ML20212H600
| ML20212H600 | |
| Person / Time | |
|---|---|
| Issue date: | 03/03/1987 |
| From: | Roberts J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Woodhall C NAC INTERNATIONAL INC. (FORMERLY NUCLEAR ASSURANCE |
| References | |
| REF-PROJ-M-40 NUDOCS 8703060160 | |
| Download: ML20212H600 (19) | |
Text
{{#Wiki_filter:{' ~ 70 S;;tgj3l l " p.gw, ;1 i pr;dt _ ' gy' " r ~~ Project No. M-40 Pi?: !t;d ' hak 1-Nuclear Assurance Corporation ATTN: Colman B. Woodhall-Oda -- -- Vice President Technical Analysis Division POR^ - 5720 Peachtree Parkway i p g;t _ _.__ Y Norcross, Georgia 30092 ! Reb 1, NdOdd Gentlemen: 1 qq #g,3 o m.; ~ I am enclosing detailed comments (see enclosure) on your't6pical report cntitled " Topical Safety Analysis Report for the NAC Storage / Transport Cask for Use at an Independent Spent Fuel Storage Installation," Revision 1, submitted by letter dated December 18, 1986. As discussed in our meeting of February 24, 1987, there are inconsistencies and omissions in the TR. Close attention should be paid to technical editing in revising this report. If ycu have any questions regarding our commerits, please call me. Sincerely, Original signed kr Join F. Roberts John P. Roberts Advanced Fuel and Spent Fuel Licensing Branch Division of Fuel Cycle and Material Safety
Enclosure:
Comments Relating to Revision 1 of TSAR DISTRIBUTION: '(Project'Fi10 40 PDR NMSS r/f^^ ~ ~ ' FCAF r/f JPRoberts LCRouse FBrown 3St.AAes dte A n 0FC: FCA
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- 03/1/[87 OFFICIAL RECORD COPY 0703060160 070303 PDR PROJ PDR M-40
COMMENTS RELATING TO REVISION 1 0F THE, TOPICAL SAR FOR THE NAC STORAGE / TRANSPORT CASK 2.0 Site Characteristics 2.3 Meterology 2.3.4 Diffusion Estimates The atmospheric diffusion factor (X/Q) is low by a factor of 10. The discrepancy is due to an error in reading the value of sigma off the Class F 7 stability curve of Fig. 2 from Regulatory Guide 1.145. The NAC quotes a value of 52 on page 7.4-10; the curve value is best interpreted as 5. 3.0 Principal Design Criteria 3.1 Purpose of Installation 3.1.1 Materials to be Stored Where are the weights of U and UO2 in Table 3.1-1 (Design Basis Fuel)? Since ORIGEN outputs are expressed per metric ton of heavy metal, this information is necessary to the verification process. What weights were used to derive the activities of the head and foot pieces? Why are the activities of the head and foot pieces not summarized in this section? How conservative is the assumption that the total head and foot 60 piece source spectrum results from Co? 3 2.6.1 Analysis Point Locations This section states that stresses and stress intensities are evaluated at the expected 55 maximum stress points for the cask. (It appears that they are provided for 54 points only.) The finite element models employed in the analyses contain many more than 54 points. Can you provide some assurance that you have in fact captured the maximum stress points with those 54 locations? i 1
3.3 Safety Protection Systems 3.3.5 Radiological Protection e 3.3.5.2 Criteria Regulatory Guide 3.48 requires estimates of the annual collective dose for various operations at the ISFSI, e.g., preparation and transfer to storage, inspection, maintenance, etc. This requirement was to be fulfilled in the TSAR revision. No new information appears nor is there any reference to information appearing in other sections. 