IR 05000263/2017004
| ML18030B023 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/30/2018 |
| From: | Kenneth Riemer Division Reactor Projects III |
| To: | Church C Nuclear Management Co |
| References | |
| IR 2017004, IR 2017501 | |
| Download: ML18030B023 (40) | |
Text
UNITED STATES January 30, 2018
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANTNRC INTEGRATED INSPECTION REPORT 05000263/2017004 AND EMERGENCY PREPAREDNESS ANNUAL INSPECTION REPORT 05000263/2017501
Dear Mr. Church:
On December 31, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Monticello Nuclear Generating Plant. On January 16, 2018, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report. The NRC also completed its annual inspection of the Emergency Preparedness Program. This inspection began on January 1, 2017, and the issuance of this letter closed Inspection Report 05000263/2017501.
Based on the results of this inspection, one self-revealed Finding was evaluated under the risk significance determination process as having very low safety significance (Green). The finding did not involve a violation of NRC requirements and is documented in this report. Further, inspectors documented a licensee-identified violation which was determined to be of very low safety significance. The NRC is treating this violation as an NCV consistent with Section 2.3.2.a of the Enforcement Policy.
If you contest the violation or significance of the NCV, 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 III; the Director, Office of Enforcement, and the NRC Resident Inspector at the Monticello Nuclear Generating Plant.
If you disagree with the cross-cutting aspect assignment or a finding not associated 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 U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region III; and the NRC Resident Inspector at the Monticello Nuclear Generating Plant. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Kenneth Riemer, Chief Branch 2 Division of Reactor Projects Docket No. 50-263 License No. DPR-22 Enclosure:
IR 05000263/2017004; 05000263/2017501 cc: Distribution via LISTSERV
SUMMARY
Inspection Report 05000263/2017004; 10/01/2017-12/31/2017; Monticello Nuclear Generating
Plant; Occupational As-Low-As-Reasonably-Achievable (ALARA) Planning and Controls.
This report covers a 3-month period of inspection by resident inspectors and announced baseline inspections by regional inspectors. One self-revealed Green finding was identified.
The significance of inspection findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609,
Significance Determination Process, dated April 29, 2015. Cross-cutting aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated November 1, 2016. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision
NRC-Identified
and Self-Revealed Findings
Cornerstone: Occupational Radiation Safety
- Green: A finding of very low safety significance (Green) was self-revealed due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees radiological work planning and work control program. Specifically, the licensee failed to properly incorporate ALARA strategies, insights while planning, and executing work activities during the 1R28 refueling outage.
The Reactor Water Cleanup (RWCU) Inlet Outboard Isolation Valve MO-2398 was scheduled for replacement during the outage. The initial dose estimate for this activity was 4.5 person-rem. However, 13.776 actual person-rem of dose was received. This issue was caused by poor radiological planning and work execution of this task. The licensee entered this issue into their Corrective Action Program (CAP) item 1558234.
The finding was more than minor because it was associated with the program and process attribute of the Occupation Radiation Safety Cornerstone. Additionally, this issue affected the cornerstone objective of ensuring the adequate protection of the workers health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Additionally, the finding is very similar to IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009,
Example 6.i. This example provides guidance that an issue is not minor if the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors determined that this finding was of very low safety significance (Green) because Monticello Nuclear Generating Plants current 3-year rolling average collective is 64.637 person-rem (2014-2016). This is less than the 240 person-rem/unit referenced within IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. This finding had a cross-cutting aspect in the area of Human Performance, related to the cross-cutting aspect of Work Management, in that the outage plan did not adequately plan, control and execute work activities to ensure the RWCU Inlet Outboard Isolation Valve MO-2398 replacement remained ALARA. [H.5] (Section 2RS2)
Licensee-Identified
- Violations of very low safety or security significance or Severity Level IV that were identified by the licensee have been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensees CAP. These violations and CAP tracking numbers are listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant Status
Monticello began the inspection period operating at approximately 100 percent power and operated at or near full power for the reminder of the inspection period, with the following exceptions:
- December 16, 2017 to December 17, 2017Power was reduced to approximately 75 percent for control rod scram time testing and quarterly turbine testing with subsequent return to 100 percent.
REACTOR SAFETY
Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity, and
1R04 Equipment Alignment
.1 Quarterly Partial System Walkdowns
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant systems:
- 12 emergency diesel generator (EDG) during 11 EDG Maintenance Work Window;
- High Pressure Coolant Injection (HPCI) during RCIC Maintenance Work Window; and
- B Residual Heat Removal (RHR) Room during A Core Spray Maintenance.
The inspectors selected these systems based on their risk significance relative to the Reactor Safety Cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could impact the function of the system and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Final Safety Analysis Report, Technical Specification (TS)requirements, outstanding work orders (WOs), condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions.
The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable.
The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies.
The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.
These activities constituted three partial system walkdown samples as defined in Inspection Procedure (IP) 71111.04-05.
b. Findings
No findings were identified.
1R05 Fire Protection
.1 Routine Resident Inspector Tours
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:
- Fire Zone 21-A; Radwaste Control Room;
- Fire Zone 21-B; Radwaste Trash Compactor Area;
- Fire Zone 21-C; Radwaste Shipping Building;
- Fire Zone 21-D; Radwaste Building;
- Fire Zone 19-C; FW Pipe Chase; and
- Fire Zone 3-02E; TIP Room.
The inspectors reviewed these areas to assess if the licensee had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant, effectively maintained fire detection and suppression capability, maintained passive fire protection features in good material condition, and implemented adequate compensatory measures for out-of-service, degraded or inoperable fire protection equipment, systems, or features in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to impact equipment which could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event.
Using the documents listed in the Attachment to this report, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees CAP.
Documents reviewed are listed in the Attachment to this report.
