IR 05000382/2014010

IR 05000382-14-010 and 07200075-14-001; on 06/24-26/2014; Waterford Steam Electric Station, Unit 3 and Independent Spent Fuel Storage Installation (Isfsi); Routine ISFSI Inspection Report

ML14211A671
Person / Time
Site:	Waterford
Issue date:	07/29/2014
From:	Keller R NRC/RGN-IV/DNMS/RSFSB
To:	Chisum M Entergy Operations
Keller R
References
IR-14-010
Download: ML14211A671 (21)

Text

July 29, 2014

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 AND INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) INSPECTION REPORT 05000382/2014010 AND 07200075/2014001

Dear Mr. Chisum:

This letter refers to a routine inspection conducted on June 24-26, 2014, of the dry cask storage activities associated with your Independent Spent Fuel Storage Installation (ISFSI).

The enclosed inspection report documents the inspection results which were discussed on June 26, 2014 with you and members of your staff. After additional in-office inspection, a final telephonic exit meeting was conducted on July 24, 2014 with Leia Milster, Regulatory Assurance.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspection reviewed compliance with the requirements specified in the Technical Specifications associated with Holtec International HI-STORM 100 Certificate of Compliance 1014, the HI-STORM 100 Final Safety Analysis Report (FSAR), and Title 10 of the Code of Federal Regulations (CFR) Part 72, Part 50, and Part 20. Within these areas, the inspection included a review of radiation safety, cask thermal monitoring, quality assurance (QA), your corrective action program, safety evaluations, and changes made to your ISFSI program since the last routine ISFSI inspection that was conducted by the U.S. Nuclear Regulatory Commission (NRC). The ISFSI facility was found to be in good physical condition.

NRC identified one finding of low safety significance during this inspection. This finding was determined to involve a violation of NRC requirements. The NRC is treating the violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.

If you contest the non-cited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Waterford Power Plant.

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E LAMAR BLVD ARLINGTON, TX 76011-4511 In accordance with title 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 Agencywide Document Access Management 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.

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/

Ray L. Kellar, P.E., Chief Repository & Spent Fuel Safety Branch Division of Nuclear Materials Safety

Dockets No.: 05000382, 07200075 Licenses No.: NPF-38

Enclosure:

Inspection Report 05000382/2014010 and 07200075/2014001

w/attachments:

1. Supplemental Information 2. Loaded Casks at the Waterford ISFSI

SUMMARY OF FINDINGS

IR 05000382/2014010; and 07200075/2014001; 06/24-26/2014; Waterford Steam Electric

Station, Unit 3 and Independent Spent Fuel Storage Installation (ISFSI); Routine ISFSI Inspection Report

The report covers an announced inspection by one regional inspector and one inspector-in-training. The significance of any Part 50 findings are indicated by their color (Green, White,

Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process. The cross-cutting aspect is determined using IMC 0310, Components Within the Cross-Cutting Areas. Findings for which the significance determination process does not apply may be Green or be assigned a severity level after the NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006. In accordance with the NRC Enforcement Policy, all of the Part 72 ISFSI inspection findings follow the traditional enforcement process and are not dispositioned through the Reactor Oversight Process.

NRC-Identified Findings and Self-Revealing Findings

SL-IV. The inspectors identified a Severity Level IV non-cited violation (NCV) of 10 CFR 72.150 for the licensees failure to prescribe activities affecting quality with appropriate procedures. Specifically, Waterford Procedure DFS-003-006 Stack-Up and Transfer of Loaded MPC, Revision 5, did not include appropriate instructions specified in the Holtec FSAR Section 8.1.6.3 affecting lift height requirements. FSAR Section 8.1.6.3 required that when using a non-single-failure-proof crane to place the HI-TRAC lid on top of the HI-TRAC, the lid shall be kept less than two feet above the top of the loaded Multi-Purpose Canister (MPC) to protect the MPC lid from a potential HI-TRAC lid drop. Waterford personnel had consistently used the non-single-failure-proof auxiliary crane to lift the HI-TRAC lid without regard to the two foot lift height restriction. This finding was more than minor because exceeding the two foot lift height restriction could create a potential of more than minor safety consequence in the event that the non-single-failure-proof auxiliary crane had a failure.

Because the licensee entered the issue into their corrective action program as condition report CR 2014-03571 this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy. Since this was a traditional enforcement violation under their Part 72 license, no cross-cutting aspects have been assigned to this violation.

Licensee-Identified Violations

None.

