IR 05000482/2015001
| ML15120A088 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 05/07/2015 |
| From: | Andrew Rosebrook NRC/RGN-IV/DRP/RPB-B |
| To: | Heflin A C Wolf Creek |
| Rosebrook A A | |
| References | |
| IR 2015001 | |
| Download: ML15120A088 (56) | |
Text
May 7, 2015
Mr. Adam C. Heflin, President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839
SUBJECT: WOLF CREEK GENERATING STATION NRC INTEGRATED INSPECTION REPORT 05000482/2015001
Dear Mr. Heflin:
On March 28, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Wolf Creek Generating Station. On April 1, 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 three findings of very low safety significance (Green) in this report. All of these findings involved violations of NRC requirements. Further, inspectors documented one licensee-identified finding which was determined to be of very low safety significance (Green) 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 the Wolf Creek Generating Station. If you disagree with a cross-cutting aspect assignment with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV; and the NRC resident inspector at the Wolf Creek Generating Station. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390Inspections, 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/ Andrew A Rosebrook, Acting Chief Project Branch B Division of Reactor Projects Docket Nos. 50-482 License Nos. NPF-42 Enclosure: Inspection Report 05000482/2015001 w/ Attachment 1: Supplemental Information Attachment 2: Request for Information for Occupational Radiation Safety Inspection cc w/ encl: Electronic Distribution
ML 15120A088 SUNSI Review By: AAR ADAMS Yes No Non-Sensitive Sensitive Publicly Available Non-Publicly Available Keyword: OFFICE ASRI:DRP/B ASRI:DRP/B RI:DRP/B C:DRS/EB1 C:DRS/EB2 C:DRS/OB C:DRS/PSB1 NAME CHenderson DDodson RStroble TFarnholtz GWerner VGaddy MHaire SIGNATURE /RA by ARosebrook Acting For/ /RA by ARosebrook Acting For/ /RA by ARosebrook Acting For/ /RA/ /RA/ /RA/ /RA/ DATE 05/07/15 05/07/15 05/07/15 05/04/15 05/04/15 05/04/15 05/05/15 OFFICE C:DRS/PSB2 C:DRS/TSS BC:DRP NAME HGepford DAllen ARosebrook SIGNATURE /RA/ /RA/ /RA/ DATE 05/01/15 05/05/15 05/07/15 Letter to Adam from Andrew A. Rosebrook dated May 7, 2015
SUBJECT: WOLF CREEK GENERATING STATION NRC INTEGRATED INSPECTION REPORT 05000482/2015001 Electronic distribution by RIV: Regional Administrator (Marc.Dapas@nrc.gov) Deputy Regional Administrator (Kriss.Kennedy@nrc.gov) DRP Director (Troy.Pruett@nrc.gov) Deputy Director (Ryan.Lantz@nrc.gov) DRS Director (Anton.Vegel@nrc.gov) DRS Deputy Director (Jeff.Clark@nrc.gov) Acting Senior Resident Inspector (Douglas.Dodson@nrc.gov) Acting Senior Resident Inspector (Christopher.Henderson@nrc.gov) Acting Senior Resident Inspector (Chris.Speer@nrc.gov) Resident Inspector (Raja.Stroble@nrc.gov) WC Administrative Assistant (Carey.Spoon@nrc.gov) Branch Chief, DRP (Andrew.Rosebrook@nrc.gov) Senior Project Engineer, DRP/B (David.Proulx@nrc.gov) Project Engineer, DRP/B (Fabian.Thomas@nrc.gov) Project Engineer, DRP/B (Steven.Janicki@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) Public Affairs Officer (Lara.Uselding@nrc.gov) Project Manager (Fred.Lyon@nrc.gov) Team Leader, DRS/TSS (Don.Allen@nrc.gov) RITS Coordinator (Marisa.Herrera@nrc.gov) ACES (R4Enforcement.Resource@nrc.gov) Regional Counsel (Karla.Fuller@nrc.gov) Technical Support Assistant (Loretta.Williams@nrc.gov) Congressional Affairs Officer (Jenny.Weil@nrc.gov) RIV Congressional Affairs Officer (Angel.Moreno@nrc.gov) RIV/ETA: OEDO (Michael.Waters@nrc.gov) NRR/DPR/PLPB Senior Project Manager (Brian.Benney@nrc.gov) Enclosure - findings of very low safety significance (Green) are documented in this report. All of these findings involved violations of NRC requirements. Additionally, NRC inspectors documented in this report one licensee-identified violation of very low safety significance. --- Green. The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the failure to properly preplan maintenance such that it would not affect safety-related equipment in accordance with procedure AP 22C--Line Qualitative . Specifically, during planning of emergent work activities on January 29, 2015, the licensee failed to recognize that when electrical cabinet doors containing safety-related under voltage and under frequency relays were opened to accomplish troubleshooting activities, the cabinet was not in a seismically qualified configuration. Thus the maintenance had the potential to impact the reliable operation of emergency diesel generator B during a seismic event. The licensee initiated Standing Order requirements for assessing operability of opening safety-related electrical cabinet and panel doors out of their seismically qualified configuration during maintenance activities and entered this issue into their corrective action program for resolution as Condition Reports 91501 and 94605. ilure to properly preplan maintenance such that it would not affect safety-related equipment during emergent work activities was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences properly preplan maintenance resulted in emergency diesel generator B being placed in a condition that did not meet its seismic design requirements. Using