IR 05000416/2014008
| ML14184B561 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 07/02/2014 |
| From: | Ray Kellar NRC/RGN-IV/DNMS/RSFSB |
| To: | Kevin Mulligan Entergy Operations |
| Kellar R | |
| References | |
| IR-14-008 | |
| Download: ML14184B561 (21) | |
Text
July 2, 2014
SUBJECT:
GRAND GULF NUCLEAR STATION, UNIT 1 AND INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) INSPECTION REPORT 05000416/2014008 AND 07200050/2014001
Dear Mr. Mulligan:
This letter refers to a routine inspection conducted on June 3-5, 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 5, 2014, with you and members of your staff.
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, 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).
Your ISFSI operations were found to be in compliance with the applicable NRC regulations and your general Part 72 license requirements. The ISFSI facility was found to be in good physical condition. No violations of NRC regulations 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 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.
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION IV
1600 E. LAMAR BLVD.
ARLINGTON, TX 76011-4511 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.: 05000416, 07200050 Licenses No.: NPF-29
Enclosure:
Inspection Report 05000416/2014008 and 07200050/2014001
w/attachments:
1. Supplemental Information 2. Loaded Casks at the Grand Gulf ISFSI
ML14184B561 X SUNSI Review By: LEB ADAMS X Yes No XPublicly Available Non-Publicly Available X Non-Sensitive Sensitive Keyword:
OFFICE RSFSrs RSFS RSFS:C
NAME LEBrookhart EJSimpson RLKellar
SIGNATURE
/RA/
/RA/
/RA/
DATE 07/02/14 07/02/14 07/02/14
Letter to from R. Kellar, dated July 2, 2014.
SUBJECT:
GRAND GULF NUCLEAR STATION, UNIT 1 AND INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) INSPECTION REPORT 05000416/2014008 AND 07200050/2014001
DISTRIBUTION:
Regional Administrator (Marc.Dapas@nrc.gov)
Deputy Regional Administrator (Kriss.Kennedy@nrc.gov)
Acting DNMS Director (Linda.Howell@nrc.gov)
Acting DRP Director (Troy.Pruett@nrc.gov)
Acting DRP Deputy (Michael.Hay@nrc.gov)
RSFS Branch Chief (Ray.Kellar@nrc.gov)
DRP/C Branch Chief (Don.Allen@nrc.gov)
Senior Resident Inspector (Blake.Rice@nrc.gov)
Senior Project Engineer, DRP/C (Ray.Azua@nrc.gov)
Project Engineer, DRP/C (Paul.Nizov@nrc.gov)
GG Administrative Assistant (Alley.Farrell@nrc.gov)
RSFS Inspector (Lee.Brookhart@nrc.gov)
RSFS Inspector (Eric.Simpson@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)
RIV/ETA: OEDO (Yen.Ju-Chen@nrc.gov OEMail_Resources@nrc.gov ROPreports
Enclosure U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Dockets:
050-00416; 072-00050
Licenses:
Report Nos.:
05000416/2014008; 07200050/2014001
Licensee:
Entergy Operations, Inc.
Facility:
Grand Gulf Nuclear Station and
Independent Spent Fuel Storage Installation (ISFSI)
Location:
P.O. Box 756
Port Gibson, MS 39150
Dates:
June 3-5, 2014
Inspector:
Lee Brookhart, Senior ISFSI Inspector
Repository & Spent Fuel Safety Branch
Accompanying Personnel:
Eric Simpson, Health Physicist, Inspector-in-Training,
Repository & Spent Fuel Safety Branch
Approved By:
Ray L. Kellar, P. E., Chief
Repository & Spent Fuel Safety Branch
Division of Nuclear Materials Safety
- 2 -
SUMMARY OF FINDINGS
IR 05000416/2014008; and 07200050/2014001; 06/03-06/05/2014; Grand Gulf Nuclear Station and Independent Spent Fuel Storage Installation; Routine (ISFSI) Inspection Report
The report covers an announced inspection by one regional inspector and one inspector-in-training. No findings or violations associated with Nuclear Regulatory Commission (NRC)regulations were identified. 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 disposition through the Reactor Oversight Process or the Significance Determination Process.
