ML20138D212
| ML20138D212 | |
| Person / Time | |
|---|---|
| Site: | 07109268 |
| Issue date: | 03/24/1997 |
| From: | Reid D NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Mcconaghy W SIERRA NUCLEAR, INC. |
| References | |
| TAC-L22232, NUDOCS 9705010064 | |
| Download: ML20138D212 (26) | |
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d UNITED STATES j j
! NUCLEAR REGULATORY COMMISSION
- WASHINGTON. D.C. 20555 0001 I %, March 24, 1997 l
l Mr. William J. McConaghy l Manager of Licensing Sierra Nuclear Corporation One Victor Square Scotts Valley CA 95066
SUBJECT:
RE0 VEST FOR ADDITIONAL INFORMATION REGARDING THE TRANSTOR SHIPPING CASK SYSTEM (TAC NO. L22232)
Dear Mr. McConaghy:
By application dated December 20. 1995, as supplemented January 12. 1996.
l Sierra Nuclear Corporation (SNC) requested that a certificate of compliance (C0C) be issued for the TranStor shipping cask in accordance with 10 CFR Part 71. By letter dated June 19. 1996, the staff requested additional information (RAI-1). On September 13. 1996. SNC responded to RAI-1 and revised the Safety Analysis Report (SAR) to incorporate the responses.
I As a result of the staff's review of the RAI-1 responses and revised SAR, we l have identified additional information that is needed for the staff to complete its review to ensure that you have demonstrated compliance with 10 CFR Part 71. This RAI is being provided so that you may address those areas ;
and is authorized in accordance with 10 CFR 71.39. Please provide a response to this letter within sixty (60) calendar days from the date of this letter.
SNC's response should address each item by the number identified in this RAI-2.
During the November 7,1996. meeting between staff and SNC. a number of structural issues were discussed, including impact limiter design analyses for )
lid bolt. lead slump. basket lifting, and neutron shield shell; impact limiter i attachment and attachment bolts: neutron shield compressible material: and dynamic load factors. This.RAI addresses these items and other technical matters.
Some topical areas do not have any RAI items at this time; however, it should not be interpreted that the staff will not have future questions in these areas. In particular, the staff does not reference a new RAI quest"on for the evaluation of coatings, which was asked in RAI-1 (No. 2-38). You r.hould supply this information at your earliest opportunity.
You are again reminded to apply the appropriate requirements of your quality l assurance and control program to any and all information provided to the NRC l to ensure accuracy and completeness of information. If your response to this l RAI contains proprietary material, it is incumbent upon you to identify that i material as proprietary pursuant to 10 CFR 2.790.
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9705010064 970324 PDR ADOCK 07109268 hf j C P DR.,; }
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Mr. William J. McConaghy The staff has not identified any proprietary information in this letter or its enclosure. However, to preclude any inadvertent release of proprietary information, this letter will not be placed in the public document room (PDR) for 30 days. If, within 20 days. Sierra Nuclear does not, identify to.the staff any proprietary .information.o this letter will- be.placed in the. PDR.
Please reference the above TAC No. in future correspondence related to this request.
The requirements of this letter affect nine or fewer respondents and, therefore, are not subject to Office of Management and Budget review under P. L.96-511.
If you have any questions or comments lease contact me at (301) 415-8556.
Sincel uy.
Ob Dennis . Reid. Project Manager Spent Fuel Licensing Section Spent Fuel Project Office -
Office of Nuclear Material S fety and Safeguards
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Docket No. 71-9268 s cc: Mr. Stephen M. Quennoz. Trojan Site Executive-Portland General Electric Compariy- -
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i REQUEST FOR ADDITIONAL INFORMATION This document titled Request for Additional Information (RAI), contains a compilation of additional information reauirements identified to date by the Nuclear Regulatory Commission staff, during its review of your application and Safety Analysis Report (SAR) for packaging. This RAI follows the same format as the applicant's SAR. This RAI is authorized pursuant to 10 CFR 71.39.
Each individual RAI describes information needed by the staff to complete its review of the application and/or the SAR and to determine whether the applicant has demonstrated compliance with the applicable regulatory requirements. Where an individual RAI relates to the applicant's failure to meet one or more regulatory recuirements or where an RAI specifically focuses on compliance issues associatec with one or more specific regulatory requirements (e.g., specific design criteria or accident conditions). such requirements are specified in the individual RAI.
This RAI is organized as follows:
Chapter 1.0 Introduction and General Information Chapter 2.0 Structural Chapter 3.0 Thermal Chapter 4.0 Containment Chapter 5.0 Shielding Chapter 6.0 Criticality Chapter 7.0 Operating Procedures Chapter 8.0 Acceptance Tests and Maintenance Program
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Chapter 1.0 Introduction and General Information The following regulatory requirements are applicable in this chapter: 10 CFR 71.0(a). 71.7(a). 71.33. 71.35(a). 71.37. 71.39. and Subparts E. F. and G. It should be noted that other regulatory requirements may be applicable to this section.
1-1 Revise the non-proprietary drawings to provide the material. component sizes and dimensions, tolerances, and detailed information with regard to the construction, including the standards and codes used.
4 The non-proprietary drawings provided in the SAR are not acceptable.
The SAR drawings define the design approved by the NRC. Upon approval.
. the packaging will be constructed in acccrdance with the SAR drawings.
and the drawings will be referenced in the NRC C0C. It should be noted that 10 CFR 71.12(c) states that the general license applies only to a
. licensee who has the drawings and other documents referenced in the approval relating to the use and maintenance of the packaging.
1-2 Provide a revision block on all drawings. For reference purposes, a i
revision block should be provided to indicate the type and date of any drawing revisions.
1-3 Revise Drawing CA-001:
(a) Specify the thickness of the weld deposit for each weld pass as
. identified in note 1.
The liquid penetrant examination (PT) can only detect surface flaws up to 1/8-inch depth. Thus, the maximum weld deposit should not be greater than 1/8 inch to ensure that no flaws are left undetected in the welds.
4 (b) Revise note 7 to delete alternate joint designs codes, and standards.
Alternate joint designs, related codes, and standards are subject to NRC review.
1-4 Revise Drawing TSP-001 to specify the maximum thickness of deposit metal for each weld pass as identified in note 2.
PT or magnetic 3 article examination (MT) of the root and final passes may not detect ]uried flaws in the welded joint. PT or MT examination can detect flaws up to 1/8-inch deep. Thus. PT or MT should be performed on each weld pass, and the thickness of deposit metal should not be greater than 1/8 inch for each pass.
