Event Notification Report for August 13, 2007

EMERGENCY OPERATIONS FACILITY VENTILATION INOPERABLE

"This event is being reported under 10CFR50.72(b)(3)(xiii) for a major loss of offsite emergency response capability.

"At 2325 Central Daylight Time (CDT) on 8/12/2007, Callaway Plant received a trouble alarm for the Emergency Operations Facility (EOF). Investigation revealed the EOF had lost one phase of incoming rural electric power which is believed to have been caused by a severe thunderstorm in the area. At 0001 CDT on 8/13/2007, the EOF diesel generator was started and aligned to provide power to all building loads. An EOF functionality evaluation was performed for a loss on normal power with the EOF being supplied from the diesel generator and pre-existing degraded air conditioning cooling capacity. The evaluation concluded that the EOF was fully functional.

"At approximately 1000 CDT on 8/13/2007, maintenance personnel reported to the Control Room that the emergency ventilation system in the EOF was not functional based on the pressurization fans rotating in reverse. Three phase electrical loads at the EOF are incorrectly polarized apparently due to improper terminations at the diesel generator. If an emergency condition occurs during the time repairs are being made, the EOF will be utilized as long as radiological conditions allow. Procedure EIP-ZZ-C0010, EMERGENCY OPERATIONS FACILITY OPERATIONS, section 4.4, directs EOF management to evaluate the need to relocate operations to the Backup Emergency Operations Facility, as required. Affected emergency responders have been notified of facility conditions and a courtesy notification has been made to the State Operations Branch Chief."

The licensee has notified the NRC Resident Inspector.

* * * UPDATE PROVIDED BY KEITH DUNCAN TO JEFF ROTTON AT 1248 ON 08/14/07 * * *

"The EOF (Emergency Offsite Facility) rural electric power was restored to the building at 1730 (CDT) on 8/13/2007. At 0059 (CDT) 8/14/2007 the EOF diesel work was completed, restored and retested satisfactorily. Building temperatures were verified to be at 85 deg F and lowering, this is within the 55-95 deg F range for functionality. The EOF is considered restored to a fully Functional Status.

The licensee notified the NRC Resident Inspector. Notified R4DO (Gaddy)

INADEQUATE FIRE PROTECTION ON SAFETY CHILLED WATER SYSTEM ELECTRICAL CABLES

"At 0900 on July 30 2007, an Engineer noted during the review of a revision to the Comanche Peak Fire Safe Shutdown Analysis that a cable associated with the control circuitry for Train B of the Safety Chilled Water System may not be adequately protected from a potential fire. By design, electrical control cables for Trains A and B of the Safety Chilled Water System are located in the same fire zone. The original design specified that the Train B electrical control cables in this zone were to be protected with fire barrier material (thermolag). However, in this case the fire barrier material was found to be missing from the Train B electrical control cables. Upon discovery of this condition, a fire impairment was implemented for the affected fire zone.

"Engineering performed an evaluation of this condition and at 0900 on August 13, 2007 concluded that if a fire occurred in the affected fire zone, the required degree of separation for redundant safe shutdown trains was inadequate (i.e. both A and B trains were affected) and this would adversely affect the control circuitry and potentially prevent the Unit 1 Safety Chilled Water System from performing its intended safety function. The Unit 1 Safety Chilled Water Systems safety function at Comanche Peak is to remove heat dissipated from engineering safety features equipment and to maintain ambient temperatures in rooms containing safety related equipment below maximum design temperatures.

"This condition is similar to an example given in NUREG 1022, Rev. 2, Section 3.2.4 for an unanalyzed condition that significantly affects plant safety (fire barrier missing such that the required degree of separation for redundant safe shutdown trains is lacking). Therefore, this condition is reportable per 10CFR50.72(b)(3)(ii)(B), 'The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.'"

The licensee notified the NRC Resident Inspector.


* * * RETRACTION ON 08/30/07 AT 1249 EDT FROM RAUL MATINEZ TO MACKINNON * * *

"CPNPP is retracting Event Notification 43562 based on the following:

"Further review of this issue by Engineering has determined that the required degree of separation for redundant safe shutdown trains was adequate and the Unit 1 Safety Chilled Water System was capable of performing its intended safety function. Therefore, this condition is not reportable per 10CFR50.72(b)(3)(ii)(B), 'The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.'

"CPNPP has informed the NRC Resident Inspector."

R4DO (R. Nease) notified.

PART 21 REPORT - AREVA MINIMUM CRITICAL POWER RATIO

The licensee provided the following information via facsimile:

"The defect is in the calculation of steady-state core Minimum Critical Power Ratio (MCPR) by the core monitoring system when the SPCB critical power correlation is used for ATRIUM-10 fuel. Specifically, the defect is in the additive constants, a parameter used by the SPCB critical power calculation and based on test data." AREVA notified the affected plants.

Affected Plants:
Browns Ferry, Units 2 & 3
Columbia
Grand Gulf
LaSalle, Units 1 & 2
River Bend
Susquehanna, Units 1 & 2

INOPERABLE RADIATION MONITORS DUE TO SETPOINT CALCULATION ERRORS

"On August 13, 2007 at 14:20 [EDT], the Susquehanna Control Room was notified that during review of calculations for the Secondary Containment Radiation Monitor Setpoints errors were identified in Tech Spec Allowable Values and TRM Trip Setpoints that rendered them non-conservative. The affected instruments are Refuel Floor Wall, Refuel Floor High, and Railroad Access radiation monitors. The condition affects both Susquehanna Units. The radiation monitors are required to be operable for conditions noted in footnotes (a), (b), and (c) in Tech Spec tables 3.3.6.2-1 and 3.3.7.1-1 (i.e. operations with a potential for draining the reactor vessel, during CORE ALTERATIONS and during movement of irradiated fuel assemblies in the secondary containment, and movement of irradiated fuel assemblies within or above the Railroad Access Shaft). The function of these instruments is to initiate systems that limit fission product release during and following certain postulated accidents and to minimize the consequences of radioactive material in the control room environment.

