Information Notice 2007-02, Recent Operating Experience with Failure of the Male Coupling of the Lead Screw on the Control Rod Drive Mechanism
| ML070100459 | |
| Person / Time | |
|---|---|
| Issue date: | 03/07/2007 |
| From: | Michael Case NRC/NRR/ADRA/DPR |
| To: | |
| Omid Tabatabai, NRR/DIRS 301-415-6616 | |
| References | |
| IN-07-002 | |
| Download: ML070100459 (4) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
March 7, 2007
NRC INFORMATION NOTICE 2007-02:
FAILURE OF CONTROL ROD DRIVE
MECHANISM LEAD SCREW MALE COUPLING
AT A BABCOCK AND WILCOX-DESIGNED
FACILITY
ADDRESSEES
All holders of operating licensees for nuclear power reactors, except those who have
permanently ceased operations and have certified that fuel has been permanently removed
from the reactor vessel.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to inform
addressees about the discovery of a failed control rod drive mechanism (CRDM) lead screw
male coupling during a maintenance activity at Oconee Nuclear Station (ONS), Unit 3. It is
expected that recipients will review the information for applicability to their facilities and consider
actions, as appropriate, to avoid similar problems. However, suggestions contained in this
information notice are not NRC requirements; therefore, no specific action or written response
is required.
DESCRIPTION OF CIRCUMSTANCES
In a letter dated November 23, 2004, Framatome ANP (the vendor) informed NRC of the failure
of two CRDM lead screw male couplings in a pressurized-water reactor, designed by Babcock
and Wilcox (B&W), at ONS. During an outage at ONS in April 2001, Duke Power (the licensee)
discovered that two tangs in two separate CRDM male couplings in ONS, Unit 3 were fractured.
CRDMs are mounted on the reactor vessel head. The lead screw and the roller nuts on the
CRDMs convert rotary motion into linear motion. The lead screw connects the CRDM to the
control rod assembly (CRA). Specifically, the male coupling at the end of the lead screw inserts
into the female coupling on top of the CRA. The four equally spaced tangs on the male
coupling must align with the four slots in the female coupling. The tangs and slots must align
for the CRA and lead screw to be coupled properly. In addition, a stop pin inside the female
coupling acts as a rotational hard stop that serves to confirm that the CRA is properly coupled
to the lead screw.
The lead screw male coupling is made of 17-4 precipitation-hardened (PH) martensitic stainless
steel (17-4 PH steel). Failure analysis of the lead screw male coupling indicated the material
had lost ductility due to a phenomenon called thermal embrittlement. The licensee determined
1 Fracture of Type 17-4 PH CRDM Lead Screw Male Coupling Tangs, by H. Xu and S.
Fyfitch, the 11th International Conference on Environmental Degradation of Materials in Nuclear
Power Systems-Water Reactors, ANS: Stevenson, WA (2003). that the root cause of the male coupling failures was the combination of thermal embrittlement
of the 17-4 PH steel and the excessive force that was used during refueling outages when
coupling and decoupling the lead screw from the CRA. The licensee noted a low load on the
male couplings during normal operation and a significant load only when the control rods are
initially withdrawn from the core prior to start-up operation. The licensee concluded that a
simultaneous failure of two or more male couplings is unlikely.
Licensee corrective actions included revising the procedure used during the refueling outage for
coupling and decoupling the lead screw to minimize the force on the couplings. Additionally, the licensee performed control rod drop analyses for ONS, Units 1, 2, and 3 and concluded that
the analyses were acceptable and bounding.
The vendor advised all B&W licensees to perform similar control rod drop analyses and to
revise their procedure for coupling and decoupling the CRDM lead screw. Some B&W
licensees have replaced the CRDM male couplings and scheduled an inspection of these
couplings during future refueling outages. Other B&W licensees have revised their coupling
and decoupling procedures to minimize the force applied to the CRDM male couplings.
DISCUSSION
When exposed to a light-water reactor temperature of 550E F, the 17-4 PH steel that has
previously been subjected to aging (heat treatment) at 1100E F can experience thermal
embrittlement and an increase in hardness (i.e., a reduction in Charpy V notch fracture
toughness values)1 . The operating experience at ONS shows that thermally embrittled 17-4 PH steel is susceptible to failure when exposed to unexpected loading conditions. These
loading conditions may not be part of the original design basis analyses and may be introduced
either during maintenance or normal operations. Therefore, synergistic effects of unexpected
loading conditions and thermal embrittlement can cause failure of 17-4 PH steel components in
the reactor coolant system (RCS).
Thermal embrittlement of 17-4 PH steel cannot be identified by typical in-service inspection
activities. However, by performing visual or other inspections, licensees can identify cracks
which could lead to failure of the embrittled component prior to component failure. Licensees, thus, can prevent the deleterious effects of thermal embrittlement in the 17-4 PH steel
components by identifying aging degradation (i.e., cracks), implementing early corrective
actions, and monitoring and trending age-related degradation.
CONTACT
S
This information notice does not require any specific action or written response. If you have
any questions about the information in this notice, please contact one of the technical contacts
listed below or the appropriate project manager in the NRCs Office of Nuclear Reactor
Regulation (NRR).
/RA/
Michael J. Case, Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contacts:
Omid Tabatabai, NRR/DIRS
(301) 415-6616 E-mail: oty@nrc.gov
Ganesh Cheruvenki, NRR/DCI
(301) 415-2501 E-mail: gsc@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
CONTACT
S
This information notice does not require any specific action or written response. If you have
any questions about the information in this notice, please contact one of the technical contacts
listed below or the appropriate project manager in the NRCs Office of Nuclear Reactor
Regulation (NRR).
/RA/
Michael J. Case, Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contacts:
Omid Tabatabai, NRR/DIRS
(301) 415-6616 E-mail: oty@nrc.gov
Ganesh Cheruvenki, NRR/DCI
(301) 415-2501 E-mail: gsc@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
Distribution: IN Reading File
ADAMS ACCESSION NO: ML070100459 OFFICE
NRR:DCI:CVIB
NRR:DIRS:IOEB
Tech Editor
BC:NRR:DCI:CVIB
TL:NRR:DIRS:IOEB
NAME
GCheruvenki *
OTabatabai*
HChang*
MMitchell*
JThorp*
DATE
1/11/2007
1/11/2007
1/11/2007
1/16/2007
1/17/2007 OFFICE
NRR:DPR/PGCB
NRR:DPR:PGCG
BC:DPR;PGCG
D:NRR:DPR
NAME
CMHawes CMH
DBeaulieu
CJackson
MCase
DATE
2/19/2007
2/28/2007
03/07/2007
03/07/2007
- See Previous Concurrence
OFFICIAL RECORD COPY