IR 05000458/2012011

IR 05000458-12-011 & 07200049-12-001, 09/25-27/2012, River Bend Nuclear Station, Unit 1, Independent Spent Fuel Storage Installation

ML12299A101
Person / Time
Site:	River Bend
Issue date:	10/24/2012
From:	Lee Brookhart NRC/RGN-IV/DNMS/RSFSB
To:	Olson E Entergy Operations
Spitzberg D
References
IR-12-011
Download: ML12299A101 (32)

Text

UNITE D S TATE S NUC LEAR RE GULATOR Y C OMMI S SI ON ber 24, 2012

SUBJECT:

RIVER BEND NUCLEAR STATION - INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) INSPECTION REPORT 05000458/2012011 AND 07200049/2012001

Dear Mr. Olson:

A routine site inspection was completed of your dry cask storage activities associated with your Independent Spent Fuel Storage Installation (ISFSI) on September 25 - 27, 2012. An exit was conducted with your staff to discuss the findings of the inspection on September 27, 2012. In addition, a telephonic discussion was held with your staff and representatives of Holtec International on October 24, 2012 to discuss an unresolved item concerning the offsite dose calculations versus the assumptions used for fuel enrichment in those calculations. This issue is discussed in Section 1.2(b) of this report. The focus of this inspection was to review the status of the stored casks, verify ongoing compliance with the Holtec Certificate of Compliance No. 1014 and the associated Technical Specifications; the Holtec Final Safety Analysis Report; the regulations in 10 CFR Part 20 and Part 72; and to review any changes that had been made to your ISFSI program since the last NRC inspection. This review included the areas of radiation safety, cask vent inspections, quality assurance, corrective action program, safety evaluations, records retention, and how you addressed industry issues that affected your ISFSI program. During the week of the inspection, one canister was in the process of being loaded at River Bend. The inspectors observed portions of the drying and transfer operations while on-site. Your ISFSI operations and program were found to be in compliance with the applicable NRC requirements and your storage casks were found to be in good physical condition. No violations of significance were identified.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response if you choose to provide one, will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal, privacy or proprietary information so that it can be made available to the public without redaction.

River Bend Nuclear Station -2-Should you have any questions concerning this inspection, please contact the undersigned at 817-200-1191 or Mr. Lee Brookhart at 817-200-1549.

Sincerely,

/RA/

D. Blair Spitzberg, Ph.D., Chief Repository & Spent Fuel Safety Branch Dockets: 50-458, 72-49 Licenses: NPF-47 Enclosure: Inspection Report 05000458/2012011; 07200049/2012001 w/ Attachment:

1. Supplemental Information 2. Condition Reports Reviewed During the Inspection 3. Loaded Casks at River Bend Nuclear Station

River Bend Nuclear Station -3-Electronic distribution by RIV:

Regional Administrator (Elmo.Collins@nrc.gov)

Deputy Regional Administrator (Art.Howell@nrc.gov)

DRP Director (Kriss.Kennedy@nrc.gov)

DRP Deputy Director, Acting (Allen.Howe@nrc.gov)

DRS Director, Acting (Tom.Blount@nrc.gov)

DRS Deputy Director, Acting (Jeff.Clark@nrc.gov)

Senior Resident Inspector (Grant.Larkin@nrc.gov)

Resident Inspector (Andy.Barrett@nrc.gov)

Branch Chief, DRP/C (Vince.Gaddy@nrc.gov)

Senior Project Engineer, DRP/C (Bob.Hagar@nrc.gov)

Senior Project Engineer, DRP/C (Rayomand.Kumana@nrc.gov)

RBS Administrative Assistant (Lisa.Day@nrc.gov)

DNMS Director (Anton.Vegel@nrc.gov)

DNMS Deputy Director (Vivian.Campbell@nrc.gov)

RSFSB Branch Chief (Blair.Spitzberg@nrc.gov)

RSFSB Inspector (Lee.Brookhart@nrc.gov)

RSFSB Inspector (Vincent.Everett@nrc.gov)

RSFSB Inspector (Eric.Simpson@nrc.gov)

Project Manager (Alan.Wang@nrc.gov)

Project Manager, SFST (William.Allen@nrc.gov)

Public Affairs Officer (Victor.Dricks@nrc.gov)

Public Affairs Officer (Lara.Uselding@nrc.gov)

RITS Coordinator (Marisa.Herrera@nrc.gov)

TSB Technical Assistant (Loretta.Williams@nrc.gov)

Regional Counsel (Karla.Fuller@nrc.gov)

Congressional Affairs Officer (Jenny.Weil@nrc.gov)

DRS/TSB (Dale.Powers@nrc.gov)

RIV/ETA: OEDO (Silas.Kennedy@nrc.gov)

OEMail Resources DRAFT: S:\DNMS\RSFS\!brookhart\River Bend\River Bend RB2012011-ISFSI-LEB FINAL: R:\REACTORS\RBS\2012\RB2012011-ISFSI-LEB SUNSI Rev Compl. X Yes No ADAMS X Yes No Reviewer Initials Publicly Avail X Yes No Sensitive Yes X No Sens. Type Initials RIV:DNMS/RSFS RIV:DNMS/RSFS R-IV/C:RSFS LEBrookhart JVEverett DBSpitzberg

/RA/ /RA/ /RA/

10/22/12 10/24/12 10/22/12 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 05000458, 07200049 Licenses: NPF-47 Report Nos.: 05000458/2012011 and 07200049/2012001 Licensee: Entergy Operations, Inc.

Facility: River Bend Nuclear Station, Unit 1 Independent Spent Fuel Storage Installation (ISFSI)

Location: 5485 US Highway 61 St. Francisville, LA 70775 Dates: September 25-27, 2012 Inspectors Vincent Everett, Senior Inspector Lee Brookhart, Inspector Approved By: D. Blair Spitzberg, Ph.D., Chief Repository and Spent Fuel Safety Branch Division of Nuclear Materials Safety Enclosure

EXECUTIVE SUMMARY River Bend Nuclear Station NRC Inspection Report 50-458/2012-11 and 72-49/2012-01 The NRC conducted a routine inspection of the licensees programs for safe handling and storage of spent fuel at their ISFSI and observed operations associated with the loading of Cask #18. Seventeen casks were currently loaded and stored on the River Bend ISFSI pad, which is located within the Part 50 reactor facility protected area. The first seven casks were loaded between 2005 and 2007. These casks were loaded to Certificate of Compliance 1014, Amendment 2 and Holtec Final Safety Analysis Report (FSAR), Revision 3. Casks #8 through

1. 18 were loaded to Certificate of Compliance 1014, Amendment 5 and Holtec Final Safety Analysis Report (FSAR), Revision 7 from 2008 through 2012. While on site for the inspection, the licensee was in the process of loading Cask #18.

The current ISFSI staff was effectively monitoring the loaded casks and keeping the ISFSI program current by monitoring industry activities at other sites and applying lessons learned to the River Bend ISFSI program. Cask loading operations including forced helium dehydration and multipurpose canister (MPC) downloading activities were observed by the NRC inspectors for storage Cask #18. River Bends staff was well trained and thoroughly understood the dry cask storage operations and requirements.

Operation of an ISFSI at Operating Plants (60855.1)

! The licensee completed an audit of the River Bend ISFSI program. The plants Quality Services organization reviewed activities associated with the ISFSI including maintenance, operations, engineering, radiation protection, fuel for loading in the canisters, annual physical inventories, maintenance of records, review of design change packages, submittal of required 10 CFR 72.48 reports to the NRC, performance of daily vent checks, radiological controls around the ISFSI, and field observation of cask loading activities. The audit and surveillance reports presented good information directed at improving organizational performance and providing management with feedback on the status of the various ISFSI related activities (Section 1.2.a).

! Radiological conditions at the ISFSI were evaluated including conducting a survey of the area around the ISFSI. Radiation levels on the dosimeters around the ISFSI pad showed decreasing radiation levels over time. Neutron exposures during cask loading operations continue to be a small portion of the overall dose due to the low heat load of the casks. Person-rem doses to load a cask have continued to improve, with current exposures in the range of 0.2 to 0.4 person-rem/cask. This is consistent with other Holtec cask users (Section 1.2.b).

! An overly conservative assumption in the offsite dose calculations required by 10 CFR 72.104 was found in the River Bend 72.212 Evaluation Report. The calculation used an incorrect value for the contribution to dose by the operating nuclear plant to demonstrate that individuals beyond the controlled area boundary would receive less than 25 mrem/year. Due to this assumption, the 72.212 Evaluation Report established an unnecessary limit for the number of casks that could be stored at the ISFSI (Section 1.2.b).

