IR 05000313/2015001
| ML15117A663 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/13/2015 |
| From: | Lantz R E NRC/RGN-IV/DRP |
| To: | Jeremy G. Browning Entergy Operations |
| References | |
| IR 2015001 | |
| Download: ML15117A663 (48) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION IV 1600 E. LAMAR BLVD. ARLINGTON, TX 76011-4511 May 13, 2015 Mr. Jeremy Browning, Site Vice President Arkansas Nuclear One Entergy Operations, Inc.
1448 SR 333 Russellville, AR 72802-
0967
SUBJECT: ARKANSAS NUCLEAR ONE NRC INTEGRATED INSPECTION REPORT 05000313/2015001 and 05000368/2015001
Dear Mr. Browning:
On March 31, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Arkansas Nuclear One facility, Units 1 and 2. O n April 2, 2015, the NRC inspectors discussed the results of this inspection with you and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report. NRC inspectors documented one finding of very low safety significance (Green) in this report.
This finding involved a violation of NRC requirements.
Additionally, NRC inspectors documented one Severity Level IV violation with no associated finding.
Further, inspectors documented a licensee-identified violation which was determined to be of very low safety significance in this report. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest the violations or significance of these NCVs, 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 Arkansas Nuclear One.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390 a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,/RA/ Ryan E. Lantz Deputy Director Division of Reactor Projects Docket Nos. 50-313, 50-368 License Nos. DPR-51; and NPF-6
Enclosure:
Inspection Report 05000313/2015001 and 05000368/2015001 w/
Attachments:
1. Supplemental Information 2. Request for Information 3. Detailed Risk Evaluation ANO-2 HELB Door 447 Issue cc w/ encl: Electronic Distribution for Arkansas Nuclear One
ML15117A663 SUNSI Review By: NHT ADAMS Yes No Non-Sensitive Sensitive Publicly Available Non-Publicly Available Keyword: OFFICE SRI:DRP/E RI:DRP/E DRS/EB1 DRS/EB2 DRS/OB DRS/PSB1 DRS/PSB2 NAME BTindell MYoung ERuesch GWerner VGaddy MHaire HGepford SIGNATURE /RA/ /RA/ /RA by TRFarnholtz Acting for/ /RA by GReplogle Acting for/ /RA by GWerner Acting for/ /RA by PElkman Acting For/ /RA by DAllen Acting for/ DATE 05/05/15 05/05/15 04/28/15 04/30/15 04/29/15 04/29/15 04/28/15 OFFICE DRS/TSS SRA:DRS/PSB2 DD:DRP NAME DAllen RDeese RLantz SIGNATURE /RA/ /RA/ /RA/ DATE 04/29/15 05/05/15 05/13/15 Letter to Jeremy Browning from Ryan E. Lantz dated May 13, 2015
SUBJECT: ARKANSAS NUCLEAR ONE NRC INTEGRATED INSPECTION REPORT 05000313/2015001 and 05000368/2015001 DISTRIBUTION: Regional Administrator (Marc.Dapas@nrc.gov) Deputy Regional Administrator (Kriss.Kennedy@nrc.gov) DRP Director (Troy.Pruett@nrc.gov) DRP Deputy Director (Ryan.Lantz@nrc.gov) DRS Director (Anton.Vegel@nrc.gov) DRS Deputy Director (Jeff.Clark@nrc.gov) Senior Resident Inspector (Brian.Tindell@nrc.gov) Resident Inspector (Matt.Young@nrc.gov) Acting Resident Inspector (Megan.Williams@nrc.gov) Branch Chief, DRP/E (Neil.Okeefe@nrc.gov) Senior Project Engineer, DRP/E (Nick.Taylor@nrc.gov) Project Engineer, DRP/E (Thomas.Farina@nrc.gov) Project Engineer, DRP/E (Brian.Correll@nrc,gov) Project Engineer, DRP/E (Jackson.Choate@nrc.gov) ANO Administrative Assistant (Gloria.Hatfield@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) Public Affairs Officer (Lara.Uselding@nrc.gov) Project Manager (Andrea.George@nrc.gov) Team Leader, TSS (Don.Allen@nrc.gov) ACES (R4Enforcement.Resource@nrc.gov) RITS Coordinator (Marisa.Herrera@nrc.gov) Regional Counsel (Karla.Fuller@nrc.gov) Technical Support Assistant (Loretta.Williams@nrc.gov) Congressional Affairs Officer (Jenny.Weil@nrc.gov) Congressional Affairs Officer (Angel.Moreno@nrc.gov) RIV/ETA: OEDO (Michael.Waters@nrc.gov) Enclosure 05000313; 05000368 DPR-51; NPF-6 05000313/2015001; 05000368/2015001 Entergy Operations Inc. Arkansas Nuclear One, Units 1 and 2 Junction of Hwy. 64 West and Hwy. 333 South Russellville, Arkansas January 1 through March 31, 2015 B. Tindell, Senior Resident Inspector M. Young, Resident Inspector S. Garchow, Senior Operations Engineer March 31One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. Additionally, NRC inspectors documented one Severity Level IV violation with no associated finding and one licensee-identified violation of very low safety