ML19312E132
| ML19312E132 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/31/1980 |
| From: | GILBERT/COMMONWEALTH, INC. (FORMERLY GILBERT ASSOCIAT |
| To: | |
| Shared Package | |
| ML19312E131 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8006030326 | |
| Download: ML19312E132 (33) | |
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DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF EQUIPMENT FOR SPACES / SYSTEMS WHICH MAY BE USED IN POST ACCIDENT OPERATIONS OUTSIDE CONTAINMENT AT CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING STATION 1 REVISION 1
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PREPARED FOR FLORIDA POWER CORPORATION
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PREPARED BY GILBERT / COMMONWEALTH GILBERT ASSOCIATES, INC.
READING, PENNSYLVANIA MAY 1980 8006080324
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INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
SUMMARY
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Shielding sources . . . . . . . . . . . . . . . . . . . . . . . . . 9 Access maps . . . . . . . . . . . .. . . . . . . . . . . . . . . 14 ,,
Dose rate maps . . . . . . . . . . . . . . . . . . . . . . . . . . . 7(y Dose rate decay time correction .. . . . . . . . . . . . . . . . . 28 Equipment list with integrated dose . . . . . .. . . . . . . . . .
J7 Reference material 257. damage dose . . . . . . . . . . . . . . . . . 23 Heplacement of.Fadlation Sea.sitive Materjaj . . . . . . . . . . . . 26 I
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1.0 INTRODUCTION
This report has been prepared in response to the September 13, 1979, letter
. from Darrel G. Eisenhut of the NRC to all operating nuclear power plants.
This letter presented an implementation schedule for tha recommendations presented in "TMI-2 Lessons Learned Task Force S:atus Report and Short-Term Recommendations" (NUREG-0578). The requirements defined in the September 13th letter were subsequently clarified in a letter f rom liarold R. Denton to all operating nuclear power plants dated October 30, 1979.
Among the requirements defined in the two NRC letters is a review to determine whether post-accident radiation fields unduly limit personnel access to areas necessary for mitigation of or recovery from an accident or unduly degrade the proper operation of safety equipment. Corrective actions for problems identified as a result of the review are also to be determined. This report presents the results of such a review for the Crystal River Unit 3 Nuclear Generating Station.
The review was based on the following guidelines:
a) The post accident dose rate in areas requiring continuous occupancy should not exceed 15 mr/hr.
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b) The post accident dose rate in areas which do not require continuous occupancy should be such that the dose to an individual during a required access period is less than 5 Rem whole body or its equivalent.
c) The integrated dose to safety equipment should be less than the dose that the equipment has been qualified for.
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d) The minimum radioactive term used in the evaluation should be equivalent to the source terms recommended in Regulatory Guides 1.3 and 1.4.
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SUMMARY
Reviews have been performed of: a) the areas requiring access for post-
- accident operations; and b) the radiation qualification of mechanical safety equipment outside containment; basically, electrical equipment is to be handled under IE Bulletin No.79-01B.
i The areas identified as requiring access are:
nuclear sample room hydrogen purge equipment containment air monitor RM-A6 diesel generator room radioactive waste disposal control board control room radiochemistry laboratory count room Considerations associated with the nuclear sample room, radiochemistry laboratory, and count room are addressed in the response to NRC Lessons Learned Task Force Short Term Recommendation 2.1.8. Considerations associated with hydrogen purging are addressed in the response to NRC Lessons Learned Task Force Short Term Recommendation 2.1.5.
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An additional result of the review is that certain modifications are planned which will make-it unnecessary to have post-accident access to certain
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areas. These modiiications include:
o chany,iny, valves pilV-7, 8, 39, and 40 to motor opera ted valves.
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o Adding a motor operated bypass valve for the Makeup and Purification System filters.
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j o Modifying the waste gas system such that venting of the holdup tanks l can be performed entirely f rom a remote, low radiation area.
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o Relocatingtheradioactivewastedisposalpa$ielorporticasthereof
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necessary during post-accident conditions to a low radiation area to assure access.
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As a result of the review, it is recommended that the Radiation Sensitive Materials in the equipment indicated in Table 5-3 he replaced with more radiation resistant materials. Teflon is not capable of withstanding even moderate radiation exposure.
