ML20056G630

Responses to Us NRC Comments Re SAR for NAC Storable Transport Cask (NAC-STC)

ML20056G630
Person / Time
Site:	07109235
Issue date:	08/31/1993
From:	NAC INTERNATIONAL INC. (FORMERLY NUCLEAR ASSURANCE
To:
Shared Package
ML19310D674	List: ML20056G626 ML20056G627 ML20056G630 ML20056G632
References
NUDOCS 9309030357
Download: ML20056G630 (19)

Text

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Responses to g

U.S. NRC Comments Regarding the Safety Analysis Report for the l

o NAC Storable Transport Cask (NAC-STC)

O Docket No. 71-9235 1

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Nuclear Assurance Corporation 655 Engineering Drive Norcross, Georgia 30092 Telephone: (404) 447-1144 Telex: 6827020,C827114 Facsimile: (404) 447-1797 4

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. Docket No. 71-9235 August 1993

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This document contains the NAC responses to the NRC O

comments dated December 23,1992. The NAC-STC SAR has been revised to incorporate all of the changes associated with these responses.

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i Docket No. 71-9235 August 1993 l

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!O STRUCTURAL L

i-NRC Comment 7

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The temperatures experienced by the basket under normal and accident l0 conditions appear to exceed the temperatures at which aluminum should be used as a structural material. Justify the use of aluminum as the material of construction for the basket, disks, rods, spacer nuts, fuel tubes, and other components. Specify the codes and standards which authorize use of aluminum at these temperatures over long periods of time and justify that j

'O these codes and standards are applicable to the basket structure and to the specific aluminum alloys used in the design. Show explicitly that the basket design conforms to these codes and standards.

O NAC Response NAC has redesigned the basket removing aluminum from all structural load paths.

The basket fuel support disks, tie rods, spacer nuts, and fuel tubes are fabricated 3

from Type 17-4 PH or Type 304 stainless steel. Use of these materials for fuel

,O basket construction is authorized by ASME Code,Section III, Subsection NG, Core Support Structures, which has been used by NAC as the fuel basket j

governing code. Structural analysis of the basket documented in NAC-STC SAR l

Revision 2 shows that the revised basket design conforms to the applicable ASME O

Code requirements.

I Heat transfer efficiency is maintained in the revised fuel basket design through the j

addition of 6061-T6 aluminum alloy heat transfer fins. Heat transfer fins are not 3

j part of the fuel support structure and are not governed by ASME structural

!O criteria. Support of the heat transfer fins is provided by the steel basket structural l

components. Structural analysis of the steel basket includes load from the weight 4

of the aluminum heat transfer fins.

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Since the fuel basket has been redesigned using ASME Code material and

O structurally analyzed with respect to ASME Code criteria, NAC is of the opirdon I

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.O Docket No. 71-9235 August 1993 STRUCTURAL Response 1 (continued) g that Comment 1 is no longer applicable and is not addressed in further detail in this response.

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Docket No. 71-9235 August 1993 O

STRUCTURAL NRC Comment 2.

Provide the following information concerning the aluminum materials used in O

the basket:

l MIL-HDBK-5E, Figures 3.2.6.4.1 (a) and (b) and the methods of a.

extrapolation to longer exposure time (e.g. 1,000,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />).

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Basis of the conclusion that the yield strength of 2219-T87 aluminum i

alloy in the 300 F to 500 F temperature range decreases as a logarithmic i

function of exposure time.

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Basis that the reduction of strength for the ultimate and yield strength is 2O c.

identical for 2219-T87 aluminum alloy. Note that this relationship is not l

true for 6351-T54 or 2024-T351 aluminum alloy.

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The extent that thermal cycling causes additional degradation in the O

aluminum strength.

NAC Response i

Since the fuel basket has been redesigned using ASME Code authorized stainless O

steel materials and structurally analyzed with respect to ASME Code criteria, NAC is of the opinion that Comment 2 is no longer applicable and is not addressed in l

this response.

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.O Docket No. 71 9235 August 1993 STRUCTURAL e

NRC Comment 3.

Verify that the Borated Type 6351-T54 modified aluminum alloy has a higher yield strength of 400 F, than at 350'F, as shown in Table 2.3.5-2.

NAC Response Borated Type 6351-T54 modified aluminum alloy has been removed from the O

NAC-STC design. Comment 3 is not applicable to the current license application.

