ML20034A965
| ML20034A965 | |
| Person / Time | |
|---|---|
| Issue date: | 04/20/1990 |
| From: | Roberts J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Massey J PACIFIC SIERRA NUCLEAR ASSOCIATES |
| References | |
| REF-PROJ-M-53 NUDOCS 9004250150 | |
| Download: ML20034A965 (21) | |
Text
,
APR 2 01990 Project No. M-53 Pacific Sierra Nuclear Associates
'ATTH:
Dr. John V. Massey General Manager 5619 Scotts Valley Drive Scotts Val ky, CA 95066
Dear Dr. Massey:
Nuclear Regulator RC) staff has reviewed Pacific' Sierra Nuclear W
Associates' (PSN)y Commission 0: Topical Report er. titled, " Ventilated Storage Cas System," PSN.89-001',- Revision 1, dated February 1990. Our review includes all the responses and calculation packages sent to the NRC in response to our initial review.-- The staff's detailed comments are enclosed.
Based on our review of the materials provided by PSN, we-believe the proposed design of PSN.89-001, Revision't, has improved appreciably over the Revision 0.
However, there are still concerns.'in some areas that need to be satisfactorily addressed by PSN in order for the VSC design to meet the requirements of 10 CFR Part 72. These concerns include:
(1)- Criticality concerns including design tolerances, and bench marking (2', shielding design calculations which are open to interpret & tion and their effect on dose estimate evaluation, (3) incomplete a
design criteria for the NSB 6nd the MSB calculation packeges and TR section-which do not provide an unambiguous Systematic and C0mplete audit trail, (4) numerous welds for components important to safety which do not comply with
-i ASME or AISC codes, (5)-the concrete aggregate for the VCC which is not-appropriate to meet the ACI Code or.NRC criteria for el::vated temperature concrete applications, and (6) the.VCC reinforcing cteel wh5ch does not meet j
the ACI code.
For details, please refer to specific comments in the ' enclosure.
NRC staff will be available to discuss these and any responses that you may 1
have at our naeting on May 3 and 4, 1990.
If you have any quer
- as, please call-K.C. Leu'at (301) 492 0696.-
.j i
Sincerelvi Original S!gned by' JohnP. Roberts John P. Roberts, Section Leader i
irradiated Fuel Section-F<uel Cycle Safety-Branch l
Division of Industrial and Medical Nuclear Safety -
l l
Enclosure:
Comments j
. Distribution:
PROJECT M-53 (KC/PSN)
PDR NRC File Center IMIF R/F
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Comments on Pacific Sierra Nuclear Associates' iopical Report for the Ventilated Storage Cask System Criticality Analysis, Design Tolerances t-1.
The' calculation of Ak, the reactivity uncertainty due to design and g
manufacturing geometric tolerances, was calculated for irradiated fuel assemblies.
Was a similar analysis performed for the'misloading of-unirradiated fuel assemblies? If it was, please provide the value of Ak.
If not, please perform the analysis for fresh fuel criticality i
g i
including the reactivity uncertainty due to design and e.anufacturing geometric tolerances.
Criticality Analysis, Axial Burnup Variation' 2.
Provide a description of. the axial model used for the calculation of Ak ' a How many axial zones were modeled?.What is the sensitivity of Ak, to the
-f number of axial zones in the model?
Criticality Analysis, Benchmarking
[
l' l-3.
Provide copies of the KENO-IV printed output for UO criticals No. 1, 3, 2
11, 17, and 21, and Pu0 -UO critical No. 1.
Are the observed 2
2 differences between the calculations and experiments statistical or systematic? How do-you know?
Hydraulic Roller Skip
-f 4.
(1.0-23 reference first question set) PSN still has not specified the maximum surface roughness.which can be accommodated by the Hillman rollers.
The concern is that the design' surface roughness as required by the Hillman Company is, in fact, satisfied by the opposing surface.
PSN should call this maximum surface roughness out so that it may later be incorporated by reference by a licensee.
PSN also did not specify the maximum load capac Ry of the rollers or the minimum hardness requirements by the opposing surface.
/7 l
2 Brittle Fracture of Ferritic Steel 5.
(3.0-11 reference first question set) PSN has not adequately addressed the issue of brittle fracture of the ferritic. steel shell of the HSB in j
the TR.
Independent assessment made by the NRC staff indicates that this
~
problem may not be significant,- but it should be properly documented..
