ML19290H022
| ML19290H022 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/15/1980 |
| From: | Brasher J, Daniel J, Neville J, Schlomer E METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19290H019 | List: |
| References | |
| RTR-NUREG-0591, RTR-NUREG-591 NUDOCS 8012220388 | |
| Download: ML19290H022 (44) | |
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METHODOLOGY AND CALCULATION OF INTEGRATED DOSE TO j
EPICOR-II PREFILTER 29 4
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METHODOLOGY AND CALCULATION OF INTEGRATED DOSE TO EPICOR-II PREFILTERS
)
J.
aniel, Minager, Radiological Analysis 3
i J.
- . Brasher, Dept. Manager, Radiological Controls 7
REVISION 1 DECEMBER 15, 1980 METHODOLOGY AND CALCULATION OF INTEGRATED DOSE TO EPICOR-II PREFILTER 29 J.A. DA'i1EL E.A. SCHLOMER J.J. NEVILLE
-._~. _ _ _
w_-4 RADIOLOGICAL ANALYSIS GROUP 1
TABLE OF CONTENTS Page Nunber 1.0 PURPOSE AND SDSIARY 1
1.1 Purpose 1.2 Su==ary 2.0 Source Ter: Generation 2
2.1 Fission Product Activity in Core 2.2 Fission Product Activity in Coolant 3.0 Material Balance and Curic Deposition Calculation 8
3.1 Curie Deposition 3.2 Determination of Specific Activity 3.3 Calculation of Gat =a Flux 4.0 Energy Absorption Calculations 17 4.1 Ga==a Energy Absorption 4.2 Beta Energy Absorption 4.3 Energy Absorption frc= Transuranics 5.0 Su==ary of Calculations 22 5.1 Total Dose Rate Comparison Attachment I,
4
1.0 Purpose and Summary 1.1 Purpose The purpose of this report is to describe the methodology of calculating the integrated dose to EPICOR-II ion exchanger material; specifically, in the prefilters of that system since the prefilters have been used to remove the most highly concen-trated radionuclides of the contaminated waste water from TMI-2.
Results of calculations for Prefilter 29 are included in order that the postulated effects of radiolytic decompo-sition might be evaluated for the actual resins and chemical additives used. The effects of radiolytic decompostion are being evaluated by the Materials Technology Section of CPU.
1.2 Summary EPICOR-II became operational on October 22, 1979, to process intermediate level waste water,i.e., waste containing less than 100 uC1/ml.
It was recognized at the time of operation that radiolytic effects from fission products could have a direct i= pact on resin stability and the degradation impact on the integrity of the carbon steel liner itself. Operational limitations were set during. e month of July, 1979, and were agreed upon by all parties involved.
(Ref.1) Specifically, curie loading limitations were set at 1300 curies for the prefilters. However, radiation surveys taken on the initial spent liners during changeout indicated that curie loading was taking place in a relatively narrow band of the resin.
In April, 1980, Radiological Analysis Group was requested to study,
the impact of this occurence.
Included in this report is a realistic analysis of energy absorption by the resin, i.e.,
neither overly conservative nor overly optimistic. Conserva-tism vs. non-conservatism is discussed in the text.
In cases where actual information and/or data was unavailable, logical assumptions were made based on accepted experimental data from the industry.
Comparison of radiation survey data to shielding calculations indicate that approximately 80% of the total curies deposited in a prefilter takes place in a relatively narrow band of the resin in the liner, i.e.,the cation layer. This specific activity results in a localized integrated dose of approximately 1.0 x 108 RADS within 1 year of removal from service, for those prefilters loaded to the maximum curie loading. The calculations in this document are for Prefilter 29 only, and are based on information available. Updates to this calcula-tion as well as integrated dose to other prefilters will be made as more detailed information becomes available.
A 1
2.0 Source Term Generation Calculation of an energy absorption rate is a simple and straight-forward procedure once the voluterr c source term, hereaf ter referred to as S, is defined. Depending upon the degree of accuracy desired, y
the generation of S varies from an estimation to a detailed fission y
product transport / chemical balance calculation. The source term generated for this calculation includes all available data with respect to the fission products identified to be present in water processed by EPICOR-II as well as fission product transport calcu-lations for those isotopes known :o be present but not identified because of limitations in the counting techniques.
