ML20148K359
| ML20148K359 | |
| Person / Time | |
|---|---|
| Issue date: | 03/23/1988 |
| From: | Roberts J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Haelsig R NUCLEAR PACKAGING, INC. |
| References | |
| REF-PROJ-M-44 NUDOCS 8803310096 | |
| Download: ML20148K359 (26) | |
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MAR 2 219R9 Project M-44 Nuclear Packaging, Inc.
ATTN: Richard T. Haelsig President 1010 South 336th Street Federal Way, Washington 98003 Gentlemen:
In response to your submittal docketed on November 6,1987, under Project No. M-44, we have initially reviewcd your Topical Report (TR) entitled "Topical Safety Analysis Report for the NuPac CP-9 Concrete Storage Cask" (TP-08), Revision 0, as subsequently revised (Revision 1) in November 1987.
Our detailed comments are enclosed (enclosure 1).
In particular, I refer you to our coments on Appendix 4, "NuPac Drop Test Program Development," and our earlier coments on this matter in our letter of March 3, 1987, which is enclosed (enclosure 2).
If you have any questions regarding our coments, please call me.
Sincerely, Origigi Signed By:
John P. Roberts, Section Leader '
Irradiated Fuel Section Fuel Cycle and Safety Branch
Enclosures:
- 1) Coments on NuPac TR
- 2) Letter of March 3,1987 to NuPac l
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1 COMMENTS ON THE "TOPICAL SAFETY ANALYSIS REPORT FOR THE NUPAC CP-9 CONCRETE STORAGE CASK" CHAPTER
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF STORAGE SYSTEM 1.0-1 Para 1.2.2. NUPAC is requested to discuss the pad on which the casks are to be placed. If it is the intent of NUPAC that the pad be site specific, it should so state, but at least describe the minimal design envelope for the pad. ;
1.0-2 Para 1.2.2. Since neither the procedures for the dry load, nor its handling equipment have been addressed in Revision 0 of the TR,.the NRC staff is not reviewing its adequacy to meet the requirements of 10 CFR 72.
4 1.0-3 Para 1.4. Figure 1.4-1 is said to show an array of 50 casks, when in fact 68 casks are depicted. Please explain.
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CHAPTER 2.0 PRINCIPAL DESIGN CRITERIA 2.0-1 Para 2.1. Table 2.2-1. NUPAC lists only a 24-hour rolling average temperature of 100 F. The NRC staff requests that NUPAC incorporate an accident design criteria with a maximum ambient temperature of 125*F. Also, discuss the consequences of this ambient temperature on the fuel cladding as well as the effect on the long term durabil-ity of the concrete. These consequences can be discussed in Chap-ter 11 of the TR.
2.0-2 Para 2.2, p. 2-3. The rainfall rate has not been included. This is a potential source of concrete spalling due to thermal gradients which might be produced. Please discuss this load condition.
2.0-3 Para 2.2.2, p. 2-5. Thermal gradients (causing potential concrete j spalling) produced by flooding are not included. ,
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2.0-4 Para 2.2.5. The combined load criteria shown in this section, does )
not correspond with ANSI /ANS 57.9-1984, Sections 6.17.3.1 and l 6.17.3.2.1 for concrete and steel structures. The basic concern that !
the NRC staff has with Section 2.2.5 of the TR is that the thermal I loads are not adequately considered in the off-normal and the acci-dent conditions. Since the NUPAC CP-9 cask is constantly in a state of thermal stress, the thermal loads should be combined with all off-normal and accident cases, as shown in the ANSI 57.9 reference. This will not require two accident cases to be postulated to occur simul- ;
taneously since the thermal condition is not accident, but rather is a constant condition.
2.0-5 Para 2.2.5, p. 2-9. Accident loads do not include a drop with rota-tion resulting in impact on upper rim of cylinder. Please discuss rationale for excluding impact on upper rim of cask. (Note that any aircraft crash impact or fire has been left to a site-specific application.)
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2.0-6 Para 2.3.3.1, NUPAC's discussion of equipment is not satisfactory.
NVPAC's positfon that the "CP-9 storage system may include several pieces of support equipment, (which) .... by its nature is site specific," While the crane, transport vehicle or trailer, lift fix-ture and support pad may be sit) specific, they must be described in enough detail to desnoastrate feasibility for operations purposes, and the minimum design envelope must be defined, i.e., load specs for the pad, lift capacity of the lift fixture and crane, and appropriate safety classification of these items, with appropriate industry star.d-ards. The "additional handling equipment" are important to safety since their operation and dimensions may define the parameters of an accident, such as the potential height of a drop. For example, the cask will be suspen M at a height greater than the 5-foot height used for the accident analysis (Section 11.2) when it is over the decontamination area (per Figure 1.2.2-1).