Inspection Manual Chapter 0609, AppendSignificance Determination Process for Finding At-determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with -cutting aspect in the area of human performance associated with work management. Specifically, the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work [H.5]. (1R13) The inspectors identified non-cited violation of 10 CFR 50, Appendix B, Criterion V, complete an adequate operability evaluation in accordance with procedure AP-bility criteria. Specifically, the licensee did not have an accurate technical basis for declaring the train A control room air condition unit operable when the minimum air flow rate was not met. The licensee operability evaluation, which declared the train A control room air condition unit operable, incorrectly applied instrument uncertainty and used a superseded minimum air flow value. When these inaccuracies were addressed, the licensee determined the train was inoperable. The licensee entered this issue into their corrective action program as Condition Report 92274. use of an inadequate technical basis for an operability evaluation of a non-conforming condition resulting in the train A control room air conditioning air condition unit being declared operable when it was actually inoperable was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associate cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage)Determination Process for Finding At-that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with -cutting aspect in the area of human performance associated with conservative bias component because the licensee did not use a decision making-practice that emphasized prudent choices over those that are simply allowable. A proposed action was determined to be safe in order to proceed, rather than unsafe in order to stop [H.14]. (1R15.1) The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion requirements of Station Procedure AP 10-Specifically, the failure initiate a breach permit and station a boundary watch when the auxiliary building emergency exhaust system boundary door 41015 was opened multiple times for transporting scaffolding from the turbine building to the auxiliary building. Opening this door without compensatory measures rendered the auxiliary building emergency exhaust system inoperable. The license entered this issue into their corrective action program for resolution as Condition Reports 92315 and 92630. a breach permit and implement required compensatory measures for when the auxiliary building emergency exhaust system boundary door 41015 was open was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the system, structure, and component and barrier performance attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to ensure the radiological barrier functionality of the auxiliary building emergency exhaust system. Specifically, without a dedicated individual in constant communication with the control room, as required by AP 10-104, opening this door required entry of Technical Specification 3.7.13 Limited Condition of Operation Condition B. The longest period door 41015 was open was approximately one hour without the required compensatory measure. Significance Determination Process (SDP) for Finding At-inspectors determined that the finding screened as having very low safety significance (Green) because the finding only involved a degradation of the radiological barrier function provided for the auxiliary building. The finding has a cross-cutting aspect in the area of human performance associated with work management. Specifically, the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work [H.5]. (1R15.2) - a. b. -- - - - -- - - --- - The inspectors directly observed the following nondestructive examinations: -- -- -- -- -- --- --- --- -- - - - - The inspectors reviewed records for the following welding activities: - - - - - - - - - - - Introduction. The inspector identified an unresolved item pertaining to 10 CFR 50 Appendix B, Criterion IX, Control of Special Processes, associated with method of performing ultrasonic examination of the reactor vessel flange stud hole threads in accordance with applicable American Society of Mechanical Engineers (ASME) Code requirements. Description. The inspector identified several issues of concern while observing the or vessel flange stud hole threads. The inspector questioned whether the licensee would be able to detect any reportable indication within the ASME Code examination zone using the technique employed. - - -The inspector questioned the following statements in the procedure: - - Procedure UT-that are orientated on a plane normal to the axis of the stud that are equal to or exceed 0.2 in, as measured radially from the root of the thread, shall be reported to the LMT Site Supervisor and recorded on the Ultrasonic Examination report Additional analysis and simulations need to be completed to determine if the licensee is meeting ASME Code requirements. URI 05000482/2015001-01, Questions Related to Ultrasonic Examination of Reactor Vessel Flange Stud Hole Threads. a. reactor vessel upper 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. During refueling outage RF19, ultrasonic examinations of all seventy-eight