Rod Control System Failure and Widthdrawal of Rod Control Cluster Assemblies, 10 Cfr 50.54(f) (Generic Letter 93-04)
June 21, 1993
TO: FOR ACTION - ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION
PERMITS FOR WESTINGHOUSE (W)-DESIGNED NUCLEAR POWER REACTORS EXCEPT
FOR INFORMATION - ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION
PERMITS FOR COMBUSTION ENGINEERING (CE)-DESIGNED AND BABCOCK AND
WILCOX (B&W)-DESIGNED NUCLEAR POWER REACTORS AND HADDAM NECK
SUBJECT: ROD CONTROL SYSTEM FAILURE AND WITHDRAWAL OF ROD CONTROL CLUSTER
ASSEMBLIES, 10 CFR 50.54(f) (GENERIC LETTER 93-04)
The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter
(1) to notify addressees about a single failure vulnerability within the
Westinghouse solid state rod control system that could cause an inadvertent
withdrawal of control rods in a sequence resulting in a power distribution not
considered in the design basis analyses, and (2) to require, in accordance
with Section 50.54(f) of Part 50 of Title 10 of the Code of Federal
Regulations (10 CFR 50.54(f)), that all action addressees provide the NRC with
information describing their plant-specific findings related to this issue and
actions taken. The NRC will use this information to assess licensee
compliance with the plant-specific licensing basis regarding single failures
in the rod control system.
The staff issued Information Notice 93-46, "Potential Problem With
Westinghouse Rod Control System and Inadvertent Withdrawal of a Single Rod
Control Cluster Assembly," dated June 10, 1993, to alert licensees to the
potential for an inadvertent withdrawal of one or more rod control cluster
assemblies in Westinghouse plants in response to an insert signal.
Description of Circumstances
On May 27, 1993, operators at the Salem Nuclear Generating Station, Unit 2,
experienced problems with the rod control system. During an attempt to
withdraw Shutdown Bank A, the operator observed that the analog rod position
indicator (ARPI) did not indicate that the control rods were being withdrawn.
The operator stopped attempting to withdraw rods at 20 steps as indicated on
the group demand indicator. At this time the ARPI indicated that all of the
rods in Shutdown Bank A were at the 0 step position. (The function of the
group demand indicator is to provide the operator with information on the
Generic Letter 93-04 - 2 - June 21, 1993
position to which the rods should have moved on the basis of the demand from
the rod control system. The function of the ARPI is to show the actual
position of each rod.) The operator then attempted to insert Shutdown Bank A.
However, one control rod (1SA3) withdrew to 8 steps as indicated by the ARPI
while the group demand indicator counted down from 20 steps to 6 steps. The
operator continued to try to insert the Shutdown Bank A control rods until the
group demand indicator showed a rod position of zero. The operator observed
that the indicated position on the ARPI for control rod 1SA3 was 15 steps.
Public Service Electric & Gas (the licensee) removed power from the rod by
pulling fuses, and rod 1SA3 dropped to the 0 step position as indicated by the
The licensee initiated troubleshooting activities on the rod control system at
Salem, Unit 2. An NRC augmented inspection team (AIT) was sent to Salem
Unit 2 to evaluate this issue and observe the investigation of this event by
The licensee, in response to NRC questions, has postulated that, for the event
that occurred on May 27, 1993, a single failure in the rod control system
caused a single rod to withdraw from the core 15 steps while the operator was
applying a rod insertion signal. The failure of an integrated circuit on a
slave cycler decoder card disrupted the normal sequence of pulses that the rod
control system sends to the rods in the selected bank. Normally, on insert
demand, the pulses are staggered in a sequence that leads to rod insertion.
With the failure, the rod control system periodically sent simultaneous pulses
to the movable gripper coil, lift coil, and stationary coil for each of the
rods in the selected bank. Under these conditions each rod in the bank may
either remain where it is or withdraw from the core when a rod movement demand
When the rod control system is in the automatic mode of operation, a rod
movement demand is generated automatically in response to changes in turbine
load and changes in the average reactor coolant temperature. Rod movement
then occurs without any operator action until the demand is satisfied. When
the rod control system is in the manual mode of operation, a rod movement
demand is generated only in response to operator manipulation of the IN-HOLD-
OUT switch, given no failures in the demand circuit.
