[Federal Register Volume 60, Number 131 (Monday, July 10, 1995)]
[Notices]
[Pages 35566-35570]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-16808]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Report to Congress on Abnormal Occurrences October-December,
1994; Dissemination of Information
Section 208 of the Energy Reorganization Act of 1974, as amended,
requires NRC to disseminate information on abnormal occurrences (AOs)
(i.e., unscheduled incidents or events that the Commission determines
are significant from the standpoint of public health and safety).
During the
[[Page 35567]]
fourth quarter of CY 1994, the following incidents at NRC licensed
facilities were determined to be AOs and are described below, together
with the remedial actions taken. The events are also being included in
NUREG-0090, Vol. 17, No. 4, (``Report to Congress on Abnormal
Occurrences: October-December 1994''). This report will be available at
NRC's Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC 20037 about three weeks after the publication date of
this Federal Register Notice.
Nuclear Power Plants
94-20 Core Shroud Cracking in Boiling Water Reactors
One of the AO reporting guidelines notes that a major deficiency in
design, construction, or operation having safety implications requiring
immediate attention can be considered an AO. A second reporting
guideline notes that recurring incidents and incidents with
implications for similar facilities (generic incidents) that create a
major safety concern can be considered an AO.
Date and Place--From October 1993 through the present, various
General Electric-designed boiling water reactors.
Nature and Probable Consequences--Intergranular stress corrosion
cracking (IGSCC) of General Electric (GE)-designed boiling water
reactor (BWR) reactor vessel internals has been identified as a
technical issue of concern by both NRC and the industry. Core shroud
cracking as a result of IGSCC was initially discovered overseas and
later identified in operating BWR plants within the United States.
Although no adverse consequences are expected at currently observed
levels of shroud cracking, it has been postulated that a 360-degree
through-wall core shroud crack in concert with a loss-of-coolant
accident has the potential to lead to core damage.
NRC has been meeting every year since 1988 with the BWR Owners
Group (BWROG) and GE to review the generic safety implications of
potential failure of reactor internals, with IGSCC as one of the
failure mechanisms of concern.
Cracking of BWR core shrouds was first observed in an overseas BWR
in 1990. It was located in the heat affected zone of a circumferential
weld in the beltline elevation of the shroud, and was reported by GE
via Rapid Information Communication Services Information Letter
(RICSIL) 054. The core shroud is a stainless steel cylinder which
performs the following functions: (1) Separates feedwater in the
reactor vessel's downcomer annulus from cooling water flowing through
the reactor core, (2) maintains core geometry, and (3) provides a
refloodable volume under postulated accident conditions. The potential
loss of a refloodable volume under accident conditions has the
potential of resulting in core damage making BWR core shroud cracking
the most significant concern related to potential failures of reactor
internals reported during 1993 and 1994.
In response to this concern, several actions were taken by NRC. In
a meeting with the BWROG in January 1992, the staff emphasized that a
comprehensive program should be developed to address internals cracking
and that the utilities should adopt an enhanced inspection program. In
September 1993, Information Notice (IN) 93-79, ``Cracking at the
Beltline Region Welds in Boiling Water Reactors,'' was issued in
response to the discovery of significant circumferential cracking of
the core shroud welds at Brunswick Unit 1. (This event was also
included in NRC's ``Report to Congress on Abnormal Occurrences,
October-December 1993.'' [NUREG-0090, Vol. 16, No. 4]). Following the
additional discovery of significant core shroud cracks at Dresden Unit
3 and Quad Cities Unit 1 in May and June 1994, respectively, NRC issued
IN 94-42 ``Cracking in the Lower Region of the Core Shroud in Boiling
Water Reactors,'' June 7, 1994; IN 94-42 Supplement 1, July 19, 1994;
and Generic Letter (GL) 94-03, ``Intergranular Stress Corrosion
Cracking of BWR Core Shrouds,'' July 25, 1994.
