPE ALUMINUM REBAR (A)
(B)
(AAWINAL)
(AAERREB) l------- pCi/gm-------I NEUTRON ACTIVATION PARAMETERS IN MATERIALS FESS FE54 5300 69000 CO60 C059 350 2800 NI63 NI62 20 7
ZN65 ZN64 1100 13 EU152 EU151 EU154 EU153 AAYYYXXX AAXXX LYYY / 100, Specific (pCi/gm)
=
activity of material, XXX, located at position YYY.
AAXXX See Table 4.4.5.
(pCi/gm)
LYYY (FL/FU) (OL/OU)
Ratio of (perc.)
neutron fluence.
FL/FU Neutron flux at YYY relative (perc.)
to Core.
OL/OU Fraction of operation at location.
0.3 percent LLWIN
=
LLER(window)=
0.01 percent OV 3-28
~
TABLE 4.4.7 NEUTRON ACTIVATION ANALYSIS SPECIFIC ACTIVITY IN CONCRETE ACTIVE l-SPECIFIC ACTIVITY-CONCRETE----l ISOTOPE EXP.RM.
POOL WEST POOL OTHER (XA)
(AAERCON)
(AAPLWCON) (APLOCON)
(------------
pCi/gm------------)
NEUTRON ACTIVATION PARAMETERS IN CONCRETE FE55 7000 18 1
CO60 800 1
0 NI63 1
0 0
ZN65 1
0 EU152 2900 4
1 EU154 240 0
0 NATURAL BACKGROUND ACTIVITY IN CONCRETE (Note 1)
K40 17 17 17 hs)
U238-P206 8.2 8.2 8.2 Th232-Pb208 11.2 11.2 11.2 Note 1)
See Tables 4.5.1 thru 4.5.5 for calculations.
2)
See Table 4.4.6 for parameter definition.
LLER
= 0.01 percent LLWPL
= 0.00002 percent LLOPL
= 0.000002 percent O
3-29
TABLE 4.4.8 NEUTRON ACTIVATION ANALYSIS SPECIFIC ACTIVITY OF CONCRETE VERSUS DEPTH NEUTRON INDUCED ACTIVATION IN CONCRETE OF EXPOSURE ROOM DEPTH EE-15 CD-60 EU-152 EU-154 TOTAL (INCHES) l----------------pCi/gm-----------------------l 0-3 7060 796 2870 234 10960 17 1300 2.92 5.0 0.6 1307 20 73.5 2.74 8.2 0
84.4 24 (see Note)
(13)
(2.5)
(8.8)
(0.7)
(25.0) 29 1.86 0.36 1.25 0
3.5 Note: Data for 24 inch depth is interpolation between 20 inch and 29 inch measurements.
O l
O 3-30 l
TABLE 4.5.1 PRIMORDIAL RADIOACTIVITY IN CONCRETE MEASURED RADIOISOTOPE SOURCES IN CONCRETE AND SOIL NATURAL BACKGROUND ACTIVITY IN NORMAL CONCRETE LOCATION K-lp RA-226 TH228 TH232 0235 l-----------------pci/gm---------------------I EXP.RM.
14.4 EXP.RM.
14.0 EXP.RM.
18.6 0.38 0.041 POOL 18.6 0.31 POOL 17.9 0.46 POOL 14.6 OUT. WALL 20.0 0.32 0.73 0.64 AVERAGE 16.9 0.37 0.73 0.64 0.041
=============================================
a V
NATURAL BACKGROUND ACTIVITY IN SOIL BELOW EXP.RM 16.1 0.42 0.038 BELOW EXP.RM 18.3 0.35 0.047 AVERAGE 17.2 0.38 0.042 Note Blanks in table indicate that particular isotope was not analyzed, and does not necessarily mean that the isotope was not present in the material.
O 3-31 i
\\
TABLE 4.5.2 O
PRIMORDIAL RADIOACTIVITY IN CONCRETE POTASSIUM-40 SOURCE Nuclear Decay Reactions K-40 = Ca-40 + beta + 1.3 MeV (89 percent)
K-40 = A-40 + El.C.+ 1.5 MeV (11 percent) 9 T1/2 = 1.26 x 10 years Soecific Activity A
" 17 PCi/gm - Total measured activity of K-40 t
Ab
= 15 pCi/gm - Beta activity, 89 percent of A
- t 2 pCi/gm - Gamma activity, 11 percent of A A
=
t*
g O
O 3-32
TABLE 4.5.3 O
PRIMORDIAL RADIOACTIVITY IN CONCRETE URANIUM SERIES SOURCES Nuclear Decav Reactions U-238 ----> Pb-206 + 8 alphas + 6 betas + ~8 gammas /xrays l/2
= 4.5 x 109 T
years D-Mass
= 51.7 MeV Rest mass difference, see Bib 4.5.1.
D-alpha
= 42.9 MeV Kinetic energy, see Bib 4.5.2.
D-beta
= 5.7 MeV See Bib 4.5.2.
D-recoil
= 0.8 MeV Momentum conservation nuclear recoils D-part.
= 49.4 MeV Sum of particle kinetic energies D-gamma
= 51.7 - (f.4 = 2.3 MeV Difference of above E
= 42.9/8
= 5.4 MeV per alpha a
Eb
= 5.7/6
= 0.8 MeV per beta E
= 2.3/8
= 0.3 MeV per gamma
()
Specific Activity A1
= 0.37 pCi/gm Measured Ra-226 activity for one alpha decay.
1 8x0.37
= 3.0 pCi/gm for the 8 alpha decays A
=
a 6x0.37
= 2.2 pCi/gm for the 6 beta decays A
=
b 8x0.37
= 3.0 pCi/gm for the 8 gamma decays A
=
g
= A +A +A = 8.2 pCi/gm for all decay processes At a
3 g
O 3-33
TABLE 4.5.4 O
PRIMORDIAL RADIOACTIVITY IN CONCRETE TBORIUM SERIES SOURCES Nuclear Reactions Th-232 ---> Pb-208 + 6 alphas + 4 betas + approx. 6 gam /xrays 10 1/2 1.41 x 10 years T
=
42.7 MeV Rest mass difference, see Blb 4.5.1.
D-Mass
=
D-alpha
=
35.8 MeV Kinetic energy, see Bib 4.5.2.
3.2 MeV See Bib 4.5.2.
D-beta
=
D-recoil 0.6 MeV-Momentum conservation, nuclear
=
recoils 39.6 MeV Sum of the particle energies D-part.
=
42.7 - 39.6 = 3.1 MeV Difference of above D-gamma
=
35.8/6
= 6.0 MeV per elpha E
=
a 3.2/4
= 0.8 MeV per beta E
=
b
()\\
S0ecific Activity3.1/6
= 0.5 MeV per gamma E
=
s_
0.7 pCi/gm Measured Th-232 single A1
=
alpha decay 6x0.7 = 4.2 pCi/gm for the 6 alpha decays A
=
a b
4x0.7 = 2.8 pCi/gm for the 4 beta decays A
=
6x0.7 = 4.2 pCi/gm for the 6 gamma decays A
=
g A
= A +A +Ag = 11.2 pCi/gm for all decay products t
a b
O 3-34
[~'
TABLE 4.5.5 PRIMORDIAL RADIOACTIVITY IN CONCRETE
SUMMARY
OF BACKGROUND RADIOACTIVITY SOURCE SPECIPIC ACTIVITY OF THE VARIOUS PROCESSES ENERGY E
E RAIJIATION ALPHA BETA GAMMA SUBTOTALS GAMMAS
(--------------pCi/gm---------------)
(MeV)
K-40 15 2
17 1.5 Uranium series 3.0 2.2 3.0 8.2 0.3 Thorium series 4.2 2.8 4.2 11.2 0.5 Subtotals 7.2 20.0 9.2 36.4 Average 0.7
\\
O 3-35
TABLE 4.5.6 PRIMORDIAL RADIOACTIVITY IN CONCRETE EXTERNAL EMISSION OF RADIATION Enerov Density Factor in Uniformiv Activated Material d
2.13 (uR/hr) / (pci/gm) / MeV K
=
Specific Enercy Deposition from Uniformlv Activated Material ITZ5 UNITS PER PARTICLE TOTAL ALPHA BETA GAMMA E
MeV/ particle 5.7 1.2 0.7 x
A pCi/gm 7.2 20.0 9.2 36.4 x
ix (pCi-MeV)/gm 41.0 24.0 6.4 71.4 D
2x uR/hr (=K xDix) 87.3 51.1 13.6 152.1 D
d D
uR/hr (=D2t) 152.1 sx BU (ul,E )
2.1 g
uR/hr (=BUxD2) 28.6 Dyg 9
ft uR/hr (=0.5xDst) 76.0 D
fg uR/hr (=0.5xDig) 14.3 D
Ex, A,
From previous table x
Dix = (E xA)
Specific energy source, pCi-MeV/gm.
x D2x = (KdxD x)
Specific energy density, uR/hr.
D
=D2t Specific energy deposited in test material, gx in small cavity in the uniformly activated material.
BU Buildup factor, for 1 relaxation depth and for gamma energy, E
( see Bib 4.5.2).
a Dig = (BU x D2g)
Specific energy, r%ceptor in large cavity, beyond range charged particles from surface.
Dft= ( 0.5 x Dst)
Specific energy, receptor near surface uniformly activated material (2pi geometry).
fg = ( 0.5 x Dig)
Specific enerc;, receptor beyond range of D
charged particles from flat surface (2pi geometry).
fg = 13.8 uR/hr Calculated value, Bib 4.5.3 method.
D 3-36
rm TABLE 4.5.7
)
PRIMORDIAL RADIOACTIVITY IN CONCRETE EXTERNAL EMISSION OF RADIATION AT SURFACE ALPHA SURFACE ACTIVITY 2
=
K X A dpm/100cm Surface alpha activity Asa a
a 2
0.005 gm/cm Range of 5.7 MeV alphas R
=
0.5 Effective depth emission factor X
=
0.5 Effective solid angle factor O
=
Kd 222 dpm/100pCi Conversion factor
=
2 RxXx0xKd = 0.28 (dpm/100cm )/(pCi/gm)
K
=
a 7.2 PCi/gm Specific activity of alphas A
a 2
2.0 dpm/100cm Surface activity from alphas A
=
sa BETA SURFACE ACTIVITY 2
d m/100cm Surface beta activity A
KbxAb P
=
sb 2
0.25 gm/cm Mean range of betas R
=
0.125 Effective depth emission factor X
=
0.5 Effective solid angle factor O
=
2 RxXx0xKd = 3.5 (dpm/100cm )/(pCi/gm)
K
=
b b
20.0 pCi/gm Specific activity of betas A
=
2 i
Asb 69 dpm/100cm Surface activity from betas
=
GAMMA SURFACE ACTIVITY 2
12 gm/cm Mean range of gammas R
=
0.083 Effective attenuation, depth, XO
=
solid angle factor 2
R x XO x Kd = 222 (dpm/100cm )/(pci/gm)
K
=
g 9.2 pCi/gm Specific activity of gammas A
=
g 2
2000 dpm/100cm Surface gamma activity A
=
sg BETA-GAMMA SURFACE ACTIVITY 2
2300 dpm/100cm Asbg = Asb+Asg
=
3-37
TABLE 4.5.8 PRIMORDIAL RADIOACTIVITY IN CONCRETE RADON COMPONENT NUCLEAR REACTIONS U238 ----> Rn222 + 4 ALPHAS + 2 BETAS + 4 GAMMAS Rn222 ---> Pb206 + 4 ALPHAS + 2 BETAS + 4 GAMMAS Th232 ---> Rn220 + 3 ALPHAS + 2 BETAS + 3 GAMMAS Rn220 ---> Pb208 + 3 ALPHAS + 2 BETAS + 3 GAMMAS K40 -----> Ca40 0.89 BETAS K40 -----> Ar40 0.11 GAMMAS The Radon isotope splits the Uranium series and the Thorium series into two approximately equal decay components.
Since Radon is a noble gas it will tend to diffuse out of the surface of the concrete, and therefor will reduce the surface activity of the concrete if the surface is exposed.
Since Rn222 has a halflife of 3.82 days, and Rn220 a halflife of 55.6 seconds, the diffusion time is relatively short.
The halflife of the remaining decay products are all very short except for Pb210, which is 22 years.
()
SPECIFIC ACTIVITY The Radon component represent 50 percent of the alphas, 12 percent of the betas, and 25 percent of the gammas.
CATEGORY l--------PARTICLE--------l ALPHA BETA GAMMA SUBTOTALS Radon independ.
3.6 17.6 6.9 28.1 pCi/gm Radon dependent 3.6 2.4 2.3 8.3 pa'i/gm TOTALS 7.2 20.0 9.2 36.4 pCi/gm SURFACE ACTIVITY CATEGORY l---------PARTICLE-------l ALPHA BETA GAMMA SUBTOTLAL 2
At 1 cm (w/o Rn) 1 250 1500 1750 dpm/100cm 2
At 1 cm (with Rn) 2 280 2000 2300 dpm/100cm At 1 m (w/o Rn) 10.2 10.2 uR/hr At 1 m (with Rn) 14.3 14.3 uR/hr In cavity (w/o Rn) 21.4 21.4 uR/hr
()
In cavity (with Rn) 28.6 28.6 uR/hr 3-38
TABLE 5.1.1 t
TRANSFER OF NORTHROP FUEL ELEMENTS FUEL ROD RADIATION LEVELS FUEL ROD ACTIVATION M y M ACTIVITY ACTIVITY RATIO TERMINATED MEASURED MEASURED CALCULATED MEASURED CALCULATED (TYPE-LOC) (DATE)
(DATE)
(mR/HR)
(mR/HR)
REG. AVER 12/84 4/85 1100+
1430 0.8 1
REG.