2.1 Fission Product Activity In The Core The fission product activity in the Un.t 2 core has been calculated by the digital computer code RIGEN (Version 2) which B&W has modified and named LOR-2.
he fission product inventory was calculated using the appropriate pocer history and boron concentratien history for Unit ?
The iqitial conditions were 83,000 Kg Uranium with av averaga enrithment of 2.63 wt.% U-235, and final burnup of i 75 ?ti MTU (93 EFPD).
This data was used as input to the computer coce RADTRAN, (Ref.2) a time dependent fission product transport cod..t, to determine the fission product release from the core. The fission products were decayed to the time of cladding failure (-130 min.) and then allowed to leach out of the fuel as a function of time.
The escape rate coefficients represent the fraction of the activ-ity in the fuel that is released, per unit time, from the fuel,
matrix. Values of these coefficients are derived from experimental data for most elements.
(Ref. 3-7) The values of the escape rate cuefficients for elements of lesser importance were estimated based on the chemical similarity with elements for which the escape rate coefficient was experi-mentally determined. The escape rate coefficiants used in this analysis are givca in Table 2.1.
2.2 Fission Product Ac_1vity In Coolant The general rate equation for the inventory of a radioactive nuclide, N, in the coolant is:
dNe = =N +fA'N '-AN -SN -yN t
C C
C C
'n'he re :
N = inventory of radioactive nuclide in coolant c
N = inventory of same radioactive nuclide in the fuel f
_1
= escape rate coefficient of N (sec )
a f = fraction of precursor which decays to N A'N '= activity of precursor which decay to N A = decay constant of N (sec 1)
S = removal rate coefficient of N by intentional c
removal ___c primary system (sec-s ) determined by:
R, Ec -
Where:
N R = letdewn rate (Kg/sec)
M = mass of primary system (Kg) y = removal rate coefficient of N by discharge out of the PORV determin,d by:
R
_E Mc where:
R = discharge rate out of PORV (Kg/sec)
M = cass of primary system (Kg)
The sequence of events was used to determine the times at which the PORV was open, and the Moody Critical Flow Tables were used to determine the actual flow out of the PORV.
Letdown flow began at 5 minutes into the accident. The assu=ption was made that letdown flow was directed to the Reactor Coolant Bleed Holdup Tanks. This assumption is based on the knowledge that this action can be initiated by an operator in the control room without hr.ving an auxiliary operator enter the Auxiliary Building,.nd normal valve alignment is for the letdown to be directed to the RC Bleed Holdup Tanks.
Prior to the accident, an entry in the auxiliary operator's logbook dated March 27, 1979 indicated that the Miscellaneous Waste Holdup Tank (MWHT) contained approximately 15,300 gal.
Usable volume of this tank is -19,600 gallons. The same entry noted that the Auxiliary Building Sump Tank was full (-3000 gal),
and that the Auxiliary Building sump was empty.
Sample data from the Reactor Coolant Bleed Tanks were used to determine the amount of dilution due to water in the tanks prior to core damage, and all isotopes were adjusted accordingly.
It is recognized that water transfers were made during the month of April, 1979, into and out of the bleed tanks. Ho:.
- r, these transfers were between tanks in the Auxiliary Building, which were contaminated with essentially the same ratios of fission products, varying only in concentration.
It is therefore felt that the concentrations shown in this document are not significantly different from what was actually in the bleed tanks. Such transfers would have introduced oxygenated water into che bleed tanks along with other possible chemicals.
