ML20205E242
| ML20205E242 | |
| Person / Time | |
|---|---|
| Issue date: | 08/07/1986 |
| From: | Roberts J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| References | |
| REF-PROJ-M-43 NUDOCS 8608180285 | |
| Download: ML20205E242 (46) | |
Text
e-DISTRIBUTION: Please return original concurrence copy to FBrown SS 396 Pygject M 43; PDR
]NMSS r/f FCAF r/f JSchneider M 0 7 W JRoberts LCRouse Project M-43 JSpraul' Combustion Engineering A. E. Scherer Director Nuclear Licensing Power Systems 1000 Prospect Hill Road P.O. Box 500 Windsor, Connecticut 06095-0500 Gentlemen:
This is in response to your letter dated December 19, 1985, submitting your topical report entitled " Topical Safety Analysis Report for the Combustion Engineering Dry-Cap Cask." Our detailed comments are enclosed (enclosure 1).
A separate coment is enclosed with respect to our quality assurance (QA) program review (enclosure 2).
We also have one general comment. There are numerous references to 10 CFR Part 71 requirements in the topical report. We are not reviewing this report under Part 71, but under Part 72.
If you have any questions concerning these consents please call James Schneider (301-427-4205), who has been assigned as contact for the review of your topical report, or call me.
I shall continue to oversee all topical report reviews.
Sincerely OMOfMAL SIGNED Bh John P. Roberts Advanced Fuel and Spent Fuel Licensing Branch Division of fuel Cycle and Material Safety
Enclosures:
1.
Coments 2.
QA Connent n
'g:FC 0FC: FCAF
/
- FCA NAME:JSch eider /jl:J berts :
Rou DATE:08/ 7 /86
- 08/ 9 /86 :08/ 7 /86:
OFFICIAL RECORD COPY 0600100205 060007 PDR PROJ PDR M-43
COM8USTION ENGINEERING DRY-CAP SPENT FUEL STORAGE CASK COMENTS 1 INTRODUCTION AND GENERAL DESCRIPTION OF INSTALLATION 1.1.2 General Description of the Installation Reference is made to the use of a ventilated roofed enclosure for multiple cask storage yet no suppon ting calculations are presented in the TSAR to justify the adequacy of this arrangement.
1.1.5 Description of the Spent Fuel to be Stored (a) The minimum burnup for the fuel should also be specified because assurance of subcriticality depends upon fuel depletion.
The operating history and cooling times for the fuel should also be specified.
(b) The thermal outputs of 0.4 KW per assembly for the BWR fuel and 1.0 KW per assembly for the PWR fuel appear to be too low for a 40,000 MWD /MTU and a five year cooling time.
Provide ORIGIN outputs for decay heat, gamma and neutron spectrum, and rapid nuclide activities for the specified fuel con-ditions.
This set of origin outputs should be used for the thermal, shielding, and containment analyses presented in the TSAR.
- 1. 2 General Description of Installation (a) The criteria for subcriticality of K,ff<0.95 should also reflect the use of fresh fuel and all biases.
(b) Somewhere in this section a description of the cask penetration should be presented.
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1.2.2 Principal Design Criteria 1.2.2.2 Design Criteria for the Cask (a) The cask maximum ambient design temperature of 38'C(100*F) is too low for a credible upper bound applicable to the entire U.S.A.
(b) Is the cask seals maximum leak rate of 10'*cc/sec for each of two seals or the two seals in combim tion? If in combination, what are the two indi-vidual seal leak rates? If for each seal, what is the leak rate of the two seals in combination?
(c) Explain how the difficul es of verifying a maximum leak rate of 10' cc/sec under field test conditions can be overcome.
(d) Radiation dote rate design criteria from 10 CFR Part 71.47 have been cited.
Why are there no references to the radiation dose limits for normal and accident conditions from 10CFR Part 72.67(a) and 72.68(b)?
(e) The radiation dose levels for nuclear operations appear high when ALARA requirements are considered.
(f) What is the basis for the limits of 380*C and 570*C on the fuel cladding for the respective normal and accident conditions? The experience data bases that lead to recommendations for maximum rod temperatures not to exceed 380*C are, as yet, insufficient.
The criteria for acceptance of the maximum clad temperatures is that there be no significant damage to the cladding at the end of storage life.
To assess the acceptability of the stated temperature limit, the applicant should provide a temperature decay curve for both the BWR and PWR fuel for at least a twenty year storage period and the maximum anticipated fuel rod internal pressure.
(g) How long could the cladding be subject to 570*C before cladding rupture occurs?
(h) The expression "shall maintain integrity" on page 1.8 needs to be clarified.
Does this refer to specific limits of stress and deformation, does it refer 1-2
to some level of physical damage or does it refer to some level of radio-active release?
(1) Why is Subsection NB of Section III of the ASME Boiler and Pressure Vessel Code universally applied to the cask and its internal structures? Would it not be more appropriate, for example, to use Subsection NG " Core Support Structures" for the fuel basket?
(j) While Chapter I need not provide stress limits for design, it should discuss the approach adopted for establishing stress limits for normal and accident conditions.
For example, which ASME Code Service Levels will be used for normal and accident conditions?
(k) Is it the intent of C.E. to qualify the storage cask for a thirty foot drop with impact limiters shown in Fig. 1.2-5?
If so, are the impact limiters to remain in place during storage or are they to be removed during some phase of the installation?
(1) What provisions are made to limit the storage cask to a height of one foot above the pad if it is not equipped with impact limiters?
1.2.3 Operating Systems The statement is made that operating systems are not applicable to a spent fuel storage cask.
Are there no systems (pressure transducers, etc.)
attached to the cask that must remain operational?
1.2.5 Structural Features (a) Why do the penetrations described in Section 3.3.2.1 (Confinement Barriers and Systems) not appear in the figures of this section?
(b) Detailed drawings of the BWR and PWR fuel baskets should be provided.
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(c) With respect to Fig. 1.2-1 (Fuel Storage Cask Design), why are there no representation of the BWR and PWR fuel assembly envelopes? Where are the dimensions for the thicknesses of the primary and secondary covers and the cask bottom, the distance from the top of the cask to the top of the radia-tion shield, the distance from the bottom of the radiation shield to the bottom of the cask, the distance from the bottom of the fuel basket to the bottom of the cask cavity, etc.?
(d) With respect to Fig. 1.2-2 (Plan View - Dry Cap B-60), is the component appearing in the upper right hand quadrant of the cask wall the penetration sealing of Fig. 3.3-2?
(e) With respect to Fig. 1.2-3 (Plan View - Dry Cap P-24), the lifting trunnions are diagrammatically different than those in Fig. 1.2-2.
Is there a dif-ference in the lifting trunnions for the BWR and PWR fuel storage casks?
(f) Figures 1.2-1 through 1.2-4 should provide sufficient dimensions to afford the reviewer early on with some conception of the size of the cask and its components.
(g) Where is the weld shown for attaching the cask base to the cylinder?
1.2.9 Safety Related Features (a) Detailed drawings of the cask that completely describe all " safety related" features (use of the term important to safety rather than safety related applies) must be provided. These drawings should show dimensions that are needed to perform the structural, thermal, containment, shielding, and criticality evaluations.