The staff has consulted several references (NUREG/CR-1815,
" Recommendations for Protecting Against Failure by Brittle Fracture in j
Ferritic Steel Shipping Containers Up to Four Inches Thick," NRC Draft ll Regulatory Guide, " Fracture Toughness Criteria for Ferritic Steel 1
Shipping Cask Containment Vessels with a Maximum Wall Thickness of Four Inches, July 1989," and Regulatory Guides 7.6 and 7.8).
However, before-the recommendations suggested in NUREG/CR-1815 can be evaluated, PSN l
should provide (1).the minimum metal shell temperature that the MSB could attain during the worst case scenario of coolest spent fuel and col sst 3
day when there is any handling of the HSB that could lead to the w up accident and (2) maximum nominal-stress as defined by NUREG/CR-1815.
In other words PSN should bound the possibility ~of brittle fracture and show that it is accounted for.
(The maximum principal tensile. stress'at a point considering primary stresses and secondary membrane-stresses-(see definition in Regulatory Guide 7.6) should bel evaluated for the appropriate load combination i.e., pressure plus verticle drop or.
pressure plus horizontal drop.)
VCC Thermal Stress Analysis
. i i
6.
(3.0-24 reference first question set) The computer. output for the VCC.
j thermal stress analysis (Appendix 4.1 of the TR) appears.to be incomplete.
The temperature distribut-lon-for nodes 1-49'is missing,~as are the stresses for nodes 1-10.
Also PSN should provide an explanation why " iteration 15" and no other iterations were provided in the output'.
i 7.
Thermal stress on the concrete due to axial expansion of the cask steel liner appears not to have been considered.
The vicinity of the outlet ducts-appears to be particularly vulnerable to higher tensile stress on j
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the concrete.
This situation will be aggravated by high concrete temperatures due to heated air flow in the outlet ducts.
i 8.
Section 3.4.4.1.2, page 3.26, from examination of the calculation package WEP-101.2101 it appears that the VCC model consists of 433 elements and 615 modes, not'as stated in the TR 432 elements and 610 modes.
If this is correct, please correct the TR, 9.
Section 2.2.6.1, page 2-9, Table 2.2-3, page 2-10, Drawing VCC-24-001, Sheet 1 of 2.
The air outlet assemblies ~ of steel _ plate' constitute j
the equivalent of " piping and equipment" in the potential development of-reactions in the concrete due to greater temperature rise in the steel l
and different coefficients of thermal expansion.
The abrupt. corners and the attachment to the steel liner indicate potential for stress concentrations in the concrete.
Stresses due to.this design'under-normal, off-normal, and accident conditions have not been submitted yet.
They should be included as either thermal loads (T) in the combination of load expressions of ANSI 57.9,' para 6.17.3.1 or as R loads in the load g
combination expressions of.ACI 349.
Since the VCC is stated as being l.
designed to ACI 349, with checks made to confirm satisfying ANSI 57.9, l
the more stringent ACI 349 requirements for Rg_should be met.
It is further noted that minimum reinforcement requirements of ACI 349 para l
-7.12 are not met in the concrete faces-adjacent the air outlets (or
\\
inlets).
These are considered to be exposed faces.
The steel plate does not constitute concrete reinforcing steel; in fact, it acts ~to increase tensile stresses in the concrete.
l 10.
If the air outlet steel plate is intended to provide the-shear resistance to a tornado automobile missile impact above the outlet level'(this L
analysis has not yet been provided) then this resistance would preferably l-be provided by strength of the reinforced concrete section.
[It appears that sheet metal left-in place forms and adequate reinforcing steel would prove more likely to be satisfactory].
l 8
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4 VCC Thermal Stresses Analysis Revision 1 i
11.
Para 4.1 and Table 4.1-1, Section 1.0, pps 1 and 2'of 8.
The worst case temperature gradient is determined as occurring at 75 F, for " normal," "1/2 inlets," and "no inlet" cases.
However the gradient.
apparently is only the worst among the 4 cmbient Semperatures for which it is computed.
Since the gradient peaks betweer the extremes, it would
.be a coincidence if that peaking occurred at the one intermediate point used (which was not selected on the basis of peal ~ gradient).
Excursions should be conducted to.either side of 75 F to deter-ine if a higher gradient occurs, and that higher gradient should be' used in the calculations.
As'the table of. temperatures and gradients demonstrates, the relationships are non-linear.