-Sa-
TABLE 2.1 ESCAPE RATE COEFFICIENTS Element Coefficient (sec-1) 8 Kr 6.5 x 10
~8 Xe 6.5 x 10 11 Sr 1.0 x 10 3
Br 1.3 x 10
~3 1
1.3 x 10 3
Cs 1.3 x 10 3
Rb 1.3 x 10
~
Cd 1.0 x 10 '
S In 1.0 x 10 Sn 1.0 x 10-9
_9 Mo 2.0 x 10
_9 Nb 2.0 x 10
_9 Tc 2.0 x 10 Ru 2.0 x 10 '
~
Rh 2.0 x 10 '
11 Ba 1.0 x 10 12 Y
1.6 x 10 12 Ce 1.6 x 10 12 Pr 1.o x 10 12 Zr 1.6 x 10 22 Nd 1.6 x 10 12 Sm 1.6 x 10 i
iAOLE 2.2 RC3T
'B' ISOTOPIC INVENTORY AS OF MAY 3, 1980 ORNL Reported EXXON
- Calculated Nuclide Activity (uci/ml)
Nuclear (uci/ml)
Activity (uCi/ml)
Cs-137 37.57 36.47 35.1 Cs-134 6.80 6.75 6.60 Ba-137m 33.34(+1)
.30
.4 Sr-89
.17 0.17 Y-90
.4 Ru-106 1.58(-3)
Ru-103 6.83(-5)
Rh-106 1.58(-3)
Rh-103m 6.83(-5)
Sb-125 5.10(-3)
Nb-95 2.90(-3)
Zr-95 1.46(-3)
Co-60 1.31(-4)
Co-58 5.56(-5)
Mn-54 3.19(-4)
Ag-110m 4.14(-5)
Ce-144 3.48(-3)
Ce-141 9.78(-6)
Sn-113 4.11(-5)
In-113m 4.11(-5)
Pr-144 3.48(-3)
Pu
.012 ppb U
13.0 ppb
- Preliminary Results, as of November 25, 1980,
3.0 Material Balance and Curie Deposition Calculation Concentration products were calculated for possible chemical cocpounds using elemental concentrations obtained from Oak Ridge National Laboratory analyses of Reactor Coolant Bleed Tank (RCBT) 'B' sample (Ref. 8).
These concentration products were compared to the correspond-ing solubility products, available in the literature, and very few of the possible compounds showed any possibility for precipitation.
Since the systems connected to the primary system are kept oxygen free to alleviate corrosion proble=s associated with oxidation, the assump-tion was =ade that oxide formation was negligible, as was radioactive-oxyanion for=ation. Thus, the nuclides present in RCBT 'B' were assumed to be pri=arily in cationic form and were f avored for c.ation exchange.
3.1 Curie Deposition The data obtained from effluent samples taken during processing was used to deter =ine radionuclide effluent concentrations. These effluent sample results were averaged to determine the mean iso-topic concentration during processing. The difference was taken between the influent and effluent concentrations of known isotopes to determine an approximation of the radionuclides deposited per el of processed water. Total curies deposited was calculated by multi-plying by the total volume of water processed. Table 3.1 is a tabulation of the curies deposited in the cation region of PF-29 by the above method. Also shown is the activity in the chemically bound water after dewatering.
3.2 Determination of Soecific Activity Attachment I containes the radiation survey data for those prefilters that were surveye4 together with the band of principle loading. The instrument was fixed at 9" from the liner, and radiation readings were re.ordud at 6" vertical intervals.
Cation exchange resins of the highly aci31c suifamated polystyrene type have swelling ranges frca 3-to-15% (3%
for highly cross-linked resins to 15% for 1(s croe -;3rked resins).
If a low degree of cross linkagc is assured.
there is greater swelling capability and :hus a lower specific activity after exchange. However, a low Jegree c: cross linkage results in less radiation stability. Al te rnatively,
if a high degree of cross linkage exists, smalter swelled volume occurs and results in higher specific activity after exchange but enhanced radiation stab.lity.
This may be graphically demonstrated by Figures 3.1, 3.2 and 3.3.
These figures are not meant to be interpreted quantitatively, they are only provided to help the reader understand the relationships between these characteristics. On a dry-weight bacis the exchange capacity of a resin is inversely proportional to its cross linkage (Ref. 10).
For the calculation of specific activity loading in PF-29, an average swelling over the total resin volume was assumed to be about 25%. This average swelling of 25% was used to find the volume of cation in which distribution of the cationic radionuclides took place. The actual distribution of the radionuclides could conceivably take place in a smaller volume.
It should be noted that this assumption is not particularly
_9_
conservative in one sense, i.e.,
it results in lower specific activity. The calculated loading was compared to the radiation survey taken on PF-29 when it was removed from service.
Com-parison of survey data to deposited curies indicate that 80%
of the principle gnmnn emitting nuclides were distributed in the layer identified as the band of principle curie loading.
It is assu=ed that this is a cation layer with even distribution.
Further, the assu=ption was made that 40% water retention took place in the resin volume, due to chemical / mechanical bonding.
The volumetric source term generated was then used as input to the computer code ANISNBW (Ref.9), a one dimensional dis-crete ordinates transport code, to calculate the ga==a flux.
3.3 Calculation of Ga==a Flux The ga==a source tern as calculated above was used as input to the ANISNBW code as volumetric source strength in 12 energy groups in units of photons /cm3 sec.