Clearances, gaps and penetrations which are significant for any of the above evaluations should also be shown on the
- drawings, t
(b) Since the pressurized inner lid space and the pressure monitoring system discussed in Section 3.3.2 (Protection by Multiple Confinement Barriers and Systems) and 3.3.2.1 (Confinement Barriers and Systems) are " safety related" features, they should be included in the introduction and general description of tne installation.
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l.3 General Systems Descriptions If the fully loaded cask weights for the BWR and PWR casks are the same it should be so stated.
If not, the actual weight of each type of cask should be given rather than a " conservative" weight.
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2 SITE CHARACTERISTICS Comments presented in Section 1.2.2.2 with respect to the maximum ambient design temperature apply to Table 2.0-1.
2.1 Geography and Demography of Site Selected According to 10 CFR Part 72.68(b), the minimum distance to the controlled area boundary is 100 m (328 ft).
Why is there no mention of this distance here?
Furthermore, based upon the projected dose rates for the BWR and PWR fuel storage casks, what are the minimum controlled area and site boundary limits for a single cask and multiple cask array?
2.3 Meteorology 2.3.4 Diffusion Estimates Cannot some generic estimate be made of the short-term atmospheric diffusion, e.g., Pasquill Type F? This is most important in evaluating the effect of hypothetical accidents involving gaseous activity release at the controlled area and site boundaries.
2.6 Geology and Seismology The statement is made "that the cask would not tip over in a seismic event up to 0.3g load applied simultaneously in a horizontal and vertical direction".
The reviewers note that the standardized requirement of 10 CFR Part 72.66 for all ISFSI sites east of the Rockies is that the ISFSI must be designed to with-stand an earthquake acceleration of 0.25g. This is calculated as two mutually perpendicular components of.25g in a horizontal direction and 2/3 of 0.35g in the vertical direction. This is a more severe requirement than stated in the TSAR and would, it appears, cause tipover.
Consequently, tipover is a credible event and not merely a conservative assumption.
The analysis for response to 2-1
earthquakes in the TSAR should reflect this both in Chapter 2 and Chapter 8.
Furthermore, Table 2.0-1 indicates that loading above 0.3g would require the tie-downs to prevent tipover implying that such tie-downs would prevent tipover during a design earthquake of 1.0g simultaneously in the horizontal and vertical direction.
Yet, no analysis for this condition appears in the TSAR.
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3 PRINCIPAL DESIGN CRITERIA 3.1 Purposes of Installation In this section it is stated that " keeping the fuel cladding intact is an impor-tant design criteria when considering future handling of the fuel". Where in the TSAR is the analysis that shows that this criterion is met?
3.1.1 Materials to be Stored (a) The active fuel length for the PWR assembly in Table 3.1.1-1 is 168 inches while in Section 1.1.5.3 and Table 3.3-4 it is shown as 144 inches. Which is correct?
(b) Why is there no indication of the amount of heavy metal (U) per assembly?
If the neutron source in Section 7.2 (Radiation Sources) is to be expressed in terms of MTU, the amount of heavy metal per assembly must be presented.
(c) What is the minimum burnup associated with the PWR fuel assembly in Table 3.1.1-2 (Characteristics of Stored Fuel)? Comments with respect to the amount of heavy metal (U) per assembly apply here as well.
(d) Is the value for minimum burnup for PWR fuel in Table 3.1.1-2 left blank intentionally?
f 3.1.2 General Operating Functions
. In this section it is stated that "the storage cask functions without the use j
of an active operating system." Is not the pressure monitoring system discussed l
in Section 3.3.2.1 (Confinement Barriers and Systems) considered active?
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3.2 Structural and Mechanical Safety 3.2.1 Tornado and Wind Loads (a) Provide an analysis to demonstrate the resistance of the cask to tipping under extreme wind speeds citing applicable formulas, values for the param-eters and references. What drawing provides information concerning the L/D ratio? A free body sketch showing the geometry and loading configura-tion would be helpful.
(b) Where in the TSAR is the evaluation of the effect of the tornado missile impacts?
3.2.2 Water Level (Flood) Design (a) Where is the analysis for submergence?
(b) Would a water velocity required for tipover be less than 18.0 mph if buoy-ancy effects were considered?
3.2.3 Seismic Design Is it possible that loaded internal structures of the cask might be less than 33HZ so that they are vulnerable to amplification due to ground acceleration?
l 3.2.5 Combined Load Criteria i
(a) Provide a table summarizing the actual stress intensity limits for all structural materials used in the cask, for all load combinations, and for normal and accident loading conditions.
(b) The limits for primary membrane stress under accident conditions are the lesser of 2.4Sm or 0.75u, not the greater.
(c) Do the limits in Tables 3.2.5-1 and 3.2.5-2 refer to all bolts used on the cask? It should be noted that trunnion bolts are lifting devices.
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3.3 Safety Protection Systems 3.3.2 Protection by Multiple Confinement Barriers and Systems Since the pressurized annulus between the primary and secondary covers is a feature of the safety protection system it should be described in Chapter I as part of the general description of the installation.
3.3.2.1 Confinement Barriers and Systems (a) In this section it is stated that "there are four other penetrations through the cask body; the drain hole, the gas input / sampling penetration, the penetration to pressurize between the primary and secondary lids, and the penetration for monitoring the pressure between the covers." These pene-trations should be graphically described in the figures of Section 1.2.5 and not left to be discovered this far into the TSAR.
Furthermore, this is also the first indication that a pressure monitoring system is a feature of the cask.
This should also be described earlier in the TSAR.
(b) With respect to Fig 3.3-1 (Cask Seal Detail). Fig. 3.3-2 (Penetration Seal-ing), and Fig. 3.3-3 (Cask Seal Detail Optional Seal Ring), the first and third figures show no dimensions and the second figures' dimension are not adequately shown.
In addition, the primary seal spring is not indicated and it is not clear whether the primary seal plate is threaded in Fig.
3.3-2.
Moreover, even though Fig. 3.3-2 appears to apply only to the drain hole, is it implied by the text that all penetrations are exactly the same in construction?
(c) With respect to the replacement of a leaking secondary seal or the instal-lation of a welded seal ring in the event of a leaking primary seal, where are these repair operations performed?
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3.3.3 Protection by Equipment and Instrumentation 3.3.3.1 Seals (a) It would be helpful if Fig. 3.3-4 (Seal Cross-Section) would show the actual dimensions.
(b) It is stated in this section that "the seals are designed to permit a leak-age value of less than 0.001 A2/ week." Where is the justification to sup-port this claim? It should either be provided in this section or in an appendix.
Furthermore, 10 CFR Part 71 requirements are not applicable to storage casks.
3.3.3.2 Instrumentation How is a failure of the pressure transducer detected? Is there any redundancy in the pressure monitoring system in the event of transducer failure? How is a failed transducer repaired?
3.3.4 Nuclear Criticality Safety 3.3.4.1 Control Methods for Prevention of Criticality l
(a) Prevention of criticality for the storage cask relies on fuel burnup and U-235 depletion.