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3.0 !!ETHODS
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3.1 SOURCE TERMS
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The activity essumed for liquid source term calculation is based on 100T. of the noble gas inventory, 50% of the halogen core inventory and 1% of all other nuclides in the core inventory. The activity assumed for gaseous source term calculation is based on 100% of the noble gas core inventory and 25% of the halogen core inventory.
Two liquid source terms were used in the evaluation. For systems which contain post accident recirculation fluid, the source term was based on diluting the liquid inventory discussed in the previous paragraph with the expected volume of fluid in the bottom of the reactor building post accident.
For systems which can contain fluid from the reactor coolant system but .
do not take suction on the recirculation sump, the source term was based on diluting the liquid inventory discussed in the previous paragraph with the volume of fluid in the reactor coolant system.
Caseous source terms were determined for containment and for the waste gas system. Ths containment airborne source term was based on diluting the gaseous inventory discussed previously with the air contained in the containment free volume. The waste gas system source term was determined by taking 100% of the noble gases and 50% of the halogens in the makeup
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diluting that inventory in one-half the makeup tank volume.
Table 3-1 presents the inventories and source terms discussed above for the time period immediately after the postulated accident (T=0). For other time periods.,,the decay parameters given in Refs. I and 2 were used to adjust the source terms for radioactive decay.
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- 3. 2 CALCULATION OF DOSE RATES Dose rates for the areas of interest in this review were calculated by
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determining the potential contributing sources at a representative location. The appropriate source term data for these sources was selected from Table 3-1 and adjusted for decay as required. The dose rate at the representative location was used as the general area dose rate. Both the SDC Code (Ref. 3) and the Gilbert / Commonwealth developed SPOTl Code were used in performing the dose rate calculations. The SPOTl Code uses the methodology originally presented by Ono and Tsuruo (Ref. 4) and developed by Shure and Wallace (Ref. 5). Energy groups required as input to the codes were determined using the gamma ray energy and intensity data in Refs. I and 2 for the nuclides in Table 3-1. The shielding sources by system for various times post-loca are presented in Table 3-2.
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3.3 CALCULATION OF DOSES TO PERSONNEL DURING POST ACCIDENT ACCESS TO VITAL AREAS Personnel doses received in performing a given operation in a given vital area are calculated as the sum of the doses received during travel to and from the vital area and the dose received while performing the given operation in the vital area.
The doses received during travel are determined by calculating dose rates ,
at selected locations (or at a single location if the dose rate along the
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travel route is' relatively uniform) along the travel route using the methodology discussed in Section 3.2 and multiplying the dose rates by f
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the appropriate travel time for each selected location along the travel route.
1)oses received while performing n given operation are determined of multiplying the done rat e l'or t he given area by t he time required to perform the operation.
Dose rates for the given vital area are determined using the methodology discussed in Section 3.2.
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3.4 CALCULATION OF INTEGRATED DOSES TO SAFETY EQUIPMENT The integrated dose to a given item of safety equipment is determined by
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integrating the dose rate appropriate for the given item over a six month time period. Dose rates were calculated using the methodology discussed in Section 3.2.
Dose rates at T=0 are shown for various regions of interest in the plant on Figures 5-1 through 5-3.
4 The dose rate in any given region for a given time post accident may be obtained by multiplying the T=0 dose rate from Figures 5-1 through 5-3 by the multiplication factor for the given time post accident from Figure 5-4.
j As an example, if the dose rate at T=0 for a given region is 1.0+7 mr/hr and it is desired to know what it would be at T=100 hrs, multiply by 3.6-2 i (from Figure 5-4) yielding 3.6+5 mr/hr.
The integrated dose was obtained by integrating under a log-log linear fit to the dose rate data at one hour and five days (120 hr). Observation 2
of the linear fit to these two points on Figure 5-4 will show the integration to be conservative.
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Table 5-1 lists the equipment and its approximate location along with the
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I calculated integrated 6 month dose in Rads.