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! The calculations for the contribution from the casks on the ISFSI pad to the dose to the public required by 10 CFR 72.104 were based on a Holtec report that had assumed certain bounding fuel characteristics. However, a review of the spent fuel stored in Cask

1. 8 determined that the cask contained spent fuel assemblies that were not bounded by the Holtec calculations. The licensee issued Condition Report CR-RBS-2012-6153.

This issue is being treated as an Unresolved Item (Section 1.2.b).

! Storage cask vent inspection requirements of Technical Specification 3.1.2 were performed daily as required (Section 1.2.c).

! Selected condition reports over the past four years were reviewed related to the ISFSI and the fuel building cask handling crane. A wide variety of issues had been documented including generic industry issues that could affect the River Bend ISFSI.

The condition reports reviewed were well documented and properly categorized based on the significance of the issue. Corrective actions taken to resolve the issues were found to be adequate. No significant trends were identified by the NRC inspectors (Section 1.2.d).

! To verify compliance with several of the 10 CFR Part 72 record requirements associated with the loaded casks, the licensee was requested to provide specific records to demonstrate the retrievability of the record. The casks loaded over the last several years appeared to have records that were more readily retrievable than the earlier casks. There was no cross reference established that linked all records to the cask that the record was associated with. At the exit briefing, a discussion was held with senior management representatives concerning the difficulty this may present at a future date when the casks are removed from the site. Condition Report CR-HQN-2012-1111 was issued as an Entergy site wide condition report to address the issue of cask specific record retrievability (Section 1.2.e).

! Helium leak testing of canisters during fabrication had been discontinued for a period of time by the cask vendor, Holtec International. Subsequently, the NRC required Holtec to re-initiate the testing on canisters during fabrication. In the interim, a number of canisters had been shipped to licensees including eight that had been loaded and placed on the River Bend ISFSI pad. NRC review of data provided by Holtec and Entergy concluded that the eight loaded canisters at River Bend were acceptable for continued use (Section 1.2.f).

! Industry related issues identified by the NRC and Holtec technical groups were reviewed for applicability to River Bends ISFSI program and facility. The licensee adequately incorporated resolutions to the issues identified in NRC Information Notice 2011-10 and issues from Holtec Information Bulletins (HIB) 51 and 53 into their program (Section 1.2.f).

Review of 10 CFR 72.212(b) Evaluations (60856.1)

! No changes to the 10 CFR 72.212 report had been made since the last NRC inspection in 2008 (Section 2).

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Review of 10 CFR 72.48 Evaluations (60857)

! All required safety screenings and safety evaluations had been performed in accordance with procedures and 10 CFR 72.48 requirements. All screenings reviewed were determined to be adequately evaluated. No safety evaluations were performed since the last inspection in 2008 (Section 3).

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Report Details Summary of Facility Status River Bends ISFSI currently stored seventeen Holtec 100 casks and was in the process of loading Cask #18 at the time of this inspection. The casks were being monitored in compliance with Technical Specification 3.1.2 for daily vent inspections and for radiation levels in compliance with 10 CFR Part 20. The casks were co-located within the plants Part 50 protected area. Six loading campaigns had been performed at River Bend since starting in 2005. The first seven casks were loaded in three different campaigns from 2005 through 2007.

The licensee then loaded four casks in 2008 and four casks in 2010. During this inspection, the licensee was in the middle of a loading campaign. Two casks had been placed on the pad and two more were planned before the end of the year. The NRC inspectors observed loading operations associated with Cask #18. All loading activities since 2008 had been performed under Certificate of Compliance 1014, Amendment 5 and Holtec Final Safety Analysis Report (FSAR), Revision 7. Previous casks had been loaded to Certificate of Compliance, Amendment 2 and FSAR, Revision 3. A tour of the ISFSI area found the casks to be in good physical condition. Dosimeters along the ISFSI fence were providing radiological dose data within the expected levels for an ISFSI with seventeen casks in storage. The current ISFSI pad can hold 40 casks with four empty areas to allow for shuffling the casks.

1 Operations of an Independent Spent Fuel Storage Installation (ISFSI) at Operating Plants (60855.1)

1.1 Inspection Scope A review of the ISFSI program was conducted which included a variety of topics associated with the loading and storage of the casks at the River Bend Station. This review covered audits and surveillances conducted by the licensee, condition reports related to the ISFSI and the fuel building cask handling crane, environmental radiological data collected around the ISFSI for the past several years, compliance with Technical Specification 3.1.2 for temperature monitoring of the casks, and current issues that relate to the Holtec cask system that have been documented at other ISFSI sites relevant to River Bend for future cask loadings. A tour of the ISFSI area was performed and radiological dose rates measured by the licensee along the complete perimeter of the ISFSI pad. The observed radiation levels were low and consistent with levels expected for a pad containing 17 casks. While on-site, the NRC inspectors observed operations associated the loading of Cask #18 including the forced helium dehydration operations and downloading of the multipurpose canister (MPC) into the HI-STORM storage cask.

1.2 Observations and Findings a. Quality Assurance Audits and Surveillances Entergy recently completed an audit of the River Bend ISFSI program. Audit QA-20-2012-RBS-1 was conducted August 13, 2012 through August 31, 2012 and included observation of activities related to the current cask loading campaign. Organizations audited included maintenance, operations, engineering, and radiation protection. Areas reviewed included selection of fuel for loading in the canisters, annual physical inventories, maintenance of records, review of design change packages, submittal of required 10 CFR 72.48 reports to the NRC, performance of daily vent checks,

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radiological controls around the ISFSI, and field observation of cask loading activities.

Records related to previous cask loadings were reviewed related to welding, non-destructive testing of the welds, equipment calibration, vacuum drying, and leak testing of the canister.

The audit included a review of the process applicability determinations associated with the ISFSI procedures. The process applicability determinations (PAD) are part of the 10 CFR 72.48 screening process. Of the 33 PADs performed, 13 were reviewed by the auditors, of which two were identified, by the auditors, as not being clear as to the potential effect on the design function. Condition Report CR-RBS-2012-5383 was issued concerning two changes to procedures that the quality assurance (QA) auditors questioned. One issue involved the deletion of a precaution of opening the fuel building doors while the in-line fuel transfer system blind flange was removed. The second issue involved a change to procedures to address a stuck/galled port cap on the cask.

Overall, the audit concluded that the ISFSI program was being effectively implemented.

No challenges to the continued performance and the ability to maintain compliance with regulatory requirements were identified.

The audit conducted by QA described above was the first audit specifically directed at the ISFSI. Previous audits had been 10 CFR Part 50 Appendix B audits of programs that were applicable to both Part 72 and Part 50, such as audits of the radiation protection program, maintenance, operations, engineering, etc. These audits, however, only looked at limited issues related to the ISFSI. In July 2011, Condition Report CR-RBS-2011-5322 was issued that recognized that audits conducted of Part 50 programs that overlapped the ISFSI programs may not meet the intent of the 10 CFR 72.176 requirements for ISFSI audits. Since this issue was generic to all of the Entergy sites, Entergy issued a fleet wide Condition Report CR-HQN-2011-0732 to address the need to develop an audit program specific to the ISFSIs at the Entergy sites. As a result of Condition Report CR-HQN-2011-0732, Entergy established a stand-alone ISFSI inspection requirement as Audit 20 in Attachment 9.2 List of Required Audits for Operating Plants to Procedure EN-QV-109 Audit Process, Revision 21. Dry fuel storage inspection audits were required on a 24-month frequency. A standardized audit template was developed as Audit Template QA-20-2012-XXX-01 which provided topics and attributes for inclusion in the ISFSI audits.

Entergy Procedure EN-QV-109 established the requirements for periodic audits at the River Bend site. Attachment 9.2 included Audit 20 ISFSI (10 CFR 72.176) as one of the required audits and referenced Note 7 which stated The audit schedule for stand-alone ISFSIs may combine audits to cover the functional areas listed. Only those areas invoked by the ISFSI technical specifications are required to be audited. The statement that only those areas invoked by the technical specifications are required to be audited was identified by the NRC inspector as a statement that could be misinterpreted. The requirement in 10 CFR 72.176 states that licensees shall carry out a comprehensive system of planned and periodic audits to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program. The program requirements in 10 CFR 72.140(b) states The quality assurance program must cover the activities identified in this subpart throughout the life of the activity. For licensees, this includes activities from site selection through decommissioning prior to termination of the license. The scope of activities covered by the quality assurance program is defined in 10 CFR 72.142(b)(2) which states that the quality assurance functions include verifying by procedures, such as checking, auditing, and inspection, that activities-6- Enclosure

affecting the functions that are important to safety have been correctly performed.