2.0-7 Para 2.3.6, p. 2-13. Fire or explosive events are credible per Regulatory Guide 3.48, para 2.2 entry 7 (aircraft impacts). They can, however, be dealt with in a subsequent site-specific license application. NUPAC should modify their statement to reflect this.
2.0-8 Para 2.4, p. 2-14. It is stated that decontaminable coatings are used on the exterior of the cask so that the exterior surface can be cleaned "to meet site-specific standards." Define the governing j bounds of the "site-specific standards" that the proposed coatings can meet. What standards (e.g., alpha and beta contamination limits l for offsite shipping containers) are being used as the design basis )
for achieving an acceptable level of decontamination?
I 2.0-9 Para 2.4. Since the decontaminable coating is an integral part of the ,
NUPAC CP-9 cask, the NRC staff requires that the coating be identified. l Also NUPAC should supply evidence regarding the suitability of the i material for the environment in question, i.e. , application to con-crete, long term durability at 190 F, survival during moisture out- 1 gassing from the concrete, etc.
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CHAPTER 3.0 STRUCTURAL EVALUATION !
l 3.0-1 Para 3.1.1, p. 3-1. It is considered that the inner concrete sur- l faces, against the liner, should be treated as "Exposed" due to the I high temperatures to which it is subject, and should therefore receive minimum reinforcement in accordance w'ith Section 7.12, ACI 349-80 (or 349-85).
l 3.0-2 Para 3.1.2, p. 3-2. Live loads due to the extreme thermal gradient 1
produced by heavy, cold ra r on a high temperature day (e.g., summer ,
thunderstorm) have not been estimated. It is considered that there j is significant potential for stresses that spalling of the concrete l may result. This would be a "live load" in combination of load j expressions, per ANSI-ANS $7.9-1984 (i.e., neither an "other-than- :
normal" nor an "accident" situation).
1 I 3.0-3 Para 3.1.2.1, p. 3-2. The concrete design is based on ACI 349-85 instead of ANSI-ANS-57.9-1984 load combinations. The principal impacts are that: T factor should be 1.3 in load combinations 8 and o ,
9 of Table 3.1-1 instead of 1.05, to conform to ANSI 57.9-1984 com-binations 6.17.3.1 (c) and (d); internal pressure due to temperature, F, on Table 3.1-1 are thermal or live loads and should have the same factor as the live load, L, or the thermal load, Tg, but not the lower
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factor of the dead load. [ Note: Use of ACI 349-80 is required to the i extent directed by ANSI /ANS 57.9-1984 (e.g., Section 6.17.2.1). The l NRC has not approved the substitution of ACI 349-85 in lieu of ACI l 349-80 for ISFSI.]
l 3.0-4 Para 3.1.2.1, p. 3-2. The use of ACI 349-85 factor F, lateral and vertical pressure of liquids or related internal moments and forces, for gas pressures is not appropriate. Normal operating gas pressures should be L, live loads, if ACI 349-85 is used (also see comment above 3.0-3). (Predictability of liquid vs. gas pressures is different.)
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3.0-5 Para 3.1.2.3, p. 3-4, Table 3.1-2, p. 3-5. NUPAC states that the steel liner and basket internals are designed to ASME B&PV Code,Section III, Div. 1, Subsection NC for Class 2 components. This is not recognized by NRC Regulatory Guide 3.60, or ANS-ANS 57.9-1984.
ASME B&PV Code Sectkn III only relates to stainless steel material and properties per Regulatory Guide 3.60.Section VIII relates to stainless steel materials and properties, and cask / silo design per ANSI /ANS 57.9-1984, Section 6.4.2.4 3, in addition'to the require-ments of Section 6.17 of the standard. The ASME is not as conserva-tive as ANSI 57.9 because the Code excludes thermal stresses from off-normal and accident cases by treating them as secondary stresses.
The staff requires that regardless of the design procedures, the stress or plastic strength load combinations of ANSI-ANS 57.9-1984 para 6.17.3 must be met.
3.0-6 Para 3.2 and 3.4.3, p. 3-3 and 3-11. The maximum weight for lif ting should include weight of fluid within liner for removal from spent fuel pool.
3.0-7 Para 3.2, p. 3-6. Please include references for criteria defined for (a) lifting devices for normal operations for cask, (b) cask vessel and cover for primary membrane and bending for accident condi-tions, (c) cask vessel for cask through-wr.ll and outer surface stress for accident condition puncture by projectile.