"The radiation monitoring instruments were declared inoperable. All movement of irradiate fuel assemblies was halted (dry fuel storage activities were in progress at the time of notification). Based on this action, these monitors are no longer within the specified Applicability, and are therefore not currently required to be operable. The event has been determined to be reportable within 8 hours under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D)."


The licensee notified the NRC Resident Inspector.


* * * UPDATE FROM G. ROBINSON TO J. KNOKE AT 1222 EDT ON 10/03/07 * * *

"On August 13, 2007 PPL Susquehanna reported that during a review of calculations for the Secondary Containment Radiation Monitor Setpoints, errors were identified in Tech Spec Allowable Values and TRM Trip Setpoints that rendered them non-conservative. The affected monitors were declared inoperable and the required LCO & TRO Actions entered.

A subsequent engineering review determined that the values used to establish the Tech Spec Allowable Values and TRM Trip Setpoints were based on a Realistic Source Term, not a Design Basis Source Term. These values were sufficiently conservative to compensate for the errors in the calculations and the dose release limits from a Design Basis Accident would not have been challenged. It was concluded that the existing Tech Spec values and setpoints remained valid and the monitors were operable in their original condition. Based on this conclusion Event Notification EN# 43563 is being retracted."

The licensee notified the NRC Resident Inspector. R1DO ( Powell) was notified.

POTENTIAL HOT SHORT ISSUE WITH CONTAINMENT SPRAY SUMP SUCTION VALVES

" A condition is being reported as a conservative measure regarding a situation involving potential fire induced circuit failures on associated circuits. Plant Vogtle is considered to be in compliance with its current fire protection licensing basis, Branch Technical Position CMEB 9.5.1 as endorsed by UFSAR Appendix 9B, however, this condition is being reported as the generic industry issues associated with multiple spurious circuit failures have not been resolved to date.

"A potential hot short issue exists involving Containment Spray Sump Suction valves 1(2)HV-9002A, 1(2)HV- 9003A, 1(2)HV-9002B, and 1(2)HV-9003B. If a fire were to occur in a location where the control circuit cables for those valves run in the same electrical raceway, a credible condition exists where either flowpath could spuriously open which would cause the Refueling Water Storage Tank (RWST) to drain down to the containment sumps.

"The fire zones where associated cables are located were placed under a compensatory hourly fire watch until compensatory measures were established. As an interim corrective action, operator manual actions have been specified and incorporated into plant procedures to deenergize and verify closed one valve in the series flowpath upon determination that a credible fire exists. These manual actions are in accordance with RIS 2006-10 for plants which are licensed to operate after January 1, 1979. Southern Nuclear will determine long term corrective actions for this condition.

"The effects of a fire sufficient to cause a spurious operation of multiple valves is not expected to be immediate and thus will allow sufficient time to perform the manual actions."

The licensee notified the NRC Resident Inspector.

* * * RETRACTION PROVIDED BY G. SAXON TO J. KOZAL ON 8/15/07 AT 1522 * * *

"An eight hour report (EN#43564) per 10CFR50.72(b)(3)(ii)(B) was conservatively reported based on a fire-induced hot short issue (multiple-spurious) which could result in the spurious opening of Containment Spray Sump Suction valves 1(2)HV-9002A, 1(2)HV-9003A, 1(2)HV-9002B, and 1(2)HV-9003B. HV-9002 and HV-9003 are two valves in series between the containment emergency sump and the containment spray pumps. If a fire were to occur in a location where the control circuit cables for those valves run in the same electrical raceway, it is postulated a condition may exist where either flowpath could spuriously open which would cause the Refueling Water Storage Tank (RWST) to drain down to the containment sumps.

"The VEGP design basis defines a spurious control signal action/inaction as being caused by a single hot short, open circuit, or short-to-ground for a single component. Since all three conditions are postulated, the assumption that all adverse conditions will happen to the subject component(s) yields a conservative evaluation. Where a hi/low pressure interlace (fire-induced LOCA) is involved, an unlimited number of spurious control signal actions/inactions are postulated. Calculations have been reviewed to ensure these design criteria have been met.

"At this time, there is no definitive new guidance (Generic Letter, etc.) by the NRC on how licensees are to handle analysis of multiple spurious shorts. The status and history of misunderstanding and confusion relative to the regulatory and design requirements is well documented in NRC EGM 98-002 Rev. 2 and EGM 07- 004. The NRC staff and the industry are currently working to resolve questions raised by the industry about the adequacy of the existing staff guidance concerning fire-induced circuit failures and the consistency of staff interpretations of both the guidance and the underlying regulatory requirements.

"In summary, VEGP is in conformance with the current design basis. However, as defense-in-depth, actions have been taken that include implementation of hourly fire watches in the affected zones until procedural guidance for performance of manual actions were added to the Annunciator Response Procedures for the Fire Alarm Computer. The manual actions consist of opening the breakers for one of the emergency sump-to containment spray pumps suction isolation valves and then verifying the valves are closed.

"Based on the above discussion, SNC is retracting EN#43564."

The licensee will notify the NRC Resident Inspector. Notified R2DO (Ogle).

AGREEMENT STATE REPORT - LOST GAS CHROMATOGRAPH

The licensee provided the following information via facsimile:

"A portable gas chromatograph with a 15 mCi source of Ni-63 was lost during Hurricane Katrina when the levees broke. This is a General License device. LDEQ was not notified of this loss until August 13, 2007."

LA Event Report - LA070025

THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL

Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks.