Headquarters Daily Report
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REPORT NEGATIVE
ATTACHED INPUT RECEIVED
HEADQUARTERS X
REGION I X
REGION II X
REGION III X
REGION IV X
PRIORITY ATTENTION REQUIRED MORNING REPORT - HEADQUARTERS NOV. 18, 2002
Licensee/Facility: Notification:
Part 21 Database MR Number: H-02-0090
Pilgrim Date: 11/18/02
Subject: Part 21 - Emergency diesel generator fuel oil pump leaking
through nameplate hole drilled through-wall in the pump body
Discussion:
VENDOR: Fairbanks Morse PT21 FILE NO: m2-25-0
DATE OF DOCUMENT: 09/18/02 ACCESSION NUMBER:
SOURCE DOCUMENT: EN 39197 REVIEWER: RORP, C. Petrone
The Pilgrim licensee reported the fuel injector pump for an ALCo
(Fairbanks Morse), model 251F emergency diesel generator leaked
excessively through a nameplate rivet hole that had been drilled
through-wall in the pump body.
Contact: C. Petrone, NRR
301-415-1027
E-mail: cdp@nrc.gov
HEADQUARTERS MORNING REPORT PAGE 2 NOVEMBER 18, 2002
Licensee/Facility: Notification:
Part 21 Database MR Number: H-02-0091
Fitzpatrick Date: 11/18/02
Subject: Part 21 - Test failure of overcurrent sensors
Discussion:
VENDOR: General Electric PT21 FILE NO: m2-26-0
DATE OF DOCUMENT: 10/02/02 ACCESSION NUMBER: ml022820676
SOURCE DOCUMENT: LETTER REVIEWER: RORP, D. Billings
The Fitzpatrick licensee reported that two new General Electric Model
EC-1 Trip Device (part number QEC10225ABCG10N00, serial numbers 77386-3A
and 77386-3B) overcurrent sensors failed during bench testing. The
magnetic (instantaneous) element of the device tripped prematurely. The
application for these specific overcurrent sensors was in the emergency
bus feeder breaker for a 600VAC Motor Control Center. This motor control
center powers various safety system loads in the Standby Gas Treatment
System, Residual Heat Removal System, and High Pressure Coolant Injection
System. The feeder breaker would have tripped prematurely on the starting
current of a standby gas treatment system exhaust fan and thus the
associated motor control center would not have remained energized. This
would have resulted in the loss of components for one train of the
associated safety systems.
Contact: D. Billings, NRR
301-415-1175
E-mail: deb1@nrc.gov
HEADQUARTERS MORNING REPORT PAGE 3 NOVEMBER 18, 2002
Licensee/Facility: Notification:
Part 21 Database MR Number: H-02-0092
General Electric Date: 11/18/02
Subject: Part 21 - Potentially delayed BWR scram because of stability
Option III period based detection algorithm
Discussion:
VENDOR: General Electric PT21 FILE NO: m2-27-0
DATE OF DOCUMENT: 10/01/02 ACCESSION NUMBER: ml022830278
SOURCE DOCUMENT: LETTER REVIEWER: RORP, V. Hodge
The vendor, General Electric, reported a potentially delayed reactor
scram because of the period based detection algorithm (PBDA). This
algorithm provides the licensing basis minimum critical power ratio
(MCPR) safety limit protection under stability Option III for anticipated
coupled thermal hydraulic-neutronic reactor instabilities. The algorithm
determines successive confirmation count (SCC) of an oscillating power
signal.
A reactor scram is only initiated by the PBDA when the SCC exceeds the
count setpoint and the oscillation amplitude exceeds the amplitude
setpoint. The licensing basis is that the SCC will exceed the count
setpoint before the amplitude reaches the amplitude setpoint. If the SCC
resets, then the amplitude could exceed the amplitude setpoint before SCC
reaches the count setpoint. This could lead to violation of the MCPR
Safety Limit. The algorithm is more susceptible to SCC resets with a
period tolerance that is near to the minimum allowed by licensing
documents (e.g., 50 milliseconds). SCC resets are less likely with higher
period tolerance values (e.g., 100 to 300 milliseconds).
If scram is delayed, boiling transition could be experienced on a portion
of some fuel bundles. This would be a violation of a Technical
Specification Safety Limit and is reportable under 10 CFR 21. However, it
would not produce a significant safety hazard or threat to public health
and safety.
The vendor communicated this concern to the Boiling Water Reactor Owners'
Group Potential Issues Resolution Team (PIRT) and Stability Detect &
Suppress Committee and is continuing to evaluate the potential for the
SCC to be reset for currently licensed reactor operating conditions.
The vendor expects to complete this effort by November 18, 2002. Affected
domestic nuclear power plants include Clinton, Brunswick 1& 2, Nine Mile
Point 2, Fermi 2, Columbia, Dresden 2 & 3, LaSalle 1 & 2, Limerick 1 & 2,
Peach Bottom 2 & 3, Quad Cities 1 & 2, Perry 1, Susquehanna 1 & 2, Hope
Creek, Hatch 1 & 2, and Browns Ferry 1, 2 & 3.
