ML20134M796
| ML20134M796 | |
| Person / Time | |
|---|---|
| Issue date: | 08/29/1985 |
| From: | Roberts J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Rouse L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| REF-PROJ-M-39 NUDOCS 8509040270 | |
| Download: ML20134M796 (132) | |
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Distribution: please return concurrence copy to FBrown SS 396 AUG 2 S 1585
/~ Pro.iectji-39 PDR NMSS R/F SCornell/GBeveridge Project F-39 FCAF R/F JCounts JRoberts FBrown JSchneider Reg 11 Leland C. Rouse Chh" l Licensing Branch NEMORANDUM FOR:
AdvancedFuelandSpenNue Division of Fuel Cycle and flaterial Safety FROM:
John P. Roberts Advanced Fuel and Spent Fuel Licensing Branch Division of Fuel Cycle and Material Safety
SUBJECT:
MEETING WITH NUTECH, INC.
Date and Time:
August 27, 1985: 9 a.m.
Location:
bth floor conference room. Willste Building Silver Spring, FD Attendees:
NUTECH NRC J. Passey J. Roberts J. Schneider W. Macnabb (SAIC)
Purpose:
A discussion of NUTECH's approach to responding to NRC staff coments on NUTECH's topical report for a concrete redule design independent spent fuel storage installation (ISFSI).
Discussion:
BUTECH had prepared preliminary responses (see enclosure), and these were examined and discussed.
Issues involving criticality, therr.a1 hydraulic, and shield penetration calculations remain open for further discussion in a full neeting between NRC and NUTECH staff tentatively scheduled for Septenter 10-11, 1985.
In the interim, NRC staff will consider NUTECH's preliminary responses. NUTECH will provide responses in finished form at our next meeting and expects to submit a revised topical report on Septerher 30, 1985
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.lelum F. Itoberta 8509040270 850029 PDR PROJ John P. Roberts M-39 PDR Advanced Fuel and Spent Fuel Licensing Branch tvWer As ted
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DETAILED. COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
1
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF INSTALLATION 1.1 Introduction A helium storage atmosphere is specified within the DSC.
What measures has NUTECH taken to assure that the helium initially placed in the DSC will not leak to the atmosphere?
What is the maximum leakage rate consid-ering the integrity of all welds and diffusion through the canister?
What measures are available to monitor the DSC atmosphere composition?
RESPONSE
NUTECH has included a number of measures in the design, fabrication, and operation of NUHOMS that will ensure the helium initially placed in the DSC will not leak to the atmosphere. Nor will the oxygen in the atmosphere leak or diffuse into the canister.
The longitudinal and bottom closure welds of the DSC are designed and will be fabricated in accordance with the ASME Boiler and Pres-sure Code, an internationally accepted code for the de-sign and fabrication of leak tight containment ves-sels.
The design of the DSC also includes redundant closure welds on the top.
All welds that are on the pressure retaining boundaries of the DSC will undergo nondestructive examination.
The nondestructive exami-nation includes:
Ultrasonic or o
All shop welds on Radiographic Testing pressure retaining boundaries o
Secondary closure welds Dye Penetration Primary closure weld Helium Leak Test Primary closure weld In regards to dif fusion, discussions with scientists in the helium industry (at Union Carbide and the U.S.
Bureau of Mines' Helium Operation) indicate that as a practical matter, helium does not diffuse through steel or stainless steel.
The scarcity of references in the NUH-001-1 1.1 l
DETAILED COMMENTS ON NUTECH TOPICAL REPCRT NUH-001 literature seems to bear this out.
It was further dis-cussed that significant helium diffusion only occures under condition of a
hard va cu um.
In the Metals
- Handbook, R.M.
Parke flatly states that " Helium does not c.if f use through solid iron".(A)
Howe ve r, since the reviewers' concern with diffusion is evident, we have-8 l ulated the diffusion rate to be on a
the order of 10 g moles / year at nominal design condi-tions.
Diffusivity for this system varies exponentially with temperature so that in an accident condition with elevated temperatures, the diffusion rate could be expected to increase by as much as 3 orders of magni-tude.
The leakage rate would rema in, howe ve r, of no practical consequence.
1 NUH-001-1 1.2 t
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DETAILED COMMENTS ON NUTECH j
TOPICAL REPORT NUH-001 1
QUESTION:
2 1.2.1 General Description of Installation Even though it may be possible to design a single stand-t alone horizontal storage module (HSM) or physical ar-
{
rangements in other than a 4x2 array, the design con-tained in the current Topical Report is limited to a 4x2 array (see p. 8.2-13).
The Topical Report should clear-ly state which arrangements (s) are being proposed for review and maintain consisency thoughout the document.
4
RESPONSE
The 4x2 array is the only design which is to be reviewed in this Topical Report.
The Topical Report will be re-vised so that consistency is maintained throughout the
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document.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i.
QUESTION:
3 1.2.2 Principal Design Criteria 1.
Are there limits on the kind of external atmosphere allowed for this system in order to control cor-rosion of exposed surf aces or to control growth of algae or other vegetative matter within the HSM?
Please discuss.
RESPONSE
Because of the corrosion resistent properties of com-ponents or the coatings used in the NUHOMS system and the hot, drying environment which exists within the HSM no limits on the kind of external a_tmosphe re ars re-quired.
All components are either 304 stainless steel or ATSM 36
- steel, galvanized or coated with Carbo Zinc 11, an inorganic, zinc based corrosion protection coating.
The interior of the HSM is all cement and steel and is void of any substances which would be conducive to the growth of organic matter.
The design has been ' changed to include a drainage line in the air ' inlet chamber.
The drainage line will drain any water which enters through the air inlets.
The air outlets located on the roof of the module are designed to be water tight.
Any water entering into the HSM cavity, by this route will evaporate long before any vegetative matter could estab-lish. itself and disrupt the air flow through the HSM.
The evaporation will be rapid because the module floor is 50 to 70*F above the incoming air temperature due to the heat. radiated from the canister.
Additionally, as i
docume nted in many studies on the environmental effects of radiation, radiation is not conducive to ' the growth of organic matter.
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QUESTION:
4 1.2.2 Principal Design Criteria 2.
No mention is made of the duty cycle allowed for the materials (particulary concrete).
Please provide some discussion of how the duty cycles are f
addressed or why they do not need to be addressed.
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RESPONSE
According to the ASHRAE Handbook, 1981 Fundamentals, the largest mean daily change of temperature in the United States occurs in Reno, Nevada and is 45'F.
Assuming l
this maximum temperature change occurs every day for 50 l
years the HSM will experience 18250 thermal cycles.
The maximum moment caused by normal operating thermal loads is 2496.2 k.
in.
This loading is only 62% of the ulti-mate strength.
Consequently, refering to the S-N curve of Figure 6-41 of the handbook of concrete engineering by Fintel, the number of cycles before failure will
- occur, is approximately 10,000,000.
Since only 18250 cycles will occur in the 50 year life, fatigue failure i
of concrete is not possible.
To show the negligible effect of the duty cycle on the fatigue strength of the DSC the fatigue analysis of section 8.2.10 is reeval-uated.
Overly conservative assumptions are made and daily thermal / pressure cycling is included in this analysis.
i Conservatively, the operating pressure loading is as-sumed to change from zero to operating pressure each day.
Secondly, accident pressure loading is assumed to occur once a year for each of the 50 years of design life.
One major seismic event is assumed to occur with 100 damaging cycles.
For seasonal -changes a wr =
175~. 7'F is used which envelopes' the UT f rom the -40 *F to
+125'F inlet temparature load cases.
As stated pre-viously the largest daily range in the United States is 45*F.
The WT is applied each day for the 50 year life-t ime to account for daily temperature cycling.
All of' the five (5) loadings are combined in a histogram.-
ASME Figure I-9.2 is used to de te rmine the fatigue usage factor for the DSC shell.
A value of 0.21 was ' calcu-lated; therefore, the DSC shell is adequate for cyclic loadings.
5 NUH-001-1 1.5
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i DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
5 1 2.4 Safety Features What are the values for cask movement, cask head and truck transport which were not included in Table 1.2-47 l
RESPONSE
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There are no values associated with cask movement, cask j
head, and truck transport.
These are operational capa-bilities which an system must include in order te be compatible with the NUHOMS system.
Table 1.2-4 will be l
revised to indicate N/A for these parameters.
Some of i
the qualitative parameters which must be met are listed below.
The combined weight of the cask 'and the loaded DSC must not exceed the lifting capacity of the sites existing j
spent fuel handling crane.
i The cask must be capable of being rotated by the sites existing crane from a vertical position to a horizontal position.
The cask lid must ' be removable while the cask is in a horizontal position.
j The size of the cask and the combined weight of the cask and the loaded DSC must not be so excessive that they l
could not be economically transported to the HSM by a
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tractor-trailer arrangement.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
6 3.0 PRINCIPAL DESIGN CRITERIA 3.1.1.1 Physical Characteristics Based on the data on pages 3.1-1 and 3.1-3 it appears that NUTECH has used the Westinghouse fuel assembly array of 15x15/204.
Examination of the data in Table 3 1-2 reveals that several fuel arrays are not enveloped by the 15x15/204 array, namely the B& W 15x15/208, the B&W 17x17/264, the CE 15x15/208, 212 and 216, and the CE 16x16/224 to 236.
The text and the tables should be consistent and any fuel assembly types not enveloped by the design criteria case should be deleted or the design modified to envelope all the listed fuel types.
In addition to these comments, are there any design para-meters of the reference fue,1 assembly physical charac-teristics which would lead to a
non-conservative analysis?
RESPONSE
The main structural criteria for choosing an enveloping fuel assembly is the weight distribution per spacer disk Spacer disks are located at fuel assembly grid spacer locations to assure proper support.
The weight per spacer disk is obtained by multiplying the distributed weight times the maximum length between spacer disks.
Utilizing this procedure, the Westinghouse 15x15/204 envelopes all fuel assemblies in Table 3.1-2 except the combustion engineering 15x15/208, 212, 216.
This PWR fuel assembly will be deleted from Table 3.1-2.
The following table shows the weight per spacer disk for the various assemblies.
Fuel Assembly Wt. Per Spacer Disk (kg)
Westinghouse 15x15/204 106.56 Babcock & Wilcox 15x15/208 87.73 B&W 17x17/264 90.41 Combustion Engin. 14x14/176 64.14 C.E.
15x15/208, 212, 216 127.46 (Env#T5 ped)
Westinghouse 14x14/179 93.15 Westinghouse 17x17/264 101.66 NUH-001-3 3.1
DETAILED COMMENTS OM NUTECH TOPICAL REPORT NUH-001 Some of the aforementioned assemblies weigh more than the Westinghouse 15x15/204, the maximum increase is 83.8 lb per assembly or 583.1 lb per canister.
This causes a dead weight increase for a loaded canister from 21236 lb to 21819 lb.
This 2.7%
increase will cause minimal stress increase on the DSC shell, DSC support assembly, i
and the HSM.
Since significant safety margin exists for these components, the 2.7% increase is negligible.
3 No other design parameters exist for the referenced fuel assemblies which could lead to non-conservative analyses.
Consequently all fuel assemblies in Table 3.1-2 are enveloped except the Combustion Engineering 15x15/207, 212, 216.
This array will be deleted from the table.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 1
QUESTION:
7 3.1.2.2 Handling and Transfer Equipment l.
If NUTECH intends to assume that the transfer cask will be handled in the reactor building and during loading of the cask onto the transfer vehicle by a single f ailure-proof _ crane,
it should state that criterion in this section.
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RESPONSE
The transfer cask 'and irradiated fuel will be handled within the fuel pool and reactor building in accordance with the requirements and criteria already appro"ed in the sites FSAR and in accordance with the sites existing procedures.
Information on cask and fuel handling pro-cedures must be provided on a site specific basis.
These procedures must show that the transfer. cask will be handled in a manner that will not present a safety hazard.
That is, the impact -energy absorbing properties of the cask in use and the limitations on the drop height are such the resulting impact deceleration is enveloped by the values given in section 8.2 of the Topical Report.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
8 1
3.1.2.2 Handling and Transfer Equipment 2.
In order for the five foot horizontal drop criterion to be
- valid, NUTECH should specify dimensional criteria for the cask skid and the transfer trailer to show that the maximum con-ce ivable horizontal drop height is five feet of less.
RESPONSE
The type of cask and transfer trailer is..s i t e specific and should be included in the site's FSAR. The Topical Report will be changed to state that the limitation is the enveloping deceleration load (48gs) for which the DSC was designed (section 8.2), not a drop height.
The five foot drop criterion previously specified corres-ponds to the maximum deceleration which the DSC is capable of withstanding in a GE IF-300 cask dropped on an unyielding surface.
Different cask, trailers, and ground conditions at various sites may significantly effect any actual drop height limitation.
Indeed, drops on actually existing surfaces on reactor sites wil yield much higher allowable drop heights.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
9 3.1.2.2 Handling and Transfer Equipment 3.
What vertical acceleration factors have been mea-i sured or will be assumed for design criteria for the transportation of the GE IF-300 cask loaded with the NUTECH DSC while being transported on the trailer?
Show that these accelerations have been enveloped by the five foot drop analysis.
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RESPONSE
The vertical acceleration which the DSC will be sub-jected to while in transport to and from the HSM will depend on the type of
- trailer, travel speed of the trailer, and type of road surface.
All these factors i
are site specific.
- However, 10 CFR71.31 (d)
(1) requires the vertical acceleration design criteria for cask tie downs to ' be 2g.
If this same criteria is assumed for the trans-portation of the DSC to and from the HSM, then this design criteria is bounded by the 34g vertical acceleration term associated with the drop accident analysis.
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l DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
10 3.2 Structural and Mechanical Safety Criteria Pages 3.2-4, 8 2-53 and 8.2-56 show inconsistent maximum internal pressurization levels of 44 and 39.7 psig.
Also, associated with these pressures are inconsistent temperature levels of 423* and 413*C.
NUTECH should state what the actual design parameters are and use them 4
consistently throughout the Topical Report.