PLANT AND ISFSI STATUS

Waterford Steam Electric Station, Unit 3 Independent Spent Fuel Storage Installation (ISFSI)stored thirteen loaded Holtec HI-STORM 100S Version B casks at the time of the routine inspection. The licensee utilized a general Part 72 license in accordance with the Holtec HI-STORM 100 System, approved under Certificate of Compliance 1014, License Amendment 5 and Final Safety Analysis Report (FSAR), Revision 7. The version of the Holtec systems used by the licensee included the MPC-32, a 32 fuel bundle multi-purpose canister (MPC), designed to hold 32 pressurized water reactor (PWR) fuel assemblies. The ISFSI consisted of one concrete pad that could accommodate 72 casks. The storage casks were located inside the Part 50 facilitys protected area (PA).

REPORT DETAILS

4OA5 Other Activities

.1 Operations of an Independent Spent Fuel Storage Installation at Operating Plants

(60855.1)

a. Inspection Scope

(1) Quality Assurance Audits and Surveillances

An on-site review of the Quality Assurance (QA) audit and surveillance reports related to dry cask storage activities at the Waterford Steam Electric Station, Unit 3 ISFSI was performed by NRC inspectors. Since the last NRC inspection in November 2011, the licensee had issued one QA audit report, three ISFSI related QA surveillance reports, and numerous QA observations documented in Oversight Observation Checklist (O2C) reports.

The QA audit report, QA-20-2012-W3-1, evaluated six elements of the licensees ISFSI operations: Licensing, Design, Fuel Selection, Campaign, Operations Surveillances, and Corrective Actions.

NRC reviewed three QA surveillance documents. The first two surveillances were follow-up surveillances performed to track the progress of corrective actions from the QA finding documented in Audit Report, QA-20-2012-W3-1. The third QA surveillance was a roll-up surveillance performed to assess nuclear oversight activities related to the ISFSI/Dry Fuel Storage Project from February 2010 through November 2012. The O2C reports documented observations made in ISFSI areas of maintenance, engineering, operations, and radiation protection.

The QA audit report and surveillances resulted in one QA finding and nine condition reports (CR) that were entered into the licensees corrective action program. NRC inspectors reviewed the corrective actions related to the finding and CRs to ensure that the identified deficiencies were properly categorized based on their safety significance and properly resolved. All identified deficiencies had been properly categorized and resolved by the licensee.

(2) Radiological Conditions Related to Stored Casks

The ISFSI was located approximately 900 feet southwest of the reactor building within the plant protected area (PA). The pad was roughly 120 feet wide and 154 feet long with the capacity to hold 72 HI-STORM 100S spent fuel storage casks configured in an 8 by 9 array. The ISFSI was protected on all sides by an exclusion area fence, which was properly posted: Caution Radioactive Materials Area. No flammable or combustible materials, debris, or notable vegetative growth were observed inside the ISFSI area. Thirteen casks were loaded with spent fuel at the time of the inspection. The last ISFSI loading campaign added four casks to the pad in 2013. The inspectors found all thirteen HI-STORM 100S casks to be in good physical condition.

Radiological conditions at the ISFSI were determined from the most recent monthly radiological survey and records from three years of optically stimulated luminescence dosimeter (OSLD) monitoring results. There were four OSLD monitoring locations that were roughly centered on the four sides of the ISFSI exclusion area fence.

An ISFSI supervisor and radiation protection (RP) technician accompanied the NRC inspectors during a walk-down of the ISFSI pad area. A radiological survey was performed by the RP technician with an ion-chamber to record gamma exposure rates in milli-roentgens per hour (mR1/h). The RP technician also carried a survey meter that measured neutron exposure in units of dose equivalent, millirem per hour (mrem/h). The NRC inspector carried a Ludlum Model 19 scintillation detector that was capable of measuring low level gamma radiation exposure rates in micro-roentgen per hour (µR/h). Survey measurements were taken at ISFSI exclusion area fence locations, around the perimeter of the ISFSI pad, at selected areas between casks, and at HI-STORM lower vent locations.