1-5 Specify on an appropriate drawing the glue used to assemble the redwood pieces for the impact limiter.
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t 1-6 Revise note 2 on Drawing TSP-009. sheet 1, to specify the minimum areal density in units of g/cm2 1-7 (a) Revise note 12 on Drawing CA-001. sheet 1. to give the minimum boron content in absolute terms, rather than a percent of weight.
(b) Specify the minimum hydrogen density in units of gm/cm3 1-8 Revise the drawings and the SAR, as appropriate. to specify that it is not intended to permit mixing materials within the same basket component.
In response to RAI-1. question 1-8(c). SNC states ". . . it is not intended to permit mixing materials within the same basket subcomponent (i.e.. shell) the fabrication specifications will specify the material selection for the specific project." A C0C. when issued, controls the s)ecifications of fabrication and use by referencing the drawings and tie SAR. Therefore, the drawings (e.g., TSP-001) and the SAR should clearly reflect the intentions of the design.
Chapter 2.0 Structural The following regulatory requirements are applicable in this chapter: 10 CFR 71.7(a). 71.35(a) 71.39. 71.71, 71.73, and Subpart E. It should be noted that other regulatory requirements may be applicable to this section. l 2-1 RE: BNFL 1.10.06.01 Weight and Center of Gravity Calculation Revise tne application to present a step-by-step calculation of the weights and the location of center of gravity ifor one cask configuration.
The application only provided a table showing the results of the calculated weights and the location of center of gravity. No hand computation was presented. To facilitate a review, the application should present step-by-step calculations for at least one of the many cask configurations.
2-2 RE: BNFL 1.10.06.02 Impact Limiter Analysis (a) Justify the redwood properties in Tables 1 and 2.
The redwood crushing stress is not consistent with the strain (i.e.. higher crushing stress at lower strain), and this is contradictory to your Reference 5.2. Sandia Report SAND 76-0087.
(b) Perform benchmark testing of the redwood to verify the crushing 1 stresses shown in Tables 1 and 2. I The crushing stresses of the redwood are one of the most important parameters for the impact limiter Jerformance. Therefore, it is important for the application to slow that the postulated redwood 3
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i crushing stresses are reasonably accurate for the redwood to be used in the construction of the impact limiters.
(c) Justify that the modification factors for dynamic and temperature effects are applicable to crushing the redwood in a direction perpendicular to the wood grain.
The modification factors for dynamic and temperature effects are based on the test data presented in Reference 5-1. However, the test data presented in Reference-5-1 are for crushing redwood parallel to its wood grain. SNC should justify that these factors derived on the test data for crushing redwood parallel to the wood
, grain are also applicable for crushing the redwood perpendicular ,
to its wood grain. This is particularly importan for the end drops.
(d) Show the crushing stress and the strain of the redwood for the end drops.
This should include 1-foot and 30-foot drops for both hot and cold ambient temperatures. Based on Reference 5.2. Figure 3. and the ,
table of average crush strength, the redwood crush strength does !
4 not change significantly until the crushing strain is above 50%. ,
Thus, the big differences in g-loads for the 1-foot (i.e. 10.39 )
hot. 17.1g cold) and the 30-foot (i.e. 49g hot. 35.69 cold) end drops do not appear reasonable. ,
(e) Develop and provide additional procedures for the verification. I testing, and acceptance of redwood for the construction of the !
impact limiters. i Visual inspection, moisture content, and density will not be sufficient to ensure the redwood properties. Sample tests of each lot of redwood should be performed to verify the redwood properties. The number of sample tests required and the maximum deviations permitted for the sample test results should be provided.
(f) Justify that the properties of redwood used in the impact limiter will not deteriorate curing its service life.
Redwood is not an equilibrium structure, and it is subject to the effects of heat and the limited environment that it is expected to l experience in service. The properties of all materials are :
dependent. in part on their structure. Therefore. the properties of redwood will vary in accordance with any alterations in its structure that bring it closer to the equilibrium state. For packages to survive the conditions specified in 10 CFR 71.71.
" Normal conditions of transport.~ and 10 CFR 71.73 " Hypothetical accident conditions." the properties are expected to be maintained throughout the service life.
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l Perform impact limiter crush tests to validate the force-i (g) deflection relationships shown in the SAR (Figures 2.3.2 thru 2c3.5).
The 30-foot drop tests can only be performed for a few representative drop orientations. Impact analyses are performed for all drop orientations considered in the desirt The force-deflection relationships of the impact limiter define the energy absorbing characteristics of the impact limiter and provide the basis for all im)act analyses. In addition to redwood properties.
the energy absoraing capability of the impact limiter depends on the construction of the impact limiter. Thus it is necessary to perform crush tests to validate the force and deflection relationships that are derived from assumed impact limiter crushing characteristics.
(h) Perform drop tests to substantiate representative impact analysis results (e.g.. computer anal.) sis cases A through T).
SNC should show that the impact limiters will perform as designed and will remain attached to the cask to protect the lid closure seal under the fire test condition. The performance of the impact limiters under the 30-foot impact condition should be validated by testing.
(i) Provide g-load computations for side-drop cases C. F. M. and T.
based on half of t1e package weight.
l This is necessary because there are two impact limiters acting l simultaneously in a side-drop condition.
l (j) Provide g-load computations, including results. for shallow angle drop (e.g. 15 to 20 degrees from the horizontal surface) to find the maximum secondary impact.
2-3 RE: BNFL 1.10.06.04 Lid Bolt Analysis Provide an evaluation of the washer (diameter and thickness) to be placed under the closure bolts.
Because of the large differences in strength between the bolt material and the forging material it appears that a larger and thicker washer i
! should be placed under the bolt head to distribute the load. t 2-4 RE: BNFL 1.10.06.09 Lead Slump Analysis Revise the lead slump analysis based on inelastic properties of the lead.
l The stress in the ' lead shielding exceeds the yielding strength in an end-drop event. Thus, inelastic lead properties should be used to j compute the lead slumps due to a 30-foot end-drop. In addition. the 5
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structural analysis needs to show that this shielding will stay in place and function as necessary under the normel and hypothetical accident
! conditions.
2-5 RE: BNFL 1.10.06.10 TranStor Basket Lifting Devices (a) Revise the calculations to provide an independent basket lifting analysis.
There are sufficient differences between the TranStor and the VSC-24 designs to warrant an inde)endent analysis. The finite element model shown in Figure 1 s1ould release the moment resistance at the perimeter of the lid because only a single 3/4-inch weld is used to connect the lid to the basket shell.