NRC-Identified Findings and Self-Revealing Findings
No findings were identified.
Licensee-Identified Violations
None.
PLANT AND ISFSI STATUS
Grand Gulf Nuclear Stations (GGNS) Independent Spent Fuel Storage Installation (ISFSI)stored twenty-three loaded Holtec HI-STORM 100S Version B casks at the time of the routine inspection. GGNS 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 at GGNS included the MPC-68, a 68 fuel bundle multi-purpose canister (MPC), designed to hold 68 boiling water reactor (BWR) fuel assemblies. The ISFSI at GGNS consisted of one concrete pad that could accommodate 44 casks with provisions for four additional spaces to allow for cask unloading and shuffling purposes. The storage casks were located inside the Part 50 facilitys protected area (PA). To date, the licensee has loaded only intact spent fuel assemblies and has maintained the total decay heat load of each cask to less than 20 kilowatts (kW) except for one cask. This strategy minimizes the thermal heat levels of the canister and the radiation levels, to facilitate a better working environment during loading, decontamination, and welding operations.
REPORT DETAILS
OTHER ACTIVITIES
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 Quality Assurance (QA) audit reports and surveillances related to dry cask storage activities at the GGNS was performed by the NRC inspectors.
Since the last inspection in August of 2012, Entergy had issued one QA audit report and three ISFSI related QA surveillance reports, known internally as Oversight Observation Checklists.
During the QA audit, six elements of Grand Gulfs ISFSI operations were evaluated by Entergy: Licensing, Design, Fuel Selection, Cask Loading Campaign, Operations Surveillances, and Corrective Actions. Fourteen condition reports (CRs) were generated from the QA audit. The three QA surveillances documented observations made during the dry fuel movement evolutions and follow-up activities resulting from ISFSI CRs. This resulted in one CR being written on the dry fuel movement activities.
All fifteen CRs associated with the QA audit report and surveillances were classified as not adverse to quality and appropriately entered into the licensees corrective action program (CAP). NRC inspectors reviewed these 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 Grand Gulf Nuclear Station ISFSI was located approximately 900 feet northeast of the reactor building within the protected area (PA). The ISFSI site consisted of one concrete pad, 61 feet wide by 196 feet in length with the capacity to hold 44 HI-STORM 100S spent fuel storage casks in a 4 by 11 array.
During a tour of the ISFSI, the inspectors observed that no flammable or combustible materials were inside the ISFSI controlled area. In addition, there were no debris or notable vegetative growth observed within the ISFSI area. Twenty-three out of a planned capacity of 44 casks were loaded on the pad at the time of the inspection.
The last ISFSI loading campaign loaded six casks in 2013. The inspectors found the 23 loaded HI-STORM 100S casks to be in good physical condition.
Radiological conditions at the GGNS ISFSI were documented in the most recently performed quarterly radiological survey and semi-annual optically stimulated luminescence (OSL) monitoring results for the ISFSI. A licensee radiation protection (RP) technician accompanied the NRC inspectors during the walk-down of the ISFSI pad to perform confirmatory measurements of the radiological conditions.
Survey measurements were taken around the perimeter of the ISFSI pad, at selected boundary locations, and at various locations on the inner ISFSI fence by the RP technician using a neutron survey instrument and the NRC inspector using a Ludlum Model 19 scintillation detector.
General area background readings around the ISFSI were measured at 0.02 mR/h gamma and < 0.01 mrem/h neutron. The dose rates near the inner ISFSI fence ranged from 0.05 mR/h to 0.23 mR/h gamma and were <0.01 mrem/h neutron. The area immediately outside of the loaded casks was roped off and properly posted with signage that read, Caution Radioactive Materials Area. The roped boundary area measured from 0.31 to 1.00 mR/h gamma and 0.01 to 0.02 mrem/h neutron. The measurements recorded by the NRC inspector and RP technician confirmed quarterly site survey measurements. The radiological conditions of the ISFSI were as expected for the age, burn-up levels, and heat-load of the 23 currently loaded spent fuel storage casks.