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Shielding Source Terms (T=0)
Liquid Caseous Containment Containment Source Source Sump Reactor Coolant Airborne Waste Cas Activity Activity Concentration Concentration Conc en t ra t ion Concenttation Isotope (C1) (Ci) (uCi/cc) (uci/ce) (uci/cc) (uCi/ct)
Br-84 7.85 + 6 3.93 +'6 4.24 + 3 2.43 + 4 6.94 + 1 2.44 + 4 Kr-83m 9.25 + 6 9.25 + 6 4.99 + 3 2.87 + 4 1.63 + 2 5.74 + 4 Kr-85m 2.19 + 7 2.19 + 7 1.18 + 4 6.79 + 4 3.87 + 2 i.35 + 5 Kr-85 5.30 + 3 5.30 + 5 2.86 + 2 1.64 + 3 9.36 + 0 3.28 + 3 Kr-87 4.00 + 7 4.00'+ 7 2.16 + 4 1.24 + 5 7.06 + 2 2.48 + 5
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Kr-88 5.60 + 7 5.60 + 7 3.02 + 4 1.74 + 5 9.89 + 2 3.43 + 5 i
sq Rb-88 5.64 + 5 -
3.04 + 2 1.75 + 3 - -
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Sr-89 7.42 + 5 -
4.00 + 2 2.30 + 3 - -
Sr-90 J.99 + 4 -
2.15 + 1 1.24 + 2 - -
Sr-91 9.72 + 5 -
5.25 + 2 3.01 + 3 - -
Sr-92 9.50 + 5 -
5.13 + 2 2.94 + 3 - -
Y-90 3.96 + 4 -
2.14 + 1 1.23 + 2 - -
Y 91 9.S5 + 5 -
5.32 4 2 3.05 + 3 - -
- 10-99 1.2S + 6 -
6.91 F 2 3.97 + 3 - -
Ru-106 2.29 + 5 -
1.24 + 2 7.10 + 2 - -
Xe-131m 4.38 + 5 4.38 + 5 2.36 + 2 1.36 + 3 7.73 + 0 2.72 + 3 Xe-133m 3.07 + 6 3.07 + 6 1.66 + 3 9.51 + 3 5.42 + 1 1.91 + 4 Xe-133 1.27 + S 1.27 + 8 6.85 t 4 3. 9 3 + 5 2.24 + 3 7.87 + 5 Xe-135m 3.26 + 7 3.26 + 7 1.76 + 4 1.01 + 5 5.76 + 2 2.02 + 5 Xe-135 2.09 i 7 2.09 +,7 1.13 + 4 6.48 + 4 3.69 + 2 1.29 + 5 Xe-138 1.17 + S 1.17 + 8 .
6.31 + 4 3.63'+ 5 2.07 + 3 7.26 + 5
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Liquid ( Caseous ( } Containment Containment Source Source Sump Reactor Coolant Airborne Waste Gas ,
Activity Activity Concentration Concentration Concentration Concentration Isotope (C1) (Ci) (uC1/cc) (UCI/cc) (UCi/ce) (PC1/cc)
I-131 3.68 + 7 1.84 + 7 1.99 + 4 1.44 + 5 3.25 + 2 1.14 + 5
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I-132 4.31 + 7 2.16 + 7 2.33 + 4 1.34 + 5 3.81 + 2 1.34 + 5 I-133 6.40 + 7 3.20 + 7 3.45 + 4 1.98 +.5 5.65 + 2 1.98 + 5
- I-134 8.00 + 7 4.00 + 7 4.32 + 4 2.48 + 5 7.06 + 2 2.48 + 5 I-135 6.35 + 7 3.18 + 7 3.43 + 4 1.97 + 5 5.62 + 2 1.98 + 5
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Cs-134 1.27 + 4 -
6.85 + 0 3.93 + 1 - -
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Cs-136 8.02 + 3 -
4.33 + 0 2.48 + 1 - -
Cs-137 4.99 + 4 -
2.69 + 1 1.55 + 2 - -
Cs-138 1.23 + 6 -
6.64 + 2 3.81 + 3 - -
I no Ba-137m 4.67 + 4 -
2.52 + 1 1.45 + 2 - -
I Ba-140 1.25 + 6 -
6.75 + 2 3.87 + 3 - -
La-140 1.27 + 6 -
6.85 + 2 3.93 + 3 - -
Ce-144 7.50 + 5 -
4.05 + 2 2.32 + 3 - -
C i - 5'1 - -
9.06 - 4 5.20 - 3 - -
Mn-54 - -
1.01 - 4 5.80 - 4 - -
Hn-56 - -
2.96 - 3 1.70 - 2 - -
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Fe-59 - -
1.01 - 4 5.80 - 4 - -
Co-58 - -
3.23 - 3 3.00 - 2 - -
Co-60 - -
6.97 - 4 4.00 - 3 - -
Zr-95 - -
8.71 - 5 5.00 - 4 - -
(1). Based on 1007. noble gas core inventory, 50% halogen enre inventory, and 1% of all others core inventory.
(2) Based on 100T; noble gas cu e inventury anil 257. halogen come inventory.
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REVIEW Of AREAS REQUIRING ACCESS FOR POST-ACCIDENT OP
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g- A review has been performed to identify areas requiring access for post-I accident operations.
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Dose evaluations for the required access to these
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areas were performed, and the results are documented below.
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- An additional result of the review is that certain modifications are planned which will make it unnecessary to have post-accident 1
areas. access tc certain These modifications include: '
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a) Changing valves DilV-7, 8, 39, and 40 to motor operated valves .
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s b) Adding a motor operated bypass valve for the flakeup and Pu' rification System filters. ,.,
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c) Modifying the waste gas system such that 4 venting of the holdup tanks can be performed entirely from a remote, low radiationfarea.
d) Relocating the radioactive waste disposa*
panel.or portions thereof
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necessary during post-accident assure access. conditions to a low radiation are'a to 4.1 AREA A
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Area A is th'e designation for the nucicar sample room located on the 95 elevation of the auxiliary building (see Figure 4-1).
associated with this area, For the evaluations refer to the responsen for NRC Lessons Learned Task Force Short Term itecoimnendat ion 2. l.8.
4.2 AREAS B AND d'
.i Areas B and C are the general location of the hydrogen purge n *equipme t which could be used for controlling hydrogen buildup in containment.
These areas are located on the 119' elevation of the Intermediate Building (see Figure 4-2)'.
Access could be required post-accident to. establish hydrogen purge operations.
For the evaluations associated with hydrogen purge,
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refer to the responsc'for NHC Lessons Learneil Task Force Short Term
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4.3 ARTA D
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Area D is the general location of containment air monitor RM-A6. For the evaluations associated with this area refer to the response for NRC Lessons Learned Task Force Short Term Recommendation 2.1.8.
4.4 AREA E
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Area E as identified on Figure 4-2 is the diesel generator room. Continuous access to this area following an accident would be required if offsite
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power is lost. Access to Area E is via Travel Route 2 presented oh
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Table 4-1. The dose rate in the diesel generator room from direc,t,, .-
radiation following an accident is negligibic. The dose for' Travel Route 2 is also negligible. Area E could thus be i:ontinuously occupied Jf required.
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4.5 CONTROL R00t! <'
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The control room is located on the 145' elevation of the control complex.
Continuous and immediate access is required post accident. dccesstothe control compicx is via Travel Route 3 shown on Tabic 4-1. The dose rate in the control room from direct radiation from containment is estimated to be 11 mr/hr at T=0. The dose for Travel Route 3 is negligible. Thesc
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4.6 RAD 10CllEtllSTRY LABORATORY AND COUNT ROOM
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The radiochemistry laboratory and count room are located on the 95' elevation of the control complex. Accenn is reyul reil for ana l ye.l:t o's duele.ir samplea.
Access from the control room is within the control complex,. ,The doue rate In these arcas from sources other than those associated with sampling is
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negligibic. The dose for travel from the control room in hldo negligible.
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1 From the control room in the control complex at
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- and west in the turbine buildint to access the f intermediate building at elevation 119.
2 Access from the southeast corner of the auxiliary I building.
, 3 Access from office building to turbine building 1
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5.0 REVIEW OF RADI Ai10N QUALIFICATION OF SAFETY EQUll' MENT Table 5-1 identifies the approximate location of the safety equipment
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Table 5-2 lists the Radiation Sensitive Materials identified in Table 5-1 and the 25% radiation damage dose for the stated Reference Material of the same or similar composition, data being available for the Reference Material as noted in the Table 5-2 footnotes.
Table 5-1 permits a simple comparison between the calculated 6 month dose and the 25% damage dose for the stated Reference Material for which data exists. This comparison with the assumptions as noted in Table 5-1 shows that various components will not be able to withstand the radiation environment.
These pieces of es,ipment with the Radiation Sensitive Material noted are shown in Table 5-3.