B-C 12/84 6/85 1610 1430 1.2 1
AVERAGE 1355 1430 1.0 1
T.C.
B-C 1/68 6/85 99 83 0.07 0.06 PARAMETER DEFINITION REG.
Standard Fuel Rods, in Reactor at shutdown T.C.
Rods with Thermocouples, removed from core in 1968.
3 MEASURED ACT.
Activity measured at 3 feet from rod j
CALCULATED ACT Calculated using,(mR/HR)/ mci = 0.5 x n x E x F n
=1 Photons per disintegration E
= 0.7 MeV Gamma Energy F
=1 Ratio Flux at Loc. to Average Flux Ci
= 4.lCi/ rod: Calculated - Table 4.2.4 Note : Activity ratio is the measured or calculated value divided by the average value.
O 3-39 1
. - -..I
TABLE 5.1.2 O
TRANSPER OF NORTHROP PUEL ELEMENTS FIRST SHIPMENT TO UNIVERSITY OF TEXAS AT AUSTIN ROD SERIAL RING INIT.
INIT.
INIT. BURNUP BURNUP FINAL FINAL FINAL TYPE NO.
NO. U235 URANIUM CONC.
U235 U238 U235 URANIUM CONC.
(GRMS)
(GRMS)
(PER) (GRMS) (GRMS) (GRMS)
(GRMS) (PER)
FUEL 2931E D10 36.15 180.75 20.00 0.66 0.06 35.49 180.03 19.71 FUEL 2938E E4 40.11 200.55 20.00 0.73 0.07 39.38 199.75 19.71 FUEL 2964E E3 37.22 186.10 20.00 0.68 0.06 36.54 185.36 19.71 FUEL 2910E E2 37.27 186.35 20.00 0.68 0.06 36.59 185.61 19.71 FUEL 2947E El 39.37 196.85 20.00 0.72 0.06 38.65 196.07 19.71 FUEL 2935E D18 40.03 200.15 20.00 0.73 0.07 39.30 199.35 19.71 FUEL 2971E C12 39.58 197.90 20.00 0.72 0.06 38.86 197.12 19.71 FUEL 2968E D17 40.02 200.10 20.00 0.73 0.07 39.29 199.30 19.71 FUEL 2957E D16 40.17 200.85 20.00 0.73 0.07 39.44 200.05 19.72 FUEL 2962E D15 37.64 188.20 20.00 0.69 0.06 36.95 187.45 19.71 FUEL 2928E D14 37.68 188.40 20.00 0.69 0.06 36.99 187.65 19.71 FUEL 2940E E24 39.41 197.05 20.00 0.72 0.06 38.69 196.27 19.71 FUEL 2912E C11 38.95 194.75 20.00 0.71 0.06 38.24 193.98 19.71 FUEL 2908E B6 40.30 201.50 20.00 0.74 0.07 39.56 200.69 19.71 FUEL 2930E D2 39.78 198.90 20.00 0.73 0.07 39.05 198.10 19.71 O FUEL 2983E F436.72 183.60 20.00 0.67 0.06 36.05 182.87 19.71 FUEL 2979E F3 36.91 184.55 20.00 0.67 0.06 36.24 183.82 19.71 FUEL 2955E F2 38.40 192.00 20.00 0.70 0.06 37.70 191.24 19.71 FUEL 2985E F1 35.74 178.70 20.00 0.65 0.06 35.09 177.99 19.71 FUEL 2943E F30 37.79 188.95 20.00 0.69 0.06 37.10 188.20 19.71 FUEL 2978E E23 39.40 197.00 20.00 0.72 0.06 38.68 196.22 19.71 FUEL 2974E E22 38.04 190.20 20.00 0.69 0.06 37.35 189.45 19.71 FUEL 2950E E21 38.34 191.70 20.00 0.70 0.06 37.64 190.94 19.71 FUEL 2977E E20 37.82 189.10 20.00 0.69 0.06 37.13 188.35 19.71 FUEL 2906E E19 37.72 188.60 20.00 0.69 0.06 37.03 187.85 19.71 FUEL 2932E E18 37.76 188.80 20.00 0.69 0.06 37.07 188.05 19.71 FUEL 2984E E17 37.91 189.55 20.00 0.69 0.06 37.22 188.80 19.71 FUEL 2944E F22 36.58 182.90 20.00 0.67 0.06 35.91 182.17 19.71 FUEL 2905E F23 38.15 190.75 20.00 0.70 0.06 37.45 189.99 19.71 FUEL 2945E F24 38.84 194.20 20.00 0.71 0.06 38.13 193.43 19.71 SUM 30 1149.80 5749.00 20.00 20.99 1.89 1128.81 5726.12 19.71 HQIE Ring Number signifies the location of the fuel element in the core at shutdown.
O 3-40
TABLE 5.1.3
\\
TRANSFER OF NORTHROP FUEL ELEMENTS SECOND SHIPMENT TO UNIVERSITY OF TEXAS AT AUSTIN ROD SERIAL RING INIT.
INIT. INIT. BURNUP BURNUP FINAL FINAL FINAL TYPE NO.
NO. U235 URANIUM CONC.
U235 U238 U235 URANIUM CONC.
(GRMS) (GRMS) (PER) (GRMS) (GRMS)
(GRMS)
(GRMS) (PER)
FUEL 2975E F25 38.50 192.50 20.00 0.70 0.06 37.80 191.74 19.71 FUEL 2960E F26 39.59 197.95 20.00 0.72 0.06 38.87 197.17 19.71 FUEL 2911E F27 38.73 193.65 20.00 0.71 0.06 38.02 192.88 19.71 FUEL 2903E F28 40.05 200.25 20.00 0.73 0.07 39.32 199.45 19.71 FUEL 2899E F29 36.17 180.85 20.00 0.66 0.06 35.51 180.13 19.71 FUEL 2976E C1 40.41 202.05 20.00 0.74 0.07 39.67 201.24 19.71 FUEL 2952E B1 42.43 212.15 20.00 0.77 0.07 41.66 211.31 19.72 FUEL 2970E D3 40.23 201.15 20.00 0.73 0.07 39.50 200.35 19,72 FUEL 2958E C3 39.26 196.30 20.00 0.72 0.06 38.54 195.52 19.71 FUEL 2925E F6 36.76 183,00 20.00 0.67 0.06 36.09 183.07 19.71 FUEL 2980E ES 36.29 181,65 20.00 0.66 0.06 35.63 180.73 19.71 FUEL 2939E D4 37.40 187.00 20.00 0.68 0.06 36.72 186.26 19.71 FUEL 2904E B2 42.72 213.60 20.00 0.78 0.07 41.94 212.75 19.71 FUEL 2959E D5 37.38 186.90 20.00 0.68 0.06 36.70 186.16 19.71 FUEL 2951E E6 39.77 198.85 20.00 0.73 0.07 39.04 198.05 19.71 FUEL 2941E F7 36.78 183.90 20.00 0.67 0.06 36.11 183.17 19.71
\\
FUEL 2948E C4 40.57 202.85 20.00 0.74 0.07 39.83 202.04 19.71 FUEL 2969E D6 39.11 195.55 20.00 0.71 0.06 38.40 194.78 19.71 FUEL 2954E E7 40.04 200.20 20.00 0.73 0.07 39.31 199.40 19.71 FUEL 2965E F8 36.34 181.70 20.00 0.66 0.06 35.68 180.98 19.71 FUEL 2929E E8 38.35 191.75 20.00 0.70 0.06 37.65 190.99 19.71 FUEL 2913E C5 40.11 200.55 20.00 0.73 0.07 39.38 199.75 19.71 FUEL 2946E C6 38.79 193.95 20.00 0.71 0.06 38.08 193.18 19.71 FUEL 2918E D8 38.68 193.40 20.00 0.71 0.06 37.97 192.63 19.71 FUEL 2902E C7 39.04 195.20 20.00 0.71 0.06 38.33 194.43 19.71 FUEL 2927E B4 40.23 201.15 20.00 0.73 0.07 39.50 200.35 19.72 FUEL 2915E C8 39.04 195.20 20.00 0.71 0.06 38.33 194.43 19.71 T.C. 5283E B5 38.00 190.00 20.00 0.69 0.06 37.31 189.25 19.71 T.C. 2990E B3 39.63 198.15 20.00 0.72 0.06 38.91 197.37 19.71 SUM 29 1130.40 5652.00 20.00 20.64 1.86 1109.76 5629.50 19.71 HQIE Ring Number signifies the location of the fuel element in the core at shutdown.
O 3-41
TABLE 5.1.4 TRANSFER OF NORTHROP FUEL ELEMENTS SHIPMENT TO KANSAS STATE UNIVERSITY ROD SERIAL RING INIT.
INIT. BURNUP BURNUP FINAL FINAL FINAL FINAL TYPE NO.
NO. U235 URANIUM CONC.
U235 U238 U235 URANIUM CONC.
(GRMS) (GRMS) (PER) (GRMS) (GRMS)
(GRMS)
(GRMS) (PER)
FUEL 2963E D12 38.60 193.00 20.00 0.70 0.06 37.90 192.24 19.71 FUEL 2933E E16 38.55 192.75 20.00 0.70 0.06 37.85 191.99 19.71 FUEL 2949E E15 38.40 192.00 20.00 0.70 0.06 37.70 191.24 19.71 FUEL 2909E E14 37.61 188.05 20.00 0.69 0.06 36.92 187.30 19.71 FUEL 2911E E13 36.72 183.60 20.00 0.67 0.06 36.05 182.87 19.71 FUEL 2986E E12 38.53 192.65 20.00 0.70 0.06 37.83 191.89 19.71 FUEL 2987E Ell 38.79 193.95 20.00 0.71 0.06 38.08 193.18 19.71 FUEL 2937E F13 36.89 184.45 20.00 0.67 0.06 36.22 183.72 19.71 FUEL 2907E F14 39.55 197.75 20.00 0.72 0.06 38.83 196.97 19.71 FUEL 2942E F15 36.18 180.90 20.00 0.66 0.06 35.52 180.18 19.71 FUEL 2953E F16 40.12 200.60 20.00 0.73 0.07 39.39 199.80 19.71 FUEL 2934E F17 38.88 194.40 20.00 0.71 0.06 38.17 193.63 19.71 FUEL 2900E F18 39.59 197.95 20.00 0.72 0.06 38.87 197.17 19.71 FUEL 2982E F19 34.70 173.50 20.00 0.63 0.06 34.07 172.81 19.72 FUEL 2966E F20 38.40 192.00 20.00 0.70 0.06 37.70 191.24 19.71 O'T.C.2989E--
FUEL 2914E F21 38.20 191.00 20.00 0.70 0.06 37.50 190.24 19.71 37.56 187.80 20.00 0.69 0.06 36.87 187.05 19.71 T.C. 2988E --
38.66 193.30 20.00 0.71 0.06 37.95 192.53 19.71 SUM 18 685.93 3429.65 20.00 12.52 1.13 673.41 3416.00 19.71 HQIE Ring Number signifies the location of the fuel element in the core at shutdown.
O 3-42
TAE)E 5.1.5 TRANSPER OF NORTHROP FUEL ELEMENTS SHIPMENT TO UNIVERSITY OF ILLINOIS ROD SERIAL RING INIT.
INIT. INIT. BURNUP BURNUP FINAL FINAL FINAL TYPE NO.
NO.
U235 URANIUM CONC.
U235 U238 U235 URANIUM CONC.
(GRMS)
(GRMS) (PER) (GRMS) (GRMS)
(GRMS)
(GRMS)
(PER)
FUEL 2972E F9 39.81 199.05 20.00 0.73 0.07 39.08 198.25 19.71 FUEL 2981E F10 38.98 194.90 20.00 0.71 0.06 38.27 194.13 19.71 FUEL 2967E E9 38.76 193.80 20.00 0.71 0.06 38.05 193.03 1 9.71 FUEL 2926E Fil 39.87 199.35 20.00 0.73 0.07 39.14 198.55 19.71 FUEL 2961E E10 40.07 200.35 20.00 0.73 0.07 39.34 199.55 19.71 FUEL 2973E F12 36.00 180.00 20.00 0.66 0.06 35.34 179.28 19.71 FUEL 2956E C9 38.53 192.65 20.00 0.70 0.06 37.83 191.89 19.71 FUEL 2916E C10 38.98 194.90 20.00 0.71 0.06 38.27 194.13 1 9.71 FUEL 2901E D9 38.70 193.50 20.00 0.71 0.06 37.99 192.73 19.71 FUEL 2936E D11 40.08 200.40 20.00 0.73 0.07 39.35 199.60 19.71 T.C. 2991E C2 40.27 201.35 20.00 0.74 0.07 39.53 200.54 19.71 SUM 11 430.05 2150.25 20.00 7.85 0.71 422.20 2141.69 19.71 O
\\--
HQIE Ring Number signifies the location of the fuel element in the core at shutdown.
3-43
TABLE 5.1.6 TRANSPER OF NORTHROP PUEL ELEMENTS
SUMMARY
OF SHIPMENTS SILIP COUNT INITIAL INITIAL INIT. BURNUP BURNUP PINAL PINAL PINAL
[1235 URANIUM CONC.