- 3. % Para 3.3, p. 3-10. The NRC staff considers that the 40 year values for concrete at elevated temperatures represent extreme extrapola-tions from very limited empirical data. Also address potential synergistic effects from radiation exposure.
3.0-9 Para 3.4.4.1, p. 3-14. The temoarature projections indicate that:
the concrete may reach 409 F at tne concrete liner interface, and even under the coldest period, the concrete will be continually over 231 F. Furthermore, the average tr'oerature in warmer seasons may be approximately 330 F (70 F 24-hour avg) to 340 F (80 F 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> avg).
All of these temperatures exceed the 150*F maximum allowed by ACI 5
349-80 (referenced in ANSI 57.9-1984), or ACI 349-85. ACI 349 80 (and 85) state that higher temperatures may be allowed if tests are provided to evaluate the reduction in strength and this reduction is applied to design allowables. Also, that evidence shall be provided which verifies that the increased temperatures do not cause deteriora-tion of the concrete either with or without load.
The NUPAC report includes Appendix 2 which specifically addresses potential reduction in strength due to extended increased tempera-tures, and makes projections for a concrete designed to minimize adverse characteristics. Appendix 2 does not present any new emeir-ical results. The projections of physical properties to the life of the proposed NUPAC cask appear to be extreme extrapolations from very limited data, which do not constitute a statistically adequate addressal of the potentially important parameters.
The projection of negligible physical degradation over the projected cask life is based on even more limited empirical data, which does not appear to include the temperature levels to be expected by the CP-9 design.
The concrete data, cited by NUPAC, on which the projections of pro-perties, and ability to survive the temperature levels for the cask for the lifetime of the cask were apparently available to the ACI in preparing ACI 349-85. Tne data may have supported higher temperatures but the ACI did not allow higher temperatures. It is ccnsidered that Appendix 2 does not adequately support use of concrete at the projected temperatures.
Conclusion- NUPAC has not provided adequato empirical bases for justifying the use of concrete at the projected temperatures. Until NUPAC can demonstrate the suitability of the material as used in the design, the design of the CP-9 cask does not meet the requirements of 10 CFR 72.
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3.0-10 Para 3.4.3, p. 3-12. It appears as though the cross section area for the 1-1/4" bolt is incorrect (0.994 ina vs. 0.8 M in2 ). Also 8 threads instead of 7 were specified. If NUPAC is not using standard UNC threads and/or cross sections, it must state that.
3.0-11 Para 3.4.3, p. 3-13. No cone and edge distance effect of the embedded reinforcing bar were considered. ACI 349-80, Appendix B should be used.
3.0-12 Para 3.4.4.2.1, p. 3-14. Please supply a listing of the ANSYS com-puter analyses, both input and output files. Also if any post pro-cessing has been done, supply these also. Please include a detailed sketch showing numbers of nodes as well as element numbers.
3.0-13 Para 3.4.4.2.2, p. 3-18. No calculations are provided to show development of in plane thermal stresses. These are necessary to !
validate the results. The studs are ignored in all calculations.
Please justify this simplification, or provide reanalysis with studs.
There are no calculations or discussion addressing the structural j design of the base of the cask. l 1
1 3.0-14 Para 3.4.4.2.2, p. 3-18. No discussior, or inclusion of effect of differential expansion of the higher temperature studs extending into the concrete 8". Do these cause vertical cracks at the end I buttons, pull the concrete away from the liner (introducing a low I thermal transmission gap), or what? The studs should exert inward ;
1 loading on the liner plates. If concrete is pulled away.from the j liner, what happens at corners; is there a potential for a signifi- I cant vertical crack?
3.0-15 Para 3.4.4.2.3, p. 3-19. Since a cracked concrete section is assumed (initial expansion of the liner before steady state conditions should assure that the concrete cracks), any concrete which has thermal expansion greater than the perimeter rebar must be restrained by tension in the rebar. Ccocrete should be added to the equations on pages 3-20 and 3-21 of the TR. Increase in rebar stress is expected.
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3.0-16 Para 3.4.4.2.3, p. 3-20, 1st equation: all temperatures should be the differences in temperature from the temperature at which there are no thermal stresses (the reference temperature, 70 F?). It is also erroneous to use the final coefficient of expansion of the final temperature to calculate expansion which has occurred getting to that temperature.
3.0-17 Para 3.4.4.2.3. The assumption made for the axial thermal stress is not conservetive. For instance, plane sections do not remain plane because the thermal gradient is very high, measured from bracket to cuter surface of concrete. The axial rebar is approximately 205 F whereas the basket is 409 F for Case 1. None of the four structural elements listed in the analysis, i.e. , the rebar, liner, or two basket elements are at the same temperature nor are they the same length, therefore, it is incorrect to assume equal thermal strains. Thus, the set of four equations discussed on pages 3-20 and 3-21 does not form a homogeneous set of linear equations, and the solutions offered are incorrect.