The licensee provided personnel dose information associated with the six casks loaded during the last ISFSI loading campaign in 2013. The loading of those six casks presented worker doses that ranged from 0.098 to 0.172 person-rem per cask, as shown in Attachment 2. The neutron dose rates for these cask loading activities showed that 19-31 percent of the total worker dose was due to neutron exposure.
The highest contributor to radiological worker dose (estimated to be 40-66%)
occurred during transfer cask decontamination and MPC lid welding activities.
- (3) Environmental Radiological Monitoring Program
The primary purpose of the Grand Gulf Nuclear Station Radiological Environmental Monitoring Program (REMP) was to evaluate the radiological impacts that the reactor site and ISFSI may have on the 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 Grand Gulf ISFSI. NRC reviewed the Spent Fuel Storage Radioactive Effluent Release Reports for 2012 (ML13042A454) and 2013 (ML14058B181) for Grand Gulf, which confirmed that the Grand Gulf ISFSI did not produce any effluent radioactive releases during 2012 or 2013.
The yearly results of the REMP are issued in an Annual Radiological Environmental Operating Report (AREOR). NRC reviewed the Grand Gulf AREORs for 2012 (ML13121A394), issued May 1, 2013, and 2013 (ML14118A172), issued April 25, 2014. The Grand Gulf AREORs did not include any additional direct radiation monitoring data in close proximity of the ISFSI. However, the NRC was provided with some direct monitoring data that had been collected by the staff.
The Grand Gulf ISFSI was located approximately 0.2 miles from the reactor building in the in the NE monitoring sector. Five of the 29 total REMP thermoluminescent dosimeter (TLD) monitoring sites were at site boundary locations in close proximity to the Grand Gulf ISFSI: M-16, 0.9 miles from the reactor in the N sector; M-23, 0.5 miles from the reactor in the NW sector; M-94, 0.8 miles from the reactor, in the NNW sector; M-96, 0.7 miles from the reactor in the NNE sector; and M-100, 0.6 miles from the reactor in the NE sector.
Table 1, Site TLD Monitoring Results Closest to GG ISFSI in mrem/yr TLD#
Station and Location 2012 2013 M-16 North Sector, 0.9 miles from reactor 43.3 42.5 M-23 Northwest Sector, 0.5 miles from reactor 34.0 34.2 M-94 North Northwest Sector, 0.8 miles from reactor 42.7 42.0 M-96 North Northeast Sector, 0.7 miles from reactor 27.6 30.4 M-100 Northeast Sector, 0.6 miles from reactor 45.5 45.8 M-14 (Control)
North Northeast, 18 miles from reactor 44.1 43.4
All TLD monitoring results in closest proximity to the Grand Gulf ISFSI (near the sites owner controlled boundaries) were at near background (control levels). The most elevated readings were from the M-100 monitoring location in both 2012 and 2013. Those values, with background subtracted, were 1.4 and 2.4 mrem/yr net gamma dose near the site boundary (see Table 1, above). Those values were well below the 10 CRF 72.104(a)(2) requirement of less than 25 mrem annual dose to any real individual located beyond the owner controlled area.
- (4) Records Related to Fuel Stored in the Casks
A site review of dry fuel storage records for the last six casks loaded at the Grand Gulf ISFSI 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 six most recently loaded HI-STORM 100S casks were recorded on form EN-DC-215, Attachment 9.2, Fuel Assembly Qualification for BWR Fuel. The fuel assembly qualification data included Multi-Purpose Canister loading maps and individual fuel assembly parameters, including type, 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 was reviewed for each of the six casks. Some of the fuel data is tracked along with other information on Attachment 2 to this inspection report. Grand Gulf was in compliance all applicable Technical Specifications for fuel stored at their ISFSI and all regulatory requirements for fuel storage records.