11215 112.33 1123.5 URANIUM Q L (UNIV) (NO. ) (GRAMS)
(GRAMS) (PER) (GRMS) (GRMS)
(GRMS)
(GRMS) (PER)
TEXA 59 2280.20 11401.00 20.00 41.63 3.75 2238.57 11355.62 19.71 KANS 18 685.93 3429.65 20.00 12.52 1.13 673.41 3416.00 19.71 ILLI 11 430.05 2150.25 20.00 7.85 0.71 422.20 2141.69 19.71 SUM 88 3396.18 16980.90 20.00 62.00 5.58 3334.18 16913.32 19.71 O
3-44 i
TABLE 5. 2.1 REACTOR HARDWARE RADIOACTIVITY AND DISPOSITION RECIPIENT ITEM DIMENSIONS RADIATION READING UNIVERSITY (INCHES)
(A)
(B)
U. OF ARIZONA COMPENSATED ION CHAMBER 85 X 12 X 4 8
16 INCORE CHAMBER COOL FUEL HANDLING TOOL 122 X 14 X 4 0
FUEL MEASURING FIXTURE 34 X 24 X 3 0.5 RM-1 STACK MONITOR 1
U.C. BERKELEY FISSION POILS REMOTE RECORDER-RM-COLD DRIVE MOTOR REG. ROD COLD DIFFUSION PUMP COLD AUTOMATIC SAMPLE CHANGER COLD CORNELL UNIV. BEAM PORT PLUG 49 X 8D.
0.6 B. PORT PLUG WITH 6"Pb 43 X 8D.
0.7 B. PORT PLUG WITH 6"Pb 43 X 8D.
1.3 I)
B. PORT PLUGS-BOR. CON.(5)30 X 8D.
0.1 k/
B. PORT PLUGS-BOR. CON.(3)30 X 8D.
COLD B. PORT PLUG-WOOD 49 x 8D.
COLD B. PORT PLUG-GRAPHITE 38 X 8D.
COLD B. PORT PLUG-GRAPHITE 37 x 6D.
COLD B. PORT BORATED COLLIM.
12 X 6D.
COLD B. PORT PLUG 39 X 9D.
COLD COMPENSATED ION CHAMB. 102 X 6X5 1
6 COMPENSATED ION CHAMB. 115 X 13 X 3 U.C.
IRVINE FISSION CHAMBER 12 X 1D.
0.2 GAS PROP. COUNTER COLD ION CHAMBER COLD POOL MONITOR-RM-2 COLD STORAGE CASKS-PIES COLD SHIELDED WELL COUNTER COLD ELECTRONIC PICO AMP SOURCE COLD COUNTING EQUIPMENT COLD DETECTOR EQUIPMENT COLD S0-U-BRIDGE COND. METER COLD DECADF RESISTOR BOX COLD POOL LIGHT COLD VACUUM HOSE AND FILTER COLD Note : Column A is radiation reading, in mR/hr at contact.
nO.
Column B is distance, in inches, at which radiation equals 0.5 mR/hr.
3-45
TABLE 5.2.1 (Cont.)
REACTOR HARDWARE RADIOACTIVITY AND DISPOSITION RECIPIENT ITEM DIMENSIONS RADIATION READING UNIVERSITY (INCHES)
(A)
(B)
KANSAS STATE PULSE / CONTROL ROD 34 X 2D.
170 REED UNIV.
FISSION CHAMBER 116 X 3 X 1 40 36 FIS. CH. SECTION 120 X 12 X 3 COLD VAN MONITOR-RM-3 CONTROL CONSOLE UNIV. OF UTAH FISSION CHAMBER 87 X 4X 4
40 36 FIS. CH. SECTION 132 X 10 X 3
COLD COMPENSATED ION CH.218 X 12 X 12 18 20 UNCOMPENSATED CH.
84 X 4X 4
19 20 UNCOMP. CH. SECT.
127 X 13 X 4
COLD REG CONTROL ROD 36 X 2D. -l SHIM CONTROL ROD 36 X 2D.
l-240 SAFE CONTROL ROD 23 X 2D. -l C.R./ GUIDE & THIMBLES 55 X 3D.
142 35 t
PNEUMATIC TUBES ETC.89 X 12 X 19 3
14 PNEUKATIC TUBES ETC.94 X 2X 2
0.5 -
PNEUMATIC TUBE BENDS 37 X 55 X 2
0.5 -
PNEUKATIC TUBE PUMPS ETC COLD CONTROL ROD DRIVES 83 X 1D.
COLD PORTABLE RAD. DET.
COLD STORAGE CASKS COLD RAD AREA MON.-RM-7 COLD PENN STATE NEUT SOURCE (3 Ci AmBe)
HOT Note : Column A is radiation reading, in mR/hr at contact.
Column B is distance, in inches, at which radiation equals 0.5 mR/hr.
O 3-46
\\
TABLE 5.4.1 EXPOSURE ROOM CAVITY DOSE RATE EXPERIMENT A Ludlum Low Level Gamma Scintilator, with a 1 inch diameter by 1 inch long sodium iodide detector was placed in the center of the exposure room.
It measured 21.3 uR/hr activity.
A2 inch thick by 8 inch wide by 16 inch long layer of lead was placed below the detector, with the center of the detector at 1.375 inches above the center of the surface.
The meter read 14 uR/br.
The detector was placed below the layer of lead at 1.875 inches below the surface.
The meter read 15.5 uR/hr.
The detector was completely surrounded with 2 inches of lead.
The meter read zero.
ANALYSIS UPPER HEMISPHERE AU = 14 uR/hr SU = 2 PI + 8 x ARCSIN ( l.4 + ( l.42+ (42+82 )).5 )
SU = 6.28 + 1.23
= 7.5 Steradians ASU= AU/SU
= 1.87 uR/(hr-ster)
C\\
'Q LOWER HEMISPHERE AL = 15.5 uR/hr SL = 2 PI + 8 x ARCSIN ( 1.9 + ( 1.92+(42+82 )).5 )
SL = 6.28 + 1.65
= 7.93 Steradians ASL= AL/SL
= 1. 95 uR/ (hr-ster)
AVERAGE ACTIVITY ASA= ( ASU + ASL ) / 2 = 1.89 uR/(hr-ster)
AT = 4 PI x ASA
= 23.7 uR/hr (calculated)
AM = 21.3 uR/hr (measured)
The minor discrepancy between the calculated and measured total activity indicates the actual solid angles for the upper and lower measurements were slightly larger than calculated.
BACKGROUND FLAT SURFACE ACTIVITY AC = AT / 2
= 12.3 uR/hr ( Calculated for 2 pi geometry)
AM = 11.6 uR/hr ( Measured background - see Table 6.1.1 )
The agreement indicates the concrete in the exposure room is at or near background level.
3-47
7-s TABLE 5.6.1 E IOCHEMISTRY LABORATORY DjCONTAMINATION DATA SURVEY LOCATION AVERAGE SURFACE READING SYMBOL IDENTIFICATION ALPHAS 2
(dpm/100 cm )
A Well Surface Plate 95 A
Well Cover Plate 157 A
Well Wall 302 A
Well Bottom 335 B
Well Surface Plate 83 B
Well Cover Plate 166 B
Well Wall 236 B
Well Bottom 389 C
Cabinet Top Surface 199 C
Cabinet Underside Floor 1702 x
~
EQIE 1)
Eberline PAC 4G, Serial Number 2114 meter.
2)
Readings are average readings for each surface.
3)
All areas decontaminated, or material removed and shipped to Hanford Site for burial.
1 3-48 l
J
TABLE 5.7.1 VENTILATION DUCTS DECONTAMINATION DATA f OVERHEAD LOCATIONS )
LOCATION REMOVABLE RADIOACTIVITY NUMBER ALPHA EETA GAMMA 2
(dpm / 100 cm 3 1
0 0.9 200 2
0.4 2.8 175 3
0 2.2 203 4
0 6.9 164 5
0.8 3.1 157 6
1.6 2.8 371 7
0.4 1.8 235 8
0 2.8 207 9
2.0 1.5 107 10 0.8 2.5 218 11 1.2 0.9 342 12 0.4 4.7 328 13 0
0.9 264 O
14 0.4 1.2 82 15 1.2 1.8 96 16 0
4.0 86 Average 0.6 2.6 204 Allowed 200.0 1000.0 1000 Note : Location number 25 and 26 are in the holdup room ventilation ducts - 25 inside vent caover and 26 inside vent pole.
O 3-49
TABLE 5.7.1 (cont.)
VENTILATION DUCTS DECONTAMINATION DATA
( WALT. LOCATIONS 1 LOCATION REMOVABLE RADIOACTIVITY NUMBER ALPHA BETA GAMMA 2
(dom / 100 cm 3 1
0 0.9 0
2 0.3 3.7 0
3 0.3 4.0 0
4 0
0.9 0
5 0.7 2.8 0
6 0
2.5 0
7 0.7 2.1 0
8 0.7 2.1 0
9 0
0.3 0
10 0.3 0
0 11 0.3 4.0 0
12 0
4.6 0
13 0
1.8 0
14 0
0.9 0
(,
15 0.3 1.8 0
16 0
5.3 0
17 1.1 2.8 0
18 0.3 0.3 0
19 0
0.3 0
20 0.3 7.5 0
21 0.3 0.3 0
22 0
1.8 0
23 0
0.6 0
24 0.7 4.0 0
25 0
2.8 0
26 0
7.8 0
Average 0.2 2.5 0
Allowed 200.0 1000.0 1000 Note : Location number 25 and 26 are in the holdup room ventilation ducts - 25 inside vent caover and 26 inside vent pole.
O 3-50
TABLE 5.7.2 O
PLUMBING DECONTAMINATION DATA LOCATION DRY SCRAPINGS RADIOACTIVITY NUMBER ALPHA BETA GMiMA l---------pci/gm-----------I 1
1 2
0 2
4 18 57 3
20 588 2106 4
6 249 177 5
1 17 0
6 3
63 110 7
1 27 60 8
2 35 0
9 1
9 38 10 0
36 223 11 0
6 0
12 2
23 68 13 0
12 64 O
HQTE Survey samples were taken at the locations indicated in Figure 5.7.2.
The.camples were scraped from the interior of'the pipes, dried, and then counted in the Tennelec counting system.
The drain pipes from the location number (3) to beyond location number (10 ) were removed and shipped to the Banford burial site.
O 3-51
TABLE 5.8.1 l
RADIOACTIVE WASTE SHIPMENTS SHIP. QAIE WEIGHT VOLUME NUMBER ACTIVITY MAJOR HQ.
RIES CONTRIBUTORS 3
(1bs)
(ft )
(mci) 1 7/15 43810 672 7LSA 211 CONC. REB. WOOD 2
7/17 43400 672 7LSA 168 CONC. REB. WOOD 3
7/19 44080 768 8LSA 142 CONC. REB. WOOD 4
7/31 44930 576 6LSA 265 CONC. REB. WOOD 4
7/31 (Included in shipment)
(21400)
(Sealed Tritium Sources)
(Ra225 foils (84uCi) 5 8/8 44470 576 6LSA 204 CONC.REBAR (Ra226 foils (66uCi)
()6 8/15 43210 480 SLSA 293 CONC.REBAR 7
8/22 39880 536 SLSA 228 CONC.REBAR SDRUMS HT.EXCH'R TOTALS 303780 4280 44LSA
.1511 CONCRETE (138 MT) 5 DRUMS REBAR HT.EXCH'R WOOD (Sealed Sources) (21400)
(Tritium)
ESTIMATES OF MATERIAL COMPOSITION LOCATION MATERIAL VOLgME DENSI{Y MASS (m )
(MT/m )
(MT)
EXPOSURE ROOM CONCRETE 44 2.4 110 EXPOSURE ROOM REBAR 2.5 7.9 20 EXPOSURE ROOM WOOD 8
0.5 4
BEAM PORTS B.P., CONC.
2 2.4 4
()
TOTAL 56.5 2.4 138 3-52
\\
~'
TABLE 5.9.1 s_-
RESIDUAL CONCRETE RADIOACTIVITY RADIOACTIVITY AT BURIAL TIME RADIO CONTROLLING l SURFACE 22 MET. TONS l REMAINING 1000 MET. TONS l ISOTOPE LIFETIME l RADIO RADIOACTIVE l RADIO RADIOACTIVE l
l l ACTIVITY MOLES l ACTIVITY MOLES l
l (YEARS) l(uCi)
(uMoles) l (uCi)
(uMoles) l (XA)
(TAX) l(AXNM)
(NXNM) l (AXNM)
(NXNM) l l
I NEUTRON CAPTURE RADIOISOTOPES l
FESS 3.9 286 0.002 0
0 l
CO60 7.6 55 0.001 0
0 EU152 19.6 194 0.007 0
0 l
EU152 12.7 15 0.001 0
0 SUBTOTAL 550 0.011
~~
0 0
PRIMORDIAL RADIOISOTOPES
()
l(AXBM)
(NXBM) l (AXBM)
(NXBM) l i
K40 1.81 E 9
374 1310000 17000 60000000 URSER 6.51 E 9
114 1440000 5200 66000000 THSER 2'.03 E 10 154 6060000 7000 66000000 SUBTOTAL 642 8820000 29200 401000000
=================================-
w============================
TOTAL 1192 8820G'A0 29200 401000000 AXNM = KN AXN M Neutron capture activity in concrete.
AXBM = KN AXB M Natural background alpha and beta activity.
KN
= 1.94 E -6 uMoles/uCi/ year -
Conversion factor AXN Specific activity from neutron capture at the surface of the remaining concretc.
AXB Specific activity from natural primordial radioactivity in the concrete.
NXNM =
AXNM TAX Number neutron capture radioisotope nuclei.
NXBM =
AXBM TAX Number primordial radioisotope nuclei.
TAX Mean lifetime of radioisotope gS t
(J 3-53
f-sg TABLE 5.9.2
\\',)
RESIDUAL CONCRETE RADIOACTIVITY LANDPILL RADIOACTIVITY LIMITS RADIO RADIO ALLOWED ALLOWED FRACTION ISOTOPE ACTIVITY ACTIVITY ACTIVITY ALLOWED LIMIT RASE (1000 MT) (EACH BUR) (PER YEAR) (PER YEAR)
!--------------uCi-------------l---percent--l Note (1)
(2)
(3)
NEUTRON PRODUCED ACTIVITY IN CONCRETE FESS 286 100,000 1,200,000 0.02 CO60 55 1,000 12,000 0.42 EU152 194 1,000 12,000 1.62 EU154 15 1,000 12,000 0.12 TOTAL 550 2.18
=======================================
NATURAL PRIMORDIAL ACTIVITY IN CONCRETE K40 17,000 100 1,200 1417 THORIUM 7,000*
50,000 600,000 1.2
(~N URANIUM 5,200*
50,000 600,000 0.9 y) 1419.1 TOTAL 29,200*
- Activities include daughter products.