The rod control system installed at Salem Unit 2 is used at all Westinghouse-
designed pressurized-water reactors (PWRs) except Haddam Neck. Initial
assessments by Westinghouse showed that a single failure in the rod control
system could result in unintended rod withdrawal movements in multiple control
rods. Although the reactor protection system is independent of the rod
control system logic and, therefore, the scram function is not compromised,
there remains a concern that a previously unanticipated single-failure
mechanism may exist in the control system that can initiate or aggravate
reactivity excursions and result in fuel failure. This is of particular
importance since the frequency of demand on the rod control system for power
adjustments is very high (daily in many plants)..
Generic Letter 93-04 - 3 - June 21, 1993
General Design Criterion (GDC) 25, "Protection system requirements for
reactivity control malfunctions," of Appendix A to 10 CFR 50 specifies that
acceptable fuel design limits not be exceeded for any single malfunction of
the reactivity control systems. The Standard Review Plan (NUREG-0800)
Sections 15.4.1, 15.4.2 and 15.4.3, discuss the specified acceptable fuel
design limits for single failures in the reactivity control system (in this
case the rod control system). One of these fuel design limits is that fuel
rods do not violate the minimum departure from nucleate boiling ratio (DNBR)
The staff requested activation of the Westinghouse Owners Group Regulatory
Response Group (RRG) on June 8, 1993. The Westinghouse Owners Group RRG met
with the staff on June 14, 1993, to discuss the RRG generic safety assessment
of the Salem event. The RRG concluded that the failure can produce a
withdrawal signal if either a manual or automatic insert command is given to
any rod control cluster assembly (RCCA) bank or overlapping banks. The RRG
also discussed analysis results showing that asymmetric RCCA withdrawal at
power and from a subcritical condition are the limiting cases. For both of
these cases conservative bounding evaluations indicate that a small percentage
(less than 5 percent) of the fuel rods experience a calculated DNBR below the
The staff believes that the safety significance of this issue is not high
based on the following information:
o All automatic safety functions will perform as designed.
o For the worst cases of single failures in the rod control system only a
small number (or none) of the fuel rods will be below the DNBR limit.
o Not all events will lead to fuel rods below the DNBR limit.
Furthermore these events do not provide a challenge to the reactor coolant
system or the containment boundary. Although the staff believes that the
safety significance of this issue is not high, it believes that compliance
with plant-specific licensing bases is in question for all action addressees.
GDC 25 specifies that acceptable fuel design limits not be exceeded for any
single malfunction of the reactivity control systems. The analyses discussed
by the Westinghouse Owners Group indicated that fuel failures could result
from single failures identified as a result of the Salem event.
Westinghouse issued a Nuclear Safety Advisory Letter (NSAL) 93-007, dated
June 11, 1993, recommending the following actions:
1. Licensed operators should continue the normal process of verifying
that rod motion is proper for required movement.
2. Licensees should confirm the functionality of rod deviation
3. Operators should review the advisory letter to ensure their
understanding of the event.
Generic Letter 93-04 - 4 - June 21, 1993
4. The Westinghouse Owners Group (WOG) survey its members regarding
rod misalignment events and provide a summary.
Implementation of the recommendations in the Westinghouse NSAL is judged by
the NRC staff to be a prudent action.
The licensee for Salem is implementing several compensatory actions prior to
the startup of either unit. These actions include:
o Enhanced rod control system surveillances prior to startup and during
o More frequent periodic surveillances of the rod control system
o Modification of the startup procedure to preclude an asymmetric rod
withdrawal from the subcritical condition by
- first pulling control rods while still highly borated to the
estimated critical position, then,
- deborating to criticality
o Classroom and simulator training addressing the effects of potential
single failures in the rod control system
o Issuance of standing orders to heighten operator awareness of potential
rod control system malfunctions
o Review of event response procedures to assure adequate guidance to
operators in the event of a rod control system malfunction
Pursuant to Section 182a of the Atomic Energy Act of 1954, as amended, and
10 CFR 50.54(f), each action addressee is required to submit written
information as follows:
1. Within 45 days from the date of this generic letter:
(a) Provide an assessment of whether or not the licensing basis for
each facility is still satisfied with regard to the requirements
for system response to a single failure in the rod control system
and provide a supporting discussion for this assessment in light
of the information generated as a result of the Salem event.