For each initial enrichment of PWR fuel clearly specify the nominal burnup expected, the minimum burnup limit and the fissile mate-rials expected to be present.
Considering that only a few utilities are currently licensed for handling and storing spent fuel with burnup credit included, provide the following information:
1.
A discussion on the benefits, limitations, and potential hazards in using credit for burnup.
Justify the use of a neutron multiplication factor of.95 limit as having a sufficient safety margin given the additional uncertainties involved with burnup.
l 2.
Specify what administrative controls are required.
Does a burnup meter have to be used? Discuss how a utility assures that specific 3-4 l
operating histories and minimum burnups of specific fuel bundles and rods are achieved.
3.
If burnup credit is required to prevent criticality, discuss how the fuel will be transported on public highways and later disposed. Will the fuel have to be transferred to another cask for transport? This issue should be addressed.
(b) The number of groups used in the cross-section calculation appears to be inadequate for licensing ccmplex cask geometries.
The codes used must be benchmarked against criticality experiments that are similar to the CE dr:y-cap cask geometry. The benchmark model and results must be included in the TSAR.
(c) Detailed drawings showing all the materials and dimensions must be provided to enable the reviewer to perform a confirmatory analysis.
(d) Computer inputs and outputs for the minimum burnup should be provided since this is the most limiting condition.
Provide a description of how the burnup and criticality calculations are made and identify all radionuclides that are used in the calculations.
Provide information on whether credit was taken for poisoning effects of the fission products and if the effect of plutonium buildup was included.
(e) It is stated in this section that "the fuel stored in the Combustion Engi-neering dry storage cask will not achieve criticality if (1) each assembly satisfies the limitations on initial enrichment and achieved depletion specified herein, and (2) each fuel assembly is kept at least that minimum distance from its neighbors provided by the storage basket walls." Is j
there any accident condition where condition 2 could be violated?
l 3.3.4.1.2 Cask and Basket Description (a) Dimensions should be shown in Fig 3.3-5, Fig 3.3-6, Fig. 3.3-7, and i
Fig 3.3-8.
Moreover, why are the components in Fig. 3.3-5 and Fig. 3.3-7 not identified by number and associated legend?
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(b) Section 1.3 does not provide a description of the CE storage cask and internal structure.
3.3.4.1.3 Fuel Parameters (BWR)
(a) The fuel rod clad material in Table 3.3-1 is described as zircalloy-4 for BWR spent fuel while in Table 3.1.1-1 it is ZR-2.
Which is correct?
(b) The fuel assembly cross section in Table 3.3-1 is g ven as 5.278 inches square while in Table 3.1.1-1 it is 5.47 inches. Which is correct?
3.3.4.1.3~ Fuel Parameters (PWR)
The fuel pellet diameter, in Table 3.3-4 is 0.3225 for the PWR while in Table 3.1.1-1 it is 0.3088.
The fuel rod active length in Table 3.3-4 is 144 while in Table 3.1.1-1 it is 168.
The fuel rod diameter in Table 3.3-4 is 0.374 while in Table 3.1.1-1 it is 0.360.
Which dimensions are correct?
3.3.4.1.4 DOT Geometry Models -(BWR)
(a) Dimensions should be indicated on Fig. 3.3-9.
Furthermore, the components in Fig 3.3-9 should be identified.
l (b) Should not the reference to Fig 3.3-6 on page 3.25 actually be Fig 3.3-5?
I 3.3.4.1.4.1 00T Geometry Models (PWR)
(a) Dimensions should be indicated on Fig. 3.3-10.
Furthermore, the components in Fig. 3.3 10 should be identified.
(b) Should not the reference to Fig. 3.3-8 on page 3.26 actually be Fig. 3.3-7?
3.3.4.1.5 Results of Criticality Analyses (a) There is an inconsistency between the text on pages 3.26 and 3.27 and the tables and figures referenced therein.
The text only addresses BWR while 3-6
the tables and figures address both PWR and BWR.
A further inconsistency occurs in the discussion of the limit on the maximum calculated K,ff.
In one paragraph the discussion does not distinguish the fuel type while in another only BWR is specifically addressed. Moreover, Section 3.3.4.1.2 (Cask and Basket Description) indicates that the limit on the maximum calculated K,7f was enforced for both BWR and PWR fuels.
These discrep-ancies need to be clarified.
(b) With respect to Table 3.3-2 (Summary of D0T Results (BWR)), are the values of K,ff for the " infinite array" or are they " cask equivalent" values?
Whichever values of K,ff they are, why are the other values not summarized as well? Where is the result for the design basis fuel of 3.5% enrichment and 40,000 MWD /MTU burnup?
(c) With respect to Table 3.3-5 (Summary of DOT Results (PWR)), are the values of K,ff for the " infinite array" or are they " cask equivalent" values?
Whichever values of K they are, why are the other values not summarized eff as well? The design basis fuel result (4.0% enrichment and 40,000 MWD /MTU burnup) is not acceptable.
Even assuming a typing error the value of K,ff appears to be above the acceptable limit of 0.95 at the 95 percent confidence level.
(d) With respect to Fig 3.3-11 (Burnup Limits for Storage BWR) and Fig.- 3.3-12 (Burnup Limits for Storage PWR), the results depicted cannot be verified from Table 3.3-2 and Table 3.3-5, respectively.
3.3.4.1.6 Accident Analysis Where is the analysis that leads to the determination that there is no cask drop that will alter the internal arrangement of the fuel storage basket?
3.3.4.2 Error Contingency Criteria Should not the reference to Fig. 3.3-9 actually be Fig. 3.3-11?
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3.3.5 Radiological Protection 3.3.5.2 Shielding (BWR)
(a) Where is the peak surface dose rate on the cask side located? Are the doses at one and two meters also at the cask side? If so, where are they located? Where are the results for the top and bottom of the cask at the surface and at 1 and 2 m?
(b) The radiation dose rate information presented in this section is not suf-ficient to determine the occupational dose due to cask handling. The dis-tances from the cask surface to the body trunk in Table 5.1-2 vary from 2 to 4 to 5 ft in air and 2 ft in water.
The distances at which the dose rates are reported in this section are 3.28 ft (1 m) and 6.56 ft (2 m) in air.
Furthermore, some of the operations of Table 5.1-2 involve exposure at the top of the cask, yet no dose rates are reported at this location.
How can the doses at 2, 4, and 5 ft in air and 2 ft in water be determined from the information supplied in this section? Much more detailed informa-tion must be presented to validate the reported dose due to cask handling.
(c) Why is there no indication of the expected annual dose at the controlled drea and site boundary limits for a single cask and a multiple cask array?
Some discussion of the expected annual doses should be provided.
3.3.5.2.1 Shielding (PWR)
Comments presented in Section 3.3.5.2 dealing with BWR shielding apply here as well.
Furthermore, why is there no discussion at all of the dose due to cask handling in this section?
3.3.6 Fire and Explosion Protection l
3.3.6.1 Fire Section 8.2.11 (Fire) shoulo be specifically referenced in the discussion of i
the detailed analysis in this section.
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3.3.6.2 Explosion (a) It is stated in this section that "the cask body and lids are constructed of thick-walled steel plates and could withstand pressure loadings caused by an explosion of well over 1,000 psi."