significance. --- - ---- - The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions and that design changes were subject to design control measures commensurate with those applied to the original design. Specifically, the Unit 2 radwaste supply fans, 2VSF-7A and B, plenum doors and turbine building fire door 447 were maintained open, which provided a potential path for steam to enter the auxiliary building and impact both safety-related dc power trains during a high energy line break event in the turbine building. On February 12, 2014, the licensee suspended the modification and corrected the procedure. The licensee documented the concern in Condition Report CR-ANO-2-2014-00345. The failure to maintain separation of safety related systems and high energy piping systems in accordance with design, as stated in the Safety Analysis Report, was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter Determination Process (SDP) for Findings At- 1, 2012, the inspectors determined that the finding required a detailed risk evaluation because the finding represented a potential loss of system and/or function of the safety-related dc motor control centers, battery chargers and inverters. A senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was less than 4.8E-7/year (Green). The dominant initiating event likelihood of a rupture of the specific section of piping needed to initiate core damage sequences was extremely low. The inspectors determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than three years ago, and was not representative of current licensee performance. (Section 1R18) - a. b. - -- - - - - -- - - The inspectors directly observed the following nondestructive examinations: ---- ---- ---- ---- ---- ---- ---- ---- ---- ---- ---- ---- ---- The inspectors reviewed records for the following welding activities: - - - - a. The inspectors Reactor Vessel Upper and Lower Head Penetrations to determine whether the licensee identified any evidence of boric acid challenging the structural integrity of the reactor head components and attachments. The inspectors also verified that the required inspection coverage was achieved and limitations were properly recorded. The inspectors reviewed that the personnel performing the inspection were certified examiners to their respective nondestructive examination method. b. a. -- b. a. - January 27, 2015, Unit 1, lowered reactor coolant system inventory January 30, 2015, Unit 1, shift trains for decay heat removal and makeup during lowered inventory - - - Introduction. The inspectors identified a Green non-cited violation of 10 CFR Part 50, applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions and that design changes were subject to design control measures commensurate with those applied to the original design. Specifically, the Unit 2 radwaste supply fans, 2VSF-7A and B, plenum doors and turbine building fire door 447 were maintained open, which provided a potential path for steam to enter the auxiliary building and impact both safety-related dc power trains during a high energy line break event in the turbine building. Description. While touring the ventilation equipment area in the Unit 2 turbine building, inspectors noted that fire door 447 was tied open. The door had a caution tag attached -2106.032, -MOD for Maintaining potential effects to battery operability. The procedure required battery room temperatures be maintained greater than 60 degrees Fahrenheit and allowed for opening door 447 and opening the inlet plenums for radwaste supply fans 2VSF-7A and 2VSF-7B. This provided a flow path for warm turbine building air to be discharged into the battery corridor which would increase the battery room temperatures. Inspectors reviewed engineering evaluations that supported the temporary modification for this alignment during periods of cold weather. The subject reviews, which were completed in 2002, stated that because the radwaste supply fans take suction from the roof of the turbine building, during wintertime operations, the outside air supply temperatures can be between 10 to 20 degrees Fahrenheit. The colder air would then be delivered into the corridor at the battery rooms, reducing the battery room temperatures to unacceptable levels. By opening up the air path to the turbine building, the colder outside air would mix with the warmer turbine building air to increase the temperature of the corridor at the battery rooms. Inspectors