It should not be presumed that the worst gradients for 1/2 inlets and no inlets will.also occur at the ambient temperature producing the worst normal case gradient.
VCC Thermal Analysis 12.
Table 3.41, page 3.25 and,Section 11.2.9, page~11-35.
It appears possible that the worst realistic thermal condition (accident condition)"
U could be at -40 F ambient temperature with 1/2 inlets blocked lor all
. inlets blocked over the 24 period.
Could you discuss or evaluate this condition and show what the delta temperature across the VCC is?-
VCC Stress Analysis 13.
Due to the expansion of the steel outlet liners under high temperature, 1
the. concrete might be subjected to tensile force in the radialTdirection.
m Please evaluate or discuss this condition.
Please add the additional reinforcing steel as needed around the outlets.
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5 VCC Accident Analysis 1
14.
11.2.4.2,.p. 11-19, etc.
The analysis of cask resistance to tornado missile isppact does not address impact of the automobile on the VCC above the air outlet level to ensure that shear forces across the air outlet sect, ion do not cause failure to such VCC deformation as to prevent ready retrieval of the MSB and contained fuel rods.
The discussion and analysis are required.
1 MTC Design Criteria 15.
(2.0-7, 3.0-23 reference first question set) The PSN response-that they consider the MTC to be a special lifting device and a sh'ielding bell to be designed by NUREG-0612 and shielding requirements is adequate provided that PSN incorporates the NUREG-0612 recommendation that ANSI N14.6, l
" Standard for Special. Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kg) or More for Nuclear Materials," 1986 be used for design.
See p. 6-4 of NUREG-0612.
The NRC. staff suggests.that PSN consider paragraphs 4, 5, 6, and 7, especially 4.l','4.2,-4.3, 4.4, 5.1.5,
.i 6.5.1, 7.1, and 7.2 of ANSI N14.6.
MTC Details, Thermal Analysis 16.
MTC drawings do not show the shaped elements which have.been added to aid heat conduction.
The geometry of these elements is required to i
validate the heat transfer analysis.
MTC Details Shielding 17.
(4.0-4 ref-
- e first question set) Since the steel angles are not present throughout the RX-277 region, what local thermal and neutron-dose rate hot spots will occur in the Transfer Cask from this configuration?
Will the presence of these angles affect the RX-277 pouring process and increase the likelihood of internal air void formation inside the RX-277-space?
6 Shielding
'f i
18.
(4.0-11 reference.first question set). Provide the vendor test data on hydrogen loss from RX-277 when being heated in a closed space.
19.
(S.0-9 reference first question set) PSN has not provided the revised SKYSHINE-II calculation package for the analysis of offsite air' scatter dose rates from the ISFSI.
Since.this calculation methodology and dose results have changed, this calculation package should be provided for review and independent verification.<
Use of ASME Code as Design Criteria for MSB 20.
References:
A.
NRC staff comments on the PSN TR for the VSC-System, September, 1989 8.
PSN Responses tv Reference (a), February 1990,iincluding Revision l'of TR and calculation packeges l
C.
American Society of Mechanical' Engineers Boiler and' Pressure Vessel Code,=Section III Division 1, Subsection 1
[
NC for Class 2 Components, 1986 Edition l-t L-D.
NUTECH Horizontal Modular Storage. System for Irradiated Nuclear Fuel NUHOMS-24P NUTECH-002 Revision 1 O
This topic concerns appropriate use of'the ASME. Code as a design code'for the-l:
MSB.
Although PSN does not intend to obtain an N-Stamp for the MSB, they have f
stated that they are using reference C as their design code.
Appendix C of a
- Reference C discusses various elements which should be included in a design 4
report.
PSN'has included some of these elements in their Topical Report Revision 1 and numerous calculation packages which they submitted in
Reference:
B.
However the NRC staff has found that many of these analyses are incomplete.
3 and do not satisfy the requirements of the chosen design criteria.
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The Topical Report, when prepared in accordance with the requirements of 10 CFR 72, serves in lieu of a " Design Report." A properly prepared TR should contain most of the elements that are required by Appendix C of
~
Reference C.
PSN has not yet achieved this level of preparation in their Revision 1 of the TR.
21.
In an attempt to facilitate the review of the PSN ISFSI design, the NRC staff nas prepared a synopsis of some = relevant sections of the ASME Code Section III
~NC-3000 Desi n.
0
-NC-Sill Loading Conoitions (Required at NC-3212)
-i The following load conditions must:be evaluated showing' design '
calculations:
a.