========================================
NATURAL PRIMORDIAL ACTIVITY IN SOIL K40 17,200 100 1,200 1430 THORIUM
not measured------------------
URANIUM 5,200 50,000 600,000 0.09 TOTAL 22,400 1430.09 Note
- 1) The radiactivities listed in the table are the measured values given in Table 5.9.1.
- 2) The allowable levels of radioactivity are listed in the California Radiation Control Regulations, Title 17, Ap B.
- 3) The neutron produced radioactivity in the concrete rubble is only 2.1 percent of the allowei buriel limit for twelve burials separated by a distance of at least six feet.
The primordial activity in the concrete rubble is less than the natural primordial activity measured in ordinary dirt, and
(~'s) is much less available since it is chemically bonded in the concrete.
3-54
TABLE 5.9.3 RESIDUAL CONCRETE RADIOACTIVITY EROSION RADIOACTIVITY ANALYSIS EROSION OF NEUTRON ACTIVATED SURFACE The dimensions of the exposure room, after the removal of 24 inches of the activated concrete, was 4.2 x 4.2 x 4.8 meters.
The remaining activity had a mean depth of 10 cm.
The total mass of concrete remaining in the concrete structure was approximately 1000 metric tons.
The rate at which activity is eroded into the soil can be estimated for a specific erosion rates of the concrete, K, and for a specific sizes of the concrete rubble, LR.
The analysis is for an erosion rates of 1 mm/100 yrs and for rubble with a uniform mass distribution from 1 cm cubes to 1 meter cubes.
EROSION OF NEUTRON ACTIVATED SURFACE 2
SN 106 m
Exposure room surface area - no window
=
wall.
0.00001 m/yr Erosion rate.
K
=
3 2.4 MT/m Density of concrete RC
=
AX Specific activity of radioisotope X at time of burial - Table 5.9.1.
y AX0= K x SN x RC x AX Activity of radioisotope, X,
at time of burial, for 1 years erosion.
TAX:
Mean life of radioisotope, X -
Table 5.9.1.
AXT= AKO x T x EXP(T/ TAX)
Activity of eroded radioisotope, X,
in soil after T years burial Table 5.9.4.
ANT = SUM (AXT)
Total radioactivity from neutron capture radioisotopes eroded into the soil after T years of burial - Table 5.9.4.
EROSION OF PRIMORDIAL ACTIVATED SURFACES 0.01 m
Smallest size cubic concrete rubble.
RL
=
1 m
Largest size cubic concrete rubble.
RH
=
1000 Total concrete mass.
M
=
AX0= K x M x LOG (RH/RL) / (RB-RLi x AX Activity ac burial - see above.
AXT and ABT As described above.
The eroded activity of the neutron-capture radioisotopes and primord-ial background isotopes, and the logarithmic values of the totals are tabulated in Table 5.9. 4, for various burial times.
These data are plotted in Figure 5.9.1.
3-55
(~T TABLE 5.9.4 V
RESIDUAL CONCRETE RADIOACTIVITY EROSION RADIOACTIVITY VERSUS TIME BURIED RADIO l-------- RADIOACTIVITY FROM ERODED CONCRETE------------l ISOTOPE FIRST FOURTH SIXTEENTH SIXTY-FOURTH 128TH YEAR YEAR YEAR YEAR YEAR l----------------------uCi-------------------------------I (XA)
(AX1)
(AX4)
(AX16)
(AX64)
(AX128)
NEUTRON CAPTURE FESS 0.029 0.05 0.01 0.00 0.000 CO60 0.006 0.02 0.01 0.00 0.000 EU152 0.024 0.08 0.18 0.06 0.005 EU154 0.002 0.01 0.01 0.00 0.000 Subtotal 0.061 0.16 0.21 0.06 0.005 PRIMORDIAL Ci K40 3.954 15.82 63.26 253.04 506.047
\\2 TH232 2.605 10.42 41.68 166.71 333.397 U238 1.935 7.74 30.96 123.83 247.648 Subtotal 8.494 33.98 135.90 543.58 1087.092
=======================================================
SUMMARY
OF TOTALS TIME l-------RADIOACTIVITY------l--------- LOG RADIOACTIVITY-----l AFTERl NEUTRON ACTIV.lPRIMOR ACTIVllTIMEllNEUT. ACTIVl [ PRIM.ACTIVl l
BUR.
lYRSl-----------uCi-------------l-----------
LOG + 3 -----------l (T) l--(ANT)-----l-----(ABT)----l-LT--l-LOG (ANT)----l--LOG (ABT)--l 1
0.061 8.5 0.000 1.733 3.969 4
0.160 34.0 0.628 2.170 4.597 16 0.210 135.9 1.256 2.293 5.225 64 0.060 543.6 1.884 1.726 5.853 128 0.005 1087.1 2.198 0.600 6.167 O
3-56
TABLE 5.9.5 RESIDUAL CONCRETE RADIOACTIVITY GROUNDWATER ANALYSIS The radioisotopes released from the eroded concrete are diluted by the rainfall infiltrating the landfill site.
The dilution by the direct flow of permeating water through the concrete rubble is estimated from the known average annual rainfall of 0.40 meters (13 inches - Bib 5.9.2) and an assumed infiltration f actor of 50 per-cent.
The concrete rubble is assumed to cover 1000 square meters.
The dilution for the rainfall over the total area of the site is based on a site area of 5,520,000 square meters (1365 acres - Bib 5.9.1).
2 SL
= 1000 m
Area of rubble 0.4 m
Annual rainfall (13 inches)
RF
=
0.5 Infiltration factor I
=
VL
= SL*RF*I = 200 m3 Annual direct dilution volume of water VLT = VL*T Total direct dilution at year T ANT :
Total neutron induced eroded R/A at T ABT :
Total primordial R/A at year T ANLT
= ANT /VLT Direct specific neutron induced activity ABLT
= ABT/VLT Direct specific primordial activity 3
VS
= SS*RF*I = 1,105,000 m Annual site dilution water 1
VST = VS*T Total site dilution at year T ANST = ANT /VST Site specific neutron induced activity ABST
= ABT/VST Site specific primordial activity 3
AGST
= 12300 pCi/m Specific activity of potassium-40 in ground water, due to 14 mg/l of element-al potassium presently existing in the groundwater solution (Bib 5.9.1).
Other primordial radioactivities may be in the groundwater, but they were not measured.
NOTE Data tabulated in Table 5.9.6.
O 3-57
TABLE 5.9.6
(
)
N/
RESIDUAL CONCRETE RADIOACTIVITY GROUNDWATER RADIOACTIVITY VERSUS TIME BURIED TIES TOTAL l
TOTAL l
LOCAL l
SITE l
l ACTIVITY l WATER VOLUMEl SPECIFIC ACT.l SPECIPIC ACTIVITY l lNEUT.
PRIM. l LOCAL E
NEUT.
PRIM.l NEtT. PRIM. GRD W.l lyrl-----uci-----l--1000m}TEl---l-------------pci/m}---------------l (T) l ( ANT) (ABT) l(VLT) (VST) l(ANLT)
( ABLT) l ( ANST) ( ABST) ( AGST) l 1
0.061 8.5 0.2 1000 305 42000 0.06 8
12300 4
0.160 34 0.8 4000 200 42000 0.04 8
12300 16 0.210 136 3.2 16000 66 42000 0.01 8
12300 64 0.060 544 12.8 64000 5
42000 0.001 8
12300 256 0.005 1090 25.6 128000 0.2 42000 0.0000 8
12300 ANT Data from Table 5.9.3 ABT Data from Table 5.9.5 VLT Data from Table 5.9.5 VST Data from Table 5.9.5 ANLT = ANT /VLT ABLT = ABT/VLT s
ANST = ANT /VST ABST = ABT/VST AGST :
Data from Table 5.9.5 NOTE Radioactivity di ssolved in the groundwater from the eroded concrete is negligible, and the neutron induced radio-activity will completely decay into stable isotopes without ever leaving the landfill site.
O 3-58
TABLE 6.
1.1 BACKGROUND
SURVEY DATA MEASUREMENT GAMMAS BETA-GAMMAS ALPHAS BETAS GAMMAS HQ, LOCATION AT 1 METER AT <1 CM l------REMOVABLE------------l (uR/hr) l------------(dpm/103cm2)--------------l 1
GATE 15 AREA 12 1612 0.20 0.40 34.3 1
GATE 15 AREA 12 1612 0.20 0.40 34.3 2
BLDG-3-55 12 1612 0.80 1.80 175.3 3
BLDG-3-55 11 1612 1.10 1.20 113.4 4
BLDG-3-55 9
1612 0.50 0.40 13.7 5
BLDG-3-55 11 1612 0.20 1.40 106.6 6
BLDG-3-55 12 3224 0.20 1.00 103.1 7
BLDG-3-10 12 1612 0.20 1.20 20.6 8
BLDG-3-10 12 1612 0.20 0.00 0.0 9
BLDG-3-10 11 1612 0.50 0.00 0.0 10 BLDG-3-61 12 1612 0.20 0.00 0.0 11 BLDG-3-61 11 1612 0.80 0.00 55.0 12 BLDG-3-61 12 1612 0.00 0.00 0.0 13 BLDG-3-7 9
1612 0.00 0.00 347.3
(-)
14 BLDG-3-7 11 1612 0.20 0.00 0.0 15 BLDG-3-7 11 1612 0.00 0.00 6.8 16 PARKLOT-747 12 1612 0.20 0.00 130.6 17 PARKLOT-74 7 12 1612 0.50 1.00 79.0 18 PARKLOT-747 12 1612 0.00 0.00 140.9 19 BLDG-1-153 12 1612 0.20 0.40 127.2 20 BLDG-1-153 12 1612 0.20 0.00 192.5 21 WINTUNLOT-1-76 12 1612 0.00 0.00 196.0 22 WINTUNLOT-1-76 12 1612 0.20 0.00 158.1 23 TRANSWHOUSE 12 1612 0.20 0.00 113.4 24 TRANSWHOUSE 12 1612 0.50 0.00 68.7 24 TRANSWHOUSE 12 1612 0.00 0.40 113.4 25 CRENSHAW-NORTHRO 12 3224 0.20 0.80 96.2 26 CRENSHAW-NORTHROP 12 1612 0.20 0.00 0.0 27 PRAIRIE-NORTHROP 12 1612 0.10 0.10 70.0 27 PRAIRIE-ELSEGUNDO 12 1612 0.10 0.10 70.0
==-
AVERAGE (30 READINGS) 11.6 1720 0.26 0.35 85.5 STANDARD DEVIATION 0.8 410 0.26 0.51 79.2 MAXIMUM LIKELYHOOD 12.7 2140 MAXIMUM LIKELIHOOD, B = exp (Inx + 1.28 ( (n-1)/1) s), Pg. 66, Bib 6.2.1 See Figure 6.1.1 - Map shows where measurements were made.
3-59
'~N TABLE 6.2.1 EXAMPLE OF HISTORICAL SURVEY RECORD f HEALTH PHYSICS SURVEY RECORD FOR JUNE 1984 )
AREA NUMBER AREA DESCRIPTOR WIPE MEASUREMENT d
( No. )
(dpm per 100 cm )
1 Lobby 1
2 Walkway
<1 3
Work Table
<1 4
Work Table
<1 5
Exposure Room Entrance 1
6 Exposure Room Plug Door 14 7
Exposure Room (Controlled) 170 8
Hot Cell Door
<1 9
Hot Cell (Controlled) 110 10 Storage Vault
<1 11 Shop Bench 13 12 Chem Lab Bood 5
13 Chem Lab Counter Top 13 14 Walkway
<1 15 Hot Cell Window
<1 16 Floor Area
<1 17 Work Area 9
x' 18 Pharmacy
<1 19 FXR Control Room
<1 20 Counting Room
<1 21 Counting Room Table 2
22 Beam Port No. 2
<1 23 Trench Area
<1 24 thru 28 Reactor Bridge Area
<1 2
Acceptable level for unrestricted use is < 20 dpm/100 gm Note:
for transuranium nuclide alphas, and < 1000 dpm/100 cm for beta-gammas (
Measured Natural background was 36dpm/100cm{able2.2.2).
O 3-60
TABLE 6.2.2
[K -}
l i
SUMMARY
OF SURVEY AREAS SURVEY AREAS BLOCK SIZE NUMBER OF BLOCKS BLOCKS SURVEYED (sq. meters)
============================....==============----========...
SURVEY GROUP - OVHD30 BLDG-OVHD 9
240 73 SUBTOTAL 9
240 73 Sf3RVEY GROUP - PLMZ30 MEZZANINE 9
27 8
UNDER-MEZ 9
47 14 UNDER-MIDN 9
30 9
SUBTOTAL 9
104 31 O
V SURVEY GROUP - DECK 50 DECK-NHANG 4
4 2
DECK-ER 4
4 2
DECK-LOWER 4
10 5
DECK-POOL 4
14 7
DECK-MIX 4
14 7
SUBTOTAL t.