3.3.6 Fire and Explosion Protection 3.3.6.1 Fire Protection With regard to the comparison of SCOPE and HEATING 5 calculations in Table 3.3-4 (Temperature Summary - Normal Operations), the NAC states that " SCOPE-calculated temperatures were approximately 30-40* higher than those calculated by HEATING 5." This is only true for the surface and outer skin locations. From the neutron shield inward, HEATING 5 produced higher or equivalent results. Why are the HEATING 5 results for the maximum fuel rod clad tempera-ture not included here? Since HEATING 5 appears to be more conservative for the inner cask region, why are SCOPE results presented in Table 3.3-5 (Maximum Post-Fire Temperatures)? 3.3.9 Heat Rejection The NAC indicates that the maximum safe operating temperature for the solid neutron shield material is 340*F. The normal operations temperature from Table 3.3-4 (Temperature Summary - Normal Operations) for 130*F ambient from HEATING 5 is 334*F. Is the solid neutron shield fully functional at 334*F? From Table 3.3-5 (Maximum Post-Fire Temperatures), the solid neutron shield temperature allowing a fire is 642*F. Since this temperature exceeds the safe operating temperature by some 302*F, how functional is the solid neutron shield with respect to shielding? Please reference the source for the listed " Safe Operating Range," on page 3.3-12. 2
y Table 3.3-14 lists the maximum fuel rod clad temperature? 2t.668*F(353fC) l' This is inconsistent with the 688CF (364*C) listed in Table f.8-19. s - 3.5 Decommissioning Considerations / 3.5.1 Storage Casks The original TSAR did not include the cask ends in the activation calculations. This omission was to be corrected in the TSAR revision, however,thereisnoindicationinthNtextthatthecaskendsareincluded. Furthermore, the activation results presented in Table 3.5-1 (NAC S/T Cask Activation) are unchanged from the original TSAR. Are the cask ends now included? i 4.0. Installation Design s 4.2 Storage Structures t! - 4.2.1 Structural Specifications ~ 4.2.1.1 Cask Description The engineering drawings provided in this section must show the lid penetrations in their entirety. What is the reference for.the AAR specification referred to on page 4.2-3? ~ 4.2.1.3 Properties of Materials 7 4.2.1.3.1 Mechanical Properties of Materials Why is the discussion on the differences'between Type 304 and Type 304L steels included in the TSAR? Were elastic plasticlikterial models used in'.the-analysis? Lead properties are discussed on pages 4.2-9 and 4.2-10. Two figures (Figs. 14 and 24) of Reference 4.9.18 are cited. These appear to bd the wrong figure numbers from this reference. ~ s 3 ,s r w y-r w-
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'f 4 O }-{ 'f ) p h. Indfvidual Onit Descriptio;3 4.2.3 4.2.3.'2 Cask Components s f (). 4.2.3.2.1' Cask Containeent Components ./ Reference is made to metal 0 ring' seals. ;However, Fig. 4.2-3 does not ? indicate that metal seals are used. If they are not, justify the use of 4 ?rbn-metal seals for at least 20 years of storage. -c 4.2.'3.3 Design _,Bajis and Safety Assurance l -;What is the third item in Table 4.2-15? 4.8 giornal Operations 4.8.1 Normal Operation Conditions - Structural Analysis y, (; i 4.8.1.1 / fuel Basket t Do the fuel assembly foct pieces rest on the cask cavity bottom? This is , implied. If not, the n.9ans of support must be explained. b,
- 4. 8.1. 2 Weights and Centers of Gravity In Table 3.1-1 (Design Basis Fuel), the weight of the design basis fuel assembly (islistedas1440lbs.
In Table 4.8-1 (Major Component Weights), the , weight of'the design basis fuel assembly is listed as 1516 lbs. Which is ! correct? 4.8.1.3 Cissk Scj Analysis i. 4.8.1.3.3 Detailed Analysis TheTSARhtatesthatnomechanicalloadsexceptbottomsupportare f ~,; applied to the cask during ncrmal operation (A similar statement is made earlier in Section 4.8.1.3.1). What about the lifting and handling loads transmitted from the trunnions to the cask body?