Impact forces - external cause i
Applicable to MSB - (1) drop service level-D-
- (ii).off normal service level C case not evaluated by PSN 2 -f t/sec impact crane.
(
Reference para 7.2.3 of ANSI-N14;6 and para 5.1.1 of NUREG-0612.
These impact = forces will bc imparted to'the MSB from the MTC.
b.
Impact forces rapidly fluctuating internal pressure
.not' applicable j
t Weight of components including pressure. Service Level A, B and c.
C for. normal and off-normal pressure cases.
Not complete not' combined by PSN d.
Superimposed loads i
Service = Level A, B, C, and D Not complete - load cases inadequately defined and inadequately J
evaluated.
Normal operation handling load was not defined by PSN.
A handling load could be due to transportation en route to the-storage pad and could be taken as +0.5g in all directions simultaneously.
Whatever PSN selects as its design 1
criteria, this particular load must be superimposed with dead load, internal pressure, and thermal loads for service level A and/or B.
This is an additional case, distinct from the impact lor.d in (a) above.
!j l,
8 e.
Wind, snow,. vibrations, earthquake Service Level C 7
o Not complete - load-cases -inadequately evaluated - no.
summarization - no load combination f.
Reactions of supporting lugsL H
Not adequate.
Why did PSN select a hoist ring with a safety factor of 9.9 instead-of.a safety factor greater,than 107 (Page 3 of 33 of WEP-101.1101 Rev. 1) Lifting lugs should be designed to meet ANSI N14.6.
J g.
Temperature effects Service Levels A, B, C j
PSN has not adequately defined the various " service" and i
" design loadings" per. NCA-2142.1 and 2142.2, nor has PSN adequately evaluated individual thermal cases nor combined the secondary stresses for all appropriate load combinations.
NC-3112 Design Loads This section of the code defines'" design loads and referencesLNCA-2142.1.
As an example of what appropriate here, see Reference 0 j
NUTECH Table 3.2-Sa and.establisn. comparable table for the PSN MSB.
I I
NC-3112.4 Design Allowable Stress Values There are basically two path's which the designer maystakeLin this d
paragraph. 'One path is NC-3300 (standard design) which uses lower stress allowables of Tables I-7.1in Appendices.
The other path '
permits use of higher stress allowables of Tables I-1.0 if' NC-3200 '
(alternate design) is used.
Based on Table 3.3-1Lof:the PSN-TR, PSN
.1 i
is using NC-3000.
See NUTECH Table 3.2-6 for appropriate designation for stress-type (i.e., primary membrane, bending and 1
secondary), definition-of service levels and allowable stress 1
intensity levels.
PSN needs to prepare' comparable tables.for the MS8.
NC-3113 Service Conditions-The parepraph.of the Code references-NCA-2142 and NCA-2142;4(b).
PSN needs to' adequatelyf define service. conditions. and limits' and -
1 then adequately show for all appropriate; service conditions,..
i 3
including load combinations, that allowables are~ met. ;See Tablest 8.1-1, 8.1-4, 8.1-7, and'8.1-7a, of Reference O for normaltand off-normal cases.
For examples of accUnt cases. see Reference D Tables-8.2-1, 8.2-7, 8.2-9, 8.2-9a, and 8.2-9b.
PSN needs to prepare comparable tables for the MSB.
-..~.x E
9 NC-3200 Alternative Design Rules for Vessels NC-322.1 Scope (a) ' States that this article may be used as an-alternative to NC-3300, and that if NC-3200 requirements are met, stress intensity values of Tables I-1.0 (Appendices) may be;used.
This-seems to be path PSN chose.
(c) If complete r,ules in'NC-3211.1(b).are not provided (they..are;
~
not) then designer ~shall. design according to loadings of NC-3212 (References 3111 above) and Design Specification which PSN has not clearly identified.
It shall beidone in accordance with Appendix XIII-for all applicable stress categories.
See Appendix XIII.
a (d) Special requirements of-the following paragraphs' apply:
(i) NC-4260 weld joints designed to NC-3200.
PSN:has not complied with this requirement for the category C weld joint at the top of the MSB.
.'It must be a full ~
See Figures f40-4265-1 or NC-4265-2' for acceptable' full penetration corner welds.
(-11) NC-5250 Examination of_ welds designed to NC-3200.