GL 94-03 requested that BWR licensees inspect their core shrouds at
the next refueling outage, and perform a safety analysis to support
continued operation of their facilities until corrective actions were
implemented. During the same period of time, the BWROG initiated the
BWR Vessels and Internals Project (BWRVIP) to facilitate industry
response to the core shroud and internals cracking issues. Licensee
responses to GL 94-03 were received during August and September 1994,
and several BWR licensees began outages in September 1994.
Cause or Causes--IGSCC of BWR vessel internals is a time dependent
material degradation process which is accelerated by the presence of
crevices, residual stresses, material sensitization, irradiation, cold
work and corrosive environments.
Actions Taken To Prevent Recurrence
Licensees--Several domestic BWR licensees performed visual
examinations of their core shrouds in accordance with the
recommendations of GE RICSIL 054 or GE Services Information Letter
(SIL) 572, which was issued in late 1993 and incorporates domestic
experience.
NRC--Because of the extent of cracking observed, NRC evaluated
safety concerns associated with the possibility of a 360-degree
circumferential separation of the shroud following a pipe break. Such
separation might either prevent full insertion of the control rods, or
open a gap in the shroud large enough so that the resulting leakage
would limit adequate core cooling by the emergency core cooling system.
The accident scenarios of primary concern are the main steam line break
and the recirculation line break, which are normally referred to as
loss-of-coolant accidents.
The most serious event associated with cracks in the upper shroud
welds is the steam line break, since the lifting forces generated may
be sufficient to elevate the top guide and potentially affect the
ability to insert rods. The most serious event associated with cracks
in the lower elevations of the core shroud is the recirculation line
break. A recirculation line break concurrent with a 360-degree through-
wall weld failure could cause a lateral displacement of the shroud or
opening of a crack, which would allow enough leakage through the shroud
and out of the break affecting the ability to adequately cool the core.
NRC performed a probabilistic risk assessment of the consequences
of shroud separation at the lower elevation for Dresden Unit 3 and Quad
Cities Unit 1. The assessment estimated the potential contribution to
core damage frequency from a cracked shroud. Assuming that severe
shroud cracking (360-degree through-wall cracking) did exist, a large
rupture of either a steam or recirculation line would have to occur to
generate sufficiently large loads to move the shroud. No events
involving a large rupture of a steam line or recirculation line have
ever occurred, and probabilistic risk assessments have shown that such
ruptures have a low probability of occurring. Furthermore, for welds in
the upper portion of the shroud, such extensive degradation in and of
itself can be detected during normal operation by a power/flow mismatch
condition.
From the above evaluations, NRC made conservative estimates of the
risk contribution to core damage from shroud cracking and concluded
that immediate corrective actions are not necessary. Although immediate
plant shutdowns to implement corrective actions are not necessary,
degradation of the core shroud does have the potential to impact plant
safety. The core shroud provides the important functions of
[[Page 35568]]
properly directing coolant flow through the core, maintaining core
geometry, and providing a refloodable volume under postulated accident
conditions. NRC therefore considers that 360-degree cracking of the
shroud is a safety concern for the long term based on: (1) The
potential to exceed the American Society of Mechanical Engineer Code's
structural margins, if the cracks are sufficiently deep and continue to
propagate through the subsequent operating cycle; and (2) the potential
effects on the ability to protect against core damage.
Even though licensees have justified (through engineering
evaluations) continued operation with significant cracks existing in
core shrouds, BWRs with core shroud materials susceptible to IGSCC will
eventually have to be repaired or modified to inhibit cracking and
thereby assure structural integrity of the shrouds in the long term.
Due to the location and the extent of the cracking recently found,
NRC and the BWROG agreed that additional attention to this issue was
warranted. BWROG met with NRC on June 28, 1994, to announce the
formation of BWRVIP, which is headed by several high level utility
executives to direct its efforts. BWRVIP has since submitted documents
which addressed an integrated safety assessment of the issue,
inspection plans for the reactor internals, and generic criteria for
repairs and flaw acceptance.