control rod drive mechanism (CRDM) penetration nozzles and the eddy current examination of the vent line in the reactor vessel head was completed. A number of thermal sleeves were found to have wear indications extending up to as much as 360 degrees around the thermal sleeve where the thermal sleeve exits the bottom end of the control rod drive mechanism head adapter tube. Wear was found in rodded and unrodded penetration locations. The wear is attributed to the thermal sleeve contacting the inside diameter of the CRDM head adapter tube due to a flow-induced impact/whirling motion of the thermal sleeve. The sleeve-to-adapter contact resulted in wear of material on the outside diameter of the thermal sleeves. A sample of the thermal sleeves were re-inspected this outage and no change in the wear indications were noted. - -- The inspectors reviewed the certification of the personnel performing the inspection to verify they were certified examiners to their respective nondestructive examination method. b. a. - --- a. - - - -- -- -- During the initial eddy current examinations in steam generator A, a single circumferential indication was identified in the hot leg tube sheet of tube (R20, C102) approximately 4 inches down from the top of the tube sheet. This indication is not located within a specified examination subset of the hot leg tube sheet (bulge or overexpansion). This primary water stress corrosion cracking (PWSCC) indication is associated with a low level (4.0 volt) bulge anomaly that is below the threshold of the bulge signal reporting criteria (18 volts) that had not previously been identified as a degradation mechanism. The tube was plugged and because the indication is 4 inches inside the tube sheet, there are no concerns with lateral movement resulting in tube severance if the indication grows. Because the tube is unpressurized, there is no pull- out force to cause vertical motion. Therefore, there was no need to stabilize the tube. The current EPRI Steam Generator Examination Guidelines for this damage mechanism require that a 100 percent inspection of affected steam generator (steam generator A) and a 20 percent inspection in the unaffected steam generators (steam generators B, C, and D) be completed. Wolf Creek expanded the hot leg top of tube sheet eddy current examinations to 100 percent in steam generator A and a minimum of 50 percent in steam generators B, C, and D. No additional indications were identified in the expanded scope inspection. -- .5 Identification and Resolution of Problems Inspection Scope b. a. - - - January 14, 2015,---- -- February 24, 2015, Failure to Assess the Operability of Emergency Diesel Generator B during Emergent Work Activities. Introduction. The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the licensee failure to properly preplan maintenance such that it would not adversely affect safety-related equipment in accordance with procedure AP 22C--. Specifically, during emergent work activities, the licensee failed to recognize that when electrical cabinet doors containing safety-related under voltage and under frequency relays were opened to accomplish maintenance, the cabinet was no longer in a seismically qualified configuration. Description. On January 29, 2015, while troubleshooting an intermittent power potential transformer fuse blown alarm for the emergency diesel generator B, maintenance personnel opened the doors to panel NE 106 to gain access to the relay NE 106160 per Work Order 15-397359-000. During the maintenance, inspectors noted that the doors were not restrained and there was not a dedicated person attending the door. The door associated with panel NE 106 contained safety-related under voltage and under frequency relays. The inspectors asked the licensee if the safety-related relays where seismically qualified with the door open. The licensee informed the inspectors that the safety-related relays where not seismically qualified with the panel door open. The inspectors were concerned that in the event of a seismic event, the doors could suddenly shut and cause the relays to change state, impacting the reliability of emergency diesel generator B at a time when it was required to perform its safety function. Thus, the inspectors concluded that the licensee should have declared the emergency diesel generator inoperable and entered the appropriate technical specification limiting condition for operation prior to the commencement of the maintenance. The emergency diesel generator was not in a non-conforming configuration for greater than the technical specification allowed outage time. The inspector reviewed Station Procedure AP 22C--Line Qualitative Risk , and determined the licensee failed to identify the worst case consequences (i.e., seismic event) and have appropriate mitigating actions for the emergent work activity in accordance with step 6.2.3 of the procedure when planning the emergent work activities for emergency diesel generator B. The licensee initiated Condition Reports 91501 and 94605 to document this issue in the corrective action program. Condition Report 91501 was initiated on February 3, 2015, for an industry concern regarding the opening of doors of operable safety related electrical cabinets