These activities constituted six quarterly fire protection inspection samples as defined in IP 71111.05-05.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program
.1 Resident Inspector Quarterly Review of Licensed Operator Requalification
a. Inspection Scope
On October 23, 2017, the inspectors observed a crew of licensed operators in the plants simulator during licensed operator requalification training. The inspectors verified that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and that training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:
- licensed operator performance;
- crews clarity and formality of communications;
- ability to take timely actions in the conservative direction;
- prioritization, interpretation, and verification of annunciator alarms;
- correct use and implementation of abnormal and emergency procedures;
- control board manipulations;
- oversight and direction from supervisors; and
- ability to identify and implement appropriate TS actions and Emergency Plan actions and notifications.
The crews performance in these areas was compared to pre-established operator action expectations and successful critical task completion requirements. Documents reviewed are listed in the Attachment to this report.
This inspection constituted one quarterly licensed operator requalification program simulator sample as defined in IP 71111.11-05.
b. Findings
No findings were identified.
.2 Resident Inspector Quarterly Observation During Periods of Heightened Activity or Risk
a. Inspection Scope
On December 16, 2017, the inspectors observed licensed control room operators during reactor downpower activities for control rod scram time testing and quarterly turbine testing. This was an activity that required heightened awareness or was related to increased risk. The inspectors evaluated the following areas:
- licensed operator performance;
- crews clarity and formality of communications;
- ability to take timely actions in the conservative direction;
- prioritization, interpretation, and verification of annunciator alarms;
- correct use and implementation of procedures;
- control board manipulations; and
- oversight and direction from supervisors.
The performance in these areas was compared to pre-established operator action expectations, procedural compliance and task completion requirements. Documents reviewed are listed in the Attachment to this report.
This activity constituted one quarterly licensed operator heightened activity/risk sample as defined in IP 71111.11-05.
b. Findings
No findings were identified.
.3 Annual Operating Test Results
a. Inspection Scope
The inspectors reviewed the overall pass/fail results of the Annual Operating Test, administered by the licensee from October 2, 2017 through November 9, 2017, required by Title 10 of the Code of Federal Regulations (CFR) 55.59(a). The results were compared to the thresholds established in Inspection Manual Chapter 0609, Appendix I, Licensed Operator Requalification Significance Determination Process, to assess the overall adequacy of the Licensees Operator Requalification Training (LORT) program to meet the requirements of 10 CFR 55.59. (02.02)
This inspection constituted one annual licensed operator requalification examination results sample as defined in Inspection Procedure (IP) 71111.11-05.
b. Findings
No findings were identified.
.4 Biennial Review
a. Inspection Scope
The following inspection activities were conducted during the weeks of October 9, 2017 and October 16, 2017, to assess:
- (1) the effectiveness and adequacy of the facility licensees implementation and maintenance of its systems approach to training (SAT)based Licensed LORT program, put into effect to satisfy the requirements of 10 CFR 55.59;
- (2) conformance with the requirements of 10 CFR 55.46 for use of a plant referenced simulator to conduct operator licensing examinations and for satisfying experience requirements; and
- (3) conformance with the operator license conditions specified in 10 CFR 55.53. The documents reviewed are listed in the Attachment to this report.
- Licensee Requalification Examinations (10 CFR 55.59(c); SAT Element 4 as Defined in 10 CFR 55.4): The inspectors reviewed the licensees program for development and administration of the LORT biennial written examination and annual operating tests to assess the licensees ability to develop and administer examinations that are acceptable for meeting the requirements of 10 CFR 55.59(a).
- The inspectors conducted a detailed review of one version of the 2017 biennial requalification written examination administered to Crew 3, the week of October 30, 2017, to assess content, level of difficulty, and quality of the written examination materials. (02.03)
- The inspectors conducted a detailed review of ten job performance measures (JPMs) and four simulator scenarios to assess content, level of difficulty, and quality of the operating test materials. (02.04)
- The inspectors observed the administration of the annual operating test to assess the licensees effectiveness in conducting the examinations, including the conduct of pre-examination briefings, evaluations of individual operator and crew performance, and post-examination analysis. The inspectors evaluated the performance of one shift crew split into two simulator operating crews in parallel with the facility evaluators during four dynamic simulator scenarios, and evaluated various licensed crew members concurrently with facility evaluators during the administration of several JPMs. (02.05)
- The inspectors assessed the adequacy and effectiveness of the remedial training conducted since the last requalification examinations and the training planned for the current examination cycle to ensure that they addressed weaknesses in licensed operator or crew performance identified during training and plant operations. The inspectors reviewed remedial training procedures and individual remedial training plans. (02.07)
- Conformance with Examination Security Requirements (10 CFR 55.49):
The inspectors conducted an assessment of the licensees processes related to examination physical security and integrity (e.g., predictability and bias) to verify compliance with 10 CFR 55.49, Integrity of Examinations and Tests. The inspectors reviewed the facility licensees examination security procedure, and observed the implementation of physical security controls (e.g., access restrictions and simulator input/output controls) and integrity measures (e.g.,
security agreements, sampling criteria, bank use, and test item repetition)throughout the inspection period. (02.06)
- Conformance with Operator License Conditions (10 CFR 55.53): The inspectors reviewed the facility licensee's program for maintaining active operator licenses and to assess compliance with 10 CFR 55.53(e) and (f). The inspectors reviewed the procedural guidance and the process for tracking on-shift hours for licensed operators, and which control room positions were granted watch-standing credit for maintaining active operator licenses. Additionally, medical records for ten licensed operators were reviewed for compliance with 10 CFR 55.53(I). (02.08)
- Conformance with Simulator Requirements Specified in 10 CFR 55.46:
The inspectors assessed the adequacy of the licensees simulation facility (simulator) for use in operator licensing examinations and for satisfying experience requirements. The inspectors reviewed a sample of simulator performance test records (e.g., transient tests, malfunction tests, scenario based tests, post-event tests, steady state tests, and core performance tests), simulator discrepancies, and the process for ensuring continued assurance of simulator fidelity in accordance with 10 CFR 55.46. The inspectors reviewed and evaluated the discrepancy corrective action process to ensure that simulator fidelity was being maintained. Open simulator discrepancies were reviewed for importance relative to the impact on 10 CFR 55.45 and 55.59 operator actions as well as on nuclear and thermal hydraulic operating characteristics. (02.09)
- Problem Identification and Resolution (10 CFR 55.59(c); SAT Element 5 as Defined in 10 CFR 55.4): The inspectors assessed the licensees ability to identify, evaluate, and resolve problems associated with licensed operator performance (a measure of the effectiveness of its LORT program and their ability to implement appropriate corrective actions to maintain its LORT Program up-to-date). The inspectors reviewed documents related to licensed operator performance issues (e.g., recent examination and inspection reports including cited and non-cited violations; NRC End-of-Cycle and Mid-Cycle reports; NRC plant issue matrix; licensee event reports; licensee condition/problem identification reports including documentation of plant events and review of industry operating experience). The inspectors also sampled the licensees quality assurance oversight activities, including licensee training department self-assessment reports. (02.10)
This inspection constituted one biennial licensed operator requalification program inspection sample as defined in IP 71111.11-05.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
.1 Routine Quarterly Evaluations
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following risk-significant systems:
- 12 EDG Maintenance; and
- MO-2034 HPCI Inboard Steam Isolation Valve Backseating.