1
RESPONSE
i New helium filling procedures have been developed.
The table below list the internal pressure of the DSC, both under both normal and abnormal conditions (100% rod fill gas release and 253 fission gas release)
The Topical Report will be revised to include these new pressures and checked to make sure the numbers used are consistent i
throughout the text.
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Table 3.2 III Air.
Average llelium lielium Partial Pressure Temperature Temperature Pressure Fission of Fill Gas Total Pressure Case
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(Psia)
(Psia)
(Psig) 1
-40 315 14.2 20.92 35.12 20.42 2
70 389 14.5 22.92 37.42 22.72 3
125 429 15.2 23.99 39.19 24.49 4
Inlets 502 16.4 25.97 42.37 27.67 Plugged 5
Complete 775 21.05 33.33 54.38 39.68 blockage of inlets and outlets 3
(1) 100% of rod fill gas and 25% of fission gas released.
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 P
QUESTION:
11 8
3.2.'3.1 Seismic Design 1.
It is stated on Page 3.2-10 that a damping value of 2 percent of critical damping should be used for a large diameter piping system under a safe shutdown earthquake.
This damping value is taken directly j
-from NRC Regulatory Guide 1.61.
Shouldn't this i
damping value be 3 percent?
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RESPONSE
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'The damping value of 2 percent has been changed to 3 percent.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
12 3.2.3.1 Seismic Design 2.
It is stated on page 3.2-10 that the maximum horizontal ground acceleration selected for the design of the NUHOMS is 0.25g and the maximum vertical acceleration component selected is 0.lg.
It is further stated in this paragraph that these ground acceleration values correspond to certain recommendations contained in 10 CFR 72.66(2)(i1).
There does not appear to be any reference to a ver-tical ground acceleration level anywhere in 10 CFR 72.66.
There is a reference in 10 CFR 72.66(a)
(6)(ii) to a standardized ISFSI design earthquake response spectrum anchored at 0.25g.
NRC Regula-tory Guide 1.60 states that, depending on excita-tion frequency, that the vertical component of the design response spectra should be either the same as or 2/3 of the corresponding horizontal design spectra.
Shouldn' t the maximum vertical acceler-ation component selected for use in designing the NUHOMS be at least 0.17g rather than 0.lg if the maximum horizontal component is 0.25g?
Additional guidance is given in NUREG-0800, Section 3.7.1 p.3.7.1-4, "To be acceptable the design re-sponse spectra should be specified for three mutually orthogonal directions; two horizontal and one vertical.
Current practice is to assume that the maximum ground accelerations in the two hori-zontal directions are equal, while the maximum vertical ground acceleration is 2/3 of the maximum horizontal acceleration."
RESPONSE
The vertical ground acceleration will be increased to 0.17g.
All associated analysis will be revised to incorporate this change.
NUH-001-3 3.9
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
13 3.2.3.2 Seismic System Analysis On page 3.2-11 it is stated that the concrete co-efficient of friction of 1.0 is taken f rom the ACI 349-80 code.
However, this is only valid if concrete is placed against hardened concrete and the interface between the concrete surfaces is clean, free of laitance and " intentionally roughened to a full amplitude of approximately 1/4 inch."
In the absence of concrete surface preparation specifications, NUTECH should use 0.6 as stated in ACI 318-83 Section 11.7.4.3.
This concern arises again in Chapter 8.
RESPONSE
The concrete coefficient of friction will be changed from 1.0 to 0.6.
The only area effected by this change is the sliding of the single HSM under various accident loads.
If necessary, a shear key type of connection located at the module walls or a tie-down
- system, depending on the type of. concrete construction (i.e.
cast in place vs. precast) will be specified.
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NUH-001-3 3.10
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
14 3.2.5.2 Dry Shielded Canister 1.
NUTECH has choseh ASME Level A service limits for normal operating conditions and Level D service limits for accident or faulted conditions.
The use of Level D service limits, " permit gross general deformations with some consequent loss of dimen-sional stability and damage requiring repair, which
. may require removal of the component from service,"
(ASME, NCA 2142.2(b)(4)).
If Level D service limits are retained for accident or faulted con-i ditions, NUTECH must state in the structural design criteria, that following an accident (such as the 5 foot drop and drop combined with pressure) which could cause damage to the DSC and DSC internal members, the DSC must be disassembled and inspected for damage.
RESPONSE
The only accident condition that service level D
allowables are used is the drop accident.
For this condition the structural design criteria has been q
changed to state:
"Following a drop accident the DSC must be opened and fuel assemblies removed and inspected for damage."
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
15 3.2.5.2 Dry Shielded Canister 2.
If Level D service limits are used, can NUTECH assure that possible deformation within the DSC following an accident will not inhibit the easy retrieval of all fuel assemblies (see Question 8,
under 3.3.4 below).
RESPONSE
For the horizontal drop accident only two components in the DSC basket, the spacer disk and the boral tube, could deform and cause interference 'with the fuel re-trieval operation.
To verify that no interference will
- occur, the maximum elastic deformation of the spacer disk and boral tubes obtained from the STARDYNE analysis reported in section 8.2 will be conservatively added to the maximum possible plastic deformation and then com-pared to the minimum gap between. the fuel assembly and the inside of the boral tube.
The maximum elastic deformation of the spacer disk is.0301" at the fuel cell centerline.
The maximum boral tube deflection is
.0072".
An upper limit _ plastic analysis was performed for the spacer beams between two fuel cells, assuming a three hinged, uniformly loaded beam.
The maximum de-flection calculated was 0.016".
The actual plastic deformation is conservatively enveloped by this value.
Since the boral tube stress is far less than yield stress, no permanent deformation occurs.
Summing the total calculated deflections yields.0533".
The minimum gap is
- 0. 324".
Consequently no interference for the horizontal drop accident is present.
For the vertical orientation drop, the.permenant defor-mation.in the lead casing is by far smaller than the 18 gap specified between top of fuel & bottom surface of top lead plug.
Clearly then, no interference will occur which could inhibit the easy retrieval of all fuel assemblies.
NUH-001-3 3.12
1 DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 d
QUESTION:
16 3.3.3.2 Instrumentation It is stated that there is no need to monitor the DSC internal pressure or temperature.
Limiting pressure and temperature values are based on calculations.
How does NUTECH assure that the calculational techniques are qualified for use in the design?
RESPONSE
The code used to calculate the temperatures is a fully
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benchmarked code and the version which NUTECH used~is fully verified and QA'ed.
It is the code recommended by i
the NRC for such analysis and is used in the same manner as used by the NRC in the evaluation of cask internal-temperatures.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
17 3.3.4 Nuclear Criticality Safety 1.
Were analyses of reactivity as a function of the number of assemblies in the canister pe rfo rmed ?
During loading of the canister in the pool, does analysis show that an arrangement of fewer than seven assemblies does not have a larger reactivity than the completely loaded canister?
Such a
situation might be possible due to reflector effects.
RESPONSE
Reactivity analyses we re not performed as a function of the number of fuel assemblies in the canister.
NUTECH's design basis for criticality control is to use neutron poison (Boral) in all seven guide slee ve s.
(Refer to NUTECH response to question 50 section 3.3.4).
When the canister is loaded with less than seven fuel assemblies, several physical events take place which prevent the reactivity from being larger than when fully loaded.
Most significantly, a major source of reactivity (a fuel assembly) is removed from the system.
To maintain the reactivity, or a higher reactivity, a certain popu-same lation of neutrons would have to pass through the vacant cell before striking an adjacent fuel assembly.
Neutrons traveling through the cell must now pass through a total of four layers of boral.
Since the cell-contains only
- water, neutrons are more efficiently thermalized resulting in their accelerated removal from the population since boron is especially ef fective for thermal neutrons.
The system's reactivity must decrease since the neutron generation rate will be
- reduced, and since propor-tionally more neutrons will expire due to the increased effectiveness of the boral guide sleeves.
If reflection is considered, the effect of the guide sleeves is amplified.
Futhermore, the unaccounted for (in the calculations) fission products in the fuel will always assure that the actual keff will be much less than the conservative ke ff calculated.
NUH-001-3 3.14
DETAILED COMMENTS ON NUTECH 1
TOPICAL REPORT NUH-001 QUESTION:
18 3.3.4 Nuclear Criticality Safety 2.
Were design clearances and fabrication tolerances considered in the criticality analysis?
What are the design clearances that have been provided to ensure insertion and removal of the fuel elements in the guide sleeves?
What is the worst case reactivity for a water flooded canister if the design clearances and fabrication tolerances are taken into account?
RESPONSE
Original criticality analyses did not consider design clearances and fabrication tolerances.
The seven fuel assemblies were assumed to be centered in the spacer disk cutouts.
i A design clearance of 0.1525" at each edge of the fuel has been provided to assure insertion and removal of the fuel elements in the guide sleeves.
Refer to question 3.2.5.2 for further details.
Additional calculat. ions-have been performed to determine the wo rs t case ' reactivity for a water flooded canister if the design clearances are taken into account.
The reactivities for cases where fuel assemblies are as far j
apart as possible and where they are as close together as possible are both lower than that calculated for the i
nominal case.
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Thus, -it is concluded that the variations in design tolerancing will not cause the reactivity to go up.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
19 3.3.4 Nuclear Criticality Safety 3.
From the hydrogen and oxygen atomic densities in t
the KENO input it appears that a water density corresponding to 20 degrees Celsius has been assumed.
How does this temperature compare to limits of the fuel pool temperature.
If the i
temperatures are different, has analysis shown that the reported KENO results are conservative?
?
RESPONSE
Fuel pool temperatures are typically no warmer than 100'-125'F (37.7*-51.7'C).
The pool water was
(
conservatively modeled at 20*C
Since hydrogen and oxygen are both effective moderators, their higher density results in more effective moderation of neutrons.
Since the fission cross section of U-235 is much higher at low energies, more moderation leads to a higher reactivity.
The modeling choice-is. therefore conservative since it results in a higher calculated reactivity than could reasonably be anticipated under actual, plant conditions.
Futhermore, the unaccounted for (in the calculations) for fission products in the fuel will always assure that the actual keff will be much less than.the conservative 1
keff calculated.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
20 3.3.4 Nuclear Criticality Safety 4.
What are the errors or biases due to the cell homogenization procedure?
If the biases have not been quantitatively evaluated, has it been shown that the biases are conservative?
RESPONSE
Biases due strictly to the cell homogenization procedure have not been quantitatively evaluated.
An overall bias was established by comparison of NUTECH criticality calculations to the benchmark results presented in NUREG/CR-0073.
The overall bias is in the nonconservative direction and has been accounted for as described in Section 3.3.4.3.
Tne homogenization process used by NUTEC s described inSection2.1.3o{yheSCALEModuleCSAS2
- Bierman, Clayton, and Durst report that fuel regions (pins) do l
not have to be discretely modeled in KENO IV calcula-tions.
The " smeared" modeling process produced more accurate results than discrete modeling.
J (1)
NUREG/CR-0200 (2)
Bierman, Clayton and Durst, " Critical Separation Between Subcritical Clusters of 2.35 wt%
U Enriched UO3 Rods in
'd3 Water with Fixed Neutron Poisons," Dattelle Pacific Northwest Laboratories, PNL-2438, October 1977.
NUH-001-3 3.17
o DETAILED COMMENTS OM NUTECH TOPICAL REPORT NUH-001 4
4 QUESTION:
21 4
3.3.4 Nuclear Criticality Safety 5.
The des ign of the DSC apparently cons ide rs that some modules may use boron sleeves, while others j
may not.
What controls are employed to ensure'that j
low burnup assemblies are not inadvertently. loaded into unborated canisters?
t
RESPONSE
1 All seven sleeves will be constructed from boral.
Any references in the Topical Report to the contrary will be removed or corrected.
i t'
I i
i I
4 l
NUH-001-3 3.18 1
1
I DETAZLED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
22 3.3.4 Nuclear Criticality Safety 6.
What are the manufacturing tolerances of the boron content of the boral guide sleeves?
Were the criticality analyses p,erformed with a nominal boron concentration or a minimum concentration?
If there I
is a difference, what is the magnitude of the effect on reactivity?
i
RESPONSE
The manufacturing tolerance of the boron content of the Boral guide sleeves is directly related to_ the tolerance in the core panel thickness.
Books & Perkins Product Performance Report 624 indicates a tolerance of i 4 mils for a 75 mil thick Boral panel (which includes the core and cladding).
It is assumed therefore that the boron content per unit area will vary no more than i 5.3%.
Criticality analyses were performed with a nominal boron concentration.
NUTECH's scoping studies demonstrated that reactivity is relatively insensitive to small changes in boron con-l centration, if other parameters are unchanged.
The data shows that when the B-10 concentration increases by 36%,
the reactivity decreases by 2.7%.
Another set of data shows a = 1.8% increase in reactivity for a 22% reduction in B-10 concentration.
I r
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NUH-001-3 3.19 4
a
,c
O DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
23 3.3.4 Nuclear Criticality Safety 7.
The dimensions shown in Figure 3.3-2 are incon-sistent with the text and the KENO input.
It is correct to assume that the Figure is in error?
RESPONSE
Figure 3.3.2 is in error.
The following shall be revised:
0.25" shall read 0.025",
0.75" shall read 0.075".
e NUH-001-3 3.20
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
24 3.3.4 Nuclear Criticality Safety 8.
If Level D service limits are retained for the' DSC, how would the multiplication factor be affected, l
should deformation of the DSC internals occur, when a
the canister is opened for inspection in the fuel j
pool?
(Refer to Questions in 3.2.5.2 above).
RESPONSE
The maximum departure 'f rom the nominal design condition is described in question #18.
No increase in system i
reactivity is expected as a result of fuel assembly A
movement.
(see question #18).
I 1
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NUH-001-3 3.21 e
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,,,n.,
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
25/26 3.3.4.3 Verification Analysis 1.
Have other calculations been performed to verify that the implementation of the KENO-IV computer program used for the criticality analysis produces results with expected statistical dispersions?