The review shows that those items in Table 5-3 should have the Radiation Sensitive Material replaced with a more radiation resistant material if possibic. If that is not possible consideration should be given to equipment replacement.
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8 (2) Cable qualified to 10 Rads (3) The Radiation Sensitive Material is not well enough characterized to warrant selection of an individual Reference Material; however, examination of Table 5-2 shows that by enlarge 5.0 x 10 6Rads is a reasonable value to select for the 25% Damage Dose,. This being the case, no further action is required for this given piece of equipment. The Radiation Sensitive Materials which fall into this category are given below: Bakelite Plastic shim washers Evalpak 187-lX Packing and Pist. Carlock 7021, 7022 Gaskets Carbon bushings Crane Super Seal #1 A-108 Asb Polyagrylate Silicon Sponge Invar Plastic .. (4) Dose <500 Rads and therefore exact Reference Material is not important. (5) See note (3); replace TFE (teflon).
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e TABLE 5-2 25% RADIATION SENSITIVE DA>fAGE DOSE COMPONENT MATERIAL REFERENCE MATERIAL (RADS) Armco-I Jaf Casket - - Asbestos Asbestos Fabric Filled Phenolic 10 (1)
- JC-187-I-CR Asbebtos Packing Asbestos Fabric Filled Phenolic 10 (1)
[ Braided Asbestos Packing JG-187 Asbestos Fabric Filled Phenolic 10 O(y) Braided Asbestos ?acking JC-187 Asbestos Fabric Filled Phenolic 10 0(1) ,,, ASB Polyacrylate Silicon - - A-108 - - 6 Crosslink Polyethylene Black XLPE Crosslinked Polyethylene 5.0x10 (3) 7 7 Elastomer Diaphragm Butyl Rubber CR-I 50, Styrene-Butadiene 2.1x10 (2), 1.2x10 (2) 8 Rubber (SBR) (GRS-50) U 7 Synthetic Rubber Butyl Rubber GR-I 50, Styrene-Ibtadiene 2.1x10 (2), 1.2x10 (2) 8 Rubber (SBR) (GRS-50) Carbon Bushings - - Crane Super Seal dl - - 6 Grade M EPT Nordel Diaphragm EPDM Ethylene Propylene Diene Monomer 5.0x10 (3) 6 EPD51 Nordel (Sulphur Free) EPDM 5.0x10 (3) 6 EPR O-Ring EPM Ethylene Propylene Monomer 5.0x10 (3) 6 Ethylene Propylene EPM Ethylene Propylene Monomer 5.0x10 (3) 6 Ethyl-Ppro Seals EPM Ethylene Propylene Monomer 5.0x10 (3) Epoxy Impregnated Cellulose Fiber Epoxy Cellulose Pulp Filled 10 (1)
-
Filter
.Evalpak 187-lX Packing - -
SAE P-3 Felt Felt 5.4x106 (4)
?
.
%
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - - - - _ - -
9
.
*
.
.
TABLE 5-2 (Cont'd) 25% RADIATION SENSITIVE DAMAGE DOSE
~
COMPONENT 3'ATERIAL REFERENCE !!ATERIAL (RADS) - 9 ' Fiberglass Insulation Fiberglass 5.0x10 (4) Garlock #7021 Casket . - - Carlock #7022 Gasket - - 9 Glass Mat Media Glass 5.0x10 (4) '
~
7 7 Buna N 0-Ring Hycar OR-15 NBR, Hycar PA-21 1.5x10 (2), 1.6x10 (2) 7 Buna N (Hycar) Seat & Seal Hycar OR-15 NBR, Hycar PA-21 1.5x10 7 (2), 1.6x10 (2) Hypalon Hypalon 5.0x106 (3) Invar - - 8 JM-398 Packing - - 5 Kerite (Brand) - - Vinyl Coated Mylar Label Marlar A 10 (1) Rubber Cronsa Natural Rubber 4.5x107 (2) 7 Rubber Seal Natural Rubber 4.5x10 (2) 7 Red Rubber Natural Rubber 4.5x10 (2) Neoprene Base Adhesive 6 Neoprene 5.0x10 (3) 6 Neoprene Gaskets Neoprene 5.0x10 (3) 0 Neoprene Neoprene 5.0x10 (3) 7 Nitrile Diaphrag: Nitrile Rubber Hycar OR-15 NBR 1.5x10 (2) Nylon 6 Nylon 5.0x10 (4) Plastic Shim Wacher - - Plastic insulatien - -
,
e
%
- _ _ _ - - - - _ _ _ _ _ . .