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iO STRUCTURAL i

4 NRC Comment i

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It is not clear why the maximum stress in the fuel basket support disk was not O

at the critical area (i.e. near the point of impact). Justify the validity of the l

analysis considering that the impact g-load increased 300% from normal to accident conditions, but the maximum stress increased only 5%.

NAC Response O

Structural criteria proposed for the all aluminum basket evaluated total stresses l

(primary plus secondary) relative to yield strength of the basket material to assure that the material did not yield under any load combination. Combining stress j

O components for impact with thermal expansion stresses in the basket support disk l

results in the highest total stress value and its location being controlled by the j

thermal tension stresses at the periphery of the support disk. Total stress is calculated by algebraically adding the thermal tension stresses to the tension I

i' stresses created by impact, opposite to the actual impact point. Compraine

,O stresses produced by the impact local to the point of impact combine with tension f

thermal expansion stresses resulting in lower total stresses.

Combining stress components from both the primary and secondary stress category l

O resulted in total stresses being controlled by the secondary thermal stress j

components. Although the impact load for normal transport condition impact, j

18 g, is only 33% of the load for transport accident condition impact,55 g, the

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resultant stress does not produce such a large difference.

O NAC has revised the design of the fuel basket removing aluminum as the material j

i for components in the structural load path. The revised basket is designed and j

structurally analyzed in accordance with ASME Code,Section III, Subsection NG.

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Stress categories of primary membrane and primary membrane plus bending are f

evaluated in agreement with ASME Code criteria.

Structural analysis results O

presented in the NAC-STC SAR Revision 2 identify maximum impact stresses local to the point of impact.

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.O Docket No. 71-9235 August 1993 STRUCTURAL g

NRC Comment 5.

Clarify whether the lead was poured from the top of the cask or from the bottom of the cask. Show the procedures to prevent lead contamination for e

the final full penetration weld of the cask.

NAC Res_ponse 8

The lead pour for the NAC-STC is performed with the cask body inverted (closure end down) and vertical. Following completion of the lead pour, the final full-penetration welds of the bottom outer forging to the outer shell and to the bottom inner forging are completed. (Refer to the NAC Response to Comment No.1 in the Acceptance Tests and Maintenance Program for fabrication details).

g As shown in the attached sketch, backing bars are used to prevent lead contamination of the full-penetration welds between the outer shell and the bottom outer forging and between the bottom outer forging and the bottom inner forging.

8 The backing bars are positioned and tack-welded to the outer shell and to the bottom inner forging prior to lead pour. With the cask inverted (closure end l

down), a temporary " dam" is installed at the outer shell/ bottom outer forging / bottom inner forging juncture to support the open end of the outer shell 1

during the lead pour and to facilitate pouring of the entire length of the lead j

O cylinder. Following cooldown after the lead pour, the " dam" is removed and the lead is machined to its final configuration, including facing off the backing bar to 4

l ensure that no lead remains on the weld site of the backing bar. The two final j

full-penetration welds of the bottom outer forging to the outer shell and to the bottom inner forging are then performed as specified in Zones B2 and B3 of NAC e

Drawing No. 423-802, Sheet 3, where backing bars, as well as PT Root and Final Pass weld examinations, are required. The backing bars prevent the lead from reaching the final full-penetration welds. The liquid penetrant (PT) examination of the root pass of those welds is the final check for contamination prior to completion of the full-penetration welds. De PT examination will indicate any ei contamination, including lead, that is present in the welds.

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Docket No. 71-9235 August 1993 O

Location of Backing Bar O

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8^cxi"o e^a OUTER

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BOTTOM O

3 INNER FORGING f

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FORGING k

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.O Docket No. 71-9235 August 1993 STRUCTURAL g

NRC Comment 6.

The shear stress on the weld at the base of the rear trunnion (page 2.5.2017) exceeds the yield stress limit specified in 10 CFR E71.45(b). In addition, the e

maximum stress at the weld and cask interface was calculated non-conservatively because Mx and My (page 2.5.2-20) were not summed correctly.

8 NAC Response The analysis of the weld / rotation trunnion recess interface that is presented on SAR pages 2.5.2-15 through 2.5.2-18 has been rev;-wed and several inconsistencies have been identified. The weld analysis has been revised to: (1) use a 11/2 J-g, groove weld in combination with a 13/8 fillet weld; (2) use the correct equation for torsional shear stress on a box section from Table 14 in Advanced..ahanics of Materials by Seely and Smith; (3) calculate the combined bending, direct shear, and torsional shear stresses properly to obtain the von Mises equivalent stress at Points A, B, and C; and (4) calculate the margin of safety based on the material yield strength of the E309 weld filler material (S = 61.0 ksi), which has a lower y

yield strength than the Type 17-4 PH stainless steel rotation trunnion recesses.