In Section 4.8.5 (Cask (External Pressure Limits)), the maximum allowable external pressure on the secondary cover is only 62 psi.
This discrepancy should be clarified.
(b) Section 8.2.2 (Cask Tip-0ver) should be specifically referenced with regard to the ability of the cask to withstand tip-over.
3.3.7 Materials Handling and Storage 3.3.7.1 Spent Fuel Handling and Storage The criteria presented in this section for the placement of spent fuel into the cask are incomplete.
If fuel assemblies with characteristics other than those of the design basis BWR and PWR fuels are to be stored, the effect of the speci-fic differences must be addressed in sufficient detail to validate a conclusion of no increase in consequences to the general public.
3.5 Decommissioning Considerations I
I (a) How was the neutron flux determined? Was it assumed to be constant throughout the cask wall? How were the a:tivation product activities cal-culated? Sample calculations should be provided.
l (b) It is stated in this section that "there is only one long half-life material produced in the cask body, and that is ssMn." The half-life of ssMn is I
2.58 h.
Was not the isotope 54Mn meant with its half life of 312.5d?
l (c) From the material components appearing in Tables 7.3-4 and -5, it is evident that a number of activation products with half lives in excess of 27 days may be produced.
Included among the possible activation products 1
i are 51Cr, 54Mn, ssesspe, soCo, and 59'88Ni.
Why are the activities of these activation products not summarized?
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(d) The discussion of the production of soCo in the cask internal structures is inadequate. An activity in the 10 4 pCi range is reported.
- However, there is no indication of the flux or the amount of the contaminant in the stainless steel.
(e) To justify the statement, "the decommissioned cask...can be shipped to a low level burial site," requires a discussion relative to 10 CFR Part 61 (Class A, B or C).
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4 INSTALLATION DESIGN 4.2 Storage Structures 4.2.3 Individual Unit Description Drawings should be provided which illustrate the description of the individual units that are contained in this section.
4.2.3.1 Cask Body (a) Type SA-516 grade 55 steel is ferritic and subject to a transition with temperature from ductile to brittle behavior.
The applicant should address the subject of resistance to brittle fracture under dynamic loads for this I
component.
It should be pointed out that if this cask is intended to be used for transporting radioactive spent fuel the acceptability of the specified cask body material should be discussed with NRC's Transportation Certification Branch.
This review does not cover offsite transportation which is covered under 10 CFR Part 71.
If this cask is not intended for transportation of spent fuel, all references to such a dual function should be deleted from the TSAR.
(b) What specification controls the fabrication of the welded cask body?
4.2.3.2 Fuel Basket (a) While type 405 stainless steel is similar in heat and corrosion resistance to comparable types of stainless with the same chromium range, the lower carbon content and addition of aluminum results in an alloy that will cool from welding temperatures without marked hardening.
However, even though the weld zone is not hard, the grain growth at the high welding temperature results in a large reduction in toughness.
Therefore, under dynamic loads, imperfections in the weld area may cause fracture in a brittle manner and 4-1
collapse of the basket.
If the applicant insists upon using this material for the basket, it is mandatory that it be qualified against brittle frac-ture under the anticipated loading environments.
This requires, at least, fracture toughness properties of the weld and heat affected zones and a failure analysis that demonstrates resistance to brittle fracture with a high level of confidence.
(b) What specification controls the fabrication of the welded basket?
4.2.3.5 Trunnions The safety factors applied to the trunnion materials are as stated in ANSI-N14.6.
However, N14.6 has the additional requirement that for a non-redundant system thc factors are twice those stated in the TSAR.
4.2.3.7 External Radiation Shield (a) How is the addition of the radiation shield and shell accomplished? Is a preformed radiation shield attached to the cask body and then enclosed in a welded steel shell, or is a welded steel shell attached to the cask body and then the radiation shield material added as a liquid and allowed to harden? If the former is correct, will welding of the steel shell cause local deformations in the radiation shielo material? If the latter is correct, what assurance is there that voids will not form in the radiation shield material during hardening?
l (b) What type of steel is used in the radiation shield shell?
1 4.3 Auxiliary Systems 4.3.9 Maintenance Systems l
(a) Means for monitoring the inter-lid pressure is provided by installation of l
a pressure transducer and an appropriate alarm system. What provisions are made to assure the operability of this system throughout the life of the ISFSI?
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(b) The location should be identified where replacement of the secondary seal and welding of the seal ring are accomplished.
(c) How is the welded seal ring removed if the cask is unloaded subsequent to this fix?
4.5 Shipping Cask Repair and Maintenance What is involved in the " replacement or servicing" of the pressure transducer?
4.8 Cask and Components - Normal Operation Analyses 4.8.5 Cask (External Pressure Limits)
(a) Comments presented in Section 3.3.6.2 (Explosion) on the maximum allowable external pressure on the secondary cover apply here as well.
(b) Where are external pressure limits for the radiation shield and radiation shield shell discussed?
4.8.6 Internals and Cruciform Loads A drawing should be provided that shows the details of the lifting arrangement for the cruciform.
The analysis should be augmented with a sketch of the load-i ing configuration.
4.8.7 Trunnion Analysis (a) The trunnion analysis should reference a drawing that completely describes the trunnion.
Figure 4.8-1 is inadequate.
Here again, the analysis should be augmented with a sketch showing the loading configuration.
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(b) The analysis is based upon the assumption of a redundant system wherein the allowable stresses are the lesser of 1/3 yield and 1/5 ultimate.
If the system is not redundant, the allowable stresses should be the lesser of 1/6 yield and 1/10 ultimate.
(c) What material is used for the trunnion? What are its mechanical properties?
It would be helpful if the allowable stresses were summarized in a table.
(d) Provide complete bibliographical information for all cited references.
(e) Substantiate the assumption that the bearing stresses on the cask are uniformly distributed over the bearing area.
(f) The relevance of 10 CFR Part 71 to the storage cask needs to be explained.
4.8.8 Cask Thermal Performance 4.8.8.2 Thermal Analysis (a) With respect to Table 4.8.3 (Mechanical and Thermal Characteristics of Cask Loaded with PWR Fuel), why is the fuel rod emissivity based upon a different material than the material of the clad? More importantly, why is the PWR thermal analysis based upon a 15X15 assembly instead of the design basis 17x17 assembly? With no specific mention of the BWR and PWR fuel assembly types in the text, the implication is that the design basis assemblies were used. This should be clarified.
(b) The statement, "the maximum clad temperature with the cask in the vertical storage position will be lower than the horizontal value since the (verti-cal) configuration will allow natural circulation of the gas," implies that there is no natural circulation of the gas in the horizontal position.
Yet, in Section 4.8.8.3 (Thermal Model), it is stated that, " natural con-vection is small in the enclosure of a cask in the horizontal position."
These statements should be reconciled.
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(c) As noted earlier, the decay heat of 0.4 KW for BWR fuel and 1.0 KW for PWR fuel appear to be too low for 40 MWD /MTU burnup and five years cooling time. The analysis referred to that shows these values to be conservative should be included in the TSAR.
(d) Drawings of the thermal models should be provided showing all materials and significant dimensions and clearances including tolerances.
