ML20211K347
| ML20211K347 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco, 07200011 |
| Issue date: | 08/31/1999 |
| From: | Hall J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Shankman S NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| TAC-L10017, NUDOCS 9909070195 | |
| Download: ML20211K347 (19) | |
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'f UNITED STATES r
- ' [ NUCLEAR REGULATORY COMMISSION 2 WASHINGTON, D.C. So88M001 August 31, 1999 MEMORANDUM TO: Susan F. Shankman, Deputy Director
' Licensing and Inspection Directorate Spent Fuel Project Office, NMSS FROM: James R. Hall, Project Manager . mg4 .7 Spent Fuel Licensing Section /
Licensing and Inspection Direct ,rpte Spent Fuel Project Office, NMSS
SUBJECT:
SUMMARY
OF JULY 28,1999, MEETING WITH SACRAMENTO MUNICIPAL UTILITY DISTRICT (SMUD) REGARDING THE APPLICATION FOR AN INDEPENDENT SPENT FUEL STORAGE INSTALLATION AT RANCHO SECO (TAC NO. L10017)
On July 28,1999, staff from the U.S. Nuclear Regulatory Commission (NRC) met with representatives of the Sacramento Municipal Utility District (SMUD) to discuss the staff's July 21,1999, request for additional information (RAI), regarding SMUD's application for an independent spent fuel storage installation (ISFSI) at the Rancho Seco site. Also attending the meeting.were representatives of Transnuclear West, Inc. (TNW), the vendor of the dry cask storage system, and Science Applications International Corporation (SAIC), the staff's technical assistance contractor for the review of the Rancho Seco ISFSI. An attendence' list is included as Attachment 1; Attachment 2 is the RAI dated July 21,1999. SMUD indicated during the meeting that this RAI did not contain any proprietary information and could be released to the public. This meeting wa's noticed on July 15,1999.
The NRC staff summarized the recent meetings, phone calls, and correspondence related to the subject application, since SMUD's submittal on January 28,1999, of Revision 2 to the i Safety Analysis Report for the Rancho Seco ISFSI. These included meetings on February 9 and March 23,1999; SMUD's submittal of proposed Technical Specifications for the ISFSI on April 30,1999, and Revision 3 of the SAR on May 28,1999; and the staff's letters transmitting the review schedule and the RAI, dated June 23 and July 21,1999, respectively.
The staff and applicant then discussed each item in the July 21,1999, RAI to ensure that there k )
was a mutual understanding of the information needed for the staff and its contractors to '
complete the review. In response to several of the RAI questions, SMUD agreed to provide more detailed discussions in the SAR. SMUD and TNW also indicated that the input and output data requested by SAIC to perform independent confirmatory structural analyses will be provided. To clarify questions related to the assumptions used in off site doses calculations, l the staff agreed to provide the current position to SMUD. Co7 i
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9909070195 990831 PDR ADOCK 05000312 y PDR CONTACT: J.R. Hall, NMSS/SFPO (301) 415-1336
I S. Shankman . .
The staff and applicant confirmed that the RAI response will be submitted by September 10, 1999, consistent with the schedule published in the staff's June 23,1999, letter. The staff reiterated that, for the duration of the review, the applicant should be available to meet with the staff on short notice to resolve any remaining issues. The staff acknowledged the need for a meeting to resolve security issues as soon as possible. Also, the staff and SMUD agreed to hold a future meeting to address the proposed Technical Specifications.
l The meeting was an exchange of information and no regulatory actions or decisions were made.
Docket No. 72-11 Attachments: 1. Attendance List
- 2. RAI dated July 21,1999 i
cc w/att: Steven Mirsky, SAIC I 1
DISTRIBUTION:
Dockets NRC File Center PUBLIC NMSS R/F SFPO R/F NJensen, OGC NRC Attendees WHodges LKokajko PEng EEaston SGagner, OPA BSpitzburg, RIV VEverett, RIV FLyon, NRR G:\SMUD\0728 SUM.wpd JPH 8/20 j OFC: SFPO 6 SFPg lE SFPO /l/
NAME: JRHall gg [arpe CRCh DATE: 8/2,f/99[ 8/[/99 8d[/99 OFFICIAL RECORD COPY
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Attachment 1 Attendance List '
July 28,1999 Meeting between the Nuclear Regulatory Comitdssion and the Sacramento Municipal Utility District ATTENDANCE LIST Name Affiliation Bill Brach NRC/SFPO, Director Ross Chappell NRC/SFPO, Section Leader Eric Leeds NRC/SFPO, Section Leader James R. Hall NRC/SFPO, Project Manager Dick Dudley NRC/PDND, Project Manager Steve Mirsky SAIC i David Williamson SAIC John Stokley* SAIC Jerry Delezenski SMUD John Walkin SMUD ]
q Bob Jones SMUD Kyle Jones Transnuclear West Ian McInnes Transnuclear West David Krohn* Precision Components Corporation Steve Schulin* lbex Group
- Participated by telephone
' Attended as a member of the public
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Attachment 2 Rec uest for Acditional Information Dated July 21,1999 l I
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\* UNITED STATES
';* y- NUCLEAR REGULATORY COMMISSION 1 WASHINGTON, D.C. 30086 4001 s M (IK ,
July 21,1999 Mr. Steve Redeker Manager, Plant Closure and Decommissioning Sacramento Municipal Utility District s 6201 S Street
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P.O. Box 15830 '
Sacramento, CA 95852-1830 i
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION (RAl)
CONCERNING THE RANCHO SECO INDEPENDENT SPENT FUEL STORAGE INSTALLATION (TAC NO. L10017)
Dear Mr. Redeker:
6
. By letters dated January 28 and May 28,1999, the Sacramento Municipal Utility District (SMUD) submitted Revisions 2 and 3 of the Safety Analysis Report for the Rancho Seco independent Spent Fuel Storage Installation (ISFSI). These revisions incorporated changes to reflect your responses to previous questions raised by the Nuclear Regulatory Commission (NRC) staff and to reflect other design or operational changes proposed by you. Based on our
- review of these most recent revisions of the SAR, the staff has prepared the enclosed Request for AdditionalInformation (RAI).