Further descriptions of the areas that fall under the quality assurance program are provided in 10 CFR 72.144. Having a statement in Procedure EN-QV-109 that the audits are limited to only areas related to the technical specifications as opposed to the areas related to important to safety functions and activities could be misinterpreted. The licensee issued Condition Report CR-HQN-2012-1109 to capture this issue and to identify the issue as an Entergy fleet wide condition. In reviewing the audit conducted at River Bend in August 2012, many areas had been included in the audit related to important to safety activities and the audit had not been limited to only the technical specification areas.

b. Radiological Conditions A tour of the ISFSI was completed and the inspectors observed dose rates being measured by the licensee at the radiologically posted area boundary around the casks, which was only a few feet from the casks. Both neutron and gamma readings were taken with readings of less than 1 mrem/hr. Neutron readings were taken with a remball survey instrument. Gamma readings were taken with an RO-20 survey instrument.

Neutron readings were also taken back away from the casks with no indication of skyshine found. The results of a recent radiological survey of the ISFSI conducted on June 14, 2012 were reviewed. The survey provided dose rates at each of the vents on the casks stored in the ISFSI. The highest dose rate on any vent was one mrem/hr contact and 0.6 mrem/hr at 30 cm. Neutron levels were all less than 0.5 mrem/hr.

The radiological environmental reports for 2009, 2010, and 2011 were reviewed which provided thermoluminescent dosimetry (TLD) results around the site. TLD readings for each year did not vary significantly from the control badges, which averaged around 15-16 mrem/quarter. In addition to the environmental TLD program, TLDs were maintained on the fence around the ISFSI, which was 100 - 200 feet from the pad. Two TLDs on each side of the fence provided dose information for the area around the casks. The north side fence was closer to the pad than the south side fence. The pad had been loaded on the east side first. The following TLD data was available:

Table 1: TLD Data for the ISFSI Fence in mrem per 6 Month Period LOCATION 7/09-12/09 1/10-6/10 7/10-12/10 1/11-6/11 Dry Fuel North - 1 99 60 54 43 Dry Fuel North - 2 106 84 77 67 Dry Fuel South - 1 38 17 19 18 Dry Fuel South - 2 26 19 14 10 Dry Fuel East - 1 52 37 31 26 Dry Fuel East - 2 105 51 40 44 Dry Fuel West - 1 78 48 55 43 Dry Fuel West - 2 57 31 21 15 Source: Data Provided by Licensee-7- Enclosure

The time period covered in the table included the loading of four casks between August 12, 2010 and September 27, 2010. The TLDs located closest to the new casks on the west and south side showed slight increases. The other doses showed decreasing values over the 2 year period reflecting the continued radioactive decay of the spent fuel.

The dose to the public from the ISFSI is limited by 10 CFR 72.104 to 25 mrem/year.

This dose limit must include the dose from both the ISFSI and the operating nuclear plant. The 10 CFR 72.212 Evaluation Report, Appendix D - River Bend Station Specific Information included calculations to show compliance with 10 CFR 72.104 in Section D.3.9 72.212(b)(2)(i)(C) - Dose Limitations per 72.104. The section referenced Holtec Report HI-2043196 Dose Versus Distance from a HI-STORM 100S Version B Containing the MPC, Revision 0 which calculated the dose to the nearest public. Holtec had reviewed the various fuel types that were approved for storage in the HI-STORM 100 canister and had established a bounding fuel type that would produce the maximum dose. For boiling water reactors (BWR), this source term was based on a 7x7 BWR fuel assembly with a burnup of 55,000 Megawatt Days/Metric Ton Uranium (MWD/MTU), a minimum cooling time of 3 years, and an enrichment of 4%. The fuel used at River Bend and the HI-STORM 100S and 100S Version B overpacks used at River Bend were considered by the licensee to be bounded by the source term in the Holtec report. River Bend had two HI-STORM 100S overpacks with the rest as HI-STORM 100S, Version B. The controlled area boundary for River Bend was assumed to be 700 meters from the ISFSI for the calculations. Table 2 Annual ISFSI Dose Rate at 700 Meters from the HI-STORM 100S Version B Containing the MPC-68 gave a calculated dose at the controlled area boundary for an occupancy time of 8,760 hours0.0088 days <br />0.211 hours <br />0.00126 weeks <br />2.8918e-4 months <br /> as 11.89 mrem/year. This was based on 40 casks on the pad.

The River Bend 72.212 Evaluation Report, Appendix D, Section D.3.9 referenced the 11.89 mrem/yr value in determining compliance with the 25 mrem/yr in 10 CFR 72.104.

For the contribution from the operating plant dose, a value of 15.22 mR annual mean dose was taken from the 2004 Environmental Operating Report (ADAMS Accession No.

ML051230399). Based on the value of 15.22 mrem/yr, the 72.212 Evaluation Report determined that fully loading the ISFSI pad with 40 casks would result in the 25 mrem/yr limit of 10 CFR 72.104 being exceeded, since 15.22 + 11.89 equaled 27.11 mrem/yr. As such, the 72.212 Evaluation Report estimated that there would be a limit to the number of casks on the ISFSI pad of 30, though it did note that the assumptions used in the calculations were based on much hotter fuel than that actually loaded in the River Bend casks.

On further review of Section D.3.9 of the 72.212 Evaluation Report and the 2004 environmental report, the 15.22 mrem value listed in Table 2.1 of the environmental report as the highest mean value (indicator dosimeter TB1) included background plus contribution from the operating plant and represented a mean quarterly value, not an annual value. The average of all the mean values for the indicator TLDs around the plant was 12.61 mR/quarter. Background dose was determined by the control TLDs and was measured at a mean value of 14.41 mrem/quarter. In reviewing Table 2.1, which listed the mean value for all the TLD locations, only two of the 16 indicator TLD near the plant exceeded the 14.41 mean background TLD value with the other 14 TLDs well below that value, typically in the 10 to 13.5 mrem/quarter range. As such, Section 2.2 of the environmental report concluded that a comparison of the indicator TLDs with the control TLDs indicated that the ambient radiation levels were unaffected by plant-8- Enclosure

operations and levels continued to remain at or near background. Based on this conclusion, the use of the 15.22 mrem value in the 72.212 Evaluation Report as an annual dose contribution from plant operations for determining compliance with 10 CFR 72.104 overestimated the dose to the public and placed an unnecessary restriction on the number of casks that can be stored at the ISFSI. Condition Report CR-RBS-2012-6443 was issued to address the dose calculations.

Review of the information in the Holtec HI-2043196 Report identified a problem with the assumptions used in the spent fuel loaded in the casks. Section 4 Assumptions and the computer files attached to the report used the design basis fuel assembly in the calculations. This assumed 4% enrichment for the spent fuel. However, River Bend had loaded several casks with enrichments below this value, such as Cask #8 where all 68 fuel assemblies were below 2.49% enrichment and Cask #16 with spent fuel assemblies as low as 2.231% enrichment. As fuel enrichments decrease, the dose from the cask increases for the same burnup. This is due to the fact that as the initial enrichment decreases, the fuel is exposed to a larger neutron fluence to achieve the same burnup.

The larger neutron fluence generates larger actinide content which results in a larger neutron source term and secondary gamma source term as illustrated in NUREG/CR-6716 Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks, Section 3.4.1.2 Enrichment. This results in a non-conservative assumption for determining compliance with 10 CFR 72.104 if higher enrichment values are used in the calculations to establish compliance. Entergy Procedure EN-DC-215, Revision 3, which became effective April 17, 2012, had recognized the need to ensure the spent fuel loaded in the casks was consistent with the assumptions in the site boundary dose calculations and identified in Section 5.1(14)(g)

that the minimum enrichment may be restricted at some Entergy sites. Several steps were placed in the current version of the procedure to ensure that the minimum enrichment limitation was verified for the spent fuel selected for storage in a cask. This guidance had not been provided in the procedure for earlier casks such as Casks #8, but was in place for Cask #16 placed on the ISFSI pad on August 25, 2012. This issue was identified to River Bend, who initiated Condition Report CR-RBS-2012-6153. The condition report referenced a 2011 Entergy corporate condition report, CR-HQN-2011-0502 which had evaluated this same issue related to enrichment values for Entergys Vermont Yankee plant. As part of the Vermont Yankee analysis, Entergy performed an extent of condition evaluation to determine if the issue affected other sites. The conclusion of the Entergy condition report was that for Entergy sites that used the Holtec design basis assembly assumptions from the FSAR, such as River Bend, that the fuel associated with the site would be bounded.