- (5) Technical Specification 3.1.2, Cask Temperature Monitoring
Technical Specification (TS) 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 126 degrees F for all casks loaded under Certificate of Compliance (CoC)
Amendment 2 (casks 1-7) and 137 degrees F for casks loaded under CoC Amendment 5 (casks 8-23). Seventeen of 23 HI-STORM casks at Grand Gulf were equipped with remote temperature monitoring equipment. The temperature surveillance or vent inspections were performed using Surveillance Procedure 06 OP-1000-D-0001, Attachment 1, Data Sheet 1 Daily Operating Logs, Revision 143. Documentation was reviewed for the (randomly selected) weeks of December 23, 2012, March 17, 2013, and June 9, 2013 for compliance with the technical specification. Of the three weeks selected for review, the TS surveillance requirement was met by performing daily vent inspections. No cask vents were reported as being blocked.
- (6) Corrective Action Program
A list of condition reports issued since the last NRC inspection in August of 2012 was provided by the licensee for the cask 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 Condition Report (CR) in the licensees corrective action program (CAP).
Of the list of CRs provided relating to the ISFSI and the cask handling crane, approximately 29 CRs were selected by the NRC inspectors for further review. The CRs related to a number of different topics including: a review of Holtec Information Bulletin 58 regarding galling of the vent port caps, evaluation of the calibration of the Supplemental Cooling System (SCS) flow meter, foreign material that was found in a loaded MPC, the crane stoppage during ISFSI preparation activities, and failure of the north mechanical load brakes on the cask handling crane during annual maintenance testing, etc.
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 cask handling crane. No NRC concerns were identified related to the condition reports reviewed.
- (7) Control of Measuring and Test Equipment
In the last routine ISFSI inspection conducted in August of 2012 (ML12303A002),inspectors found that GGNS had planned to eliminate the calibration requirement for the flow meter on the SCS. This was identified by the NRC inspectors as inconsistent with the requirements in the Holtec FSAR Table 8.1.7 which summarizes some instrumentation used to load and unload the HI-STORM 100 system. A footnote to the table stated that all instruments required calibration and flow meters were listed in the table. GGNS had stated that the flow meter was not required for SCS operations. The use of the flow meter was only necessary to satisfactorily perform the thermal validation test of License Condition 9 of the CoC, which specifically stated that coolant flow rate was to be measured as part of the test. The FSAR Appendix 2.C, Supplemental Cooling System, described the SCS function, and Figure 2.C.1 provided a drawing of the system showing the system components. Neither the discussion of the system nor the drawing required a flow meter. GGNS had completed License Condition 9 of the CoC in July of 2011. NRC inspectors reviewed the results of that test in the previous routine inspection. All flow meters and other equipment had been properly calibrated and no issues were identified with the test results. During the routine inspection conducted in August of 2012, GGNS issued CR 12-10254 to seek Holtec clarification of the FSAR intent for the calibration requirement listed in Table 8.1.7 to determine if it related to the supplemental cooling system prior to the next loading campaign.
The results of the CR 12-10254 were reviewed as part of this routine inspection.
Holtec provided a technical response to GGNS stating that after the CoC License Condition 9 activities were completed for the SCS, the only instruments that are required to be calibrated are those specified in the license basis for the system, which are limited to the temperature monitoring instruments. The response was documented in the Response to Request for Technical Information (RRTI) 2005-005R1, dated February 22, 2013. The response further clarified that the instrumentation that is identified for use in the functional test of the SCS per the CoC License Condition 9 is the flow meters and temperature gages. Since the flow meters are not discussed nor shown in the Appendix 2.C of the FSAR, a calibration of those flow meters would only be required if the site needed to demonstrate compliance with the License Condition 9. Following the test per the CoC, the flow meters are technically not required and therefore not subject to calibration. The temperature gages are the only equipment necessary to demonstrate that water temperature in the HI-TRAC/MPC annulus is below the design limit of 180 degrees F under steady-state conditions for the design basis heat load. The temperature gages are therefore required to be calibrated to ensure compliance with the Technical Specification associated with the SCS system, not the flow meters. The NRC inspectors did not find any issues with the Holtec justification or GGNSs plan to eliminate the calibration of the flow meters in the SCS system operations since they have already completed the CoC License Condition 9 test.