Your response to the enclosed RAI is expected by September 10,1999, as provided in my letter to you, dated June 23,1999. You are requested to subm!! 6 copies of your response,in order to expedite our review.-If you are unable to meet the September 10 milestone, please .
- notify us in writing, at least 2 weeks in advance, of your new submittal date and the reasons for the delay. The staff will then assess the impact of the new submittal date and publish a revised !
schedule.
. Although parts of this request were derived from proprietary information previously submitted to the NRC, the staff has not identified any proprietary information in this letter or its enclosure.
To preclude any inadvertent release of proprietary information, the enclosed RAI will not be l placed in the public document room (PDR) for 30 days. If SMUD does not identify to the staff any proprietary information within that time, the enclosed RAI will be placed in the PDR. If any i proprietary information is included in your response, please provide an affidavit supporting your !
request for withholding such information from public disclosure pursuant to 10 CFR 2.790. l G
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, July 21, 1999 l*
S. Redeker . .
A meeting between our staffs to discuss the enclosed RAI is scheduled for Wednesday, July 28,1999, at NRC headquarters in Rockville, Maryland. If you have any comments or questions concerning this request, please contact me at (301) 415-1336.
Sincerely, (Original Signed by:)
Jarnes R. Hall, Senior Project Manager Licensing Section Spent Fuel Project Office Office of Nuclear Material Safety !
and Safeguards Docket No.: 7211
Enclosure:
Request for Additional information l
l Distribution:
Docket NRC File Center PUBLIC NMSS r/f SFPO rM i
EWBrach SShankman WHodges RDudley PEng EEaston LKokajko l DSpitzberg, RIV VEverett, RIV SMirsky, SA!C G:\SMUD\R AI_99_R1.WPD OFC SFPO E SFP W E SFPO [ SFP h / 8 NAME JRHall/ f-8 VL pe EJLeeds CRC p DATE 07$/9 07/$99 07M/99 07h99 k C = Cover E = Cover & Enclosure N = No copy OFFICIAL RECORD COPY I
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Enclosure REQUEST FOR ADDITIONAL INFORMATION By application dated October 4,1991, the Sacramento Municipal Utility District (SMUD) requested issuance of a license for the proposed independent Spent Fuel Storage Installation (ISFSI) at the Rancho Seco Nuclear Generating Station. SMUD also submitted the initial Safety Analysis Report (SAR) with the application and revised the SAR in letters dated October 27,1993, and January 28 and May 28,1999. This request for additional information (RAI) identifies additional information needed by the U.S. Nuclear Regulatory Commission (NRC) staff in connection with its review of the application and specifically, Revision 3 of the SAR, dated May 28,1999. NUREG 1567," Standard Review Plan for Spent Fuel Storage Facilities" (draft version), is the primary regulatory guidance document used by the staff in its review of the application.
The questions in this RAI are grouped by chapters in the format of NUREG 1,567, as listed below. Questions addressing specific calculation packages are listed under the calculation package number and not under the associated chapter number. Calculation packages reviewed for this RAI are listed in Table 1 below.
Each individual RAI describes information needed by the staff for it to complete its review of the application and/or the SAR and to determine whether the applicant has demonstrated compliance with the regulatory requirements.