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7 di a n, 3 e f _4.8.1.3.4 Inner Shell Stability Analysis In this section, the statement is made that the shell temperature extremes are 10*F and 350*F, however, the HEATING 5 results shown in Table 3.3-4 indicate that'the'shell.is at 360*F for Normal Operations, and the post-fire temperatures provided in Table 3.3-5 suggest that the shell temperature will be much higher still. Please clarify. In this section on pages 4.8-18 and 4.8-24, some cylinder crippling and buckling equations are provided, and the reference is listed as 4.9.50 (" United Metallic 0-rings"). Please provide the correct reference. 4.8.1.4.1 Lifting Trunnion Analysis The dimensions given in the table on page 4.8-33 for width or, diameter do not appear to correspond with those on page 4.8-27. . 4. 8.1. 6 Neutron' Shield Shell Analysis 4.8.1.6,24 Impact Analysis It appears from the calculation on page 4.8-73 that only the weight of the BISCO is used to calculate the load on the neutron shield due to a side impact. What about the weight of the cask body? l 4.8.2 Normal Operation Conditions - Thermal Analysis ) 4 8 2.1 Discussion "A two-dimensional (x y) model of the fuel basket is used to calculate l l the maximum basket fuel channel wall temperature for the center assembly." What is the axial level at which the result of this model is represented. 4.8.2.2 Thernal Properties of Materials "The surface to surface radiation was not modeled for any other surfaces because its effect at the temperatures of interest was found to be negligible." According to Table 4.8-16, the surface radiative heat transfer l coefficients for the neglected surfaces are more than 50% of the surface from l stainless steel to the environment. Please elaborate on the justification. l j, 4.8.2.3 Thermal Analysis Models l 5 l \\
4.8.2.3.1 Cask Body For the composite body wall and the composite ends, " radiation is effective across this gap, but only conduction is considered in the analysis for conservatism." Is this conservative for both the' therm'al stress and the fuel rod clad temperature, or just one of them? For the neutron shield tank, " Twenty-four 0.25-inch copper fins are distributed radially for heat transfer purposes. How are these fins modeled in the HEATING 5 R-Z model? With what thermal properties? 4.8.2.3.2 Fuel Storage Basket The thermal expansion analysis to determine gap sizes between the basket and inner containment shell assumes that the inner shell is free to expand without restraint. Actually this shell is backed by a thickness of lead and an outer shell at low temperature. These restraints should be taken into account. 4.8.2.3.3 Maximum Fuel Rod Clad Temperatures "Once the temperature difference across the basket is known, it may be added to the maximum exterior surface temperature.... Finally by combining the temperatures calculated by the three models....." Please explain more about the three models by describbig each model's input, thermal boundary conditions and output. Is one model's output used as another model's input? Or, do you really ' add' and ' combine' the results? 4.8.2.7 Models for Maximum Fuel Rod Clad Temperature A heat flux of 0.150 BTU /hr-in2 is derived for insolation on the curved surface. Where is the derivation for heat flux on the flat surface? 4.8.2.8 Other Thermal Conditions In Table 2.7-1 (Bounding Site Characteristics), the maximum flood water height is listed as 240 ft. Here the maximum flood water height is listed as 150 ft. Which is correct? The maximum post-fire temperatures listed in Table 4.8-24 (Maximum Post-Fire Temperatures) are slightly different than those listed in Table 3.3-5 (Maximum Post-Fire Temperatures) for most locations. Rounding may explain most of the differences; nevertheless, consistency must be maintained. 6
From which model are the temperatures of the 0-Ring and Valves calculated for Tables 4.8-19, 4.8-21 and 4.8-23? Are the 0-Ring and Valves represented in that model? How are the SS upper and lower end fittings (Fig'. 4.B"17) modeled? How is the void at the bottom modeled? What does that void represent? What are the thermal boundary conditions for this model? What are the thermal boundary conditions for the model in Fig. 4.8-18? What axial-level temperatures do Figures 4.8-26 and 4.8-27 represent? Are there any axial variations? To facilitate checking the models, please send a copy of References 4.9.35 and 4.9.42, including a description of their inputs and outputs. 5.0 Operation Systems 5.1 Operation Description 5.1.2 Flowsheets In Table 5.1-4 (Estimated Operation Time and Personnel), the times required to perform fuel-loading operation numbers 6 (tighter closure lid bolts above water) and 14 (emplace upper neutron shield end cap, seal weld, leak check) have increased by a factor of 2.5 and decreased by a factor of 8, respectively. Likewise, the time required to perform preparation-and-transfer-to-ISFSI-storage operation number 4 (place cask, vertical storage) has decreased by a factor of 8. Why? 5.1.3 Identification of Subjects for Safety Analysis 5.1.3.5 Maintenance Techniques The NAC states that, " repair operations upon the NAC S/T cask would be required only as part of the recovery from consequences of off-normal or accident event conditions. These could range from replacement of a leaking penetration 0-ring seal to repair of the portion of the neutron shield outer skin which might be dented or punctured by a tornado missile." The NAC assumed consequences of accident events are much more severe than those depicted in this statement. Section 8.2.1.1 (Cause of Accident - Loss of Neutron Shield) assumes that the upper neutron shield end cap can be lost 7
following a tipover or cask drop accident. Section 8.2.4.2.2 2 (Side Drop) assumes that the neutron shield shell is crushed with negligible energy absorption. /
- 6. 0 Waste Confinement and Management The estimate of the dose due to gaseous radioactive waste must be increased by a factor of 10 to account for the previously identified error in the calculated value of the atmospheric diffusion factor (X/Q).
