Morning Report for July 20, 2001
Headquarters Daily Report
JULY 20, 2001
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REPORT NEGATIVE NO INPUT
ATTACHED INPUT RECEIVED RECEIVED
HEADQUARTERS X
REGION I X
REGION II X
REGION III X
REGION IV X
PRIORITY ATTENTION REQUIRED MORNING REPORT - HEADQUARTERS JULY 20, 2001
MR Number: H-01-0048
NRR DAILY REPORT ITEM
GENERIC COMMUNICATIONS
Subject: Issuance of Regulatory Issue Summary 2001-14
NRC Regulatory Issue Summary 2001-14: Position on Reportability
Requirements for Reactor Core Isolation Cooling System Failure, dated
July 19, 2001
The U.S. Nuclear Regulatory Commission is issuing this regulatory issue
summary (RIS) to notify BWR addressees of its position regarding the
reportability of reactor core isolation cooling (RCIC) system failure.
Technical contacts: Dennis Allison, NRR
301-415-1178
E-mail: dpa@nrc.gov
Accession Number: ML011940145
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HEADQUARTERS MORNING REPORT PAGE 2 JULY 20, 2001
Licensee/Facility: Notification:
Part 21 Database MR Number: H-01-0049
Engine Systems Date: 07/20/01
Subject: Part 21 - Failure of electrolytic capacitors in Woodward 2301A
control devices
Discussion:
VENDOR: Engine Systems PT21 FILE NO: m1-21-0
DATE OF DOCUMENT: 06/19/01 ACCESSION NUMBER: ML011770294
SOURCE DOCUMENT: LETTER REVIEWER: REXB, N. Fields
The vendor, Engine Systems, reports failures of electronic controls with
electrolytic capacitors used in diesel generator and turbine control
systems. The devices were manufactured by Woodward. In August 2000, at
the Turkey Point Nuclear Power Station, a 2301A control device capacitor
failed. Its designation is C17, part number 1660-111, with manufacturing
date code 8804. In December 1995 and June 1994, at the Columbia
Generating Station, similar failures occurred with 2301A capacitors, both
designated C17, part number 1660-111, with manufacturing code 8634.
The Turkey Point licensee had a failure analysis performed on the failed
capacitor by Seal Laboratories, who concluded that the most likely cause
of failure to be electrolyte contamination that may have originated with
solvent cleaning of the circuit board during manufacturing. The Columbia
Generating Station licensee had a failure analysis performed on one of
its failed capacitors by HI-REL Laboratories, resulting in a similar
conclusion. The vendor learned in its investigation that all three
devices were installed beyond the manufacturer's recommended replacement
interval of 5-7 years. The vendor believes that the capacitors simply
reached their end of life while not discounting the possible contribution
of the suspected contamination to the failures. The vendor has not
received reports of other failures of 2301A controls.
The NRC will post ensuing reports on this subject on its Web site at
http://www.nrc.gov/NRC/PUBLIC/PART21/2001.
Contact: N. Fields, NRR
301-415-1173
E-mail: enf@nrc.gov
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HEADQUARTERS MORNING REPORT PAGE 3 JULY 20, 2001
Licensee/Facility: Notification:
Part 21 Database MR Number: H-01-0050
General Electric Date: 07/20/01
Subject: Part 21 - Potentially inadequate minimum critical power ratio
safety limit protection with respect to BWR power oscillations
Discussion:
VENDOR: General Electric PT21 FILE NO: m1-23-0
DATE OF DOCUMENT: 06/29/01 ACCESSION NUMBER:
SOURCE DOCUMENT: EN 38104 REVIEWER: REXB, V. Hodge
The vendor, General Electric, reports a potential for nonconservative
reload licensing calculations for plants that implemented stability
detect and suppress trip systems that may result in inadequate minimum
critical power ratio (MCPR) safety limit protection.
Optional stability solutions requiring these calculations are defined as
Options I-D, II, and III in the vendor's document NEDO-32465-A, "Reactor
Stability Detect and Suppress Solutions Licensing Basis Methodology for
Reload Applications," August 1996. This document specifies two generic
curves (Delta CPR/Initial CPR Vs. Oscillation Magnitude (DIVOM curves)),
one for core-wide mode oscillations and one for regional mode
oscillations, relating normalized critical power ratio to hot bundle
oscillation magnitude.
In Option I-D, the generic core-wide curve is used to confirm that the
flow-biased average power range monitor (APRM) flux trip provides
adequate mcpr safety limit protection for a core-wide mode oscillation
initiating on the rated flow control line.
Option II is not specifically addressed in the vendor's document, but the
vendor states that the generic regional mode curve has been used at Nine
Mile Point Unit 1 to confirm that the APRM trip gives adequate MCPR
protection for a regional mode oscillation initiating on the rated flow
control line.
In Option III, the generic regional mode curve is used to determine
setpoints for the implemented stability detect and suppress trip system
to provide adequate MCPR protection. At plants using this option, these
systems may be called oscillation power range monitors.
In recent evaluations, the vendor identified a nonconservative deficiency
for high peak bundle power-to-flow ratios in the generic regional DIVOM
curve and for high core average power-to-flow ratios in the generic
core-wide DIVOM curve. As a result, the Option III system trip setpoint
is overpredicted by the generic regional DIVOM curve and MCPR safety
limit protection is overpredicted for the flow-biased APRM flux trip by
the generic core-wide DIVOM curve.
The vendor states that it informed all affected nuclear power plants,
which includes General Electric boiling water reactor nuclear power
HEADQUARTERS MORNING REPORT PAGE 4 JULY 20, 2001
MR Number: H-01-0001 (cont.)
plants, and the industry boiling water reactor owners' group. The vendor
described compensatory actions taken by the Hatch licensee for both Units
1 and 2. The Perry, Nine Mile Point Unit 2, and Fermi Unit 2 licensees
notified the NRC of their compensatory actions (Event Notifications
38099, 38106, and 38119 respectively).
The vendor expects to provide updated generic DIVOM curves and a
corresponding figure of merit to determined curve applicability by
August 17, 2001.
The NRC will post ensuing reports on this subject on its Web site at
http://www.nrc.gov/NRC/PUBLIC/PART21/2001.
Contacts: V. Hodge, NRR Tai L. Huang, NRR
301-415-1861 301-415-2867
E-mail: cvh@nrc.gov E-mail: tlh1@nrc.gov
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