Further, material properties were used which approximate a mixture of wetted resins. Also, the physical dimensions of the liner and its location in the Long Ters Storage Module was modeled to account for those photons being scattered back into the liner fron the wall of the storage cell, The contribution from liners in adjacent cells was neglected.
'.he weight percent compositions of the dewatered cation as well as the energy group source ter=s used are shown in Tables 3.2 and 3.3 respectively.
It was assu=ed that a sulfonated polystyrene resin (Na-form) was used. This assumption has some effect on the calculation of ga==a flux, in that a higher density material would increase energy absorption, and thus increases the absorbed dose. -
A total of 51 intervals in the source volume (from the center-line to the edge of the liner) were used to increase the ac-curacy of the calculation. The geometry of the liner in the cell is sho"'n in Figure 3.4.
ANISNBW was then used to calcu-late the total photon flux at each of the 51 intervals. The code uses energy dependent scattering and absorption cross sections for each element in the material specified according to its weight percentage. The output of ANISNBW is a photon flux at each interval broken down into discrete energy groups. -
TABLE 3.1 CALCULATFD CURIE DEPOSITION IN PF-29 (5/9/80)
Activity In Chemically Isotope Cation Deposition (Ci)
Sound Water (Ci)
Co-58 5.84 E-4 7.12 E-6 Co-60 2.05 E-3 1.68 E-5 Sr-89 1.49 E+1 7.77 E-2 Sr-90 4.88 E+1 2.56 E-1 Y-90 4.88 E+1 2.56 E-1 Mn-54 6.70 E-3 4.08 E-5 Zr-95 3.14 E-2 1.87 E-4 Nb-95 6.31 E-2 3.71 E-4 Ru-103 3.71 E-4 8.14 E-6 Rh-103
- 3.71 E-4 8.74 E-6 Ru-106 4.91 E-3 2.02 E-4 Rh-106 4.91 E-3 2.02 E-4 Ag-110m*
0 5.30 E-6 Sn-113 0
5.26 E-6 In-113m*
0 5.26 E-6 Sb-125 5.84 E-2 5.53 E-4 Te-125=*
1.34 E-2 1.50 E-4 Cs-134 1.91 E+2 9.97 E-1 Cs-137 8.64 E+2 4.49 E0 Ba-137m*
8.16 E+2 4.26 E0 Ce-141 1.31 E-4 1.25 E-6 Ce-144 4.64 E-2 4.45 E-4 Pr-144 4.64 E-2 4.45 E-4
- Note:
International Commission on Radiation Units and Measurements Report, ICRU REPORT 19, July 1, 1971 specifically defines an isomeric tran-sition as a nuclear transformation, and is measured in units of curies..
T
TABLE 3.2 WEICHT *. COMPOSITION OF DE'='ATERED RESIN Ele =ent Weight Percentage H
7.4 0
44.8 C
37.2 S
6.2 Na 4.4 Assu=ption: Resin used is sulfonated polystyrene (Na-for=)
3 Density:
0.705 gn/cm.
TABLE 3.3 ENERGY GROUP SOURCE TERM 3
Groun Averace Enercy (MeV)
Photons /cm see 2
29 2.25 2.47 x 10 30 1.83 4.05 x 10 5
31 1.495 5.33 x 10 5
32 1.165 4.42 x 10 5
33
.9 1.43 x 10 7
34
.7 8.53 x 10 6
35
.5 1.27 x 10 2
36
.35 5.13 x 10 2
37
.25 1.05 x 10 3
38
.15 2.28 x 10 39
.075 1.84 x 10 3
40
.03 3.16 x 10 ti ui u
Z<ez
-v:
W.
x E
H<
Ti EI x;
CROSS LINKAGE Figure 3.1 l
1 N
s N
2a:
m N
CROSS LINKAGE 0
Figure 3.2 NN\\
cC Ec.<
U N
e s
x
=
\\
uxu CROSS LINKAGE
?
Figure 3.3 -
FIGURE 3.4 DIMENSIONS OF LINER IN STORACE CELL
-a.
- +--. 6 35 cm i
- s e.,
i
- -- 4 5. 7 2 L " 60.325 W 9*
CD Cm
.e l
iS.'
.,,$h 4 *. '.' 9
..c ~,
.1. '
e.. o,
l
- ,8 ea
.,,.n.