If only three benchmark calculations have been calculated for the verification analysis, can a statistically significant determination of the proper and expected operation of the computer code be made.
RESPONSE
KENO-IV has been fully verified by Boeing Computer Services in accordance with their QA program.
The intent of NUTECH's benchmark calculations using KENO-IV was to verify an adequate analytical technique (i.e.,
homogenization procedure, appropriate com-putational flow path and data, etc.) and to establish the bias produced by that technique.
No effort was made to determine the proper operation of KENO-IV by exe-cuting a
" statistically significant" number of calcu-lations.
The operation and statistical significance of the KENO program has been verified numerous times by others.
NUH-001-3 3.22
DETAILED COMMENTS OM MUTECH TOPICAL REPORT NUH-001 1
i l
QUESTION:
27 3.3.4.3 Verification Analysis 2.
What is the justification for the selection of the BNL critical experiments for the verification analysis and in particular the selection of the three experiments that were calculated?
Were other series of critical experiments considered?
If not, why not?
t
RESPONSE
l The - three BNL experiments were chosen because of their similarity to the NUHOMS system, and in particular to the reactivity control materials used in NUHOMS.
4 Other series of experiments were not chosen due to their lack of similarity to the NUHOMS design.
1 i
h NUH-001-3 3.23
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
28 3.3.4.3 verification Analysis 3.
Table 3.3-5 shows the results of the NUTECH calculations of the benchmarks as 'the mean value plus two standard deviations.
While this is the 1
accepted conservative value for criticality design calculations, it is not the correct value to i
dete rmine the computational bias.
The mean value should be compared with the experimental. results.
What are the mean value and standard deviation of the benchmark calculations?
t
\\
RESPONSE
1 The mean value of the benchmark calculations is 0.964.
The, average standard deviation is 0.00544.
1 Table 3.3.5 shall be corrected.
The reported bias shall be based on the benchmark mean values, rather than the i
mean plus two standard deviations, i
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NUH-001-3
-3.24
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 4
QUESTION:.29 3.3.4.3 verification Analysis 4.
The reported results for the three experiments that were calculated dif fered from measurements by more than 2.3 to 2.4%
(see comment in question 2
above).
Is this computational error consistent.
with errors experienced in calculation of these experiments that have been performed by others?
What are the reasons for the reported errors in the present calculations?
RESPONSE
A-reasonable bias of -3. 6% _ was obtained by acceptable methods (See previous question) and applied to the i
computational results.
The NUHOMS design is critically i
safe. as indicated by the resulting K gg which is less e
than 0.92 after inclusion of all statistical and operational biases.
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1 DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 1
QUESTION:
30 3.3.7.1 Irradiated Fuel Handling and Storage The temperature limits on fuel cladding in order to prevent its degradation in storage are based principally on tests in inert atmospheres.
What critoria are applied in the design regarding storage atmosphere inside the DSC to prevent degradation and gross rupture per 10CFR72.72(h)?
RESPONSE
NUHOMS System is designed for the storage of irradiated fuel in a helium a tmosphe re at temperatures well below those that would cause cladding failure.
The criteria applied in the DSC design to assure inert atmosphere are the vaccum level prior to helium backfilling and the closure weld integrity.
i-I 4
1 3
f NUH-001-3 3.26 i
. =.
c, i
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
31 y
3.4 Classification of Structures, Componer.ts, and Systems The safety and quality assurance classifications for each of the components of NUHOMS 0708 important to safety should be provided.
f
RESPONSE
The utility must assign the quality assurance classification and requirements to each of these components in accordance with the utilities established
}
quality assurance policies and practices.
The quality assurance classification for each of ~ these components should be included in the sites FSAR.
As listed in Table 3.3-1 the following items are those which NUTECH considers to be important to the safe
-operation of the NUHOMS System:
1 Cask I
Canister Basket spacer disk Support rods 1
p Lead plug End closer plates and closure welds Canister body
~
Concrete Module DSC support assembly Concrete Shielding The exact nomenclature and classification system used by a specific utility at a specific site will depend on the utilities existing way of doing business.
i i
1 i
NUH-001-3 3.27 1
1 i
I DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i
QUESTION:
32 4.0
. INSTALLATION DESIGN 4.2.3.1 Dry Shielded Canister 1.
It is stated the canister body consists of a 0.5 inch thick rolled and welded stainless steel plate.
What is the maximum leakage rate for the longitudinal weld?
.What, inspections and/or tests
.will be pe rf o rmed to assure that actual leakage is below this rate?
RESPONSE
The-longitudinal weld will be radiographed or ultrasonic
' tested to assure that the integrity of the weld is equal to the 0.5 inch' thick stainless steel plate.
Therefore, the leak rate of the longitudinal weld will be equal to that of the parent. me tal.
Under the NUHOMS system conditions the diffusion will be insignificant (see response to Question 1).
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.m,,
m
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i
4 QUESTION:
33 l
4.2.3.1 Dry Shielded Canister 2.
The canister is partially coated with a dry film lubricant to reduce friction during loading.
Will this film attract an insulating layer of dust or dirt which could interfere with heat transfer?
How much of the canister will be coated with this I
material?
a i
RESPONSE
A dry film lubricant has a thin, hard, and dry finish i
which is bonded to the surface.
The lubricant has a film thickness between 0.0003 and 0.0005 inches.
The film does not attract dust or dirt and nor does dirt
]
adhere to the finish.
I For economic reasons, the dry film lubricant coating has been removed f rom the DSC design and transferred to the l
cask liner and.T-section rails.
By coating these components with a solid film lubricant the coefficient of friction between the sliding surfaces will only be between 0.04 and 0.08, far below the design value of 0 25.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 l
QUESTION:
34 5.0 OPERATION SYSTEM 5.1.1 3 Cask Drying Process 1.
When backfilling the canister with
- helium, the valves are closed when the pressure reaches 14.7 psia.
What is the temperature of the helium when the valves are closed?
How is the temperature measured or inferred?
If the helium is not at the final equilibrium temperature, then the pressure will continue to rise as the gas heats up.
RESPONSE
4 The procedures have been revised to include hydrostatic testing of the DSC and monitoring the pressure until it reaches equilibrium pressure.
The following procedures will be incorporated in the Topical Report o
Attach a self priming pump to valve #2 of the siphon line and drain more than 15 gallons from the DSC.
o Remove the self priming pump.
o Dry any water from top lead plug and DSC interface
-and then seal weld the upper stainless steel
~ cladding plate of the top lead plug to the canister body.
o Connect 0-75 psig pressure gauge to valve #2 on vent tube, and open valves #1 and #2.
.o Connect demineralize water supply to intake side of hydro-pump and connect hose from discharge of hydro-pump to valve #2 of siphon tube.
o Open valves #1 and #2 on siphon tube.
o Activate hydro-pump and pressurize the DSC to 50 psig as read'on pressure gauge.
NUH-001-5 5.1
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001
RESPONSE
(Continued) o Once internal pressure of the DSC has reached 50 psig close valves #1 and #2 on siphon tube and disengage hydro-pump, o
Monitor pressure for 10 minutes, o
After 10 minutes examine the primary closure weld-(weld between top lead plug and DSC shell) for leaks.
Continue monitoring pressure for throughout examination.
If a leak is detected or if internal pressure drop is detected release pressure, remove closure weld, and place fuel assemblies back in fuel pool.
o If no leaks are detected disconnect siphon hose from hydro pump discharge and connect to plant's low-level radioactive waste system.
Open valves on siphon
- tubes, allowing-pressure to drop to atmospheric pressure.
o Remove pressure gauge from valve #2 on vent tube and connect the sites compressed air supply to valve 92.
o Dry canister in accordance with the existing procedures #53 through # 58.
o Open valve number 2
and allow the premeasured quantity of helium to flow into the DSC cavity.
o Close valve numbers 1 and 2 on helium.
Monitor pressure until it is stable.
l NUH-001-5 5.2
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
35 5.1.1.3 Cask Drying Process 2.
What is the differential pressure across the canister wall when the helium sniffer leak test is performed?
RESPONSE
The differential pressure across the canister wall when helium sniffer leak test is performed is 0.5 atm.
J NUH-001-5 5.3
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 J
t 4
QUESTION:
36
!~
5.2.2.1 Safety Features What are the features of the spent fuel storage system which' are important to safety that provide for safe operation under both normal and abnormal. conditions?
What, if any, are the limits selected for a commitment to action?
RESPONSE
Section 5.2.2 of Regulatory Guide 3.48 is intended to describe the operation and safety features associated with the fuel handling. system.
The safety features associated with the fuel handling system are described j.
in Section 5.2.1.2 of the Topical Report.
i The safety features for the DSC and HSM are described in Sections 1 3.1.1 and 'l.3.1.2 respectively.
Chapter 10 lists the operating controls and limits for the safe operation of NUHOMS and specify commitments to action.
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. w DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
37 5.3.2 Component / Equipment Spares In the event that components are damaged during the life
.of the installation, what provision is made for instal-lation of spare or alternative equipment to provide for continuity of safety under normal and off-normal conditions?
RESPONSE
As the analysis has shown, the HSM protects the DSC during normal operation or from any. credible accident.
The only component of the NUHOMS system that can be damaged is the precast shielding blocks located on the i
roof of the HSM.
The consequence of. losing one or. more I
shielding blocks is an increase in the skyshine scattered dose in the vicinity of the HSM.
In order to reduce the scattered dose, new shielding blocks could replace - the damaged shielding blocks or portable shielding could be placed over the air out-lets.
Therefore, NUTECH recommends that spare shielding blocks be retained on site.
The quantity is site specific and may depend on the number of modules. in the installation, frequency of tornadoes, etc.
The utility may wish to retain forms for casting shielding or plan to use existing portable shielding.
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NUH- 001-5 5.5
.,...,-,--c-.
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nn-er r-,.n,..m=.,
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 l
QUESTION:
38 7.2.1 Characterization of Sources "he radia tion sources for the 1.
The design basis of t
shielding analysis is fuel assemblies with an average specific power of 37.5 MW( t) /MTHM, 33000 mwd /MTHM and five years post irradiation time.
Based on these parameters a design basis neutron and gamma ray source strength have been calcu-la t'ed.
Since the burnup parameters are. average values, it may be that some combinations of seven fuel assemblies selected for loading into the DSC may have a higher source strength.
Is the use of the mean values intended to imply that an arbitrary selection of fuel assemblies from a batch of as-semblies satisfying the burnup parameters can be loaded into the DSC?
If not, explain how limiting the source strength in ' any given canister to the design basis source strength will be accomplished in practice.
RESPONSE
~
An arbitrary selection ~of fuel assemblies will not be loaded into the DSC.
Selection and insertion into the DSC will be controlled by existing plant records and procedures.
Section 5 1.1.1, Item 1
ensures that assemblies will be. screened such that they satisfy the radiological requirements specified in Chapters 3 and 10.
Furthermore, health physics surveys of the module l
are also ; specified after DSC insertion.
Any, violation of design bases dose rates will require DSC removal or the use'of; portable shielding.
s 4
3 NUH-001-7 7.1 1
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001
^
QUESTION:
39 1
4 7.2.1 Characterization of Sources l
2.
The flux-to-dose conversion factors shown in Figure
~
7.2-1 seem to be consistent with the multicollision tissue -dose response function at energies greater than 1 Mev, but are a factor of 2 to 10 lower than i
that-response function at lower neutron' energies.
What is the reference for the flux-to-dose conver-sion factors that were used?
RESPONSE
1 Neutron flux-to-dose conversion factors were obtained i
from ANSI /ANS 6.1.1-1977 (N666)
"American National Standard Neutron and Gamma-Ray Flux-to-Dose-Rate Factors."
Table
- 7. 2-1 will be revisedL to reflect the correct
)
response function for the 7.1 kev through 0.41 kev
]
neutron groups.
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DETAILED COMMEMTS OM NUTECH 4
TOPICAL-REPORT NUH-001 i
1 i.
)
QUESTION:
40 7.3.2 Shielding 1.
Have verification analyses been performed for the l
computer. codes used in the shielding analysis (ORIGEN, ANISN, DOT-IV, QADMOD-G, and SKYSHINE-II)?
l Please discuss any shielding benchmark validations that have been performed.
l 4
4
RESPONSE
j verification analysis has been performed on the subject t
computer codes by their respective computing vendors, f
They are, specifically:
ORIGEN2 - Babcock & Wilcox Computer Services ANISN; DOTIV-Boeing Computer.. Services i
QADMOD-G; SKYSHINE II-UCCEL Corporation 4
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o DETAILED COMMENTS OM NUTECH i
TOPICAL REPORT NUH-001 QUESTION:
41 7.3.2 Shielding 2.
What was the concrete composition assumed in the-shielding analysis?
What variation in water i
content is considered over the service life of.the HSM?
RESPONSE
The concrete composition assumed in the shielding analysis is (by weight percent):
Hydrogen 0.5294 %
Oxygen 47.45 i
Sodium 1.629 Magnesium 0.2441 %
Aluminum 4.474 Silicon 30.02 Sulphur 0 1222 %
,1.833 i
Copper 12.53 i
Iron 1.181 The initial water content of the module concrete is expected to be 7.00%.
Therefore, since a
very i
conservative value was chosen for the water content, (4 2%,
a '40%
reduction) no allowance was made for variation in water content over the service life of the i
HSM.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
42 7.3.2 Shielding 3.
What is the relative contribution of primary and secondary gamma rays for the gamma ray surface dose rate on the HSM wall or roof?
RESPONSE
Secondary gamma-rays account for approximately 6% of the total gamma ray dose on the HSM wall or roof.
NUH-001-7 7.5 j
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
43 l
7.3.2 Shielding 4.
Page 7.3-9 third paragraph defers discussion of the HSM penetration calculation.
The deferred dis-cussion is not found later in the report.
Please provida a discussion of the penetration calcula-tion.
What was the quadrature order used in the DOT-IV calculation?
Justify that the quadrature is adequate to accurately calculate streaming through the HSM air exhaust penetration.