(b) Specify on the drawing and in the operating procedures that the height between the basket and the lifting point is not less than 11 feet. Also, the minimum thread engagement length of the lifting lug should be at least 1.95 inches.
The analysis is based on a maximum lifting angle of 12 degrees with the vertical that requires a minimum height of 11 feet between the basket and the lifting point. The minimum thread engagement length is required by the design code. I 2-6 RE: BNFL 1.10.06.24 Neutron Shield Shell Analysis (a) Provide an analysis to show that the heat dissipating fins will not be separated from the neutron shield shell due to material expansion.
(b) Show the effects of a gap between the shield material and the shell on the heat dissipating capability of the cask.
Paragraph 6.3 of the calculation package states that "After the first cask heat-up and initial yielding of the neutron shield shell, the neutron shield will simply separate from the shell during cool-down and further heat-up/ cool-down cycles will cause i expansion and contraction within the gap without loading the !
shell." The requested analysis is needed to ensure that the !
neutron shield will work as stated above and the heat dissipating !
capability of the c6sk is acceptable. I 2-7 RE: BNFL 1.10.06.30 Impact Limiter Attachment Bolts ,
(a) Revise the attachment block welds to meet the requirements of ASME Code. Sectior. III, Subsection NF-3324.5. and Table NF-3324.5(a)-1.
The attachment block is welded to the cask body by fillet welds.
Based on Table NF-3324.5(a)-1, the stress limits of the weld i should be 0.3 times the minimum tensile strength of the weld I
metal, except. shear stress on base metal will not exceed 0.4 6
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times the yield stress of the base metal. The attachment block should be designed to have the capability to develop the tensile strength of the attachment bolt (b) Show that the torque applied to the attachment bolt is sufficiently large to prevent loosening of the attachment bolt during transit.
- The bolt torque and the clamping force should be large enough to prevent vibratory motions of the impact 1.imiter during transport so that the bolt will not be loosened over time.
2-8 RE: BNFL 1.10.06.34 TranStor PWR Basket 30-Foot Drop Analysis and BNFL 1.10.06.39 TranStor PWR 3asket One-Foot Drop Analysis (a) Revise the basket analysis for the oblicue-drop condition to show that the combined stresses (vertical anc horizontal components) are less than the end-drop or side-drop condition.
The application stated that the 30-foot oblique-drop condition has been determined to produce lower deceleration than that calculated for the vertical or horizontal 30-foot drops. However, the combined stresses of an oblique drop may exceed the stresses produced by either vertical or horizontal drop.
(b) Revise the basket analysis to show that a shallow angle oblique dro) (slap down near horizontal orientation) will not generate higler stresses in the basket structure.
The evaluated sla)-down condition was near a vertical drop orientation, whic1 is not as severe as a slap down near a '
horizontal drop orientation.
(c) Show the initial ga) sizes assumed for the ga) elements at the interface between t1e mainframe-to-shell in t1e finite element model.
(d) Provide justification for using both a gap element and a spring element at each nodal point of the basket shell interfacing with the cask inner shell.
The use of low stiffness spring elements in conjunction with the gap elements is confusing. The ga) element is a spring that transmits compressive force when t7e gap is closed.
(e) Provide justification for the use of A516. Gr 70 material for basket cells.
The maximum tem)erature for this material per ASME Code Section Ill is 700 F, w1ich is lower than the maximum cell temperature shown in the SAR.
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(f) Provide justification for the stiffness of the gap elements used in the finite element models.
A gap element and a spring eleaient are used at each basket shell l nodal point interface with the cask inner shell. An arbitrary i value of stiffness (9 x 10 lb/in ) 2is used for the gap element.
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and a very low spring stiffness K is used for the spring element.
It seems that the gap element stiffness should be determined by
, the stiffness of the basket shell, and the ground node of the gap I l element should be fixed to simulate the support of the cask shell. l It is unnecessary to use both gap and spring elements at the same location (see item 2-8(d) above).
, (g) (1) Provide a discussion or, i,vw loads are supported and l transmitted throughout the model.
(2) Show the thermal stresses generated for normal and accident conditions for the 2-D model.
The corner cell and the frame are not connected to the basket shell and there is no discussion about the gaps that may exist between the cells.
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i (h) Provide an evaluation of the welds in the basket shell.
(i) Provide a sketch showing the finite element model of the basket and the gap and spring elements.
(j) Provide a table showing the displacements of the basket shell for the drop. orientations considered.
SNC should verify. and so state, that the reaction forces are within reasonable limits and are also in equilibrium with the applied load. Reaction forces at the gap elements should be shown on the table. :
(k) Show a sketch of the basket, indicating the locations and the magnitude of maximum stresses for each drop orientation analyzed.
2-9 RE: BNFL 1.10.06.40 TranStor Shipping Cask ANSYS Structural Analysis j (a) Describe the forces in the gap elements at the end of the cask and the crushed geometry assumed for the various off-center oblique drops. j Force distribution in the axial direction should not be uniform for the oblique-drop orientation. j (b) Justify that stresses for the 1-foot drop conditions are obtained l by taking the ratio for the 30-f.oot drop conditions for impact loads.
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It appears that the impact limiter crushed area may be different l for t1e 1-foot and 30-foot drops. This may have some effect on '
the distribution of the reaction forces at the gap elements.
Since the impact limiter crushed area changes are not linear, justifications should be provided if stresses are obtained by l direct ratio.
(c) Provide a discussion of how the stresses for buckling analysis
- were obtained and show that the stresses obtained for buckling are consistent with the inner shell stresses obtained in 30-foot drops.
i l It is not clear how the stress outputs were obtained for inner l shell buckling evaluation.
l 2-10 RE: BNFL 1.10.06.43 MPB-BWR Structural Analysis 30-Foot Side-Drop l Accident and BNFL 1.10.06.44 MPB-BWR Structural Analysis 1-Foot l Drop (a) Provide the mechanical properties of A500 Gr C.
l (b) Justify the use of poor aspect ratios for some elements (i.e., top
! and cottom plates) in the multi-purpose basket (MPB) finite j element model.
l (c) (1) Provide a sketch of the shell (circumferential and axial cross sections) which shows that the calculated stresses are within the allowable stress ranges.
Sufficient sections should be taken to show the stresses at different locations along the axial and circumferential directions.
(2) Verify that the reaction forces are reasonable and in equilibrium with the applied loads.