This RAI is organized by section as follows:
Chapter 1.0 General Description Chapter 2.0 Site Evaluation Chapter 3.0 Operation Systems Evaluation Chapter 4.0. SSC and Design Criteria Evaluation Chapter 5.0 installation and Structural Evaluation Chapter 6.0 Thermal Evaluation Chapter 7.0 Shielding Evaluation Chapter 8.0 Criticality Evaluation Chapter 9.0 Confinement Evaluation Chapter 10.0 Conduct of Operations Evaluation Chapter 11.0 Radiation Protection Evaluation Chapter 12.0 Quality Assurance Evaluation Chapter 13.0 Decommissioning Evaluation Chapter 14.0 Waste Confinement and Management Evaluation Chapter 15.0 Accident Analysis Evaluation Chapter 16.0 Technical Specifications Evaluation Calculations Drawings 1
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' Table 1 List of Calculation Packages Reviewed Caleu'lation i2069.0201 Rev. 3 FC Structural Analysis (FO/FC internals)
Calculation 2069.0202 Rev.1 SMUD HSM Analysis Calculation 2069.0203 Rev. 2 MP-187 Cask 10 CFR 72 Structural Analysis Calculation 2069.0204 = Rev. 5 MP-187 Guide Sleeves (FO/FC)
Calculation 2069.0205 Rev. 3 FF-DSC Structural Analysis (FF)
Calculation 2069.0216 Rev. O FO DSC Shell Assembly Structural Analysis Calculation 2069.0217 Rev. O FO/FC DSC Basket Assembly Structural Analysis Calculation 2069.0219 Rev. O FC/FF DSC Shell Assembly Structural Analysis Calculation 2069.0220 Rev.2 FC/FF Shield Plug Analysis Calculation 2069.0400 Rev. 3 Rancho Seco HSM Thermal Analysis Calculation 2069.0401 Rev. 3 MP 187 Cask Thermal Analysis for Onsite Transfer and Storage Conditions Calculation 2069.0453 Rev. 3 MP-187 Cask Peak Pressure Calculation for DSC Cavity and Cask DSC Annulus During Storing DSC Calculation 2069.0500 Rev.1 Radiological Source Term Calculation for Rancho Seco Calculation 2069.0501 Rev. 3 Rancho Seco Shielding Material Densities Calculation Calculation 2069.0502 Rev. 4 Rancho Seco Site Dose Calculation .
Calculation 2069.0503 Rev. 2 Rancho Seco Occupational Exposure Calculation
' Calculation 2069.0505 Rev. 2 Rancho Seco Fission Gas Release Dose Assessment Calculation Z SFC M2557 Rev. O Decay Heat Value of Spent Fuel and Control Components Calculation NUH005.0450 Rev. O Dry Storage Clad Temperature Acceptance Limit for Rancho Seco Fuel Assemblies Calculation NUH005.0552 Rev. 2 MP-187 FO-DSC Radiological Source Term Calculation Calculation NUH005.0553 Rev. 2 MP 187 FC-DSC Radiological Source Term Calculation 1 Chapter 1.0 General Description
- The staff has no questions at this time.
Chapter 2.0 Site Evaluation
- The staff has no questions at this time.
Chapter 3.0 Operation Systems Evaluation The following regulatory requirements are applicable to the RAls in this chapter: 10 CFR Parts 20 and 50, and 10 CFR 72.11,72.24(e),72.24(f),72.24(I)(2),72.40(a)(5),72.104,72.106, 72.122(f),72.122(l),72.128(a)(5),72.150,72.212(b)(9),72.234(f),72.236(h), and 72.236(l). It should be noted that other regulatory requirements may be applicable to this section.
Re: SAR Volume'll, Rev. 3, Chapter 5 3-1 Provide the detailed operation sequence for the installation of the impact limiters onto the MP 187 cask and for final placement of the MP 187 cask onto the transportation vehicle. Explain how those operations satisfy the lifting height restrictions of the Technical Specifications (TS).
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The evaluation of the impact limiter analysis is reviewed under 10 CFR Part 71; however, the Part 72 review must verify that the operation, prior to having the impact limiters in functional condition, does not involve a lift exceeding the 80-inch lifting height restriction. Lifting the loaded MP-187 greater than 80 inches, such as might be required to install the impact limiters, would be an unanalyzed loading condition. Also, information on the operation sequence for placement of the MP-187 onto a rail car or truck is necessary to determine if any equipment is used which could be classified as important to safety and which has not been previously reviewed. .
Chapter 4.0 SSC and Dealgn Criteria Evaluation The following regulatory requirements are applicable to the RAls in this chapter: 10 CFR 72.2(a)(1), 72.11, 72.24(c), 72.24(d), 72.24(e), 72.24(f), 72.24(q), 72.24(m), 72.24(l)(2),
72.30(a), 72.44 (d), 72.102, 72.104, 72.106, 72.120(a), 72.122, 72.124, 72.126, 72.128, 72.130, and 72.236. It should be noted that other regulatory requirements may be applicable to this i section.
Re: SAR Volume I, Rev. 3, Chapter 3 41 Provide the number of curies of plutonium in each Rancho Seco damaged fuel assembly, if there are greater than 20 curies of plutonium in any damaged fuel assemblies, then, l for future transportation reasons, these assemblies would have to be placed into a sealed confinement can within the dry shielded canister (DSC), not a gross particle ,
confinement can, as specified in 10 CFR 71.63(b).
42 Provide design information for all Rancho Seco Fuel Assembly Integral Components intended to be stored in the ISFSI and explain how this data is incorporated into the thermal, shielding, DSC pressure, and criticality analyses for the ISFSI.
The SAR does not provide design data for the neutron source assemblies. Calculation 2069.0500, Rev.1, Radiological Source Term Calculation for Rancho Seco, pages 25 through 27, should be revised to include the contribution of reactor startup neutron source assemblies. In addition, the applicant should describe how all fuel assembly integral components intended for storage have been adequately evaluated in terms of their impact on design safety parameters.
Chapter 5.0 Installation and Structural Evaluation .
The staff has no questions at this time. 4 Chapter 6.0 Thermal Evaluation The staff has no questions at this time. .
i Chapter 7.0 Shielding Evaluation The staff has no questions at this time. .
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Chapter 8.0 Criticality Evaluation The fo!Iowing regulatory requirements are applicable to the RAls in this chapter: 10 CFR 72.11,72.24(c)(3),72.24(d),72.124(b),72.236(c), and 72.236(g). It shtuld be noted that other regulatory requirements may be applicable to this section. .
Re: SAR Volume I, Rev. 3 Chapter 3 81 Provide the calculation (s) that support the nuclear criticality safety of the Rancho Seco ISFSI as described in SAR Section 3.3.4.
The SAR does not provide specific information on the criticality safety of the ISFSt. The SAR should specify the method for criticality safety analysis; assumptions regarding enrichment, moderation, and fixed neutron absorbing material composition; maximum calculated value of kW and any requirements for spent fuel pool water soluble boron i
concentration.
. Chapter 9.0 Confinement Evaluation
. The staff has no questions at this time.
Chapter 10.0 Conduct of Operations Evaluation l
The staff has no questions at this time.
Chapter 11.0 Radiation Protection Evaluation l
The following regulato,ry requirements are applicable to the RAls in this chapter: 10 CFR l 20.1201,20.1207,20.1208,20.1101, Subpart F; and 10 CFR 72.11,72.24(e),72.104, 72.104(b),72.106, and 72.126(a). It should be noted that other regulatory requirements may be applicable to this section.
i Re: SAR Volume 1, Revision 3, Table 7-4 11-1 Explain the technical basis for footnote 3 of Table 7-4, which revises the maximum annual dose to a member of the public at the nearest site boundary below the value
- l. delineated in the table.
The table shows that the dose rate at the west site boundary (the highest offsite dose rate) is 4.31x10'8 mrem /hr, which equates to an annual dose of 37.8 mrem, which exceeds the 10 CFR 72.104 limit of 25 mrem per year. However, a footnote to this location states that the application of a source term correction factor reduces the calculated maximum annual dose to 18.3 mrem. The supporting basis for this reduced source term should be provided in the SAR and referenced in this footnote. ,
l Chapter 12.0 Quality Assurance (QA) Evaluation The following regulatory requirements are applicable to the RAls in this chapter: 10 CFR 72.11 and Subpart G. It should be noted that other regulatory requirements may be applicable to this section.
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Re:~ SAR Volume I, Rev. 3, Chapter 11, and the Rancho Seco Quality Manual (RSOM) -
12-1 Clarify the relationship between the Director, Power Generation, Quality / Licensing / Administrative Superintendent, and Plant Manager.
RSOM Policy," Quality Assurance Policy," states,"The Director, Power Generation, has delegated the responsibility and authority for development and execution of the Rancho Seco Quality Assurance Program to the Quality / Licensing / Administrative Superintendent." RSOM Section I," Organization" states,"The Plant Manager has the responsibility and authority for development and execution of the Rancho Seco Quality Assurance Program... The Plant Manager has delegated authority and responsibility for development and execution of the Quality Assurance Program to Nuclear Quality."
Section 72.142 requires that the licensee shall clearly establish and delineate the -
3 authority and duties of persons and organizations performing functions associated with I quality.
12 2 Describe how OA program criteria applicable to the ISFSI are determined. Specifically, clarify the intent of paragraph 4.3.2 of RSOM Section ll," Quality Assurance Program,"
and explain why criteria Vill, IX, XI, and XIV on Attachment Il 1 of RSQM Section 11 are not fully applicable for 10 CFR Part 72.
Paragraph 4.3.2 of RSOM Section Il states," Appropriate elements / criteria of this manual, as indicated in Attachment 11 1, apply to ISFSI structures, systems, and components which are determined by Technical Services to be impor%nt to safety in accordance with 10 CFR 71 and 10 CFR 72." Criteria Vill, IX, XI, and XIV on Attachment 111 are not marked with an "X." No "X" means not all requirements of that RSOM criterion apply.