46 23 SURVEY GROUP - ROOM 50 FANROOM 4
30 15 COUNT ROOM 4
104 52 DECON-HEAD 4
72 36 ISOMED 4
20 10 FRONT-RESTROOM 4
24 12 SUBTOTAL 4
250 125
()
Note Plan view of survey areas - Figure 6.2.1 3-61
TABLE 6.2.2 (cont.)
g-~g
(_)
SUMMARY
OF SURVEY AREAS SURVEY AREAS BLOCK SIZE NUMBER OF BLOCKS BLOCKS SURVEYED (sq. meters)
===============================================================-
SURVEY GROUP - WALL 50 WALL-REACTOR 4
68 34 WALL-HOLDUP 4
24 12 WALL-HOTCELL 4
16 8
SUBTOTAL 4
108 54 SURVEY GROUP - WKWY50 WALKWAYS 4
284 142 SUBTOTAL 4
284 142 SURVEY GROUP - CHEM 75 f
(
CHEMLAB 1
141 110 SUBTOTAL 1
141 110 SURVEY GROUP - EXRM75 ER-EWALL 1
12 9
ER-SWALL 1
24 18 ER-FLOOR 1
16 12 ER-WWALL 1
25 19 ER-OVE RHEAD 1
16 12 ER-PLUG 1
8 6
SUBTOTAL 1
101 76 SURVEY GROUP - HUHC75 HOLDUPRM 1
80 60 HOTCELL 1
31 23 SUBTOTAL 1
111 83 0
3-62
TABLE 6.2.2 (cont.)
')
\\~/
SUMMARY
OF SURVEY AREAS SURVEY AREAS BLOCK SIZE NUMBER OF BLOCKS BLOCKS SURVEYED (sq. meters)
======================================================
SURVEY GROUP - PDOL75 POOL-EWALL 1
15 11 POOL-SWALL 1
60 45 POOL-NWALL 1
39 29 POOL-FLOOR 1
23 17 POOL-WWALL 1
29 22 SUBTOTAL 1
166 124 SURVEY GROUP - SUMP 75 SUMP 1
68 52 SUBTOTAL 1
68 52 O
SURVEY GROUP - SLOT 75 (Note)
EXPOSURE ROOM SLOT 1
123 93 SUBTOTAL 1
123 93 SURVEY GROUP - BKGD BACKGROUND 1
30 30 SUBTOTAL 1
30 30
==============================================
TOTAL 1772 1016 Note The SLOT 75 surface areas are the new areas created when the concrete walls were demolished to open up the expos-ure room so that the exposure room gamma radiation level would conform with the background gamma radiation level measured in a 2 pi geometry.
O 3-63
TABLE 6.3.1 k.
BLOCK SURVEY DATA PROCESSING DATA PROCESSING OUTLINE The large volume of survey data - approximately 20,000 measure-ments - are assembled into a database for computer processing.
The data was processed by an IBM-XT computer, using the dBASE III for the database program, and the ABSTAT program to do the statistical analysis and for random number selection of the blocks to be survey.
A brief outline of the processing steps follows.
SURVEY AREA SELECTION 1)
Facility areas are divided into hazard categories and named (SN,TPA - Table 6.2.2 and Figure 6.2.1).
2)
Blocks are assigned a size (1,2,3 meters) and the fraction of blocks to be surveyed designated (75,50,30 percant).
is gridded and blocks assi 3)
Specific area
( Al, A2..., B1, B2... etc - e. g. Figu re 6. 3.1.gnned pair numbers SURVEY BLOCKS SELECTED 1)
An area file number (SN) is created in ABSTAT program.
p) 2)
Block pair numbers are entered in block pair field.
(#
3)
Random numbers are assignned to random number field (1-1000).
4)
Lowest random numbers, up to the percent defined by the category (30,50,75 percent), are selected for measurement.
5)
Selected block pairs are given to Health Physics team for measurement.
DATA MEASUREMENT RECORDED 1)
Reference data recorded on data sheet (see Table 6.3.2 for entries and symbols).
2)
Unbiassed and biassed measur:ments made and recorded on data sheet at each of the selected blocks (e.g. Table 6.3.3).
3)
Unbiassed data averaged for each block and entered on data sheet (Reml, Rem 2, Rem 3 - see Table 6.3.3).
4)
Copies of data sheets assembled for data processing.
Ov 3-64
TABLE 6.3.1 (cont.)
U BLOCK SURVEY DATA PROCESSING DATA PROCESSING OUTLINE DATA PROCESSING 1)
An area file number (SN) created in dBASE for data (see Table 6.3.2).
2)
Reference data entered into assignned fields (SN,TPA,DATE,-
MSN,SBN,BGSN,MCF,ACF,BGCF,MDB,ADB,BGDB).
3)
Data from each block transfered to the assignned fields (SBN,AGER,FBG, ALPHA, BETA, GAMMA, REM 1, REM 2, REM 3).
4)
Measurement data processed to give radiation levels, and results placed in assignned fields ( A, BG, G, BGR, G R, ALPHA BETA, GAMMA).
STATISTICAL PROCESSING 1)
An area file number (SN) created in ABSTAT program to accept radiation level data from dBASE files.
2)
Measurement data recorded in DATA PROCESSING 4) transfered to ABSTAT file (A,BG,G,BGR,GR, ALPHA, BETA, GAMMA).
3)
Statistical analysis of each set of radiation level data is made by ABSTAT program (Mean, standard deviation, variance, standard error of mean, coefficient of variance, minimum p()g value, maximum value, range of values, total, median, mode, skewness, and kurtosis - e.g. Table 6.3.5).
4)
Histogram of each set of measurements plotted (
e.g. Table 6.3.6).
5)
Statistical analysis of data transfered to dBASE file and a summary of the results are made into a report text file for use in preparation of the report by the word processing program.
WORD PROCESSING
SUMMARY
1)
Statistical data for all areas (SN,TPA) assembled into summary tables for presentation in report (see Table 6.4.1-4).
O 3-65
TABLE 6.3.2 BLOCK SURVEY DATA PROCESSING DATA SHEETS WITH SYMBOLS SURVEY BLOCK FORM DATE o...
P.g. No Of b
b c
........,c 1,... a,o,.c,-.,.. TPA s-..... e,
I
- "*8 C.:
C ort.c t O
8.e.ctor d
instruments Mfg /uod.
S.r.. No Due 0.ie E tt F.cior chorou o n
MSN MC F:
MOG a
u c,o n yn,n,
<<.im.. n, ASN ACF ADB cpu sS.,s"ioc.
av i,
%CF 860 B c,,
- e..-o.mm.
- d cil!.*o*,.
t? 'I.,.~,*.
o'.',i.
GIO~.
^ 'o a.
Aioa.
a'.".'#i S a". '
e'un n....
a.i,.i c,o,a', y - cowar a.i., coaism
,a,., gg e '.
a.-ara.
S
<j *n, >
c,ain, 3 copui (c ' * )
(cpu) l (o'u) io<<o'u>
AGER lA LA 3A I
O ggg n-te to 30
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CALCULATED DATA ENTERED IN REMARKS BLOCK REMI =
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(3A+3B+3C+3D+3E) / 5 CALCULATED DATA ENTERED IN COMPUTER PRINTOUT SHEET G
MCF x REMI
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A
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m 3-66
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l TABLE 6.3.3 V
BLOCK SURVEY DATA PROCESSING WALKWAY DATA SHEET EXAMPLE SURVEY BLOCK FORM 8/22/65 oe,e I
Page No-Of 50
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TABLE 6.3.5 V
BLOCK SURVEY DATA PROCESSING WALKWAY STATISTICAL ANAI.YSIS EXAMPLE EXAMPLE OF FACILITY WALKWAY ANALYSIS AB STr" 4. 0 9
SUMMARY
OF WALKWAY DATA FILES WKWY50A REV8 0 PAGE 5 COMMAND: DESC MISSING VALUE TREATMENT: VARWISE THERE ARE 8 VARIABLES AND 130 CASES IN THE DATA SET 130 CASES (100.0%) ARE VALID STD ERROR COEFF OF VARIABLE MEAN STD.DEV.
VARIANCE OF MEAN VARIATION 1 ALPHA 0.627692 0.771490 0.595196 0.0676641 122.909 2 BETA 1.92154 2.40281 5.77349 0.210740 125.046 3 GAMMA 60.2469 92.0803 8478.79 8.07598 152.838 4A 0.9'52:1 0.690420 0.476679 0.0605538 120.444 5 BG 1320.73 301.767 91063.0 26.4667 22.8484 p
6G 9.59462 1.63860 2.68501 0.143715 17.0783 5
7 BGR 1378.35 548.061 300371 48.0681 39.7622
\\
8 GR 9.42308 1.80381 3.25373 0.158205 19.1425 VARIABLE MINIMUM MAXIMUM RANGE TOTAL 1 ALPHA 0.00000 3.20000 3.20000 81.6000 2 BETA 0.00000 15.3000 15.3000 249.800 3 GAMMA 0.00000 502.600 502.600 7832.10 4A 0.00000 2.48400 2.48400 74.5200 5 BG 130.108 2129.04 1998.93 171695.25 6G 6.20000 12.2000 6.00000 1247.30 7 BGR 0.00000 2957.00 2957.00 179185.00 8 GR 0.00000 13.0000 13.0000 1225.00 VARIABLE MEDIAN MODE SKEWNESS RURTOSIS 1 ALPHA 0.000000 0.00000 1.27393 4.14422 2 BETA 1.20000 0.00000 2.34301 10.7933 3 GAMMA 0.00000 0.00000 1.85908 6.78983 4A 0.414000 0.00000 1.10844 3.61205 5 BG 1182.80 1182.80 0.563352 5.44670 6G 10.2000 10.0000
-0.777981 2.40152 7 BGR 1182.80 1182.80
-0.113378 4.25908 8 GR 10.0000 10.0000
-1.52955 7.44421 O
3-70
TABLE 6.3.6 p
BLOCK SURVEY DATA PROCESSING WALKWAY HISTOGRAM PLOT EXAMPLE ABSTAT 4.09
SUMMARY
OF WALKhAY DATA FILE: WhWY50A REVf 0 PAGE 6 COMMAND: HIST MISSING VALUE TREATMENT: VARWISE VARIABLE: 1 ALPHA AUT0/ 5 AT LEAST 0.00000 10 20 30 40 50 BUT NOT OVER: FREQ
+---------+---------+---------+---------+---------+
0.640000 63 48.5 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 1.28000 44 33.8 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 1.92000 13 10.0 IXXXXXXXXXX 2.56000 8
6.2 IXXXXXX 3.20000 2
1.5 IXX
+.........+.........+........
4.........,.........,
TOTAL 130 100.0 10 20 30 40 50 MISSING VALUE TFEATMENT: VARWISE l'V)
VARIABLE: 2 BETA AUT0/ 5 AT LEAST 0.00000 10 20 30 40 50 60 70 80 BUT NOT OVER: FREQ
+-----+------+-----+-----+-----+------+-----+-----+
3.06000 102 78.5 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 6.12000 20 15.4 IXXXXXXXXXX 9.18000 6
4.6 IXXX 12.2400 1
0.8 I
15.3000 1
0.8 I
+....,......+.....,....
4....
4......
4.....+.....+
TOTAL 130 100.0 10 20 30 40 50 60 70 80 MISSING VALUE TPEATMENT: VARWISE VARI ABLE: 3 GAMMA AUT0/ 5 AT LEAST 0.00000 10 20 30 40 50 60 70 80 BUT NOT OVER: FREQ
+-----+------+-----+-----+-----+------+-----+-----+
100.520 101 77.7 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 201.040 18 13.8 IXXXXXXXXX 301.560 8
6.2 IXXXX 402.080 2
1.5 IX 502.600 1
0.8 I
4......+.....+.....+....
4.....
4....
4.....,
TOTAL 130 100.0 10 20 30 40 50 60 70 80 0
3-71
pg TABLE 6.3.6 (Continued)
U BLOCK SURVEY DATA PROCESSING WALKWAY HISTOGRAF PLOT EXAMPLE ABSTAT 4.09
SUMMARY
OF WALKWAY DATA FILE: WKWY50A REV6 0 PAGE 7 COMMAND: HIST MISSING VALUE TFEATMENT: VAkhISE VARIABLE: 4A AUTO / 5 AT LEAST 0.00000 10 20 30 40 50 BUT NOT OVER: FREQ t
+---------+---------+---------+---------+---------+
0.496800 65 50.0 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 0.993600 46 35.4 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 1.49040 0
00.0 I
1.98720 13 10.0 IXXXXXXXXXX 2.48400 6
4.6 IXXXXX
+........
4........
4.........+.........+..........,
TOTAL 130 100.0 10 20 30 40 50 MISSING VALUE TREATMENT: VARWISE
/
(
)
VARI ABLE: 5 BG AUTO / 5 AT LEAST 130.108 10 20 30 40 50 60 70 BUT NOT OVER: FREQ t
+------+------+------+-------4------+------+------+
$29.894 1
0.8 IX 929.681 2
1.5 IX 1329.47 89 68.5 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 1729.25 24 18.5 IXXXXXXXXXXXXX 2129.04 14 10.8 IXXXXXXXX
+......+......+.....
4......
4......+.....
4......+
TOTAL 130 100.0 10 20 30 40 50 60 70 MISSING VALUE TREATMENT: VARWISE VARIABLE: 6G AUTO / 5 AT LEAST 6.20000 5
10 15 20 25 30 35 40 DUT NOT OVER: FREQ
+-----+------+--
i-----+-----+------+-----+-----+
7.40000 23 17.7 IXXXXXXXXXXXXXXXXXXXXXX 8.60000 11 8.5 IXXXXXXXXXXX 9.80000 18 13.8 IXXXXXXXXXXXXXXXXX 11.0000 52 40.0 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 12.2000 26 20.0 IXXXXXXXXXXXXXXXXXXXXXXXXX
+.....+......+.....+.....,....
4.....
4....