The calculated margin of safety is +1.13. so 10 CFR 71.45(b) is easily satisfied.

The-analysis of the section at the weld / cask body interface that is presented on pages 2.5.218 through 2.5.2-20 has been reviewed and several inconsistencies have been identified. The weld / cask body interface section analysis has been revised to:

(1) use a 1 1/2 J-groove weld in combination with a 13/8 fillet weld; (2) use the correct equation for torsional shear stress on an open-box section from Table 14 in e

Advanced Mechanics of Materials by Seely and Smith; (3) calculate the combined bending, direct shear, and torsional shear stresses properly to obtain the von Mises equivalent stress at Points A, B, and C ;and (4) calculate the margin of safety based on the material yield strength of the Type 304 stainless steel forging material (S = 22.5 ksi), which has a lower yield strength than the E309 weld filler material.

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The calculated margin of safety is +0.03. so 10 CFR 71.45(b) is satisfie.d.

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J Docket No. 71-9235 August 1993 r

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O STRUCTURAL Comment 6 (continued)

The NAC-STC license drawing No. 423-802 has been revised to specify a 1 1/2 J-l groove weld in combination with a 13/8 fillet weld for the attachment of the rotation trunnion recesses to the cask body.

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Chapter 9, References, has been revised to include the text, Advanced Mechanics j

of Materials by F.B. Seely and J.O. Smith.

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.O Docket No. 71-9235 August 1993 STRUCTURAL e

NRC Comment 7.

Revise Drawing No. 423-019, sheet 1 and sheet 3, to clearly specify the actual thickness of the steel plate used to construct the outer shell of the quarter scale model. Note that if a thickness different than 0.665 inches was used, it is not clear that the model would be representative of a full scale prototype with a shell thickness of 2.65 inches.

NAC Response The 3/4-inch plate identified in the Description block of the List of Materials for the Outer Shell on Sheet 1 of Drawing No. 423-019 is the rough stock plate thickness that was to be used to fabricate the Outer Shell. The 3/4-inch plate was a

rolled and welded to form a cylinder. That cylinder was then machined on both the inner and outer surfaces to obtain the 21.68-inch outside diameter and the 20.35-inch inside diameter as specified on Sheet 3 of Drawing No. 423-019.

Therefore, the thickness of the Outer Shell of the NAC-STC quarter-scale model is 8

(21.68 - 20.35)/(2) = 0.665 inch and the model is fully representative of a full-scale prototype with an Outer Shell thickness of 2.65 inches.

Drawing No. 423-019 has been revised to clarify the outer shell thickness: Sheet 1 -

In the description of Item 1 in the Bill of Materials, "3/4" has been deleted; and Sheet 3, Zone E5 - a reference dimension of (.665) for the outer shell thickness has been added.

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Docket No. 71-9235 August 1093 O

STRUCTURAL NRC Comment 8.

Revise Section 2 7.5.0 to show there will be no inleakage of water under O

hypothetical accident conditions. Note that contrary to the statement in Section 2.7.5.0, the criticality safety analysis for an array of packages did not consider inleakage of water.

NAC Response O

Section 2.7.5.0, Immersion - Fissile Material, has been revised to demonstrate that there will be no inleakage of water under hypothetical accident conditions. A discussion similar to that presented in Section 2.7.6.0, Immersion - All Packages, O

has been prepared. The discussion utilizes the detailed analysis results in the SAR and the metallic o-ring manufacturer's specifications to demonstrate that the NAC-

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STC easily withstands the external pressure due to the immersion event, which l

j follows the free drop, puncture, and fire events in the hypothetical accident l

sequence. Therefore, there will be no inleakage of water into the cask and the

,O requirement of 10 CFR 71.73(c)(4) is satisfied for the NAC-STC.

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.O Docket No. 71-9235 August 1993 CONTAINMENT g

NRC Comment 1.

Revise the containment analysis to calculate the allowable release rate for normal and accident conditions based on respective quantities determined 9

from the rcleasable inventory (given in Table 4.2-2) rather than the total inventory (given in Table 4.2-1). Assume that 30% of the total fission gas inventory is releasable. Note that only 10% of the tritium was considered releasable (see page 4.3-1).