(d) (1) Provide a discussion of how loads are transferred and the most critical loadings for these components in various ,
loading conditions. '
- I (2) Provide a discussion of how the inserts and cells are put into the MPB shell. including the required gaps and clearances.
l (3) Show that the 2-D models of the tube and the inserts are in l equilibrium with the applied loads and the reaction forces.
NOTE: The inserts and the cells for the BWR MPB are not connected.
(e) (1) Provide a sketch showing the components evaluated for buckling for the fuel cells and support inserts.
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t (2) Show the derivation of loads applied to the component for l buckling evaluations (e.g. , P. M. and lateral loads if applicable).
(3) Justify that the most critical case is considered.
2-11 RE: BNFL 1.10.06.45 MPB-BWR Structural Analysis-Pressure Condition l Describe the boundary conditions for the finite element model of. the MPB. indicating what constraints were used at the perimeter of the top and bottom plates that join the shell. The stresses at sections away from the ends (e.g. , mid-length of the shell) should be checked with hand calculations to verify the computer analysis results. Moment constraint should be released for small fillet or partial penetration
! welds.
2-12 RE: BNFL 1.10.06.46 MPB-BWR Structural Analysis 30-Foot End-Drop (a) Describe how loads are applied to the model. .
(b) Show and verify the analysis results (e.g., applied load equals to resultant stresses).
2-13 RE: BNFL 1.10.06.51 TranStor Shipping Cask Cradle Clarify whether or not the cradle is an integral part of the package.
If so. the following information must be provided for NRC review:
(a) Show the correlation between the design computation and the figures.
(b) Provide a sketch showing how the cask and cradle are lifted together.
(c) Show the design is based on combined 2. 5 and 10g force applied at the centroid of the package.
(d) Show that the shearing capacity of the bolts is based on the threaded part of the bolt and the bearing stress will not exceed the maximum allowable bearing stress (i .e. . max. F, = 1.35 F,) .
Shipping cradles are normally not reviewed by NRC. Because shi) ping cradles are generally not considered as a part of the paccage. there are no regulatory requirements specified under 10 CFR Part 71. However, if the cradle is designed to perform a l
safety function, in conjunction with the package, then it must be considered as a part of the package and reviewed in accordance with the requirements of 10 CFR Part 71.
2-14 RE: BNFL 1.10.06.52 Inner Shell Buckling Analysis Show that the buckling analysis is based on the maximum stresses
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obtained in the cask design.
The stresses shown in the stress table on sheet 6 do not agree with the f cask analysis (BNFL 1.10.06.40) The oblique drop may result in higher ,
stresses in the shell (e.g., BNFL 1.10.06.40. Table C-79).
l 2-15 RE: BNFL 1.10.06.54 Neutron Shield Expansion (a) Provide the material properties of the compressive material at the j end plate, including specifying the material on the drawing.
l (b) Justify that the thermal transfer fin welds will not fail as a result of differential thermal expansion.
2-16 RE: BNFL 1.10.06.57 TranStor foke Brittle Fracture (a) Provide the design calculations of TranStor shipping cask lifting yokes.
(b) Show that the yoke material meets the nil-ductility-transfer-temperature (NDTT) requirement in Regulatory Guide 7-11.
Brittle fracture criteria in Regulatory Guide 7-11 are based on material NDTT. not charpy V-notch impact tests.
2-17 RE: BNFL 1.10.06.60 Dynamic Load Factor For TranStor Cask Drop (a) Demonstrate that the stress intensities throughout the cask are within the allowable stress intensities based on elastic analysis.
(b) Show that the stress intensities throughout the cask are under the most critical loading condition. including dynamic load factors for impact loads.
The approach presented in this section does not appear to be acceptable.
l The ASME stress intensity criterion is based on elastic analysis. To
- decermine that the design meets the ASME Code. the analysis must include l the requirement stated above.
2-18 RE: BNFL 1.10.06.63 Evaluation of Slapdown Effect (a) Provide an analysis to show the most critical drop orientation for the slap-down impact condition.
l (b) Revise the application to show that the slap-down impact energy is i absorbed by the impact limiter at the slapped down end of the cask.
l It is not reasonable to assume that the slap-down impact energy is uniformly distributed and thus, absorbed along the length of the cask. The slap-down impact energy is absorbed by crushing the impact limiter at the slapped.down end of cask.
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2-19 RE:
BNFL 1.10.06.65 Tiedown Reactions Show that the design tie-down forces are based on combined 10g ,
longitudinal. 2g vertical and Sg lateral forces applied at the center I of gravity of the package. j 2-20 RE: TranStor Failed Fuel Cans and Fuel Debris Cans Show the structural evaluations performed on the Failed Fuel Cans and Fuel Debris Cans. l 2-21 RE: Greater Than Class C (GTCC) Waste Basket Provide a structural evaluation of the GTCC basket including how the additional shielding is held in place under normal and hypothetical accident conditions. ,
The structural analysis needs to show that this shielding will stay in i place and function as necessary under the normal and hypothetical 1 accident conditions.
Chapter 3.0 Thermal The following regulci,ary requirements are applicable in this chapter: 10 CFR l 71.7(a). 71.35(a), il.39, 71.43(g) 71.71(a). 71.71(b) 71.71(c). 71.71(c)1. '
71.73(a), and 71.~3(c)(4). It should be noted that other regulatory requirements moy be applicable to this section.
3-1 Evaluate the concainment pressure analysis for accident conditions with 100% rod rupture. assuming a 30% release fraction for all significant fission gases (sr.able and radioactive) and 100% of the rod fill gas.
SNC should use ORIGEN run data or ORCWM data for determining fission gas inventory.
BNFL 1.10.06.07 used ~0.25 atoms / fission" to estimate the amount of fission gas produced based on Olander's " Fundamental Aspects of Nuclear Reactor Fuel Elements." 1976 report. The use of this number in the l applicant's equation is not appropriate since it underestimates total I fission gas produced. Page 202 of Olander's text identifies 0.25 atoms / fission based on total fission vield of stable xenon and kry) ton.
The number does not appear to account for neutron reactions with s1 ort-lived isotopes.
Short-lived isotopes are discounted and would be appropriate for total yield: however, not for total gas produced from reactor operations, especially 236 Xe (stable). Although production of 23*Xe from fission is accounted for, production from other mechanisms does not appear to be l included. The interaction of 35Xe + n 236Xe (stable)36has not been I accounted for. If not included in total yield, total Xe inventory l must also include the chain decay of Te (elemental apor) into 35 Xe 35 during reactor operations and subsequent neutron inter m I ton. Other l 12
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( fission and decay gases are not accounted for such as iodine, bromine.
tritium. helium, etc.