(b) If the assessment in 1(a) indicates that the licensing basis is
o provide an assessment of the impact of potential single
failures in the rod control system on the licensing basis of
Generic Letter 93-04 - 5 - June 21, 1993
o describe any compensatory short-term actions taken or that will
be taken to address any actual or potential degraded or
nonconforming conditions (see Generic Letter 91-18,
Reference 1) such as
- additional cautions or modifications to surveillance and
preventive maintenance procedures
- additional administrative controls for plant startup and
- additional instructions and training to heighten operator
awareness of potential rod control system failures and to
guide operator response in the event of a rod control system
2. If the assessment in 1(a) indicates that the licensing basis is not
satisfied, within 90 days from the date of this generic letter provide a
plan and schedule for the long-term resolution of this issue.
Address the required written reports to the U.S. Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington, D.C. 20555, under oath
or affirmation under the provisions of Section 182a of the Atomic Energy Act
of 1954, as amended, and 10 CFR 50.54(f). In addition, submit a copy to the
appropriate regional administrator. This generic letter requires information
that will enable the NRC to verify that the licensee is complying with its
current licensing basis regarding single failure vulnerability within the rod
control system. Accordingly, an evaluation justifying this information
requirement is not necessary in accordance with 10 CFR 50.54(f).
This generic letter does not involve any backfitting. It only requires (under
the provisions of 10 CFR 50.54(f)) the submittal of information needed by the
NRC staff to assess the compliance by the action addressees with existing NRC
rules and regulations.
Although the staff believes that the safety significance of the issue
addressed by this generic letter is not high, there is an urgency to the
information requirement involved based on the consideration that plants may be
currently operating outside of their licensing bases and the information is
needed promptly to allow the staff to assess this situation. Therefore, a
notice of opportunity for public comment on this generic letter was not
published in the Federal Register.
Generic Letter 93-04 - 6 - June 21, 1993
Paperwork Reduction Act Statement
This generic letter contains information collection requirements that are
subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.).
These requirements were approved by the Office of Management and Budget,
Approval Number 3150-0011, which expires June 30, 1994.
The public reporting burden for this collection of information is estimated to
average 40 hours per response, including the time for reviewing instructions,
searching existing data sources, gathering and maintaining the data needed,
and completing and reviewing the collection of information. Send comments
regarding this burden estimate or any other aspect of this collection of
information, including suggestions for further reducing reporting burden, to
the Information and Records Management Branch (MNBB-7714), U.S. Nuclear
Regulatory Commission, Washington, D.C. 20555; and to the Desk Officer, Office
of Information and Regulatory Affairs, NEOB-3019, (3150-0011), Office of
Management and Budget, Washington, D.C. 20503.
Compliance with the request for the following information is purely voluntary.
The information would assist the NRC in evaluating the cost of complying with
this generic letter.
(1) the licensee staff time and costs to perform requested inspections,
corrective actions, and associated testing
(2) the licensee staff time and costs to prepare the requested reports and
(3) the additional short-term costs incurred as a result of the inspection
findings such as the costs of the corrective actions or the costs of
(4) an estimate of the additional long-term costs that will be incurred in
the future as a result of implementing commitments such as the estimated
costs of conducting future inspections or increased maintenance
The NRC is issuing this generic letter to the information addressees to alert
them to a problem with the Westinghouse rod control system and inadvertent
withdrawal of a control rod. It is expected that recipients will review the
information for applicability to their facilities and consider actions, as
appropriate, to avoid similar problems. However, the requested actions and
reporting requirements applicable to the action addressees are not applicable
to the information addressees; therefore, no specific action or written
response is required from them.
Generic Letter 93-04 - 7 - June 21, 1993
If you have any questions about this matter, please contact one of the
technical contacts listed below or the appropriate Office of Nuclear Reactor
Regulation project manager.
ORIGINAL SIGNED BY
James G. Partlow
Associate Director for Projects
Office of Nuclear Reactor Regulation
List of Recently Issued NRC Generic Letters
Technical contacts: Margaret Chatterton, NRR
Timothy Collins, NRR
Lead project manager: Thomas Alexion, NRR
Generic Letter 91-18, "Information to Licensees Regarding Two NRC
Inspection Manual Sections on Resolution of Degraded and Nonconforming
Conditions and on Operability," issued November 7, 1993
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