3.0-18 Para 3.4.4-3.1, p. 3-22. Pressure loading stresses are assumed to be )
wholly withstood by liner, and not to add to stress in tangential rebar. This appears to ignore the compressively prestressed studs !
supporting the liner on an approximately 5" x 18" grid. The only reaction to forces transmitted to the button heads of the studs is in the rebar, with the assumption of no concrete tensile stresses and no ;
tensile bo7d between concrete and liner.
3.0-19 Para 3.4.4.3, p. 3-23. The lid is not only subjected to internal l pressure but also the bending stress due to the expansion of the i i
liner as well as thermal stress due to the elevated temperature.
l 3.0-20 Para 3.4.4.3.2, p. 3-25. Load equation should be U=1.3T . (See n
comment 3.0-3).
3.0-21 Para 3.4.4.3.2, p. 3-25. See comments 3.0-13 and 3.0-15 regarding calculation of stresses for rebar.
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l 3.0-22 Para 3.4.4.3.2, pp. 3-26 thru 3-31. See comment 3.0-5 regarding use l of ASME B&PV Code Section III in lieu of ANSI 57.9-1984. l 3.0-23 Fara 3.4.4.4.2, pp. 3-28, 3-29. Rebar-Axial: strength reduction l
factor was not applied per ACI 349-80 para 9.3. Compare against required strength, see comment 3.0-3.
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3.0-24 Para 3.4.5.1.1, p. 3-32. Do the basket stresres include stress due l to inward bending of liner due to ternperature gradient across the liner, and expansion of the studs?
I 3.0-25 Para 3.4.5, p. 3-31 refers to thermal analysis as it is driven by cold ambient temperatures. See comments 3.0-12 through 3.0-23 above as they pertain to the "Heat" case, Section 3.4.4 of the TR.
3.0-26 Section 3.0. Qualification of welds was not addressed. No analyses of removable lifting lugs and cask lifting device were performed. !
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u CHAPTER 4.0 THERMAL EVALUATION 4.0-1 Para 4.0. For the canistered design, what percentage of heat is transferred to the steel liner by radiation and what percentage is transferred by conduction in gas? and in steel?
4.0-2 Pc a 4.1, What is the basis for the choice of solar insolation values? Following 10 CFR 71, one would expect to use 2 x 61.5 = 123 Btu /hr - ft2 for curved surfaces and 2 x 123 = 246 Btu /hr - f t2 for the top? These are twice the values used in the TR.
4.0-3 Para 4.1. Since 10 CFR 71 is being used as a basis for the 100 F initial temperature, why not also use the cold value of -40 F?
4.0-4 Para 4.1. Address the effect of sudden rainfall / snowfall on thermal stresses, including the effect of repeated cycling over the life of the cask.
4.0-5 Please provide a copy of Reference 15.
1 4.0-6 Para 4.1. What thermal stresses will occur durine heatup following loading of the fuel? In particular, the steel liner will heat up much faster than the concrete. The concern is the transient thermal response of the system. The thermal conductivity of steel is two orders of magnitude higher than concrete. Please evaluate this problem.
4.0-7 Para 4.4.1.1.1. Is axiai conduction in the fuel modeled? If so, what is the value of thermal conductivity used?
4 4.0-8 Para 4.4.1.1.1, What is the basis for the axial power profile used?
4.0-9 Para 4.4.1.1.1, p. 4-13. Boundary condition 5 states that the side cask surfaces radiate to the other casks. How is this reconciled with the statement at the bottom of page 4-11 and top of 4-12 which states "the cask side surfaces radiantly transferred heat to the sky only."
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4.0-10 Para 4.4.2. What is the maximum abnormal temperature for which approval is sought? This has not been addressed.
4.0-11 Para 4.4.1.1.1. What contact resistance was assumed between the steel liner and the concrete of the canister? What is the basis for this choice? What would be the impact if a gap occurred between the liner and the concrete (see Comment 3.0-14)?
4.0-12 Para 4.4.1.1.1. In the analysis of cask burial, how is the effect of neighboring casks accounted for?
4.0-13 Para 4.4.1.1.1. How is the asymmetric nature of the solar insolation load addressed in the thermal stress analysis?
4.0-14 Table 4.4-8, p. 4-35. The table shows calculated temperatures for Case 1 level 23 remaining at 100 F for sixteen cells within the cask.