A review of the Holtec FSAR, Revision 7, Section 5.2 Source Term was performed to evaluate Entergys conclusion related to River Bend. The FSAR described the process used by Holtec to determine the bounding conditions for fuel loaded in a HI-STORM 100 cask system. Holtec determined that for the BWR fuel approved for loading in the HI-STORM 100 casks, the GE 7x7 fuel assembly, which has the highest Uranium Dioxide content of the BWR assemblies, produces the highest neutron and gamma source terms and the highest decay heat load for a given burn-up and cooling time. FSAR Section 5.2.5.2 BWR Design Basis Assembly provided details for the design basis assembly identified by Holtec as the assembly that bounds the other designs. The second paragraph in this section stated that the initial enrichment used in the analysis is consistent with Table 5.2.24. Table 5.2.24 Initial Enrichments Used in the Source Term Calculations provided a table of various burn-up ranges and the associated enrichment-9- Enclosure

values that were used for the calculations. For the 55,000 to 60,000 MWD/MTU range, the 4% enrichment value can be found consistent with the Holtec offsite dose calculation report (HI-2043196). For the range of 30,000 to 35,000 MWD/MTU, the table lists an enrichment value of 2.6%. However, Cask #8 contained fuel with a maximum 2.492%

enrichment and maximum burn-up of 31,010 MWD/MTU. This value is lower than the enrichment value used in Table 5.2.24 and the Holtec calculations and as such could result in a non-conservative dose calculation. Eleven of the 68 assemblies were not bounded by Table 5.2.24. The burnup values ranged from 30,250 MWD/MTU to 31,010 MWD/MTU with enrichment values for these eleven assemblies ranging from 2.471% to 2.486%.

Failure to recognize the enrichment limit for the casks resulted in at least one cask being loaded that was not bounded by the enrichment values used in the calculations in Holtec Report HI-2043196 and as a result failed to fully demonstrate compliance with 10 CFR 72.104. A written evaluation to demonstrate compliance with the requirements of 10 CFR 72.104 is required by 10 CFR 72.212(b)(5)(iii). Contrary to this, River Bend dose calculations may not be conservative in demonstrating compliance and may not meet the requirements of 10 CFR 72.212(b)(5)(iii). The licensee issued condition report CR-RBS- 2012-6153 to document the issue and address the adequacy of the offsite dose calculation. A telephonic discussion was held between the NRC, Entergy, and Holtec International on October 24, 2012 to further discuss the issue. Holtec and Entergy agreed to perform additional analysis to demonstrate compliance with 10 CFR 72.104 for the casks loaded at River Bend. This has been designated as an Unresolved Item (URI).

River Bend was moving Cask #18 to the ISFSI pad during the week of this inspection.

The final cask (Cask #19) for this loading campaign was planned for the following week.

Attachment 3 provides a list of the casks with the man-rem dose per cask. The first cask loaded incurred 1.028 person-rem cumulative dose. As time progressed, River Bend continued to show progress in reducing the dose. The current loading campaign and the previous 2010 loading campaign were performed within the cumulative dose values similar to other Holtec cask users of 0.2 to 0.4 person-rem. The reduction in dose per cask over time was attributed by the licensee to the use of new resins to clean-up the spent fuel pool water which was causing high background levels in the work area, improved familiarity of the loading operations by the staff over time, strictly limiting access to the cask to those required to perform work, and reducing the time the welders were around the cask by using cameras to observe the welding process.

The contribution from neutron dose to workers was reviewed. When the canister had been drained of water, workers were required to wear neutron dosimetry while around the canister. This included both the CR-39 chip in the dosimeter of official record and an electronic neutron dosimeter. The electronic neutron dosimeter dose readings for the period of July 2010 through December 2010 were reviewed. This period coincided with the loading of Casks #12 - #15. The total of the electronic neutron dosimeter readings during this period was 0.186 rem cumulative exposure among 30 individuals. Not all individuals working around the cask received a neutron reading. The gamma dose during the loading of Casks #12 - #15 was 1.089 rem total for all workers. A review of the dosimeter of official record found no neutron values recorded. This can be contributed to the neutron flux being very low with readings on the electronic dosimeters individually being 1 to 14 mrem and the CR-39 dosimeter of record having a threshold that appears to be around 30 mrem for this energy spectrum. Of the nine individuals

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that each received the highest neutron dose in the range of 9 to 14 mrem for the 2010 loading campaign, which involved four casks, four individuals were radiation protection personnel, three were maintenance, and two were welders. This averaged 2 - 3 mrem/cask/person for these nine individuals that received the highest neutron dose.

The neutron dose was not included in the permanent record for each individual since it was not detected by the dosimeter of official record.

c. Technical Specification 3.1.2 Temperature Monitoring Technical Specification 3.1.2 required either a daily inspection of the inlet and outlet vents for blockage or daily verification that the temperature difference between the HI-STORM outlet temperature and the ISFSI ambient temperature was 126 oF for casks loaded under Amendment 2 (Casks 1-7) and 137 oF for casks loaded under Amendment 5 (Casks 8-19). At River Bend, none of the HI-STORM casks were equipped with temperature monitoring equipment. River Bend performed daily monitoring of the inlet and outlet vents for compliance with Technical Specification 3.1.2.

During normal operations, Surveillance Test Procedure STP-000-0001 Daily Monitoring Logs, Revision 70 was utilized for vent inspections. For periods when the reactor was in cold shutdown, Surveillance Test Procedure STP-000-0004 Daily Cold Shutdown Logs, Revision 41 was used. Documentation was reviewed for the months of December 2008, June 2009, April 2010, October 2011, February 2012, and June 2012 for compliance with the Technical Specification. Of the six months selected for review, vent inspections were performed daily. For all the days of the selected months reviewed, no cask vents were reported as being blocked.

d. Corrective Action Program A list of condition reports issued since the last NRC inspection in 2008 was provided by the licensee for the crane and ISFSI. The condition reports were processed in accordance with Procedure EN-LI-102 Corrective Action Process. Attachment 9.1 Condition Report Classifications/Category provided a description of the safety categorization for the condition reports. Those classified as A or B were considered adverse conditions, with the A condition reports considered significant. Category C were non-significant issues that needed to be investigated and corrected. Category D conditions were minor issues. Selected condition reports were identified by the NRC for further review. The condition reports and a brief description of each are provided as Attachment 2 to this report.

The condition reports reviewed were well documented and properly categorized based on the significance of the issue. The corrective actions taken were appropriate for the situation. No NRC concerns were identified related to the condition reports reviewed.

e. Permanent Record Storage Permanent records related to the dry cask storage program are required by 10 CFR Part 72. Records documenting the spent fuel placed in each cask are required by 10 CFR 72.212(b)(12). Records related to the fabrication of the cask are required by 10 CFR 72.234(d) and (e). Records of written procedures and appropriate tests related to dry fuel storage activities are required by 10 CFR 72.234(f). Records related to compliance with the offsite dose limitations of 10 CFR 72.104 are required by 10 CFR 72.212(b)(7). Records of activities affecting quality are required to be maintained per

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10 CFR 72.174. These are a few of the records specified as required documents in 10 CFR Part 72. River Bend maintained a set of records generated for the casks and used Grand Gulf as the back-up record repository. Procedure EN-DC-160 Dry Fuel Storage Document Control Revision 1 provided instructions for record retention. Cask document records were defined in the procedure along with responsibilities for generating and storing the records. Examples of required records and examples of how to code the records was provided in the procedure.

To verify compliance with several of the Part 72 record requirements, the licensee was requested to provide specific records to demonstrate the retrievability of the record. The casks loaded over the last several years appear to have records that are more readily retrievable than the earlier casks. As such, Cask #1 was selected for review of the records. The work order package, characteristics of the spent fuel loaded in the cask, and Holtec records related to the fabrication of the cask were requested. There was a list of various records associated with the cask that was found, but the list was not all inclusive. The effort to find the records required several people over two days looking for records related to the cask. There was a River Bend specific database of records and an Entergy database. Different records were in the two systems. Only a small number of the records, required by 10 CFR Part 72, were located during the time frame of the search. The conclusion reached was that all required records were probably in the permanent record system but were not easily retrievable in a short time frame. One exception was the enrichment values for the fuel loaded in the casks. These records were found by contacting the licensees reactor engineering group who had the enrichment value for each assembly on their computer. The inspectors were told that these records were backed up at Entergy headquarters on CDs maintained by the reactor engineers that operated the Caskloader software for Entergy. The reactor engineering group also noted that enrichment information was required to be included in the DOE/NRC Form 741 Nuclear Materials Transaction Report that was submitted to the Department of Energy and the NRC whenever new fuel was received onsite and as such, DOE could locate the enrichment values for the fuel albeit with some considerable effort. Though the records may all be in the permanent record system, the effort to prove this would be extensive. In many cases there was no cross reference established for the various records that tied the individual record to the particular cask that the record was associated with. A search was required to locate each record.