and panels and Condition Report 94605 was initiated for the inspectors issue identified on January 29, 2015. In response to Condition Report 91501 the licensee initiated Standing Orexpectations for opening safety related electrical cabinets. Specifically the standing order required; (1) control room permission prior to opening any safety related cabinets; (2) the doors shall be attended at all times; (3) the doors shall be restrained, and the doors to be shut immediately if a seismic event were to occur. Analysis. The failure to properly preplan maintenance such that it would not affect safety-related equipment was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associate cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage). preplan maintenance resulted in emergency diesel generator B being placed into a condition that did not meet its seismic design requirementsDetermination Process for Finding At-determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-rule program. The finding has a cross-cutting aspect in the area of human performance associated with work management. Specifically, the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work [H.5]. Enforcement. Technical Specification 5.4.1.a requires, in part, that written procedures be established, implemented, and maintained covering the applicable procedures e Program requires maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Work Order 15-397359-000 provided work instructions for troubleshooting inside an electrical cabinet associated with emergency diesel generator B. Contrary to the above, on January 29, 2015, the licensee performed maintenance that affected safety-related equipment was not performed in accordance with documented instructions that were appropriate to the circumstances. Specifically, troubleshooting activities performed under Work Order 15-397359-000 caused the safety related emergency diesel B to be rendered non-conforming to its seismic requirements. Because the finding was of very low safety significance Condition Report 94605, it is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000482/2015001-02Assess the Operability of Emergency Diesel Generator B during Emergent Work Activities - - Failure to Complete an Adequate Operability Evaluation for Declaring the Train A Control Room Air Conditioning Unit Operable Introduction. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, failure to complete an adequate operability evaluation in accordance with procedure AP-, following the failure to meet a surveillance test acceptance criteria. Specifically, the licensee did not have an accurate technical basis for declaring the train A control room air condition unit operable when the minimum air flow rate was not met. The licensee operability evaluation, which declared the train A control room air condition unit operable, incorrectly applied instrument uncertainty and used a superseded minimum air flow value. When these inaccuracies were addressed, the licensee determined the train was inoperable. Description. The inspectors reviewed Condition Report 92109 based on its risk ring a design basis accident. This condition report documented that the train A control room air condition unit did not meet the minimum air flow acceptance criteria in accordance with Station Procedure STS PE-Revision 3A, and included an operability determination which the licensee completed on February 19, 2015. The inspectors reviewed the operability determination, Station Procedure STS PE-010A, and Design Calculation GK-M-d Heating Load Calculation for Control Room HVAC System Capabilities during Normal Plant Operation and Accident Conditions During a surveillance test on February 25, 2015, the air flow rate measured was 20,760 cfm, which was less that the minimum air flow rate acceptance criterion of 21,012 cfm stated in the surveillance procedure. Step 4.2 of STS PE-010A required technicians to notify operations if the control room air conditioner flow rate is less than 21,012 cfm, and to refer to Limited Condition of Operation 3.7.11 for applicable action. The inspectors determined that, rather than declaring the train inoperable and taking the actions required by technical specifications, the licensee had performed an operability evaluation that incorrectly applied instrument uncertainty and used a minimum flow rate value that had been superseded. Specifically, the inspectors determined that the surveillance acceptance criteria already accounted for instrument uncertainty, so the operability determination incorrectly applied the instrument uncertainty factor twice. Additionally the operability evaluation used a minimum air flow value of the air condition units that was taken from Revision 2 of Calculation GK-M-001. Revision 2 was no longer the current version of the calculation and the minimum air flow value had been revised to a higher value in Revision 3 of Calculation GK-M-001. When the inspectors brought this to the attention of the licensee, Condition Report 92274 was written to documeThe licensee subsequently concluded that control room air condition system train A was inoperable on February 25, 2015. The flow rate was corrected by adjusting flow dampers and re-performing the test, and returned to