noted that the evaluations did not account for the impacts of a high energy line break inside the turbine building on any equipment supplied by the radwaste fans, specifically, the effects of humidity on electrical equipment. ned the design basis and stated that separation between redundant safety-related components and separation between these components and high energy piping systems has been provided in the design and layout of this plant. This separation provides the primary means of assuring safe plant shutdown capability following a postulated high energy pipe break. Inspectors reviewed Drawing M-2263, Sheet 6, Instrumentation Diagram Air Flow Diagram HVAC Auxiliary Building Miscellaneous the battery room corridor, but they also supplied both direct current equipment rooms. Inspectors were concerned that, in the event of a main steam or feedwater line break in the turbine building, steam could be drawn into the open plenums of the radwaste supply fans and adversely impact both trains of safety-related batteries, direct current switchgear, battery chargers, and inverters. Since 2002, the licensee had been using the temporary modification and the periods of alignment varied based on the battery room temperatures. Inspectors noted that the licensee installed the modification for the radwaste supply fans on January 4, 2014, and the modification was suspended on February 12, 2014, after inspectors identified that the engineering evaluation that supported the temporary modification did not consider high energy line break effects on the direct current power system. The licensee documented the concern in Condition Report CR-ANO-2-2014-00345. Analysis. The inspectors concluded that the failure to maintain separation of safety related systems and high energy piping systems in accordance with design, as stated in the Safety Analysis Report, was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the Unit 2 radwaste supply fans, 2VSF-7A and B, plenum doors and turbine building fire door 447 were maintained open, which provided a path for steam to potentially adversely impact both dc power trains during a high energy line break event in the turbine building. Using Inspection Manual Chapter - 1, 2012, the inspectors determined that the finding required a detailed risk evaluation because the finding represented a potential loss of system and/or function of the safety-related dc motor control centers, battery chargers and inverters. A senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was less than 4.8E-7/year (Green). The The initiating event likelihood of a rupture of the specific section of piping needed to initiate core damage sequences was extremely low. See Attachment 3 for the detailed risk evaluation. The inspectors determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than three years ago, and was not representative of current licensee performance. Enforcement. Title 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies, are correctly translated into specifications, drawings, procedures, and instructions. Design changes shall be subject to design control measures commensurate with those applied to the original design. Unit 2, Safety Analysis Report, separation between redundant safety-related components and separation between these components and high energy piping systems has been provided in the design and layout of this plant. This separation provides the primary means of assuring safe plant shutdown capability following a postulated high energy pipe break. Contrary to the above, between January 3, 2002, and February 12, 2014, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions and that design changes were subject to design control measures commensurate with those applied to the original design. Specifically, Unit 2 Procedure OP--MOD for Maintaining Battery Operability -7A and B, plenum doors and turbine building fire door 447 to be maintained open, which provided a path for steam to potentially adversely impact both dc power trains during a high energy line break event in the turbine building. The licensee corrected the conditions by closing the plenum doors and removing the temporary modification allowance from the procedure. This violation is being treated as a noncited violation (NCV), consistent with Section action program as Condition Report CR-ANO-2-2014-00345. (NCV 05000368/2015001-01, Failure to Protect Safety Equipment From Potential High