PSN has l
t complied with this except for;the, top lid /shell. weld which will.be tested via helium leak test.
configuration for the top weld is' currently inadequate.
The Code is predicated 'on examination of full penetration.
and/or a suitable double seal 1 weld configuration.
1 (iii) NC-6221, 6222 pressure requirements.
(e) Design Report comply with NCA-3550.
The PSN TR does not s x cly with this presamably because the TR shall be substituted ir lieu of this. 'However the basic intent and' content of NCA-3550' should be covered in the TR.
The PSN submittal Revision 1 does not fulfill the intent and content of NCA-3550.
22.
The following excerpts from Appendix XIII are provided to indicate general definitions which should be incorporated:in TR.
1 XIII-1440 XIII-1142 General Primary Membrane Stress intensity produced by design-internal pressure and other specified Design Mechanical Loads exclur"ng all secondary and peak stress 4
P, < kSm using Table I-1.0 for Sm
- j
4 N
10 XIII-1143 Local Membrane Stress intensity produced by' design pressure and 4
other specified design mechanical: loads excluding thermal and peak P 5 1.5 kSm L
' XIII-1144 General or local primary membrane plus primary bending produced by design pressure and specified design mechanica1 loads q
excluding all secondary and peak stresses
-j 1
(P,'or P ) + PE 1 1.5 kSm g
See NC-3217(b) also note the k factor is-not permitted in. Level D Service Limits.
J XIII-1345 Primary Plus Secondary Stress. Intensity i
Highest value derived from general or local _ primary membrane.
l stress plus primary bending stress plus' secondary stresses produced by pressure and specified mechanical. loads and general thermal effects Pg+Pb + Q < 35m Note k factors are given in Table NC-3217-1.
i Fron. Table NC-3217-1 Service Limit
.k 1
Design 1.0 Level A
.1.0 3
Level B 1.1
~
l Level C 1.2 i
Level D 2.0 Test 1.25-hydrostatic test From Table NC-3217-1, when an analysis is performed in accordance with NC-3211.1(c) the stress limits =of Appendix F may be applied.
See Table F-1322.2-1 for more-information.-
~0
- 23. The following comments on welding'are offered to help resolve the currently unsatisfactory weld joint at the top 1id/ shield of the MSB.
~
NC-3257 Welded Joints Subject to Bending Stress (Applicable to PSN corner weld due to pressure and drop cases etc.)
NC-3357.
The subject is fillet velds to reduce stress concentration ;(may or may not be applicable to PSN design).
i
,8
i(
j 11' NC-4243 Category C Joints in' Vessels Shall be full penetration joints as shownLin Figures NC-4243-1 or NC-4243-2.
The PSN weld design for the top lid /shell does' not comply with the various examples shown in.the above figures as well as the Figures NC-4265-1 or'NC-4265-2.
However.4243.1 '
1(b) shows a possible configuration which would require, minimal changesEto the currentsPSN, design.
This configuration could be incorporated by the basic PSN design in order to avoid using a-double seal weld.
Conclusion to Concern About' Inappropriate Use of ASME Code 24.
In tiew of the.many deficiencies in.the use of the ASME' Code cited above, the NRC staff.cannot currently approve the-Revision 1 of-the-PSN-T!L This is not to imply that the' design is inadequate (except forf the top lid /shell weld design).
In many instances PSN has generated adequate.models, but has not-followed through in methodology as required by the Code.
In other cases the modeling is incomplete or inaaequate (see following discussions on vertical drop;and pressure models).
+
.The NRC staff suggests that other approved ISFSI TopicalLReports, which are in the public domain, could be used by PSN as examples of appropriate-
~
use of the ASME Code.
I Design Criteria for Ambient Temperature 25.
Criteria should be stated which a site must satisfy in terms lof ambient temperature.
For example, yearly average temperature less'than 75" and-the maximum temperature for the 50 percent probability level (two year U
I recurrence) less than 100 F wor:ld be considered for acceptance criteria based on the analysic presented in the Topical Report.
(See NUREG/CR-1390, Probability Estimates of Temperature Extremes for the' Contiguous-4 United States).
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12 Vertical Drop of MSB 26.
The NRC staff has reviewed WEP-101.1101 for the 30 foot drop analysis as well as the Revision 1 of the Reference TR and finds that the' treatment s
is superficial. 'The ASME Code is very specific about what is required
'or qualification'under Service Level D, and PSN has not satisfied these
-.i requirements.