- (8) Heavy Loads Operations
The Grand Gulf spent fuel cask handling crane was a Whiting crane rated at 150 tons. The crane had been installed in the late 1970s. The crane was constructed to meet the requirements of Regulatory Guide 1.104, Overhead Crane Handling Systems for Nuclear Power Plants. This regulatory guide was later withdrawn in July 1981 and superseded by NUREG-0554, Single Failure Proof Cranes issued May 1979. Grand Gulf Engineering Change (EC)-12920 changed the commitment for the crane from Regulatory Guide 1.104 to NUREG-0554.
In the last routine ISFSI inspection conducted in August of 2012, inspectors documented that the cask handling crane north mechanical load brakes had failed to pass brake tests conducted during the annual crane maintenance operations.
GGNS had placed the issue in their corrective action program as CR-09-03632. The licensee prepared EC-16388 to accept the north mechanical load brake as inoperable but placed requirements to eventually replace the brake. The EC also documented that the crane could still be considered single failure proof without the north mechanical load brake being operational. This was based on an evaluation of the crane against Section 4.9 of NUREG-0554, which stated that the minimum hoisting braking system should include one power control braking system and two holding brakes. The spent fuel cask handling crane consisted of a single wire rope drum with dual, independent gear trains, each utilizing a shoe type holding brake.
Each of the two shoe type holding brakes were spring applied, electrically released.
There was one shoe type holding brake and one mechanical load brake for each of the two redundant gear trains and one eddy-current brake on the hoist main drive shaft. Each shoe type holding brake was rated at 220 percent of the rated motor torque. Each of the gear boxes was equipped with a mechanical load brake. These redundant brakes were intended to control the descending load. In consultation with the crane manufacturer, Whiting Corp., the licensee determined that the crane had sufficient redundant and operational braking systems to be considered single failure proof even with the north mechanical load brake inoperable. The two shoe type holding brakes met the holding brake requirement and the eddy-current brake met the power control braking system requirement. This conclusion was confirmed during consultation between the NRC regional inspectors and NRC headquarters staff during the routine inspection in 2012. At the conclusion of the 2012 routine inspection, work orders were in place at GGNS to still repair the north mechanical load bake even though the brake was not needed for the crane to meet single failure proof requirements.
Since the last inspection GGNS had decided to forgo replacing the north mechanical load brake and continue using the crane in a use-as-is configuration. GGNS performed EC-37635 dated January 28, 2013 to evaluate not repairing the north mechanical load brake. As stated before, the GGNS cask crane braking system consists of two holding brakes, an eddy current brake, and two mechanical load brakes (north and south). The mechanical load brakes are redundant to the eddy-current brake. The evaluation continued to state that operation of the cask crane with the north load brake inoperative/degraded will not adversely affect the ability of the crane to perform its design intent and maintain the minimum requirements of NUREG 0554. Additionally, in the event of an emergency both the eddy-current brake and the two shoe-type holding brakes can be utilized to manually lower a load.
The mechanical load brakes are not utilized for this condition. This was confirmed by inspectors by reviewing GGNS Procedure 07-S-14-282 Cask Crane Manual Lowering, Revision 1. The EC concluded that continued operation of the cask handling crane without taking credit for the mechanical load brakes will not adversely affect the single-failure-proof classification of the crane nor be an impact to the current operational basis.
- (9) Cask Handling Crane Annual Maintenance
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. GGNS utilized Procedure 07-S-14-226 General Maintenance Instruction Spent Fuel Cask Crane Periodic Inspection Safety Related, Revision 8 and Work Order (WO) 52349462, 1 Year Electrical Inspection, dated August 6, 2013 and WO 52463162, 1 year Mechanical Periodic Inspection for Cask Handling Crane, dated September 30, 2013 to fulfill this requirement.
- (10) HI-STORM 100 Cask Yearly Maintenance
Holtec 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 overpacks external surfaces and identification markings for signs of damage or degradation. NRC inspectors reviewed the work-orders and records related to the annual visual examination of Grand Gulfs HI-STORM casks for 2012, 2013, and 2014. Those documents included filled out copies of Procedure 20-S-02-001, Attachment XIV, Dry Fuel Storage Preventative Maintenance/HI-STORM Annual Inspection, Revisions 3, 4, and 5. Those visual examination checklists included 16 points of inspection that were verified as either satisfactory or unsatisfactory. For the years reviewed, one instance of an unsatisfactory condition was identified: one HI-STORM cask was found to have had a vent screen stud, nut, and washer missing from the top east vent. A CR and WO were generated to initiate the repair of that outlet vent. The NRC inspectors determined that the licensees yearly maintenance activities and records met the requirements of the Holtec FSAR.
b. Findings
No findings were identified.