NRC has reviewed these documents and concurs with the BWRVIP
recommended generic repair criteria and flaw assessment methodology.
Inspection scope and methodology are still under consideration.
In addition to the aboves actions, in order to verify compliance
with the structural integrity requirements of 10 CFR 50.55a and to
assure that the risk associated with core shroud cracking remains low,
NRC concluded that it is appropriate for BWR licensees to implement
timely inspections and/or repairs, as appropriate, at their plants. To
implement this position, NRC issued GL 94-03 (July 25, 1994) which
requested BWR licensees to inspect their core shrouds by the next
outage and to justify continued safe operation until all appropriate
corrective actions have been implemented.
* * * * *
Other NRC Licensees
(Industrial Radiographers, Medical Institutions, Industrial Users,
etc.)
94-21 Recurring Incidents of Administering Higher Doses Than
Procedurally Allowed for Diagnostic Imaging at Ball Memorial Hospital
in Muncie, Indiana
One of the AO reporting guidelines notes that a serious deficiency
in management or procedural controls in a major area can be considered
an AO.
Date and Place--October 1988 through June 1993; Ball Memorial
Hospital; Muncie, Indiana.
Nature and Probable Consequences--On July 19, 1993, NRC was
notified that nuclear medicine technologists employed by the licensee
had increased the dosages of radiopharmaceuticals used in diagnostic
studies. NRC was also informed that the technologists had falsified the
required records of the dosages administered.
On July 21 through August 9, 1993, NRC conducted an inspection of
the licensed facility. The inspection revealed that since 1988, nuclear
medicine technologists employed by the licensee have been administering
radiopharmaceutical dosages above the approved dose ranges for
diagnostic image studies by as much as 40 percent. The inspection also
verified that subsequent to administering high doses, the technologists
entered false information in NRC-required records. The doses were
increased for imaging studies of the lung, liver, bone, and
gastrointestinal tract using technetium-99m and xenon-133.
NRC inspectors did not identify any medical misadministrations, as
defined in 10 CFR 35.2, as a result of this practice of administering
higher than approved doses for diagnostic imaging.
Cause or Causes--According to the licensee, one technologist told
licensee officials that dosages were increased to minimize patient
discomfort, to reduce imaging time for critically ill patients and to
enhance the clarity of images for studies performed on obese patients.
Action Taken To Prevent Recurrence
Licensee--The licensee conducted an internal review. Based on the
findings from this review, the licensee initially suspended two nuclear
medicine technologists from all NRC-licensed activities. Subsequently,
the licensee terminated one of the two individuals and the other
individual was allowed to continue to perform duties that do not
involve NRC-licensed activities.
The licensee also committed to a number of corrective actions. Some
of the corrective actions include: Assigning a pharmacist or a
radiologist to verify all radioisotope dosages; implementing a unit
dose system; obtaining the services of an assistant radiation safety
officer; and conducting monthly and quarterly audits of the Nuclear
Medicine Section for at least one year.
NRC--A special safety inspection was conducted by NRC from July 21
to August 9, 1993. Subsequent to that inspection, NRC conducted a
followup review.
NRC issued a Confirmatory Action Letter on July 26, 1993, and a
Confirmatory Order Modifying License on October 20, 1993. These
documented specific procedures and verifications to prevent any further
unauthorized increases in patient doses.
On May 23, 1994, NRC issued an Order against a former nuclear
medicine technologist of the licensee. The Order required the
following: (1) Prohibited the technologist from involvement in NRC-
licensed activities for a period of one year; (2) required the
technologist to provide a copy of the Order to any prospective employer
who engages in NRC-licensed activities for a three-year period; and (3)
required the technologist to notify NRC within 20 days of accepting
employment involving NRC-licensed activities.
On May 27, 1994, the technologist requested a hearing and on
September 26, 1994, a settlement agreement was reached. The settlement
was reviewed and approved by the Atomic Safety and Licensing Board on
October 3, 1994. The agreement resulted in the withdrawal of the
requirement for the technologist to provide a copy of the Order to any
prospective employer who engages in NRC-licensed activities. The
settlement retained provisions (1) and (3) of the Order.
