[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