At the exit briefing, a discussion was held with senior licensee management concerning the difficulty this may present at a future date when casks are removed from the site. It is expected that not all casks will be removed at the same time and the schedule may spread over many years. As each cask is removed, the records specific to only that cask will be requested. These requests will be well after all those currently working with the records will be gone. Without some type of cross reference, it is likely that providing the complete set of records may not be possible, since the person searching for the records will not have any way of knowing if they have located all the records for the particular cask. Condition Report CR-HQN-2012-1111 was issued as an Entergy site wide condition report to address the issue.

Also noted during the inspection, many of the documents being generated for the dry cask storage project that were required to be submitted to the NRC, such as the 30-day notice of cask loading and the biennial 72.48 report, were being submitted to the NRC Document Control Desk with only the Part 50 license number. Correspondence related to the dry cask storage program should also include the 72-49 ISFSI docket number

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assigned to River Bend. River Bend initiated two condition reports related to the issue, one for River Bend as CR-RBS-2012-6078 and one for Entergy wide as CR-HQN-2012-1129.

f. Industry Issues Impacting River Bends ISFSI Program (1) Thermal Issues During Canister Loading, NRC Information Notice (IN) 2011-10 NRCs Information Notice (IN) 2011-10 Thermal Issues Identified during Loading of Spent Fuel Storage Casks (ADAMS Accession # ML111090200) was distributed to all holders of a Part 72 license on May 2, 2011. The purpose of the notice was to inform the addressees of an incident that occurred during the loading of spent fuel storage canisters at the Byron Generation Station. The NRC expected recipients to review the information for applicability to their facilities and take appropriate actions to avoid similar problems. Bryon, using the HI-STORM 100 system, experienced a malfunction of the canister cooling system during a cask loading. The circulating water system for the annulus area between the canister and transfer cask (annulus cooling), used to keep the fuel cladding temperatures below allowable limits, was found to be inoperable after being left unattended during the night shift. The annulus cooling system was required when loading higher kW canisters using the vacuum drying option. The information notice discussed six potential issues related to the incident, five of which related to vacuum drying. Since River Bend used the forced helium dehydration system instead of vacuum drying, those issues were not applicable. The sixth issue related to the use of nitrogen to blow down the canisters to remove water prior to backfilling with helium and sealing the canister. River Bend used helium for the blow-down and as such this issue was also not applicable. Regardless, River Bend issued condition report CR-HQN-2011-00498 on May 12, 2011 to address one of the five vacuum drying issues as it could relate to forced helium dehydration (FHD) operations. The issue related to preventing or mitigating air ingression into the canister containing fuel, which could cause fuel oxidation, if certain failures of the system occurred, such as a hose rupture. The condition report verified that the existing site procedures for cask loading and unloading contained the appropriate actions to prevent air ingression. The condition report stated that all the River Bend procedures are Continuous Use procedures and as such are required to be manned during operations. The licensees Procedure DFS-0140 Forced Helium Dehydration Operations contained provisions that if the FHD system became inoperable for any reason, the MPC was required to be filled with helium to a pressure of 59 psig to maintain the fuel in an analyzed condition and prevent air ingression.

(2) Helium Leak Testing of Canisters During Fabrication, Holtec Information Bulletin (HIB) - 39 During a 2009 inspection of Holtec International, the NRC identified an issue related to the helium leak testing of the canisters during fabrication. Holtec had modified their FSAR to eliminate the requirement for a shop helium leak rate test after fabrication on the canister confinement boundary welds (shell seam weld and shell-to-baseplate weld). This test was one of the tests to demonstrate the integrity of the canister confinement boundary. The requirement had been in Revision 3 of the FSAR and was removed in Revision 4. The NRC issued a non-cited violation

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to Holtec in a letter dated August 6, 2009 (ADAMS Accession # ML092180140 incorrectly dated as August 5, 2009) entitled Exercise of Enforcement Discretion -

Holtec International. Holtec agreed to re-instate the helium leak test requirement.

Between the time the helium leak testing was discontinued during canister fabrication until the time the testing was re-initiated, several canisters had been manufactured and sent to reactor sites. Holtec informed the affected users of the issue in Holtec Information Bulletin (HIB)-39.

When River Bend became aware of the NRC finding, they issued Condition Report CR-RBS-2009-3386. Eight loaded canisters were affected (MPC Serial Nos. 48, 49, 55, 56, 57, 216, 217, and 218). The heat load of the affected canisters ranged from 9.85 kW to 18.98 kW. Holtec provided a letter to the NRC dated September 2, 2009 (ADAMS Accession # ML092470363) in response to the non-cited violation issued in the August 6, 2009 NRC letter. Holtecs response included analysis that supported continued use of the loaded canisters that had not been leak tested during fabrication.

On November 2, 2010 (ADAMS Accession # ML103090653), Entergy Operations, Inc. provided information to the NRC related to the eight loaded River Bend canisters. The Entergy letter provided information that had been requested by the NRC during a teleconference on December 1, 2009 (ADAM Accession #

ML093510008). Entergy noted that the maximum heat load for the canisters was 18.98 kW which was bounded by the thermal and overpressure helium analysis provided by Holtecs letter to the NRC dated September 2, 2009, that no discernable increase in offsite dose had been detected since the loaded casks had been placed in the ISFSI, and that the non-conforming condition of the canisters had been entered into the site corrective action program and an operability determination had been performed which concluded that the affected canisters continued to perform within their designed safety function. The NRC responded to Entergy by letter dated January 26, 2011 (ADAMS Accession # ML110270139)

stating that the NRC had reviewed the information provided by Holtec and the information provided by Entergy and had determined that the affected canisters currently stored at the River Bend ISFSI were acceptable for continued use. No further actions were necessary.

(3) Determination of Total Heat Load, Holtec Information Bulletin (HIB) - 51 Holtec issued HIB-51 Revision 0, on October 25, 2011 and Revision 1 to the bulletin on December 14, 2011 to the Holtec user group. The bulletin applied to users of CoC Amendment 5 or greater. The bulletin provided an operational issue that was discovered at Tennessee Valley Authority (TVA) by Holtec, when providing technical support to that user. Holtec determined that TVA had calculated the total heat load of their MPCs by using a simple summation of the individual assembly heat loads. This practice was later found to be the case for many Holtec users, including River Bend. Casks #8 through #15 had been loaded to CoC Amendment 5 prior to the issuance of HIB-51.

Summing the individual fuel assemblies heat loads to determine compliance with MPC threshold total heat load limits is not consistent with the HI-STORM 100 FSAR Subsection 2.1.9.1.2. The FSAR subsection contained a discussion on the total MPC heat load (QCoC) and how to calculate it for compliance to various

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Technical Specifications. For uniform loading (which is how River Bend loaded fuel) QCoC is defined as the highest heat load assembly multiplied by the number of locations in a MPC (rmax * n). River Bend uses the MPC-68 with sixty eight fuel storage locations. Holtec stated that the conservative method of calculating total heat load was necessary as otherwise the system designer would have to consider an infinite number of heat load distributions and that would be practically impossible. It was easier for the system designer to assume all locations were generating a heat load of rmax to run the thermal model to ensure peak fuel cladding temperature limits were not exceeded. The subsection 2.1.9.2 method of calculating heat load (QCoC) applied to the helium backfill requirement (TS Table 3-2), the supplemental cooling system requirement (TS 3.1.4), the use of forced helium dehydration requirement (TS Table 3-1), the heat removal system requirement (TS 3.1.2), and vacuum drying time limits requirement (TS 3.1.1).

At the time of the issuance of HIB-51, River Bend had loaded all eight Amendment 5 canisters using the uniform heat load criteria but had calculated each canisters heat load using the simple summation method of summing all the individual assemblys heat loads versus using the method described in FSAR Subsection 2.1.9.1.2. Upon receiving HIB-51 from Holtec, River Bend initiated Condition Report CR-RBS-2011-7738 on October 26, 2011. The condition report documented the issue and re-performed the total heat load calculation using QCoC

= rmax * n for the eight canisters. The condition report documented that the new total heat load values for the eight canisters did not increase to a level that caused a violation of any of the Technical Specifications identified in HIB-51.