O NAC Response NAC has revised the containment analysis for normal and accident conditions based on the releasable inventory given in Table 4.2-2. A 30% release rate for tritium was used in the revised analysis. The revised analysis results in a maximum allowed leak rate of 7.18 x 10 std em'/second.

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Docket No. 71 9235 August 1993 3

O CONTAINMENT NRC Comment i

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Revise Section 4.1.3 to specify that a periodic leak test (not an assembly i

,O verification leak test) shall be performcd on the contaimnent seals after i

each loading of the cask. Note that the containment seals are metallic and must be replaced after each use. Also, revise the operating procedures and maintenance program accordingly.

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NAC Response l

t Section 4.1.3 has been revised to specify that a periodic leak test shall be l

performed on the containment seals each time the seal is replaced, which O

because the seals are metallic, will mean each time the cask is loaded with fuel. All references to an assembly verification leak test for the NAC-STC i

containment have been removed from the SAR.

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.O Augue 1993 Docket No. 71-9235 CONTAINMENT gj NRC Comment 3.

Show that the sensitivity of the fabrication leak test is sufficient to e

support the statement in Section 4.1.3.1.1 that, "This test verifies that the combined leakage rate of the containment vessel seals does not exceed

[the allowable leak rate calculated under item 1 above]... "

NAC Response O

The statement regarding the purpose of the fabrication verification leak test mistakenly referred to the combined leakage of the containment vessel. In accordance with the requirements of ANSI N14.5, the statement has been changed to say that, "This leak test verifies that the leakage rate of the e

assembled containment does not exceed the maximum allowable leakage rate i

of 7.74 x 10 std cm'/second [the allowable leak rate calculated under item 1 4

above].

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Docket No. 71 9235 August 1993 O

SHIELDING NRC Comment Evaluate the stability and effectiveness of the NS4FR material over long periods of O

time in service. Show that after the maximum expected storage period, at the normal operating temperature of 315*F, the degradation of the NS4FR would not cause the normal condition dose rates to exceed the limits in 10 CFR Part 71.

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O The NAC-STC neutron shield materials are designed to provide sufficient neutron shielding during cask transport immediately following fuel loading or folkswing extended storage to satisfy the dose rate criteria for normal conditions of transport.

O The BISCO NS4FR material was selected for use because of its stability at elevated temperatures.

In tests performed by the manufacturer, the long-term performance and stability of the material has been demonstrated. As discussed in Section 3.3.2 of the NAC-O STC SAR Revision 1, BISCO performed a test of material stability at 338*F for 145 days. During this test, a maximum wieght loss of less than 4 percent occurred, with a significant fraction of the weight loss occurring in the first 30 days at temperature.

O Due primarily to the incre-ase in the minimum cool time of the design basis fuel, the design basis heat load of the NAC-STC has been reduced to 22.1 kW. Based on the revised thermal analysis, the peak neutron shield temperature has been recalculated. The revised peak calculated temperature for the NAC-STC neutron O

shield under transport conditions is 285 F. This maximum temperature is 53 F lower than the BISCO test temperature. In addition, the peak temperature occurs only at a localized area at the aial center adjacent to the inner shell with the remainder of the neutron shield material well below the 285*F value. Therefore, it is expected that the degradation to the shield performance will be much less than O

the 4 percent loss experienced at 338'F.

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.O Docket No. 71-9235 August 1993 SIIIELDING Comment (continued)

Additional tests were performed by Hitachi Zosen Corporation (HZ) on the BISCO NS4FR material, where it was found that at a temperature of 175"C (347*F), a weight loss of less than 1.5 percent was measured after 73 days at temperature. A separate test performed at a temperature of 150 C (302*F) e produced a weight loss of less than 0.5 percent after 73 days. After 56 weeks at 150 C, the weight loss was approximately 1.2 percent. In an extrapolation of the test data performed by HZ, a weight loss of less than 2 percent was predicted for a 20-year exposure period.

O Both of the material performance tests completed by BISCO and HZ were carried out at a constant bulk mass temperature. During extended storage prior to transport, temperatures will decrease as the storage period increases, resulting in less limiting neutron shielding conditions than those predicted by the test cases described above.