operable status on March 6, 2015. Analysis. use of an inadequate technical basis for an operability evaluation of a non-conforming condition resulting in the train A control room air conditioning air condition unit being declared operable when it was actually inoperable was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associate cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage). Using Inspection Manual -pectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licen-cutting aspect in the area of human performance associated with conservative bias component because the licensee did not use a decision making practice that emphasized prudent choices over those that are simply allowable in that they did not determined the proposed action to be safe in order to proceed [H.14]. Enforcement. lity be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, and drawings. Station Procedure AP--conforming conditions to be evaluated for operability. Station Procedure STS PE-010A, e control room air conditioner flow rate is less than 21,012 cfm then refer to Limited Condition of Operation 3.7.11 for applicable action for the limiting condition of operation. Specifically, the licensee completed an inadequate operability evaluation due to using incorrect data and assumptions which resulted in an inoperable system being declared operable. Because the finding was of very low safety significance (Green) and was entered into the rogram as Condition Report 92274, it is being treated as a non-cited violation in accordance with Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000482/2015001-03, Failure to Complete an Adequate Operability Evaluation for Declaring the Train A Control Room Air Conditioning Unit Operable 2. Failure to Station Boundary Watch for Opening Auxiliary Building Emergency Exhaust System Boundary Door Introduction. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, follow the requirements of Station Procedure AP 10-boundary watch when the auxiliary building emergency exhaust system boundary door 41015 was open greater than three-quarters of an inch for other than entry and exit through the door for transporting scaffolding from the turbine building to the auxiliary building. Description. On February 24, 2015, the inspectors identified door 41015 was opened multiple times during plant status walk down of the auxiliary building and turbine building while scaffolding material was moved from the turbine building to the auxiliary building. The inspector reviewed the requirements of Station Procedure AP 10-that door 41015 was a fire boundary and pressure boundary for the auxiliary building emergency exhaust system. This procedure required that if a single door opening in the auxiliary building emergency exhaust system barrier envelope was planned to be open more than 3/4-inch, it would require obtaining a breach permit. Where auxiliary building emergency exhaust system barrier envelope integrity was affected, compensatory measures were required, including stationing a dedicated individual to act as a Boundary Watch to maintain barrier operability/functionality. The plant was operating in Mode 1 at the time, so the auxiliary building emergency exhaust system was required to be operable. Thus, with the door open and no compensatory measure taken, an entry into the technical specification limiting condition for operation should have been made. The door was not breeched for greater than the technical specification allowed outage time. The inspectors informed the licensee of the issue with door 41015 and asked if a breach authorization permit was issued on February 24, 2015. The licensee determined that a breach authorization permit was not issued and initiated Condition Report 92315 into their corrective action program. The inspectors reviewed the events leading up to the door being opened, and found that maintenance and security personnel had requested a breach permit, but operations and fire protection personnel had incorrectly concluded that a breach permit was not needed. The inspectors determined that the licensee had not addressed all the boundary functions of door 41015 (specifically the pressure boundary function for auxiliary building emergency exhaust system), and had incorrectly applied the requirements of procedure AP 10-104, Section 6.7.1. The licensee entered this issue into their corrective action program for resolution as Condition Report CR 92630. The inspectors determined that the performance deficiency did not impair the high energy line break or fire protection functions. Analysis. The failure to initiate a breach permit and take required compensatory measures prior to opening auxiliary building emergency exhaust system boundary door 41015 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the system, structure, and component and barrier performance attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to ensure the radiological barrier function of the auxiliary building emergency exhaust system. Specifically, without a dedicated individual in constant communication with the control room, as required by AP 10-104, opening this door rendered the emergency exhaust system inoperable. The longest period door 41015 was open was approximately one