Twenty cells remain at 100 F at level 24. These do not appear to be boundary cells so one would expect some temperature rise above 100 F.
This same apparent anomaiy also occurs for the other two cases.
Please explain.
4.0-15 Para 4.4.2. Please identify the relative magnitude of heat transferred from the cask surface by each heat transfer mechanism.
4.0-16 Figure 4.4-7, p. 4-30. From the temperatures given in the figure, the average heat flow through the canister was calculated using the formula.
2RkL 0= (T1 T) 2 in(r2/rt)
The average heat flow was calculated to be 384 watts per foot of cask height at axial level 11. However, using 0.6 kW per assembly and a 1.2 axial peaking factor, the average heat flow should be 540 watts per foot. Please explain this discrepancy. Can this be due to heat losses through the top surface?
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CHAPTER 5.0 SHIELDING EVALUATION 5.0-1 Table 5.2-6, p. 5-12. Provide additional information regarding derivation of ANISN model source normalizations given in the table.
5.0-2 Table 5.3-1. p. 5-15. What is the free or bound water content assumed in the concrete composition given in the table? Has any variation in moisture content with concrete temperature been considered?
5.0-3 Table 5.3-1, p. 5-15. Is the fron content of the concrete composition given in the table uniformly distributed in the aggregate, or does it represent iron in structural reinforcing (rebar)?
5.0-4 Para 5.4. Why wasn't an estimate of the neutron dose made with QAO-CG?
5.0-5 Para 5.4. Hava any error or uncertainty estimates of the doses been made?
5.0-6 Para 5.4, p. 5-18. On the basis of the results shown in Table 5.4-1, please explain the statement that the neutron dose rate from the bottom of the cask is expected to be relatively insignificant. Rela-tive to what?
5.0-7 Para 5.4, p. 5-29. Clarify the logic leading to the statement that the "loss of the RX-277 shielding from the top of the NuPac CP-9 cask will be of no consequence."
5.0-8 Para 5.4, p. 5-29. It is stated that the thickness of concrete required for a dose rate of 1,000 mrem /hr at 1 m from the cask side is found from linear interpolation of Table 5.4-4 values to be 48 cm. From the Table, however, it appears that the corresponding dose rate for 48 cm is about 2,000 mrem /hr. Please explain how the 48-cm value was derived.
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5.0-9' Para 5.5.1, p. 5-31. It is stated that the estimate of offsite dose j from normal operation is. based on an array of 50 casks "as illustrated '
in Figure 1.4-1." That figure, however, depicts an array of 68 casks. ,
Please explain the discrepancy. l
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CHAPTER 6.0 CRITICALITY ANALYSIS 6.0-1 Para 6.1, p. 6-1. Have any criticality calculations been performed.-
for the reference design? What analysis has been performed to indi- I cate that the design is expected to satisfy criticality safety crite-ria even taking credit for burnup?
6.0-2 Para 6.1, p. 6-1. Even if NUPAC is deferring burnup credit considera-tions pending NRC resolution, what verification calculations of your methodology, codes, and cross section libraries have been performed in order to determine calculational and library uncertainties?
6.0-3 Para 6.3.1. What is the basis for the selection of the 15 x 15 PWR array for the criticality analysis? Are other fuel designs to be con- l sidered in the Topical Report? Are other fuel designs candidates for storage?
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CHAPTER 8 OPERATING PROCEDURE 8.0-1 Para 8.1, p. 8-1. NUPAC has stated that detailed operating procedures will be developed for a site-specific application. In order to show feasibility of the procedures, NUPAC is requested to provide details of some of the equipment which is necessary for operation.
Examples of these details include:
(a) Description of the trailer or special transport vehicles defin-ing load capability of the trailer, height of the bed, any restrictions in use, such as minimum density of roadbed for supporting the loaded vehicle.
(b) Equipment for removing the cask lid.
(c) Procedure for checking the integrity of the decontaminable coat-ing for cask, i.e. , what are the criteria for integrity?
(d) Description of the lifting device, including availability from commercial supplier, standards to which device is designed, method of determining the two degree angle for the lifting lugs.
(e) Method for collecting the contaminated water removed from the decontamination procedure.
(f) Description of alternative fuel loading technique for dry transfer.
(g) Specify the areas above the shield plugs to be avoided by operating personnel (p. 8-2).
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8.0-2 Para 8.1, Item 5, p. 8-1. Removal and installation of lifting lugs are not shown anywhere in the report.
8.0-3 Para 8.1, Item 7, p. 8-1. Configuration, and the operation of the lifting device are not shown and explained anywhere in the report.