RESPONSE
a)
The words "which is discussed later" will be deleted from the third paragraph on page 7.3-9 to avoid any future confusion.
Only one DOT-IV model was constructed to find the
" axial" and " air exhaust" dose rates.
Note that in Figure 7.3-7, " DOT-IV Model Axial and HSM Penetra-tion Shielding Analysis," the model includes. both I
axial points of interest as well as the air exhaust outlets.
The words "See Figure 7.3-7" will be added to the end ' of. the first paragraph on page 7.3-14 to emphasize the DOT-IV modelling choice.
b) 58 is the level of quadrature used in DOT-IV (48 solid angles in each hemispherical space).
It was chosen in accordance with the guidelines given in ANSI /ANS-6.4-1977 as well as other technical literature, and sample problems provided by the computer vendor.
A comparison between the degree of convergence obtained with Sg and lower orders of quadrature was made during analysis for NUHOMS.
On the basis of those comparisons, the additional computational expense of higher order quadrature s
did not offer significantly greater accuracy and was therefore unjustifiable.
I NUH-001-7 7.6 i.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 OPERATION:
44 7.3.2 Shielding 5.
The results in Table 7.3-2 indicate that the neutron attenuation through the HSM air exhaust penetration is greater than that for gamma ra ys.
Can this result be explained?
RESPONSE
i The results in Table 7.3-2 do not indicate that the neutron attenuation is greater.
The table below shows the resulting attenuation f actors.
Neutron Gamma Dose Rate
- Dose Ra te Gamma to Location
( mrem /hr)
-(mrem /h r)
Neutron Ratic,
1)
Air Exhaust 1 (4.17 10-4)(1) 11.5 (6.05.10-7) 11.5 Outlet 2)
HSM 0.03 (1.25.10-5) 8.2 (4.32.10~7) 273 Roof 3)
Surf ace of the 2.4x103 (-)
1.9x107 (-)
7920 DSC (Midplane of fuel) 4 1
't 3
j
- 1) Attenuation tactor; dose rate in rows 1 & 2 divided by dose rate in row 3.
NUH-001-7 7.7 4
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
45 1.2 Shielding 6.
Were any measures taken to minimize ray effects in the DOT-IV calculations?
We re ray effects evident in the results?
RESPONSE
It is well known that ray effects associated with two-dimensional discrete ordinates codes are inevitable in most cases.
A careful examination of NUTECH's results indicates that the flux oscillations ( ray ef fect) occurs only to a minimum extent.
The results were also checked by examining flux variations from interval to interval.
Large gradients in any direction (more than 503) were not evident.
NUH-001-7 7.8
DETAILED COMMENTS ON NUTSCH TOPICAL REPORT NUH-00i QUESTION:
46 7.3.2 Shielding 7.
Has a verification analysis been performed for the application of the DOT-IV code to this type of shield penetration problem?
RESPONSE
A verification run for streaming through a pipe chase (a similar penetration to the NUHOMS air outlet) was performed by Boeing Computer Services as part of their verification program.
t1U H-0 01-7 7,9
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
47 7.3.2 Shielding 8.
Are the
" Cask DSC Annular Gap" dose rates reported in Table 7.3-2 averaged over the cask surface, or are they actually dose rates calculated in the gap?
l
RESPONSE
The dose rates in Table 7.3-2 are the actual dose rates in the gap.
V NUH-001-7 7.10
o a
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 j.
QUESTION:
48 1
7.3.2 Shielding
?
e 9.
What was the angular quadrature used in the Cask DSC annular gap calculation?
Justify that the angular quadrature was fine enough to adequately-model radiation transport through the gap.
Were ray ef fects observed in the. calculated fluxes?
If j
the dose rate on the surface of the canister is plotted as a function of distance from the canister centerline, is streaming through the _ gap observed f
in the results?
1 l
RESPONSE
d l
The angular quadrature used was S Due to the minor i
improvement in convergence ;noted ' g.between Sg and lower order quadratures, an order higher than S8 a9Peared i
unjustifiable in view of the increased computational I
costs (See the discussion in question 4 of section 7.3.2.
1 No significant ray effects were noticed.
Streaming through. the annular gap is apparent in the results if the surface dose rate at the end of the canister is plotted as a function of distance from the i
canister centerline.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
49 l
7.3.2 Shielding 10.
Is there any analysis or data to benchmark the validity of the QADMOD/ albedo procedure used for
]
estimating gamma ray penetration through the' HSM air exhaust?
RESPONSE
NUTECH is not aware of any publishe'd benchmark studies which validate the QADMOD/ albedo procedure used for estimating gamma ray penetration through the HSM air j
exhaust.
cADMOD, a modification of QAD, is an industry-standard point-kernel shielding code --which is fully capable of determining incident dose rates on concrete surfaces within the penetration.
The albedo method is recom-l mended by ANSI /ANS - 6.4-1977, Section 8.4.4.2, and was j
applied in strict accordance with reference number 45 therein.
It is therefore NUTECH's opinion that benchmark studies of this particular analytical method are not appro-priate.
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4 DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
50 7.3.2 Shielding j
11.
Were secondary gamma rays in the concrete con-sidered in the vicinity of the air exhaust i-penetration?
i i
RESPONSE
l Secondary gamma rays produced in the concrete were not considered in the albedo analysis of the air exhaust penetration.
In view of the small neutron flux along the predominant reflecting
- surfaces, secondary gamma j
rays were considered to contribute a negligible amount to the overall gamma ray dose rate outside the i
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DETAILED COMMENTS OM NUTECH TOPICAL REPORT NUH-001 QUESTION:
51 7.3.2 Shielding 12.
With reference to the proper axial pos itioning of the DSC within the HSf1 both during normal operation and in case of seismic event (refer to Question 6.
under 8.1.1.5 and Question 6 uner 8.2.3.2),
it should be noted that there could be shielding implications.
RESPONSE
A seismic retaining assembly has been added to the NUHOMS design.
The seismic retaining assembly consists of a stopping block attached to the back end of the DSC support rails and seismic blocking device to be placed between the bottom cover plate and the door prior to closing the door.
This retaining assembly will prevent the axial movement of the DSC.
NUH-001-7 7.14
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
52 7.4 Estimated On-Site Dose Assessment 1.
Figure 7.4-1 indicated that the Total Dose drops slightly below the Scattered Dose at approximately 1000 ft.
Is this a plotting error or a problem in the computer code?
RESPONSE
There is a plotting error in Figure 7.4-1.
This figure will be revised.
1 NUH-001-7 7.15
DETAILED COMMENTS ON NUTECH 4
j TOPICAL REPORT NUH-001 l
1 1
QUESTION:
53 1
7.4 Estimated On-Site Dose Assessment i
2.
The basis for the dose rates in Table 7.4-1 is not evident in the text.
Please explain how the ' dose
{
rate values were obtained.
i i
RESPONSE
The number of personnel and time span assigned to each 7
of the operational steps in Figure ~ 7. 4-1 is based on r
experience with simila r operations and engineering judgement.
An average distance from the cask surface was assumed and used to estimate the ambient dose rate.-
The distances are intended to reflect time-averaged values for the given activity, and are taken
~
from the torso to the cask surface nearest the operator.
1-The dose rate is obtained using the average distance and dose rate maps constructed from the results of NUTECH shielding calculations.
Dose rates are typically con-servative and do not assume that portable shielding has i
been used.
NUTECH anticipates, however, that in actual practice on a site specific basis portable shielding and i
other normal plant-health physics procedures will be i
employed wherever possible to keep total exposure even 1
lower than indicated on Table 7.4-1.
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NUH-001-7 7.16 i
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_i DETAILED COMMENTS OM NUTECH 1'
TOPICAL REPORT NUH-001 i
j QUESTION:
54 8.0 ANALYSIS OF DESIGN EVENTS r
8.1.1 Normal Operation Structural Analysis i
1.
It. is stated (Page 8.1-1) that "the mechanical properties of materials employed in the --structura l analysis of the NUHOMS system are presented in 2
Table 8.1-2."
This table only contains properties l
for stainless steel, carbon steel and lead.
A number of other materials are used in the NUHOMS system including
- concrete, steel reinforcing,
- boral, etc.
The properties for each of these i
materials should be included in Table 8.1-2 to j
insure that a consistent set of properties are used
~
for all analyses and that appropriate consideration is given to the effect of temperature in each j
specific analysis.
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RESPONSE
4 Table 8.1-2 will be revised to include the mechanical I
properties of concrete, reinforcing steel and boral at
{
various temperatures.
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NUH-001-8 8.1 i
t l
m c
DETAILED COMi4ENTS ON NUTECH TOPICAL' REPORT NUH-001 4
I QUESTION:
55 8.1.1 Normal Operation Structural Analysis h
2.
There is lack of-experience on long herm storage of i
spent fuel ho rizontally.
Please address the ef-(
fects on the fuel of its horizontal storage ove r i
long times in the HSM and any related effects on its retrievability.
i
RESPONSE
}
Although there is-lack of experience on long ~ term storage of spent fuel horizontally, there is con-e j
side rable experience on the shipment of.the fuel in the i
horizontal position.
A summary of shipping experiences i
wo rld wide is listed in the document " Technical Basis for Storage of Zircaloy-Clad Spent Fuel in Inert Gases" by Battelle, and is reproduced here.
a "Approximately 1000 shipments. ( 2000 tU) of LWR fuel l
have been, transported ; dry in Europe and from Japan to j
Europer these shipments have included 4550 BUR and PUR assemblies.
Within the United States there have been j
1500 LWR fuel shipments -(900. tU; 4100 fuel assem-blies).
During sea shipments the fuel remains hori-zontal for up 2' to 3 months, sometimes at cladding i
tempe ratures estimated at up to 385'C.
Thus, the' fuel j
resides horizontally for significant periods at condi-i tions approaching those in dry storage without evidence of significant damage.*
If the fuel is not damaged during ' emplacement and retrieval, horizontal storage i
does.not appear to offer a
threat to cladding integrity."
.It is worth noting that the environment fo r snipping i
fuel, given the transportation ~ shock loadings, ' vibra-l ticasietc., _is much, harsher than the passive environment-1 under ' dry storage.
.Also, on theoretical
- grounds, j
studies have shown cladding creep is acceptable with regard to retrievability, at temperatures below 38 5 'C i
l over the time spans associated with fuel heat decay.
The Battelle report 1 refers.to the document:
" Fry Post-S, Pile Creep:
Experimental Procedure, Test ' Samples and First Results", by Pechs, - M, et,al. 1983, where by the i
studies on the cladding creep is reported.
From the discussion above it can reasonably be concluded that'the cladding behavior in a horizontal position' will' not differ significantly than in the vertical direction.
i i
NUH-001 8.2 i
o-DETAILED COMMENTS OM NUTECH TOPICAL REPORT NUH-001 QUESTION:
56 q
8.1.1.1 Normal Operation Loads 1.
What is the basis for the assumptions of 25%
fission gas release in the design internal pressure calculations?
s
RESPONSE
The selection of a 25% ' fission gas release fraction 'is based on an EPRI fuel performance data base.
The fraction represents a value subs tantia lly higher than that normally seen in LWR fuel operation.
The study of fission gas release fraction is of some importance to the nuclear power industry since large release fractions can have a deleterious effect on fuel performance.
As part of this study, EPRI has developed a data base of fission gas release measurements for 124 well-char-acterized fuel rods.*
Some 102 of these rods were irra-diated in domestic LWRs.
These rods encompass a large range of linear heat ratings, burnups, enrichments fuel pellet densities and operating histories.
None of these samples exceeds a' 25% release fraction.
Most release fractions are between 0.33% and 12%.
1
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- Reference E. Rumble, J. Simpson, S. Lee and A. Woodis, "The EPRI Fuel Performance Data Base; General Description", EPRI NP-1489, October 1980.
i NUH-001-8 8.3 4
l
o DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
57 8.1.1 1 Normal Operation Loads 2.
On page 8.1-5, the normal average canister gas temperature is stated to be 450*F.
However, Table 8.1-13 gives an average helium temperature of 429'F. at 125'F.
HSM inlet temperature.
Which value is correct?
RESPONSE
At 125*F ambient temperature the helium temperature is 429'F as shown in Table 8 1-13.
Page 8.1-5 will be revised.
NUH-001-8 8.4
4 l
l DETAILED COMMENTS'ON NUTECH i
TOPICAL REPORT NUH-001 i
i QUESTION:
58 I
'8.1.1.1 Normal Operations Loads 3.
The temperature distribution for the. spacer disc l
(Figure 8 1-2) shows that the temperature at the point of closest approach between the canister and i
the guide tube is lower than its neighboring points.
Is this correct?
Please explain.
4 t
i
RESPONSE
i Those locations have incorrect temperatures.
The figure i
-l and analysis will.be revised.
.However, the error will i
have a. negligible. ef f ect on the results of analysis, j
because of the small temperature dif ference between the correct temperature value and the temperature shown in I
the figure, and the maximum tempera ture of the spacer disk.
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i DETAILED COMMENTS ON MUTECH l
TOPICAL REPORT NUH-001 t
QUESTION:
59 8 1.1.1 Normal Operation Loads 4.
Please explain how the three steady-state temper-ature distributions were used to determine the effects of thermal cycling.
On page 8.2-51, the heat up rate (100*C temperature change) of the i
insulated loaded DSC is given as 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
Since
}
this is about one-half of a diurnal temperature cycle, have such cycles been included in the cyclic fatigue analysis?
If not, how would their inclu-l sion modify the results?
I.
1
RESPONSE
The heat up rates on page 8.2-51 are for an insulated loaded DSC.
These heat up rates were used to determine the temperature of various components when the air inlets and outlets are blocked, an accident condition.
The fatigue analysis due to seasonal thermal cycling is explained on pages 8.2-57 and 8.2-60.
This analysis is based on the conservative assumption that the range of alternating thermal stress-is from zero to the maximum normal steady state temperature of the DSC ( i.e. 228*F t
at 70*F ambient).