The simple P/A approach ignores bending at the shell/ lid interface at both the top and bottom lids.
Furthermore, no consideration s
has been given.to the weld design at the top.
No loao combinations were evaluated.
PSN should evaluate primary membrane stresses and primary-membrane plus primary bending stresses at. top,and bottom of the shell at support ring junction, for the bottom plate, for theisupport ring, and=
structural lid. 'Alsosee-followingcorhmentsonmeshsizefor.FEMmodel.
-The mesh size PSN has selectel will not properly predict stresses in j
4 vacinity of the weld at the top of the shell.
j 27.
Please discuss NC-3133.6 as it relates to the PSN design for vertical y
drop case or other axial compression loading.
28.
If PSN elects not to use the Code to eva,1uato the storage sleeve assembly, it should choose an appropriate approach, cite an; appropriate
}
code and criteria,-and evaluate buckHng as well as. compressive stresses, and weld requirements.
s t
Horizontal Drop of MSB 29.
Please send a copy of WEP 109.002-15, the methodology PSN used to determine deceleration levelt.for the drop accident.
Apparehtly the NRC staff does not have this package.
P 30.
The coarse element grid in the vicinity.of discontinuities.does.not model stresses accurately enough.
There is no possibility of-predicting stresses in.either top or bottom weld considering1the element size for, element numbers 6 or 386, (the bottom or top plates elementA) or for element-number L
369 (a shell element which joins the top plate).
PleasedeviseamearIs.
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+
13 of predicting primary membrane as well as primaty plus bending stresses for the welds and for areas of discontinuity.
{
MSB Critical Pressure Analysis 3
31.
The NRC staff has reviewed the calculation package WEP-101.1101 for the MSB pressure analysis and find that the element mesh is too coarse to predict stresses in the welds at the top and bottom of the MSB.
The upper joint is nat full penetration, as discussed elsewhere.
However even provided that an appropriate weld joint is specified, the element mesh size for the shell and for the structural lid in the vicinity of the welds is too large.
Please devise a neans of predicting primary membrane as wni as primary plus bending stresses for weld and for areas of discontinuity.
MSB Dry-out Procedure (12.0-3) 32.
Reference para 12.2.2.2 in TR.
PSN bas apparently decided not to follow the guidance as suggested by the NRC staff in the first set of questions.
The PNL-6365 document was suggested as a reference by the staff because it cites empirical data about procedures which produced a storage cask environment which proved to be benign to spent fuel assemt,1ies.
The second evacuation and back fill with helium provides additional assurance of low oxygen concentration.
Considering the fact that the PSH design for the MSB specifies carben steel insteed of stainless steel, the staff find a second reason for recommendir.g a second evacuation to 3 Torr to ensure purity of the cover gas.
p-DRWG MTC-24-008 Sheet 1/1 N
33.
The cover plate design for the lifting trunnion does nat follow recommendations of ANSI N14.6-86 paragraph 4.4.1 in that holes and l
pockets should be avoided in order to reduce radioactive contaminants.
t 2) 4 k
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14 1
DRWG NTC-24-008 Sheet 1/1 t
34.
Paragrar <.3.1(2) of ANSI N14.6-86 discusses galling as a design consideration.
PSN has specified a " cylinder cover," but has not specified what it is. Please discuss and show that galling has been adequately addressed.
Table 1.2-1 and Appendix 2.2 Paragraph 3.4.4 35.
PSN cites that the principal design criteria for=the MSB is-ASME Section III, Division 1, Subsection NC in Table 1,2-1.
However, PSN' deviates fro:a
~!
this in paragraph 3.4.4 of Appendix'2.2, where PSN citesSection VIII for f
material requirements.
The NRC' staff does not accept this' mixing of. criteria.
i Further exception to the stated design-criteria is taken by PSN in paragraph 3.5.2 of Appendix 2.2.
DRWG MSB-24-002 Sheet 1/2 Note 4 and Statement-3.6.2 of Appendix 2.2 36.
PSN states in Note 4 that all girth and' longitudinal. welds of the'MSB be-radiographed and also states "See' Specification.for MSB-24-001."- However Statement 3.6.2 omits to mention the type of nondestructive testing of the longitudinal and girth welds.
Longitudinal welds'shall be.
9
- r categorized as A (see NC-3351.1)'.
Girth. welds shal.1 be categorized as C i
j (see NC-3352.3).