8.0-4 Para 8.1, Item 12. Removing the shield plugs before lowering the cask into the pool is essential to reduce the stay time in the high radiation areas and is a good ALARA practice.
8.0-5 Para 8.1, Item 26, p. 8-3. Is reference to Section 9.0 meant?
8.0-6 Para 8.2, Item 7, p. 8-4. Where will the air exit and how will the temperature of the air be measured?
8.0-7 Para 8.2, Item 8, p. 8-4. How will the cask cavity be flooded with water?
8.0-8 Para 8.2, Item 11, p. 8-5. To practice ALARA principle, the three remova' ale cask lifting lugs should be installed before the removal of the lid (Item 6, page 8-4).
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CHAPTER 9.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM 9.0-1 Para 9.1.1.2, p. 9-1. NUPAC states that they intend to visually inspect the welding per the ASME Code requirements of Section V, Sub-section A, Articles 1 and 9, and Subsection B, Article 28. There is no Article 28, perhaps NUPAC intended 28 instead of 2B?
9.0-2 Para 9.1.1.2, p. 9-1. The NRC staff does not find that the visual method of weld inspection for the confinement liner to be satisfac-tory. Baced on Article NC-5000 of Section III, Division 1, Subsec-tion NC for Class 2 components, NUPAC's choices are limited to radio-graphic examination per Article 2 of Section V, ultraronic examination per Article 5 of Section V. liquid penetrant examination per Article 6 1 of Section V or magnetic particle examination per Article 7 of Section V.
After NUPAC selects one of the above acceptable methods of weld inspection, they must then select the corresponding required docu- l mentation of Section V. Article 28 for visual examination is not one of the choices.
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1 CHAPTER 10 RADIATION PROTECTION 10.0-1 Para 10.2.4, p. 10-6. There is an error in the estimate of the "per assembly" occupational dose. It should read 0.0179 rem / year, not 0.179 rem / year.
10.0-2 Para 10.1, p. 10-1. It is stated that one of the ALARA considerations is "thick concrete walls to reduce the side surface dose to below 50 mrem /hr." On the following page, however, it is stated that the design dose rate for the side surface is 20 mrem /hr. Please explain this apparent inconsistency.
10.0-3 Para 10.3, p. 10-6. NUREG/CR-0446 is cited as the basis for ALARA achievement. This report should be listed as a reference.
10.0-4 Para 10.3.2, p. 10-3. "Working Dose Rates" are used as a basis for estimated occupational doses. This parameter is used in Tables 10.3-2 and 10.3-3. In some cases, but not all, the values for Working Dose Rate are one-fourth the values for Design Dose Rate. Please explain (1) exactly what is meant by this parameter; (2) how the values were derived; and (3) whether or not the values include the dose contribu-tions from multiple casks in an array.
10.0-5 Para 10.3. In the onsite dose assessment presented in Chapter 10, no attempt has been made to estimate the dose to "other" site workers (i.e., workers other than those directly involved in cask handling and surveillance activities). It is recognized that such an assess-mer.t depends strongly on site-specific factors (e.g., site layout, construction and occupancy of buildings, etc.). However, from the dose-versus-distance data presented in Figures 5.5.1-1 and 5.5.1-2, it is apparent that the collective dose to other site workers is likely to be much higher than that for those working directly on cask-related tasks. A worker whose work location is 100 meters from 18
the array would receive about 140 mrem /yr. Does the data used in the evaluation of concrete as a neutron shield account for the decrease occurring in the hydrogen content of the concrete? It would be highly desirable for the TR to provide not only a general assessment of the total onsite cose, but also identify (1) site-specific considerations affecting the assessment, and (2) means which could potentially be used to reduce this dose.
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CHAPTER 11.0 ACCIDENT ANALYSES 11.0-1 Para 11.1.2.2, p. 11-3. The principal problem would be the loss of effectiveness of the concrete for radiation shielding. The principal concerns are the readily observed physical loss of spalled or frac-tured concrete; or, the formation of opened fractured planes within the concrete allowing a reduced-attenuation radiation path, which may not be readily detectable.
11.0-2 Para 11.1.2.3, p. 11-3. Editorial, 2d para: A 5-foot free drop results in a velocity of approximately 18 fps. (Does not affect 1 1
conclusion.) j l
11.0-3 Para 11.1.2.3, p. 11-3. The NRC staff doubts that maximum velocity of cask moved by crane can be held to 1 fps or less; however, the staff concurs that 5-foot drop is a more severe situation.
11.0-4 Para 11.2, p. 11-4. Observation. Potential aircraft impact case, per Regulatory Guide 3.48, para 2.2.7, has apparently been left to the site specific application.