-12 3 Revise paragraph 3.0 of RSOM Section XVil to reference the correct paragraph in Section 11 of the manual. Section 11, Paragraph 4.2 does not appear to be the appropriate reference.
Chapter 13.0 Decommissioning Evaluation !
The following regulatory requirements are applicable to the RAls in this chapter: 10 CFR 72.11, 72.24(q),72.30,72.54,72.130, and 72.236(i). It should be noted that other regulatory requirements may be applicable to this section.
Re: .SAR Volume 1, Rev.3, Chapters 3 and 9 13 1 The SAR should be revised to reference, and to summarize, the proposed Rancho Seco ISFSI decommissioning plan, submitted by letter dated June 27,1995. Explain how this plan complies with the financial assurance requirements of 72.30(c).
. Chapter 14.0 Waste Confinement and Management Evaluation The staff has no questions at this time.
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Chapter 15.0 Accident Analysis Evaluation The following regulatory requirements are applicable to the RAls in this10chapter: CFR 72.2(a)(1 ), 72.11, 72.24 (b), 72.24 (d), 72.24 (m), 72.104 (a), 72.106(b), 72.122(b), 72.122(c),
72.122(d), 72.122(g), 72.122 (h), 72.122(i), 72.122(l), 72.124(a), 72.126(a), 72.236(c),
72.236(d), and 72.236(l). It should be noted that other regulatory requirements may be applicable to this section.
Re: SAR Volume I, Rev. 3, Chapter 8.2; Volume ll, Rev. 3, Chapter 8.3; Volume lil, Rev. 3 Chapter 8.3 15-1 Explain the basis for not presenting and evaluating a postulated fire accident scenario for both the Horizontal Storage Module (HSM) and the MP-187.
Since the vehicle which moves the MP-187 transfer cask from the spent fuel pool to the HSM for loading a DSC into the HSM carries flammable fuel, it is possible for fuel leakage and subsequent combustion to occur in or around the MP-187 or the HSM.
Chapter 16.0~ Technical Specifications Evaluation .
The following regulatory requirements are applicable to the RAls in this chapter: 10 CFR 72.11, 72.24(g),72.26,72.44(c),72.104,72.106,72.234(a),72.236, and Subparts C, E, F, G, H, and I.
It should be noted that other regulatory requirements may be applicable to this section.
Re: SAR Volume 11, Rev. 3, Operating Controls and Limits 10.3.3, and proposed TS 5.5.3.1 and 5.5.3.2, dated April 30,1999 16 1- Define what is meant by the phrase,".. a significant unexplained difference (or changc.)..." in the SAR and in proposed TS 5.5.3.1. Specific acceptance values should be provided for the HSM roof concrete temperature, and for the differential temperature between the roof vents and ambient in proposed TS 5.5.3.2.
CALCULATIONS Re: Calculation 2069.0201, Rev. 3, FC Structural Analysis (FO/FC internals) c-1 Describe the difference between load combinations A3 and A4 as given in Table 3.
Table 3 shows intemal pressure loads, which cannot affect the internals since the internals do not retain any differential pressure. This statement is also given in section 3.3 of the calculation.
c2 Clarify the temperature conditions assumed for load combination A4.
Table 11 lists the normal design condition temperatures for various DSC cornponents.
Explain how these temperatures are considered in the subject calculation for load combination A4.
c-3 Provide electronic format input and output files for the ANSYS runs (so that they can be evaluated on a computer using ANSYS 5.5.2) for the following load combination taken from Table 30 of the referenced calculation: A4 (deadicad horizontal, thermal, handling horizontal, pressure assumed to be zero).
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j The objective is to verify stresses in the components for the load combination listed
,- above as follows: (1) FO/FC spacer disc and (2) FO/FC support rod.
Re: Calculation 2069.0202, Rev.1, SMUD HSM Analysis c- 4 Transmit Reference 3 for the HSM, NUH004.0200, and if FEM analyses are used to support the conclusions in 004.0202, send the ir.put and output files.
The objective is to verify the results in the above reference calculation.
Re: Calculation 2069.0203, Rev. 2, MP-187 Cask 10 CFR 72 Structural Analysis.
c-5 Provide electronic format input and output files for the ANSYS runs (so that they can be evaluated on a computer using ANSYS 5.5.2) for the following load combinations taken from Tables 31 and 32 of the calculation: Load combination C4 (DW + accident pressure of 50 psig, + thermal loading, + tomado wind /tomado missile), and load combination D1/D3 (DW + thermal effects + drop).
The objective is to verify stresses in the components for the load combi, nations listed above as follows: (1) outer and inner shells for LC 4, and (2) outer and inner shells, top cover plate, top corner forging, bottom comer forging for the D1/D2 case.