The inspectors reviewed events such as where ineffective equipment maintenance had resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:
- implementing appropriate work practices;
- identifying and addressing common cause failures;
- scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;
- characterizing system reliability issues for performance;
- charging unavailability for performance;
- trending key parameters for condition monitoring;
- ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and
- verifying appropriate performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2), or appropriate and adequate goals and corrective actions for systems classified as (a)(1).
The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the CAP with the appropriate significance characterization. Documents reviewed are listed in the Attachment to this report.
These activities constituted two quarterly maintenance effectiveness samples as defined in IP 71111.12-05.
b. Findings
No findings were identified.
.2 Quality Control
a. Inspection Scope
On October 23, 2017, the inspectors reviewed the licensees quality control activities through a review of the licensees control of quality parts during maintenance associated with V-EF-18A Standby Gas Treatment Off Gas Stack Motor Replacement. The following work package was reviewed for proper quality controls:
- WO 700005779; V-EF-18A Standby Gas Treatment Off Gas Stack Motor Replacement.
This activity constituted completion of one quality control maintenance effectiveness sample, as defined in Inspection Procedure 71111.12.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
.1 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed the licensee's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:
- WO 700026387 RCIC Steam Line Outboard Isolation Valve MO-2076.
These activities were selected based on their potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate and complete. When emergent work was performed, the inspectors verified that the plant risk was promptly reassessed and managed. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.
Documents reviewed during this inspection are listed in the Attachment to this report.
This activity constituted one maintenance risk assessment and emergent work sample as defined in IP 71111.13-05.
b. Findings
No findings were identified.
1R15 Operability Determinations and Functional Assessments
.1 Operability Evaluations
a. Inspection Scope
The inspectors reviewed the following issues:
- CAP 501000003873 - Unexpected Annunciator 125 VDC Bus Ground;
- CAP 501000006215 - Air Leak on #11 Air Start Solenoid.
The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and updated safety analysis report (USAR) to the licensees evaluations to determine whether the components or systems were operable.
Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations.
Documents reviewed are listed in the Attachment to this report.
These operability inspections constituted three samples as defined in IP 71111.15-05.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing
.1 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed the following post-maintenance (PM) activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:
- Work Order (WO) 700012684-0020; V-EF-12 Charcoal Filter Heater for 480 volts Supply;
These activities were selected based upon the structure, system, or component's ability to impact risk. The inspectors evaluated these activities for the following (as applicable):
the effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed; acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate; tests were performed as written in accordance with properly reviewed and approved procedures; equipment was returned to its operational status following testing (temporary modifications or jumpers required for test performance were properly removed after test completion); and test documentation was properly evaluated. The inspectors evaluated the activities against TSs, the Final Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with post-maintenance tests to determine whether the licensee was identifying problems and entering them in the Corrective Action Program (CAP) and that the problems were being corrected commensurate with their importance to safety.
Documents reviewed are listed in the Attachment to this report.
These activities constituted five post-maintenance testing samples as defined in IP 71111.19-05.
b. Findings
No findings were identified.
1R22 Surveillance Testing
.1 Surveillance Testing
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether risk-significant systems and equipment were capable of performing their intended safety function and to verify testing was conducted in accordance with applicable procedural and TS requirements:
- 0089, Boron Concentration Standby Liquid Control System [ROUTINE];
- OSP-EFT-0596 Control Room Envelop In-Leakage Test [ROUTINE]; and
- 0058 HPCI Steam Line High Area Temperature Test and Calibration Procedure
[ROUTINE].