4.....,
TOTAL 130 100.0 5
10 15 20 25 30 35 40 i(
3-72
m TABLE 6.3.6 (Continued) i BLOCK SURVEY DATA PROCESSING WALKWAY HISTOGPAM PLOT EXAMPLE ABSTAT 4.09
SUMMARY
OF WALKWAY DATA FILE: WKWYSOA REVO O PAGE 8 COMMAND: HIST MISSING VALUE TREATMENT: VARWISE VARIABLE: 7 BGR AUTO / 5 AT LEAST 0.00000 10 20 30 40 50 60 70 BUT NOT OVER: FREQ
+------+------+- ----+-------+------+------+------+
591.400 8
6.2 IXXXX 1182.80 79 60.8 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 1774.20 2
1.5 IX 2365.60 26 20.0 IXXXXXXXXXXXXXX 2957.00 15 11.5 IXXXXXXXX
+.....
4......,......,.......+.....
4.....
4.....
4 TOTAL 130 100.0 10 20 30 40 50 60 70 MISSING VALUE TREATMENT: VARWISE
[\\
VARI ABLE: 8 GR AUTO / 5 AT LEAST 0.00000 10 20 30 40 50 60 BUT NOT OVER: FREQ
+-------+--------+-------+-------+--------+-------+
2.60000 1
0.8 IX 5.20000 0
00.0 I 7.80000 26 20.0 IXXXXXXXXXXXXXXXXX 10.4000 74 56.9 IXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX XX 13.0000 29 22.3 IXXXXXXXXXXXXXXXXXXX
+.......,.......
4.......+......
4........+.......,
TOTAL 130 100.0 10 20 30 40 50 60 o
3-73
TABLE 6 121 4
x_/
PINAL SURVEY OF MORTHROP REACTOR FACILITY HIGH CONTAMINATION POTENTIAL AREA DATA SURVEY UNITS MEAN STANDARD STANDARD MAXIMUM MEDIAN AREA VALUE DEVIATION ERROR OF READING VALUE MEAN TYPE RADIATION - REMOVABLE ALPHAS CHEM 75A dpm/100cm2 0.20 0.44 0.04 2.00 0.00 EXRM75A dpm/100cm2 0.39 0.48 0.05 1.60 0.20 HUHC75A dpm/100cm2 0.13 0.29 0.03 1.60 0.00 POOL 75A dpm/100cm2 0.99 1.06 0.10 6.50 0.80 SUMP 75A dpm/100cm3 0.11 0.42 0.06 2.40 0.00 dpm/100cm 0.26 0.34 0.04 1.40 0.20 SLOT 75A
=
TYPE RADIATION - REMOVABLE BETAS CHEM 75A dpm/100cm2 1.40 1.57 0.14 7.50 1.20 EXRM75A dpm/100cm2 2.66 3.81 0.44 18.40 1.20 HUHC75A dpm/100cm2 1.85 2.75 0.30 19.10 1.20
{()% POOL 75A dpm/100cm2 13.44 24.67 2.40 189.80 6.30 SUMP 75A dpm/100cm3 5 17 14.99 2,08 87.70 1.80 SLOT 75A dpm/100cm 0.79 0.73 0.08 3.50 0.60 TYPE RADIATION - REMOVABLE GAMMAS CHEM 75A dpm/100cm2 32.16 62.18 5.68 294.50 0.00 EXRM75A dpm/100cm2 33.39 63.48 7.28 276.60 1.20 HUHC75A dpm/100cm2 172.26 115.38 12.66 653.50 175.50 POOL 75A dpm/100cm2 70.96 137.71 13.38 732.40 0.00 SUMP 75A dpm/100cm3 27.75 74.53 10.33 402.20 0.00 dpm/100cm 109.18 92.73 9.67 367.70 91.85 SLOT 75A TYPE RADIATION - BIASSED BETA GAMMA (BGR)
CHEM 75A dpm/100cm2 3538.54 306.52 27.90 4139.80 3548.40 EXRM75A dpm/100cm2 3119.24 1287.84 147.72 6448.00 3200.00 HUHC75A dpm/100cm2 3345.56 1509.71 165.71 5914.00 3200.00 POOL 75A dpm/100cm2 2637.99 773.91 75.17 4836.00 2418.00 SUMP 75A dpm/100cm3 3653.13 362.48 50.27 4139.80 3548.40 dpm/100cm 1566.76 1141.46 119.00 3204.00 1039.50 SLOT 75A 3-74
<-w TABLE 6.4.1 (cont.)
v)
(
FINAL SURVEY OF NORTHROP REACTOR FACILITY HIGH CONTAMINATION POTENTIAL AREA DATA SURVEY UNITS MEAN STANDARD STANDARD MAXIMUM MEDIAN AREA VALUE DEVIATION ERROR OF READING VALUE MEAN TYPE RADIATION - UNBIASSED GAMMA AT 1 METER (GR)
CHEM 75A uR/hr 8.43 1.66 0.15 10.00 9.00 EXRM75A uR/hr 18.43 6.15 0.70 30.00 20.00 HUHC75A uR/hr 13.84 1.53 0.17 16.00 14.00 POOL 75A uR/hr 14.53 2.53 0.25 21.00 15.00 SUMP 75A uR/hr 13.60 0.72 0.10 15.00 14.00 SLOT 75A uR/hr 12.27 1.89 0.20 16.00 12.00 TYPE RADIATION - UNBIASSED GAMMA AT 1 CM (G)
CHEM 75A uR/hr 8.44 1.85 0.17 10.00 8.80 EXRM75A uR/hr 18.65 6.80 0.78 32.60 20.30
^'
/
HUHC75A uR/hr 13.49 2.94 0.32 16.00 14.20
( >)
POOL 75A uR/hr 14.36 2.48 0.24 26.80 14.40 SUMP 75A uR/hr 13.68 0.53 0.07 14.60 13.80 SLOT 75A uR/hr 13.44 2.05 0.21 10.50 13.40 TYPE RADIATION - UNBIASSED BETA GAMMA (BG)
CHEM 75A dpm/100cm2 3007.76 667.35 60.92 3666.68 3193.56 EXRM75A dpm/100cm2 2744.92 3069.17 352.06 2740.40 2418.00 HUHC75A dpm/100cm2 2729.16 1408.99 154.66 5559.16 2240.00 POOL 75A dpm/100cm2 2065.96 572.22 55.58 4513.60 1920.00 3250.43 227.68 31.57 3666.68 3252.70 dpm/100cmg 1409.26 SUMP 75A dpm/100cm 987.67 102.97 2867.58 1037.61 SLOT 75A TYPE RADIATION - UNBIASSED ALPHAS (A)
CHEM 75A dpm/100cm2 5.32 9.14 0.83 44.71 0.00 EXRM75A dpm/100cm2 1.41 3.03 0.25 17.39 0.00 HUHC75A dpm/100cm2 0.28 1.09 0.12 8.28 0.00 POOL 75A dpm/100cm2 0.23 0.98 0.95 8.28 0.00 SUMP 75A dpm/100cm3 0.96 1.76 0.24 4.14 0.00 SLOT 75A dpm/100cm 14.83 14.13 1.47 59.70 9.95
( )
Note : EXRM75A surfaces eliminated and replaced by SLOT 75A data.
3-75
r-TABLE 6.4.2
(]/
FINAL SURVEY OF NORTHROP REACTOR FACILITY MEDIUM CONTAMINATION POTENTIAL AREA DATA SURVEY UNITS MEAN STANDARD STANDARD MAXIMUM MEDIAN AREA VALUE DEVIATION ERROR OF READING VALUE MEAN TYPE RADIATION - REMOVABLE ALPHA DECK 50A dpm/100cm2 0.43 0.45 0.09 1.60 0.40 ROOM 50A dpm/100cm2 0.60 1.03 0.10 9.00 0.80 WALL 50A dpm/100cm2 0.46 0.59 0.08 1.60 0.00 WKWYSOA dpm/100cm2 0.63 0.71 0.07 3.20 0.80 TYPE RADIATION - REMOVABLE BETA DECK 50A dpm/100cm2 1.69 1.50 0.31 5.40 1.50 ROOM 50A dpm/100cm2 1.88 2.79 0.26 24.00 1.20 W ALL50 A dpm/100cm2 1.32 1.52 0.21 5.70 0.60 WKWYSOA dpm/100cm2 1.92 2.40 0.21 15.30 1.20
)
N/
TYPE RADIATION - REMOVABLE GAMMA DECK 50A dpm/100cm2 12.91 47.00 9.80 218.00 0.00 ROOM 50A dpm/100cm2 91.85 110.80 10.42 489.00 50.20 WALL 50A dpm/100cm2 116.00 117.00 15.96 480.00 82.40 WKWYSSt.
dpm/100cm2 60.25 92.08 8.08 503.00 0.00 TYPE RADIATION - BIASSED BETA GAMMA (BGR)
DECK 50A dpm/100cm2 3051.00 1086.00 226.00 4731.00 3200.00 ROOM 50A dpm/100cm2 2440.00 923.00 86.82 4731.00 2418.00 WALL 50A dpm/100cm2 27.80 738.00 100.50 4731.00 2418.00 WKWY50A dpm/100cm2 1378.00 548.00 48.06 2957.00 1183.00 Ov 3-76
TABLE 6. 4. 2 (cont.)
(\\-
FINAL SURVEY OF NORTHROP RE&CTOR FACILITY MEDIUM CONTAMINATION POTENTIAL AREA DATA SURVEY UNITS MEAN STANDARD STANDARD MAXIMUM MEDIAN AREA VALUE DEVIATION ERROR OF READING VALUE MEAN TYPE RADIATION - UNBIASSED GAMMA AT 1 METER (GR)
DECK 50A uR/hr 9.70 2.57 0.54 14.00 9.00 P.OOM50A uR/hr 9.75 1.51 0.14 14.00 10.00 WALL 50A uR/hr 10.74 1.62 0.22 14.00 11.00 WKWY50A uR/hr 9.42 1.80 0.16 13.00 10.00 TYPE RADIATION - UNBIASSED GAMMA AT 1 CM (G)
DECK 50A uR/hr 9.10 2.09 0.44 13.00 9.00 ROOM 50A uR/hr 9.71 1.41 0.13 14.20 9.40 WALL 50 A uR/hr 10.80 2.04 0.28 15.80 11.10 WKWY50A uR/hr 9.71 2.15 0.19 12.20 10.20 (x
TYPE RADIATION - UNBIASSED BETA GAMMAS (BG) i
)
'~'
DECK 50A dpm/100cm2 2892.00 988.00 206.00 4613.00 3075.00 ROOM 50A dpm/100cm2 2052.00 744.00 70.00 4494.00 1773.00 WALL 50 A dpm/100cm2 2212.00 683.00 92.94 3903.00 1972.00 WKWY50A dpm/100cm2 1320.00 302.00 26.47 2129.00 1183.00
===-
TYPE RADIATION - UNBIASSED ALPHAS (A)
DECK 50A dpm/100cm2 0.00 0.00 0.00 0.00 0.00 ROOM 50A dpm/100cm2 0.42 d.92 0.09 8.28 0.00 WALL 50 A dpm/100cm2 0.00 0.00 0.00 0.00 0.00 WKWYSOA dpm/100cm2 0.57 0.69 0.06 2.48 0.41 3-77
TABLE 6.4.3 FINAL SURVEY OF NORTHROP REACTOR FACILITY LOW CONTAMINATION POTENTIAL AREA DATA SURVEY UNITS MEAN STANDARD STANDARD MAXIMUM MEDIAN AREA VALUE DEVIATION ERROR OF READING VALUE MEAN TYPE RADIATION - REMOVABLE ALPHA OVHD30A dpm/100cm2 0.28 0.45 0.05 1.60 0.00 PLMZ30A dpm/100cm2 0.48 0.87 0.16 3.20 0.00 TYPE RADIATION - REMOVABLE BETA OVHD30A dpm/100cm2 0.95 1.41 0.17 8.20 0.30 PLMZ30A dpm/100cm2 3.25 2.79 0.50 10.10 2.50 l
TYPE RADIATION - REMOVABLE GAMMA OVHD30A dpm/100cm2 31.38 65.50 7.67 301.10 0.00
()e PLMZ30A dpm/100cm2 60.52 84.09 15.10 272.40 0.00 y_,
TYPE RADIATION - BIASSED BETA GAMMA (BGR)
DVHD30A dpm/100cm2 2900.29 732.08 85.68 3548.40 2957.00 PLMZ30A dpm/100cm2 2102.90 403.06 72.39 2418.00 1773.20 O
3-78
N TABLE 6.4.3 (cont.)
e l FINAL SURVEY OF NORTHROP REACTOR FACILITY LOW CONTAMINATION POTENTIAL AREA DATA SURVEY UNITS MEAN STANDARD STANDARD MAXIMUM MEDIAN AREA VALUE DEVIATION ERROR OF READING VALUE MEAN TYPE RADIATION - UNBIASSED GAMMA AT 1 METER (GR)
OVHD30A uR/hr 5.59 0.64 0.08 7.00 6.00 PLMZ30A uR/hr 8.81 1.54 0.28 11.00 9.00 TYPE RADIATION - UNBIASSED GAMMA AT 1 CM (G)
OVHD30A uR/hr 5.69 0.49 0.06 6.80 5.80 PLMZ30A uR/hr 8.54 1.44 0.26 10.80 8.80 TYPE RADIATION - UNBIASSED BETA GAMMAS (BG)
[ )'
OVHD30A dpm/100cm2 2488.74 589.24 68.97 3430.12 2602.16 PLMZ30A dpm/100cm2 1778.94 269.56 48.41 2418.00 1773.20
= _ _.