hour without the required compensatory measure. Significance Determination Process (SDP) for Finding At-inspectors determined that the finding screened as having very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the auxiliary building. The finding has a cross-cutting aspect in the area of human performance associated with work management. The organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities [H.5]. Enforcementdocumented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, and drawings. Station Procedure AP 10-Revision 32, section 6.3.3.2.c, Where Auxiliary Building Emergency Exhaust System barrier envelope integrity is affected, one of the following compensatory measures shall be applied: (1) Utilize a Boundary Watch to maintain barrier operability/functionality in accordance with Section 6.11 of AP 26C- licensee failed to follow Station Procedure AP 10-104 while breaching the auxiliary building emergency exhaust system boundary, and activity affecting quality. Specifically, the licensee did not station a boundary watch in continuous contact with the control room to be able to rapidly close the door when the auxiliary building emergency exhaust system boundary door 41015 was open. Because the violation was of very low safety Condition Reports 92315 and 92630, it is being treated as a non-cited violation, in accordance with Sec05000482/2015001-04 - - -- - - ---- - -- -- -- - - - a. -- -- b. -- - - - The inspectors evaluated whether the licensee controlled in-plant airborne radioactivity concentrations consistent with as low as reasonably achievable (ALARA) principles and that the use of respiratory protection devices did not pose an undue risk to the wearer The engineering controls quality assurance of NIOSH certified equipment, qualification and training of personnel, and user performance 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. - a. -- -- - - b. - .1 Temporary Diesel Generator Fire a. Inspection Scope On March 11, 2015, the inspectors were informed by the control room that a temporary diesel generator inside the protected area had an approximately two foot flame emitted from the exhaust stack. The inspectors responded to the site and monitored the mporary diesel generator fire, reviewed station logs, and reviewed NUREG-compliance. b. Findings No findings were identified. .2 Event Notification 50744 Retraction a. Inspection Scope On March 18, 2015, the licensee retracted Event Notification 50744 reported on January 19, 2015, that missile door 33012 protecting Class 1E engineered safety features, buses NB01/NB02 switchgear rooms was discovered misaligned on its hinge and stuck partially open. This was reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The inspectors reviewed the basis for the retraction and reviewed NUREG-Reportin, to ensure licensee compliance. b. Findings No findings were identified. .3 (Closed) Licensee Event Report (LER) 05000482/2014-003-00: Failure of Safety Injection Accumulator Vent Line Due to Low Stress High Cycle Fatigue Results in Degraded Reactor Coolant Boundary a. Inspection Scope -Cycle Outage 20, a health physics technician observed water leaking approximately 2.5 gallons per hour from the 3/4-inch line upstream of safety injection system valve EPV0109. The leak was determined to be coming from a through-wall crack in the vent line for the combined safety injection and residual heat removal outlet piping to safety injection accumulator tank D. The cause of the through-wall cracking was determined to be to low stress high cycle fatigue. The same weld had experienced a previous failure. The evaluation of the November 2003 failure at this location had failed to include margin for vibrational impacts and variance in operational parameters resulting in inadequate corrective action to reduce vibration on the EPV0109 vent line. The immediate corrective actions called for the flawed socket weld and vent valve assembly to be replaced on April 25, 2014. Dye penetrant examinations were performed in Mid-Cycle Outage 20 on similar unsupported socket weld vent/drain assemblies connected to ASME Code Class 1 piping with no indications identified. The long term corrective action was to install a support on the EPV109 vent line during Refueling Outage 20 to reduce vibration. On October 10, 2014, NRC Problem Identification and Resolution Inspection Report 05000482/2014007 (ML14283A612), documented NCV 05000482/2014-007-to Preclude Repetition of a Significant Condition Adverse to Quality to Prevent Reactor This licensee event report was closed. b. Findings No findings were identified. .4 (Closed) Licensee Event Report (LER) 05000482/2015-001-00: Personnel Error Causes Two Inoperable Residual Heat Removal Trains a. Inspection Scope On January 28, 2015, the nightshift operations crew implemented a clearance order to support planned maintenance on residual heat removal valves EJHV8716A and EJHV8809A. At 5:34 a.m. on January 28, 2015, the oncoming crew identified that closing these valves rendered both trains of the emergency core cooling system to be inoperable. Operators entered Limiting Condition for Operation 3.0.3 and action was taken to restore valves EJHV8716A and EJH8809A to the open position. The cause of the event