I
.g a '
e
,.e"
' 'f ','4
_v 3
p.-
e
- i. *.
p e,. g
,,e j ( -[s.' f, '. r ". #. ;.l ' 9.'e..','. I.# [ 4.,. 5 .' e,"h. - It.?'.44,,'p:*'[n',]a*,<-n,f.#[-
'!' # '* I'. f.
[
' 2
' 'o
, a e e 6 s.s ', aj
- 4. f 4_p_
s J,e
's.
., 6_
- ' B.' 4. e. *. e -ej.. [v f y = a
.5 L,. q '#
5 4,'
__q_
4.0 Enercy Absorption Calculations 4.1 Ganna Energy Absorption The photon flux calculated by ANISNBW at each interval was cultiplied by the appropriate energy absorption coefficient at that particular energy and summed over all energies to determine the energy deposition rate at that interval.
This calculation was perforced by the co=puter code INTDOSE, developed for this analysis.
Expressed mathematically, f = j[ 1.602 x 10 ' y*/0 i Y
~
?E i=1 where D is the dose rate in rads /sec at the it.terval, and n = number of energy groups. A more detailed explanation is given below. The ga==a dose rate as a function of radius in the cation bed is shown in Figure 4.1.
4.2 Beta Energy Absorption The dose rate from beta emitting nuclides calculated to be in the cation bed was calculated essentially the same as that for the ga==a flux, with the assumption that the beta particle could not escape the media in which it was located.
The average energy of the beta particle was calculated by assuming 1/3 of the maximum energy of the beta particle.
Expressed mathematically, the average energy of the beta is given by:
E
=
8 S max 3
- where, ESmax = maximum energy of Beta particle (Ref. 11)
- 1 7-
The above energy balance assumes 2/3 of the remaining energy of disintegration is attributable to the neutrino. A detailed description of beta energy absorption is given. below, as well as the detailed calculation results for Prefilter 29.
4.3 Energy Absorption from Transuranics Energy absorption from transuranics is not included in this report at this tire due to lack of quantitative laboratory analyses. When these analyses are completed, this calculation will be incorporated in this report by revision.
The equation to be used is essentially that for beta energy absorption, with the exception that f is replaced by E '
g T
where E is the energy of the emitted particle.
In the case T
of beta emitting transuranics, the same equation is used, again assuming 1/3 of E as the energy of the particle.
ax,
CA>S!A ENERGY DEPOSITION RATE:
ENERGY FLL'X IS GIVEN BY:
?E Y
ENERGY DEPOSITION RATE IS GIVEN BY:
_e tE
_o _
2 (cm /ga)
Energy Absorption coefficient
'a'he re : w,j 0
2 0
Photon Flux (y/cm *sec)
E Energy of Photon (MeV)
ABSORSED DOSE RATE IS GIVEN BY:
hCE RAD /sec = 1.602 x 106 o
~6 UNIT BALANCE:
c8 y
MeV RAD gf. 1.602 x 10 e/g gd ed'*see 100 eig Eiv Note that u is (p - I,)
e 1.e.,
scattering is not included
~8 D RAD /sec = 1.602 x 10 e
- E D
_ 10
FIGl'RE 4.1 GA.T!A ENERGY ABSORPTION RATE vs RADIt'S 2.4 2.3 2.2 I
2.1 2.0 1.9 1.8
_i 1 7 _ _... _ _
n, d 1. 6 s
== t-
Sc W 14.__
E_ l. 3 E-L E 1. 2
?.:c
< 1.1 3
E
$ 1.0 g
E 0. 9 0.7 0.6 0.5 2.
0.4
(
O.3 i
- 0. 2 _
~... _.... _..