.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.
GGNSs 10 CFR 72.212 Evaluation Report at the time of the inspection was Revision 9, dated June 26, 2013. Two revisions had been performed to the 72.212 Evaluation Report since the last NRC routine ISFSI inspection and were reviewed during this inspection. The associated 10 CFR 72.48 screenings for the revisions were also reviewed.
In Revision 8, the licensee added a limitation on MPC 353 due to an inadequate helium backfill that had been reviewed in the last routine inspection (ML12303A002), correction of minor reference errors, and clarification on limits associated with the forced helium dehydrator. Revision 9, was issued to document information that pertained to the casks loaded in 2013. The revision updated various tables to include specific information relating to the casks loaded during that campaign.
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 cask handling crane were provided by the licensee. The licensee had not performed any 72.48 ISFSI evaluations or any 50.59 cask handling crane evaluations since the last NRC inspection in August of 2012. However, Holtec had provided GGNS with a full 72.48 safety evaluation for a small piece of string that was discovered in MPC-383 during loading operations. That full evaluation was still in GGNS formal review process awaiting final review at the time of the NRC inspection.
During loading operations on August 24, 2013 and upon disengagement of the fuel grapple from fuel assembly XNC 617 in MPC cell location 14, a piece of string approximately 6 in length was noticed on the bale handle. The string settled between cells 14 and 13. The string became dislodged and floated to the bottom of cell 13. Upon discovery GGNS placed this issue in their corrective action system as CR 13-05843.
GGNS had Holtec perform an evaluation documented in RRTI 2005-10, Revision 0 to review and document the effect of leaving the string in the MPC. The piece of string was conservatively assumed to be 1/8 in diameter and 6 long, having a volume of 0.074 in3, and a weight of 0.004 lbs. The volume and weight were evaluated to be extremely insignificant in comparison to the MPC-68 and HI-STORM. The Holtec RRTI evaluation reviewed Normal Storage Conditions, including change in dead weight, effects on handling, pressure, temperature in the MPC, and effects from snow and ice. The RRTI found all of those conditions to have no adverse effect on the dry cask storage system.
The RRTI also documented an evaluation on the Off-Normal Conditions of Storage, which included pressure and temperature of the system, leakage of a seal, partial blockage of the HI-STORM air inlets and off-normal HI-TRACK handling. All of those conditions were found to have no adverse effects on the dry cask storage system. The evaluation showed the foreign material had extremely little to no effect to the system under the new conditions and showed the MPC was fully capable of providing storage for the spent fuel. Based on the results of Holtecs RRTI evaluation, GGNS performed EC-46362 to continue loading the MPC-383 and place the canister on the ISFSI pad.
The EC documented that Holtec would be performing a full safety evaluation to properly document the condition. The EC allowed continuing loading operations at risk without the full 72.48 evaluation completion, with the knowledge that if the 72.48 evaluation cannot adequately address the foreign material condition, the MPC would have to be opened to remove the foreign material. The MPC and canister were placed on the ISFSI pad on September 9, 2013. Holtec issued the full safety evaluation in 72.48 Full Safety Evaluation Number 1048, dated January 15, 2014. The evaluation reviewed the foreign material against all Accident Conditions of Storage, structural integrity, thermal, shielding, criticality, confinement, and operations. The full safety evaluation concluded that the presence of the foreign material did not adversely impact the design function of the MPC or the HI-STORM 100 dry cask system.
