ML20196B595
| ML20196B595 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/09/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20196A806 | List: |
| References | |
| 50-361-99-04, 50-361-99-4, 50-362-99-04, 50-362-99-4, NUDOCS 9906230234 | |
| Download: ML20196B595 (5) | |
See also: IR 05000361/1999004
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Potential backflow of watar from the recirculation sumps into the RWST tanks was
l prevented by the outlet check valves in the piping from these tanks. Section 6.3.2.5.5 of
the UFSAR stated that failures of check valves in the safety injection system are not
considered credible failures The check valves in question are part of the safety injection
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' system. Therefore, these check valves alone were sufficient to preclude any significant
' loss of recirculation water inventory through the piping into the RWSTs.
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!The inspe,ctors concluded that the ECCS would have been capable of performing its
safety function in the recirculation mode.
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Radioloaical Containment Safetv Function
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- The NRR staff evaluated the radiological consequences of having the RWST isolation
valves open during a design basis event.~ The licensee documented its assessment of the
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. issue in Calculation N04060-024, " Radiological Consequences of Valve Leakage
Following a Loss of Coolant Accident."' A review of the calculation revealed that no credit
for the RWST isolation valves was assumed; only the outlet check valves were credited
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for preventing backflow through the RWST piping. As stated above,- the facility was
licensed based on the assumption that failures of the check valves in the safety injection
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system are not considered credible. The 'value for backleakage through the outlet check
valves assumed by the licensee was consistent with NRC staff guidelines contained in
' Standard Review Plan Section.15.6.5, Appendix B. The licensee's radiological analysis
maximized the RWST water volume, since minimizing the RWST air volume would
minimize dilution of radioactive material prior to release to the environment. The staff
concurred with the licensee's determination that minimizing the R'#ST air volume
maximizes the radiological consequences.
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In addition, the staff reviewed the impact on the radiological consequences of draining the
RWSTs and concluded that, as long as the outlet check valves remained submerged, the
leak rate assumptions used in the licensee's analysis would remain valid. The elevation
' of the subject check valves was compared to the minimum water level in the RWST
l piping, as calculated by the Reactor Systems Branch. This comparison showed that there
would be at least 20 feet of water over the check valves at all times. Therefore, the staff
concluded that having the RWST isolation valves open does not affect the results of the
~ licensee's radiological analysis and concurred with the licensee's determination that
- radiation levels were acceptable.
~ However, the staff also noted that the inability of the RWST isolation valves to close upon
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! a recirculation actuation signal causes operation to be degraded from its original design
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. purpose because the water level in the RWST could be pumped down to the RWST exit
line. The reduced water level did not agree with the assumption of a 18.5% +3.8% RWST
water level used,in the radiological calculation. In addition, if the ECCS pumps are not
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. running and there is no isolation between the sump and the RWST, the large containment
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accident pressure could push water (and containment gases) from the sump into the tank.
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The staff noted that this illustrates the importance of the two isolation valves to perform an
important redundant containment boundary function.
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in summary, since the license basis documented in the UFSAR is based on the
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assumption that check valva failure is not credible, and the check valves were found to be
~ functional at the time of the evnt, the ECCS function remained within the design basis for
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preventing water flow back to the RWST and the estimated radiological consequences
remained acceptable. However, with the RWST isolation valves left open, the redundant
containment boundary and the ability to prevent any leakage past the check valves and
back to the RWST was degraded,
e.
Reportability
in an August 13,1992, memorandum to file, the licensee documented an active failure
exemption justification for the RWST outlet check valves. The operability assessments
dowmented by the licensee in ARs 950300186 and 950500087, addressing the failure of
the Unit 3 RWST outlet isolation valves, extrapolated this exemption justification to mean
that the outlet check valves fully satisfied the design basis function of isolating the RWST
after a recirculation actuation and that the outlet isolation valves had no design function
and were not part of the desigr: basis. The licensee determined in 1995 that the valve
failures were not reportable.
Based on the recent NRC staff's determination that the ability of the RWST outlet isolation
valves to close is part of the design basis and that the RWST outlet isolation valves
perform an important redundant containment boundary function when the ECCS pumps
are not running during the course of an accident, the inspectors determined that the
licensee may have been required to have reported the event in accordance with
This issue is in the licensee's corrective action program as AR 990400496.
Based on the
additional information from the NRC staff, the licensee planned to reevaluate their original
reportability determination. This item is unresolved to give the licensee an opportunity to
provide their perspective on the NRC staff's determination prior to NRC making a final
determination regarding whether a violation occurred (URI 362/99004-04)
E8.2 (Closed) Unresolved item 361: 362/99001-04: review of reportability assessment
regarding control room emergency air cleanup system operability.
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The inspectors reviewed the licensee's reportability assessment regarding the licensee's
determination that a cable for the Train A control room emergency air cleanup system had
insufficient ampacity while a Cerablanket fire barrier was installed over a section of the
cable raceway. Subsequent review by Nuclear Engineering Design determined that
environmental qualification testing had demonstrated that the aging factors for the cable in
question were not as severe as had been assumed in the cable ampacity calculations.
Additionally, the ambient temperature of the room in which the cable was installed was,
historically, significantly lower that had been assurred in the calculations. The licensee's
subsequent evaluation concluded that the original
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Cerabianket installed, was adequate under all desie onditions. The evaluation also
considered the future aging of the cable to ensure that the cable would remain operable
for the remainder of the facility license. The licensee's initial action to declare the system
inoperable and to take corrective measures, although appropriate at the time, proved to
be conservative. No noncompliance with NRC requirements was identified.