(e) Copies of the ANSYS code inputs and outputs for the thermal analysis should be provided.
Does the ANSYS model assume uniform radiosity for radiant heat transfer? Show how the shape factors were derived and list them.
(f) Were only single values used for the thermal conductivity and emissivity for each material in the thermal analysis? If so, provide justification.
(g) Where are the containment analyses?
(h) Is the neutron shielding material qualified to provide its safety function for at least twenty years at elevated temperatures?
(i) Table 4.83 cited in this section should be 4.8-3.
(j) How was the information plotted in Fig 4.8-3 obtained?
(k) The finite element model for heat transfer is shown in Figs. 4.8-4 and 4.8-5.
The mesh representing the fuel assemblies and the basket channels should be shown.
4.8.8.3 Thermal Model i
(a) With respect to Fig. 4.8-6 (2-0 Model of Fuel Assembly (BWR)) and Fig. 4.8-7 (2-D Model of Fuel Assembly (PWR)), provide dimensions for the figures.
(b) The discussions in this section indicates that the finite element of the model of the cask (for BWR fuel) has a much higher number that illustrated in Fig. 4.8-4, only 70.
4-5
i 5 OPERATION SYSTEMS 5.1 Operation Descriptions While it is true that the detailed operating procedures for plant personnel will be written by the ISFSI applicant, those procedures will be based in large part on what is available in this TSAR.
It is important, therefore, that this chapter provide sufficient details on handling operations.
For example, para-graph (1) of Section 5.1.1.1 " Initial Cask Handling Operations" says: " unfasten the hold down rigging, and using trunnions, lift the cask from the transportation s kid. " This paragraph needs reference to a sketch which identifies exactly which rigging to unfasten, how to attach a crane to the trunnions, which trun-nions to use (two or four), whether to set it down vertically or horizontally, etc. This might be brief, but it should be complete.
5.1.1 Narrative Description The requirement to " describe the means that will be routinely used during storage to evaluate the condition of the" cask (s) is not met.
l 5.1.1.4 Fuel Removal I
l (a) The introduction to this section indicates that the cask will have been shipped to a facility with the capacity to handle a wet removal operation or a dry handling facility.
Since the DRY-CAP cask is not licensed as a shipping cask, this paragraph should be rephrased.
l (b) What is the significance of " slowly" increasing water flow rate into cask?
What are the consequences of doing this too quickly?
(c) Paragraph (5) indicates that a " prescribed amount of water" should be introduced into the cask. This instruction should be more specific.
l 5-1
(d) Paragraph (22) states that depending on the residual radiation level of the cask and internals, both will be able to be reused for storage of additional fuel assemblies. What are the criteria for this action?
(e) Section 8.1.1.1.4 lists corrective action for possible seal leaks. Two actions listed are installing a welded seal ring to primary cover, and polishing sealing surfaces of secondary seal.
How long will these tasks take?
(f) With respect to procedure 2, is the internal gas from the cask simply vented to the surroundings?
If so, why are there no precautions described?
5.1.2 Flow Sheets (a) Organize Table 5.1-1 (Cask Handling Flow Sheet), so that the various pro-cedures follow the narrative description; e.g., cask handling, fuel loading, storage, and fuel unloading.
(b) How were the times and distances in Table 5.1-2 (Tasks Requiring Operator Radiation Exposure) determined? Are they conservative? For those pro-cedures requiring more than one man, how many men are actually required?
Comments presented in (a) with respect to Table 5.1-1 apply here as well.
5.1. 3 Identification of Subjects for Safety Analysis 5.1.3.4 Instrumentation (a) The need to discuss the " testability, redundancy, and failure conditions" of the pressure transducer is not met.
Whether Combustion Engineering considers the pressure transducer to be a " safety" system or not, such information is relevant to the requirements of this section.
(b) It is stated in this section that "the cask is at a negative pressure and a leaking seal would only allow helium to enter the cask." This not com-pletely accurate.
A leaking secondary seal would vent the interspace helium to the environment.
This should be clarified.
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(c) It is also stated that, "unless the cask sustained a severe drop accident, the fuel cladding would be intact." This implies that the fuel assem-blies will not remain intact as a result of a severe drop accident? Fur-thermore, it is stated that " particulate matter would not pass through the seal unless there were a leak and there would be no event to create this leak other than a dropping accident." Does this imply that the damage to the fuel assemblies may be sufficient to generate particulate matter and that the particulate may be released to the environment? There is no indication of any such possibilities in Section 8.2.1 (Cask Drops).
If such possibilities exist, they should be addressed.
5.1.3.5 Maintenance Techniques (a) What is the indication that the pressure transducer needs to be replaced?
What are the procedures for doing so?
(b) Comments presented in Section 4.5 (Shipping Cask Repair and Maintenance) with respect to repainting apply here as well.
5.2 Fuel Handling Systems (a) Spent fuel storage operations that include placement, storage surveillance, and removal should be summarized.
Reference to and use of materials in Section 5.1.1 (Narrative Description) and Section 5.1.2 (Flow Sheets) is appropriate here.
(b) Any features important to safety for the spent fuel storage system that provide for safe operation under both normal and off-normal conditions should also be summarized.
So too, should any limits that are selected i
for a commitment to action.
5.3 Other Operating Systems Where is the ru ponse to the component and equipment spares issue? The need to " describe in detail design features that include installation of spare or alternative equipment to provide continuity of safety under normal and abnor-mal conditions" is not met.
Much of the information in Section 5.4.2 (System 5-3
and Component Spares) seems more appropriate to this section.
Reliability of the pressure monitoring system is also of interest here.
Design provisions to minimize exposure to radiation during repair operations should also be addressed.
5.4 Operation Support Systems 5.4.1 Instrumentation and Control Systems (a) We do not agree with the absolute statement that "an indication of a leak-ing cask is not a safety issue." There are circumstances where a leaking cask may have radiological consequences.
(b) The discussion of the seal repairs seems more appropriate in connection with the component and equipment spares issue of Section 5.3 (Other Operat-ing Systems).
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(c) Monitoring during normal, off-normal, and accident conditions, and redun-dancy of safety features should be addressed in this section.
5.4.2 System and Component Spares l
The requirement to " describe in detail the installation of spare or alternative instrumentation designed to provide continuity of operation under normal and abnormal conditions" is not met.
As mentioned earlier, much of the information presented here is more appropriate to Section 5.3 (Other Operating Systems).
Reliability of the pressure monitoring system is also of interest here as well.
5.6 Analytical Sampling I
l It is stated in this section that, "there are two occasions where analytical sampling would be advisable.
These are: after the cask has completed a storage cycle and is going to be opened, and after an accidental cask drop." Why is there no discussion of the occupational dose consequences to operating personnel?
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5-4
a 7 RADIATION PROTECTION 7.1 Ensuring that Occupational Radiation Exposures Are as Low as Is Reasonably
~
Achievable (ALARA) 7.1.1 Policy Considerations The requirements of this section are, for the most part, site specific.
In addition, Regulatory Guides 8.8 and 8.10 center on management policies, adminis-trative procedures, and design features associated with an installation.