From the list of 72.48 screens (also known as Process Applicability Determinations (PADs)) provided by the licensee, nine screenings were selected for further review. At GGNS the 72.48 screens or PADs were placed in the Engineering Change (EC)associated with the change or modification. The licensee utilized Procedure EN-LI-100, Process Applicability Determination, Revision 13 and Procedure EN-LI-112 10 CFR 72.48 Evaluations, Revision 10 to perform the 10 CFR 72.48 screenings or perform a full safety evaluation. None of the additional screenings (PADs) reviewed required a full 10 CFR 72.48 safety evaluation. All screenings were determined to be adequately evaluated.
b. Findings
No findings were identified.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On June 5, 2014, the inspectors presented the inspection results to Mr. Kevin Mulligan, Site Vice President Operations, and other members of the licensee staff. 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
R. Benson, Radiation Protection Manager B. Carroll, Reactor Engineer Supervisor D. Ellis, EP Planner J. Nadeau, Regulatory Assurance Manager E. Riggs, Project Manager Dry Fuel Storage R. Scarbrough, Regulatory Assurance Specialist
INSPECTION PROCEDURES USED
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
None
Discussed
None
Closed
None 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
REVISION / DATE
Survey #1404-
0165
Grand Gulf ISFSI Pad Survey
4/11/14
E-7185-010
North Cask Storage Pad
5/14/06
E-0631
Site Lighting Plan
3/30/06
ER-GG-2003-
018-009-001
DFS Cask Construction Slab
Rev. 0
Procedures
NUMBER
TITLE
REVISION / DATE
Corrective Action Process
Rev. 23
Process Applicability Determination
Rev. 13
CFR 72.48 Evaluations
Rev. 10
07-S-14-282
Cask Crane Manual Lowering
Rev. 1
07-S-14-226
Cask Crane Periodic Inspection
Rev. 8
Fuel Selection for Holtec Dry Cask Storage
Rev. 5
Special Nuclear Material Control
Rev. 11
06-OP-1000-D-
0001
Data Sheet 1 Daily Operating Logs
Rev. 143
20-S-02-001
HI-STORM Annual Inspection
Revs. 3, 4, and 5
Design Basis Documents
NUMBER
TITLE
REVISION / DATE
Grand Gulf Nuclear Station 10 CFR 72.212 Report HI-
STORM 100 System
Rev. 9
Certificate of Compliance 72-1014 HI-STORM 100
Cask System
Amendment 5
Holtec International Final Safety Analysis Report for
the HI-STORM 100 Cask System
Rev. 7
- 3 -
Miscellaneous Documents
NUMBER
TITLE
REVISION / DATE
RRTI 2005-
0005R1
Report for Request for Technical Information
2/22/13
QA-20-2012-
GGNS-01
Quality Assurance Audit Report
10/11/12
283
Oversight Observation Checklist
10/25/13
0387
O2C Report
08/22/12
0381
O2C Report
08/17/12
AWO-
1314410584
Landauer Radiation Dosimetry Report
05/31/13
GNRO-
2013/00009
Letter: Spent Fuel Storage Radioactive Effluent
Release Report for 2012
2/14/13
GNRO-
2014/00019
Letter: Spent Fuel Storage Radioactive Effluent
Release Report for 2013
2/27/14
Grand Gulf Offsite Dose Calculation Manual
Rev. 35
Engineering Changes
EC-37635
EC-13502
EC-41939
EC-41939
Work Orders
WO 52349462
WO 00204609
WO 330016-01
WO 00364709-01
Condition Reports
CR 12-09533
CR 12-10418
CR 12-12030
CR 13-06849
CR 12-10010
CR 12-10692
CR 13-01461
CR 14-03606
CR 12-10036
CR 12-10731
CR 13-03307
CR 12-10254
CR 12-10254
CR 12-10892
CR 13-04541
CR 12-10326
CR 12-11061
CR 13-05589
CR 14-02238
CR 12-10342
CR 12-11076
CR 13-05758
CR 12-12871
CR 12-10360
CR 12-11085
CR 13-05843
CR 12-02238
CR 12-09949
CR 12-11069
CR 13-00695
CR 12-10692 CR 12-10078
- 4 -
LIST OF ACRONYMS
Agencywide Documents Access and Management System
ANSI
American National Standards Institute
Annual Radiological Environmental Operating Report
American Society of Mechanical Engineers
Boiling Water Reactor
Corrective Action Program
Certificate of Compliance
CR
Condition Report
CFR
Code of Federal Regulations
Division of Nuclear Material Safety
EC
Engineering Change
Final Safety Analysis Report
Grand Gulf Nuclear Station
IMC
Inspection Manual Chapter
Independent Spent Fuel Storage Installation
kW
kilo-watt
mR/h
milliroentgen per hour
mrem/h
millirem per hour
µR/h
microroentgen per hour
Multi-Purpose Canister
MWd/MTU
megawatt days per metric ton uranium
NRC
U.S. Nuclear Regulatory Commission
Protected Area
Optically Stimulated Luminescence
Quality Assurance
Radiological Environmental Monitoring Program
radiation protection
RRTI
Response to Request for Technical Information
SCS
Supplement Cooling System
thermoluminescent dosimeter
ATTACHMENT 2:
LOADED CASKS AT THE GRAND GULF NUCLEAR STATION ISFSI
LOADING
ORDER
SERIAL No.