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E8.3 L (Closed) LER 361: 362/1998-008-00: 4.16 kV supply cable exceeds ampacity rating.
' On March 6,1998, the licensee determined that the feeder cables from a unit auxiliary
transformer to the Class 1 E 4.16 kV buses for both Units 2 and 3 could exceed their
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' maximum allowable conductor temperature when a unit is in the backfeed alignment, is
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. supplying power to its shutdown loads, and is also providing power to the maximum
postaccident loads in the other unit via the 4.16 kV bus crosstie. Subsequent calculations
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revealed that the Unit 2 ampacity was acceptable. However, Unit 3 would require a-
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design change to correct the ampacity deficiency.
In the LER, the licensee stated that the cables would have been able to supply the
- maximum calculated amperage, in spite of exceeding the allowable temperature limits. In
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addition, all connected loads would have been able to perform their intended functions.
' Based on the licensee's analysis, the inspectors concluded that there was negligible
safety consequence associated with this condition and no actual safety consequence.
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To return the Unit 3 feeder cables to an acceptable ampacity rating, the licensee
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implemented Field Change Notice F14774E. The field change notice removed sections of
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the top cable tray cover, providing greater heat dissipation of the feeder cables, and
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installed appropriato firr barriers on the Class 1E raceways that did not meet the required
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separation distances. The licensee completed the field change notice on March 15,1999.
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The ' licensee determined that, under worst case loading conditions while backfeeding, the
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. feeder cable temperatures could exceed the allowed 130*C and might not return to 90*C,
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or less, within the allowed 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. This condition did not comply with the design
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information of UFSAR Section 8.3.1.1.3.10 CFR Part 50, Appendix B, Criterion Ill,
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requires, in part, that measures shall be established to assure that the design basis is
correctly translated into specifications, drawings, procedures, and instructions. The failure
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. of the licensee to assure the design basis was correctly translated into specifications was
a violation of 10 CFR Part 50, Appendix B, Criterion lil. This Severity Level IV violation is
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being treated as a noncited violation, consistent with Appendix C of the NRC Enforcement
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Policy (NCV 362/99004-05). This violation was in the licensee's corrective action program
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as AR 980300480.
IV. Plant Support
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Conduct of Emergency Preparedness Activities
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P1.1
Alert Declared because of a Suspicious Lookina Pioe/ Potential Bomb - Units 2 and 3
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2 a.
Ir16Dection Scope (71750. 93702)
. The inspectors monitored the licensee's performance during a declaration of an Alert.
The inspectors reviewed Procedure SO123-Vill-10. " Emergency Coordinator Duties,"
Revision 9, and Procedure SO123-Vill 1, " Recognition and Classification of
Emergencies," Revision 11. The inspectors reviewed ARs 990300467,496,503,506,
509,597, and 991.
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Observations and Findinas
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On March 15,1999, at 10:15 a.m., the licensee declared an Alert because of a potential
bomb that had been discovered in the protected area. The suspicious looking device was
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an approximately 12-inch long,2-inch diameter copper pipe that was capped on both ends
and was discovered behind a large storage container on the turbine deck 70 feet elevation
by a contract employee.
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The inspectors responded to the control room and then to the technical support center.
The licensee activated the technical support, operations support, and emergency
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operations centers and performed an evacuation of local plant areas surrounding the
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device. Security contacted the United States Marine Corps Explosive Ordnance Disposal
Team for assistance. The Explosive Ordnance Disposal Team x-rayed the device and
ultimately determined that it was not a bomb. The event was terminated at 12:27 p.m.
The licensee initiated several ARs, as a result of the event, to capture lessons-learned
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and areas for improvement. The inspector reviewed the ARs and concluded that the
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licensee's actions were self-critical.
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Conclusions
The licensee's declaration of an Alert and response to a potential explosive device (pipe
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' bomb) were conservative. Licensee performance in the technical support center was
good and included appropriate personnel, communications, and briefings. The licensee's
assessment of the event was self-critical.
V. Manaaement Meetinas
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Exit Meeting Summary
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The inspectors presented the inspection results to members of licensee management at
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the exit meeting on April 7,1999. The licensee acknowledged the findings presented.
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The inspectors asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified
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San Onofre Nuclear Generating Station -
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E-Mail report to T. Frye (TJF) ..
E-Mail report to D. Lange (DJL)
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E-Mail. report to NRR Event Tracking System (IPAS)
E-Mail report to Document Control Desk (DOCDESK)
E-Mail report to Richard Correia (RPC)
. E-Mail report to Frank Talbot (FXT)
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. bec to DCD (IE01) ~
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bec distrib. by RIV:
Regional Administrator
Resident inspector
DRP Director '
DRS-PSB
DRS Director
MIS System
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Branch Chief (DRP/E)
RIV File
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Senior Project inspector (DRP/E)
Branch Chief (DRP/TSS) .
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DOCUMENT NAME: R:LSO23\\SO904RP.COR '
To recohm copy of document, Indicate in box: "C" = Copy without enclosures *E' = Copy with enclosures "N" = No copy
RIV:SPE:DRP/F#
C:DRP/EC/n
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OFFICIAL RECORD COPY
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