Energy Line Breaks) - - -- ---- - -- -- - - - -- January 30, 2015, Unit 1, borated water storage outlet check valve BW-4B full flow test January 30, 2015, Unit 1, low pressure injection pump 34A full flow test - - -- - - - The inspectors evaluated whether the licensee controlled in-plant airborne radioactivity concentrations consistent with ALARA principles and that the use of respiratory protection devices did not pose an undue risk to the wearer engineering controls The licenseequality assurance of National Institute of Safety and Health (NIOSH)-certified equipment, qualification and training of personnel, and user performance lling and transporting self-contained breathing apparatus (SCBA) air bottles to and from the control room and operations support center during emergency conditions, status of SCBA staged and ready for use in the plant and associated surveillance records, and personnel qualification and training -- - - - - - - a. b. - .1 (Closed) Licensee Event Report 05000368/2014-002-00, Automatic Reactor and Main Generator trip with a Subsequent Emergency Feedwater Actuation and Start of an Emergency Diesel Generator a. On April 3, 2014, Unit 2 tripped from 100 percent power due to a lightning strike on an offsite power line that resulted in a momentary undervoltage condition on startup transformer 3. This under voltage condition initiated a fast transfer of nonvital buses to startup transformer 2. However, by design two of the buses did not transfer, which caused loss of power to two reactor coolant pumps and a reactor trip. All safety equipment operated as expected. No additional deficiencies were identified during the review of the licensee event report. This licensee event report is closed. b. (Closed) Licensee Event Report 05000368/2013-003-00, Inoperable Offsite Power Supply Transformer Arkansas Nuclear One - Unit 2 a. Inspection Scope On August 20, 2013, the licensee identified an undocumented wiring configuration associated with the Unit 2 startup transformer 3 voltage regulator circuit. Startup transformer 3 is one of two offsite power source transformers designed to supply offsite power for Unit 2. The wiring configuration would have prevented the startup transformer 3 voltage regulator from operating as designed. The voltage regulator has an automatic tap-changer designed to step up transformer voltage in response to a low voltage condition after a twenty second time delay to maintain a pre-defined voltage control band. This twenty second time delay is designed to be bypassed for three minutes in the event of a main turbine generator lockout, to allow immediate voltage adjustments as Unit 2 station loads are fast transferred from the unit auxiliary transformer to the offsite startup transformer 3 during worst case accident load sequencing. The discovered wiring configuration prevented the bypass of the twenty second time delay, resulting in the inoperability of the offsite power source. The inspectors reviewed the data and determined that startup transformer 3 did not exceed its technical specification allowed outage time due to new calculations performed by the licensee since the issuance of the LER. The incorrect wiring configuration appeared to have been introduced in the 2005-2006 timeframe. On August 21, 2013, a temporary modification was installed to remove the startup transformer 3 voltage regulatory tap-change controller twenty second time delay, which restored the startup transformer 3 operability. This licensee event report was reviewed and no violations of NRC requirements were identified. This licensee event report is closed. b. .3 (Closed) Licensee Event Report 05000368/2014-004-01, Technical Specification 3.0.4 Violation due to a Mode Change with an Inoperable Emergency Feedwater Pump a. Inspection Scope The licensee changed Unit 2 modes from Mode 4 to Mode 3 with the turbine driven emergency feedwater pump inoperable, which is a violation of technical specifications, due to a human performance error. Maintenance personnel had failed to follow the work instructions for governor calibration, which resulted in unstable operation of the governor and the turbine overspeed trip during surveillance testing. This performance deficiency is minor because the error was detected by a post-maintenance test before the equipment was returned to service. This failure to comply with technical specifications constitutes a minor violation that is not subject