HI-STORM
No.
DATE
ON PAD
HEAT LOAD
(kW)
BURNUP
MWd/MTU (max)
MAXIMUM FUEL
ENRICHMENT %
PERSON-REM
DOSE
MPC-47
Serial No.
11/18/06
2.26
34,320
3.42
0.252
MPC-44
Serial No.
2/01/06
15.48
35,866
3.42
0.236
MPC-46
Serial No.
2/08/06
19.11
43,546
3.71
0.218
MPC-45
Serial No.
2/15/06
21.82
46,297
3.71
0.246
MPC-214
Serial No.
269
04/21/08
19.14
46,215
3.71
0.350
MPC-215
Serial No.
270
04/28/08
19.35
46,301
3.71
0.279
MPC-69
Serial No.
271
05/05/08
19.03
45,941
3.71
0.262
MPC-225
Serial No.
333
09/01/09
19.73
44,933
3.71
0.177
MPC-226
Serial No.
334
09/10/09
19.81
44,653
3.71
0.208
MPC-227
Serial No.
335
09/16/09
19.86
44,882
3.71
0.218
MPC-224
Serial No.
2
2/09/09
18.92
44,622
3.71
0.191
MPC-223
Serial No.
331
2/16/09
18.94
44,943
3.71
0.159
MPC-350
Serial No.
2
06/29/11
18.01
44,945
3.71
0.197
MPC-351
Serial No.
533
07/01/11
16.29
44,857
3.71
0.126
MPC-352
Serial No.
534
07/16/11
18.74
48,290
3.87
0.161
MPC-353
Serial No.
535
07/23/11
19.09
49,591
3.87
0.115
- 2 -
LOADING
ORDER
SERIAL No.
HI-STORM
No.
DATE
ON PAD
HEAT LOAD
(kW)
BURNUP
MWd/MTU (max)
MAXIMUM FUEL
ENRICHMENT %
PERSON-REM
DOSE
MPC-354
Serial No.
536
07/29/11
18.93
49,431
3.87
0.116
MPC-381
Serial No.
644
08/16/13
15.17
44,929
3.87
0.172
MPC-382
Serial No.
645
08/23/13
15.65
44,725
3.87
0.142
MPC-383
Serial No.
646
09/03/13
15.25
44,699
3.87
0.098
MPC-384
Serial No.
647
10/11/13
15.51
44,908
3.87
0.151
MPC-385
Serial No.
648
10/28/13
15.47
44,818
3.87
0.121
MPC-386
Serial No.
649
11/01/13
15.67
44,869
3.87
0.117
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 - 7 were loaded under Certificate of Compliance, Amendment 2; Holtec Final Safety Analysis Report, Revision 3
Casks 8 - 23 were loaded under Certificate of Compliance, Amendment 5; Holtec Final Safety Analysis Report, Revision 7
Casks 4 - 7 Values of heat load and burnup reflect corrections to the CASKLOADER data files in 2008. Burnup values were from the CD file
Burnup_adjustments.xls of NEAD-SR-08/021. The heat load values were from the same CD in the files Real_case_N4.xls through
Real_case_N7.xls