Re: Calculation 2069.0205, Rev. 3, FF-DSC Structural Analysis (FF) c-6 Provide electronic format input and output files for the ANSYS runs (so that they can be evaluated on a computer using ANSYS 5.5.2) for following load combinations taken from Tables 51 and 5 3: TR 9, TR-10, and TR-11 for the drop analyses; T=117'F. both horizontal and vertical orientations; and Service Level A for the 20*F and 117'F cases.
Also provide Table 5-1 of the calculation, which was missing from your submittal.
The objective is to verify stresses in the components for the load combinations listed above as follows: (1) TR 9 and/or TR 10 and/or TR 11 spacer disc with the 64.8 kai stress intensity and (2) Service Level A for the two thermal extremes for the spacer disc. I Re: Calculation 2069.0216, Rev. O, FO DSC Shell Assembly Structural Analysis i
c7 Provide electronic format input and output files for the ANSYS runs (so that they can be evaluated on a computer using ANSYS 5.5.2). The following load combinations are taken from Table 71 of ie referenced calculation:- (1) TR-11 (75 G side drop,10 psi, 100*F), (2) HSM 8 (seNic,10 psi,100*F), and (3) UL-7 (unload 80 kip,41 psi, 100'F).
The objective is to verify stresses in the components for the load combinations listed above as follows: (1) TR-11; top half shell, outer top cover plate, inner top cover plate, weld top outer cover, bottom half shell, bottom inner cover plate, upper weld support ring base metal; (2) HSM 8; top half shell and weld inner top cover plate; and (3) UL-7 bottom outer cover plate. .
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c4 Explain how the elements are defined which represent the void between the upper shall and top outer cover plate in Figure 8-3.
The top outer cover plate has two beveled edges as a part of the weld prep. The top groove is filled with weld material and the bottom groove remains void. Two elements are shown in the figure cited above.
Re: Calculation 2069.0217. Rev. O, FO/FC DSC Basket Assembly Structural Analysis c-9 Provide electronic format input and output files for the ANSYS runs (so that they can be evaluated on a computer using ANSYS 5.5.2) for the following load combinations taken from Tables 101,10-2, and 10-3 of the referenced calculation: (1) TR 11 (45' aide drop,10 psi,100*F), (2) TR-9 or TR 10 (and drop,10 psi,100'F), and (3) TR4.
The objective is to verify stresses in the components for the load combinations listed above as follows: (1) TR 11 for the spacer disc with the 78.3 ksi stress intensity, (2) TR 9 or TR 10 for the spacer sleeve for which the interactive equation 22 has a value of 0.94 stress ratio, and (3) TR 8 for the support rod for Service Level A. If there is no FEM analysis for (2) and/or (3), then there is no need to send them.
Re: Calculation 2069.0219, Rev. O, FC/FF DSC Shell Assembly Structural Analysis c-10 Provide electronic format input and output files for the ANSYS runs (so that they can be evaluated on a computer using ANSYS 5.5.2.) taken from Table 71 of the referenced i calculation: (1) TR-11 (75 G side drop,10 psi,100'F), (2) HSM-8 (seismic,10 psi, 10'F), (3) UL-7 (unload 80 kip,41 psi,100*F).
The objective is to verify stresses in the components for the load combinations listed above as follows: (1) TR 11; top half shell, outer top cover plate, weld inner top cover ;
plate, bottom half shell, and upper weld support ring base metal; (2) HSM-8; top half shell, weld inner and outer top cover plates, and inner and outer top cover plates; and (3) UL-7 shell bottom.
Re: Calculations 2069.0400, Rev. 3; 2069.0453, Rev. 3; 2069.0500, Rev.1; 2069.0502, Rev. 4; and NUH005.0450, Rev. 0 -
c 11 Provide the following references from the above calculations:
(1) NUH004.0421, Rev. 3; (2) NUH004.0423, Rev. 2; (3) NUH005.0350, Rev. 8; (4) NUH004.0100, Rev. 2; (5) EPRI TR 104329, May 1995; (6) NUH004.0509, Rev.0; and (7) NUH002.0203, Rev. O.
- Re: Calculation 2069.0401, Rev. 3 c-12 Justify the use of the 1.08 axial power peaking factor assumed for Rancho Seco fuel and presented on page 7 of this calculation. This thermal axial peaking factor is smaller
- than the shielding analysis axial peaking factor.
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c-13 (a) Explain the trend of carbon steel temperature dependent thermal conductivity in Table 4 5 on page 18.
(b) Justify the single value of aluminum thermal conductivity delineated in Table 65 on page 19.
Sources of thermal properties show that the thermal conductivity of carbon steel
, continuously decreases with increasing temperature, whereas this calculation lists j carbon steel thermal conductivity as increasing up to a temperature of 200*F, staying l constant to 300*F, and then decreasing at temperatures above 300*F. In addition, l thermal property sources indicate an aluminum thermal conductivity which is about 29% I lower than the value used by the applicant.