From an analysis of these test results, it is expected that the maximum weight loss of the neutron shield will be less than 2 percent after a 20-year period. Based on the dose rate contributions for neutrons and gammas presented in Table 5.1-5, a 2 percent reduction in the effectiveness of the neutron shield will not result in dose rates exceeding the normal transport dose rate limits of 10 CFR 71. As noted in Chapter 8 of the SAR, both neutron and gamma dose rates will be measured and recorded prior to transport to verify that they are less than the 10 CFR 71 dose e

rate limits.

The previous discussion documenting the long term temperature stability of the BISCO NS4FR has been added to Chapter 3 of the SAR.

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Docket No. 71-9235 August 1993 O

CRITICALITY NRC Comment 1.

Specify the following parameters for e.ach of the four major fuel assembly O

classes (i.e.14x14,15x15,16x16, and 17x17) to be snipped in the Model No.

NAC-STC package:

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• fuel fonn cladding material O

maximum initial uranium content per assembly (kg) l l

maximum initial uranium-235 enrichment (wt%)

maximum bundle cross-section (cm) j

• fuel pellet diameter range (cm) i

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= maximum active fuel length (cm) 4 i

NAC Response 4

l Tne following table of assembly class parameters will be included in the SAR.

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.O Docket No. 71-9235 August 1993

'lable I Fuel Assembly Class Data Assembly Class 14x14 15x15 16x16 17x17 Fuel Fonn Intact fuel Intact fuel ktact fuel Intact fuel UO pellets e

UO2 Pellets UO2 Pellets UO2 Pellets 2

Cladding Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Uranium Content (kg/ assembly) 457 464 426 461 Initial Enrichment 4.2 4.2 4.2 4.2 Maximum Cross-Section g

(cm) 20.96 21.68 20.96 21.68 (in) 8.25 8.536 8.25 8.536 Fuel Pellet Diameter Range (cm)

Minimum 0.819 0.890 0.826 0.770 Maximum 0.966 0.940 0.826 0.821 Maximum Active Fuel Length (cm) 369 366 366 366 (in) 145.20 144 144 144 e

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Docket No. 719235 August 1993 i O CRITICALITY NRC Comment 2.

The footnote on page 6.2-2 states that the criticality analyses used a O

uranium mass of 469 kg per assembly. Show how this value was obtained. From the computer input / output in Appendix 6.6, it appears that the calculations were based upon a uranium mass of 462 kg per assembly.

NAC Response As a result of the basket modification, the criticality analyses for the NAC-STC have been revised. He revised calculations are based on the 469 kg O

uranium mass as specified in the footnote on page 6.2-2.

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.O Docket No. 71-9235 August 1993 CRITICALITY NRC Comment 3.

Show that the K rr for a single package is less than 0.95 (when corrected e

for bias and uncertainty) assuming 75% credit for the poison, optimum e

water moderation, and optimum reflection. Also, assume the contents have the most reactive configuration consistent with the accident conditions of 10 CFR 671.73.

NAC Response The revised criticality evaluation presented in Chapter 6 of the SAR for normal conditions of transport and the hypothetical accident shows that k, is less than 0.95 based on 75 percent boron effectiveness, optimum water moderation, and optimum reflection.

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Docket No. 71 9235 August ?993 O

OPERATING PROCEDURES NRC Comment 1.

Show that the leak test procedures can reliably detect the leakage of O

small amounts of helium, considering the distance from the seal to the detector probe. Show how the sensitivity of the leak detection system will be established.

NAC Response O

Due to the distance from the edge of the inner lid to the centerline of the outermost inner lid metallic o-ring (2.75 inch). the helium leak test procedure to be utilized to verify the leakage rate from the lid o-rings has been changed from O

the helium mass spectrometer leak detector (MSLD) sniffer probe method, to the I

evacuated envelope method. This change has been incorporated into the procedures of Chapter 7, and into the leak test requirements of Chapter 8.

f In the modified procedure, the outer lid, which is also sealed with a metallic i

O o-ring, will be installed. The test chamber (envelope) thus created by the interlid volume will be evacuated through the interlid ODVN. Helium pressure in the inner lid seal interspace will be allowed to leak past the outer o-ring into the isolated evacuated interlid volume for an established time period. The isolated v lume will then be evacuated through a MSLD to determine the helium leakage O

from the inner lid seals. This leak test method provides a high sensitivity test l

which is readily reproducible as the isolated and evacuated volume is provided by normally installed cask components. A certified calibrated helium leak will be used prior to and following the leak testing to assure proper equipment O

performance and system sensitivity.