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3-2 (a) Evaluate the decomposition of vrotective coatings for the carbon steel basket when the cask is exposed to fire conditions.
l (b) Assess impact on gas temperature and containment pressure.
Predicted basket temperatures (BNFL 1.10.06.21. sheet 94) for accident l conditions exceed the non-continuous rating (800*F) for Carbo ZincZinc i as specified on the manufacturer data sheet. The partial pressure '
contribution (if vaporized) to overall containment pressure should be l included. The staff understands that alternative protective coatings
! ar being evaluated. The behavior of these materials under accident conditions should also be examined.
l l 3-3 Provide figures in the SAR for carbon steel basket temperature
- distribution similar to that of SAR Figures 3.4-10 and 3.4-11.
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3-4 Calculate containment pressure assuming shipment of fuel assemblies with control rods.
SNC should assume rupture of 100% of tae control rods and release of all internal rod gases. The thermal section did not prevent shipment of control rods with fuel and was not accounted for in BNFL 01.10.06.07.
Control rods would reduce the free volume inside the cask and also i increase the total gas' released to the cavity during an accident for cladded control rods. Control rod backfill gases should be included if
, used during manufacturing and any helium produced from B 4C/ neutron
! interaction in reactor operations.
3-5 Identify lead purity requirements in the SAR.
The fire analysis indicates very little margin to lead service limit.
593 F versus a limit of 600 F specified in the SAR.
l Chapter 4.0 Containment l
The following regulatory requirements are applicable in this chapter: 10 CFR 71.7(a). 71.35(a). 71.39. 71.43(c). 71.43(e). 71.43(f). 71.51. 71.61. 71.63.
i 71.71(a). 71.73(a). and Appendix A. It should be noted that other regulatory l
requirements may be applicable to this section.
l l 4-1 Revise the SAR and associated calculations to reflect that only 3% of
! the spent fuel tubes are postulated to fail under normal conditions of l transport. in lieu of the 100%.
4-2 RE: BNFL 01.10.06.05. Revision 2. and BNFL 01.10.06.06. Revision 2 Use the unchoked flow correlations of ANSI 14.5-1987 in lieu of the choked flow correlations.
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i The basis for this practice is NUREG/CR-5403 " Predicting the Pressure Driven Flow of Gases Through Micro-Casillaries and Micro-Orifices" which demonstrates to the satisfaction of t7e staff that the unchoked flow correlations better approximate the true measured flowrate for the leakage rates associated with transportation packages. The choked flow correlations tend to overestimate the leakage rate or underestimate the hole diameter (see also NUREG/CR-6487 " Containment Analysis for Type B Packages Used to Transport Various Contents").
l 4-3 Evaluate the requirements of 10 CFR 71.51(a)(2) separately for the I hypothetical accident condition.
For the hypothetical accident condition, the referenced air leakage rate <
should be determined separately for "no escape of krypton-85 exceeding I 10 A, in 1 week" and "no escape of other radioactive material exceeding a total amount A2 in 1 week." The calculations should separately analyze
- each of these requirements.
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! 4-4 Re-evaluate the pressure-drop test to ensure that it meets the maximum I
assembly verification test sensitivity of 1 x 10 3 std cm3/sec.
The acceptable maximum assembly verification test sensitivity is 1 x 10 3 std cm3/sec. Values larger than this were inappropriately used l to determine the parameters of the pressure drop test.
4-5 Demonstrate that the assumption of using the 8-year cooled fuel with a burnup of 45.000 mwd /MTU is bounding for the containment analysis.
Specifically, additional calculations should be submitted that analyze the controlling leakage rate (i.e.. BWR. NCT) for the 5-year cooled fuel with 30.000 mwd /MTU.
l 4-6 Address the following comments to improve the overall quality of, and consistency between. the containment calculations and Chapter 4.0 of the SAR.
(a) The format of ANSI 14.5 should be followed. For PWR and BWR fuels and for normal conditions of transport and hypothetical accident conditions, the following items need to be clearly identified, including the details of their derivation: A, values, release rates (Rs . R4), source terms (Cy. C4 ), leakage ' rates (Lu. L3 ),
reference air leakage rates (L g u. Lt ,) . maximum allowable leakage rate for the package. fabrication verification test sensitivity l for the package, and assembly verification test sensitivity. A l
tabular format of these results would aid in the review.
(b) The calculations and SAR should utilize units consistent with ANSI Code Section 14.5 (e.g.. leakage should be in cm3/sec). Even though the calculations present metric unit conversions, the SAR is written primarily using English units with no conversion. For clarity, the SAR should utilize metric units as does the standard for leakage tests (i.e. ANSI Code Section 14.5).
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(c) The calculation details in addition to the narrative description should be provided. For example. the containment calculations do not provide the details of converting the gas leakage rates into standard air leakage rates, nor do they provide the details of determining the leak test conditions.
(d) The introduction section of the calculations mis-state the permissible release rates for Krypton 85 for accident conditions.
Apparently, applicant used 1995 edition in lieu of the revised (April 1,1996) edition of 10 CFR Part 71.
(e) The containment calculations reference the 1995 edition of 10 CFR
. Part 71. This is inappropriate since the 1996 edition (effective A)ril 1.1996) is currently applicable to this application. Also, i t1e subject containment calculations were both revised twice 4
subsequent to the 10 CFR Part 71 effective date and should have
, used the latest edition.
I NOTE: If true, the staff reminds SNC that this was a concern in i RAI-1 and cautions SNC to correct its material.
(f) Use of the term "off-normal' is not ap)ropriate for transportation packages. For example Calculation Taale 2.1 and SAR Section 4.3
- use the term off-normal. Only normal and accident conditions apply.
2 (g) SAR Section 6.4 of the PWR and BWR containment calculations list the maximum bulk average temperature of the helium as 750 F and 714 F, respectively. However. Table 4.2-2 of the SAR reverses the aforementioned values and lists the design basis temperature of
(h) SAR Section 4.1.3.1.2 references SAR Section 7.2.7 for leak testing acceptance criteria. However, no such criteria are provided in that section. Provide the correct cross reference.
(i) SAR Reference 2.2 is for the 1992 Edition of the ASME Section III.
SAR Reference 4.2 is for the 1994 Edition of the ASME Section III.