The inspectors observed in-plant activities and reviewed procedures and associated records to determine the following:
- did preconditioning occur;
- the effects of the testing were adequately addressed by control room personnel or engineers prior to the commencement of the testing;
- acceptance criteria were clearly stated, demonstrated operational readiness, and were consistent with the system design basis;
- plant equipment calibration was correct, accurate, and properly documented;
- as-left setpoints were within required ranges; and the calibration frequency was in accordance with TSs, the USAR, procedures, and applicable commitments;
- measuring and test equipment calibration was current;
- test equipment was used within the required range and accuracy; applicable prerequisites described in the test procedures were satisfied;
- test frequencies met TS requirements to demonstrate operability and reliability; tests were performed in accordance with the test procedures and other applicable procedures; jumpers and lifted leads were controlled and restored where used;
- test data and results were accurate, complete, within limits, and valid;
- test equipment was removed after testing;
- where applicable for in-service testing activities, testing was performed in accordance with the applicable version of Section XI, American Society of Mechanical Engineers code, and reference values were consistent with the system design basis;
- where applicable, test results not meeting acceptance criteria were addressed with an adequate operability evaluation or the system or component was declared inoperable;
- where applicable for safety-related instrument control surveillance tests, reference setting data were accurately incorporated in the test procedure;
- where applicable, actual conditions encountering high resistance electrical contacts were such that the intended safety function could still be accomplished;
- prior procedure changes had not provided an opportunity to identify problems encountered during the performance of the surveillance or calibration test;
- equipment was returned to a position or status required to support the performance of its safety functions; and
- all problems identified during the testing were appropriately documented and dispositioned in the CAP.
Documents reviewed are listed in the Attachment to this report.
These inspections constituted three routine surveillance testing sample(s) as defined in IP 71111.22, Sections-02 and-05.
b. Findings
No findings were identified.
1EP4 Emergency Action Level and Emergency Plan Changes
a. Inspection Scope
The regional inspectors performed an in-office review of the latest revisions to the Emergency Plan, Emergency Action Levels (EALs), and EAL Bases document to determine if these changes decreased the effectiveness of the Emergency Plan.
The inspectors also performed a review of the licensees 10 CFR Part 50.54(q) change process, and Emergency Plan change documentation to ensure proper implementation for maintaining Emergency Plan integrity.
The U.S. Nuclear Regulatory Commission review was not documented in a safety evaluation report, and did not constitute approval of licensee-generated changes; therefore, this revision is subject to future inspection. The specific documents reviewed during this inspection are listed in the Attachment to this report.
This EAL and Emergency Plan Change inspection constituted one sample as defined in Inspection Procedure 71114.04.
b. Findings
No findings were identified.
RADIATION SAFETY
2RS2 Occupational As-Low-As-Reasonably-Achievable Planning and Controls
.1 Radiological Work Planning (02.02)
a. Inspection Scope
The inspectors selected three to five work activities of the highest exposure significance or involve work in high dose rate areas.
The inspectors reviewed the radiological work planning as-low-as-reasonably-achievable (ALARA) evaluations, initial and revised exposure estimates, and exposure mitigation requirements. The inspectors determined if the licensee had reasonably grouped the radiological work into work activities.
The inspectors assessed whether the licensees planning identified appropriate dose reduction techniques; appropriately considered alternate reduction features; and defined reasonable dose goals. The inspectors evaluated whether the licensees ALARA assessment had taken into account decreased worker efficiency from use of respiratory protective devices and/or heat stress mitigation equipment. The inspectors determined if the licensees work planning considered the use of remote technologies and dose reduction insights from industry and plant-specific operating experience. The inspectors assessed whether these ALARA requirements were integrated into work procedure and/or radiation work permit documents.
The inspectors compared the results achieved with the intended dose established in the ALARA planning. The inspectors compared the person-hour estimates provided by work groups to the radiation protection group with the actual work activity time results, and evaluated the accuracy of these time estimates. The inspectors evaluated the reasons for any inconsistencies between intended and actual work activity doses.
The inspectors evaluated whether post-job reviews were conducted to identify lessons learned and entered into the licensees corrective action program.
These inspection activities supplemented those documented in Inspection Report 05000263/2017002 and constituted one complete sample as defined in IP 71124.02-05.
b. Findings
Introduction:
A finding of very low safety significance (Green) was self-revealed due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees radiological work planning and work control program. Specifically, the licensee failed to properly incorporate ALARA strategies, insights while planning, and executing work activities during the 1R28 refueling outage.
The Reactor Water Cleanup (RWCU) Inlet Outboard Isolation Valve MO-2398 was scheduled for replacement during the outage. The initial dose estimate for this activity was 4.5 person-rem. However, 13.776 actual person-rem of dose was received. This issue was caused by poor radiological planning and work execution of this task.
Description:
During the 1R28 refueling outage, numerous work tasks were performed.
One of these was the replacement of RWCU Inlet Outboard Isolation Valve MO-2398.
The original estimate for this task was 4.5 person-rem. The licensee performed a number of Work In Progress reviews for this valve replacement activity between 04/13/2017 and 05/02/2017. During this time period, it became apparent to the licensee that the percentage of work completed was not tracking with original dose estimates.
The Station ALARA committee revised the dose estimate for this task as follows:
- 04/22/2017 9.410 person-rem
- 04/25/2017 11.3 person-rem
- 05/02/2017 14.2 person-rem The task received 13.776 person-rem. On July 21, 2017, the licensee completed an Apparent Cause Evaluation (ACE) to determine the inaccurate dose estimate for the valve replacement project. The ACE listed a number of causes for this task exceeding the dose goal. These included incorrect welds, which required re-work, changes in the scaffold design to allow work completion, incomplete equipment to perform tests, incomplete work order instructions and installed shielding which was less than the approved design.
Analysis:
Procedure FP-RP-SEN-02 established requirements to incorporate ALARA or insights while planning work. Contrary to this, the licensee failure to appropriately plan and coordinate outage activities, together with the failure to properly incorporate ALARA strategies or insights while planning and executing the replacement of the of RWCU Inlet Outboard Isolation Valve MO-2398 during the 1R28 refueling outage, was a performance deficiency that was within the licensees ability to control and should have been prevented. The finding was more than minor because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone.
Additionally, this issue affected the cornerstone objective of ensuring the adequate protection of the workers health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Additionally, the finding is very similar to IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009.