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8 August 1993 Docket No. 71-9235 OPERATING PROCEDURES gi NILC Comment 2.

Revise Section 7.1.2, " Preparation of Cask for Loading," to include an 9

inspection for damage to the inner lid and coverplate bolts.

NAC Response The procedures of Section 7.1.2 have been revised to include inspection O

requirements for the inner lid and coverplate bolts.

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Docket No. 71-9235 August 1993 i

O OPERATING PROCEDURES NRC Comment 3.

Section 7.1.3, " Loading Fuel" j

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Revise Step 1 to specify that fuel assemblies with missing fuel pins may a.

be shipped only when dummy fuel pins are used to displace an amount of water equal to that displaced by the original pins.

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Revise Steps 16,21. 22,23 and 24, to require a fabrication verification leak test on all containnsent seals.

Revise Steps 16,18 and 21 to give the pressure values as absolute.

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When giving an observation period for a pressure rise test, also give the j

necessary pressure gauge sensitivity (e.g. see Step 23.b.).

NAC Response O

The note of Step 1 was_ expanded to include requirements that fuel s

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r semblies may be loaded and shipped only when dummy fuel pins are j

installed to replace missing pins, and they displace an amount of water equal to that displaced by the original pins.

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The loading procedure has been modified to require the performance of a containment system periodic verification leak test on all containment seals.

f The series of leak tests on the various containment boundary components

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,O and sents has been modified to incorporate the leak test procedure changes j

referred to in the NAC Response to Operating Procedures Comment 1.

The pressure values provided in Steps 16,18 and 21 of Section 7.1.3 have c.

been clarified as absolute values, as appropriate. The sequence of steps has

.O been changed to incorporate leak test procedure changes.

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.O Docket No. 71-9235 August 1993 OPEPATING PROCEDURES Response 3 (continued) d.

The current drying procedure gives an observation period of ten minutes.

When the cask cavity is not dry (e.g., a significant amount of unevacuated water and water vapor remains), the pressure will rise appreciably after isolation of the vacuum pump. Therefore, a test sensitivity of 5 mbar will e

be sufficient to validate the cavity dryness. The NAC-STC Operations Manual will describe and detail the specific test equipment and systems to be utilized in the operation and testing of the cask. This procedure wih require the use of a calibrated vacuum gauge with a range of between 0-50 mbar and 0-100 mbar with a minimum gauge readability of 2.5 mbar, which will ensure that the minimum test sensitivity is met.

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Docket No. 71-9235 August 1993 O

OPERATING PROCEDURES NRC Comment 4.

Section 7.2.2, " Preparation for Transport (after extended storage)"

O Revise the section to discuss the steps that will be taken to ensure that a.

after extended storage a sufficient amount of helium remains in the

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cask cavity to inert and cool the contents.

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Revise Steps 10.h and 12.h to state that the pressure is absolute pressure.

Revise the section to specify the action to be taken if a containment c.

O seal fails the leak test.

NAC Response a.

Section 7.2.2 has been revised to include a discussion of the steps taken O

prior to and during spent fuel storage to ensure that helium will be present at the end of an extended storage period, hence prior to transport.

Additionally, should the helium be lost by some unspecified means, thermal analyses have been performed with either helium or air in the cask cavity to ensure that even without helium, cask temperatures remain acceptable and O

cask margins of safety remain positive.

b.

Procedures have been revised throughout to identify pressures as absolute.

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A new Section 7.5.3 " Corrective Action" has been aJded to the chapter to describe the actions to be taken when it is determined that the inner lid containment o-ring requires replacement.

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.O Docket No. 71-9235 August 1993 f

OPERATING PROCEDURES NRC Comment 5.

Resise Section 7.3.2, " Preparation of Cask Unloading," Step 8 to include installation of the lifting eye bolts.

e NAC Response Section 7.3.2, Step 8 has been revised to include the installation and removal of 8

the lifting eyebolts at the appropriate steps in the procedure.

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Docket No. 71-9235 August 1993 O

ACCEPTANCE TESTS AND MAINTENANCE PROGRAM NRC Comment 1.

Add a section on ft.brication to Chapter 8 of the application. Include O

the procedures for pouring and fabricating the lead shielding. These l

procedures, temperatures, times, etc. (e.g. the information contained in Section 2.6.11.0) should be consistent with the values used in the structural analysis in Sections 2.6.11.1 through 2.6.11.3. Also include a specification of the code sections that will be used for performing, O

examining, and accepting the welds for qualifying the welders.