SAR Reference 8.2 just references the ASME Code without any mention to section edition, or addenda. The code for construction of these transportation packages should be clearly identified and should be the same edition and addenda.
(j) SAR Page 1-17 is not properly continued on Page 1-18. Some text appears to be missing. Provide this information as appropriate.
4-7 Clarify the following discrepancy between the desi~gn drawings and the SAR, including an explanation of why the subject drawing was recently changed to present information different from the SAR:
Drawing CA-001. Revision 1. note 25, indicates that the cask containment boundaries are to be leak tested using atmospheric 15
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air, evacuating the cavity to 0.15 psia with a leak rate not to exceed 2.5 X 10~8 in3/sec. This proposed testing does not agree with the helium leak testing described in SAR Section 8.1.3 that involves pressuring the cask cavity with helium to 20 psig and measuring for leakage not to exceed 6.4 X 10 4 in2/sec.
l 4-8 Clarify the following discrepancy between the design drawings and the SAR:
Drawing CA-001. Revision 1. note 24, indicates that the cask l containment is to be hydrostatically tested to a minimum of 62.5 psig. This test pressure does not agree with SAR Section 8.1.2.2 that states that the hydrostatic test pressure should be 75 psig.
l l 4-9 Correct the following discrepancies between the design drawings and the ;
calculations regarding 0-ring material specifications:
Drawing CA-005 dated September 13. 1996, indicates different specifications for items 3 and 4 (lid 0-rings) than are indicated l in SAR Section 6.1 of BNFL 1.10.06.17 dated August 26, 1996.
Similarly. Drawing CA-011 dated September 13. 1996, indicates different specifications for items '4 and 5 (port 0-rings) than are indicated in SAR Section 6.1 of BNFL 1.10.06.17 dated August 26, i 1996.
4-10 Revise the drawing as noted in your response:
l The response to RAI-1. question 4-7(b), states that the PT requirement for the seal weld is shown on Drawing CA-012.
However, no such requirement could be identified on the drawing.
1 i 4-11 Revise BNFL 1.10.06.05 and BNFL 1.10.06.06 as noted in your response:
! The response to RAI-1. question 4-9. states that BNFL 1.10.06.05 and BNFL 1.10.06.06 were revised to include laminar flow '
methodology. However, contrary to this statement. the choked flow l
l l correlations are still used. Revise the associated calculations l accordingly. Also, refer to the above containment Comment No. 4-2 for basis of omitting choked flow.
l 4-12 Revise the containment section of the SAR to describe your proposed j methods for shipping damaged and failed fuel, specifically addressing how the requirements of 10 CFR 71.63 are being met.
TranStor drawing Series TBD. TBF. TPD, and TPF show arrangements for failed fuel cans and fuel debris cans. However no explanation is provided in the SAR containment section that addresses the packaging requirements of 10 CFR 71.63. "Special requirements for plutonium shipments." SNC should demonstrate that the aforementioned configurations meet the containment requirements of 10 CFR 71.63.
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Chapter 5.0 Shielding The following regulatory requirements are a plicable in this chapter: 10 CFR 71.7(a) 71.35(a) 71.39. 71.41. 71.43(f). 1.47, 71.51(a). 71.71(a). 71.73(a) and Sub] art G. It should be noted that other regulatory requirements may be applica]le to this section. ,
5-1 RE: The following question relates to SNC's response to RAI-1 question 5-1:
Show that the personnel barrier will survive the normal conditions of transport in 10 CFR 71.71.
The response to RAI-1 states that the personnel barrier is an integral part of the package. All parts of the package that perform a function important to safety, such as constituting the package surface fnr external dose rate determinations must remain effective under all normal conditions specified in 10 CFR 71.71.
l 5-2 Specify whether the curves between the points in Figures 5.0-1 and 5.0-l
- 2. as noted in the SAR Revision 1. dated September 13. 1996, are to be l determined by linear interpolation or some other method.
5-3 Remove the statement in the last paragraph on page 5-1 regarding I
assemblies with an initial enrichment below the minimum values presented may be shipped if a specific analysis shows their source terms are bounded by the values.
Contents with parameters outside the limits used in the SAR evaluations, and subsequently specified in the C0C. must be ap] roved by the NRC prior to shipment. The fuel parameter limits used in t1e analysis will be :
incorporated into any associated C0C.
l 5-4 RE: SAR Section 5.4.1.1 (a) Describe the method used to account for the multiplication of source neutrons by the fissile isotopes in the fuel region.
(b) Specify and justify the amount of fissile material used.
l- The multiplication of source neutrons in the fuel region will increase the dose due to secondary gammas, as well as the neutron dose itself.
5-5 RE: SAR Section 5.5.2 (a) Explain the basis for the statement in the first paragraph on page 5-69 that the gamma and neutron sources from the fuel material do not differ significantly between stainless steel clad and zircaloy
- clad fuels.
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(b) Discuss the differences in activation rates between the two cladded fuels. ;
5-6 RE: SAR Section 5.5.4 If any contents other than activated steels are to be shipped as GTCC j waste, provide an analysis for each material matrix.
The conclusions in this section of the evaluation appear to apply to activated steels only. Any other conclusions should be based on physical facts and quantitative information.
5-7 Provide the basis for the statement at the top of page 5-73 that ,
directly above a concentrated clump of waste the dose rates are nearly I the same as they would be if the entire cavity were filled with solid activated metal.
The sentence after the statement cited, characterizes it as an assumption.
5-8 o ;fy how to determine the content limit for radioactive isotopes with )
L . iple gamma-ray emissions at various energies and yields per disintegration.
5-9 Provide calculations, and the basis for all assumptions. to justify the limits for GTCC waste containing neutron emitters. :
1 The SAR established neutron source limits by comparison with the I calculations performed for a content of spent fuel. The rationale presented does not address the special assumptions in the spent fuel analysis that may not apply to GTCC waste, such as: (1) the self ,
shielding effects of the fuel: (2) the assumed source distribution )
(i.e., the neutron emitters are concentrated away from the edge of the '
neutron shield): and (3) not assuming any shifting or concentration of the source volume during the hypothetical accident.
5-10 RE: Proprietary supplement BNFL 1.10.06.50 and SAR Section 3.1.4 Provide the basis for the total weight of cobalt initially present in the fuel region that was used for the activation calculations for PWR and BWR assemblies.
5-11 RE: Proprietary supplement BNFL 1.10.06.50 and SAR Section 3.2.1 Provide a quantitative justification for the modeling assumption of L smearing the mass of the eight 0.5-inch thick support plates around the
^ inside wall of the package cavity.