11.0-5 Para 11.2.1.2, p. 11-7. ANSI-ANS 57.9-1984 load combinations and of allowable stress should be used. [1.7S > D + L + H (0) + T + A].
11.0-6 Para 11.2.1.2, p. 11-7. Provide copies of input and output of ANSYS PC/ Linear Finite Element Analyses. As well as the mathematical models with node and member numbers.
11.0-7 Para 11.2.2.2.1, p. 11-15. Provide qualification reports of computer programs CONC 0P and OBLQUE.
11.0-8 Para 11.2.2.2.1, p. 11-18. The TR does not appear to address pre-stressed condition produced by temperature, or the extent of cracking produced by the initial transient and long term temperature gradients.
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The drop could occur near beginning or near end of life, and during life if land use or storage regulations change.
11.0-9 Para 11.2.2.2.2, p. 11-22. Provide copies of input and output of ANSYS PC/ Linear Finite Element Analyses, as well as the mathematical ,
model with node and member numbers.
I 11.0-10 Para 11.2.3.1, p. 11-29. Tip over could also occur if the drop from suspension by a crane occurred while cask was over corner of a load-ing dock or trailer. Loading dock heights are typically 5 feet. :
Drop involving rotation about such a corner with impact on upper rim of cylinder is considered potentially more serious than impacts i involving the base.
1 11.0-11 Para 11.2.3.3, p. 11-30. See comment 11.0-10.
11.0-12 Para 11.2.5, p. 11-40. Worst flood situation may be the steep thermal gradients produced by "apid immersion of a hot cask in cold water, resulting in potential spalling, and successive spalling if the water is able to progressively contact the surfaces created by spalling (to be expected, depending on the extent of internal reinforcing steel).
11.0-13 Para 11.2.6.2, p. 44. Frequency of the cask should be 33.61 Hz instead of 29 Hz. (Calculation or typing error).
11.0-14 Para 11.2.6.2, p. 45. For consistency, the velocities for the same damping value, should be used for the stress combinationc There is only one vertical velocity. Velocities for .5% and 10% damping should not be combined.
11.0-15 Para 11.2.6.2, p. 45. Since hand calculations were performed, the Guidance of Standard Review Plan 3.7.2 titled: Seismic System Analyses Rev.1, dated July 1981, page 3.7.2-5 "Equivalent Static Load Method" should be used. To be conservative, the peak values of the lowest 21
damping (.5%) may be used. Although, the peak values for 2% damping are also acceptable based on the Regulatory Guide 1.61, October 1973, Table 1.
l 11.0-16 Para 11. 2. 8. 2, p. 11-51. Load combinations should be in accordance j
with ANSI-ANS 57.9-1984.
I I
l l
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- - . . =. . . .
APPENDIX 1.0 NUPAC CPQ STORAGE CASK DESIGN DRAWINGS i I
A1.0-1 A Bill of Materials table should be prepared to account for each item in the analysis and for ordering and purchasing.
A1.0-2 Drawing X-104-105-SP Rev. None, Page 4 of 5. It is not clear how the three (3) removable lifting lugs are installed.
A1.0-3 No drawing for the cask lif ting device.
1 A1.0-4 There are no apparent provisions for ensuring proper alignment of the lid.
1 l
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I 23
APPENDIX 2.0 CONCRETE MATERIAL PROPERTIES A2.0-1 App 2, p. vii. 2d Para, 1st Sentence: It is considered that the study (App 2) does not establish that the concrete will not degrade over the time and temperatures anticipated.
A2.0-2 App 2, p. 2. 2d Para, Last Sentence: Wall thickness actually varies from 25" to 32.5" of concrete.
A2.0-3 App 2, para 1.0, pps. 2-3. Last para on p. 2, continuing to p. 3:
Sources not given for statements.
A2.0-4 App 2, para 2.1, p. 5. Last 3 sentences of para 2.1: Versus what time factors? Time is extremely critical for a steady state high temperature situation.
A2.0-5 App 2, para 4.3.3, p. 41. Last 2 sentences on page: Appear to be subjective extrapolations of a limited data sample, especially the final sentence. ,
A2.0-6 App 2, para 6.3.1, p. 56. Editorial, 3d line from bottom: Believe intent is "eight weeks" ("months" is in error).
A2.0-7 Type II concrete subjected to high temperature should only b6 quali-fied by testing. The conclusions reached by research and review of reports which may or may not be applicable for this application are not accepted, as lacking in empirical support. (See Comment 3.0-9).
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APPENDIX 4.0 NUPAC DROP TEST PROGRAM DEVELOPMENT A4.0-1 App. 4, p. 4-1. There are two pages with the same page number (4-1),
the title page and the introduction page.