How-ever, discussion of the policies during periodic inspection / surveillance and any maintenance operations is in order.
Indicate how the guidance of (or alternatives to) the regulatory guides will be followed.
7.1.2 Design Considerations (a) With respect to design consideration 1, should not Section 1.2.2.2 design limits of less than 10 mres/hr at 2 m for normal operations and less than 1000 mrem /hr at the surface following an accident also be addressed? What about the radiation dose limits for normal and accident conditions from 10 CFR Part 72.67(a) and 72.68(b)?
(b) With respect to design consideration 2, since both burnup and decay affect the neutron and gamma spectra, the effect on the dose due to any differences between the design basis fuels must be addressed.
~(c) Does design consideration 3 imply that radioactive contaminants are present within the cask? If so, these contaminants should be addressed as either particulate or gaseous sources.
(d) With respect to design consideration 4, once the pressure has stabilized, could gases escape due to pressure buildup from a fire?
i 7-1
(e) With respect to-design consideration 7, where are the dose calculations that validate the claim of "almost no personnel radiation exposure for maintenance activities?"
(f) Indicate how the guidance in (or alternatives to) regulatory position 2 of Regulatory Guide 8.8 will be followed.
It is suggested that Combustion Engineering address each section of regulatory position 2 separately and sequentially.
Those sections that are site specific or irrelevant can be so identified.
Those sections that are relevant should be addressed.
7.1.3 Operational Considerations Supply a description of the methods used to develop the detailed plans and procedures for ensuring exposure is ALARA and the criteria and conditions under which various procedures and techniques are implemented for ensuring that exposure is ALARA.
The information presented is insufficient.
7.2 Radiation Sources (a) What code was used to calculate the neutron and gamma radiation sources, and what are the irradiation conditions associated with the calculations?
(b) With respect to Table 7.2-1 (Principal Isotopes in Gamma Ray Source Term, 5 Years Decay, 40,000 MWD /MTU), why are the source strengths in MeV/ Watt-sec instead of photons /sec? Are the results based upon one assembly or one full cask? Why are the BWR and PWR gamma source distributions not expressed in terms of the gamma group structure of Table 7.3-8 (Multigroup Energy Group Structure from DLC-23E)?
l (c) With respect to Table 7.2-2 (Combustion Engineering Dry Storage Cask Neutron Source Term, 5 Year Decay (BWR), 40,000 MWO/MTU), why not express the neutron source term on the same basis as the gamma source, i.e., per assembly or per cask? Where is the neutron source term for the PWR assemblies? Why is the BWR neutron source distribution not expressed in terms of the neutron group structure of Table 7.3-8 (Multigroup Energy Group Structure from DLC-23E)?
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(d) Where are the gamma radiation source terms for the activation products in the steel of the head and foot pieces?
(e) It is requested that "the sources of radiation...be described in the manner needed as input to the shielding design calculations." Both in this sec-tion and Section 3.1.1 (Materials to be Stored), the information presented is inadequate.
(f) Where are the sources of airborne radioactive material? Is there no con-dition or circumstance where leakage of gaseous fission products from the cask cavity is possible? Furthermore, how can compliance of the seals be verified without airborne radioactive material source terms?
- 7. 3 Radiation Protection Design Features 7.3.1 Installation Design Features (a) The statement, "the C-E ORY-CAP will meet the design requirements associated with 10 CFR Part 72, 10 CFR Part 20, and the guidelines of Regulatory Guide 3.48," is too general.
To which specific design requirements do they refer?
(b) Since the storage is AR and there are no structures, many requirements are negated.
Specific activity, physical and chemical characteristics, and expected concentrations of all sources in Section 7.2 (Radiation Sources) are to be provided. Other information needed includes the design radia-tion dose rate for the storage area and the maintenance and repair activity, and estimates of the radioactive materials that might be discharged during l
storage.
Reference to specific sections of the TSAR where this information may be found would be acceptable.
Indicate how the guidance in (or alter-l natives to) regulatory position 2 of Regulatory Guide 8.8 will be followed.
i 7-3 i
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, -.. ~ - -
7.3.2 Shielding 7.3.2.1 Radiation Shielding Design Features (a) With respect to Table 7.3-1 (Design Parameters for the Combustion Engineer-ing Dry Storage Cask with Intact Spent Fuel Assemblies), why is the neutron shield and snell not included in the overall diameter of the cask? Moreover, why is the word " neutron" now used in place of the word " radiation" in the description of the shield and shell? Why are the. thicknesses of the wall, bottom, and secondary cover inconsistent with the specifications in Sec-tion 4.2.3.1 (Cask Body) and Section 4.2.3.4 (Secondary Cover)? The metric equivalents of the primary and secondary cover thicknesses appear to be in error.
In general, descriptions and dimensions of cask components through-out the TSAR should be consistent.
(b) With respect to Table 7.3-2 (Characteristics of BWR Spent Fuel), why are the fuel rod clad material and fuel assembly cross section inconsistent with the information in Table 3.1.1-1 (Design Basis Fuel)? The metric equivalent of the active fuel length appears to be in error.
(c) With respect to Fig. 7.3-1 (Geometry Model 45* Sector (BWR)) and Fig. 7.3-3 (Geometry Model 45* Sector (PWR)), why aren't legends provided to identify the various areas as done for Figs. 7.3-2 and 7.3-47 l
(d) With respect to Fig. 7.3-2 (R-Z Geometry Model of the Upper Half of Dry Storage Cask - BWR) and Fig. 7.3-4 (R-Z Geometry Model of the Upper Half of Dry Storage Cask - PWR), what is zone 2? The term " outer wall of steel monolith" is confusing.
How was the thickness for zone 8, the " fuel assem-bly upper end steel structure," determined? Is the material in this zone made of carbon or stainless steel? There is no description of the " fuel zone for the gas plenum region." Where are the radial dimensions in each figure? From Table 3.3-1 (Characteristics of BWR Spent Fuel), the BWR fuel assembly length is 447 cm.
Sitting on the bottom of the cask, the top of the assembly would be at an elevation of 224.1 cm. According to Fig. 7.3-2, the top of the assembly is at an elevation of 184.34 cm.
To what is this large difference due? From Table 3.3-4 (Characteristics of i
7-4
l PWR Spent Fuel), the PWR fuel assembly length is 406 cm.
Sittirg on the bottom of the cask, the top of the assembly would be at an elevation of 183.5 cm.
According to Fig. 7.3-4. the up of the assembly is at an elevation of 185.06 cm. Allowing only 1.56 cm for a 1.27 cm sunport plate and void does not seem consistet:t with the drawing in Fig.1.2-1 (Fuel Storage Cask Design).
7.3.2.2 Shielding Analysis (a) With respect to Table 7.3-4 (Material Spccifications and Properties (BWR))
and Table 7.3-5 (Material Specifications and Properties (PWR)), the cask body and end shields, neutron shield canting, and fuel storage rack and cruciform carbon steel appear to have the same composition.
The first items are made of carbon steel, the second is made from an unspecified steel, and the third ate made from stainless steel. We want the composition of each component presented.
(b) Why is the fuel rod clad material in Table 7.3-4 inconsistent with the information in Table 3.1.1-1?