The SAR states that this approximation is not expected to reduce cask exterior dose rates. but does not provide quantitative support.
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l l Chapter 6.0 Criticality l
l The following regulatory requirements are applicable to this chapter: 10 CFR 71.7(a). 71.35(a) 71.35(b). 71.39, 71.55. 71.59. 71.71(a). and 71.73(a). It should be noted that other regulatory requirements may be applicable to this section. l l
L 6-1 Provide a copy of your responses to RAI-1. questions 1-18 thru l-26 and 1 l 6-1. for docketing under 71-9268.
l The criticality-related items of the RAI-1 submitted to SNC on ,
l December 17. 1996. regarding the 10 CFR Part 72 TranStor storage review, i l are equally relevant to the 10 CFR Part 71 TranStor transportation i l review. i l' 6-2 Remove the references to Fissile Class to be consistent with the current revision of 10 CFR Part 71 (effective April 1.1996).
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Fissile Class I is still mentioned in the SAR (e.g.. Sections 6.0 and ;
6.4.1). Part 71 no longer uses fissile class designations. SNC was !
reminded of this in RAI-1. l 6-3 Remove or revise the reference to " flux traps" in the second paragraph of Section 6.1.1.
l The calculated results show a significant reduction of reactivity when the four center sleeves are vacant. While it is clear that this effect I may be partly attributable to a flux trapping effect the present !
l wording is not clear in distinguishing between the designed basket !
- components called " flux traps" and the flux trapping effects that arise in absorber-plated sleeves when they contain only water. Furthermore.
any statements attributing a calculated result to a presumed aredominant 1
! phenomenon, like a flux traaping effect, should be supported ]y specific analysis. The statements s1ould either be omitted or clearly characterized as speculative. i 6-4 Explain why no channel is modeled for the GE 10x10 AC fuel assembly and :
, state the eff ct of this modeling approximation on the computed l k-effective for a loading of such assemblies.
6-5 Verify that the assumed fuel density (95% of theoretical) is bounding for all BWR fuel assembly types analyzed.
For PWR fuel. Table 6.2-1 shows that the as-modeled fuel densities are always greater than or equal to the densities calculated from the specifications for respective assembly types. Table 6.2-2 does not
- provide similar information for BWR fuel.
6-6 Justify the selection of GE 8x8 R over GE 9x9 ANF as the most reactive
. BWR assembly type for use in the design basis calculations.
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Tables 6.4-7 and 6.4-8 show calculated k-effectives for GE 9x9 ANF fuel
! that are greater than those for equally enriched GE 8x8 R fuel by more l than twice the Monte Carlo standard error, sigma.
l 6-7 Perform the required array analyses for closest packings of undamaged i and damaged casks with no water between the casks.
Section 71.59(a) requires array analyses for: (1) undamaged packages I with nothing between the packages, and (2) damaged packages with optimum !
interspersed hydrogenous moderation. ,
i Results presented in the SAR for the calculated cuboidal array (with lateral pitch fixed at 280 cm: basis not stated) indicate that neutronic !
interaction between undamaged or damaged casks is not enhanced by the l l neutron shield or by water between the casks. From this, it follows that the most reactive stacking arrangement is that of a tightly packed
, array without neutron shield and without interstitial water. Therefore.
the analyses should determine or estimate the bounding k-effective for a packed hexagonal (i.e.. triangular-pitched) array of touching casks. It I is noted that each cylindrical package in a three-dimensional hexagonal infinite array is in outer-shell-to-outer-shell contact with six lateral ;
neighbors one or three to) neighbors, and one or three bottom neighbors i
(i.e., coaxial contact wit 1 one neighbor on a flat top or bottom package surface, as is usually the case. or offset contact with three neighbors ;
l on a hemispherical top or bottom package surface). One method for using l KENO-Va to approximate close hexagonal packing of flat-ended cylindrical I packages entails modeling a cuboidal array with the areal packing factor !
artificially enhanced by reducing the lateral pitch and package outer radius by 7% (i.e. 1 - SORT [cos(30)]) while increasing the outer-shell material density to preserve mass.
To indicate the degree of cask interaction and to 3rovide a partial check on the adequacy of Monte Carlo sampling of tie cask interactions, the array analyses should be compared against corresponding calculations of a single cask without neutron shield, external moderation, or I
reflective boundary conditions. It is noted that, where necessary, a further check on Monte Carlo sampling adequacy in arrays of partially l shielded units can be provided by comparing the results of Monte Carlo and deterministic discrete-ordinates calculations on simplified models of single casks and cask arrays.
Note: The statement in Section 6.4.3 that "... the effects of varying cask pitch can be modeled by varying the cask external moderator density" is incorrect. Varying the external moderator density cannot adequately reproduce the effects of decreasing the cask pitch to represent a close-packed array. While the reactivity effects of l
neutronic interactions between spent fuel casks are usually small, they
! must be evaluated correctly as indicated above.
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6-8 Correct the statements in the SAR concerning the requirements of 10 CFR 71.55 and 71.59.
SAR Section 6.4.1 incorrectly state that 10 CFR 71.55 requires the analysis of infinite arrays. 10 CFR 71.55 concerns the analysis of individual packages, whereas analysis requirements for arrays of 4
packages are covered by 10 CFR 71.59. Infinite arrays must be analyzed to determine if the transport index is zero.
, 6-9 Confirm that the input files for all CSAS25 calculations specify
, different mixture numbers for water inside and outside the fuel pin cladding.
i A long-standing SCALE-CSAS coding error was recently discovered in the '
, calculation of Dancoff factors that are used in preprocessing problem-1 . specific resonance self-shiePiing information in the cross-section 4 libraries used by KENO-Va. This processing error. which can significantly affect the computed k-effective, arises only when the moderator inside and outside the fuel pin cladding is specified by the 4
same input mixture number. In checking a small sample of the input files provided in the latest submittal, the NRC reviewer has found no cases where the same mixture number was used for water inside and outside the fuel pin cladding. The applicant should verify that this is true for all calculations used in the criticality evaluation. I 6-10 Clarify the identification of the calculational method used and validated.
References to the calculational method in Sections 6.4.1 and 6.5 should identify the SCALE version and release, the CSAS sequence, and the cross-section library (e.g.. "The critical benchmark experiments were calculated by KENO-Va, using SCALE-4.1 PC with the CSAS25 analysis sequence and the 27-group cross-section library, 27GROUPNDF4.") The SAR should clearly indicate that the method used in the package calculations is consistent and is identical to the method used consistently in the benchmarking calculations. Any inconsistencies in the calculational method applied to packages and benchmarks should either be eliminated or stated and justified as inconsequential (see also item 6-11 below).