O.1 __
10 20 30 40 50 60 DISTANCE FROM CENTERLINE (c=)
BETA ENERGY DEPOSITION RATE:
ENERGY FLUX IS CIVEN 3Y:
3.7 x 1010 S E, I3 u
E';ERGY DEPOSITION RATE IS GIVEN SY:
Tl k 3 3.7 x 1010 gg g
p Where:
k unit balanace constant determined below e
energy absorption coefficient (assumed =1)(1/cm)
D densityofmedia(g/cm) 3 p
volumetricsourceterm(ci/cm) 3 Sy E
average energy of beta praticle MeV g
,,Etee Unit Balance:
RAD cm 1.602 x 10 erg I
cm 3.7 x 1010 Ci 6
3 cc 100 erg MeV
_cm _
.705 gm_
see Ci _ coi
[g 8.4 x 102 g RAD /sec
=
g and the infinite,0 dose is given by:
De (RAD /sec)
Af (1/sec)
Where:
A Decay constant of isotope (1/sec),
PREFILTER 29 BETA DOSE RATE (CATION BED)
Isotope 1(1/sec) v(Ci/cc)
S(MeV)
(RAD /sec)
( ADS)
Co-58 1.12x107 1.36x10~9
.43 4.9x107 4.39 Co-60 4.18x103 4.80x10~9
.497 2.0x106 4.79x102 Sr-S9 1.55x107 3.44x105
.496 1.43x102 9.25x10' Sr-90 7.83x10~10 1.13x10
.182 1.72x102 2.20x107 4
Y-90 3.01x106 1.13x10
.763 7.24x102 2.41x104 4
Zr-95 1.26x107 7.26x108
.373 2.28x105 1.81x102 Nb-95 2.29x10~7 1.46x107
.308 3.78x10~5 1.65x102 7
Ru-103 2.02x10 8.73x10 10
.242 1.77x107 8.77x101 Ru-106 2.20x103 1.20x108
.013 1.32x107 6.00 Rh-106 2.31x103 1.20x108 1.18 1.19x105 4.42x102 Ag-110m 3.17x108 1.22x10 Il
.5 5.12x103 1.61x10~1 Sb-125 8.1.'x109 1.36x107
.208 2.37x105 2.92x103 4
Cs-134 1.07x10 e 4.41x10
.485 1.79x101 1.68x107 Cs-137 8.25x10 10 2.00x10~3
.392 6.58x101 7.98x109 Ce-144 2.77x108 1.08x10~7
.109 9.89x10 3.57x10' 4
Ce-141 2.51x107 3.04x10 10
.194 4.95x108 1.97x101 Pr-144 6.60x10 1.ORx107 4
.999 9.06x105 3.27x103 TOTALS 9.42x101 8.36x108 9
5.0 Su==arv of Calculations 5.1 Total Dose Rate Conparison The total rate is determined by summing the individual dose rates at each interval, i.e. Figure 4.1.
Thus, at the cen-terline, the dose rate is:
6
(.
+
+ ) A sec 6+
TRU
=
Total y
2.97 RAD /see
=
which is approximately 1100 Rads /hr. The beta dose rate is comparable to that calculated by Georgia Tech, i.e.,.59 vs
.94 for this calculation, which is attributable to source term difference. However, the ga=ma dose rates differ considerably, 0.3 for Georgia Tech vs. 2.03 for this calculation. Aside from the difference in the source terms, this calculation took into account liner geometry in the cell, density of media, as well as scattering into lower energy groups which increases the probability for energy absorption. As can be seen by Figure 4.1, the maxi-mum ga==a dose rate occurs at the centerline, and decreases as a function of distance away from the centerline, as one would expect.
REFERENCES 1.
Letter to John T. Collins, dated July 1980 (TLL-316), Evaluation of Epicor II Waste.
2.
RADTRAN, A Ti=e Dependent Fission Product Transport Code, B & W Conputer Services.
3.
- Allison, G.M.,
and Roe, H.K., The Release of Fission Gases and Iodines From Defected UO, Fuel Elenents of Dif-ferent Lenghts, XECL-2206, June, 1965.
4.
Allison, G.M.,
and Robertson, R.F.,
The Behavior of Fission Products In Pressurized Water Systems. A Review of Defect Tests on UO Fuel 3
Elements at Chalk River, AECt-1338, 1961.
5.
Eichenberg, J.D., et. al., Effects of Irradiation of Bulk UO, WAPD-183, 2
October, 1957.
6.
Fletcher, W.D. and Picone, L.F., Fission Products from Fuel Defect Test at Saxton, WCAP-3269-63, April, 1966.
7.
- Frank, P.W., et.al., Radiochemistry of Third PWR Fuel Matrix Test--X-1 Loop NRX Reactor, WAPD-TM-29, February, 1957.
8.
Letter to R.J. McGoey, dated April 2, 1980 from J.A. Carter, RE:
ERD-79-058, Task 7, RCBT B Sample 9.
ANISNEW, A One Dimensional Discrete Ordinates Transport Code, NPCD-TM-491, Rev. 1, May, 1979.