7.0 Radiation Protection 7.2 Radiation Sources 7.2.1 Characterization of Sources It is requested that "the sources of radiation...be described in a manner needed as input to the shielding design calculations,." Whether in this section or Section 3.3.1 (Materials to be Stored), the information presented remains inadequate. Where are the source terms for the head and foot piece regions? What is the XSDRNPM 27n/18g group structure and why are the neutron and gamma sources not presented in terms of this group structure? 7 2.2 Airborne Radioactive Material Sources The reference given in the NAC Q1 response for the various release 131 3 85 fractions (SAND 83-0867) does not show Xe or H and gives Kr as 10% and 134,137Cs as 0.02L These numbers do not agree with the numbers cited in this section. Which are correct and what is the correct reference? The NAC has yet to describe the " provisions made for personnel protective measures." Are any provisions made? If not, why not?
- 7. 3 Radiation Protection Design Features 7.3.2 Shielding On page 7.3-3, the NAC states that "seven rectangular parallelepiped bodies" were used for the 3-D QAD-CG calculations.
Figure 7.3-5 (QAD-CG Three Dimensional Source Region) shows only six. Which is correct? 8
7 3.2.1 Analysis Source Description It is assumed that the materials listed in Table 7.3-2 (Source Material Compositions) are for the active fuel region only. Where are the materials for the heaa and foot pieces? Was the plenum region hodeled? If so, where is the materials description? If the plenum region was not modeled, why not? 7.3.2.2 Shielding Analysis Dose Points With regard to Figure 7.3-6 (Detector Locations), was the point directly below the bottom of the neutron shield considered? If not, why not? 7.3.2.3 Shielding Analysis Models How are the dose rates from the fuel and the head and foot pieces determined through the top and bottom with the QAD-CG and XSDRNPM 1-D_models? If the fuel is to be stored in a vertical position with the cask supported by an impact limiter, why is the limiter not modeled? The dose rate effect of this storage configuration must be considered. With regard to the radius of the equivalent circularized source, the details of the calculation must be provided. In Figure 7.3-7 (One Dimensional Calculational Model), the NAC shows a radius of 68.19 cm. We calculate a radius of 72.62 cm. Upon comparing the dimensions in the drawings from Section 4.2.1.1 (Cask Description) with the dimensions in Figure 7.3-9 (QAD-CG Three Dimensional Model), many discrepancies were found. The differences are summarized below. Radial Axial Figure Drawing Diff Figure Drawing Diff 90.10 91.38 -1.28 6.99 7.12 -0.13 111.70 111.12 +0.58 18.50 16.50 +2.00 121.22 118.68 +2.54 24.80 24.89 -0.09 122.20 119.32 +2.88 31.70 54.05 -22.35 398.90 388.12 +10.78 416.60 413.51 +3.09 429.20 428.75 +0.55 435.50 435.41 +0.09 438.10 438.91 -0.81 9 1 -,m, .-------r- ---=4-, - - - - - - - ~ - - - - - - - - - - - - - - - - -
Radial Axial -Figure Drawing Diff Figure Drawing Diff 441.40 441.45 -0.05 446.50 446.53 -0.03 Axial drawing dimensions 388.12 and 413.51 represent the extremes of the neutron shield shell. From Table 3.1-1 (Design Basis Fuel) and Table 7.3-1 (Discrete Axial Source Distribution), axial dimensions 419.85 and 428.75 represent the extremes of the active fuel and fuel assembly, respectively. Based upon these comparisons, 6.34 cm of the active fuel has no neutron shielding. This is in direct conflict with previous NAC claims and must be resolved. With regard to Table 7.3-3 (Shield Material Densities and Compositions), why are the atom densities still not provided for the shield materials, why is the stainless steel density composition shown with the aluminum, where is the lead density, and why does the cumulative density for the stainless steel (7.306 g/cc) remain below the density for 304 stainless steel (7.97 g/cc) presented in Section 4.2.1.3.1 (Mechanical Properties of Materials)? The NAC Q1 response to the cumulative density issue is that "the density of stainless steel conservatively included only iron." Does this mean that the stainless steel was assumed to be all iron (density of 7.87 g/cc) or that only the iron composition (density of 5.7 g/cc) was assumed? If the former is true, why show an iron density of 5.7 g/cc and why list the chromium and nickel compositions? If the latter is true, why list the chromium and nickel compositions? Furthermore, assuming a density of 5.7 g/cc for the stainless steel seems too conservative. This table and section need clarification. 7.3.2.4 Shielding Analysis Results - Surface Dose Rates With regard to the copper fins, the NAC states that " ducting through the fins resulted in a slight change in the dose rate on the average; the increased neutron dose rate is outweighed by the decrease in gamma contribution." Was streaming through the copper fins modeled and calculated? If so, how and why is it not discussed? i 7.4 Estimated On-Site Collective Dose Assessment 1 l l 10