This range is greater than the range between the two extreme ambient temperature (132*F at.-
40*F ambient and 278'F at 125*F ambient) and as such envelops.this condition.
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NUH-001-8 8.6 l
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l
. -=
J DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i
i QUESTION:
60 8.1.1.1 Nortaal Operation Loads 5.
On page 8.1-5 the density of normal concrete is given as 150 pounds per cubic foot.
This value differs from that shown in Tables 8.1-4 and 8.1-4.
Please clarify.-
RESPONSE
i The.upit weight of concrete varies between 140 and 150 lb/ft A 10 lb. Variance in the unit weight of con-crete is not ufusual.
The more conservative unit weight of 150 lbs/ft was used in the structural analysis such as dead weight.
Hoyever, in shielding analysis a unit weight of 145 lbs/ft was assumed In the site specific y
application a minimum 145 lb/ft will be specified in the specification.
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- I
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o DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
61 8.1.1.1 Normal Operation Loads 6.
Equations 8.1-9 and 8.1-10 on page 8.1-20 have missing terms.
If the equations in the Topical Report were used, the results may be incorrect.
Please clarify.
RESPONSE
Equation 8.1-9 and 8.1-10 on page 8.1-20 are incorrectly typed.
The analysis contained missing v and 1 terms.
Page 8.1-20 will be revised.
NUH-001-8 8.8
t DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
62 8.1.1.1 Normal Operation Loads 7.
27.0x10ylus The mod of elasticity is i taken as psi instead of 26.6xl gcorrectly psi for the design basis temperature of 400*F.
Please correct.
RESPONSE
6 The Yougg's Modulus value of 27. 0 x10 will be revised to 26.6x10 psi.
NUH-001-8 8.9 1
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i
QUESTION:
63 8.1.1.1 Normal Operation Loads I
8.
The temperature distribution within the DSC depends upon the azimuthal orientation of the DSC within the HSM.
That is, the analysis assumes that three (3) assemblies are centered on a vertical line passing through the center of the DSC.
How will this orientation be assured during loading?
RESPONSE
It.is highly unlikely that any variation in the canister i
orientation of the DSC within the HSM will cause any significant differences in the internal temperature distribution.
- However, to assure that the physical orientation is identical to that of the thermal model, 4
markings will be placed on the top cover plate that will indicate the azimuthal orientation of the canister.
The DSC will be loaded into the cask in the proper 4
orientation.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
64 8.1.1.2 Dry Shielded Canister Loads Analysis 1.
Thermal variations in the circumferential and axial direction of the canister will result in thermal stresses due to self constraint by the colder can-ister ends.
On page 8.1-22 the statement is made that, "Also, the thermal variation in the circum-ference and longitudinal directions -of the shell are not considered significant since in the unres-trained condition the shell will expand radially and longitudinally to an equilibrium state."
NUTECH should show what the overall deflected shape is in the axial direction, and the deflected shapes of a cross section at the end and the center of the DSC.
NUTECH should show what the ASME stress in-tensities are fo r the following locations on the DSC:
(1) top and (2) bottom elements of shell at top cove r plate edge (3) top and (4) bottom ele-ments of shell at center section; (5) top and (6) bottom elements of top cover plate at shell edge.
Please include effects of_ discontinuity between shell and top cover plate.
RESPONSE
The temperature along the length of the DSC does not vary significantly (
20*F).
This temperature change is not as great as the thermal variance that occurs around the circumference of the DSC shell.
As shown in Figure 8.1-1, the temperature of the DSC shell varies from 229'F to 285'F.
Thermal stresses due to this variance were evaluated by means of finite element analysis, the maximum thermal stress inter.sities of 11.24 ksi were determined which occurred at the DSC and T-rail inter-face.
This stress is far lower than the thermal stress of the DSC shell reported in Table 8.1-7 of page 8.1-15 of.the Topical Report.
The stress values reported in Table 8.1-7 is located at the region of the DSC and Spacer Disk interface and is based on the differential thermal expansion of the spacer disk and the DSC shell.
Based on the above observation the the rmal stresses at other locations requested are envoloped by those reported in Table 3.1-7.
NUH-001-8 8.11
_- =
_ _ _.. _ = _ _. _ _
DETAILED COMMENTS ON NUTECH' TOPICAL REPORT NUH-001 QUESTION:
65 2
8.1.1.2 Dry Shielded Canister Loads Analysis 2.
The drawings provided ( ADV001'. 0 20 4 sheets 1 and 2) i do not provide enough dime ns ions to be able to determine what minimum gaps might exist. between the boral baskets and the top lead plug.
Please supply actual shop drawings instead of conceptual design drawings.
This relates to p. 8.1-22 and 8.1-24.
RESPONSE
The drawings provided are neither conceptual nor shop drawings.
They are design drawings.
Clearances will be.
j added to the drawings.
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NUH-001-8 8.12 i
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
66 8.1.1.3 Dry Shielded Canister Internal Basket Loads Analysis Please provide information showing where the 0.75 inch irradiation growth of the fuel assembly came from.
The drawings do not permit a verification of calculations of this section of the topical report.
RESPONSE
The irradiation growth of 0.75 inches was based on data obtained from the following report:
Examination of High.Burnup fuel at the H.
B. Robinson Reactor End of Cycle 8 Report No. XNNF8 317 Exxon Nuclear, Inc.
NUH-001-8 8.13
DETAILED COMMEMTS ON NUTECH TOPICAL REPORT NUH-001 t
1 QUESTION:
67
{
8.1.1.4 DSC Support Assembly Loads Analysis The first sentence of the Thermal Analysis section p.
8.1-32 is acceptable if an assembly procedure specifying the torque of the nuts is included.
t
RESPONSE
As specified in Table 3,
p.
5-214 of the AISC Steel Manual the minimum tension for a 3/4 inch diameter bolt is 28 kips.
This corresponds. to a tightening torque on the nuts -of 4200 in. lb.
The amount of friction force
{
created by the tensioning of the bolt is 21 kip.
How-ever this force can easily be overcomed by the thermal expansion force created temporarily. by the heat up of the rail during normal operating condition.
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i NUH-001-8 8.14
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
68 8.1.1.5 Horizontal Storage Module Loads Analysis 1.
A reference in the third paragraph of this section (page 8.1-35) is made to NUREG 0880.
Shouldn't this reference to be NUREG 0800?
RESPONSE
The reference should be NUREG 0800.
This page will be revised.
NUH-001-8 8.15 i
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i
QUESTION:
69 8.1.1.5 Horizontal Storage Module Loads Analysis 2.
Table 8.1-10 (Page 8.1-38) includes v'alues for the ultimate moment and shear capacity of the HSM.
The value computed for the ultimate moment capacity, i
4052.0 in-kips, is given on Page 8.1-49.
- However, no calculation is presented for the ultimate shear value, 59.2 kips.
The details of this calculation i
should be included.*
l
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RESPONSE
The ultimate shear capacity was based on the following i
equation from equation 11-28 in American Concrete Institute (ACI) 349-80.
1 Y
=2
' fl; b wd i
c The ultimate shear value calculation will be incorporated'into the Topical Report.
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1 NUH-001-8 8.16
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 d
QUESTION:
70 8.1.1.5 Horizontal Storage Module Loads Analysis 3.
Figure 8.1-11 (page 8.1-39) includes moment and force loads that are applied to the reinforced concrete corbel that supports the DSC support Assembly.
Calculations should be provided for the stresses in the corbel due to these loads and for the ultimate strength capacity of the corbel.
RESPONSE
a Figure 8.1-11 shows the dead loading and live loading at the centerline of the module walls, not the. corbel cen-terline.
Table 8.2-12 show the maximum end loads (at corbel) for each load combination.
As stated on page 8.2-63 the corbel, bearing plate and bolts were designed for these load combinations.
Additionally, for the cor-bel design loads were " multiplied by ' 1.7 to envelope all 4
load combinations specified in ACI 349-80.
4 Design and analysis were performed using the criteria of Section 11.9 of ACI 349-80 and the recommendations of Chapter 5 of Wang & Salmon ~ (Reinforced Concrete Design, 1
Third Edition).
The nominal shear strength, vn, using equation (11-31) of'ACI 349-80 yields 29.4 kips.
Since i
the factored shear load is only 17k the corbel is ade-quate per ACI 349-80.
Additional corbel ~ reinforcement and welded' plates were added to allow the direct trans-i fer of.the horizontal forces to the tens ion ' re i nf orue-ment per suggested practices in Wang.
Therefore, the concrete corbel is designed to take all DSC support assembly-load combinations per'ACI 349-80.
Table 8.2-12 will be revised to incorporate the capacity Jof the. corbel, bearing pads and the bolts at both normal operating and accident condition.
i NUH-001-8 8.17
DETATLED COMMENTS ON NUTECH 4
TOPICAL REPORT NUH-001 4
l
-QUESTION:
71 8.1.1.5 Horizontal Storage Module Loads Analysis i
4.
Please state your interpretation of ACI 349-80 acceptance criteria for concrete temperatures.
An air inlet temperature of 70*F is used to evaluate normal operating temperatures.
Please relate this number to meteorological information which could be used to determine site acceptability.
a
RESPONSE
ACI 349-80 Appendix A,
Section A.4 states for normal operation or any long term period the concrete tem-perature shall be limited to 150*F except at local areas, where it is limited to 200*F.
For accident con-ditions or periods of short time duration, the temper-1 ature may not exceed 350*F except at local areas, where j
it is limited to 650*F.
Section A-4-3 of the code allows for even higher temperatures than those specified i
if tests are performed to evaluate the loss of strength j
due to sustained exposure to elevated temperatures.
On page 0.1-44 of the Topical Report a number of references are listed which show concrete strength is not effected by temperature below 212*F for both short and long term duration.
Although tempera ture does impact the mechanical prop-erties of the concrete, the critical factor to consider is the - temperature gradient in the concrete and its associated stresses.
As shown in Table 8.1-12 the maximum temperature gradient in the concrete occurs at i
normal ambient temperatures of 70*F, which produces ' an enveloping thermal stress in the concrete walls & slabs.
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NUH-001-8 8.18 i
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DETAILED.COMMEMTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
72 8.1.1.5 Horizontal Storage Module Loads Analysis 5.
The analysis of local concrete temperatures does not include conduction through the thermal shield support bolts or through the support rail to the corbels.
How will consideration of this direct heat transfer path affect local concrete temper-atures and satisfaction of ACI 349-80 acceptance criteria?
Also, how will concrete thermal con-ductivity be affected by thermal cycling over the license period, and by the presence of rebar in the concrete?
RESPONSE
The heat shield support bolt, located at the mid-section of the roof slab has been removed, figure 4.2-6 will be revised to incorporate this change.
Furthermore assuming that the heat shield support bolts or corbels do provide a direct heat transfer path, the local concrete temperatures, under nortal operating conditions, do not exceed. the' acceptable criteria esta-blished in ACI 349-80.
The remaining heat shield sup-port bolts are located at the extreme ends of the heat shield.
HEATING 6 analysis shows that the temperatures in these areas does not exceed 170*F.
Additional heat transfer analysis shows that under normal operating tem-peratures, the-temperature of the DSC support assembly is less than 170*F.
Conservatively, assuming that the bolts or corbels are at the same temperature as the base metal and that the local temperature of the concrete is at the same temperature as the bolts or corbels, the local temperature of the concrete is below the criteria
' in ACI=349-80.
Thermal cycling of the concrete over the license period of the HSM has been discussed in response to question 1.2.2-4.
s i+-
NUH-001-8 8.19
o DETAILED COMMEMTS ON NUTECH TOPICAL REPORT NUH-001 1
QUESTION:
73 8.1.1.5 Horizontal Storage Module Loads Analysis l
6.
If the end of the DSC were to contact the concrete wall, a concrete temperature of the same magnitude as the calculated DSC temperatures would result.
What is the clearance between the end of the DSC and the concrete wall and what is the concrete temperature adjacent to the end of the DSC?
2
RESPONSE
The clea rance between the end of the DSC and the concrete wall is 3 inches.
i Additional heat transfer analysis indicates that the average. concrete temperature adjacent to the end of the DSC is approximately 70*F lower than the average surface 4
temperature of the DSC end.
The concrete temperature for the end walls of the HSM is
{
not as critical as the side walls and roof slab, since t
the frame action in the transverse direction of the HSM is considered the main structural resisting system.
Also, the air outlet opening at the ends of the HSM roof
- provides the freedom for the thermal movement of the end
. walls.
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NUH-001-8 8.20 l
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
74 8.1.2 Off-Normal Events Failure of the supports for the thermal radiation shield (due to seismic events, corrosion, etc.) has not been addressed.
Please justify that this event is not feasi-ble, or provide an analysis of the consequences.
How would such as failure be detected?
RESPONSE
The heat shield is designed to withstand all applicable loads and environmental conditions during the lifetime of the horizontal storage module.
The heat shield will experience only dead weight and seismic loads.
Thermal loads will not be experiencad since oversized holes will allow unrestrained thermal expansion.
The heat shield will be galvanized to preve.it corrosion.
The shield and the embedded anchor bolts were analyzed for maximum dead weight plus seismic loadings.
Conservatively, the heat shield was treated as a simply supported beam with a uniform loading.
A maximum bending stress of 12.43 ksi was calculated.
Axial and shear stresses were negli-gible.
The tensile and shear stress of 1.08 ksi and 0.53 ksi were calculated respectively for the anchor bolts by conservative assumptions.
A bearing stress of 1.68 ksi was calculated for the anchor bolt / heat shield contact point.
Clearly then, these minimal stresses 2
show that the radiation shield is withstanding any postulated loading.
Therefore, no failure detection system.is necessary.
NUH-001-8 8.21
o-1 DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i
QUESTION:
75 a
8.1.2.2 Blockage of the Horizontal Storage Module Inlet - Cause of Event and Detection 1.
Air flow through the HSM with both inlets blocked appears to be a random process with zero mean value.
Please provide experimental confinnation s
based on scaled.or appropriate data to demonstrate the assumed magnitude of flow for this case.
s
RESPONSE
The air flow is not a zero mean, random process.