. .
.
.
.
TABLE 5-2 (Cont'd) 25% RADIATION SENSITIVE DAMAGE DOSE COMPONENT M.ATERIAL REFERENCE MATERIAL (RADS) 6 Silicon Rubber Silicon Rubber Silastic 7-170 6.8x10 (2) Sponge : - - - Teflon Teflon 3.7x10'(4) . Tef1 n Seat & Seal Teflon 3.7x10'(4) - 0 '" Glass-Teflon Insulation Teflon 3.7x10 (4) - 4 TFE Teflon 3.7x10 (4) 6 Viton Viton 1.0x10 (5) 6 Viton 0-Ring Viton 1.0x10 (5) ' Bakelite (Phenol Formaldehyde) - - U, I (1) Exposure which the given material received and upon subsequent testing retained 80% or more of its preirradiation mechanical properties of clongation and tensile strength. Engineering Compendium on Radiation Shielding, *'e l . II, Springer-Verlag, Berlin /Heidelberg, pp. 311-314, (1975).
.
(2) The 25% damage dose for the mechanical properties of tensile strength or elongation. Engineering Compendium on Radiation Shielding, Vol. II, Springer-Verlag, Berlin /Heidelberg, p. 316 (1975). (3) Exposure which the material received and upon subsequent testing retained 75% or more of its pre-irradiation mechanical preperties of tensile strength and elongation. " Insulations and Jackets for Control and Power Cable in Ther=al Peactor Nuclear Generating Stations," R. B. Blodgett and R. G. Fisher, Okonite Co. , Passaic, NJ, 07055, Institute of Electric and Electronic Engineers - June (1969) (4) Dose at which a physical or mechanical property changes by 25%. Remote Handling of Mobile Nuclear Systems, D. C. Layman and G. Thornton; U.S. Atomic Energy Commission /Div. of Tech. Information, Oak Ridge, Tennessee, pp. 506-515, Jan. (1966). (5) Dupont Data Sheet, " Radiation Resistance of VITON," Data Sheet V-A-4-303, E. I. duPont de Nemours & Co., Inc. Wilmington, Delaware 19898. .
-
.
-
,
'
.
*
,
O
.
Table 5-3
,
Equipment Requiring Replacement of Radiation Sensitive Material 1
- Radiation Sensitive Material Name Identification Requiring Replacement
,
WC Compressor Control Valve WD-V465F TFE Packing Makeup Tank Level Trans MU-14-LT Viton 0-ring DCil Cooler Control Valve (2) DC-V17,18 Carbon filled TFE bushings, TFE packing DCil Cooler Control Valve (2) DC-V177,178 Carbon filled TFE bushings, , TFE packing Demineralizer to Bleed lloldup MU-V112F TFE RC Pump Seal Control Valve MU-V16F TFE packing CRD Discharge Isolation (2) SW-V203,381 Teflan seat & seal RC Pump Temperature Control Valve (4) SW-V87,88,89,90 TFE packing Waste Cas Compressor (2) WD-PlA,B TFE
!
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- I *
'
. REFEREt!CES ,
,
'
.
'
- 1. C. M. Lederer, J. M. Hollander, and J. Perlman, " Table cf Isotopes,"
Sixth Edition, John Wiley and Sons, Inc., 1967.
- 2. A. Tobias, " Data for the Calculation of Camma Radiation Spectra and Beta
,
Meating from Fission Products (Revision 3), RD/B/M2669. June 1973.
- 3. E. D. Arnold and B. F. Maskewitz, "SDC - A Shield Design Calculation Code for Fuel Handling Facilities," ORNL -3041, March 1966.
- 4. H. Ono and A. Tsuruo, "An Approximate Calculation Method of Flux for I Spherical and Cylindrical Source with a Slab Shield," Journal of Nuclear Science and Technology, j! (6), pp. 229-235, June 1965.
_
- 5. K. Shure and O. J. Wallace, " Compact Tables of Functions for Use in Shielding Calculations," Nuclear Science and Engineering, 56, pp. 84-94, January 1975.
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