Re: Calculation 2069.0453, Rev. 3, MP-187 Cask Peak Pressure Calculation for DSC Cavity and Cask DSC Annulus During Storing DSC c 14 Recalculate the r.:;rmal, off normal, and accident DSC internal pressures to account for the contribution of helium; present as both a backfill gas in burnable poison rod control components and produced by boron absorption in the burnable poison rods.
The calculation of DSC pressure should include all sources of gas present within the DSC, including the control components, which are part of the Rancho Seco ISFSI design.
c 15 Evaluate the impact on calculated peak DSC pressures due to the reduccd intemal free i
volume from the optional spacer disk materials snd revised spacer disk intervals which are discussed in Appendix A of this calculation.
The spacer disk material and intervals may reduce the available free volume in the DSC, which would result in an increase in the calculated peak DSC internal pressures. As stated in NUREG 1567, a bounding minimum design specific DSC internal free volume should be used for the calculation of peak DSC pressures.
Re: Calculation 2069.0500, Rev.1, Radiological Source Term Calculation for Rancho Seco c 16 Revise the gamma source term calculation of Rancho Seco fuel assembly top nozzles, bottom nozzles, gas plenum regions, and axial power shaping rod assemblies using cobalt impurity levels that are bounding for the inconel and stainless steel in these components.
The applicant assumed a stainless steel and inconel cobalt impurity level of 1000 ppm, as presented on page 23 of this calculation. The staff has accepted a cobalt impurity of 1000 ppm for stainless steel, but has previously required a minimum cobalt impurity level of 4700 ppm for inconel. The applicant should either provide fabrication specification documentation that justifies a specific cobalt impurity level for all the Rancho Seco fuel or use 4700 ppm cobalt for allinconel fuel components. Also note that calculation NUH005.0553, Rev. 2, MP 187 FC DSC Radiological Source Term Calculation, page 5, assumes a 1200 ppm cobalt impurity level for inconel and stainless steel. The assumptions used in both calculations should be consistent.
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.. i, c-17 Justify the use of 141.8-inch B&W 15 x 15 active fuellength on pages 8 and 9 of this
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calculation. The ' active fuellength is used to calculate the mass of components, which are subsequently part of the source term calculation.
Some fuel design references indicate that the B&W 15x15 fuel design has a 144-inch 3 active fuel length rather than the 141.8-inch length used by the applicant. The applicant should provide the technical basis for the 141.8-inch active fuel length for all Rancho Seco nuclear fuel or use the more bounding active fuel length value.
l Re: Calculation 2069.0501, Rev. 3, Shielding Material Densities Calculation c 18 Explain the difference in DSC cavity dimensions and volume presented in Table 6 on !
page 12, as compared to Calculation 2069.0453, Rev. 3, pages 7 and 9, as well as Rancho Seco ISFSI SAR Volume I, Rev. 3, Section 1.2.2.
Calculation 2069.0501 uses a DSC cavity inside radius and length of 27.78 inches and 186.2 inches, respectively, whereas calculation 2069.0453 uses a radius of 32.97 inches and a cavity length of 167 inches. SAR Section 1.2.2 lists DSC cavity lengths of 167 inches or 173 inches, depending on the contents of the DSC, and an inside radius of 33 inches. All calculations in support of the Rancho Seco ISFSI should use appropriate consistent input parameters that reflect the design.
Re: Calculation 2069.0502, Rev. 4, Site Dose Calculation c-19 Revise the calculation of nearest site boundary total annual exposure on pages 27 through 28 to account for the ratio of design basis fuel assembly neutron and gamma source term rather than the spent fuel pool total decay heat ratio.
In this calculation, the basis for accounting for the additional decay time to meet the 25 mrem / year site boundary dose limit in 10 CFR 72.104 is predicated on a ratio of Rancho )
Seco spent fuel pool total decay heat. Although relative decay heat is a measure of the reduction in rediation emanating from a fuel assembly, radiological source term is a more direct comparison parameter since the source term determines the calculated off-site dose rate, not its decay heat. Both magnitude and spectrum of the revised design fuel source term should be compared to account for the increased decay time, c 20 Explain how the assumed " universe" sphere of 10 gamrna mean free paths surrounding the ISFSI, presented on page 4, relates to the mean free paths and scattering of neutrons outside of the ISFSI.
The MCNP model of the ISFSI for off site dose calculations should encompass a sufficient distance to adequately simulate particle scattering and back scattering to all points where the dose rate is calculated. The nature of the MCNP computer code method requires that a statistically significant number of particles scatter and pass through locations where the dose rate is calculled.
Re: Calculation 2069.0503, Rev. 2, Occupational Exposure Calculation c 21 Justify the axial flux shape assumed for Rancho Seco fuel, which is presented on page 14 of this calculation and is based on Westinghouse fuel assembly data.
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The axial flux shape is used to calculate peak dose rates and should be based on the applicant's fuel design. Rancho Seco used Babcock & Wilcox (now called Framatome) design nuclear fuel, not Westinghouse design fuel.
c-22 Evaluate the effect of axial power shaping rod assemblies (APSRAs), thimble plugging assemblies (TPAs), and neutron startup source assemblies, which could be present in .
the fuel assembly, on the assumed axial peaking.