* * * * *
94-22 Medical Therapy Misadministration at Veterans Affairs Medical
Center in Long Beach, California
One of the AO reporting guidelines notes that a therapeutic
exposure to any part of the body not scheduled to receive radiation can
be considered an AO.
Date and Place--August 9, 1994; Veterans Affairs Medical Center;
Long Beach, California.
Nature and Probable Consequences--On August 9, 1994, the licensee's
radiation safety officer (RSO) notified NRC of a misadministration
involving a therapeutic dose of strontium-89 (Sr-89).
The RSO reported that a patient scheduled to receive 185
megabecquerel
[[Page 35569]]
(MBq) (5 millicurie [mCi]) of thallium-201 (a radiopharmaceutical not
regulated by NRC) for a myocardial perfusion study was mistakenly
administered 148 MBq (4 mCi) of Sr-89 (which is regulated by NRC).
Based on the misadministration of the Sr-89, the licensee estimated
that the patient received 250 centigray (250 rads) to the surface of
the bone. The RSO reported that no action was taken to mitigate the
consequences of the dose (i.e., administration of calcium as a blocking
agent) because the patient had a preexisting heart condition which
could have been exacerbated by administering calcium. The licensee also
stated that medical experts were contacted to assist in an assessment
of potential health effects to the patient. In addition, the licensee
reported that with the exception of emergency procedures, it had
voluntarily suspended all nuclear medicine procedures involving the
intravenous administration of radiopharmaceuticals and had initiated an
internal review of the misadministration.
On August 10, 1994, NRC issued a Confirmatory Action Letter to
confirm the licensee's actions as stated above.
Cause or Causes--The cause of the misadministration was attributed
to the administering technologist's failure to verify the isotope as
well as the dosage (by reading the label on the syringe) prior to
injection.
Actions Taken To Prevent Recurrence
Licensee--Corrective actions initially proposed by the licensee
included the following: (1) Physically separating diagnostic unit
dosages from therapeutic radiopharmaceutical dosages in the licensee's
hot lab; (2) packaging unit dosages received from a local radiopharmacy
in different containers, according to isotopes; and (3) retraining
technologists in requirements for identifying radiopharmaceuticals
prior to injection.
NRC--Two NRC inspectors conducted a special safety inspection on
August 10-12 and 17-19, 1994, to review the circumstances associated
with the misadministration and to review the licensee's corrective
actions. In addition, NRC contracted a medical physician consultant to
assist in its evaluation of the potential consequences of the patient's
radiation exposure. The consultant stated that there were no adverse
health effects to the patient.
An Enforcement Conference was held with the licensee on November
30, 1994, to discuss an apparent violation involving the failure of an
individual working under the supervision of an authorized user
physician to follow the licensee's written radiation safety procedures.
Additional concerns discussed during the conference included the
licensee's use of an informal labeling system for unit
radiopharmaceuticals which was identified as a potential programmatic
weakness. The licensee presented information during the conference
which supported its view that the error which led to the August 9,
1994, misadministration was an isolated failure rather than a
programmatic problem.
Based on its review of information developed during the inspection
and information provided during the Enforcement Conference, NRC
concluded that the misadministration was the result of an isolated
failure. A Notice of Violation was issued on December 29, 1994, for a
violation involving the failure of an individual working under the
supervision of a physician authorized user to follow the licensee's
written procedures for verifying a radiopharmaceutical dose prior to
administration to a patient. The violation was categorized as a
Severity Level IV violation.
* * * * *
94-23 Medical Brachytherapy Misadministration at North Memorial Medical
Center in Robbinsdale, Minnesota
One of the AO reporting guidelines notes that a therapeutic
exposure to any part of a body not scheduled to receive radiation can
be considered an AO.
Date and Place--August 3, 1994; North Memorial Medical Center;
Robbinsdale, Minnesota.