(c) What is the last entry for the composition of the U0 fuel in Table 7.3-5?
2 (d) Define the fission products that account for 3.89% of the composition of the fuel pellets.
(e) With respect to Table 7.3-6 and Table 7.3-7, how were the various atom densities in region number 1 computed? Volume fractions and their deriva-tions must be presented.
Since Mn, Ni, Cr, etc., were shown in Tables 7.3-4 and 7.3-5 for the carbon steel cruciform and storage rock, where are the atom densities for these elements in region number 1?
(f) Is the steel in region number 8 carbon or stainless?
(g) As indicated under Section 7.2 (Radiation Sources), where are the neutron and gamma source distributions that correspond to the group structures in Table 7.3-6 (Multi group Group Structure from DLC-23E)?
7-5
7.3.2.2.1 Shielding Analytical Methods One cannot assume results calculated for the cask upper end are applicable to the bottom.
This is not correct as the active fuel region will be closer to the cask bottom than the top.
7.3.2.2.2 Shielding Results (a) Why are the dose rates calculated at 1.5 cm from the cask surface and not at the cask surface? The design limit of 200 mrem /hr is supposed to be at the surface.
(b) With respect to Table 7.3-11 (External Surface Dose Rate Contributions, 5 Year Decay, 40,000 MWD /MTU), where are the peak dose rates located?
Furthermore, where are the results for the cask bottom and where are tho results at 2 m from the cask surface?
(c) With respect to Fig. 7.3-6 (Neutron Dose Rate - BWR), Fig. 7.3-8 (Neutron Dose Rate - PWR), Fig. 7.3-11 (Top Surface Gamma Dose Rates - BWR), and Fig. 7.3-13 (Top Surface Gamma Dose Rates - PWR), where is the dose point?
(d) With respect to Fig. 7.3.-7 (Neutron Does Rates Through Top Shield - BWR) and Fig. 7.3.-7 (Neutron Dose Rates through Top Shield - PWR), what are the units of the abscissa?
(e) How was the neutron subcriticil multiplication included in the neutron dose rate calculations?
i (f) There should be calculations of the dose rate for the cask bottom.
See Section 7.3.2.2.1 (Shielding Analytical Methods) above.
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=
7.4 Estimated On-Site Collective Dose Assessment Why does this section not appear in the TSAR?
7.5 Health Physics Program Why does this section not appear in the TSAR?
7.6 Estimated Off-Site Collective Dose Assessment Why does this section not appear in the TSAR?
\\
7-7
8 ACC'IDENT ANALYSIS 8.1 Off-Normal Operations List off-normal events that are applicable to the storage cask.
8.1.1 Event Why is a, failure of the pressure monitoring system transducer not considered here?
8.1.1.2 Increased and Reduced External Pressure 8.1.1.2.3 Analysis of Effects and Consequences Where is the analysis?
8.1. 2 Radiological Impact from Off-Normal Operations Why does this section not appear in the TSAR?
8.2 Accidents i
8.2.1 Cask Drops (a) The requirements of 10 CFR 71 are not necessarily relevant to a storage cask. Where are the cask drawings that may be used for the accident analysis?
(b) The statement is made that the cask weight used in the analysis differs slightly from the actual cask weight. What weight was actually used?
8-1
8.2.1.1 Cause of Accident The statement is made that shock absorbers will be in place on the cask and trunnions for lifts greater than one foot. Why is this not mentioned in the operation systems section of the TSAR? How are the hooks applied to the trunnions with the shock absorbers in place?
8.2.1.2 Accident Analysis 8.2.1.2.1 0.3 m (1 ft) Drop without External Shock Absorbers (a) Provide more information about the computer code CESHOCK.
Has it been benchmarked? What types of elements are used? What types of material models are used? What mechanical properties were used for all materials?
Provide a free body diagram showing loading on bolts, covers, and fuel.
Provide stress levels (principal values preferably) at all critical loca-tions on the cask.
How were the fuel loads determined? Does the 59,700 psi axial load cause a buckling concern in the fuel? Explain why the F.E.
model in Fig. 8.2-4 corresponds to a corner drop.
A sketch showing what mass goes where might help.
(b) Provide the spring stiffnesses and a description of the mass distribution for the models, and discuss how these values were determined. The develop-ment of the models shown is far from trivial, especially for the corner drop model.
(c) We do not agree that the most conservative cask side drop necessarily occurs with the cask side and not the trunnions hitting the impact surface.
This assumption should be substantiated.
(d) What happens to the radiation shield and shell following the three drop configurations analyzed?
8.2.1.2.2 9 m (30 ft) Drop with External Shock Absorbers (a) Describe in greater detail the analysis that shows that the impact on the trunnion is equivalent to a one foot drop due to the cushioning effect of 8-2
the shock absorbers.
In any case, the one foot drop on the trunnions was not analyzed.
(b) Provide mechanical properties Tor all materials as a function of applicable temperature ranges.
How were the limiter properties determined? How were they modeled? Has the model been substantiated? What is the basis for using an energy absorbing capability of 2200 inches 1b/in3 for the shock absorber?
(c) Provide stress levels at all critical locations on the cask.
(d) What happens to the radiation shield and shell following the three drop configurations analyzed?
8.2.1.2.3 1 m (40 inch) Puncture Drop without External Shock Absorber Once again we note that this review is not being conducted under 10 CFR Part 71, but under 10 CFR Part 72.
Resolution of comments, such as those included below does not constitute a review pursuant to Part 71.
You may wish to delete this section from your topical report and concentrate on accident scenarios consid-ered more credible for an ISFSI site under Part 72, such as tornado missile impact.
(a) Figure 8.2-2 shows the spike striking the cask bottom, yet this section indicates that it is striking the center of the cover.
(b) What are the lid deflections due to the puncture case?
(c) For the horizontal case, how are the effects of cask bending around the punch considered?
(d) What happens to the radiation shield and shell following the two drop con-figurations anclyzed? Also, what damage to the secondary cover is sustained?
8-3
8.2.1.3 Accident Dose Calculations (a) Provide verification that the seal has been maintained under accident conditions. What pressure is required to maintain the seal? What deformations are permitted?
(b) If there is any local change in the material thicknesses of the radiation shield and shell or covers, dose calculations must be provided.
8.2.2 Cask Tip Over 8.2.2.1 Cause of Accident As indicated in the response to Section 2.6, cask tip over is not as unlikely as implied by this paragraph.
8.2.2.2 Accident Analysis (a) What velocities were actually used for tip-over analysis? What elastic impact spring stiffnesses were used? How were they applied? Provide stress levels at all critical points on the cask and components.
(b) What happens to the radiation shield and shell following a cask tip over?
(c) If there is any local change in the material thicknesses of the radiation shield and shell, dose calculations must be provided.
8.2.2.3 Accident Dose Calculations (a) See comments under 8.2.1.3(a).
(b) In Table 8.2-1 stress limits are given for cask components and fuel. Why are unirradiated values used for the fuel? What about Zircalloy-2?
8-4
8.2.3 Earthquake 8.2.3.2 Accident Analysis What happens to the radiation shield and shell following an earthquake?
8.2.3.3 Accident Dose Calculations If there is any local change in the material thicknesses of the radiation shield and shell, dose calculations must be provided.