7.4.1 Analysis Methodology Was the dose rate at distance R determined by multiplying the total XSDRNPM surface dose rate times the ratio of the distance R to surface IS0 SHIELD dose rates? 7.4.3 Occupational Dose The reference to Table 5.4-4 is incorrect. It should be Table 5.1-4 (Estimated Operation Time and Personnel). With regard to the dose rates listed in Table 7.4-5 (Occupational Doses), which dose point detector location (s) is assumed for each operation? Fuel loading operation numbers 11 through 15 are in error. They should be 10 through 14 to agree with Table 5.1-4. The individual and all-individuals occupational doses for fuel-loading operation number 15 (emplace upper neutron shield end cap) are incorrectly calculated. They should be 4.3 and 8.6 mrem, respectively. The total individual and total all-individuals occupational doses for fuel loading are also incorrect. They should be 132.7 and 265.5 mrem, respectively. The estimated time for storage operation number 1 (optional-internal cavity gas inspection) should be 0.25 hr instead of 0.17 hr. The reference to Table 4.1-4 is incorrect. It should be Table 5.1-4. 7.4 4 Airborne Radiation 7 4.4.4 Boundary Dose Typographical errors in the calculation of Q (1.99 vs 7.99 and 223.123 vs 223,123) must be corrected. Comments made earlier in Section 2.3.4 (Diffusion Estimates) with regard to the atmospheric diffusion factor (X/Q) apply here as well. The error must be corrected. 85 and 3 Whole body dose rates due to Kr H are low by a factor of 10 because of the error in the atmospheric diffusion factor (X/Q). The errors must be corrected. The atmospheric diffusion factor and the dose results for the single and multiple-cask arrays in Table 7.4-6 (Site Boundary Dose Airborne Radioactivity Contribution) must be increased by a factor of 10. The assumption that all 11
85 the dose is due to Kr is not correct. Furthermore, the dose rates presented 85 are not consistent with the assumption that all the dose is due to Kr. To be consistent, the dose rates should be reduced by about a factor of 2. / 7.7 NAC S/T Cask Performance 7.7.2 Controlled Area Boundary Dose Rates - Direct and Airborne Radiation The single cask and 140 cask whole body annual dose rates (airborne contribution) must be increased by a factor of 10 to correct for the error in the atmospheric diffusion factor (X/Q). 7.7.3 Cask Leakage The reference to Section 7.4.4.2 (Analysis - Normal Operation Conditions) is incorrect. It should be Section 7.4.4.3 (Calculated Leakage Rate). 7.7.4 Occupational Dose The total occupational doses for an individual and all individuals are incorrect. They should be 152 and 298 mrem, respectively. 8.0 Accident Analysis 8.1 Off-Normal Operations 8.1.1 Event - Leakage Through a Cask Closure 8.1.1.3 Analysis of-Effects and Consequences - Leakage Through a Cask Closure 8.1.1.3.2 Boundary Dose 85 3 Whole body dose rates due to Kr and H are low by a factor of 10 because of the error in the atmospheric diffusion factor (X/Q). The errors must be corrected. 8.1.2 Event - Fission Product Gas Release 12
8.1.2.3 Analysis of Effects and Consequences - Fission Product Gas Release Even though this section focuses on the structural concerns of fission product gas release into the cask cavity, subsections on the calculated leakage rate and boundary dose should also be included. 8.1.2.4 Corrective Action - Fission Product Gas Release The NAC statement that "since the NAC cask safely contains all fission products released from the fuel rods, no corrective actions are required," is not correct. There is a finite leakage rate established in Section 7.4.4.3 (Calculated Leakage Rate). The NAC statement that "since the release of 100 percent of all fission products from the fuel rods results in dose rates at the Controlled Area Boundary less than the 10 CFR 72.67(a) criterion (see Table 7.4-6), no corrective actions are required," is not substantiated. The dose results must be presented or reference made to the section where they may be found. 