to enforcement action in accordance with This licensee event report is closed. b. Findings No findings were identified. - - -- -- -- - ---- -- ------ - - Title 10 CFR 71.5, Section (a), Transportation of Licensed Material, requires each licensee who transports licensed material outside the site of usage, shall comply with the applicable requirements of the DOT regulations in 49 CFR. 49 CFR Part 172.800(b) requires, in part, that the licensee must develop and adhere to a transportation security plan. The licensee implemented Procedure EN-RW-these requirements. Contrary to the above, on December 18, 2014, the licensee identified that they failed to follow their Transportation Security Plan (TSP). Specifically, licensee personnel shipped a radioactive quantity of Category 2, RAM-QC, on the public highways to a waste processor without acknowledging the shipment as a RAM-QC shipment or making appropriate notifications as required by Procedure EN-RW-106. Six shipments were identified as being shipped in violation of the TSP requirements because they failed to identify the material as RAM-QC due to inadequate Category 2 threshold values. February 12, 2008, the inspectors determined the finding has very low safety significance (Green) because the licensee had an issue involving transportation of radioactive waste, but it did not involve: (1) a radiation limit being exceeded, (2) a breach of package during transport, (3) a certificate of compliance issue, (4) a low level burial ground nonconformance, or (5) a failure to provide emergency information. The licensee documented this issue in their corrective action program as Condition Report CR-ANO-C-2014-03341 and made corrections to Procedure EN-RW-106 to prevent this issue from reoccurring. There is no cross-cutting aspect with this violation due to it being licensee-identified.
Attachment 1 - 05000368/2015001-01 Failure to Protect Safety Equipment From Potential High Energy Line Breaks (Section 1R18)05000368/2015001-02 05000368/2014-002-00 Automatic Reactor and Main Generator trip with a Subsequent Emergency Feedwater Actuation and Start of an Emergency Diesel Generator (Section 4OA3.1) 05000368/2013-003-00 Inoperable Offsite Power Supply Transformer Arkansas Nuclear One - Unit 2 (Section 4OA3.2) 05000368/2014-004-01 Technical Specification 3.0.4 Violation due to a Mode Change with an Inoperable Emergency Feedwater Pump (Section 4OA3.3) -
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A1-8 - - - --- - - -1104.005 Reactor Spray System Operation - Decay Heat System Operations - - - - ---- -- -- -- -- -- -- -- -- --
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Attachment 2 The following items are requested for the Occupational Radiation Safety Inspection at Arkansas Nuclear One (February 2 6, 2015) Integrated Report 2015001 Inspection areas are listed in the attachments below. Please provide the requested information on or before January 16, 2014. Please submit this information using the same lettering system as below. For example, all contacts and phone numbers for Inspection Procedure 71124.01 should be in a file/folder titled - - If information is placed on ims.certrec.com, please ensure the inspection exit date entered is at least 30 days later than the on-site inspection dates, so the inspectors will have access to the information while writing the report. In addition to the corrective action document lists provided for each inspection procedure listed below, please provide updated lists of corrective action documents at the entrance meeting. The dates for these lists should range from the end dates of the original lists to the day of the entrance meeting. If more than one inspection procedure is to be conducted and the information requests appear to be redundant, there is no need to provide duplicate copies. Enter a note explaining in which file the information can be found. If you have any questions or comments, please contact the lead inspector, Natasha Greene at (817) 200-1154 or Natasha.Greene@nrc.gov. Currently, the other inspector will be Pete Hernandez. He may be contacted at (817) 200-1168 or Pete.Hernandez@nrc.gov. PAPERWORK REDUCTION ACT STATEMENT This letter does not contain new or amended information collection requirements subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing information collection requirements were approved by the Office of Management and Budget, control number 3150-0011.