The 7 3
kw of heat produced by the fuel will heat the air and cause a significant driving force to_ push the air out of the HSM.- As the air is forced out, the pressure. in the HSM will decrease and air from the outside will be drawn in.
Exact flow paths will depend on existing meter-logical conditions, but because. of the constant heat source, the air will flow.
As part of the demonstration program, extensive testing will be performed on the first prototype module to be built at Carolina Powe r and Light.
Evaluation of the sirflow through the HSM when both air inlets are blocked will' be included in the electrical heater. testing program.
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I NUH-001-8 8.22
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 2.
QUESTION:- 76 8.1.2.2 Blockage of the Horizontal Storage Module Inlet - Cause of Event and Detection 2.
Please clarify the pressure given on page 8.1-60.
The values are not consistent.
According to the 4
ASME text booklet "SI Units in Heat Transf9e" ist Ed.,
14.504 psi - 1 bar.
Also, refer to subsequent units conversions between psi and bar.
RESPONSE
The pressure given on page 8.1-60 will be revised.
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NUH-001-8 8.23
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
77 8.1.3.1 Thermal-Hydraulics of the Horizontal Storage Module Principles of HSM Cooling System 1.
Reference 8.44 referred to here is not given in the reference list.
RESPONSE
Reference 8.44 was inadvertently omitted.
Reference 1
8.44 is I.E.
Idel'Chik, Handbook of Hydraulic Resistance, Coefficients of Local Res is tance and ot Friction, the U.S.
Atomic Energy Commission and the National Science Foundation, Washington, D.C.,
1960.
This reference will be included in the revision of the Topical Report.
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NUH-001-8 8.24
)e
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
78 8.1.3.1 Thermal-Hydraulics of the Horizontal Storage Module Principles of HSM Cooling System 2.
The thermal analysis is based on an assumed axial peaking factor of 1.08.
This is an
- dditional limiting design parameter which is not cited in Table 3.1-1.
If it is not intended to impose this limit on stored fuel, please justify its conser-vatism for proposed fuel types, burnup histories, etc.
RESPONSE
In most heat transfer analysis of decay heat during dry storage the heat generation rate of a fuel assembly is assumed to be uniformly distributed along the active fuel length.
However, in NUTECH's analysis the axial-peaking factor of 1.08 was conservatively applied to the uniform heat generation rate in order to determine the worst case temperature distribution in a two dimensional cross section of the DSC and HSM.
Even wi';h the conser-vatism the maximum ' fuel cladding temperature was calcu-lated to be 338*C, under the worst ambient temperature of 125*F.
This maximum cladding temperature is 62*C less than the accepted limi ting fuel rod temperature of 400*C.
i As stated in Section 10.3.1.1, the decay power per fuel assembly is limited to 1 kw.
i i
NUH-001-8 8.25
o DETAKLED COMMENTS OM NUTECH TOPICAL' REPORT NUH-001 QUESTION:
79 8.1.3.1 Thermal-Hydraulics of the Horizontal Storage Module Principles of HSM Cooling System 3.
Equations 8.1-37 through 8.1340 are applicable within restricted domains of L AT.
Please show that these restrictions are satisfied.
RESPONSE
I Equations 8.1-37, 8.1-39, and 8.1-40 are valid for flow in the turbulent range.
As the table below indicates the flow is in the turbulent ra nge.
Gr:Pr falls between 108 and 10 12 The AT is based on the average steady state temperature of the particular surface and the average air temperature in the region.
Gr:Pr for HSM for Various Ambient Temperatures i
Ambient Temoerature Flow Region
-40*F 70*F 125*F Canister 2.8 x10 10 2.53x1010 2.38x1010 4
Circumference 9
Floor 2.3x10 2.68x109 2.92x109 Wall 1.83x1010 2.0x10 10 2.29x1010 l
3 8.2-138 is valid over any L at domain.
NUH-001-8 8.26
,m.+
-m n--
1rme,-
-y+w*w---T-ww--
t-
- w 1gww
_ - m O
DETAILED COMMENTS OM NUTECM TOPICAL REPORT NUH-001 i
1 QUESTION:
80 8.1.3.1 Thermal-Hydraulics of the Horizontal Storage Module 4
Principles of HSM Cooling System i
4.-
What levels of solar radiation were considered in the heat transfer analysis?
Of particular concern is. the levels associated with the hottest days (Tair=125'F).
i i
i
RESPONSE
The following equivalent solar heat flux were included in the HSM thermal model.
i Ambient Solar i
Temperature Heat Flux
(*F)
(BTU /hr.ft2)
-40 23.5(1) 70 93.8(2) 1 125 187.6(3) 4 (1) 1/12 of the half day solar heat gain for 40 degree i
north. latitude on December 21.
(2) 1/12 of the half day solar heat gain for 40 degree north latitude on June 21.
(3)
Twice the solar heat flux for 70*F case.
f 4
i NUH-001-8 8.27 e
3
- - + -
r-m.
-.w.-m.--
+-
p.
w g_.,-tyw.Wyi.--.=-.g
-w-
O DETAILED COMMENTS ON MUTECH TOPICAL REPORT NUH-001 QUESTION:
81 8.1.3.1 Thermal-Hydraulics of the Horizontal Storage Module Principles of HSM Cooling System 5.
Page 8.1-69 gives the maximum concrete ceiling temperature as 244*F, whereas Table 8.1-12 gives 249'F.
This analysis does not consider direct conduction through the supports to the concrete.
Is the severe summer condition considered to be an of f-normal or transient condition?
If so, what is the limiting ambient condition to be used for determination of whether acceptance criteria on concrete temperature are met?
The value of 70*F is quite low for a limiting ambient condition.
RESPONSE
The maximum concrete ceiling temperature is 249'F page 8.1-69 will be revised to read 249'F.
The question of " Direct Conduction through the Supports" is responded in Section 8.1.1.5, question 5.
The severe _ summer condition is consid red to be a
transient condition, since the time required for the concrete temperature to reach steady state is by far greater than the short period associated with extreme 125*F ambient temperature.
The selection of 70*F as the limiting ambient tempera-ture is based on the fact that the the rmal gradient through concrete thickness is greater for the 70*F ambient temperature than the other ambient extremes.
Comparison of cases 1 through 3 of Table 8.1-12 will indicate this fact.
Since the concrete thermal stress is maximized by the application of maximum thermal gradient, the selection of-70*F ambient condition is considered appropriate.
Furthermore, because of the long time constants associ-ated with the thermal masses of the HSM and DSC, they will, over. a daily bases, be at temperatures which are more representative of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> average air temper-ature than any
- 30 to 40*F daily temperature swings.
Thus, again, 70*F is _ a good value of the normal oper-ating ambient temperature.
NUH-001-8 8.28
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
82 8.1.3.3 Thermal-Hydraulic Analysis of the Canister Inside the Transfer Cask Model 1.
In equation 8.1-40, the emissivity of the stainless steel transfer cask exterior is taken as 0.80, whereas a value of 0.587 was used for stainless steel surfaces within the DSC (page 8.1-74) and transfer cask (Equation 8.1-43).
Please justify this use of different values for the emissivity.
RESPONSE
The the rmal-hydraulic analysis of the canister inside the transfer cask assumed that the DSC was seated within the cavity of the GE IF-300 shipping cask.
The emissiv-ity value, 0.80, for the stainless steel exterior of the transfer cask, correspon,ds to the effec:ive emissivity for the finned surface of the GE IF-300 shipping cask, 0.83 rounded down.
The ef fective emissivity is reported on page 6-41 of the GE IF-300 Shipping Cask Consolidated Saf ety Analysis Report.
NUH-001-8 8.29
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
83 8.1.3.3 Thermal-Hydraulic Analysis of the Canister Inside the Transfer Cask Model 1
2.
What is the sensitivity of the calculated results to variation of input parameters, e.g.,
amissivity, axial peaking factor, apparent theemal conductiv-ity, decay heat, etc.?
How is the margia of safety to limiting conditions established?
4 r
RESPONSE
4 i
The purpose of the reported analysis is to show that,the safe ' operation of the system is not jeopardized under various ope ra ting scenarios.
- The margin of safety was established through conservatisms assumed in the calcu-lations and the extreme conditions,which were assumed.
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NUH-001-3.
8 30
, ~.
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i
QUESTION:
84 8.1.5 Design Basis Internal Pressure Loads Please describe any hydrostatic testing to be performed to assure integrity of the DSC.
Has. NUTECH considered pressurizing the loaded DSC with helium as a test of DSC integrity?
In such a. test how would the test pressure relate to the design pressure and the accident pressure discussed in Section 8.2.9?
RESPONSE
As stated in response to Question 5 1.1.3 the DSC will be hydrostatically tested in accordance with the requirements of ASME,Section III, Subarticle NB-6220.
The test will be conducted at a pressure of 1.25 the accident pressure, discussed in Section 8.2.9.
The hydrostatic test should be more than adequate in assuring the DSC integrity.
1 i
I aa 1
e NUH-001-8 8.31
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 A
QUESTION:
85 8.2.2.2 Accident Analysis 1.
The coefficient of friction between a
- single, unanchored module is stated to be 1.0 on Page 8.2-7.
The ACI 349-80 code states in Sections 11.7.5 and 11.7.9 that the coefficient of friction shall be 1.0 if concrete is-placed against hardened concrete and the interface between the concrete surfaces is clean, free of laitance. and "inten-tionally roughened to a full amplitude of approxi-mately 1/4 inch."
The ACI-3t8-83 code states in Section -11.7.4. 3 that the coet'ficient.of friction shall be 0.6L for - the case wherein. concrete is placed against hardened concrete not intentionally roughened
(:L:
is this relation - 1. 0 for normal weight concrete).
shouldn't coefficient of fric-tion of no greater than 0.6 be used between the NUHOMS system and the supportingi. concrete pad, especially since no specific mention is made of the need for roughening the surface of the concrete pad under the NUHOMS system?
(See question on Section 3.2.3.2).
l
RESPONSE
As stated in
-response to que s tion 3.2.3.2, the coefficient of Eriction will be Jchanged from 1.0 to 0.6.
The only area effected by ~ this change is the sliding. of the single HSM. under various accident loads.
A shear key type of connection located at the module walls or a tie-down system will be specified.
NUH-001-8 8.32 w
w-3
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
86 8.2.2.2 Accident Analysis 2.
It is noted on Page 8.2-13 that a single HSM will require tie downs for resistance against overturn-ing and sliding under the action of a massive missile impact (i.e.,
a 3,976 lb automobile).
It is further noted on this page that the NUHOMS 0708 concept, which has 8 modules either poured in place or pre-cast, will not slide under the postulated impact.
Since it is understood that actual, site specific configurations for the NUHOMS system may have any number of modules, wouldn't it be appro-priate to either require tie downs, to permanently anchor the NUHOMS system to its supporting concrete pad or to set the HMS in a key-way embedded into the pad rather than rely on the use of friction to restrain the HMS from sliding on the pad and potentially exposing contaminaced surfaces?
RESPONSE
As stated in response to question 3.2.3.2, a shear key type of connection located at the module walls on a tie-down system will be specified.
i NUH-001-8 8.33
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
87 8.2.3.2 Accident Analysis 1.
The values of E on pages 3.2-14 and 8.2-16 are incorrectly taken at room temperature.
Please use the value of E
at the maximum calculated temperature.
RESPONSE
The values of E on pageo 8.2-14 and 8.2-16 of the report will be changed from 27.0 E5 to 26.6 E6.
NUH-001-8 8.34
. g.
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 1
l QUESTION:
88 8.2.3.2 Accident Analysis 2.
The calculations in this section assume the maximum vertical earthquake acceleration to be 0.lg.
As noted in Question 3.2.3.2, this acceleration should be increased to at least 0.17g.
Each of the calculations in Section 8.2.3.2 that are based on the maximum vertical acceleration should be ' redone to reflect this increased acceleration and each reference to a vertical-acceleration of 0.lg should be changed to 0.17g.
RESPONSE
As stated in response to Question 3.2.3.1-2, the vertical ground acceleration will be increased to 0.17.
All associated analysis will be revised to incorporate this change.
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NUH-001-8 8.35 t
,e
-4
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
89 8.2.3.2 Accident Analysis 3.
The term V in Equation 8.2-20 (Page 8.2-19) only w
includes the volume of one wall of the HSM whereas the term V includes the entire weight of the roof.
Furtfie rmore, the value for I (74,088 in 4) given on Page 8.2-20 represents the moment of inertia of only one wall.
Shouldn't the calcula-tion for mass presented in Equation.8.2-20 include 1
only one-half of the mass of the roof?
- Also, shouldn't this calculation include the effect of the DSC and its support?
RESPONSE
T As stated on page 8.2-19 the mass was taken as the entire top slab plus one half of the side walls.
Therefore, the term V is the volume of one half of two w
walls or one total wall.
The mass of the lower half portion of both walls is assumed. to be reacted at ground level.
This approach of lumping mass based on tributary area is commonly practiced throughout the industry.
Additionally the moment of inertia is based on one wall.
Equation 8.2-21, however, contains a factor of 2 to account for the stiffness of both walls.
Hence, the module frequency calculations are correct as shown on pages 8.2-19 and 8.2-20.
The effects of the DSC and the support assembly were 3'
omitted ' since they act below the halfway point on the module walls and will not have a significant effect on the results.
The mass of the DSC and suppoyt assembly 1
per 12 inch section of the HSM is 3.9 If this 1
mass were conservatively lumped at the HSM roof the frequency would be 35.9 Hz.
This conservative value is still beyond the control point A of 33 Hz shown in Tables 1 and 2 of NRC' Regulatory Guide 1.60.
Therefore, the ' max imum amplification factor of 1.0 is still valid for the analysis.
NUH-001-8 8.36.
k i
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 t
. QUESTION: - 90 j
8.2.3.2 Accident Analysis
-r 4.
Shouldn' t the parameter v in Table 8.2-6 (page u
8.2-21) actually'be labeled "M " as in Table 8.2-3?