8.2. Accidents 8.2.1 Accident - Loss of Neutron Shield 8.2.1.1 Cause of Accident - Loss of Neutron Shield The NAC states that "the neutron shield cannot be lost from the shield shell due to... rupture...from a drop of the cask during handling." In Section 8.2.4.2.2.2 (Side Drop), the shield and shell are " crushed with negligible energy i absorption." What is the correct condition of the neutron shield? This is a major inconsistency that must be corrected. 8.2.1.2 Accident Analysis-Loss of Neutron Shield 1 8.2.1.2.3 Shielding Analysis - Loss of Neutron Shield The shielding analysis results for the assumed loss of the upper end cap neutron shield appears in Section 7.3.2.4 (Shielding Analysis Dose Rates - Surface Dose Rates). The dose results must be presented or reference made to the table where they may be found. 13 I ._,__.c-_.- _ _ _ ~,. -..
8.2.1.3 NAC S/T Cask Performance - Loss of Neutron Shield Corrective action to recover from the cause of the accident (tipover or drop) could result in increased occupational exposure. This must be addressed. i 8.2.2 Accident - Fire '8.2 2.2 Accident Analysis - Fire How functional is the neutron shield? The maximum neutron shield temperature is some 302* above its safe operating temperature. 8.2.3 Accident - Cask Tipover In this section and throughout the TSAR, it is unclear which "g" loads will be used for the quasi-static analyses for impact. A few examples are listed here. pg. 8.2-6 refers to a 55 g side drop and a 21 g tipover pg. 8.2-11 refers to a 55 g end drop and a 14 g tipover pg. 8.2-12 refers to 36.7 in any orientation pg. 8.2-14 contains the statement that the " impact limiters are used to limit the impact loads to 44.7 g's in end and corner and 14 g's in the tipover" pg. 8.2-35 states that the upper side limiter will keep g loads below 24.1 g's 8.2.3.2 Accident Analysis,. Cask Tipover What happens to the shielding properties of the neutron shield following i a tipover accident? With regard to the assumption of 100 percent fuel rod failure, the NAC statement "that the radiological consequences of this are evaluated in Section 7.4 considering that all barrier seals perform normally, Section 8.1.1 considering that a barrier seal leaks, and Section 8.2.6 considering that all i barrier seals simultaneously fail" is incu. rect. Section 7.4 (Estimated l Onsite Collective Dose Assessment) assumes 1 percent failure, Section 8.1.1 (Event - Leakage Through a Cask Closure) assumes 10 percent failure, and 8.2.6 i. (Accident - Cask Seal Leakage) assumes 100 percent failure. Radiological consequences must include the consequences of the loss of the neutron shield. 14 i
The NAC states that there is "no significant shielding consequence" to the 0.6-inch lead slump following an end drop accident. This statement must be substantiated. What about the fuel assembly head pieces as significant streaming sources? Was this modeled and calculated? / 8.2.4 Accident - Cask Drop 8.2.4.2 Accident Analysis - Cask Drop 8.2.4.2.2 Detailed Analysis There is a typo in this section: 55 x 250,000 = 1.375 x 108 With regard to the side drop condition the TSAR is confusing. On pages 8.2-7 and 8.2-22-34 side drops are discussed. In fact, on page 9-8.2-24 three " credible cask loadings in a side drop event" are listed. However, on page 8.2-14, where impact load conditions are discussed, side drop is not included. And on page 8.2-67 three credible impact accident conditions are discussed, side drop is not included. Why not? Why is the side drop case with two limiters included in the TSAR (see pages 8.2 8.2-34), even though it appears that there is only one side limiter, and it is not attached prior to cask handling? What is the basis for assuming that the six axisymmetric loading modes as applied do actually approximate the actual side drop or tip impact loads? 8.2.4.2.2.1 Bottom End Drop Comments made earlier with regard to the 0.6-inch lead slump following an end drop accident apply here as well. 8.2.4.2.2.3 Oblique Drop l On page 8.2-34 the statement is made that the only rotating cask impact (slap-down) condition that may occur is a cask tip over. What will happen to the cask after a corner drop? 8.2.4.2.4 Impact Limiters The NAC states that the upper side impact limiter is installed on the l cask after the cask is placed on the storage pad." Why was the upper impact limiter not modeled in the shielding calculations? i l l l 15