Example 6.i provides guidance that an issue is not minor if the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors determined that this finding was of very low safety significance (Green) in accordance with IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. This was a finding in the ALARA Planning and Work controls but not greater than the 240 person-rem/unit for a boiling-water reactor. The licensees current collective 3-year rolling average was 64.637 person-rem (2014-2016). This finding had a cross-cutting aspect in the area of Human Performance, Work Management in that the outage plan did not adequately plan, control and execute work activities to ensure the MO-2398 replacement remained ALARA. [H.5]
Enforcement:
No violation of regulatory requirements occurred. This is considered a finding of very low safety significance (Green) (FIN 05000263/2017004-01, Failure to Maintain Radiation Exposure ALARA). The licensee has entered this issue into their Corrective Action Program as CAP 1558234.
.2 Verification of Dose Estimates and Exposure Tracking Systems (02.03)
a. Inspection Scope
The inspectors reviewed selected occasions with inconsistent or incongruent results from the licensees intended radiological outcomes to determine whether the cause was attributed to a failure to adequately plan work activities, or failure to provide sufficient management oversight of in-plant work activities, or failure to conduct the work activity without significant rework, or failure to implement radiological controls as planned.
These inspection activities supplemented those documented in Inspection Report 05000263/2017002 and constituted one complete sample as defined in IP 71124.02-05.
b. Findings
No findings were identified.
.3 Implementation of As-Low-As-Reasonably-Achievable and Radiological Work Controls
(02.04)
a. Inspection Scope
The inspectors compared the radiological results achieved with the intended radiological outcomes and verified that the licensee captured lessons learned for use in the next outage.
These inspection activities supplemented those documented in Inspection Report 05000263/2017002 and constituted one complete sample as defined in IP 71124.02-05.
b. Findings
No findings were identified.
.4 Problem Identification and Resolution (02.06)
a. Inspection Scope
The inspectors reviewed self-assessments and/or audits performed of the ALARA program and determined if these reviews identified problems or areas for improvement.
The inspectors assessed whether problems associated with ALARA planning and controls were being identified by the licensee at an appropriate threshold and properly addressed for resolution.
These inspection activities supplemented those documented in Inspection Report 05000263/2017002 and constituted one complete sample as defined in IP 71124.02-05.
b. Findings
No findings were identified.
2RS7 Radiological Environmental Monitoring Program
.1 Site Inspection (02.02)
a. Inspection Scope
The inspectors assessed whether the meteorological instruments were operable, calibrated, and maintained in accordance with guidance contained in the Final Safety Analysis Report, U.S. Nuclear Regulatory Commission Regulatory Guide 1.23, Meteorological Monitoring Programs for Nuclear Power Plants, and licensee procedures. The inspectors assessed whether the meteorological data readout and recording instruments were operable.
These inspection activities supplemented those documented in Inspection Report 05000263/2017003 and constituted one complete sample as defined in IP 71124.07-05.
b. Findings
No findings were identified.
OTHER ACTIVITIES
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security
4OA1 Performance Indicator Verification
.1 Mitigating Systems Performance Index-Residual Heat Removal System
a. Inspection Scope
The inspectors sampled licensee submittals for the Mitigating Systems Performance Index (MSPI) - Residual Heat Removal System performance indicator for the period from the fourth quarter 2016 through third quarter 2017. To determine the accuracy of the Performance Indicator (PI) data reported during those periods, PI definitions and guidance contained in the Nuclear Energy Institute (NEI) Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013, were used. The inspectors reviewed the licensees operator narrative logs, issue reports, MSPI derivation reports, event reports and NRC Integrated Inspection Reports for the period of October 1, 2016 through September 30, 2017, to validate the accuracy of the submittals. The inspectors reviewed the MSPI component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance.
The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the Attachment to this report.
This activity constituted one MSPI residual heat removal system sample as defined in IP 71151-05.
b. Findings
No findings were identified.
.2 Mitigating Systems Performance IndexCooling Water Systems
a. Inspection Scope
The inspectors sampled licensee submittals for the MSPI - Cooling Water Systems performance indicator for the period from the fourth quarter 2016 through third quarter 2017. To determine the accuracy of the PI data reported during those periods, PI definitions and guidance contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 31, 2013 were used. The inspectors reviewed the licensees operator narrative logs, issue reports, MSPI derivation reports, event reports and NRC Integrated Inspection Reports for the period of October 1, 2016 through September 30, 2017 to validate the accuracy of the submittals. The inspectors reviewed the MSPI component risk coefficient to determine if it had changed by more than 25 percent in value since the previous inspection, and if so, that the change was in accordance with applicable NEI guidance. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator and none were identified. Documents reviewed are listed in the Attachment to this report.
This activity constituted one MSPI cooling water system sample as defined in IP 71151-05.
b. Findings
No findings were identified.
.3 Reactor Coolant System Specific Activity
a. Inspection Scope
During third quarter 2017 inspection activities, the inspectors sampled licensee submittals for the reactor coolant system specific activity PI for Monticello Nuclear Generating Plant for the period from the third quarter 2016 through the second quarter 2017. The inspectors used PI definitions and guidance contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 8, 2013, to determine the accuracy of the PI data reported during those periods. The inspectors reviewed the licensees reactor coolant system chemistry samples, TS requirements, issue reports, event reports and U.S. Nuclear Regulatory Commission Integrated Inspection Reports to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the PI data collected or transmitted for this indicator. In addition to record reviews, the inspectors observed a chemistry technician obtain and analyze a reactor coolant system sample. Documents reviewed are listed in the Attachment to this report.
This inspection constituted one reactor coolant system specific activity sample as defined in IP 71151-05.
b. Findings
No findings were identified.