10.
- Gangwer, T.E., Goldstein, M. and Pillay, K.K.S., Radiation Effects on Ion Exchange Materials, BNL 50781, April, 1978.
11.
Table of Isotopes, 7th Edition, Lederer, Shriley, et.al, Wiley and Sons, Inc., 1978. 9
ATTAC'rDIENT I
Pre filter-28 (Survey Date-5/2/80)
Dose Rate Dose Rate
@ 22.86 c 0 22.86 cm (ren/hr)
(re /hr)
HEICHT HEICHT (cm)
(cm) 137.16 137.16 OP OF WR 121.92 121.92 l
106.68 _
_ 106.68 i
91.44 _
91.44 l
\\
76.20 76.20
\\
60.96 -
60.96 l
45.72 -
- 45.72 30.48 _
30.48 15.24 _
15.24 de 0.00 0.00 L
c
- 8. c
=
m, - s
$5$$
Four-Foot (121.92 cm) diameter liner drawn to scale
P refilter-29 (Survey Date-5/9/80)
Dose Rate Dose Rate
@ 22.86 cm
@ 22.86 cm (rem /hr)
(re=/hr)
HEIGHT 11EIGHT (cm)
(cm) i 137.16 137.16 i
OP OF LI E 121.92 121.92 106.68
_ 106.68
/
91.44 _
91.44
/
\\
76.20 -
- 76.20 60.96
- 60.96
\\
45.72 -
45.72 N
/
30.48 _
30.48 15.24 _
15.24 I
0.00 0.00 i
o e
o com o
- .~^"
jfy8 ooco Tour-root (121.92 cm) diameter liner drawn to scale 6
N
Prefilter-30 (Survey Date-5/13/80)
Dose Rate Dose Rate
@ 22.86 cm
@ 22.86 cm (rem /hr)
(rem /hr)
HEIGHT I!EIGHT (cm)
(cm) 137.16 137.16 I
j 9
OP OF LI m 121.92 121.92 I
i 106.68 _
._ 106.6P l
l 91.44 _
_ 91.44 76.20 -
76.20 60.96 -
- 60.96 45.72 -
- 45.72 30.48 _
30.48 i
15.24 _
- 15.24 f
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oo=c Four-Foot (121.92 c=) diameter liner drawn to scale
Prefilt e r-31 (Survey Date-5/14/80)
Dose Rate Dose Rate 0 22.86 cm 0 22.86 cm (ren/hr)
(rem /hr)
HEIGHT llEIGHT (cm)
(cm) 137.16 137.16 TOP OF LIE R 121.92 121.92 1
106.68 _
_ 106.68 91.44 _
_ 91.44 76.20 -
- 76.20 60.96 -
- 6C.96
\\
45.72 -
- 45.72 N
/
30.48 _
30.48 15.24 _
- 15.24 l
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ueu o eooe Four-Foot (121.92 cm) diameter liner drawn to scale W
Prefilte r-38 (Survey Date-5/27/8'J)
Dose Rate Dose Rate
@ 22.86 cm 0 22.86 cm (rem /hr)
(ren/hr)
HEIGHT llEIGHT (c=)
(c=)
137.16 137.16 j
l TOP OF LISER I
121.92 121.92 106.68 _
_ 106.6P i
l g
91.44 _
91.44 i
76.20 -
76.20 60.96 -
- 60.96 l
45.72
- 45.72 30.48 _
30.48
- 15.24 15.24 _.