General area gamma background readings outside the power plant prior to entry into the ISFSI pad area were 5 µR/h. The 13 storage casks were situated along the eastern edge and southeastern corner of the ISFSI pad. Radiation readings taken along the fence-line showed radiation levels ranging from 28 µR/h, at distant corner locations, to 180 µR/h on the east fence-line near the storage casks. General area measurements on the pad ranged from 60 µR/h to 2 mR/h. Several cask lower vent measurements were taken. The vent measurements ranged from 1.4 to 2.8 mR/h gamma and 0.8 to 1.2 mrem/h neutron. The measurements recorded by the NRC inspector and RP technician confirmed the most recent monthly ISFSI site survey results. The radiological conditions of the ISFSI were as expected for the age and heat-load of the 13 currently loaded spent fuel storage casks. The licensee was properly posting and controlling the ISFSI pad area consistent with 10 CFR Part 20 requirements.

1 For the purposes of making comparisons between NRC regulations based on dose-equivalent and measurements made in Roentgens, it may be assumed that one Roentgen equals one rem. (http://www.nrc.gov/about-nrc/radiation/protects-you/hppos/qa96.html)

The licensee provided personnel dose information associated with the ten casks loaded during the previous two ISFSI loading campaigns performed in 2012 and 2013. The loading of the last ten casks presented worker doses that ranged from 0.039 to 0.172 person-rem per cask, averaging 0.106 person-rem (shown in 2 of this report). The casks loaded during the 2013 loading campaign showed the overall lowest doses, averaging 0.048 person-rem. The low doses were due to a combination of relatively low heat load of spent fuel being loaded and good dose management practices implemented by the licensee. Personnel radiation doses were estimated using electronic personnel dosimeters (EPDs) that were sensitive to both gamma and neutron radiation. The neutron dose estimates provided by the EPDs did not exceed the neutron dose threshold associated with the OSLDs worn by personnel and were not recorded on the personnels dose of legal record (DLR).

(3) Environmental Radiological Monitoring Program

The primary purpose of the Waterford Radiological Environmental Monitoring Program (REMP) was to evaluate the radiological impacts that reactor operations and stored radioactive materials may have on the local environment. The REMP was focused on measuring airborne (gaseous and particulate), liquid effluent, and direct radiation levels onsite, at the site boundary, and at offsite locations. By design, there were no airborne or liquid effluents released from the Waterford ISFSI.

However, the ISFSI did contribute to the measurable direct radiation levels produced on site.

The REMP monitors direct radiation impacts using thermoluminescent dosimeter (TLD) monitoring. The ISFSI is monitored using OSLDs. TLDs and OSLDs report in the same units and are considered to be functionally equivalent. The OSLD data for the ISFSI and the TLD data for the REMP were handled by two different programs for different purposes. The ISFSI OSLD monitoring was performed by the RP department to track radiological exposures at the ISFSI boundary to ensure that NRC occupational limits for unmonitored individuals were not exceeded. The REMP monitoring was performed to provide an annual assessment of Waterfords total impacts to the environment and includes the contributions from the ISFSI.

OSLDs were placed at four monitoring locations on the fence around the ISFSI. The monitoring results from 2011, 2012, and 2013 for those locations were reviewed by NRC inspectors (see Table 1, below). The OSLDs were collected and replaced yearly.

Table 1 OSLD Monitoring Results for Waterford ISFSI in mrem/yr OSLD #

Location 2011 2012 2013 West Fence

69 South Fence 209 209 East Fence 926 939 North Fence 149 147

NRC Inspectors verified the radiation exposure rates at each OSLD monitoring location during the ISFSI walk-down. The highest reported dose rate, 939 mrem per year, was measured at the ISFSI east fence monitoring location. Applying a occupational occupancy factor of 2.3 X 10-1 (2,000 work hours per year onsite divided by 8,760 total hours in a year) to that dose yielded an occupational value of 214 mrem. The accessible boundaries of the ISFSI were below the CFR 20.1502(a)(1) limit, which is 500 mrem per year.

The yearly results of the REMP were issued in an Annual Radiological Environmental Operating Report (AREOR). The NRC reviewed the Waterford AREORs for 2011 (ML12122A016), issued April 30, 2012; 2012 (ML13123A030),issued April 30, 2013; and 2013 (ML14120A312), issued April 28, 2014. The Waterford AREORs did not include the OSLD direct radiation monitoring results for the ISFSI, but provided reporting and analyses for the REMP TLD monitoring results.

The ISFSI was located in the southwest (SW) REMP monitoring sector. Three out of the 30 total REMP TLD monitoring sites were at site boundary locations in reasonably close proximity to the ISFSI: K-1, 1.1 miles from the reactor in the south-southwest (SSW) sector; L-1, 1.1 miles from the reactor in the SW sector; and M-1, 0.8 miles from the reactor, in the west-southwest (WSW) sector. M-14 was a control monitoring station located 18 miles north-northwest of Waterford. The M-14 dosimeter was treated as the control TLD for radiological background purposes.