8.2.4 Tornado Wind Loading 8.2.4.2 Accident Analysis (a) Provide references or derivation for any equations used.
(b) The reference to NUREG-0800 Part 3.2.2 should be to Part 3.3.1.
(c) What are the cask weights used in the tip analysis?
8.2.5 Tornado Missiles
- 8. 2. 5. 2 Accident Analysis (a) Provide more detail for the analyses performed.
(b) What happens to the radiation shield and shell following the three tornado missile events analyzed? Also, what sort of damage is done to the secondary cover?
8.2.5.3 Accident Dose Calculations If there is any local change in the material thicknesses of the radiation shield and shell or covers, dose calculations must be provided.
See comments under 8.2.1.3(a).
8-5
8.2.6 Explosion 8.2.6.2 Accident Analysis Comments presented in Section 3.3.6.2 (Explosion) on the maximum allowable external pressure on the secondary cover apply here as well.
Also, what happens to the radiation shield and shell following an explosion?
8.2.6.3 Accident Dose Calculations If there is any local change in the material thicknesses of the radiation shield and shell or covers, dose calculations must be provided.
8.2.7 Flood 8.2.7.2 Accident Analysis What happens to the radiation shield and shell following a ficod?
8.2.7.3 Accident Dose Calculations If there is any local change in the material thicknesses of the radiation shield and shell, dose calculations must be provided.
8.2.8 Fuel Rod Rupture Support your contention that the fuel rods remain intact during all postulated accidents.
8.2.8.2 Accident Analysis (a) Where is the analysis which shows that the rupture of 100% of the fuel
(
rods will result in an increased internal pressure of one atmosphere?
(b) What would happen if a secondary seal were to fail or a fire were to follow the fuel rod rupture accident?
8-6
8.2.8.3 Accident Dose Calculations Accident dose calculations for the above scenarios would provide much insight into the worst case consequences of fission product gas inhalation.
8.2.9 Compression 8.2.9.2 Accident Analysis What happens to the radiation shield and shell following the compression accident?
8.2.9.3 Accident Dose Calculations If there is any local change in the material thicknesses of the radiation shield and shell, dose calculations must be provided.
8.2.10 Penetration 8.2.10.2 Accident Analysis What happens to the radiation shield and shell following the penetration accident?
8.2.10.3 Accident Dose Calculations If there is any local change in the material thicknesses of the radiation shield and shell, dose calculations must be provided.
8.2.11 Fire (a) Why does the discussion of fire not follow the format of the previous sec-tions? Where is Section 8.2.11.1 (Cause of Accident), Section 8.2.11.2 (Accident Analysis), and Section 8.2.11.3 (Accident Dose Calculations)?
8-7
(b) According to Section 7.3.2.2 (Shielding Analysis), the radiation shield
" material is suitable for use at temperatures up to 160*C (330*F).
During the transient, the radiation shield inner surface temperature is 200*C."
Furthermore, Fig. 8.2-7 indicates that the radiation shield inner surface temperature is still above 170 C after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. What effect does this excess temperature have on the shielding and material properties of the radiation shield material? If the properties are reduced, what are the
~
dose consequences?
(c) With such high clad temperatures (707*C), should there be concern about pressure buildup in the cask cavity due to expansion of the cavity gas?
Is leakage of the cask cavity gas possible? If so, what are the dose consequences?
(d) Was ANSYS used to perform the analysis for fire? In any event, details of the analysis should be provided.
(e) Will the neutron shielding thermal conductivity change as a result of the high temperature fire? Will the neutron shield materials outgas?
(f) List the thermal properties including heat capacities and material densities used in the analyses.
(g) For equations provided from any reference (such as the " Formulas for Stress and Strain," Roark and Young), provide also the table and/or equation number in that reference.
8.2.12 Cask Burial Under Debris l
l (a) Was ANSYS used to perform the burial analysis? In any event, details of the burial analysis should be provided.
(b) Why does the discussion of cask burial under debris not follow the format of the previous sections? Where is Section 8.2.12.1 (Cause of Accident),
Section 8.2.12.2 (Accident Analysis) and Sections 8.2.12.3 (Accident Dose Calculations)?
8-8
.,.ew--r-w------'--
r
-v--
er rn+--e---- ' - - - ' - - - " -
r
(c) According to Fig. 8.2-8, the radiation shield outer surface temperature is above 160'C within approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the accident.
Comments presented in Section 8.2.11 (Fire) with respect to the effect of excess temperature on the properties of the radiation shield and the possibility of pressure buildup in the cask cavity apply here as well, i
i I
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8-9
10 OPERATING CONTROLS AND LIMITS 10.1 Proposed Operating Controls and Limits 10.1.1 Contents of Operating Controls and Limits (a) Shouldn't the Section 1.2.2.2 design limit of less than 1000 mrem /hr at the surface following an accident be included? Where are the radiation dose limits for normal and accident conditions from 10 CFR Part 72.67(a) and 72.68(b)?
Is the maximum leak rate for each seal or both in combination?
(b) The one foot handling limit without impact limiters should be included in this section.
10.1.2 Bases for Operating Controls and Limits What about the radiation dose limits for normal and accident conditions from 10 CFR Part 72.67(a) and 72.68(b) in Section 10.1.2.2?
10.2 Development of Operating Controls and Limits i
10.2.2 Limiting Conditions for Operation 10.2.2.1 Equipment In the discussion of leak rates through other penetrations, it is stated that, "the cask is at a negative pressure and a leak at a penetration would not allow gas to flow outward." Air could, however, leak into the cask.
Would such a leak be indicated by the pressure monitoring system?
If so, how?
4 10-1
i 10.2.2.2 Technical Considerations and Characteristics (a) What is the missing PWR fuel value for the increase in the cask internal pressure?
(b) Why are IAEA " Safety Series #6" requirements relevant to a Part 72 review?
i 10-2
Combustion Engineering Dry-Cap Spent Fuel Storage Cask Comment on Quality Assurance Program 7-~.--.-.
ByletterdatedMarch26,lh8Ei(LD-86-021),thePowerSystemsDivisionof Combustion Engineering, In.',"(C-E) responded to our request for additional information (RAI) of January 46,1986 which was transmitted to C-E by our letterofFebruary18,198d l
We have reviewed the C-E 1 tieMnd-find that it acceptably responds to the RAI in this area of reviewhherethe C-E Quality Assurance (QA) program for the C-E Dry-Cap Cask as deperibed-in CENPD No. 273-NP, Rev. 1-NP.
In responding to the RAI, C-sfaled that_the following components of the Dry-Cap Cask are important to saf ty_(pe u he_1,0 CFR Part 72 definition) and therefore controlled under its QA pr,ogram:
the cask body, the primary cover and bolts fastening it to the cask,!and tho'se components in the gas sampling and cask drain cask penetration which function as a boundary between the interior of the cask and the environment. Our technical reviewers have determined the acceptability of this list, but also find that the cask trunnions and trunnion bolts should be included in this list. C-E should include this QA program scope in chapter 11 of its TSAR at its next revision.