.4 Occupational Exposure Control Effectiveness
a. Inspection Scope
During third quarter 2017 inspection activities, the inspectors sampled licensee submittals for the Occupational Exposure Control Effectiveness PI for the period from the third quarter 2016 through the second quarter 2017. The inspectors used PI definitions and guidance contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 2013, to determine the accuracy of the PI data reported during those periods. The inspectors reviewed the licensees assessment of the PI for occupational radiation safety to determine if indicator related data was adequately assessed and reported. To assess the adequacy of the licensees PI data collection and analyses, the inspectors discussed with radiation protection staff, the scope and breadth of its data review and the results of those reviews. The inspectors independently reviewed electronic personal dosimetry dose rate and accumulated dose alarms and dose reports and the dose assignments for any intakes that occurred during the time period reviewed to determine if there were potentially unrecognized occurrences. The inspectors also conducted walkdowns of numerous locked high and very-high radiation area entrances to determine the adequacy of the controls in place for these areas. Documents reviewed are listed in the to this report.
This inspection constituted one occupational exposure control effectiveness sample as defined in IP 71151-05.
b. Findings
No findings were identified.
.5 Radiological Effluent Technical Specification/Offsite Dose Calculation Manual
Radiological Effluent Occurrences
a. Inspection Scope
During third quarter 2017 inspection activities, the inspectors sampled licensee submittals for the radiological effluent Technical Specification(TS)/off-site dose calculation maual (ODCM) radiological effluent occurrences PI for the period from the third quarter 2016 through the second quarter 2017. The inspectors used PI definitions and guidance contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, dated August 2013, to determine the accuracy of the PI data reported during those periods. The inspectors reviewed the licensees issue report database and selected individual reports generated since this indicator was last reviewed to identify any potential occurrences such as unmonitored, uncontrolled, or improperly calculated effluent releases that may have impacted offsite dose. The inspectors reviewed gaseous effluent summary data and the results of associated offsite dose calculations for selected dates to determine if indicator results were accurately reported. The inspectors also reviewed the licensees methods for quantifying gaseous and liquid effluents and determining effluent dose. Documents reviewed are listed in the Attachment to this report.
This inspection constituted one Radiological Effluent TS/ODCM radiological effluent occurrences sample as defined in IP 71151-05.
b. Findings
No findings were identified.
4OA2 Identification and Resolution of Problems
.1 Routine Review of Items Entered into the Corrective Action Program
a. Inspection Scope
As discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify they were being entered into the licensees corrective action program at an appropriate threshold, adequate attention was being given to timely corrective actions, and adverse trends were identified and addressed. Some minor issues were entered into the licensees corrective action program as a result of the inspectors observations; however, they are not discussed in this report.
These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure they were considered an integral part of the inspections performed during the quarter.
b. Findings
No findings were identified.
.2 Annual Follow-Up of Selected Issues: [Circulating Water intake Basin Icing]
a. Inspection Scope
The inspectors selected the following condition reports for in-depth review:
- CAP 501000006364; Icing in the Intake.
The inspectors selected this issue for review to determine whether the licensee had previous opportunities to enhance procedures from a pre-emptive environmental condition perspective.
As appropriate, the inspectors verified the following attributes during their review of the licensee's corrective actions for the above condition reports and other related condition reports:
- complete and accurate identification of the problem in a timely manner commensurate with its safety significance and ease of discovery;
- consideration of the extent of condition, generic implications and previous occurrences;
- classification and prioritization of the resolution of the problem commensurate with safety significance;
- identification of corrective actions, which were appropriately focused to correct the problem; and
- completion of corrective actions in a timely manner commensurate with the safety significance of the issue.
The inspectors discussed the corrective actions and associated evaluations with licensee personnel.
This activity constituted one in-depth problem identification and resolution inspection sample as defined in IP 71152.
b. Observations and Assessments On December 12, 2017 at approximately 0530 the circulating water basin began to accumulate ice causing various intake and basin level alarms. Operators had observed a large amount of ice moving down the Mississippi River via a camera monitoring system. A cooling tower was not in service at the time of the event. At 525 and 0528, Operators adjusted the deicing line valve to allow more warm water to be directed to the circulating water basin entrance. At 0533 the control room received an unexpected annunciator alarm for high differential pressure on the intake bar rack. In response to the intake ice buildup, and per Operations Manual B.06.04-05, Operators placed the screenwash system in full service along with the 11 Cooling Tower Pump aligned to the circulating water system to return warm water to the intake due to the icing problems.
The operator actions resulted in warmer intake water alleviating the ice buildup and restoring circulating water basin level and pressure conditions. No other significant plant impacts were experienced as a result of the sudden influx of river ice to the intake. Plant Technical Specifications require both the Essential Service Water system and ultimate heat sink to be operable in Mode 1. The licensee determined operability requirements were met since two ESW subsystems were operable with one that will operate and the ultimate heat sink was maintained by having a minimum water level in the pump well of the intake structure of 899. The lowest Circulating Water Pump Basin level observed during this intake icing transient was 903.25. The licensee documented this issue in the corrective action program via CAP 501000006364. Inspectors reviewed CAP 501000006364 and determined the licensee established the following corrective actions:
- initiated CAP 501000006364;
- immediate corrective actions consisted of operators fully opening the deicing line and placed a cooling tower placed in service per B.06.04-05 E.13. Winter Operation;
- increased intake monitoring was implemented by Operations; and
- a revision was initiated to Operation Manual B.06.04-05 E.13 Winter Operations:
Circulating Water to enhance exiting guidance regarding environmental conditions leading toward intake icing and preemptive operator actions in response to those conditions.
Inspector review of the licensees condition evaluation determined the licensees actions associated with CAP 501000006364 were satisfactory. Additionally, Inspector review of the corrective action process for an approximate previous five years did not identify any areas of concern regarding previous opportunities for the licensee to enhance its procedures from a pre-emptive environmental condition perspective.
c. Findings
No findings were identified.
4OA3 Follow-Up of Events and Notices of Enforcement Discretion
.1 Loss of Control Room Annunciator Sound During Downpower
On December 15, 2017 the licensee performed a downpower evolution for planned control rod drive and turbine valve testing. During this evolution the control room experienced a loss of annunciator sound which brought added challenge to the operating crew. Inspectors observed activities associated with both the downpower and operator actions relative to the annunciator sound card issue.
The sound card that failed was located in control room panel C-04B and resulted in the alarm clear sound sounding continuously. Immediate actions of operations personnel resulted in implementing a decision making process which led to actions to remove the sound card. Inspector review confirmed removal of the sound card did not impact the ability for the annunciator to illuminate when an alarm condition occurs, confirming the control room maintained capability to receive annunciation of adverse conditions.
Inspectors also confirmed the licensee adequately assessed the loss of annunciation sound. Specifically, procedure A.2-101 was used by operators and requires that when approximately 75 percent of the safety system annunciator function is lost a declaration of an Notice of Unusual Event declaration would occur. Licensed operators determined that criteria had not been met given the loss of annunciator sound condition.
Additionally, licensed operators determined the condition did not impact the ability for the annunciator to alarm and therefore it was not considered a loss of safety system annunciation, or a loss of assessment capability for EAL SU4.1. The licensee documented this issue via corrective action program item CAP 501000006495.
Inspectors review of the licensee immediate actions, CAP, and EAL assessment did not identify any concerns.
a. Inspection Scope
This inspection constituted one event follow-up sample as defined in IP 71153-05.
b. Findings
No findings were identified.
.2 Intake Basin Icing
a. Inspection Scope
On December 12, 2017 Monticello was operating in Mode 1 at full power with external temperatures near 0 degrees Fahrenheit and experienced icing issues in the circulating water basin (intake).
On December 12, 2017 at approximately 0530 the circulating water basin began to accumulate ice causing various intake and basin level alarms. Operators had observed a large amount of ice moving down the Mississippi River via a camera monitoring system. A cooling tower was not in service at the time of the event. At 525 and 0528, Operators adjusted the deicing line valve to allow more warm water to be directed to the circulating water basin entrance. At 0533 the control room received an unexpected annunciator alarm for high differential pressure on the intake bar rack. In response to the intake ice buildup, and per Operations Manual B.06.04-05, Operators placed the screenwash system in full service along with the 11 Cooling Tower Pump aligned to the circulating water system to return warm water to the intake due to the icing problems.
The operator actions resulted in warmer intake water alleviating the ice buildup and restoring circulating water basin level and pressure conditions. No other significant plant impacts were experienced as a result of the sudden influx of river ice to the intake. Plant Technical Specifications require both the Essential Service Water system and ultimate heat sink to be operable in Mode 1. The licensee determined operability requirements were met since two ESW subsystems were operable with one that will operate and the ultimate heat sink was maintained by having a minimum water level in the pump well of the intake structure of 899. The lowest Circulating Water Pump Basin level observed during this intake icing transient was 903.25. The licensee documented this issue in the corrective action program via CAP 501000006364 which established actions to enhance procedures regarding pre-emptive conditions and operator actions.
Inspector review of immediate licensee actions, operability determination, and follow-up action to enhance procedures did not identify any concerns.
This inspection constituted one event follow-up sample as defined in IP 71153-05.
b. Findings
No findings were identified.
.3 (Closed) Licensee Event Report 05000263/2017-002-00, Main Steam Isolation Valve
Closure Time Outside of Technical Specification Requirements
a. Inspection Scope
The inspectors reviewed the event and circumstances detailed in this Licensee Event Report (LER) associated with the failure of the D outboard Main Steam Isolation Valve (MSIV) (AO-2-86D) to meet closure time requirements during testing per the requirements of USAR 05.02, Primary Containment (PCT) System and TS 3.6.1.3.6.
In their review, the inspectors determined that the issues documented in the LER were adequately addressed in the licensees CAP and noted repairs and subsequent closure time testing for the MSIV were completed satisfactorily in a timely and effective manner.
The inspectors noted that D inboard MSIV AO-2-80D (installed in series with the D outboard MSIV) had been tested for both leak rate and closing time over the past operating cycle and each test was completed satisfactorily. As a result, the primary containment isolation capability of the main steam lines remained operable which ensured the required isolation safety function was maintained. Lastly, the inspectors verified based on a review of work orders, the CAP, and surveillance performance history that the failure of the D outboard MSIV to pass closure timing requirements was not within the licensees ability to foresee and correct and therefore no performance deficiency was identified. Documents reviewed as part of this inspection are listed in the to this report. This LER is closed.
This inspection constituted one event follow-up sample as defined in IP 71153-05.
b. Findings
No findings were identified.
.4 (Closed) Licensee Event Report 05000263/2016-002-00; 05000263/2016-002-01;
Inadequate Appendix R Fire Barrier Impacts Safe Shutdown Capability
a. Inspection Scope
On August 4, 2016, while performing a Fire Protection/Appendix R self-assessment, the licensee discovered that the floor between the Cable Spreading Room (CSR) and the Plant Administrative Building (PAB) basement is not a credited Appendix R fire barrier.
Specifically, the licensee discovered that the CSR floor was not an adequate fire barrier, because the structural steel supporting the CSR floor did not have a 3-hour fire rating as required by 10 CFR 50, Appendix R, Section III.G.2.a. Since the CSR and PAB are located in the same fire area without a compliant fire barrier, a fire in the PAB could spread to the CSR and challenge the safe shutdown of the plant. By plant procedure, a fire in the CSR requires evacuation of the control room. Upon evacuation of the control room, plant procedures direct the operators to perform shutdown of the reactor using Division II equipment on the Alternate Shutdown System (ASDS) panel that is located in the Emergency Filtration Train (EFT) Building. However, the credited path used by the operators to access ASDS panel traverses the PAB fire area where a fire event is occurring.
b. Findings
The licensee documented the non-compliant Appendix R condition in their CAP 1530637 and implemented fire watches as a compensatory measure and performed an exposed steel fire simulation and evaluation. Enforcement aspects of this Licensee Event Report are discussed in Section 4OA7. Documents reviewed as part of this inspection are listed in the Attachment. Licensee Event Reports 05000263/2016-002-00 and 05000263/2016-002-01 are closed.
This event follow-up review constituted two samples as defined in IP 71153-05.
4OA6 Management Meetings
.1 Exit Meeting Summary
On October 20, the inspectors presented the Licensed Operator Requalification Program Biennial Program inspection results to Mr. C. Dieckmann, Plant Manager, and other members of the licensee staff.
On January 16, 2018, the inspectors presented the 2017 fourth Quarter Integrated inspection results to Mr. C. Dieckmann, Plant Manager, and other members of the licensee staff.
The licensee acknowledged the issues presented during each exit. The inspectors confirmed that none of the potential report input discussed was considered proprietary.
.2 Interim Exit Meetings
Interim exits were conducted for:
- The Radiation Safety Program inspection results with Mr. C. Church, Site Vice President, on October 5, 2017;
- The 2017 Annual Operating Test results with Mr. R. Becker, Fleet Training and Examination Developer, on November 9, 2017; and
- The 2017 Emergency Preparedness Program inspection results with Mr. B. Carberry, Emergency Preparedness Manager, conducted over the telephone on November 22, 2017.
Each inspector confirmed that none of the potential report input discussed was considered proprietary.
4OA7 Licensee-Identified Violations
The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements, which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation (NCV).
- The licensee identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix R, Section III.G.2.a, Fire Protection of Safe Shutdown Capability. Specifically, the licensee identified that the structural steel located in the plant administrative building (PAB) basement supporting the cable spreading room (CSR) floor did not have a 3-hour fire rating as required by 10 CFR 50, Appendix R, Section III.G.2.a. Title 10 CFR 50, Appendix R, Section III.G.2.a, requires, in part, that where separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating is provided, structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier. Contrary to the above, since 1982, the licensee failed to protect the structural steel supporting the fire barrier between the cable spreading room and fire area IV. This failure was identified by the licensee on August 4, 2016 during an Appendix R self-assessment and addressed in CAP 1530637. The licensee issued LER 2016-002-01 in response to this Appendix R non-compliance and implemented the immediate corrective action (compensatory measure) of an hourly fire watch in the affected fire area. The licensee conducted an exposed steel fire simulation and evaluation to understand the performance of the unprotected steel in the event of a fire in the PAB. The inspectors reviewed the licensees simulation and evaluation.
This finding was determined to be more-than minor because the performance deficiency was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the structural steel supports the fire barrier between the CSR and the PAB basement and a failure to protect the structural steel from fire damage would degrade the fire barrier separating the CSR and PAB. This could result in a fire in the PAB spreading to the CSR due to the degraded fire barrier resulting in an evacuation of the control room. The only means for operators to shutdown the reactor using the ASDS panel would require travel through the PAB fire area where a fire event is occurring. Therefore, this finding impacted the safe shutdown capability of the plant. After review of the licensees exposed steel fire simulation and evaluation, the finding was determined to be of very low safety significance (Green) because the licensee demonstrated that the unprotected structural steel would provide at least one hour of fire endurance rating under a fire event in the PAB.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- C. Church, Site Vice President (Incoming)
- P. Gardner, Site Vice President (Outgoing)
- K. Scott, Site Operations Director
- C. Dieckmann, Plant Manager
- M. Antony, Operations Manager
- M. Lingenfelter, Director of Engineering
- R. Olson, Maintenance Manager
- S. Quiggle, Chemistry Manager
- C. England, Radiation Protection Manager
- A. Ward, Regulatory Affairs Manager
- P. Young, Nuclear Program Engineering Manager
- R. Garding, Program Engineering Supervisor
- T. Jones, Engineering Analyst Principal
- R. Deopere, Engineering Analyst Principal
- R. Loeffler, Regulatory Affairs
- G. Huff, Chemist
- R. Becker, Fleet Training and Examination Developer
- D. Bosnic, Business Support Director
- G. Hernandez, Regulatory Affairs Manager
- B. Carberry, Emergency Preparedness Manager
- P. Beyers, Director of Emergency Preparedness (Corporate)
- S. Sollom, Regulatory Affairs Senior Engineer
- P. Albares, General Supervisor Operations Training
- D. Cox, Operations Training Supervisor
- D. Murphy, Training Instructor
- J. Yarbrough, Simulator Engineer
U.S. Nuclear Regulatory Commission
- K. Riemer, Chief, Reactor Projects Branch 2
- B. Dickson, Chief, Reactor Projects Branch 5
- P. Zurawski, Senior Resident Inspector
- T. Ospino, Resident Inspector (Rotation)
- K. Pusateri, Reactor Engineer
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened
- 05000263/2017004-01 FIN Failure to Maintain Radiation Exposure ALARA
Closed
- 05000263/2017004-01 FIN Failure to Maintain Radiation Exposure ALARA
- 05000263/2016-002-00 LER Inadequate Appendix R Fire Barrier Impacts Safe Shutdown Capability
- 05000263/2016-002-01 LER Inadequate Appendix R Fire Barrier Impacts Safe Shutdown Capability
- 05000263/2017-002-00 LER Main Steam Isolation Valve Closure Time Outside of Technical Specification Requirements