The Categories A and C shall be examined by-radiography 1
L per NC-5251 and NC-5253(b) respectively.
PSN should specify that.
i radiographic tests shall be carried out in conformance;with the ASME L
Codes.
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a 15 Verification of Fuel Inventory t-37.
Para 8.1(4), p. 8-5, Para 9.2, p. 9-2, Section.12.0 The system as described doec not provide for compliance with the 4
- requirement for physical inventory of all spent. fuel at intervals not to exceed 12 months (10 CFR 72.72(b)).
This might be accomplished by:
- 1) inventory of ir. tact VCC-by serial number and verification that a seal.
l I
weld of the cask lid or welds of its bo'at heads were intact, if such welds were inclitded (not currently); or, 2) another alternative if found to be acceptable to the NRC,.
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i DRWG VCC-24-006iReinforcing Stee1 in VCC
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- 38. -There should be 3 lengths of vertical rebars instead'of only 2 (Items 3 and 4).
About 1/2 of' Item 4 should extend to VCC bottom instead of only 3
U+ top of skid channels.
Note 11 should.only have to.pply to the longer bars extending from air outlets to' VCC bottoni. /The cask corners are not "hearily reinforced" per paragraph 1.2.1.2, p.1-12 of TR except the Item 3 rebars.
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- URWG VCC-24-006 Reinforcing Steel 'in VCC-3
- 39. The reinforcing steel.does: rot' meet the requirements'for minimum i
reinforcement (ACI549,' Section-7.12) in that-there is no reinforcement at the inner face of the concrete cylinder cr its base or adjacent the air I
outlet surfaces.
'[ Note:
a face cast against steel plate is still W
'" exposed" and thereby requiring minimum reinforcement in accordance vrith the code.3 The. design must include two-way reinforcement at these inner surfa:es.
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16 Welding 40.
(1.0-20 reference first question set) The ANSI 57.9 standard is not at issue on the subject of radiographic weld inspection.
Use_of radiographic testing is required by the design code which PSN has stated that they will use.
The applicable paragraph in ASME Code for~Section III are NC-5320 in general, and NC-3351.1 for longitudinal category A welds and NC-3352.3 category C welds.
See comment relative;to drawing.
MSB-24-002 in a later chapter.
41.
The weld between the trunion and the outer shell of the HTC was not evaluated.
I 42.
The weld connections (see DRWG HRS-24-001) an'd calculations (WEP 109.002-
- 17) between the 5 W 16 beam and the plate of the rollers have E
omittedthebendingmomentknducedbythelineofactionofthetowing vehicle and the line of reaction through the weld joints.
Please draw a freebody diagram of the HRS and tow vehicle and reanalyze to resolve this
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Concern.
43.
The tield design for the structural lid valve cover-(Reference DRGW MSB-24-002 Sheet 1 of 2) is not satisfactory.
This.olug is part of the structural lid and according.to the ASME Code it must be a full penetration weld.
This joint is as important to the structural integrity of the MSB as all the other welds which form the confinement boundary.
It must be full penetration or a double seal design.
Note that the weld design for the top lid is currently unsatisfactory also.
e 17 44.
The following welds which are in components not covered by the ASME B&PV Code do not comply with AISC Specification 1.17.2 (Manual of Steel-Construction) for minimum size fillet welds or minimum effective throat thickness for partial penetration groove welds:
Drawing Weld Min Size / Throat VCC-24-001 Sh 1/2, Detail A 1/8" Fillet
_5/16"'
VCC-24-002 Sh 3/3, Detail A 1/8" Fillet S/16" 1/8"-VEE 5/16" VCC-24-002 Sh 3/3,- Detail B 1/8" Fillet.
,5/16"
- VCC-24-002 Sh 3/3,. Detail B 2-6 1/8" Square Groove 3/8" VCC-24-003 Sh 1/1, " Inlet Assy - Top View" 1/8" Fillet 5/16" VCC-24-003 Sh 1/1, " Inlet Assy - Side View" 1/8" Fillet-5/16" VCC-24-003 Sh 1/1, " Items 1,2,3,8 and 9 - Top. View"-
1/8" VEE 3/16"L
- VCC-24-003 Sh 1/1, " Item 4 with Item 5 - Side View" 1/8" VEE..
-3/16"
- VCC-24-004 Sh 1/1, Side View 1/8" Fillet 3/16" VCC-24-004 Sh 1/1, End View 1/8" Fillet-3/16" VCC-24-008 Sh 1/1, Detail A 1/8" VEE.