A2-2 1. Date of Last Inspection: May 16, 2014 A. List of contacts (with official title) and telephone numbers for the radiation protection organization staff and technicians B. Applicable organization charts C. Audits, self-assessments, and licensee event reports (LERs) written since date of last inspection, related to this inspection area D. Procedure indexes for the radiation protection procedures E. Please provide specific procedures related to the following areas noted below. Additional specific procedures may be requested by number after the inspector reviews the procedure indexes. 1. Radiation Protection Program Description 2. Radiation Protection Conduct of Operations 3. Personnel Dosimetry Program 4. Posting of Radiological Areas 5. High Radiation Area Controls 6. RCA Access Controls and Radworker Instructions 7. Conduct of Radiological Surveys 8. Radioactive Source Inventory and Control 9. Declared Pregnant Worker Program F. List of corrective action documents (including corporate and subtiered systems) since date of last inspection a. Initiated by the radiation protection organization b. Assigned to the radiation protection organization c. Identify any condition reports that are potentially related to a performance indicator event NOTE: The lists should indicate the significance level of each issue and the search criteria can perform word searches. If not covered above, a summary of corrective action documents since date of last inspection involving unmonitored releases, unplanned releases, or releases in which any dose limit or administrative dose limit was exceeded (for Public Radiation Safety Performance Indicator verification in accordance with Inspection Procedure 71151) G. List of radiologically significant work activities scheduled to be conducted during the inspection period (If the inspection is scheduled during an outage, please also include a list of work activities greater than 1 rem, scheduled during the outage with the dose estimate for the work activity.) H. List of active radiation work permits I. Radioactive source inventory list A2-3 3. In-Plant Airborne Radioactivity Control and Mitigation (71124.03) Date of Last Inspection: May 16, 2014 A. List of contacts and telephone numbers for the following areas: 1. Respiratory protection program 2. Self-contained breathing apparatus B. Applicable organization charts C. Copies of audits, self-assessments, vendor or NUPIC audits for contractor support, self-contained breathing apparatuses (SCBAs) and LERs, written since date of last inspection related to: 1. Installed air filtration systems 2. SCBAs D. Procedure index for: 1. use and operation of continuous air monitors 2. use and operation of temporary air filtration units 3. Respiratory protection E. Please provide specific procedures related to the following areas noted below. Additional specific procedures may be requested by number after the inspector reviews the procedure indexes. 1. Respiratory protection program 2. Use of SCBAs 3. Air quality testing for SCBAs F. A summary list of corrective action documents (including corporate and subtiered systems) written since date of last inspection, related to the airborne monitoring program including: 1. Continuous air monitors 2. SCBAs 3. Respiratory protection program NOTE: The lists should indicate the significance level of each issue and the search criteria G. List of SCBA-qualified personnel - reactor operators and emergency response personnel H. Inspection records for SCBAs staged in the plant for use since date of last inspection I. SCBA training and qualification records for control room operators, shift supervisors, shift technical advisors, and operational support center personnel for the last year A selection of personnel may be asked to demonstrate proficiency in donning, doffing, and performance of functionality check for respiratory devices.
Attachment 3 Attachment 3 Detailed Risk Evaluation ANO-2 HELB Door 447 Issue A Region IV senior reactor analyst performed a detailed risk evaluation of the high energy line break concern and concluded it to be of very low safety significance (Green). Total change in core damage frequency was 4.8E-7/year. Internal Events Assumptions: 1. Only 20 feet of the main steam piping, if ruptured, could produce steam in sufficient quantities and flow that would progress through Door 447 to impact the electrical DC buses. 2. If the steam were to be directed into Door 447 from the 20 foot section of interest, the failure of all Unit 2 vital battery chargers and DC Buses 2D01 and 2D02 would result. 3. The break in the steam piping is too small to initiate a main steam isolation signal which would shut the main steam isolation valves and negate adverse effects to plant safety equipment. Analysis: First, the analyst identified the approximate frequency for a steam line piping break. Updated 2010 data from NUREG/CR--Average Performance for Components and specified the mean frequency for a large leak pipe fault as 2.5E-11/ft-hour. Second, the analyst used the Arkansas Nuclear One - Unit 2 SPAR model, Revision 8.26, to calculate the conditional core damage probability for a high energy line break which resulted in a subsequent failure of all Unit 2 vital battery chargers and DC Buses 2D01 and 2D02. The results of the loss of this equipment produced a plant state conditional core damage probability of 1.0 in the analysis. The analyst used a cutset truncation of 1.0E-11 and assumed an exposure interval of 39 days, which was the time the vulnerability was present in the plant. The delta-core damage frequency (delta-CDF) for internal events was estimated as follows: 2.5E-11/foot-hour * 20 (feet of piping) * 39 days* 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s/day * 1.0 = 4.7E-7 External Events Tornados Assumptions: 1. Only a tornado of F2 or above would have been able to damage the section of pipe that would produce a plume of steam through Door 447. 2. Tornadoes of intensity F2 or above in the 39 day exposure period of January 4 to February 12 would have occurred at the same frequency as they had historically from 1950 2006 within a 100 kilometer radius of the plant. 3. Operators would be well aware of impending tornadic activity. When a tornado hit the plant, operators would diagnose the break in the main steam line with nominal success (a human error probability of 1.0E-2). 4. Operators would take action after diagnosis of the break in the main steam line with nominal success (a human error probability of 1.0E-3).