0.00 0.00 l
i c c c
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5 ON N
c c, e y ee c oocc Four-Foot (121.92 cm) diameter liner drawn to scale
Prefilter-40 (Survey Date-5/31/80)
Dese Rate Dose Rate
@ 31.1 cm 0 29.8 cm (rem /hr)
(re /hr)
HEIGHT HEIGHT (cm)
(cm) 137.16 137.16 OP OF LI m 121.92 121.92 106.68 _
_.106.6P
}l
/
91.44
_ 91.44 76.20 -
76.20 60.96 -
- 60.96 45.72 -
- 45.72 30.48 30.48 15.24 _
_ 15.24 de 0.00 0.00 i
~
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Four-Foot (121.92 cm) diameter liner drawn to scale
Prefilter-41 (Survey Date-6/2/80)
Dese Rate Dose Rate G 31.1 cm 3 29.8 cm (re=/hr)
(rem /hr)
HEIGHT HEIGHT (en)
(cm) l l
137.16 137.16 TOP OF M R l
121.92 121.92 I
i l
8 i
106.68 _
_ 106.6P I
i 91.44 _
91.44 76.20 -
- 76.20 60.96 -
60.96 45.72 -
~ 45.72 30.48 30.48 15.24 _
.- 15.24 eIc 0.00 0.00 L
3=
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Four-Foot (121.92 cm) diameter liner drawri to scale
Prefilter-42 (Survey Date-6/4/80)
Dose Rate Dose Rate
@ 31.1 cm 0 29.8 cm (ren/hr)
(rem /hr)
HEIGHT HEIGHT (cm)
(cm) 137.15 137.16 TOP OF LINER 121.92 121.92 l
106.68 106.6P I
i 91.46 _
91.44 76.20 76.20 -
60.96 -
60.96 45.72 -
- 45.72 30.4S 30.48 15.24 _
_ 15.24 d[
0.00 0.00 i
i e
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c, y
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=
v=w c
= ccc Four-Foot (121.92 cm) diameter liner drawn to scale
Prefilter-43 (Survey Date-6/6/80)
Dose Rate Dose Rate G 31.1 ce 9 29.8 cm (ren/hr)
(rem /hr)
HEIGHT HEIGHT (c=)
(cm) 137.16 137.16 TOP OF LINER 121.92 W.92 1
106.68 _
_ 106.6P I
L.
91.44 _
- 91.44 l
76.20 -
76.20 60.96 -
- 60.96 45.72 -
45.72 30.48 _
_ 30.48 15.24 _
15.24 de 0.00 0.00 i
L c
=
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Four-Foot (121.92 cm) diameter liner drawn to scale
Prefilter-44 (Survey Date-6/27/ 80)
Dose Rate Dose Rate
@ 22.86 cm
@ 22.86 cr (re=/hr)
(ren/hr)
HEIGHT HEICHT (c=)
(cm) 1 l
137.16 137.16 j
OP OF WR 121.92 121.92 I
i 106.68
_ 106.6P 91.44 _
91.44 76.20 -
- 76.20 t
60.96 60.96 -
45.72 -
- 45.72 30.48 _
30.48 15.24
- 15.24 I
0.00 0.00 l
ct e c
c eCe c C
N ON N
v' N O v=w =
~~
C C O C Four-Foot (121.92 cm) diameter liner drawn to scale
- P
Prefilter-45 (Survey Date-6/30/80)
Dose Rate Dose Rate
@ 22.86 cm 0 22.86 cm (re=/hr)
(ren/hr)
HEIGHT HEICHT (cm)
(cm) 137.16 137.16 l
TOP OF LINER 121.92 121.92 f
106.68_.
106.6P I
91.44 91.44 76.20 76.20 60.96 60.96 45.72 -
- 45.72 30.48 _
30.48
- 15.24 15.24 -
t l
0.00 0.00 l
i e eee e non e
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c, y u, y c
$S$5
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Four-Foot (121.92 cm) diameter liner drawn to scale N
Prefilter-46 (Survey Date-7/2/80)
Dose Rate Dose Rate
@ 22.86 cm 0 22.86 cm (re=/hr)
(re=/hr)
HEIGHT HEIGHT (cm)
(cm) 137.16 137.16 l
TOP OF T. M R 121.92 121.92 j
l 106.68 _
_ 106.68 I
91.44 _
_ 91.44 76.20
- 76.20 60.96 60.96 45.72
- 45.72 30.48 _
30.48 15.24 _
.- 15.24 I
J k1c 0.00 0.00 L
e mer a C,
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cc=c Four-Foot (121.92 c=) diameter liner drawn to scale
Prefilter 47 (Survey Date-7/14/80)
Dose Rate Dose Rate
@ 22.86 cm 3 22.86 cm (re=/hr)
(rem /hr)
HEIGHT HEICHT (c=)
(cm) 137.16 137.16 j
TOP OF LINER I
121.92 121.92 l
106.68 _
_ 106.6P 1 1
(
91.44
_ 91.44 I
76.20 -
76.20 60.96 -
60.96 45.72 -
- 45.72 30.48 _
30.48
,_ 15.24 13.24 _
l d-0.00 0.00 l
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Four-Foot (121.92 cm) diameter liner drawn to scale