was that licensed operators involved with the preparation and implementation of the clearance order did not recognize that current plant conditions could not support the proposed maintenance activity. The licensee implemented the following corrective actions: (1) Individuals involved with this event had their qualifications removed until remediation occurred; (2) On January 29, 2015, the licensee issued Standing Order 36, to provide specific guidance that affect equipment operability; (2) On February 10, 2015, the electronic clearance order database was modified to identify the valves in Station Procedure AP 26C-Section A.16, that can cause entry into Limiting Condition for Operation 3.0.3; (3) On February 19, 2015, Station Procedure AP 21D-was revised to identify the use of red switch boxes for the valves in AP26C-004 entry into Limiting Condition of Operation 3.0.3. The red switch boxes were placed on the control room boards in the control room to provide awareness to the operator of the significance of the valve. b. Findings. One licensee identified finding was identified and documented in Section 4OA7 of this report. - - - - - Technical Specification Section 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures y Guide 1.33 requires procedures for equipment control (e.g. locking and tagging). Station Procedure AP 21E-, Revision 37, requires that the shift manager, ensure that plant conditions can support establishing the clearance order boundaries, including activities such as removing equipment from service. Contrary to the above, on January 28, 2015, the licensee failed to ensure that plant conditions could support the clearance order boundaries during preparation and implementation of clearance orders. Specifically, the preparation and implementation of clearance order EJ-A-005 unintentionally rendered both trains of the residual heat removal system inoperable and necessitated an unplanned entry into Technical Specification 3.0.3 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e. core damage). Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-inspectors determined a detail risk evaluation was required because this finding represented a loss of system and/or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst noted that the isolation of valve EJ HV8716A would only affect the reliability of hot leg injection for train B. Hot leg injection is a necessary function to ensure that there will not be unacceptably high concentrations of boric acid in the core region (resulting in precipitation of a solid phase) during the long-term cooling phase following a postulated large-break loss of coolant accident. Consequently, valve alignments affecting hot leg injection are only of concern during large-break loss of coolant accidents. Using the simplified plant analysis risk model, the analyst noted that the frequency of a large-break loss of coolant accident (LLOCA) was 2.5 x 10-6 /year. As stated above, the exposure period was two hours or 2.28 x 10-4 years. The analyst then calculated the upper bound risk impact of the performance deficiency to be 5.7 x 10-10. Therefore, this finding is of very low safety significance (Green).
A-1 Attachment 1 P. Black, Human Resources T. Damashek, Simulator Fidelity Coordinator P. Deblonk, Superintendent, Instrumentation and Control T. East, Supervisor, Emergency Planning J. Edwards, Manager, Operations D. Erbe, Manager, Security R. Flannigan, Manager, Nuclear Engineering A. Heflin, President and Chief Executive Officer J. Knapp, Superintendent, Operations Training S. Koenig, Manager, Regulatory Affairs C. Reasoner, Site Vice President B. Ryan, Licensed Operator Supervising Instructor M. Skyles, Manager, Health Physics S. Smith, Plant Manager A1-2 05000482/2015001-02 NCV Failure to Assess the Operability of Emergency Diesel Generator B during Emergent Work Activities (1R13)05000482/2015001-03 Failure to Complete an Adequate Operability Evaluation for Declaring the Train A Control Room Air Conditioning Unit Operable (1R15.1)05000482/2015001-04 Failure to Station Boundary Watch for Opening Auxiliary Building Emergency Exhaust System Boundary Door (1R15.2) 05000482/2014-003-00 Failure of Safety Injection Accumulator Vent Line Due to Low Stress High Cycle Fatigue Results in Degraded Reactor Coolant Boundary (4OA3.3) 05000482/2015-001-00 Personnel Error Causes Two Inoperable Residual Heat Removal Trains (4OA3.4) - -
A1-3 - - - - M-12EF01 M-12EF02 M-12AL01 M-12EN01 M-12BB01 M-12BB02 - - - -
A1-4 - -- -- -- - - - - - - - - - - - - - - - - - - -
A1-5 - - - - - - - --- -- - --- -- -- -
A1-6 - - - - - - - - - - - - - -
A1-7 - - - - - - - -- -- - -- - - -- -- GK-M-001 Cooling and Heating Load Calculation for Control Room HVAC System Capabilities During Normal Plant Operation and Accident Conditions (SGK04A/B) - Operability Evaluations A1-8 AP 26C-004 Operability Determination and Functionality Assessment - Breach Authorization ALR 00-047E RWST Level HILO ALR 00-077E SR HI Volt Fail SYS EC-121 Recirc of the RWST through the Fuel Pool Cleanup System M-10BN Borated Refueling Water Storage M-12BN01 Piping and Instrumentation Diagram Borated Refueling Water Storage System - - - - - -
