05000282/FIN-2013005-04
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Finding | |
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| Title | Licensee-Identified Violation |
| Description | Technical Specification 3.7.10 requires that two trains of the control room special ventilation system (CRSVS) must be operable when the reactor is operating in Modes 1 through 4 or during the movement of irradiated fuel assemblies. Contrary to the above, the licensee identified on August 9, 2013, that one or more CRSVS trains had been inoperable since December 10, 2010, due to the failure to properly perform control room envelope unfiltered air in-leakage testing as required by TS Surveillance Requirement 3.7.10.5. The inspectors determined that this issue was more than minor because it was associated with the Barrier Integrity cornerstones attribute of barrier performance and affected the cornerstones objective of maintaining the barrier functional integrity of the Control Room. Specifically, the licensee failed to ensure that the measured control room envelope unfiltered air in-leakage remained less than or equal to the in-leakage rate assumed in the accident analysis. As a result, the licensee was not able to demonstrate that operations personnel located within the control room would have been adequately protected from the radiological consequences of a design basis accident. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Barrier Integrity cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3 for the Barrier Integrity cornerstone. For the control room envelope (CRE) finding, Section C of Exhibit 3 was applicable. The inspectors answered no to the first screening question and yes to the second screening question. As a result, a detailed risk evaluation was required to be performed by a Senior Reactor Analyst (SRA). The exposure time of the finding was assumed to be the maximum of one year. To evaluate the effects of the degradation of the CRE against a toxic atmosphere, the SRA determined that there was no delta risk from toxic gas events. The outside air damper which was identified to be leakage during subsequent testing did not auto-close and was not intended to preclude toxic gas from entering the Control Room from a design basis perspective. To evaluate the effects of the degradation of the CRE against smoke, the SRA identified that the Electric Power Research Institute (EPRI) Fire Probabilistic Risk Assessment Implementation Guide (report number EPRI TR-105928), Appendix B, stated that while actions performed from the control room should generally be unaffected by a fire outside the control room, a January 1989 event at an eastern nuclear plant indicated that fires outside the control room may increase the stress on the operators if smoke reaches the control room. In addition, NUREG/CR-6883, The SPAR-H Human Reliability Method, indicated that physical stress (such as that imposed by difficult environmental factors) may increase the stress and impede the operator from easily completing a task. Based on this, the SRA conservatively assumed that with the degradation of the CRE, any fire in the plant would result in smoke entering the control room and represent a risk increase due to the performance deficiency with the following exceptions: A fire initiated in the control room (since a fire in the control room would not represent additional risk given the performance deficiency); and A fire in containment (since smoke from a fire in containment would not reach the control room). This evaluation was considered bounding because of the tortuous path that would be required for smoke from a fire in the plant to enter the CRE through the leaking outside air damper. The delta risk due for the finding was attributed to high stress on the control room operators (e.g., the control room operators may need to wear self-contained breathing apparatus (SCBAs)). Using SPAR-H, a Performance Shaping Factor (PSF) with a multiplier of 2 was used due to High Stress while performing the affected control room actions. From the Prairie Island IPEEE, Appendix B, Revision 2, Internal Fires Analysis, the core damage frequency (CDF) for fires was spread across five accident classes: (TEH) - early core melt with the reactor at high pressure; (TLH) - late core melt with the reactor at high pressure; (SEH) - early core melt with the reactor at high pressure in conjunction with a small loss-of-coolant-accident (SLOCA); (SLH) - late core melt with the reactor at high pressure in conjunction with a SLOCA; and (BEH) - early core melt with the reactor at high pressure in conjunction with a station blackout. In the Prairie Island IPEEE, Appendix B, Attachment 8, Fire PRA Dominant Cutsets, the dominant cutsets for each of the five accident classes is given. The SRA's evaluation of the ?CDF associated with the finding for each of the five accident classes is provided below: Accident class TEH - of the top 100 cutsets for this accident class, cutsets 12, 58, 80, and 85 were found to contribute to a ?CDF associated with the finding. The ?CDF associated with this accident class was evaluated to be 8.8E-8/yr due to high stress. Accident class SEH - of the top 100 cutsets for this accident class, cutset 93 was found to contribute to a ?CDF associated with the finding. The ?CDF associated with this accident class was evaluated to be 7.1E-9/yr due to high stress. Accident class TLH - of the top 100 cutsets for this accident class, cutsets 17 and 70 were found to contribute to a ?CDF associated with the finding. The ?CDF associated with this accident class was evaluated to be 5.9E-9/yr due to high stress. Accident class BEH - involves fires that cause a loss-of-offsite-power (LOOP). In the IPEEE, Appendix B, Revision 2, it states that only one fire was determined to lead to a LOOP and involved a large fire in the control room G control panel. Since a control room fire does not represent a ?CDF associated with the finding, accident class BEH was eliminated from further consideration. Accident class SLH - of the top 100 cutsets for this accident class, 98 were associated with control room fires. The other two fires (cutsets 85 and 86 on the list of 100) involved relay room fires that did not contribute to the ?CDF associated with the finding because these cutsets did not contain basic events involving operator actions. The total ?CDF associated with the effects of the degradation of the CRE against smoke is the sum of the ?CDF for each of the five accident classes or 1.0E-7/yr. Taking into account the exposure time of the finding, the ?CDF associated with the effects of the degradation of the CRE against smoke was 1.0E-7/yr. Since the total estimated change in core damage frequency was greater than or equal to 1.0E-7/yr, the potential risk contribution from large early release frequency (LERF) was evaluated for risk significance. Appendix H, to IMC 0609, Containment Integrity Significance Determination Process was used to determine the potential risk contribution due to LERF. Prairie Island is a 2-loop Westinghouse Pressurized Water Reactor (PWR) with a large dry containment. Sequences important to LERF include steam generator tube rupture events and inter-system loss-of-coolant-accident events. These were not the dominant core damage sequences for this finding. Based on the Detailed Risk Evaluation, the Senior Reactor Analysts determined that the finding was of very low safety significance (Green). |
| Site: | Prairie Island  |
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| Report | IR 05000282/2013005 Section 4OA7 |
| Date counted | Dec 31, 2013 (2013Q4) |
| Type: | NCV: Green |
| cornerstone | Barrier Integrity |
| Identified by: | Licensee-identified |
| Inspection Procedure: | |
| Inspectors (proximate) | P Laflamme P Voss A Shaikh C Tilton D Oliver J Bozga J Laughlin K Riemer K Stoedter M Jones M Phalen |
| Violation of: | Technical Specification |
| INPO aspect | |
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Finding - Prairie Island - IR 05000282/2013005 | ||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||
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Finding List (Prairie Island) @ 2013Q4
Self-Identified List (Prairie Island)
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