TYPE RADIATION - UNBIASSED ALPHAS (A)
OVHD30A dpm/100cm2 0.24 0.95 0.11 4.14 0.00 PLMZ30A dpm/100cm2 0.21 0.52 0.09 1.65 0.00 0
3-79
3 TABLE 6.4.4 PINAL SURVEY OF NORTHROP REACTOR FACILITY ALL CATEGORIES, BACKGROUND AND PERMISSIBLE LEVELS SURVEY ALPHA BETA GAMMA E
EB G
GB AREA l----- removable -----l------- < 1 cm ------------l-1 m -l l----------------dpm/100cm2-----------------------l-uR/hr-l 75 PERCENT BLOCK SURVEY AREA CHEM 75A 0.20 1.40 32 5.32 3010 3540 8.4 8.4 (EXRM75A) (0.39)
(2.66) (33)
(1.41) (2740)
(3120)
(18. 7 ) (18. 4 )
HUHC75A 0.13 1.65 172 0.28 2730 3350 13.5 13.8 POOL 75A 0.99 13.44 71 0.23 2070 2640 14.4 14.5 SUMP 75A 0.11 5.17 28 0.96 3250 3650 13.7 13.6 SLOT 75A 0.26 0.68 109 14.83 1409 1567 13.4 12.3 AVERAGE 0.33 4.50 82 4.32 2494 2949 12.7 12.5 BACKGROUD 0.26 0.35 85 0.57 1720 1720 11.6 11.6 DIFFERENCE 0.10 4.15
-3 3.75 774 1229~
1.1 0.9 ALLOWED 20.00 1000.0 1000 100.0 5000 15000 5.0
>5.0
======================================================
/~')
Note: EXRM75A surfaces eliminated and replaced by SLOT 75A data.
V 50 PERCENT BLOCK SURVEY AREAS DECK 50A 0.43 1.69 13 0
2890 3050 9.1 9.7 ROOM 50A 0.60 1.88 92 0.42 2050 2440 9.7 9.8 WALL 50A 0.46 1.32 116 0
2210 2710 10.8 10.7 WKWYSOA 0.50 1.63 94 0.14 2380 2730 9.5 10.1 AVERAGE 0.51 1.63 79 0.14 2380 2730 9.8
10.1 BACKGROUND
0.26 0.35 85 0.57 1720 1720 11.6 11.0 DIFFERENCE 0.25 1.28
-6
-0.43 660 1010
-1.8
-1.5 ALLOWED 20.0 1000.0 1000 100.0 5000 15000 5.0
>5.0
======================================================
30 PERCENT BLOCK SURVEY AREAS OVHD30A 0.28 0.95 31 0.24 2490 2900 5.7 5.6 PLMZ30A 0.48 3.25 61 0.22 1780 2100 8.5 8.8 AVERAGE 0.38 2.10 46 0.23 2140 2500 7.1
7.2 BACKGROUND
0.26 0.35 85 0.57 1720 1720 11.6 11.6 DIFFEREEEE 0.12 1.75
-39 0.34 420 780
-4.5
-4.4
~
7-s ALLOWED 20.0 1000.0 1000 100.0 5000 15000 5.0
>5.0
( j 3-80
FIGURES FIG. NO SUBJECT PAGE 2.1.1 DECOMMISSIONING ORGANIZATION CHAPT............
4-2 2.2.1 DECOMMISSIONING SCHEDULE /MANLOADING CHART.....
4-3 2.2.2 SITE MODIFICATIONS PRIOR TO DISMANTLING 4-4 3.0.1 LOS ANGELES AND HAWTHORNE MAP.................
4-5 3.0.2 NORTHROP HAWTHORNE COMPLEX MAP................
4-6 f
3.0.3 NCRTHROP REACTOR FACILITY FLOOR PLAN..........
4-7 3.0.4 REACTOP-ARTIST VIEW..........................
4-8 3.0.5 REACTOR-VERTICAL SECTION.....................
4 -9 3.0.6 REACTOR-PLAN VIEW............................
4-10 3.1.1 REACTOR-CORE CONFIGURATION...................
4-11 3.1.2 REACTOR-FUEL ELEMENT ASSEMBLY................
4-12 4.2.1 FISSION FRAGMENT MASS DISTRIBUTION............
4-13 4.3.1 ENVIRONMENTAL RADIOLOGICAL SURVEY LOCATIONS...
4-14 4.4.1 THERMAL NEUTRON FLUENCE IN POOL...............
4-15 4.'4.2 FAST NEUTRON FLUENCE IN POOL..................
4-16 4. '4.' 3 REACTOR ACTIVITY FACTOR.......................
4-17 g
4.4.4 SPECIFIC ACTIVITY VERSUS DEPTH, EXPOSURE BOOM CONCRETE...............................
4-18 i
5.1.1 NORTHROP TRANSFER CASK........................
4-19 I
5.1.2 BMI-1 SHIPPING CASK...........................
4-20 5.4.1 EXPOSURE ROOM DEMOLITION......................
4-21 5.6.1 RADIOCHEMISTRY LABORATORY DECONTAMINATION.....
4-22 5.7.1 VENTILATION DUCTS SURVEY LOCATIONS............
4-23 5.7.2 PLUMBING DECONTAMINATION SURVEY LOCATIONS.....
4-24
~
5.9.1 ERODED CONCRETE ACTIVITY VERSUS BURIAL TIME...
4-25 5.9.2 GROUNDWATER SPECIFIC ACTIVITY VERSUS BURIAL TIME........................................
4-26 6.
1.1 BACKGROUND
SURVEY AREA LOCATIONS..............
4-27 6.2.1 FINAL SURVEY AREA CLASSIFICATION..............
4-28 6.3.1 FINAL RADIOLOGICAL SURVEY - EXAMPLES OF BLOCK SURVEY................................
4-29
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_ __ _- _ = aJ i---Q v
r?y t % ~ m. E i
- 9 8
l; i.;^
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'j 47 :
i l==;; ll. $-)
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)
ew of the reactor facility showing the fenced-in area to the th of the building for temporary storage of decontaminated
+erial. The Flash X-ray area inside the building was removed to
=e additional space available for the dismantling operations.
.o, the Flash X-ray instrument room was converted to an of: tce ce for the subcontractor.
OV 4-4
FIGURE 3.0.1 LOS ANGELES AND HAWTHCRNE MAP O
1 MMMU
- gggggg, h
.. =.....
\\
/
NORTHROP - VENTURA w.
i C
GLEND P ASADE N A s.
-,f
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/
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/
's, f
,i 3
CORPORATE.
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e r.= a.oi%o
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ERMOSA BE ACH j
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' ^ % j,/
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s-The Northrop TRIGA Reactor Facility is located in the Northrop O
Hawthorne Complex on Northrop Avenue, between Prairie Boulevard and Crenshaw.
4-5
1 FIGURE 3.0.2 NORTHROP HAWTHORNE COMPLEX MAP
{
tv i
y s. w ~,,
-I [
[
m ".... s, >
l _ i __JJ cc 09 lI 5
e >.
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1
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c y;
q t L, b
20: -
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The Northrop TRIGA Facility is located near Gate 15 on Northrop Avenue.
Iv 4-6
FIGURE 3.0.3 NORTHROP REACTOR FACILITY FLOOR PLAN g e s * * **6
- e s e e aG= s e * * *
- G D o e s *
- e t e s e s s e s e * *
- e * * * * * * * * * * ** * * * * *
- j i
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il
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c.7-wg,-
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4-7
FIGURE 3.0.4 REACTOR - ARTIST VIEW g=s N TRANSIENT ROD DRIVE N
.. s CONTROL ROD DRIVE
'~
i[N N
REACTOR BRIDGE
- x
=
- kg
%{pp4 D..sg BEAM PORT 7'
j,,
,N h
/,
O
/
N EXPOSURE i
/
.\\,!
A ROOM PNEUMATIC
,J q
1 SYSTEM j 3 18 g
RE ACTOR CORE /
i i
REMOVABLE PLUG The reactor is shown at the exposure room operating position.
O 1
i 4-8
FIGURE 3.0.5 REACTOR - VEPTICAL SECTION REACTOR BRIDGE l
l l ems I
l
}
REMOV AB LE REACTOR POOL %
PLUG N..,.i 'I g
EXPOSURE
/-
[ ROOM
_ :n,
au n f BEAM PORT -
,.ounui
/
RE ACTOR CORE i
l l
The reactor is shown at the exposure room operating position.
!O l
4-9
1 l
FIGURE 3.0.6 REACTOP - PLAN VIEW (w
3 a 3 FT 3 = 7 FT STEEL 00CR p;uc
, e,,, p p Two 8 INCH DIAu[T[q DCCR Q
CONDU!TS i
i 5 FT 8 !N. a
/
e
6 FT 4 IN.
'I f'/
HIGd
,/ o l' l
/
/
/
/'
/_ '
o Tm0 2 NCH D:AVETEE 7a7=10 D
- Je,. _ - _
c
-*7 c --
---o CONDJITS gy
~-[h.
,,,-j FCUR 3 INCH O!wE'. E R
[
E X P F COM U b :: /
e, p,
q..j CONDUIT 5 m
cr -
c:r.,
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J.-,, - - - - - - -
@ e.8 i:
l f L g FT H:0H
\\_2 ACCESS PORTS FOUR 6 tNCH 8 8 8 8 I
I0" I
DlwETER PuuG5
/
l O
21 NCH DlwETER CENTER PLUG C::
q._*,*,
241NCH O!AMETER CJTER P LG 12004 d
O L
/
.3 _,
r Too 8.tsCH D!wETE R INSULA T ED DUCTS REACTOR O
CORE OPE R ATINO h
- 6FT POSIT Oss 10 F T o
\\
PM uAT.C ST5TEv
\\
F l
/
THREE FE Au DCRTS 6 tNCH O Auf TER isNES PLUG
- 8. NC-C.wE TER OUTE 4 PLUO O
4-10
FIGURE 3.1.1 REACTOR - CORE CONFIGURATION
,\\v l
LArosure Room Mood Limng M
A
/,
~
e' 1
y L
si II
[
l'ul se n 1
-j a
Cure Shroud
[
(Outso!c 1(at!aud =
II/ 8 is 5 in)
. S tiets
-l
/ /
/
~ 1 in.
Iteactor Tark (Da t
- nle Itu f au > =
11 in.;
,n.
(;f CONTItO L lt0;b I L E l. E L E M E N 'I.s
. N m,,_,,c m 0
IXi51MI:lltY lif el.ES The core is shown at the window of the exposure room.
O 4-11
FIGURE 3.1.2 REACTOR - FUEL ELEMENT ASSEMBLY v
Stainless Steel Top End-Fixture Graphite Stainless
/ pacer i
S b @#
c r Burnable Poison Stainless rSteel
/
Tube s
ll Zirconium i'
ilydride -
ll 8WTi l8 Uranium g ji il 28.31 IN Claddmg il Thickness 0.25 IN.
0 0.02 IN.
Zirconium \\llll 15 IN.
p Rod I
U il 11 1.4 3 IN.----
l,' -- l 11 1
1' l s
~ 1.47 IN.
i a
1 Oy s.,,,,a m e ss Poison Graphite a
g tt O
l g
3.44 IN.
Stainless Steel
/g
-l Bottom End-Fixture d I
v 4-12
FIGURE 4.2.1 "m
FISSION FRAGMENT MASS DISTRIBUTION Sr00 Cs137 k
f 10 1
10'I s-w
)h O
8,g-2 o
a 52 u.
39-3 10 h 30-5 I
I I
I I
I I
I I
I I
70 80 90 100 110 120 130 140 150 16 0 170 M ASS NUMBE R Strontium 90 is a principal radioisotope created by the light O
the heavy fraction fragments.
fission fragments.
Cesium-137 is a principal radioisotope for 4-13
FIGURE 4.3.1 ENVIPONMENTAL RADIOLOGICAI. SURVEY LOCATIONS p
\\s ji'
& sep * % Q. {f_ g p u _.-'.+j;"fr]ip[7p.,g,i S
- I
.A (k
- 4... _-4"* -, m.-p"$g,J' q.a...'. (. a,k y.y
- -. t + *
- d.- !
g
- i.,
g;;
g
- m..-.,., e e n
fi g
e 3.' u A.[
g) y d k*
Q.bncedn,[ f m&=
g@f * $ '&?h
., kmh-A'.?. I f'
y.
%l
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.au qg 7
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x v
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m3 A i
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+
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CIFIC
\\
f O C E A.\\.
Environmental samples were taken at the locations indicated in Table 4.3.5 during the operating life of the reactor.
The pd radioactivity observed at these locations correlated with the climatic conditions and showed no relationship to the reactor operation.
4-14 i
FIGURE 4.4.1 THERMAL NEUTRON FLUENCE IN POOL 0
10 20 30 40 50 60 I
i l
l l
100 - r
- 100 h
e REFERENCE BIB 4.1.1 0
- REFERENCE BIB 4.1.2 FCORE = 2.2 X 10l9 'cm2
=
\\
n.
10 0
10 0
\\
g
\\
e
\\
9 i
\\
i a
\\
5
\\
g
\\
0.1
\\
01
~
u.
z
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e%
zw e
f.
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\\'%
e 0 001 0.001 i
\\Sp
\\
l i
i i
\\.
1 0
10 20 30 40 50 60 DISTANCE FROM REACTOR CORE (INCHES) l l
O 4-15
.=
FIGURE 4.4.2 FAST NEUTRON FLUENCE IN POOL O
i 100 O BIB 4.4.1 e BIB 4.4.2 19 FCORE = 5 X 10 n/cm2 0
I N
10 B
\\
\\
\\
M
\\
j _
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Oo H
\\g
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s Ng%,
d
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a
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e
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0.001
\\
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0 O
10 20 30 40 50 DISTANCE F RC41 CORE (INCHES) 4-16
FIGURE 4.4.3 REACTOR ACTIVITY FACTOR C'\\
U 1,0 100 to 80 E u' ' '
Co p,ss E u' '
- eo Sr C s' 8 #
,5
.1 -
8 G:
C 5
[
\\
/
.01-
.co:
.co,
ISOTOPE LIFETIME (YRS.)