Since the issuance of HIB-51, the licensee has revised their fuel selection Procedure EN-DC-215 Fuel Selection for Holtec Dry Cask Storage to Revision 3.

The new revision is now consistent with the requirements described in the FSAR Subsection 2.1.9.1.2.

(4) Isolation of Loaded Canisters, Holtec Information Bulletin (HIB) - 53 Holtec Information Bulletin HIB-53 was issued to the Holtec users group on December 6, 2011. The bulletin described an issue that was observed by NRC inspectors at the Waterford nuclear plant (ADAMS Accession # ML12124A387).

While Waterford was loading their first canister, operators isolated the canister by closing both the vent and drain port caps during installation of the remote valve operating actuators (RVOAs). Having both port caps closed at the same time isolated the canister without having any release path or relief valve available while the canister was filled with water and fuel. This could have pressurized the canister due to the thermal heat of the spent fuel. The Holtec bulletin reminded users that the vent and drain port caps should not be closed simultaneously and that the RVOAs must be installed one at a time in the open position when the canister is filled with water. River Bend addressed this in an Operating Experience (OE) Action Item dated September 28, 2011. The OE evaluation stated that procedures for RVOA installation directed users to install the drain RVOA first and required the drain port to be vented prior to installation of the vent RVOA. This required sequence eliminated the possibility of isolating the canister.

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(5) Stack-Up for Canister Downloading Into the HI-STORM Storage Cask River Bend does not utilize a seismic restraint system when downloading a MPC from the HI-TRAC to the HI-STORM. River Bend performs the stack-up outside the fuel building underneath the cask handling crane structure. The fuel building cask handling crane passes from inside the fuel building to outside the fuel building where the stack-up operations are conducted. The licensee performed a seismic evaluation in Calculation G13.18.1.2-042 and Evaluation 01-E-0012-03 to demonstrate that lateral seismic restraints were not necessary. The licensees fuel building cask handling crane was not a single failure proof crane. The use of this crane for dry cask storage operations was approved by the NRC in Amendment

1. 149 to Facility Operation License No. NPF-47 for the River Bend Station, Unit 1 in a letter entitled River Bend Station, Unit 1 - Issuance of Amendment Re: Use of Fuel Building Cask Handling Crane for Dry Spent Fuel Cask Loading Operations dated December 1, 2005 (Adams Accession No. ML053410490).

The River Bend calculation G13.18.1.2-042 Dry Cask Stack-Up Analysis for River Bend assumed a bolted connection between the HI-TRAC and HI-STORM during stack-up operations. River Bend installs the bolts that span from the HI-TRAC through the mating device, through the spacer, and into the HI-STORM per Procedure DFS-0003 MPC Transfer Operations and HI-STORM Transport. The bolts provide a rigid stack-up body and the analysis demonstrated that the seismic restraints were not needed to ensure the stack-up would not tip over during a seismic event.

Where the stack-up operations take place, there is no safe shutdown equipment located under the outdoor crane truck bay. The stack-up configuration and location was evaluated by the licensee against the same design basis drop criteria as that of an outside cask transfer facility (CTF) allowed by the HI-STORM 100 FSAR.

1.3 Conclusions The licensee completed an audit of the River Bend ISFSI program. The plants Quality Services organization reviewed activities associated with the ISFSI including maintenance, operations, engineering, radiation protection, fuel for loading in the canisters, annual physical inventories, maintenance of records, review of design change packages, submittal of required 10 CFR 72.48 reports to the NRC, performance of daily vent checks, radiological controls around the ISFSI, and field observation of cask loading activities. The audit and surveillance reports presented good information directed at improving organizational performance and providing management with feedback on the status of the various ISFSI related activities.

Radiological conditions at the ISFSI were evaluated including conducting a survey of the area around the ISFSI. Radiation levels on the dosimeters around the ISFSI pad showed decreasing radiation levels over time. Neutron exposures during cask loading operations continue to be a small portion of the overall dose due to the low heat load of the casks. Person-rem doses to load a cask have continued to improve, with current exposures in the range of 0.2 to 0.4 person-rem/cask. This is consistent with other Holtec cask users.

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An overly conservative assumption in the offsite dose calculations required by 10 CFR 72.104 was found in the River Bend 72.212 Evaluation Report. The calculation used an incorrect value for the contribution to dose by the operating nuclear plant to demonstrate that individuals beyond the controlled area boundary would receive less than 25 mrem/year. Due to this assumption, the 72.212 Evaluation Report established an unnecessary limit for the number of casks that could be stored at the ISFSI.

The calculations for the contribution from the casks on the ISFSI pad to the dose to the public required by 10 CFR 72.104 were based on a Holtec report that had assumed certain bounding fuel characteristics. However, a review of the spent fuel stored in Cask

1. 8 determined that the cask contained spent fuel assemblies that were not bounded by the Holtec calculations. Previous experience by NRC Region 4 with this issue has found that the corrected calculations result in a minimal change to the offsite dose. The licensee issued Condition Report CR-RBS-2012-6153. This issue is being treated as an Unresolved Item.

Storage cask vent inspection requirements of Technical Specification 3.1.2 were performed daily as required.

Selected condition reports over the past four years were reviewed related to the ISFSI and the fuel building cask handling crane. A wide variety of issues had been documented including generic industry issues that could affect the River Bend ISFSI.

The condition reports reviewed were well documented and properly categorized based on the significance of the issue. Corrective actions taken to resolve the issues were found to be adequate. No significant trends were identified by the NRC inspectors.

To verify compliance with several of the 10 CFR Part 72 record requirements associated with the loaded casks, the licensee was requested to provide specific records to demonstrate the retrievability of the record. The casks loaded over the last several years appeared to have records that were more readily retrievable than the earlier casks. There was no cross reference established that linked all records to the cask that the record was associated with. At the exit briefing, a discussion was held with senior management representatives concerning the difficulty this may present at a future date when the casks are removed from the site. Condition Report CR-HQN-2012-1111 was issued as an Entergy site wide condition report to address the issue of cask specific record retrievability.

Helium leak testing of canisters during fabrication had been discontinued for a period of time by the cask vendor, Holtec International. Subsequently, the NRC required Holtec to re-initiate the testing on canisters during fabrication. In the interim, a number of canisters had been shipped to licensees including eight that had been loaded and placed on the River Bend ISFSI pad. NRC review of data provided by Holtec and Entergy concluded that the eight loaded canisters at River Bend were acceptable for continued use.

Industry related issues identified by the NRC and Holtec technical groups were reviewed for applicability to River Bends ISFSI program and facility. The licensee adequately incorporated resolutions to the issues identified in NRC Information Notice 2011-10 and issues from Holtec Information Bulletins (HIB) 51 and 53 into their program.

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2 Review of 10 CFR 72.212(b) Evaluations at Operating Plants (60856.1)

2.1 Inspection Scope The 10 CFR 72.212 Evaluation Report was reviewed to verify site characteristics were still bounded by the Holtec HI-STORM 100 design basis.

2.2 Observations and Findings The Entergy 10 CFR 72.212 Evaluation Report was currently Revision 7. No changes had been made since the last NRC inspection in 2008. In addition to the generic 72.212 Evaluation Report, River Bend specific information was included in an Appendix D. The appendix was currently Revision 2 and had not been revised since the last inspection.

2.3 Conclusions No changes to the 10 CFR 72.212 report had been made since the last NRC inspection in 2008.

3 Review of 10 CFR 72.48 Evaluations (60857)

3.1 Inspection Scope The licensees 10 CFR 72.48 screenings and evaluations since the 2008 NRC inspection were reviewed to determine compliance with regulatory requirements.

3.2 Observations and Findings A list of modifications to the ISFSI program and changes to the fuel building crane was provided by the licensee. Four engineering changes (EC) that required a 10 CFR 72.48 screening or 10 CFR 50.59 screening for the crane were selected for reviewed. The licensee utilized Procedure EN-LI-100 Process Applicability Determination, Revision 11 to perform the 10 CFR 72.48 safety screenings. The procedure contained the required questions in compliance with the 10 CFR 72.48 regulations to determine if an evaluation was warranted. If any of the safety screenings required an evaluation, the licensee utilized procedure EN-LI-112 10 CFR 72.48 Evaluations, Revision 9. None of the screenings required a full 10 CFR 72.48 safety evaluation. The issues discussed in the screenings are summarized in the following table. All screenings were determined to be adequately evaluated.