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RESPONSE
i The term V in Table 8.2-6 will be replaced by the u
correct term M in the next revision of the Topical u
Report.
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I NUH-001-8 8.'37 4
a
,= -,
a e.,--
-.-~m,-.
,.,ng
.,-a_
.w,
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
91 8.2.3.2 Accident Analysis 5.
No mention is made in the sub-sections entitled "DSC Stress Analysis" (Page 8.2-16) and " Horizontal Seismic" (Page 8.2-24) of combining horizontal seismic -loads from two directions.
Seismic loads from two horizontal directions and the vertical direction should be combined unless a statement can be made ' that loads from a particular direction can be considered to be negligible (i.e.,
as was done on Page 8.2-20).
RESPONSE
The DSC is retained axially for a seismic event as described in the response to Question 7.3.2 (Number 12).
Consequently, an axial membrane stress will be experi-
-enced due to the 0.5G horizontal loading.
Assuming an equal stress distribution throughout the DSC yields a stress of 0.19 ksi.
Compared to the local bending i
stress of 9.58 ksi calculated by a horizontal loading perpendicular to the DSC, this axial stress is negli-gible.
Additionally, no increase in combined stress will occur when summed with the other horizontal stress using the SRSS method.
Therefore, stress due to an axial horizontal acceleration.along the DSC shell is minimal and can be. considered negligible.
A statement of this effect will be added on page 8.2.24 of the Topical Report.
For the support assembly seismic evaluation the stress due to horizontal acceleration in the axial direction can be calculated directly by factoring the results from the friction load case.
The maximum bending and shear stress in the W8 x 28 are 0.56 ksi and 0.68, respec-tively.
The maximum bending and shear stress in the W4 x 33 5 are 0.76 ksi and 0.34 ksi.
These values will be combined by the SRSS method with the other horizontal stress and then added absolutely with the vertical seismic stress.
This revised value will be included in the subsequent load combination reported in Table 8.2-11.
The seismic load combinations in Table 8.2-11 will be revised to reflect the results listed below.
Load Calculated Stress (ksi)
Component Combination Axial Bending Shear W 8 x 28 DWg + DWC + SSE 0.40 8.19 5.08 i
WT4 x 33.5-DW3 + DWC + SSC 0.77 17.37 5.~01 NUH-001-8 8.38
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
92 8.2.3.2 Accident Analysis 6.
As a follow-on to the above question, it is noted that no mention is made of the possibility of the cask sliding longitudinally along the rails when loaded seismically.
If such a possibility exists, could some part of the HSM structure be impacted by the DSC (e.g.,
the door), resulting in damage to either the HSM or the grapple on the end of the DSC7 If the DSC cannot slide, shouldn't some sort of load be applied to the rails, and consequently to the DSC support assembly?
RESPONSE
As described in the response to question 7.3.2 (number
- 12) a seismic retaining assembly will be installed prior to closing the HSM door.
This assembly will prevent longitudinal sliding of the DSC along the rails during a seismic event.
Therefore, no possibility of impact damage is present.
As described in the answer to question 8.2.3 2 (number 5) the DSC support assembly will experience seismic axial loads along the WT 4x33.5 rails.
Consequently, the Topical Report will be revised to include stresses caused by this loading.
See answer to question 8.2.3.2 (number 5, for revised stresses due to this additional horizontal seismic loading.)
NUH-001-8 8.39
f.
s.
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 t-t OUESTION:
93 l
4 8.2.4.2 Accident Analysis 1
1.
Reference to Table 3.2-2 for flood loadings is incorrect (Page 8.2-24).
Also reference to Table 3.2 is incorrect (Page 8.2-28).
1 i
-RESPONSE:
i i
The reference to Table 3.2-2 on page 8.2-24 and Table l
3.2 on page 8 2-28 will be changed to Table 3.2-3.
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4 NUH-001-8 8.40 i
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s.
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i
QUESTION:
94 8.2.4.2 Accident Analysis 2.
A flood level which caused the bottom portion of l
the DSC to be in contact with water while the upper portion remained in air could cause considerable j
thermal stress in the canister.
Have these load-ings been included with the pressure loadings?
5 1
RESPONSE
A flood level precisely high enough to block air flow j
but is not considered a credible situation and therefore the analysis has not been performed.
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.NUH-001-8 8.41
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
95 8.2.5.2 Accident Analysis 1.
The discussion on page 8.2-29 implies that the decelerations used by NUTECH were obtained by multiplying actual drop-test data (obtained by ORNL) by one-sixth to reflect a 5 foot drop' height instead of a 30 foot drop height.
Examination of the GE Consolidated Safety Analysis Report of the
'IF-300 shipping. cask reveals that NUTECH used decelerations which the Stearns-Rogers Company calculated using a
dynamic analysis
- program, "Dyrec".
Since the Stearns-Rogers results envelope the ORNL
- results, this is a
conservative approach.
However NUTECH should revise the discussion to reflect their actual process.
Also NUTECH should reference the data by appendix.
RESPONSE
The discussion on page-8.2-29 will be revised to reflect i
the actual process.
Additionally, data will be referenced by appendix.
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NUH-001-8.
8.42
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
96 8.2.5.2 Accident Analysis 2.
A second related question concerning drop decelera-tions has to do with the data in Appendix V2 of the GE document NEDO-10084-3.
Why did NUTECH not use the deceleration for top end vertical drop as reported in this appendix?
It is more conservative than the Stearns-Rogers data.
RESPONSE
4 GE document NEDO-10084-3, Appendix V2, September 1984, was reviewed for possible higher vertical drop de-celeration.
This document presents the 30 ft drop impact analysis for the fuel rods and not the reeval-uation of the impact time history.
As a matter of fact, the-time history, Figure. 5-1, page V2-47, presented in this document is identical to the time history developed by Stearn-Roger and presented in Appendix V-1 of the GE 4
Safety Analysis Report.
Additionally, as stated ' in response to previous ques-tions the maximum deceleration forces which the DSC is the 48g vertical and 34g horizontal.
These are to be considered as the governing l'im i ts and not the drop height of the cask.
t NUH-001-8 8.43
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
97 8.2.5.2 Accident Analysis i
3.
NUTECH has not provided a
discussion on the possibility of the cask dropping vertically through 5 feet and then tipping over and slapping down.
Logically, if the cask.could experience a 5 foot vertical drop, then it would almost certainly also experience a tip-over and slap down.
In fact, the slap-down case would be more severe than the 5 foot vertical drop because the centers of gravity of the shipping cask and the DSC both exceed 5 feet in the vertical orientation.
Please discuss the reason for not including the slap-down case.
RESPONSE
As explained in response to question 2 of Section 3.1.2.2, the limiting drop deceleration for the DSC is 34g in the horizontal orientation and 48g in the vertical orientation._ The users of NUHOMS Systems need to ensure the DSC is ' not subject to a drop which could 4
cause deceleration values in excess of these values during handling operations.
Nevertheless as shown in Figure 1
(Page 8.45) the effective height of the drop during a slap down is the difference in the cask center of gravity' while standing on its corner and in the horizontal position which is less than 5 feet, and as such the deceleration associ-ated by the slap down is enveloped by the deceleration of the 5 ft horizontal drop.
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NUH-001-8 8.44
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001,
i i
QUESTION:
98 I
+
8.2.5.2 Accident Analysis 4.
In order for the statement of page 8.2-31, "The liner eliminates any secondary impact forces generated by the impac t of the two surfaces" to be evaluated, NUTECH should specify what the liner material-is and what the dimensions of the liner i
are.
Without these data it is not possible to evaluate what the attenuation or amplification i
characteristics of the liner might be.
RESPONSE
The liner material is stainless steel 304.
The
- dimensions of the liner will be dependent on the cask dimensions and fabrication tolerances of the DSC.
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NUH-001-8 8.45 i
-a n- - -,.
v s.
-c--
- - =,
--e-----,...,---m
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 I
1-QUESTIONS:
99 l
l 8.2.5.2
~ Accident Analysis 5._
It appears ' that the effect of the 2 inch diameter i
rods was not considered for the horizontal drop case.
NUTECH should discuss this omission, or recalculate the stresses in the spacer disc.
l 4
RESPONSE
J The weight of the rods. in comparison to the weight of the seven fuel assemblies and spacer disk.are negli-gible.
Their additional weight would only have a
. minimal impact on the stresses within the spacer disk.
The total weight of the 4x26 inch segments of the 2 inch diameter rods is 96 pounds while the total weight of the
.26 inch segment of the 7 fuel assemblies and spacer disk is 1818.1 pounds.
Additionally, the rod locations are 3
~
in an area away from the fuel assemblies, areas con-sidered to be critical stress locations.
Because of their small incremental weight and location, the 2 inch diameter rods were not included in the stress analysis of the spacer disk.
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J NUH-001-8 8.46 l
l
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 j
QUESTION:
100 t
8.2.8 Dry Storage Canister Leakage
[
The statement - that the DSC is designed for no leakage under any conditions cannot be substantiated since very small leakage rates cannot be detected or measured.
i Helium is difficult to contain, so it would be expected-that some decrease in Helium concentration would occur over the lifetime of the canister.
Please provide an i
estimate based on experimental data or analysis of the leakage rate or concentration versus time.
RESPONSE
j Please see the answer to Question 1, Section 1.1.
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NUH-001-3 8.47 i
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e-.,,,4-mme e w.w-,-er
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,,wr*y,,,,w.-t-.,m p.%~,ew..w-g
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4 DETAILED COMMENTS ON NUTECH l
TOPICAL REPORT NUH-001 QUESTION:
101 8.2.9.1 Accident Analysis l
l 1.
What is the basis for using a value of 25% fission I
gas release fraction as opposed to the 30% upper bound cited from reference 8.42 (should read 8.43) and also quoted on page'8.2-54 in Section 8.2.8?
RESPONSE
I As stated in response to question 8.1.1.1, an EPRI study of 124 fuel rods with a large range of linear heat ratings, burnups, enrichment, fuel pellet densities, and operating histories has shown only 0.33% to 12% of the fission gas is released.
Futhermore, the EPRI document is much more conclusive than reference 8.43.
Also, the 30% upper bound cited on page 8 2-54 is used for radio-logical assessment and not for presqure calculations.
l Therefore, the 25%
fission gas release exceeds the maximum documented fission gas release of 12%.
The 30% release fraction of reference 8.43 is not an
" upper bound" of actual release f ractions 'but is the authors' selected conservative value for radiological j
assessment.
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NUH-001-8 8 48 l
}
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_ = _ _ _
5 DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 1
i i
QUESTION:
103 l
8.2.9.1 Accident Analysis i
3.
As previousl'y mentioned in comment. 4 for Section 8.1 1.1, the effect of diurnal temperature i
variation on DSC fatigue should be considered in I
addition to the 50 seasonal cycles.
/
)
RESPONSE
DSC fatigue analysis due to diurnal temperature variation was included in the response to question 2 of Section 1.2.2.
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i NUH-001-8 8.49 1
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.. _ _ _ _ _ _. _ _.,.. _ _.. ~ _ _. -.. _ _. _. _ _. _ _ _
DETAILED COMMENTS ON NUTECH J.
TOPICAL REPORT NUH-001 j
i QUESTION:
104 1
8.2.9.1 Accident Analysis
}
4.
What is the ' maximum gas inventory within one fuel assembly including fill and fission gases?
How was this maximum inventory established?
Do criteria need to be established on maximum gas inventory of fuel to be stored?
Please discuss.
i 1
4
RESPONSE
The fill pas for one fuel assembly was assumed to be 9669.6 in at 72*F and 1 atm while the maximu}.- fission gas contained in one fuel assembly was 25071 in at 72*F and 1 atm.
The volume of fill gas based on a free volume of a fuel assembly being filled with helium at 1
500 psi and 72*F.
The fission gas inventory is based on a fuel assembly burnup of 35,000 MWD /MTM and a fission gas production rate of 1 atom per 4 fission.
Because of the high fill gas pressure, 500 psi, that was assumed no criteria needs to be established on the i
maximum gas inventory.of. the fuel to be stored.
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NUH-001-8 8.50 4
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l DETAILED COMMENTS ON NUTECH TOPICAL. REPORT NUH-001 QUESTION:
105 8.2.10 Load Combinations 1.
The load combinations presented in Table 8.2-10 (Page 8.2-61) do not consider the effects of either wind, tornado or flood on the HSM.
Shouldn't these loads be considered as required by ACI 349-80.
Section 9.2.1?
Also shouldn' t' the material pro-perties be evaluated at the worst thermal case
- examined, i.e.,
ambient T=125'E with inlets plugged?
(See page 8.1-70).
RESPONSE
As stated on page 8 2-60 many of the general event combinations, as shown in Table 3.2-5 are enveloped by l
the load combinations shown in Table 8.2-10.
Hence the effects of wind, tornado and flood were considered, but were enveloped by the cases listed in Table 8.2-10.
Results of. these accident analyses presented in Chapter i
8 verify this statement.
The material properties for the concrete are taken at the temperature of. the individual load combination.
For example, material properties for load combinations 2, 4,
and 10 are taken at the the rmal operating temperature since 'these combinations are -for operating conditions.
Load case 7,
however, contains thermal accident loads..
4 Therefore, material properties for this combination are taken at the HSM temperature when air-inlets are blocked and ambient tempera t.tre is 125*F.
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i NUH-001-8 8.51 m.-,_
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
106 8.2.10 Load Combinations 2.
Please provide an analysis of the HSM corbels, bearing plates, bolts and reinforcing steel for the load combination cases #7 and 10.
The tempera ture of the corbels should be specified and the material properties for temperature should be used.
RESPONSE
As stated in page 8.1-32 of the Topical Report, slotted holes are used in the wideflange support beams for the
. purpose of allowing thermal movement, and as such there is no direct the nnal force transferred to the corbels.
- Furthe rmore as stated in the response to question 3 of Section 8.1 1.5 a multiplier of 1.7 was used for various load combinations presented in Table 8.2-12.