The Activity Factor, HA, is the product of the fraction of the radioisotopes surviving at dismantling time and the fractional approach ~to equilibrium production of the specific radioisotope over the operating period.
The falloff for the short halflife radioisotope is due to the exponential decay during the operating O
- period, while the falloff for the long halflife radioisotopes indicate there was not sufficient time to reach a steady state condition during the operating period.
4-17
FIGURE 4.4.4 SPECIFIC ACTIVITY VS DEPTH c
EXPOSURE ROOM CONCRETE O
10 20 30 40 50 60 l
l I
i l
l 100 SPECIFIC RADIOACTIVITY VERSUS DEPTH IN CONCRETE
_C CRETE
=
CONCRETE REMAINING ED 10 4
- 10
- c, t
e Q
O O
e E
Q Os E
O 4
o 0.1 0.1 o
//
NATURAL BACKGROUND ACTIVITY 0.01 0 01 t
x\\
h 0
10 20 30 40 50 60 70 DEPTH IN CONCRETE (INCHES)
NEUTRON NATURAL REMAINING INDUCED BACKGROUND NEUTRON ACTIVITY ACTIVITY ACTIVITY WA M
The radioactivity created by neutron capture decreases expon-entially with depth and becomes less than the natural primordial radioactivity beyond 20-inch depth.
The concrete was removed to a depth of 24 inches and shipped to the Hanford Site for burial.
The neutron produced radioactivity remaining at greater depth is small compared to the primordial activity, and will decay in a relatively short time to a negligible level.
4-18
FIGURE 5.1.1 NORTHEOP TRANSFER CASK i
SCHEM ATIC CUTAWAY TOP VIE W
-~
~ -
/
' I '81N EYE BOLTS h
l
)
MATERIAL Z[g C
f
,, y --
TYPE C-1018 STEEL A
'v A
by, 3-9 16 IN.
TENSILE STRENGTH i
s O?
'/.,
70.000 LB IN i
h
' [/ -
iin i
7-7 81N. 00-K CAPACITY LEAD l
j
/
2 re iN -
4 400 LB E ACH l
-)
EYE BOLTS s
1 m
N
'g -
s I
s
^
O i-7;s N 1'2 lA[}
N
+
- .~J l
~
l D,'A l
8 i
l I 24IN i
1
{ LEAD L
SPECIFIC ATIOr.S f;
VATE RIAL
'/
i
/
FUEL ELEMENT TOP SHELL TYPE C-1018 STEEL j ; j {
TR ANSFER TUBES FUEL ELEMENT TR ANSFER TUBES 1-14 IN IPS TYPE 304 STAINLESS STEEL N
f 1
I f '
ViEIGH T 3-E L EYE NT 1-ELEMENT
~*~
jl lt i
f)
LOADED 423 LB 44{ L E i i LEAC l > 1 !i o!
e i,
'6 k
W
', 1 i 1 IN]
P
'C'.
_A
- v. ELD
\\K M
s
't 4 [q IF R ADIA'f 0N LEVELS 5HOULO BE Eh CESS:VE f
w TH THE THREE-ELEVENT CASK A ONE-ELEW ' '
j L16 IN -
u 3 4 IN f
CASA H AS BE EN CONSTRICTED 17 iS IDEr ? C AL i
- 5 8 IN M" TO THE CASA SHDAN EXCEPT IT PAS ONE FLIL s
ELEMENT TLBE AND THREE INCHLS Or LE AD 4
\\
CROSS SEC*lON VIE AS A 4 The transfer cask was fabricated by Northrop and used to transfer fuel elements from the pool to the shipping cask. The transfer cask minimized the exposure of the personnel during the transfer operation.
O 4-19
FIGURE 5.1.2 BMI-3 SHIPPING CASK O
\\j
'li F
?
' b'
..t-,
x
-s.
..x 9-y, #
..,.w
- C.~ U."**,0ch'*~*.*"
/
~~
~ ~ ~
)y f
f urTma v0xt c /., f n
n
'==m sa
- Lt0 6=,.
sa.
=
so-,ie-
=
U -
p3/g-
=
l
( -j-c'd % l_,3 STUDS 1* - 0 THRE AD g
L\\b l.1.
{
h~
t 8* LE AD + 0.78* STEEL
=
ff' $ jg
! 4 T-6-17A* - LONG BOLTS
, [(
1-114' - 7 THRE AD
[
.l gg
[
- S.
5-l.
h u
+
f O',
1 p,, d' Q/
g
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8' s
e n/ e
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u
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(-
O
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- O g
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j --
\\,, i 1*
- 0 THREAD J
7 SA$$ PLATE
- e. a As a i-i
//
s
\\
l
,/
\\
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C ASn PLATE (ATT ACHED)
DRAM v4Lyr BOu 38* 8 1 PALLET 9t*a 72*a 6*
t-ti2* - LONG SOLTS 7
B ATTE LL E COLUVBL 5 L APOR ATOhtE 5 1*=8 THREAD Casa evi i usa ges? a is nacecac m e wa ve o.at sw.et.% c w a Ae.
The shipping cask is required for the shipping of the fuel. It was reserved as soon as the decommissioning operation schedule was known. The cask, because of the extreme weight, cannot be easily moved inside the building.
O 4-20
FIGURE 5.4.1 EXPOSURE ROOM DEMOLITION h
e......g.......e.......e O......q..........q...........g.......q
\\
Casa +es Less FTagis Q s
I.
i.
t i
g
.theseo nata 1
.,0.......m
^
b
..x.
(
.-----,-~-~,m,we,
= = ~
a t.
= =
l
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y i
m p
l a
4"v"v u a a X A Vy"y NN)..,,.: p"..
c-*= ue l9j
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. 44.
- 7
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1 4
g G.
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==7 ~l
(
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s a4.'
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c:.. a Omcs c o....,
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_ _ _, _ _ _,. ~
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r 4...y.
l p
wa e*. -
n..... r Y'-.., ;i s -' - - - - "p i
j s.,...s
.D
- 6 0 0. D. a.=,,,,,,,,_
&*44W410..
f.
- tt' R;SSSS.1 WALLS REMOVED F<
tl CONCRETE REMOVED 0
\\
4-21
FIGURE 5.6.1 RADIOCHEMISTRY LABORATORY DECONTAMINATION f) v FLOOR AREA DECONNED C
A l
B O
O "ooo O
HOOD O
i The storage wells, A and B, and the cabinet, C, were contaminated (see Table 6.5.2.1 for data).
The area was decontaminated and the radioactive material shipped to the Hanford Site for burial.
O 4-22
FIGURE 5.7.3 VENTILATION DUCTS SURVEY LOCATIONS O
3&4 7&8 9 & 10 1h & 12 W
E 08 S
5&6 13 & 14 1&2 15 & 16 r
3 1
I I
I I
SURVEY OF CEILING EXHAUST DUCTS i
--4 E \\ST WALL 6
5 3
2 0
8 12 11 4
9 7
to 1
3 4
WEST WALL 15 14 18 17 21 20 24 23 33 16 39 22 SURVEY OF WALL SUPPLY AIR DUCTS The personnel transport bucket was required to reach the venti-lation ducts in the high overhead of the building. Table 5.7.3 shows there is no contamination in the ducts.
O 4-23
FIGURE 5.7.2 PLUMBING DECONTAMINATION SURVEY LOCATIONS
(-
ge **e sed e**** =D=e se***4 G e * * *
- e 4. = * * * ** * *
- e e....... ees... s e q c
== everse
\\
f 9
t g
..3.
t,
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The radiological survey data listed in Table 5.7.2 show some con-tamination in the drainline from the Padiochemistry Laboratory to the sump tank. The portion of the drain showing contamination was removed and shipped to the Hanford site.
p s
4-24
FIGURE 5.9.1 EPODED CONCPETE ACTIVITY VERSUS BURIAL TIME O
100,000 10,000 c#
0 *D 0
B cgo E 1,000 Bo G
sgc6
=
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100 5s O
o 10 o
O g
i _ NEUTRON ACTIVATION ERODED ACTIVITY cw y
0.1 O.01 0.001 -
I 10 100 TIME AFTER BURIAL (YEARS)
The radioactivity in the eroded concrete created by neutron A
capture is always small in comparison to the natural prirrordial h
radioactivity, and is also small compared to the natural radio-activity already present in the soil.
4-25
FIGURE 5.9.2
_ GROUNDWATER SPECIFIC ACTIVITY VERSUS BUPIAL TIME m
100,000 EXISTING PRIMORDIAL RADIOACTIVITY 10.000 TAP WATER (SEE TABLE 4.3.6) 1.000 8
100 b
W t-10 ERODED CONCRETE PRIMORDIAL RADIOACTIVITY 2
U<
9 b
1
(~h E
Q y
EHODED CONCRETE NEUTRON PRODUCED RADIOACTIVITY 0.1 E
Sco 0.01 -
0.001 -
f f
I I
f f
i 1
10 100 1.000 TIME AFTER BURIAL (YEARS)
The neutron induced radioactivity in the eroded concrete that is dissolved in the groundwater by infiltrating rainwater is extremely small in comparison to the natural primordial radio-activity already existing in the groundwater. The neutron induced ps radioactivity will completely disappear, through decay, before V) the groundwater diffuses from the burial location to any offsite i
location. For comparison, radioactivity in ordinary tapwater is also shewn.
4-26
FIGURE 6.
1.1 BACKGROUND
SUPVEY AREAS LOCATIONS O
r
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PR AIA'E A W E NUN' D
The construction in the areas surveyed was done during the same time period as the reactor facility. The measured radiation at these locations is consistent with measured natural primordial radioactivity in the concrete and soil, and with the measured radioactivity in the reactor facility.
(-
4-27
8 FIGURE 6.2.1 FINAL SUP.VEY AREA CLASSIFICATION O
==
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'/ / MEDIUM hN$ HIGH The survey areas were assigned their categories based on the specific operating activity in the area, as described in the text.
O) x /
4-28
FIGURE 6.3.1 FINAL RADIOLOGICAL SURVEY EXAMPLE OF BLOCK SURVEY O
20 l
19 18 i
l D E C O N- --CHEM LAB 17 COUNTING I
I j
. HOT 11 C E L'L 10 f
h HOLD 9
l[
UP ROOM L.
8 REACTOR
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L M N O P Q R S T R ANDOM SELECTED BLOCKS The walkway area is shown with the random block selections indi-cated.
Fifty percent of the blocks were measured.
The blank area in the figure, such as the reactor, etc., were not part of
(
the walkway area, and were surveyed separately.
i 4-29
/m PHOTOGRAPHS s
L]
i PHOTO SUBJECT PAGE 3.0.1 EXTERIOR VIEW OF FACILITY.....................
5-2 3.0.2.1 INTERIOR VIEWS OF FACILITY (2 PHOTOS).........
5-3 4.3.1 SUMP TANK.....................................
5-5 5.1.1 NORTHROP TRANSFER CASK........................
5-6 5.1.2 BMI-1 SHIPPING CASK...........................
5-7 5.1.3 FUEL TRANSFER AREA............................
5-8 5.1.4.1 FUEL TRANSFER OPERATION (6 PHOTOS)............
5-9 5.2.1.1 REACTOR HARDWARE (3 PHOTOS)...................
5-15 5.3.1.1 EXPOSURE ROOM DECONTAMINATION CONTROL (2 PHOTOS)..................................
5-18 5.3.2 EXPOSURE ROCM PLUG DECONTAMINATION............
5-20 5.3.3 EXPOSURE ROOM CONCRETE SAMPLE BORING..........
5-21 5.3.4.1 EXPOSUPE ROOM LINER REMOVAL (2 PHOTOS)........
5-22 5.3.5.1 RADIOACTIVE CONCRETE REMOVAL (3 PHOTOS).......
5-24 5.4.1.1 EXPOSURE ROOM DEMOLITION (2 PHOTOS)...........
5-27
(
)
5.5.1.1 COBALT-60 HOT CELL COMPONENTS (3 PHOTOS)......
5-28 5.6.1.1 RADIOCHEMISTRY LAPORATORY DECONTAMINATION
( 2 P H OT O S )..................................
5-31 5.7.1.1 DUCTS AND PLUMBING DECONTAMINATION (3 PHOTOS)..................................
5-33 5.8.1.1 CONTAMINATED WASTE DISPOSAL (2 PHOTOS)........
5-36 6.3.1 SURVEY BLOCK MARKINGS - EXAMPLE...............
5-38 6.4.1.1 FINAL SURVEY, HOLD-UP ROOM ( 2 PH OTOS ).........
5-39 6.4.2 FINAL SUPVEY, COUNTING ROOM...................
5-41 6.4.3 FINAL SURVEY, OVERHEAD........................
5-42
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5-1 9
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PIIOTn 3.O.2.1 Ig EF:OF VIEW OF FACILITY F
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View of the interior of the Peactor Facility, photograph taken in 1963.
The reactor is located at the mid-pool position.
The Cobalt-60 llot Cell is seen in the left foreground.
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NORTIIPOP TPANSFEP CASK O
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Fuel transfer area showing the plastic O
skirt, draped between the pool structure and the shipping cask, to collect any water that may have dripped from the transfer cask during transport of the fuel elements.
Personnel are in the process of inserting transfer cask into the shipping cask.
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PHOTO r>.i.4.3 FUEL 'IkAI SFI:R OPEPATION l
7 jM3.,' i
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O Fuel transfer operation showing the lifting of fuel elements from the transfor cask into spaces in the BMI-1 shipping cask, using the transfer tool.
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l EXPOSUPE FOOM DECONTAMINATION CONTROL O
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.2 The plastic cover to the reactor pool is r.hown in this photograph.
The concrete demolition equipment is being lowered into the pool.
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EXPOSURE ROOM DEMOLITION
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DUCTS AND PLUMBING DECONTAMINATION I
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