Table 2, TLD Monitoring Results near Waterford ISFSI in mrem/yr TLD#

Station and Location 2011 2012 2013 K-1 SSW Sector, 1.1 miles from reactor

L-1 SW Sector, 1.1 miles from reactor

M-1 WSW Sector, 0.8 miles from reactor

M-14 (Control)

NNW Sector, 18 miles from reactor

All REMP TLD monitoring results in closest proximity to the ISFSI were at near background (control levels). The most elevated readings were from the L-1 monitoring location for all years reviewed. Those values, with background subtracted, were 14-19 mrem per year net gamma doses near the site boundary.

Those values were below the 10 CRF 72.104(a)(2) requirement of less than 25 mrem annual dose equivalent to any real individual located beyond the site controlled area. It should be noted that the radiological values for TLD monitoring location L-1 were essentially the same in 2010 prior to the initial placement of loaded casks in the Waterford ISFSI. Therefore, the radiological influence of the ISFSI at the site boundary locations was minimal.

(4) Records Related to Fuel Stored in the Casks

A site review of dry fuel storage records for the last ten casks loaded by the licensee was performed to determine whether adequate descriptions of the spent fuel was documented as a permanent record as required by 10 CFR 72.212(b)(12). The spent fuel contents of the ten most recently loaded HI-STORM casks were recorded in a document titled Cask Load Composite Report. These reports included MPC loading maps, fuel assembly qualification data, assembly identification, decay heat (kW), cooling time (years), initial assembly average U-235 enrichment (%), burn-up values (MWd/MTU), and other information. A complete set of forms were reviewed for each of the ten MPCs. Some of that fuel data is tracked along with other information on Attachment 2 of this inspection report. The licensee was in compliance with all applicable license and FSAR requirements for fuel stored in at the ISFSI and met all regulatory requirements for storage of spent fuel records.

(5) Technical Specification 3.1.2, Cask 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 ISFSI ambient temperature was 155 degrees F for all casks loaded under Certificate of Compliance 1014, Amendment 5. The vent inspections were performed using Procedure OP-903-001, Attachment 11.21 Dry Fuel Storage Technical Specification Surveillance Daily Logs, Revisions 50-53. The licensee performed daily vent verifications to ensure that the inlet and outlet vents were clear. NRC reviewed documentation of these surveillances for three randomly selected months: December 2012, March 2013, and June of 2013.

Of the months selected for review, the technical specification surveillance requirement was performed as required. No cask vents were reported as being blocked in any of the surveillances reviewed.

(6) Corrective Action Program

A list of CRs issued since the last NRC inspection in November of 2011 was provided by the licensee for the fuel handling crane and the ISFSI. Issues were processed in accordance with Procedure EN-LI-102, Corrective Action Process, Revision 23. When a problem was identified the licensee documented the issue as a CR in the licensees corrective action program (CAP).

Of the list of CRs provided relating to the ISFSI and the fuel building crane, approximately 40 CRs were selected by the NRC inspectors for further review. The CRs related to a variety of issues. The CRs reviewed were well documented and properly categorized based on the significance of the issue. The corrective actions taken were appropriate for the situations. Based on the level of detail of the corrective action reports, the licensee demonstrated a high attention to detail in regard to the maintenance and operation of their ISFSI program and the fuel building crane. No NRC concerns were identified related to the CRs reviewed.

(7) Heavy Loads Program

A review of licensees heavy loads procedures and operations were reviewed to ensure the site complied with requirements listed in the Holtec FSAR. One finding was identified during this review, which is discussed below (Section

b. Findings

).

Documents were reviewed that demonstrated that the cask handling crane was inspected on an annual basis in accordance with American Society of Mechanical Engineers (ASME) B30.2, Overhead and Gantry Cranes. The licensee utilized Procedure MM-004-401,Fuel Handling Building Crane Maintenance Mechanical, Revision 7, Procedure ME-004-525, Fuel Handling Crane Maintenance Electrical and Work Order (WO) 52430191, Inspect Crane Per MM and ME Procedures, dated April 10, 2013 to fulfill this requirement.

(8) HI-STORM 100 Cask Yearly Maintenance

Holtec Certificate of Compliance (CoC), 1014 FSAR, Section 9.2, Maintenance Program, specified the HI-STORM maintenance schedule in Table 9.2.1. Among other tasks, the schedule called for annual visual inspection of the storage casks external surfaces and identification markings for signs of damage or degradation.