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Table 2: Safety Screening Screen # and Date Description This engineering change restricted movement of the fuel building cask crane during loaded cask moves. The EC # 00143 remote and manual controls of the crane were changed to prevent the movement in more than one plane of Installation of cask motion during loaded cask movement. During loaded crane electrical cask movement, only horizontal (main trolley) or only interlocks.

vertical (the main hoist) movement of the crane would be possible at any given time due to interlocks installed.

08/27/2008 This engineering change was screened as not requiring a safety evaluation.

EC # 12944 The engineering change qualified the auxiliary crane to lift floor plugs that weighed 6500 lbs due to a Evaluation of the modification of installing an oil drip pan. Additional lifting 15 ton auxiliary by the 15 ton crane would require engineering approval, crane on the spent unless a full rated load test was performed to re-qualify fuel building crane the crane to the rated capacity due to the drip pan modification. This engineering change was screened as 01/30/2009 not requiring a safety evaluation.

This engineering change provided the design and EC # 00262 installation requirements for a 28 ft. wide reinforced concrete slab that was constructed adjacent to the west Installation of an side of the existing ISFSI pad. This slab had a depth of apron adjacent to 18 in. and extended 4 ft. beyond each edge of the ISFSI the ISFSI pad.

pad to facilitate the turning of the transporter. This engineering change was screened as not requiring a 02/03/2009 safety evaluation.

EC # 36818 Engineering Change (EC)-32683 repaired eroded Evaluation of ISFSI portions of the ISFSI pad slope and installed new slope repairs erosion protection over these areas. EC-36818 provided performed under drawing mark-ups and final configuration of the slope EC-32683. and its erosion protection. This engineering change was screened as not requiring a safety evaluation.

04/24/2012 Source: Screenings Selected by NRC from a List of Plant Modifications Provided by the Licensee The licensee is required by 10 CFR 72.48(d)(2) to submit a report to the NRC containing a brief description of any changes, tests, and experiments at an interval not exceeding 24 months. Reports were reviewed back to December 2005. The last report was for the period of June 30, 2010 through June 30, 2012. No 10 CFR 72.48 evaluations were reported for the period from December 29, 2005 through June 30, 2012.

3.3 Conclusions All required safety screenings and safety evaluations had been performed in accordance with procedures and 10 CFR 72.48 requirements. All screenings reviewed were determined to be adequately evaluated. No safety evaluations were performed since the

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last inspection in 2008.

4 Exit Meeting The inspector reviewed the scope and findings of the inspection during an exit conducted on September 27, 2012.

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ATTACHMENT 1:

SUPPLEMENTAL INSPECTION INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee Personnel J. Blair, Reactor Engineer J. Boulanger, Manager, Maintenance G. Bush, Manager, MP&C J. Campbell, Senior Project Manager J. Clark, Manager, Licensing P. Ellis, Radiation Protection Technician R. Gadbois, General Manager, Operations G. Hackett, Manager, Radiation Protection J. Houseman, Reactor Engineer K. Huffstatler, Sr. Licensing Specialist B. Mashburn, Director, Engineering M. Mella, Reactor Engineer E. Olson, Vice President, Operations B. Raines, Admin Services, Supervisor J. Roberts, Director, Nuclear Safety Assurance J. Soileau, Health Physics Specialist J. Vukovies, Supervisor, Reactor Engineering L. Woods, Manager, Quality Assurance D. Yoes, QA Supervisor J. Yokovics, Reactor Engineering INSPECTION PROCEDURES USED IP 60855.1 Operations of an ISFSIs at Operating Plants IP 60856.1 Review of 10 CFR 72.212(b) Evaluations at Operating Plants IP 60857 Review of 10 CFR 72.48 Evaluations LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 72-49/1201-01 URI Fuel assemblies minimum enrichment is not bounded by offsite dose calculation required by 10 CFR 72.104 Discussed None Closed None Attachment 1

LIST OF ACRONYMS ANSI American National Standards Institute BWR Boiling Water Reactor CFR Code of Federal Regulations CoC Certificate of Compliance CR condition report CTF Cask Transfer Facility Delta T delta temperature F Fahrenheit FSAR Final Safety Analysis Report HIB Holtec information bulletin IP inspection procedure ISFSI Independent Spent Fuel Storage Installation kW killo-watt MPC multi-purpose canister mR milliRoentgen micro(µ)R microRoentgen MPC multipurpose canister mrem milliRoentgen equivalent man MWD/MTU megawatt days/metric ton uranium NRC Nuclear Regulatory Commission OE operational experience psi pounds per square inch psig pounds per square inch gauge RBS River Bend Station RVOA remote valve operating actuators TLD thermoluminescent dosimeter-2- Attachment 1

ATTACHMENT 2:

CONDITION REPORTS REVIEWED DURING THIS INSPECTION Condition Severity Description Report Level INPO issued a Significant Event Notification SER3-08 which discussed several industry issues that could affect ISFSIs.

2008-6602 C These related to fuel selection, cask cooling, and hydrogen gas monitoring. As a result, procedure changes and training were initiated at River Bend to prevent occurrence of the issues.

During lowering of the canister lid into the spent fuel pool, an unexpected noise was heard coming from the crane hoist. The 2008-6730 D lift was stopped and the crane placed in a safe configuration.

The vendor was contacted to provide service to the crane.

While performing an inspection of the fuel building bridge crane, the load block sheaves bearings and pin were found to have 2009-0108 D excessive wear. The recommended action was to replace the pin and bearings. The condition report was closed to a work order.

This condition report documents the Holtec helium leak test 2009-3386 C issue related to Holtec Information Bulletin (HIB)-39. This issue is discussed in this inspection report in Section 1.2 g (2).

During loading of the spent fuel into Cask #12, the channel fastener leaf spring was bent while loading the fuel assembly into cell 47. The problem was attributed to tight cell tolerances and channel bowing. The affected assembly was removed and 2010-3435 C placed back into its original spent fuel pool location. An alternate assembly was selected for placement into the canister and the problem occurred again on the alternate assembly. The orientation of the fuel assembly during insertion was changed and the third assembly was successfully inserted.

The fuel grapple fits very tightly in the canister basket and frequently snags on the top edges of the basket resulting in multiple attempts being required to seat the spent fuel assemblies. Options reviewed included modifying the grapple, use of space plates in the cell, and changing the orientation of 2010-3449 C the assembly in the cell if release of the assembly was difficult.

If the problem was complicated by bowing of the assembly, an alternate assembly could be used. Engineering recognized that the cells were very tight and insertion of the assembly was sometimes difficult and required centering the assembly carefully. No modifications were made to the grapple.

During the loading of Cask #12, a channel fastener leaf spring was bent while inserting a spent fuel assembly into cell 63. The spent fuel assembly was returned to its original spent fuel pool 2010-3494 C location and an alternate assembly selected. The orientation of the alternate assembly was changed and the assembly was successfully inserted.

Attachment 2

Condition Severity Description Report Level During loading of Canister #14, a channel fastener leaf spring was bent while inserting the spent fuel assembly into cell 14.

The assembly was returned to its original spent fuel pool location and an alternate assembly selected. The orientation during insertion was changed and the assembly successfully inserted. This condition report also documented that during the 2010-4359 C loading of Cask #13, the fuel handlers had identified a bent channel fastener clip (leaf spring) on fuel assembly GGE204 during the removal of the assembly from its spent fuel pool location. Reactor Engineering was aware that the assembly already had a bent clip. All the assemblies with bent channel fastener leaf springs during the dry loading campaign have been GGE fuel assemblies.

During the placement of the vent port cap on Cask #14 after drying was completed, the cap became stuck. The canister had been dried and backfilled with helium. A monitoring plan was established to continue to re-circulate helium in the canister using the forced helium dehydration system and take readings every three hours to ensure the pressure remained above 30 psig. Temperature monitoring of the inlet and outlet temperatures was also performed at the same time the pressure readings were taken. Inlet and outlet temperatures were around 120 to 130 degree F. Pressures were typically maintained around 49 to 50 psig. The re-circulation process continued for several days. Procedure DFS-0140 MPC Forced Helium Dehydration Operations, Revision 3, Step 8.6.11.4 stated that if the vent port cap will not fully close or becomes stuck, go to 2010-4411 C Attachment 4 Removal of a Stuck/Galled MPC Port Cap. The licensee implemented Attachment 4 which involved depressurizing the canister through the plant gaseous waste system, establishing a minimum helium flow through the canister at less than 1 psig, removing the removable valve operator assembly (RVOA), placing a loose fitting foreign material exclusion (FME) plug on the vent tube, and replacing and tightening the vent port cap. The cap material was Nitronic 60. Cask # 14 (14.87 kW) was loaded during the 2010 loading campaign which included four casks. The dose to load Cask

1. 14 was 0.281 person-rem. The average for the four casks loaded in 2010 was 0.272 and ranged from 0.199 to 0.362 person-rem. The effort to deal with the stuck vent port cap did not result in a significant dose contribution.