Contact: V. Hodge, NRR
301-415-1861
E-mail: cvh@nrc.gov
HEADQUARTERS MORNING REPORT PAGE 4 NOVEMBER 18, 2002
Licensee/Facility: Notification:
Part 21 Database MR Number: H-02-0093
General Electric Date: 11/18/02
Subject: Part 21 - Potential nonconservative minimum critical power
ratio in new designs of fuel for boiling water reactors
Discussion:
VENDOR: General Electric PT21 FILE NO: m2-28-0
DATE OF DOCUMENT: 10/04/02 ACCESSION NUMBER: ml022820162
SOURCE DOCUMENT: LETTER REVIEWER: RORP, R. Caldwell
The boiling water reactor (BWR) nuclear fuel vendor, General Electric
Nuclear Energy/Global Nuclear Fuel (GE/GNF), reported that the technical
specification safety limit for minimum critical power ratio (CPR) may be
exceeded in BWR/6 plants using newer fuel designs supplied by the vendor.
Recent calculations have shown that the CPR responses for newer fuel
designs are more sensitive than for older fuel designs. The CPR response
depends on the location of the bundle with respect to core support beams.
In a BWR/6 reactor, bundles may be located adjacent to 0, 1, or 2 core
support beams and therefore have different side entry orifice loss
coefficients in the core monitoring system supplied by the vendor. For
the bundles adjacent to 2 core support beams, the loss coefficients are
highest and the CPR is most sensitive to reduced flow in the bundles.
The core monitoring systems in affected plants use average side entry
orifice loss coefficients, one for central bundles and a different one
for peripheral bundles. This was previously evaluated for GE/GNF 8x8 fuel
designs and found to be acceptable. However, the recent calculations
indicate that the core monitoring system overpredicts CPR by about 0.01
for bundles near 2 core support beams and therefore may underpredict the
margin to the operating limit minimum CPR.
The vendor communicated this concern to the Boiling Water Reactor Owners'
Group Potential Issues Resolution Team (PIRT) and Stability Detect &
Suppress Committee and is continuing to evaluate the potential for the
SCC to be reset for currently licensed reactor operating conditions.
The vendor expects to complete this effort by November 18, 2002. Affected
domestic nuclear power plants include Clinton, Brunswick 1& 2, Nine Mile
Point 2, Fermi 2, Columbia, Dresden 2 & 3, LaSalle 1 & 2, Limerick 1 & 2,
Peach Bottom 2 & 3, Quad Cities 1 & 2, Perry 1, Susquehanna 1 & 2, Hope
Creek, Hatch 1 & 2, and Browns Ferry 1, 2 & 3.
Contact: R. Caldwell, NRR
301-415-1243
E-mail: rkc1@nrc.gov
HEADQUARTERS MORNING REPORT PAGE 5 NOVEMBER 18, 2002
Licensee/Facility: Notification:
Part 21 Database MR Number: H-02-0094
C&D Technologies Date: 11/18/02
Subject: Part 21 - Incorrect welds found in recently fabricated
fuel racks
Discussion:
VENDOR: C&D Technologies PT21 FILE NO: M2-29-0
DATE OF DOCUMENT: 10/11/02 ACCESSION NUMBER: ml022900283
SOURCE DOCUMENT: LETTER REVIEWER: RORP, J. Dozier
C&D Technologies, a vendor of fuel racks, reported that a series of welds
joining the horizontal cross beam to the vertical member of the frame
were not welded correctly during fabrication of two single tier fuel
racks for the Prairie Island nuclear facility. The incorrect welds may
affect the seismic qualification of the rack. These defects were
identified before installation, thus the racks were not installed at
Prairie Island. The vendor is undertaking corrective action to ensure
that future racks will not have this defect.
Contact: J. Dozier, NRR
301-415-1014
E-mail: jxd@nrc.gov
HEADQUARTERS MORNING REPORT PAGE 6 NOVEMBER 18, 2002
Licensee/Facility: Notification:
Part 21 Database MR Number: H-02-0095
Engine Systems Date: 11/18/02
Subject: Part 21 - Leakage from water pump housing on EMD emergency
diesel generator
Discussion:
VENDOR: Engine Systems PT21 FILE NO: m2-30-0
DATE OF DOCUMENT: 10/04/02 ACCESSION NUMBER: ml022900613
SOURCE DOCUMENT: LETTER REVIEWER: RORP, J. Foster
Engine Systems, a vendor of EMD emergency diesel generators, reported
that after installation of an engine-driven water pump, a seal within the
pump assembly failed prematurely, resulting in leakage of primarily
cooling water and possibly lubricating oil from the weep hole on the
water pump housing. To extend pump service life, EMD changed the geometry
of the shaft roller bearings from spherical to tapered. A seal was
installed at the opposite-drive end of the shaft to prevent oil leakage
into the cavity and out the weep hole. The vendor stated that the seal
fails because of insufficient preload, allowing chattering of the rolling
parts nd that the problem is not the result of incorrect assembly. The
vendor learned of this problem from locomotive and marine applications
and stated that no pump failures have been reported. Affected domestic
nuclear power plants are Surry, Arkansas Nuclear One, St. Lucie, Turkey
Point, and Kewaunee.
Contact: J. Foster, NRR
301-415-3647
E-mail: jwf@nrc.gov