NRC inspectors reviewed the documentation related to the annual visual examination of the licensees HI-STORM casks for 2012 and 2013. Those documents included the work orders and filled out copies of Procedure DFS-007-001, Attachment 12.1, Annual External Surface Examination on Accessible Surfaces of the HI-STORM 100S Overpack, Revision 0. Those visual examination checklists included 15 points of inspection that were verified as satisfactory, unsatisfactory, or not applicable (N/A). For the years reviewed, no instances of unsatisfactory conditions were identified. Minor weathering of the casks surface finish was noted and two casks had minor scratches that required touch-up paint. Work orders were initiated to make those repairs. The NRC inspectors determined that the licensees yearly maintenance activities and records met the requirements of the FSAR Section 9.2.

b. Findings

• Failure to Perform Heavy Lifts in Accordance with FSAR Requirements

Introduction:

The inspectors identified a Severity Level IV NCV of 10 CFR 72.150 for the licensees failure to prescribe activities affecting quality with appropriate procedures that included lift height restrictions as specified in Holtec FSAR Section 8.1.6.3.

Description:

On October 18, 2013, the licensee identified that a violation of the Part 50 Technical Specification (TS) 3.9.7 requirement prohibiting the use of a non-single-failure-proof load handling system to lift a load greater than 2,000 pounds over irradiated fuel assemblies in the Fuel Handling Building had occurred. Technical Specification 3.9.7, Limiting Conditions of Operations, required loads in excess of 2,000 pounds to be prohibited from travel over irradiated fuel assemblies in the Fuel Handling Building, except over assemblies in a transfer cask using the single-failure-proof handling system.

The licensee had been using the non-single-failure-proof auxiliary hoist to place the HI-TRAC lid, weighing 2,500 pounds, over the irradiated fuel located in the dry fuel canister.

The licensee initiated Licensee Event Report (LER) 2013-007-00 and submitted the report to the NRC (ML13353A136).

The licensee had been using the Fuel Handling Building auxiliary crane hoist to perform placement of the HI-TRAC lid (approximately 2,500 pounds) over the loaded transfer cask. The auxiliary crane hoist is not a single-failure-proof, but has the rated capacity of 30,000 pounds (15 tons). The rigging equipment that was utilized in the evolution had a lifting capacity of 32,000 pounds exceeding NUREG 0612 recommended safety factors for the lift. Additionally, the spent fuel was shielded inside the MPC by a 9.5 inch thick lid welded in place which would have provided protection to the spent fuel from the HI-TRAC lid if a drop were to have occurred.

The Holtec FSAR does allow the use of a non-single-failure-proof crane or hoist to place the HI-TRAC lid on the loaded transfer cask. However, the Holtec FSAR placed restrictions on this lifting operation. Holtec FSAR Section 8.1.6.3 required that when using a non-single-failure-proof crane to lift the HI-TRAC lid, the lid shall be kept less than 2 feet above the top of the MPC to protect the MPC lid from a potential HI-TRAC lid drop. Waterford Procedure DFS-003-006, Stack-Up and Transfer of Loaded MPC, Revision 5 did not include the Holtec FSAR Section 8.1.6.3 minimum lift height requirement. Licensee personnel had consistently used the non-single-failure-proof crane hoist to lift the HI-TRAC lid without regard for the two foot lift height restriction.

Analysis:

Failure to provide sufficient and appropriate procedures to not exceed the two foot lift height restriction when utilizing a non-single-failure-proof crane, as required by Holtec FSAR Section 8.1.6.3 was identified by inspectors as a violation of 10 CFR 72.150 requirements. Consistent with guidance in Section 2.2 of the NRC Enforcement Policy, ISFSIs are not subject to the significance determination process and, thus, traditional enforcement was used. Therefore, the violation was dispositioned per the traditional enforcement process using Section 2.3 of the Enforcement Policy.

This finding was determined by inspectors to be more than minor because exceeding the two foot lift height restriction could create a potential of more than minor safety consequence in the event that the non-single-failure-proof auxiliary crane hoist had a failure. Since traditional enforcement was used to disposition the violation, there is not a cross-cutting aspect.

Enforcement:

Title 10 CFR 72.150, Instructions, Procedures, and Drawings, states, in part, that licensees shall prescribe activities affecting quality by documented procedures of a type appropriate to the circumstances and shall require that these procedures be followed. The procedures must include appropriate quantitative criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, the licensee failed to prescribe activities affecting quality by documented procedures of a type appropriate to the circumstances and require that these procedures be followed.

Specifically, Waterford Procedure DFS-003-006 Stack-Up and Transfer of Loaded MPC, Revision 5 did not include the Holtec FSAR Section 8.1.6.3 requirement to limit the height of the HI-TRAC lid to two feet above the top of the MPC when using the non-single-failure-proof crane hoist. Upon identification, the licensee entered the issue into their corrective action program as CR 2014-03571. This condition was evaluated as having low safety significance since the auxiliary crane hoist and rigging had an adequate capacity to lift the load, including exceeding the minimum recommended safety factors in NUREG 0612. Because the licensee entered the issue into their corrective action program, the safety significance of the issue was low, and the issue was not repetitive or willful, this Severity Level IV violation was treated as a NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy.

NCV 07200075/201401-01, Failure to have quality procedures required by 10 CFR 72.150 which contained lift height requirements.

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

a. Inspection Scope

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

The licensees 10 CFR 72.212 Evaluation Report at the time of the inspection was Revision 3, dated January 9, 2014. Two revisions had been performed to the 72.212 Evaluation Report since the last NRC routine ISFSI inspection.

The two revisions to the 72.212 Report included changes to reflect the casks that were loaded and placed on the ISFSI pad from the loading campaigns performed in 2012 and 2013. The other changes that made in the 72.212 report included changes made to Table 5.10.1, Holtec-Implemented Deviations from HI-STORM FSAR Revision 7. This table listed the Holtec approved deviations from the FSAR. Each deviation listed in the table contained the associated Holtec Engineering Change Order (ECO) number. The Holtec ECOs contained a description of the change and contained a 72.48 evaluation that had been performed by Holtec. The inspectors reviewed the ECOs and 72.48 screenings that Holtec had performed which had been added to Waterfords 72.212 Evaluation Report in Revisions 2 and 3. The ECOs were for items such as: change of the contour finish of the non-structural weld surface on MPC split lid, editorial changes for various component drawings, changes of various welds from fillet weld to groove weld of equal strength performed during fabrication of a HI-STORM 100S Version B, etc.

The changes associated with Revision 2 and Revision 3 of the 72.212 Evaluation Report did not cause site characteristics or loading operations to be no longer bounded by the HI-STORM cask systems design basis.

b. Findings

No findings were identified.

.3 Review of 10 CFR 72.48 Evaluations

a. Inspection Scope

The licensees 10 CFR 72.48 screenings and evaluations since the last NRC routine ISFSI inspection were reviewed to determine compliance with regulatory requirements.

A list of modifications to the ISFSI programs and changes to the fuel building crane were provided by the licensee. The licensee had not performed any 72.48 ISFSI evaluations or any 50.59 fuel building crane evaluations since the last NRC inspection in November of 2011. The licensee identified that only one 72.48 screening had been performed other than 72.48 screenings conducted for procedure changes.

On November 21, 2011, during stack-up and downloading operations of MPC Serial # 0155 into HI-STORM Serial # 0464 from the HI-TRAC, the Mating Device was damaged and a small piece of Teflon and the heads of four 1/4 socket head screws fell under the MPC. A visual inspection was performed of the annulus between the HI-STORM and the MPC which did not reveal any of the missing items from the Mating Device. The licensee initiated Engineering Change (EC) 33277 and a 72.48 screen to accept the condition as is for continued use and storage at the ISFSI pad. The licensee contacted Holtec and received Response to Request for Technical Information (RRTI)1874-003 dated November 21, 2011. The Holtec evaluation reviewed Normal Storage Conditions, Off-Normal Storage Conditions, Accident Conditions of Storage, Thermal, Shielding, Criticality, Confinement, and Operations. EC 33277 concluded that the existence of the foreign material beneath the MPC did not have an adverse impact on the design functions or methods of performing or controlling the design functions of the HI-STORM system as described in the FSAR. The HI-STORM was acceptable for storage at Waterfords ISFSI. The 72.48 screening reviewed was determined to have been adequately evaluated.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On June 26, 2014, the inspectors presented the inspection results to Mr. Mike Chisum, Site Vice President Operations, and other members of the licensee staff. After additional in-office inspection, a final telephonic exit meeting was conducted on July 24, 2014 with Leia Milster, Regulatory Assurance. The licensee acknowledged the inspection details presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

SUPPLEMENTAL INSPECTION INFORMATION KEY POINTS OF CONTACT

Licensee Personnel

S. Anders, Maintenance Mods Supervisor K. Crissman, Maintenance Manager G. Ferguson, Project Manager J. Laque, Project Manager L. Milster, Regulatory Assurance L. Murray, Project Manager C. Rich, Regulatory Assurance Director

INSPECTION PROCEDURES USED

IP 60855.1

Operations of an ISFSI 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

72-075/201401-01 NCV Failure to have quality procedures required by 10 CFR 72.150 which contained lift height requirements.

Discussed

None

Closed

72-075/201401-01 NCV Failure to have quality procedures required by 10 CFR 72.150 which contained lift height requirements.

LIST OF

DOCUMENTS REVIEWED

The following is a list of documents reviewed during the inspection. Inclusion on this list does

not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that

selected sections of portions of the documents were evaluated as part of the overall inspection

effort. Inclusion of a document on this list does not imply NRC acceptance of the document or

any part of it, unless this is stated in the body of the inspection report.

4OA5.1 Other Activities

Drawings

NUMBER

TITLE

DATE

HP-SM-109

WF3 Survey Data Template - Blank

N/A

HP-SM-307

Dry Fuel Cask Storage Pad Template - Blank

N/A

HP-SM-307

Dry Fuel Cask Storage Pad Survey

05/27/14

HP-SM-307

Dry Fuel Cask Storage Pad Survey

03/31/14

HP-SM-307

Dry Fuel Cask Storage Pad Survey

10/26/13

Procedures

NUMBER

TITLE

REVISION

EN-LI-102

Corrective Action Process

Rev. 23

EN-LI-100

Process Applicability Determination

Rev. 13

EN-LI-112

CFR 72.48 Evaluations

Rev. 10

MM-004-401

Fuel Handling Building Crane Maintenance

Mechanical

Rev. 7

ME-004-525

Fuel Handling Crane Maintenance Electrical

Rev. 9

EN-DC-215

Fuel Selection for Holtec Dry Cask Storage

Rev. 5

Waterford 3 ISFSI/Dry Fuel Storage Project Quality

Plan

Rev. 0

OP-903-001

11.21 (numerous)

Revs. 50-53

DFS-007-001

Annual External Surface Examination on Accessible

Surfaces of the HI-STORM (numerous)

Rev. 0

Design Basis Documents

NUMBER

TITLE

REVISION

Waterford Steam Electric Station Unit 3 10 CFR 2.212 Report Utilizing the HI-STORM 100 System

Rev. 3 and Rev. 2

Certificate of Compliance 72-1014 HI-STORM 100

Cask System

Amendment 5

Holtec International FSAR for the HI-STORM 100

Cask System

Rev. 7

- 3 -

Miscellaneous Documents

NUMBER

TITLE

REVISION / DATE

RRTI 1874-003

Report for Request for Technical Information

11/21/11

EN-QV-109

Audit Process, Attachment 9.1

Rev. 21

QA-20-12-W3-1

QA Audit Report

09/25/12

QS-2012-W3-20

QA Follow-up Surveillance of the 2012 ISFSI Audit,

QA-20-2012-W3-1

2/01/12

QS-2012-W3-004

QA Second Follow-up Surveillance of the 2012 ISFSI

Audit, QA-20-2012-W3-1

2/19/12

QS-2012-W3-03

ISFSI/Dry Fuel Storage Project Roll-up Surveillance

01/30/13

Attendance List, Entrance Meeting, ISFSI Audit, QA-

20-2012-W3-1

08/13/12

Attendance List, Exit Meeting, ISFSI Audit, QA-20-

2012-W3-1

09/25/12

Cask Loading Dose/Task # Spreadsheet

N/A

Audit Notification Memorandum

08/06/12

QA ISFAI Program Audit Plan

08/06/12

Standardized Audit Template Checklist

N/A

OSLD Monitoring Data for ISFSI Pad

N/A

Cask Load Composite Report (numerous)

Oversight Observation Checklists

O2C-W3-2010-0048

O2C-W3-2010-0108

O2C-W3-2010-0122

O2C-W3-2010-0141

O2C-W3-2010-0143

O2C-W3-2010-0162

O2C-W3-2010-0183

O2C-W3-2010-0244

O2C-W3-2010-0048

O2C-W3-2010-0108

O2C-W3-2011-0012

O2C-W3-2011-0013

O2C-W3-2011-0174

O2C-W3-2011-0175

O2C-W3-2011-0177

O2C-W3-2011-0199

O2C-W3-2011-0200

O2C-W3-2011-0214

O2C-W3-2011-0215

O2C-W3-2012-0048

O2C-W3-2012-0060

O2C-W3-2012-0061

O2C-W3-2012-0145

O2C-W3-2012-0156

O2C-W3-2012-0157

O2C-W3-2012-0170

O2C-W3-2012-0495

Engineering Changes and Engineering Change Orders

EC-33277

ECO-152

ECO-065

ECO-064

ECO-063

ECO-148

ECO-146

ECO-151

ECO-066

Work Orders

WO 52430191

WO00343956

WO52475858

- 4 -

Condition Reports

CR 12-03507

CR 12-02812

CR 12-01560

CR 13-03723

CR 13-03226

CR 13-05520

CR 12-01562

CR 13-04455

CR 13-05106

CR 12-00487

CR 12-01726

CR 13-04498

CR 13-03571

CR 12-00793

CR 12-01994

CR 13-04852

CR 12-00172

CR 12-00983

CR 12-02063

CR 13-05154

CR 13-02848

CR 12-01003

CR 12-04350

CR 13-05520

CR 12-03994

CR 12-01084

CR 13-03314

CR 14-03272

CR 12-04380

CR 12-01152

CR 13-03571

CR 12-01637

CR 12-04381

CR 12-01414 CR 13-03591

CR 12-07253

- 5 -

LIST OF ACRONYMS

ADAMS

Agencywide Documents Access and Management System

AREOR

Annual Radiological Environmental Operating Report

ASME

American Society of Mechanical Engineers

CAP

Corrective Action Program

CoC

Certificate of Compliance

CR

Condition Report

CFR

Code of Federal Regulations

DLR

Dose of Legal Record

DNMS

Division of Nuclear Material Safety

EC

Engineering Change

ECO

Engineering Change Order

EPD

Electronic Personnel Dosimeter

FSAR

Final Safety Analysis Report

IMC

Inspection Manual Chapter

ISFSI

Independent Spent Fuel Storage Installation

kW

kilo-watt

mR/h

milliroentgen per hour

mrem/h

millirem per hour

µR/h

microroentgen per hour

MPC

Multi-Purpose Canister

MWd/MTU

megawatt days per metric ton uranium

NCV

Non-cited Violation

NRC

U.S. Nuclear Regulatory Commission

PA

Protected Area

PWR

Pressurized Water Reactor

O2C

Oversight Observation Checklist

OSLD

Optically Stimulated Luminescence Dosimeter

QA

Quality Assurance

REMP

Radiological Environmental Monitoring Program

RP

radiation protection

RRTI

Response to Request for Technical Information

TLD

thermoluminescent dosimeter

ATTACHMENT 2:

LOADED CASKS AT THE WATERFORD ISFSI

LOADING

ORDER

MPC

SERIAL No.

HI-

STORM

No.

DATE

ON PAD

HEAT LOAD

(kW)

BURNUP

MWd/MTU (max)

MAXIMUM FUEL

ENRICHMENT %

PERSON-REM

DOSE

MPC-32-154

DFS-463

11/08/11

14.549

44,327

4.601

0.188

MPC-32-153

DFS-462

11/15/11

15.413

2,766

4.59

0.155

MPC-32-155

DFS-464

11/21/11

16.373

43,039

4.61

0.184

MPC-32-225

DFS-599

03/25/12

19.785

44,783

4.614

0.171

MPC-32-222

DFS-598

04/03/12

19.432

44,960

4.591

0.168

MPC-32-226

DFS-597

04/10/12

19.617

44,910

4.613

0.173

MPC-32-223

DFS-596

04/21/12

19.350

44,964

4.614

0.141

MPC-32-221

DFS-600

04/28/12

19.185

44,905

4.609

0.117

MPC-32-224

DFS-601

05/04/12

18.292

44,950

4.602

0.102

MPC-32-297

DFS-738

10/08/13

13.381

2,837

3.951

0.063

MPC-32-294

DFS-735

10/14/13

13.263

2,776

3.948

0.047

MPC-32-295

DFS-736

10/19/13

13.010

2,442

3.942

0.042

MPC-32-296

DFS-737

10/26/13

2.411

39,632

3.955

0.039

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 average initial enrichment per cent of U-235

Casks 1 - 13 were loaded under Certificate of Compliance, Amendment 5; Holtec FSAR, Revision 7