A3-2 Analysis: The analyst observed that 4 tornados of F2 intensity or greater had historically hit in the dates of interest. These tornados accounted for 1.45% of the total yearly probability of a strike by a tornado of intensity of F2 or above with a probability of 4.8E-6. The analyst applied the exposure time of 39 days (0.107 years) and a plant condition CCDP of 1.0 to obtain: 4.8E-6/year * 0.107 years * 1.1E-2 * 1.0 = 5.7E-9 The analyst used this estimate as the risk posed by a tornado striking the plant and breaking the 20 foot section of piping that would then discharge into Door 447. Seismic Assumptions: 1. The seismic frequency and fragilities information were used from the Risk Assessment Standardization Project Manual, Volume 2, External Events. 2. The 20 foot section of steam piping that would have to break is 42-inch piping and a section of moisture separator reheater piping. The piping is not seismically qualified, but is of robust construction that is adequately supported. The analyst used the surrogate of component cooling water piping for the seismic event. 3. During a seismic event, the analyst modeled the turbine building as failing with Risk Assessment Standardization Project Manual fragilities and causing damage to the piping. 4. Operators would be well aware of seismic activity. When an earthquake hit the plant, operators would diagnose the break in the main steam line with nominal success (a human error probability of 1.0E-2). 5. Operators would take action after diagnosis of the break in the main steam line with nominal success (a human error probability of 1.0E-3). Analysis: The seismic analysis for piping failure analysis yielded a seismic initiating event frequency of 2.30E-6/year. Applying this to the 39 day exposure period and 1.1E-2 operator human error probability yielded: 2.3E-6/year * 0.107 years * 1.1E-2 * 1.0 = 2.7E-9 The seismic analysis for turbine building failure which would impact the piping yielded a seismic initiating event frequency of 3.8E-6/year. Applying this to the 39 day exposure period and a 1.1E-2 operator human error probability yielded: 3.8E-6/year * 0.107 years * 1.1E-2 * 1.0 = 4.5E-9 Combining the seismic analyses yielded a total seismic risk of 8.2E-9. Other External Events The analyst reviewed the Individual Plant Examination for External Events and screened other external events qualitatively as insignificant for this performance deficiency.
A3-3 Total External Events The total external events risk was obtained by summing the tornado and seismic risks. Tornado 5.7E-9 Seismic 8.2E-9 Total External 1.4E-8 Total Risk The total risk was obtained by summing the external and internal events risks. External 1.4E-8 Internal 4.7E-7 Total Risk 4.8E-7 Since the change in core damage frequency was less than 1E-6, the finding was of very low safety significance (Green). Large Early Release Frequency (LERF) The analyst used Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, to determine that this condition was not a significant contributor to the large early release frequency (LERF) because steam generator tube rupture and intersystem LOCA sequences were not affected by the performance deficiency.