The use of 1.7 multiplier conservatively envelopes all the' load combination cases specified by ACI 349-80.
-Table 8.2-12 will be revised to incorporate the capacity of the corbel, bearing ' pads and the bolts at both the operating and accident temperatures.
NUH-001-8 8.52-i
DETAILED COMMENTS ON NUTECH j
TOPICAL REPORT NUH-001.
1 QUESTION:
107 l
1 8.2.10 Load Combinations 3.
Note 7 in Table 8 2-10 on page 8.2-61 implies that i
the material properties for the the rmal accident load are taken at 300*F, however Table 8.1-2 shows that the maximum concrete temperature for the inside roof is 321*F, and the side wall is 344*F.
4 Please clarify.
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RESPONSE
Note 7 in Table 8.2-10 on page 8.2-61 will be changed to 350*F to conservatively bend all accident temperature throug hout the HSM.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
108 8.2.10 Load Combinations 4.
Table 8.2-11 reports DSC support assembly stresses for various load combinations.
The note (3) indi-cates allowable stresses for the SA 36 structural steel at 200*F.
NUTECH should show what the temperature of tne DSC is where it and the Tee support assembly rails are in contact for the worst thermal case examined, i.e., ambient T=125*F with inlets plugged (see page 8.1-70).
RESPONSE
As stated on page 3.2-12, only one accident is con-sidered to occur at one time.
Therefore, the HSM and DSC are considered to be at temperatures associated with the normal operating conditions, 70*F ambient tempera-ture during a seismic event or when friction loads are applied.
The jammed DSC event is considered to be only credible during the HSM loading operation when the HSM is at ambient temperature 70*F.
The allowable stresses for note 3 will be changed to 250*F.
This bounds the conservative assumption that the temperature of the DSC support assembly is equal to the exterior surface temperature of the DSC, 230*F.
NUH-001-8 8.54
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
109 10.0 OPERATING CONTROLS AND LIMITS 10 1 Proposed Operating Controls and Limits The fuel stored should also have an axial power peaking of equal or less than 1.08 for decay heat, since this parameter was used in the thermal design.
RESPONSE
As stated in response to Question 8.1.3.1-2, the axial peaking power factor was intended to be a conservatism in the thermal analysis not a limiting parameter on the fuel that may be stored in NUHOMS.
The only thermal parameter for fuel is the hea*. output per fuel assembly must be limi ted to 1.0 kw, as indicated in Section 10.3.1.1.
NUH-001-10 10.1 4
i DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
110 10.2.2.2 Technical Conditions and Characteristics 1.
Condition 3 regarding the DSC helium leak rate of the primary weld should apply to all welds on the canister including the axial seam weld.
Leak rate as a function of temperature also needs to be addressed.
RESPONSE
Please see the answer to Question 1, Section 1.1.
s NUH-001-10 10 2
j DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
111 10.2.2.2 Technical Conditions and Characteristics 2.
Why is the limiting heat load of 1 kw/ assembly not included in the list of seven technical ennditions and characteristics?
RESPONSE
The heat load of 1 kw per assembly is considered to be a functional limit as discussed in Sections 10.2.1 and 10.3.1.
NUH-001-10 10 3
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
112 10.2.3 Surveillance Requirements 1.
What provisions have been made for inspection of internal air passages for blockage due to buildup of organic matter, insect activity, etc.?
RESPONSE
Visual inspection of the internal air passages will be conducted from outside of the HSM.
Any regions of the internal air passages that are not visible from outside are in an area which is not conducive to the growth of organic matter due to radiation and temperature levels.
See response to Question 1.2.2.1 for additional information.
e NUH-001-10 10 4 f.
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 4
QUESTIONS:
113 i
10.2.3 Surveillance Requirements 2.
What provisions have been made for drainage of any 1
water which enters the HSM through the inlets or outlets?
Buildup of water could also cause partial or complete blockage of inlet flow in ' the absence of drains.
What inspections are performed to as-sure that water does not accumulate, in the bottom of the HSM?
RESPONSE
Two 1 inch diameter pipes will be installed in the front wall of the HSM.
In the event that any water should enter through the air inlet the pipes will allow for water to drain from the air inlet chamber.
This will elimina te the possibility of water building up and
)-
causing a partial or complete blockage of the inlet flow and the necessity of inspecting passage ways.
The' air outlets are designed to be water tight and therefore water will not be allowed to flow into the HSM.
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NUH-001-10 10.5
e DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
114 10 2.3 Surveillance Requirements 3.
Surveillance is performed every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to assure e.
that air is flowing through the module.
How will acceptability of the flow rate be determined?
How will the flow rate be measured, or how will it be determined that the flow criteria are satisfied.
RESPONSE
Routine surveillance or inspection of the air inlets conducted on a weekly basis and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after an unusual events.
NUH-001-10 10.6
j DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 1
' QUESTION:
115 a
10.3 1.1 Fuel Specifications 4
Please clarify which parameters are for an assembly and i
which are for the
- canister, e.g.,
neutron
- source, seight.
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RESPONSE
1 1
The following lis t clarifies which parameters are for the assembly and which are for the canister.
This table will be incorporated into the revised Topical Report.
1 Burnup i 33,000 mwd /Mt Post irradiation time
> 5 years Initial enrichment i 3.5 %
g 235 Weight per distance between any 1 665 kg adjacent spaces Decay power per assembly i 1 kw r.
8 Neutron source-per canister 9.98x10 n/sec 4
Gamma source for canister 7.76x1915 photons /sec i
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NUH-001-10 10.7 i
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 l
QUESTION:
116 10.3.2.1 DSC Vacuum Pressure During Drying What is the basis for the one hour time at pressure criteria for the drying operation?.Has the possibility of a waterlogged rod been considered?
1 S
RESPONSE
Based on engineering judgement one hour time at the specified pressure should be sufficient.
A water logged fuel assembly has not been considered.
Prior to inserting the fuel assemblies into the DSC, a i
fuel assembly will be visually examined for structural j
and mechanical integrity.
Additionally, at the temperature and vacuum pressure which the fuel rods will be exposed to during the drying operations, any water in the fuel rods should evaporate and be suctioned from the DSC cavity.
If any water does not come out under the vacuum conditions it is not likely to come out with 1 Atm of He in the canister.
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5 NUH-001-10' 10.8 i
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
117 10.3.2.2 DSC Helium Backfill Pressure It appears that a hold time at this pressure will need to be specified to assure that the rmal equilibrium has been reached.
RESPONSE
As discussed in the response to Question 5.1.1.3-1, the backfilling procedures have been changed to include monitoring the pressure.
NUH-001-10 10.9
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
118 10.3.2.3 DSC Helium Leakage Rate of Primary Weld 3
1.
Calculation of 631 cm of He appears to correspond to 2
atmosphere of pressure rather than 1.5 atmosphere.
RESPONSE
3 of He does correspond to 2 atmosphere and 473 6313, cm 3
cm corresponds tp 1.5 atmosphere.
631 cm will be revised to 473 cm.
This change has no impact on the leak rate specification of the primary weld.
NUH-001-10 10.10
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 i
i 00ESTION:
119 10.3.2.3 DSC Helium Leakage Rate of Primary Weld 2.
The leak rate corresponds to the detection level of the helium sniffer.
- However, the calculation assumes a single leak at this level.
In reality, i
leakage may occur at several locations along the many feet of weld and by diffusion through the metal matrix of the caniser itself.
Has this type of leakage been considered?
Please address.
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RESPONSE
Please see the answer to Question 1, Section 1.1.
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I NUH-001-10 10.11 l
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DETAILED COMMENTS ON NUTECH
]
TOPICAL REPORT NUH-001 QUESTION:
120 10.3.2.7 Maximum Air Exit Temperature l.
If the temperature rise is greater than 100'F and i
small fans are provided as stated in Item 5,
the passive nature of the cooling mechanism will be lost.-
Since.the design is based on achieving passive cooling, the addition of fans should either be included in the design or eliminated as-a corrective action.
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RESPONSE
The addition of fans will be eliminated as a corrective action in the report.
If the temperature rise is greater than 100*F, the DSC will be removed from the HSM or additional information and analysis will be provided
}
to'show. that the existing condition does not represent i
an unsafe situation.
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o NUH-001 10 12
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
121 10 3.2.7 Maximum Air Exit Temperature 2.
The cooling air temperature rise is to be checked twenty-four hours after placement of the canister i
in the HSM.
What features of the design preclude the need for later measurements to assure that the 100*F temperature rise criterion is satisfied.
If j
such assurance cannot be
- provided, then more frequent measurements may be necessary.
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RESPONSE
There are no moving components that could significantly alter the air flow or cause an increase in the temper-ature change.
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NUH-001-10 10,13 i
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DETAILED COMMENTS ON NUTECH I
TOPICAL REPORT NUH-001
(
QUESTION:
122 10.3.3.1 Surveillance of the HSM Air Inlets 1.
How does this surveillance assure that there is no blockage of the outlets or internal blockage?
Does surveillance include any measurement of flow or temperature rise of the cooling air?
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RESPONSE
There is no reason to inspect any other HSM air pathways other than the air inlets.
Bird screen - on the air inlets and outlets will prevent any debris which may significantly alter the air flow through the HSM. -There are no lose or moving parts within the HSM that could fail and block the air. flow.
Additionally, the non-visible portions of the air pathways are not conducive to the growth of organic matter.
Therefore, inspection of the air inlets and outlets, the only points where air flow could be restricted, will assure that there is no blockage of the outlets or intervals.
Surveillance does not include measurement of air flow or temperature rise of the cooling air.
4 NUH-001-10 10.14
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 l
QUESTION:
123 10.3.3.1 Surveillance of the HSM Air Inlets 2.
Page 10.3-17 state:s that for " normal operations, inspection of air inlets once per week will assure that any local obstructions can be removed".
If the HSMs are inspected once per week, ' how can NUTECH asse:ae that an air blockage would be no longer than -
hours?
This question also relates
'~
to Section 8.2.7.2.
RESPONSE
Analysis in Chapter 8 showed that no temperature limits are exceeded if all inlets of the HSM are completely blocked.
The complete blockage of all inlets and i
outlets is not expected to occur except during an accident condition.
As stated in Section 10.3.3.1, the air inlets will be inspected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of an abnormal event.
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NUH-001-10 10 15
DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
124 DETAILS OF HEAT TRANSFER ANALYSIS OF THE NUHOMS SYSTEM 1.
On page B.4, it is assumed that the gas behaves the same as the gas in a horizontal annulus.
- However, on page B.5, a formula said to be applicable for a vertical annulus is used to determine apparent thermal conductivity.
Please explain this dis-crepancy.
The gas
" annulus" is horizontal for drying and vertical for storage.
RESPONSE
Reference to the vertical annulus on page B-5 is incorrect.
The words vertical on page B-5 and B-6 will be changed to horizontal.
The equation listed in Appendix B is-used to determine the thermal conductivity of gas in a horizontal annulus.
When the cask is in the vertical position, the most sever temperature conditions will be encountered during vacuum, dry of the DSC.
As stated on page B.6, the apparent thermal conductivity of the gas in the thermal model was approximated by dividing the' apparent thermal conductivity of_ air in a horizontal position by 1/760.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 4
1 4
QUESTION:
125 DETAILS OF HEAT TRANSFER ANALYSIS OF THE NUHOMS SYSTEM I
2.'
The yangth>of the fuel assembly L is given on page i
v.
B '. 5, but does not appear to be used in any of the equations.
Is this correct?
Why is the number s
c i t e s ?,,.'
i
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RESPONSE
l 1
J Length of*the fuel assembly, L,
is not required and will j
be removed in the revised Topical Report.
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j-NUH-001-N 11 2-4-
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001
~ QUESTION:
126 DETAILS OF HEAT TRANSFER ANALYSIS OF THE NUHOMS SYSTEM 3.
Table 8.1-13 gives the average helium temperature for air inlet temperatures of -40, 70 and 125'F as
- 315, 389 and 429*F.
Table B-2 gives apparent helium conductivity at three values of delta T for average gas temperatures.of 80, 160, 440 and 620*F.
A value of 1.6 was determined for apparent thermal conductivity by averaging the values in Table B-2.
In light of the range of helium temperatures for which analyses were performed, it does not seem appropriate to include the values-at 620*F, and particularly at 80*F in arriving at the average value.
Justify that 1.6 - is a representative value for apparent thermal conductivity.
RESPONSE
The average value of the apparent helium conductivity for the two temperatures values of 160 and 440*F is I
' l.55.
This 0.05 difference in average value will have an insignificant: dif ference-in the temperature.
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J NUH-001-N 11.3 f
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o DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
127 DETAILS OF HEAT TRANSFER ANALYSIS OF THE NUHOMS SYSTEM 4.
For the calculation of h
on page B.13, the assumption of delta-T 300gg o F is stated, but does
=
not appear to be needed.
RESPONSE
The assumption of delta-T 300*F is necessary.
This
=
assumption is used later in the derivation on page B-15.
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 I
f QUESTION:
128 DETAILS OF HEAT TRANSFER ANALYSIS OF THE NUHOMS SYSTEM 5.
Contrary to the statement on page B.16, the two-i step approach to using the Wcoten-Epstein Formula (WEF) appears to be closer to the actual i
application.
The second step is ' essentially the manner in which the result is used.
RESPONSE
i The first step is deriving the b' oral sleeve temperature considering the DSC wall temperature and a
pseudo assembly of 7 rods.
The second step goes from the boral i
guide sleeve temperature to the central rod temperature I
using a single assembly.
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I NUH-001-N 11.5
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DETAILED COMMENTS ON NUTECH TOPICAL REPORT NUH-001 QUESTION:
129 DETAILS OF HEAT TRANSFER ANALYSIS OF THE NUHOMS SYSTEM 6.
Corrected values for Kfuel for the alternative WEF approach should be provided.
RESPONSE
The correct value for Kfuel using the WEF on a single 15x15 fuel assembly
- "9"
- E' H
temperatures is 5.23 k e, NUH-001-N 11 6