Nature and Probable Consequences--On August 15, 1994, a licensee
informed NRC that a patient received 1380 centigray (cGy) (1380 rads)
to a wrong treatment site during a brachytherapy treatment for
metastatic lung cancer.
On August 3, 1994, a catheter was inserted into the patient's
bronchus and a ribbon containing 20 seeds of iridium-192 having a total
activity of 673.4 megabecquerel (18.2 millicuries) was then inserted
into the catheter and moved to the proper treatment location. The
treatment plan was intended to deliver a prescribed dose of 2000 cGy
(2000 rads) to the intended target. The treatment began at 11:15 a.m.
on August 3, 1994, and continued until its scheduled completion at
10:15 a.m. on August 4, 1994.
At about 7 p.m. on August 3, 1994, a nurse informed the physician
that the visible portion of the catheter appeared to be protruding
approximately 25.4 to 30.5 centimeters (10 to 12 inches) from the
patient's nose. This was a significantly greater protrusion than
previously observed, indicating that the catheter had moved from its
initial placement. The nurse secured the catheter in place with
additional tape. The physician stated that, based on the information
available to him at that time, he determined that the catheter and
ribbon had moved but that the tumor was receiving some radiation dose
and therefore he continued the treatment. The iridium-192 seeds were
removed on August 4 as planned. On August 4, 1994, a staff radiologist
read the portable x-ray film taken on August 3, 1994, and indicated
that the iridium implant was not seen.
Due to catheter displacement, the tumor dose was significantly
reduced and estimated to be 620 cGy (620 rads) or 31 percent of the
intended dose. The remaining dose of 1380 cGy (1380 rads) was delivered
to an unintended site.
The patient was notified of the event by the treating physician on
August 4, 1994, and again by another physician on August 17, 1994. The
referring physician was informed by the treating physician on August 4,
1994.
An NRC medical consultant was retained to perform a clinical
assessment of this misadministration. The medical consultant concluded
that it is improbable that the patient will experience any long term
consequences as a result of the exposure to the unintended treatment
site.
Cause or Causes--The licensee has determined that the catheter
movement caused a misadministration of the intended dose. Two possible
explanations for the catheter movement could be the following: (1)
Failure to properly secure the catheter in place with tape; or (2)
nasal discharge decreasing the adhesive capability of the tape.
Action Taken To Prevent Recurrence
Licensee--The licensee's corrective actions include: amending the
nursing staff procedure so that the attending physician will be
contacted if there are further questions; directing nurses to follow
the standing protocol for obtaining an administrative consult;
providing additional inservice training; documenting the final length
of the catheter in the patient chart; and documenting the catheter
position on each visit to the patient's room.
NRC--NRC conducted a safety inspection from August 15 through
September 7, 1994, to review the circumstances of the
misadministration. One apparent violation and one area of concern were
identified. An Enforcement Conference was held with
[[Page 35570]]
the licensee on October 11, 1994. Enforcement action is pending. NRC is
continuing its review.
* * * * *
A copy of NUREG-0090, Vol. 17, No. 4 is available for inspection or
copying for a fee at the NRC Public Document Room, 2120 L Street NW.,
(lower level), Washington, DC 20037, or at any of the nuclear power
plant Local Public Document Rooms throughout the country.
Copies of this report (or any of the previous reports in this
series), may be purchased from the Superintendent of Documents, U.S.
Government Printing Office, Post Office Box 37082, Washington, DC
20013-7082. A year's subscription to the NUREG-0090 series publication,
which consists of four issues, is also available.
Copies of the report may also be purchased from the National
Technical Information Service, U.S. Department of Commerce, 5285 Port
Royal Road, Springfield, VA 22161.
Dated at Rockville, MD this 3rd day of July 1995.
For the Nuclear Regulatory Commission.
John C. Hoyle,
Secretary of the Commission.
[FR Doc. 95-16808 Filed 7-7-95; 8:45 am]
BILLING CODE 7590-01-M