The axial distribution of gamma radiation from APSRAs,TPAs, and neutron radiation
_ from startup sources may affect the axial distribution of these sources for the entire fuel assembly. If a higher peak resulted, the effect of peak source on calculated dose rate may not be bounding in the current calculation.
c 23 Justify the adequacy of the radial and axial mesh size selected for the DORT-PC models used in this calculation.
The calculated dose rate can be affected by the selected mesh size. A sufficiently fine mesh size will acceptably simulate material attenuation of neutron and gamma radiation.
If the mesh size is too coarse, the calculated dose rate may not be bounding for a specific source and shielding geometry configuration.
Re: Calculation 2069.0505, Rev. 2, Fission Gas Release Dose Assessment c 24 Revise this calculation to conform with either NUREG 1536 Section 7 or NRC Spent Fuel Project Office Interim Staff Guidance (ISG) 5, Revision 1.
This calculation does not consider all the radioisotopes which are delineated in NUREG-196 or ISG 5, Revision 1. In addition, the regulatory guidance provides details regarding which fission product isotopes need to be included, atmospheric dispersion meteorology, population sectors, time frame, and DSC leakage rate.
Re: . Calculation Z SFC-M2557, Rev. O, Decay Heat Value of Spent Fuel and Control Components c 25 Provide Attachments B, C, D, G, H, I, K, N, P, and Q to this calculation.
These attachments are identified in the calculation table of contents but were not provided to the staff by the applicant. These attachments contain information about the j actual 493 spent fuel assemblies at Rancho Seco which is important to the technical l safety evaluation of the Rancho Seco ISFSIlicense application.
Re: Calculation NUH005.0450, Rev. O, Dry Storage Clad Temperature Acceptance Limit for Rancho Seco Fuel Assemblies c 26 Justify the calculated value of long-term storage cladding temperature limit in light of the actual cooling time of Rancho Seco spent nuclear fuel.
This calculation assumes a 5.5 year cooling time for Rancho Seco fuel whereas, based i on plant shutdown in 1989 and the planned loadino in 1999, the shortest cooling time would be approximately 10 years. The CSFM me ...d from PNL-6189, which the applicant uses, results in significantly lower cladding temperature limits for the same fuel 1
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which has a longer cooling time. Using the same dimensions and cladding hoop stress, the Rancho Seco fuel long-term cladding temperature limit would be lower for the 10 year cooling time as compared to the 5.5 year cooling time. The appropriate ~
temperature limit associated with this cooling time should be compared to the calculated maximum cladding temperature for normal conditions with the concomitant 10 year cooling time maximum DSC decay heat.
c-27 Compare the actual Rancho Seco spent fuelinitial rod backfill helium gas pressure (including manufacturing tolerances), cladding dimensions, cladding oxidation, rod internal gas temperature, and calculated peak rod intemal pressure with the values assumed in the generic referenced B&W 15x15 fuel assembly design fuel clad temperature limit calculation.
The cladding temperature limit is based on an interpolation of values calculated in a generic NUHOMS calculation which assumes specific fuel design values for B&W 15x15
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fuel. The applicant should demonstrate that the specific fuel design and operational parameters of the Rancho Seco fuel are equivalent to or bounded by the parameters assumed in the generic calculation in terms of their effect on calculated long-term cladding storage temperature limits.
Re: Calculation NUH005.0552, Rev. 2, MP-187 FO DSC Radiological Source Term Calculation c 28 Justify the 1.17 and 1.12 gamma peaking factors listed on Tables 16 and 17 on pages 42 and 43 in comparison to actual measured Rancho Seco fuel peaking factors and different axial peaking factors used in Calculations 2069.0401, Rev. 3, and 2069.0503, Rev. 2.
Axial peaking factors for ISFSI calculations should be consistent and either based on actual plant specific fuel data or shown to be bounding for actual plant specific fuel data. ;
Axial peaking factors affect calculated peak dose rates and maximum calculated '
temperatures.
c 29 Demonstrate that the selected enrichment burn up combination is bounding for Rancho Seco fuel. The minimum enrichment for the same bum-up and cooling time results in a significantly higher neutron source arid concomitant neutron dose rate.
DRAWINGS Re: Drawings NUH-05 4001, Rev. g; NUH 05-4003, Rev. 7; NUH 05 4004, Rev.10; NUH-05-4005, Rev. 8 d1 Explain how drawing design changes are incorporated into calculations that Were completed prior to the latest drawing revision.
Although the latest revisions of these drawings are dated August and September of 1998, many Rancho Seco ISFSI calculations, including 2069.0503,2069.0501, and 2069.0401 were completed before these drawing revisions. The applicant's calculations must incorporate the current ISFSI design.
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