'5/16"-
1/8" Bevel 5/16"-
MTC-24-001 Sh 1/2, Top View 1/8" Fillet 1/4" MTC-24-001 Sh 1/2, Detail A
-1/8" Fillet.
1/4" MTC-24-001 Sh 2/2, Section C-C 1/8" Fillets (2)-
1/4" 1/2" Bevel 5/8" M1C-24-001 Sh 2/2, Section D-D 1/2" Bevel 5/8" MTC-24-002 Sh 1/1, Section A-A
.1/8" Bevel' 1/4" 1/16" Square, NLT Pipe 1/16" Fillet Thickness (0.109")
MSB[ SIC]-24-007 Sh 1/1, Section View 1/2" VEE 5/8" MTC-24-008 Sh 1/1, Section View 1/8" Fillet Item 3'to Item 2 5/16" 1/8" Fillet Item 4-to Item 1 3/16" DRWG VCC-24-008, Sh 1/1, Detail A 45.
The symbol for a full penetration square butt weld is shown for-a 6" deep joint.
This should be corrected.
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18 DRWG VCC-24-008, Sh 1/1, Detail A i
46.
It appears that item 4 should be joined to item 4, but is not.
~i DRWG MSB-24-002, Sh 1/2, Detail A
Specifically, full i
i penetration welds are required for the pressure vessel and its covers and-
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the pressure seal.
Components and welds should be redesigned to: comply-with the Code.
-l DRWG MTC-24-002 Revision 1 48.
Please provide the thickness of the plates'shown on the drawing ~at the.
f bill of material tables.
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Surveillance i
- 49..It is possible that over the long term debris or dirt may accumulate
_l within the inner air passages which cannot be observed by periodic-inspections of the air inlets and outlets.
More comprehensive j
inspections at'less frequent intervals ever the lifetime of the caskl are desirable.
PSN should propose such less frequent, but more-comprehensive inspections as a part of the conditions for operation.
50.
(Reference p. 12-4 of TR Revision-1) The statements in this section are.
somewhat contradictory.
If Section 4.0 of the TR has conservative thermal analyses, why would a "somewhat poorer than predicted" cask performance result if the air outlet temperature exceeds the ambient temperature by more than 110 F?
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L 19 Radiation Exposure 51.
Revised calculations were performed and used by PSN to construct a new dose-versus-distruce curve (Figura 5 4-3).
Although the previous results were considered " conservative", the revised results are substantially higher.
Obviously, these results are very sensitive to assumptions, data, and methods; and the "conservativism" of the previous results must be questioned. -What_are the specific changes in assumptions,' data, and methodology which resulted in'this increase? It would be extremely useful if a calculation package could be provided in which all assumptions, data, and-methods are clearly stated.
Manual calculations which are based on code results must also be included.
For good examples-of ISFSI arrayfdose-versus-distance calculation packages, please -refer to.
resent submissions by Nutech Engineers and Carolina Power and Light for the Brunswick ISFSI application.
Decommissioning 52.
PSN states that the HSB will-be placed in a decon pit for cutting of the lid.
However, no mention is made of the control of effluent from the HSB.
See p. 8-5,6 of Revision 1.
Concern is there is good reason to suspect contamination inside the MSB.
P. 2-17 of Revision 1 states,~"the MSB interior is expected to be highly contaminated with fuel crud."'
Fuel Clad Temperature 53.
There is an inconsistency in temperatures expected for the fuel.
In Section 3.5 (p. 3-44)-the temperature is given in the range of 378-450 C, U
while in several other places the range is given as 370-400 C.
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20 Protection of Fuel Assemblies
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54.
(Reference p. 2-13 of TR Revision 1) The helium leak appears to be low.
What is the void volume that PSN used for the MSB?-
MSB Thermal Analysjs
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i 55.
Thermal profiles within the MSB were not provided.
(p. 11-11) Please provide.
Appendix 2.3 MTC-89-001, Revision 1
.56.
Please provide a drawing MTC-24-011, Revision 1, Cask Lifting Yoke.
It-was apparently omitted in previous transmittals to the NRC, Also_ note, no mention of the yoke was made in Section-1.1 of this appendix.
Why was it omitted?
Appendix 2.2 MSB-87-001, Revision 2 i
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57.
Section 3.2 of the above appendix has'all drawings incorrectly identified l
both as to drawing number and title please correct.
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