Additionally the discussion in Section 6.5 should be revised to reflect the fact that the validation and bias analysis apply, not to the KENO-Va code per se. but rather to the more specific calculational method. l namely KENO-Va used with SCALE-4.1, CSAS25. and 27GROUPNDF4. l 6-11 Resolve or justify the apparent inconsistency of using SCALE-4.3 for MOX fuel loadings. while using SCALE-4.1 for UO 2 fuel loadings and for all benchmark analyses.
The cask analyses for M0X fuel described in BNFL 1.10.06.66. Revision 0, uses SCALE-4.3/CSAS25/27GROUPNDF4, while the cask analyses for UO 2 fuel 21
and for all experimental benchmarks a)parently uses SCALE-4.1/CSAS25/27GROUPNDF4 (i .e. . t1e exact method is not clearly identified in all instances: see item 6-10 above). The_ validation and specific bias determination developed for one method are generally not applicable to other methods. The apparent inconsistency of using SCALE-4.1 versus SCALE-4.3 may be resolved by rerunning sets of l calculations using a single consistent method or justified by clearly
, demonstrating that either: (a) the two methods are essentially L identical (i.e., they use essentially identical codes and data and give the same k-effective values over an ap3ropriate range of numerical and experimental benchmark problems): or (3) their results are used in a way that is rigorously shown to preserve the conservatism of assumed biases and bias uncertainties for all cases.
6-12 Provide a discussion of intact fuel criteria.
The SAR should include a discussion of intact fuel criteria that l
addresses such considerations as the following: An intact spent fuel assembly is one that does not have gross structural or cladding defects.
A gross structural or cladding defect is a known or suspected condition of the assembly structure or fuel ain cladding that results in the fuel not meeting its transportation casc design basis criteria. The cask shielding. criticality, thermal, and radiological design enalyses typically assume that the assembly structure and pin cladding provide sufficient structural integrity to retain the fuel pellets in the fuel assembly geometry for normal and accident conditions. In addition, both individual fuel rods and fuel assemblies should have adequate structural integrity to preclude fuel handling or operational safety problems during loading and unloading operations. It is the responsibility of the licensee to ensure that fuel placed in the transportation package meets the design basis conditions. Alternate means. such as canning, are required for fuel that does not meet the design basis conditions.
6-13 Clarify the descriptions of the basket interior parameters considered.
including a discussion of their relationships to structural analysis i results. i The second paragraph of Section 6.4.2.1 is not clear in its designation ;
of fuel sleeve locations. Part of the confusion lies in the fact that I 3 arts of the discussion refer to the calculational model that shows only 1alf of the basket. The description should reference the "CEL#~
designation from Figure 6.3-1 or use a better illustration for designating fuel sleeve locations.
The eighth paragraph of Section 6.4.2.1 is not clear in describing:
(a) the effect of central cross displacements on " flux traps" (or on presumed flux trapping effects? - see item 6-2 above); and (b) the KENO-Va modeling of vertical and horizontal dis)lacements of the central l
cross. .An explanation is needed to eliminate t1e apparent inconsistency between the statements about flux trap effects found here and on sheet 4 l
of BNFL 1.10.06.14 Revision 1.
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, . - l The SAR's descri3 tion of the parameter configurations considered should I also include a s1 ort discussion of how they relate to the results of the structural analyses for normal and accident conditions.
l 6-14 Provide BNFL 1.10.06.16. Revision 0 and SNC Document No. CCV-1.1.3 4
BNFL 1.10.06.16 is referenced in BNFL 1.10.06.14. .15. and .66, and SNC Document No. CCV-1.1.3 is referenced in BNFL 1.10.06.66. Both l documents are needed for assessing the adequacy of the benchmark validation and bias analysis.
The title of BNFL 1.10.06.16 is " Benchmark Calculation and Validation of the SCALE 4.1 Code Package (PC Version) for Analysis of Cask Systems with i Highly Enriched Fresh Fuel. Fixed Paisons, and Unborated Water." Here. '
the term " highly enriched." generally reserved for uranium enrichments i well in excess of 20% U-235. appears to be inappropriate for the loading l enrichments of this cask design and should be explained or revised. )
6-15 Justify the assumed calculational bias and bias treatment by comparison against the results of more rigorous statistical methods for analyzing ,
and treating bias, such as those described in NUREG/CR-6361. l Chapter 7.0 Operating Procedures The following regulatory requirements are applicable in this chapter: 10 CFR ,
71.7. 71.35(a). 71.39. 71.71(a). 71.73(a) 71.111. and Subparts A and G. '
including U.S. Department of Transportation regulations referenced in 10 CFR 71.5 and Subpart G. It should be noted that other regulatory requirements may be applicable to this section.
7-1 Provide the basis for your position that ultrasonic testing (UT) of the closure lid welds cannot be readily performed.
1 The response to RAI-1. question 7-9. states that no UT examination could l be performed on the lid closure welds of the basket due to l inaccessiblity and ALARA. The staff believes that a UT examination of the final 3/4-inch closure weld on the basket can be performed with the i use of angle beam probe (for checking weld fusion in the tapered weld l prep area and for checking weld volume integrity) and a thin. 1/4-inch !
thick, straight beam 3 robe (for checking the weld fusion at the shell).
ALARA concerns could 3e minimized by the shield lid assembly and the cask being full of water. Provide the ALARA analysis and a detailed explanation as to why you believe UT examination techniques could not be performed.
7-2 Submit Drawing SA-001 The response to RAI-1. question 7-18, states that Drawing SA-001 was provided as part of the submittal. However, it is not a part of the Tran$ tor drawing package nor is it identified in the list of 10 CFR Part 71 drawings.
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i Chapter 8.0 Acceptance Tests and Maintenance Program ;
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The following regulatory requirements are applicable in this chapter: 10 CFR j 71.7. 71.35(a). 71.39. 71.71(a). 71.73(a). and Subparts G and H. It should be '
noted that other regulatory requirements may be applicable to this section.
i 8-1 Place inspection acceptance criteria, as identified in the response to RAI-1 question 8-13. into the SAR.
8-2 Correct inspection criteria of RAI-1. question 8-13. to read: "no" significant galling, visible cracks, etc.
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