The limiter descriptions are not clear enough. We need a drawing showing how the limiters are attached to the cask body. We need stress-strain data for the limiter materials (in three directions). And we need a discussion of how the stress-strain data were developed. 8.2.4.2.5 Impact Limiter Attachment Are the two layers of honeycomb bonded to the aluminum separator sheet with the same epoxy which is used to join the honeycomb assemblies to the stainless steel shells? Will the epoxy retain its bonding capability for the life of the package under the expected environmental conditions? The impact limiter attachments appear to be designed for normal handling loads. What about accident loads? 8.2.5 Accident - Flood 8.2.5.2 Accident Analysis - Flood 8.2.5.2.2 Structural Analysis With regard to the puncture of the neutron shield shell, the NAC statement "that the shield's effectiveness is not diminished" is incorrect. With puncture, there is some local loss of shielding and a subsequent reduction in effectiveness. What does the statement that " rupture of the shell is not expected, only wrinkling or partial collapse" mean with regard to the neutron shield shell? Is some local loss of the neutron shield possible? 8.2.6 Accident - Cask Seal Leakage 8.2.6.2 Accident Analysis - Cask Seal Leakage 8.2.6.2.2 Boundary Dose 85 3 Whole body dose rates due to Kr and H are low by a factor of 10 because of the error in the atmospheric diffusion factor (X/Q). The errors must be corrected. l 16 l L
Witn regard to the conditions in Table 8.2-41 (Sit.e L(sundary Dose), why S 3 are the Kr and H inventories not listed separately and where are the 3 inhalation rate and the H dose factor? / -8.2.8 Accident - Tornado Missiles 8.2.8.1 Cause of Acccident - Tornado Missiles Even though the massive missile is assumed to deform on impact, there is no indication that some deformation of the neutron shield shell is expe::ted. 8.2.8.2 Accident Analysis - Tornado Missiles 8.2.8.2.1 Analysis - Massive Missile For the tipover impact, the NAC states that "some deformation of the neutron shield shell would be expected," yet no shielding or radiological consequences are considered. Why? With deformation, the local dose rates will increase. For the top end impact, the NAC statement that "the sealing effectiveness of the closure lid is not significantly affected," implies a reduction in -sealing effectiveness. Is the cask leak rate changed? If so, the dose consequences must be addressed. 8.2.8.2.2 Analysis - Penetrant - Missile For the side impact, the NAC states that "the effect of crushing the neutron shield material over a local region eight inches in diameter is discussed in the radiological consequences section." There is no such section, and there is no discussion of the consequences of an eight inch diameter local crush anywhere in the report. 8.2.8.3 NAC S/T Cask Performance - Tornado Missile Accident Why are the shielding and radiological consequences of the various tornado missile accidents not summarized? 8.2.10 Accident - Earthquake 17
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8.2.10.3 NAC S/T Cask Performance - Earthquake Accident Why are the shielding and radiological consequences of the earthquake accident not summarized? J e 8.2.11 Accident - Lightning 8.2.11.2 Accident Analysis - Lightning The analysis presented in this section is incomplete. 8.5 Appendices 85.1 Lead Slump Study The lead slump study in Appendix 8.5 does not provide enough information on the lead model. Why was a " mini model" used instead of a realistic full size model? What mesh was used? 8.5.2 RBCUBED - A Program to Calculate Impact Limiter Dynamics Does the quasi-static computer code which is used to analyze the limiters (RBCUBED) implement the dynamic load factor? 10.0 Operating Controls and Limits 10.1 Proposed Operating Controls and Limits 10.1.2 Bases for Operating Controls and Limits 10.1.2.1 Fuel Characteristic Limits The burnup and radiation source characteristics must acknowledge the spectral dependence as well. The atmospheric diffusion factor in Table 10.1-1 (Operating Controls and Limits) is low by a factor of 10. The error must be corrected. 18}}