NAC Response O

This chapter has been revised to include Section 8.4 to describe the fabrication sequence for the NAC-STC, including a detailed description of lead pouring methods, equipment and procedures.

Section 8.1.1 has been revised to incorporate the applicable ASME Code O

requirements for the performance, examination and acceptance of welds, and for the qualification of welding personnel.

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.O Docket No. 71-9235 August 1993 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM g

NRC Comment 2.

Revise Section 8.1.1, " Visual Inspection", to add the statement that non-containment welds specially marked "PT root and final pass" will be 8

examined with PT per Section III, Article NB-5350, of the ASME Code as indicated on Drawing No. 423-802, Sheet 1.

NAC Response O

Section 8.1.1 " Visual Inspection" has been retitled " Weld Procedures, Examination, and Acceptance" and has been revised to identify the inspection requirements for non-containment welds marked "PT root and final pass" on the License Drawings (Section 1.3.2). The section has also been revised to identify ASME Section III, o

Subsection NB, Article NB-5350 as the acceptance criteria requirement for these welds.

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ACCEI"FANCE TESTS AND MAINTENANCE PROGRAM NRC Comment 3.

Revise Section 8.1.3, " Leak Tests," to be consistent with any changes in O

the leak test procedures made in Chapter 7.

NAC Response The leak test procedure section has been revised to incorporate changes in the O

acceptance criteria and test sensitivity. The section has been revised to incorporate a reference to the detailed leak testing procedures in Chapter 7.5.

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.O Docket No. 71-9235 August 1993 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM g

NRC Comment 4.

Revise Section 8.1.5, " Tests for Shielding Integrity," to include a fabrication acceptance test for the neutron shield similar to that 9

described in Section 8.1.5.1 for the gamma shield.

NAC Response A fabrication acceptance test has been added to Section 8.1.5.2 for the verification of the integrity of the radial neutron shield.

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Docket No. 71-9235 Augun 1993 O

ACCEPTANCE TESTS AND MAINTENANCE PROGRAM NRC Comment n

5.

Revise Section 8.1.5.3, " Neutron and Gamma Shield Effectiveness O

Tests," to describe and justify the procedure for correcting neutron dose rate measurements based on the decay heat of the loaded fuel.

l'AC Response O

.iection 8.1.5.3 has been revised to discuss the correction factors to be addressed in the field performance of shield tests on casks containing specific spent fuel assemblies which have lower source terms and cooling periods greater than the L

minimum for the design basis content condition.

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.O Docket No. 71-9235 August 1993 ACCEI'FANCE TESTS AND MAINTENANCE PROGRAM g

NRC Comment 6.

Revise Section 8.1.6, " Thermal Acceptance Tests," to include procedures to ensure that the aluminum basket is not over heated during the 9

thermal acceptance test.

NAC Response The main structural components of the NAC-STC basket are fabricated from stainless steel components, which will be less affected by the test heat load than the previously proposed all-aluminum basket. Also, the thermal power level for the test has been established as 20 + 0.0/-2.0 kilowatts. This limits the maximum heat load for the test to 80 to 90 percent of the full design basis decay power g

levels, but provides sufficient heat load to verify the performance of the heat dissipation components of the NAC-STC. The test power level provides a margin to ensure that the cask components are not thermally overloaded.

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Docket No. 71-9235 August 1993 L

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ACCEPTANCE TESTS AND MAINTENANCE PROGRAM NRC Comment 7.

Revise Section 8.2.2.2, " Containment Periodic Verification Leakage O

Test," to make it clear that the metallic o-rings in the containment seal must receive a periodic level leak test.

NAC Response O

Section 8.2.2.2 has been revised to clarify that the containment system periodic verification leakage test is required to be performed following replacement of containment boundary metallic o-ring seals.

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NRC Comment 8.

Revise Section 8.2.7, " Miscellaneous," to include a check of the cask cavity bottom to detect any crud build-up that could clog the drainage 9

paths.

NAC Response O

Section 8.2.7 has been revised to incorporate a visual inspection of the cavity prior to each fuel loading to observe for build-up of foreign matter in the cavity which could block the cavity drainage path. Also, the cask loading procedure in Section 7.1.2 includes a visual inspection requirement for the internal cavity, fuel basket and drain line. to be performed prior to fuel loading (Step 15).

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