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 

ML070100459
 “Fracture of Type 17-4 PH CRDM Lead Screw Male Coupling Tangs,” by H. Xu and S. 1

Fyfitch, the 11 International Conference on Environmental Degradation of Materials in Nuclear th

Power Systems-Water Reactors, ANS: Stevenson, WA (2003).

 IN 2007-02 

Page 2 of 3 

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) . The operating experience at ONS shows that thermally embrittled 17-4 1

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. 
 IN 2007-02 

Page 3 of 3 

CONTACTS

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 NRC’s 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.