ENS 43499
ENS Event | |
|---|---|
12:00 Jul 17, 2007 | |
| Title | Unanalyzed Condition - Vent Space Less than Design Basis |
| Event Description | During a review of the temporary repair of the steam line drain bypass line in the Unit 1 Reactor Building Steam Chase, two storage gangboxes were noted to be on the grated opening in the floor of the Steam Chase (elevation 129 ft). These grated openings are designed to be open to provide pressure and temperature relief between the steam chase and the torus room for high energy steam line breaks.
Appendix N to the Unit 1 FSAR credits the openings for venting the steam chase to the torus room through the openings for a main steam line break, and for venting the torus room to the steam chase for a HPCI steam line break in the torus room. The most limiting event is the HPCI steam line break in the torus room and the vent path associated with that event. Original assumptions used in the calculation for the vent opening did not adequately account for the grating itself and for louvers installed in a previous plant modification. As a result the vent area was further reduced. Upon further review of the above condition, it has been determined that a non-conforming and unanalyzed condition exists In that the vent area between the torus room and main steam chase in the reactor building is less than the area assumed in the analysis, even without gangboxes covering a portion of the grating. As such, for a HPCI steam line break in the torus room, the short term pressure between the torus room and the corner rooms (diagonals) is greater than 2 psid, which is the stated limit in Appendix N of the Unit 1 FSAR. The corner rooms contain ECCS components in the RHR and core spray systems. Based on engineering judgment there is reasonable assurance that the present nonconforming condition does not prevent safety systems and structures from fulfilling their safety function. This is based on the following information: Structural Steel floor elevation platforms do not appear to have been credited in the structural design capability of the walls. These platforms should act to help maintain the wall intact with increased pressure. The increased pressure transient is a very short term transient, approximately 2-3 seconds in duration, after which the pressure will return to within 2 psid. It is expected that the wall would withstand this transient without degrading the performance of the low pressure ECCS systems or other structures and components. Lastly, the probability of occurrence of a steam leak leading to an instantaneous line break is very small. There is currently no report of steam leaks from the HPCI line, and although a probability evaluation has not been performed, it is likely that the probability of occurrence of such a break is very small. Thus, there is no known immediate threat that would prevent safety systems from performing their safety function. More detailed review is continuing at this point. Short term corrective action will be required to increase the open 'vent' area between the torus room and the reactor building 130 ft elevation and restore at least the assumed vent path from the torus room. This can be accomplished by removing the gangboxes over the vent area in the steam chase and/or completing a floor plug evaluation of vent area needed between the torus room and the reactor building 130 ft elevation which will restore compliance with the 2 psid criteria. Analysis is currently underway to assess the pressure and temperature effects on the safety related structures and equipment by these short term actions. Regarding reportability, based on engineering judgment as previously discussed, the unanalyzed condition does not represent a condition that significantly degraded plant safety; however, additional information is needed in order to more conclusively determine this. For this reason this condition is being conservatively reported under 10CFR50.72(b)(3)(ii)(B) until such time as more conclusive information is provided to make the final determination. The licensee notified the NRC Resident Inspector.
Upon further review of the above 'as found' conditions, it has been determined that there are existing conservatisms in the current analysis which bound the flow restriction caused by the gang boxes on the grating. The evaluation concluded that the gang boxes found on the grated opening in the floor of the steam chase would not increase the pressures in the Unit 1 reactor building as a result of HELB conditions. Thus the pressure between the torus room and the corner rooms (diagonals) which is limited to 2 psid as stated limit in Appendix N of the Unit 1 FSAR is not affected. In addition, an additional open floor plug (the 3 ft by 3 ft floor plugs between Elevation 130 and the torus room below found to be covered by a hinged metal plate) is acceptable since it causes less differential pressure across reactor building compartments during the HELB's evaluated. The results of this additional review confirmed the original engineering judgment that there was reasonable assurance that the as found nonconforming condition did not prevent safety systems and structures from fulfilling their safety function. Short term corrective actions were completed upon discovery of he 'as found' condition to further increase the open 'vent' area between the torus room and the reactor building 130 ft elevation and restore at least the assumed vent path from the torus room. This was accomplished by removing the gang boxes over the vent area in the steam chase. Based on this review of the design calculations white taking the 'as found' conditions into consideration, the conclusion reached is that the nonconforming 'as found' conditions did not represent a condition that significantly degraded plant safety. For this reason this condition that was initially reported under 10CFR50.72(b)(3)(ii)(B) is being retracted. The licensee will notify the NRC Resident Inspector. R2DO (Shaeffer) notified. |
| Where | |
|---|---|
Hatch  Georgia (NRC Region 2) | |
| Reporting | |
| 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | |
| Time - Person (Reporting Time:+3.35 h0.14 days <br />0.0199 weeks <br />0.00459 months <br />) | |
| Opened: | Steve Brunson 15:21 Jul 17, 2007 |
| NRC Officer: | Jason Kozal |
| Last Updated: | Aug 30, 2007 |
| 43499 - NRC Website
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Unit 1 | |
|---|---|
| Reactor critical | Critical |
| Scram | No |
| Before | Power Operation (100 %) |
| After | Power Operation (100 %) |
WEEKMONTHYEARENS 526502017-03-30T13:22:00030 March 2017 13:22:00
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