A1-9 - - - - - - - - AP 29G-001 RCS Unidentified Leak Rate Monitoring Program STS BB-006 RCS Water Inventory Balance Using the NPSI Computer STS BB-201B Cycle Test of PORV Block Valve BB HV-8000B STS CR-001 Shift Log for Modes 1, 2, & 3 STS EJ-201 Train B RHR System Inservice Valve Test STS EM-201B Safety Injection System Train B Inservice Valve Test STS KJ-015A Manual/Auto Fast Start, Sync & Loading of EDG NE01 M-12BB01 Piping & Instrumentation Diagram Reactor Coolant System - Piping & Instrumentation Diagram Reactor Coolant System M-12EJ01 Piping and Instrumentation Diagram Residual Heat Removal System M-12EM01 Piping & Instrumentation Diagram High Pressure Coolant Injection System A1-10 - - Emergency Action Level and Emergency Plan Changes - - - - - - - - - -
A1-11 Audits and Self-Assessments Number Title Date -- 2015-0736 QA Surveillance: RP Refueling Outage 2015-0738 QA Surveillance: RP Department Rebuilding Plans December 26, 2014 2015-0740 QA Surveillance: 4th Quarter RP 2014-0816 Quick Hit Assessment: PWR Power Entries June 17, 2014 2014-0890 Quick Hit Assessment: HP 5-Year Plan September 26, 2014 2014-0894 Quick Hit Assessment: Protection of Category 1&2 RAM October 2, 2014 - - - - - --
A1-12 - - - - - - - - - -- - -- - - - - - - -- - -- - - - --
A1-13 - - - -- - - - - - - --
A1-14 - - Operability Determination and Functionality Assessment A2-1 Attachment 2 The following items are requested for the Occupational Radiation Safety Inspection at Wolf Creek March 9 thru 13, 2015 Integrated Report 2015001 Inspection areas are listed in the attachments below. Please provide the requested information on or before February 19, 2015. 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 1- 1- If information is placed on ims.certrec.com, please ensure the inspection exit date entered is at least 30 days later than the onsite 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 at (817) 200-1441 or john.odonnell@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 Date of Last Inspection: March 17, 2014 A. List of contacts and telephone numbers for the Radiation Protection Organization Staff and Technicians B. Applicable organization charts C. Audits, self-assessments, and 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 sub-tiered systems) since date of last inspection a. Initiated by the radiation protection organization b. Assigned to the radiation protection organization NOTE: The lists should indicate the significance level of each issue and the search criteria used. Please provide in document formats inspector 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 IP 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 a. All radioactive sources that are required to be leak tested b. All radioactive sources that meet the 10 CFR Part 20, Appendix E, Category 2 and above threshold. Please indicate the radioisotope, initial and current activity (w/assay date), and storage location for each applicable source.
A2-3 J. The last two leak test results for the radioactive sources inventoried and required to be leak tested. If applicable, specifically provide a list of all radioactive source(s) that have failed its leak test within the last two years K. A current listing of any non-fuel items stored within your pools, and if available, their appropriate dose rates (Contact / @ 30cm) L. Computer printout of radiological controlled area entries greater than 100 millirems since the previous inspection to the current inspection entrance date. The printout should include the date of entry, some form of worker identification, the radiation work permit used by the worker, dose accrued by the worker, and the electronic dosimeter dose alarm setpoint used during the entry (for Occupational Radiation Safety Performance Indicator verification in accordance with IP 71151).
A2-4 3. In-Plant Airborne Radioactivity Control and Mitigation (71124.03) Date of Last Inspection: February 4, 2013 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 (SCBA), and LERs, written since date of last inspection related to: 1. Installed air filtration systems 2. Self-contained breathing apparatuses 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 self-contained breathing apparatuses 3. Air quality testing for SCBAs 4. Use of installed plant systems, such as containment purge, spent fuel pool ventilation, and auxiliary building ventilation F. A summary list of corrective action documents (including corporate and sub-tiered systems) written since date of last inspection, related to the Airborne Monitoring program including: 1. Continuous air monitors 2. Self-contained breathing apparatuses 3. Respiratory protection program NOTE: The lists should indicate the significance level of each issue and the search criteria used. Please provide in document formainspector can perform word searches. G. List of SCBA qualified personnel - reactor operators and emergency response personnel H. Inspection records for self-contained breathing apparatuses (SCBAs) staged in the plant for use since date of last inspection. I. SCBA training and qualification records for control room operators, shift supervisors, STAs, and OSC 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 J. List of respirators (available for use) by type (APR, SCBA, PAPR, etc.), manufacturer, and model.