This condition report was issued to address the human and equipment performance issues related to the bent channel fastener springs encountered during the 2010 loading campaign 2010-4552 B (This issue relates to CR-2010-3435, CR-2010-3494, and CR-2010-4359). An apparent cause evaluation was performed.

During the 2010 loading campaign, four General Electric GE9 fuel assemblies had their channel fastener leaf springs bent-2- Attachment 2

Condition Severity Description Report Level during insertion when the bottom edge of the spring made contact with the corner edge of the basket cell. In addition, a fifth channel fastener was identified to have one spring that protruded out and was considered to be at risk of damage during insertion. The fuel assemblies were returned to their original spent fuel pool locations and alternate assemblies selected. The insertion procedure was revised to change the orientation of the assembly during insertion and all alternate assemblies were successfully inserted. Possible causes of the problem included spring protrusion, canister cell design, and channel bowing. Inspection of the springs prior to insertion using an underwater camera was identified as the best method to prevent recurrence.

During insertion of the spent fuel assembly in cell 58 of Cask

1. 15, the fuel assembly could not be fully disengaged. One side of the fuel grapple would release but the other side would not.

The problem was attributed to the tight tolerance of the cell. A 2010-4705 D 3/8 spacer was placed in the bottom of cell 58, which provided additional height to the fuel assembly and provided additional room for the grapple to disengage. The spacer had originally been designed in case problems with disengaging the grapple occurred.

This was an Entergy headquarters condition report related to ISFSI audits and stated that 10 CFR 72.176 audits were not being conducted as required. The issue was a result of an industry evaluation conducted at Vermont Yankee. Prior to this time, audits conducted of the Part 50 programs were credited to 2011-0732 B meeting the 72.176 requirement in that the same program elements were being audited. However, ISFSI specific aspects of the Part 50 programs were not necessarily being audited. As a result, audit procedure EN-QV-109 was revised to establish the requirement for stand-alone ISFSI audits on a 24-month cycle at the Entergy sites.

An inspection conducted of the Perry nuclear plant by NRC Region 3 identified an issue concerning the stackup of the HI-TRAC transfer cask on the HI-STORM storage cask during lowering of the loaded canister. The issue related to seismic 2011-2951 C stability of the stack-up configuration. River Bend evaluated their stack-up configuration and determined that no lateral seismic restraints were required. This issue is discussed in this inspection report in Section 1.2 g (5).

Periodic audits required by 10 CFR 72.176 should be included in the master audit plan or as a stand-alone audit scope. A 2011-5322 D determination was made to conduct the ISFSI audits as stand-alone audits. This issue was closed to the headquarters condition report CR-HQN-2011-0732.

-3- Attachment 2

Condition Severity Description Report Level Erosion of the shotcrete on the west slope of the dry fuel storage pad was occurring. Evaluation by a geotechnical firm 2011-6233 D determined that the area near the ISFSI pad was still stable.

The slope was repaired.

This issue dealt with the method of calculating the cask heat load to determine compliance with the Certificate of 2011-7738 C Compliance. Holtec issued Holtec Information Bulletin (HIB)-51 concerning this problem. This issue is discussed in this inspection report in Section 1.2 g (3).

Records identifying the alternate fuel assemblies that had been placed in casks MPC Serial No. 295, 296, 297, and 304 had not been provided to administrative services or updated in the 2011-7739 C permanent files. This was discovered while reviewing the cask loading records in response to HIB-51. The permanent records were updated.

The cask document records for the last loading campaign had not been submitted to Echelon Administrative Services for 2011-8546 C processing as required by Procedure EN-DC-160. The holdup was due to the records not receiving final signature. The issue was resolved.

During the inspection of vendor supplied rigging equipment used to upend and lift the canister, the shackle pin holes for two lifting lugs on the lifting frame were observed to be oval as opposed to being round. Measurements were taken and provided to Holtec for evaluation. Holtec determined that the 2012-1854 C ovalization was within acceptable tolerances. As a precaution, the licensee performed a nondestructive examination (NDE) of the lifting lugs load bearing surfaces to verify no surface indications were present. The results were satisfactory.

Ongoing monitoring of the lugs was implemented.

While the spent fuel cask handling crane was outside the fuel building moving the lift yoke and being operated by remote control, the main hoist breaker tripped. The following day during functional testing of the crane, the crane stopped during trolley 2012-4303 D movement outside the fuel building with no load on the hook.

The crane was being operated from the main cab. The problem was identified as an electrical coil in the trolley control cabinet.

The crane was reset and operated properly.

During a quality assurance audit, the auditor identified that the Process Applicability Determination (which relates to the 2012-5383 C required 72.48 review) was not clear as to the potential effect on the design function concerning changes made to two loading procedures.

During the drying of Cask #17, the forced helium dehydration (FHD) skid tripped due to loss of electrical power to the unit.

2012-5711 D The system had been operating for approximately 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> using helium to dry the canister. The FHD operator pressurized the canister to 59 psig helium and isolated the canister from the-4- Attachment 2

Condition Severity Description Report Level FHD skid in accordance with Procedure DFS-0140 MPC Forced Helium Dehydration Operations, Revision 3, Step 6.35.2. The FHD operator continued to monitor the canister pressure until station electricians reset the breaker.

-5- Attachment 2

ATTACHMENT 3:

LOADED CASKS AT THE RIVER BEND NUCLEAR STATION ISFSI LOADING MPC HI-STORM DATE HEAT LOAD BURNUP MAXIMUM FUEL PERSON-REM ORDER SERIAL No. No. ON PAD (kW) MWd/MTU (max) ENRICHMENT % DOSE Serial No. Serial No.

1 12/29/05 16.55 39,391 3.34 1.028 43 26 Serial No. Serial No.

2 06/22/06 16.99 39,402 3.33 1.036 26 25 Serial No. Serial No.

3 07/19/06 19.16 41,203 3.35 0.845 32 31 Serial No. Serial No.

4 07/20/07 18.03 41,568 3.35 0.901 48 34 Serial No. Serial No.

5 08/01/07 17.43 40,412 3.35 0.570 49 58 Serial No. Serial No.

6 08/17/07 18.76 41,886 3.35 0.580 55 60 Serial No. Serial No.

7 11/15/07 18.98 41,109 3.35 0.637 56 52 Serial No. Serial No.

8 11/06/08 9.85 31,010 2.49 0.720 57 236 Serial No. Serial No.

9 11/21/08 12.34 34,446 3.01 0.477 216 313 Serial No. Serial No.

10 12/09/08 12.32 33,670 3.07 0.504 217 314 Serial No. Serial No.

11 12/16/08 15.97 40,496 3.36 0.503 218 315 Serial No. Serial No.

12 08/12/10 14.81 38,983 3.54 0.362 295 442 Serial No. Serial No.

13 08/26/10 14.76 38,973 3.54 0.246 296 443 Serial No. Serial No.

14 09/15/10 14.87 40,944 3.54 0.281 297 444 Serial No. Serial No.

15 09/27/10 14.85 39,320 3.54 0.199 304 460 Serial No. Serial No.

16 08/25/12 17.51 44,253 3.88 0.307 375 602 Attachment 3

LOADING MPC HI-STORM DATE HEAT LOAD BURNUP MAXIMUM FUEL PERSON-REM ORDER SERIAL No. No. ON PAD (kW) MWd/MTU (max) ENRICHMENT % DOSE Serial No. Serial No.

17 09/14/12 17.37 43,604 3.87 0.225 376 603 Serial No. Serial No.

18 09//28/12 17.40 43,608 3.88 0.215 377 604 Serial No. Serial No.

19 10/10/12 17.34 43, 480 3.88 0.213 378 605 NOTES: Heat load (kW) is the sum of the heat load values for all spent fuel assemblies in the cask Burn-up is the value for the spent fuel assembly with the highest individual discharge burn-up Fuel enrichment is the spent fuel assembly with the highest individual initial enrichment per cent of U-235 Casks 1 - 7 were loaded under Certificate of Compliance, Amendment 2; Holtec Final Safety Analysis Report, Revision 3 Casks 8 - 19 were loaded under Certificate of Compliance, Amendment 5; Holtec Final Safety Analysis Report, Revision 7-2- Attachment 3