A4.0-2 App 4, para 1.1, p. 4-1. Tip over could also occur if drop from suspension by a crane occurred while the cask was over an edge of a loading dock or trailer. Loading dock heights are typically 5 feet.
Drop involving rotation about such a corner with impact on upper rim of cylinder is considered potentially more serious than impacts involving the base.
A4.0-3 App 4, p. 4-14, 3d para: Address the potential effect of the much hotter studs, which try to push the concrete at the plane of the button heads outward of the concrete within that plane. The shape of the buttons could foster the initiation of cracks in the plane of the buttons which could be propagated due to the thermal gradient.
The restraint on concrete movement would be shear and tension resist-ance at the corners and any bond (considered to be nagligible) at ,
the liner. Cracks at the corners could provide a radiation path.
Cracks in button plane would affect thermal conductivity. Opening a gap at the liner (which is also probable due to the gradient across the liner plate) could increase maximum internal temperatures.
A4.0-4 App 4, pp. 4-14 to 4-16. Address whether the cask should have a thermal gradient to simulate the prestressing which would exist in
, the real-life situation. If testing is performed, it should include an appropriate thermal gradient.
9 A4.0-5 App 4, p. 4-17. It is considered credible that the side of the upper end of the case strike first. See Comment A4.0-2, above.
A4.0-6 App 4, pars 6.0, p. 4-37. The NRC staff does not concur with state-ment of projected model test validity, due to lack of a rotation onto upper rim and lack of thermal gradient during test.
25 g _ __
F.C 3 'C2 Project No. H-44 , ,
Nuclear Packaging, Inc.
ATTN: Richard T. Haelsig President 1010 South 336th Street Federal Way, Washington 98003 Gentlemen, We have examined the NuPac Internal Report "Impact l'erfomance and Testing of the NuPac CP-9" submitted by letter dated December 17, 1986, in support of the proposed NuPac CP-9 Scale Model Test Program. This document and two others also submitted are material in support of your topical report (TR) for the CP-9 concrete cask design. From discussions during our meeting of December 17, 1986, this report is to be an appendix to the TR to be submitted. The CP-9 Scale Model Test Program which is to be developed, is also to be submitted in conjunction with and as a part of the TR.
We have concluded that we see nothing incorrect in your understanding of scale model testing. Clearly, however, we do not have sufficient infomation to judge whether your test program and TR will be adequate. At such time as you have fully developed the CP-9 Scale Model Test Program in conjunction with the TR, we would expect to review these prior to any actual testing of a CP-9 cask scale model.
We also want to take this opportunity to make a few connents regarding your submitted report and your proposed test program. Hopefully, these will be helpful to you in addressing issues concerning our prospective review of the test program and TR.
e The report appears to reflect an appropriate understanding of scale model design and test design suited to experiments with sub-scale reinforced concrete structures.
e The report, however, provides insufficient information on the actual prototype and model designs, concrete design, and potential handling scenarios to be replicated by test (i.e., elements of your test program) to support concurrence in the approach.
e The role (s) and relationship of the model testing to your analytical model or code should be explained. Also, explain these with respect to your design analysis, design verification, and/or validation of proof against accidental impact. In conclusion, the objectives of the test program are not yet clear and must be carefully developed.
<tBEMP MP
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o The report does not address the themal enviropment,. with the exception of describing potential use of a deliberately weak concrete. The analysis of themal effects on the potential structural response and i design suitability of the actual cask will be closely examined in the '
topical report. It should be noted that the NRC requires a high degree of proof prior to accepting designs exceeding the limits of the appropriate codes, regulations, and standards (such as ACI 349-75). l The potential thermal environment and temperature-produced metamor-phosis which may occur with your design may require such proof of design suitability. l e Be aware that additional potentially critical impact events may be !
postulated on the basis of the actual design and proposed usage '
scenarios, when developed.
s The acceptance criteria indicated should refloct both comprehensive review of potential failure modes corresponding to the hypothesized accident, and adequate measurement and/or tests to verify meeting the criteria.
original signed hF Jolm F. Roborte John P. Roberts Advanced Fuel and Spent Fuel Licensing Branch Division of Fuel Cycle and Material Safety Di e 'eri'y; tion :
TJFe?i.'~if T4 File POR fI.T.c File NHSS R/F JPRoi. M s LCRouse F8rown
.75his dee n _a OFC: :F : : : : :
NAME: oberts/jp/cr:LCRous'e : : : : :
DATE:03/.2 /87 :03/%/87: : : : :
/ OFFICIAL RECORD COPY
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