ML20215L616
| ML20215L616 | |
| Person / Time | |
|---|---|
| Issue date: | 09/30/1986 |
| From: | NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-0090, NUREG-0090-V09-N01, NUREG-90, NUREG-90-V9-N1, NUDOCS 8610290171 | |
| Download: ML20215L616 (71) | |
Text
.-.
NUREG-0090 Vol. 9, No.1 J
Report to Congress on
- Abnormal Occurrences 4
January - March 1986 U.S. Nuclear Regulatory Commission Offica for Analysis and Evaluation of Operational Data p,p'"*%,
e, 2
\\...../
"eA ";8Ala "' =
0090 R PDR
I Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication.
Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161
i NUREG-0090 Vol. 9, No.1 Report to Congress on Abnormal Occurrences January - March 1986 Dita Published: September 1986
.[
Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission W:shington, DC 20665 f*"'%
Previous Reports in Series NUREG 75/090, January-June 1975, NUREG-0090, Vol.3, No.4, October-December 1980, published October 1975 published May 1981 NUREG-0090-1, July-September 1975, NUREG-0090, Vol.4, No.1, January-March 1981, published March 1976 published July 1981 NUREG-0090-2, October-December 1975, NUREG-0090 Vol.4, No.2, April-June 1981, published March 1976 published October 1981 NUREG-0090-3, January-March 1976, NUREG-0090, Vol.4, No.3, July-September 1981, i
published July 1976 published January 1982 NUREG-0090-4, April-June 1976, NUREG-0090, Vol.4, No.4, October-Deceder 1981, published March 1977 published May 1982 NUREG-0090-5, July-September 1976 NUREG-0090 Vol.5, No.1, January-March 1982, pubitshed March 1977 published August 1982 NUREG-0090-6, October-December 1976, NUREG-0090 Vol.5, No.2, April-June 1982, published June 1977 published December 1982 NUREG-0090-7, January-March 1977, NUREG-0090, Vol.5, No.3, July-September 1982, pubitshed June 1977 published January 1983 NUREG-0090-8, April-June 1977 NUREG-0090, Vol.5, No.4, October-December 1982, published September 1977 published May 1983 NUREG-0090-9, July-September 1977, NUREG-0090, Vol.6, No.1, January-March 1983, published November 1977 published September 1983 NUREG-0G90-10, October-December 1977; NUREG-0090, Vol.6, No.2, April-June 1983, published March 1978 published November 1983 NUREG-0090 Vol.1, No.1, January-March 1978, NUREG-0090, Vol.6, No.3, July-September 1983, published June 1978 published April 1984 NUREG-0090, Vol.1, No.2, April-June 1978, NUREG-0090, Vol.6, No.4, October-December 1983, published September 1978 published May 1984 NUREG-0090, Vol.1, No.3, July-September 1978, NUREG-0090, Vol.7, No.1, January-March 1984, published December 1978 published July 1984 NUREG-0090, Vol.1, No.4 October-December 1978 NUREG-0090 Vol.7, No.2, April-June 1984, published March 1979 published October 1984 NUREG-0090, Vol.2, No.1, January-March 1979 NUREG-0090, Vol.7, No.3, July-September 1984, published July 1979 published April 1985 NUREG-0090, Vol.2, No.2, April-June 1979, NUREG-0090 Vol.7, No.4, October-December 1984, published November 1979 published May 1985 NUREG-0090 Vol.2, No.3, July-September 1979, NUREG-0090, Vol.8, No.1, January-March 1985, published February 1980 publie.hed August 1985 NUREG-0090, Vol.2, No.4, October-December 1979, NUREG-0090, Vol.8, No.2, April-June 1985, published April 1980 published November 1985 NUREG-0090, Vol.3, No.1, January-March 1980, NU9EG-0090, Vol.8, No.3, July-September 1985 published September 1980 published February 1986 NUREG-0090, Vol.3, No.2, April-June 1980, NUREG-0090 Vol.8, No.4, October-December 1985, pubitshed November 1980 published May 1986 NUREG 0090 Vol.3, No.3, July-September 1980, published February 1981
ABSTRACT Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety and requires a quarterly report of such events to be made to Con-gress.
This report covers the period from January 1 to March 31, 1986.
The report states that for this reporting period, there were two abnormal occurrences at the nuclear power plants licensed to operate.
The events were (1) a loss of power and water hammer event and (2) a loss of integrated control system power and overcooling transient.
There were five abnormal occurrences at the other NRC licensees.
The events were (1) a rupture of a uranium hexa-flueride cylinder and release of gases, (2) a therapeutic medical misadminis-tration, (3) an overexposure to a member of the public from an industrial gauge, (4) a breakdown of management controls at an irradiator facility, and (5) a tritium overexposure and laboratory contamination.
There were four abnormal occurrences reported by the Agreement States.
Three of the events involved radiation injuries to people working either as radiographers or assistant radi-ographers; the other event involved contamination of a scrap steel facility.
The report also contains information updating some previously reported abnormal occurrences.
l 11.i
CONTENTS Page ABSTRACT........................................................
iii PREFACE........................................................
vii INTRODUCTION.............
vii THE REGULATORY SYSTEM.....................................
vii REPORTABLE OCCURRENCES.....................................
viii AGREEMENT STATES....
ix FOREIGN INFORMATION.....................
x REPORT TO CONGRESS ON ABNORMAL OCCURRENCEF 9Y-MARCH 1986...
1 NUCLEAR POWER PLANTS...........
1 86-1 Loss of Power and Water Hammer Event...............
1 86-2 Loss of Integrated Control System Power and Overcooling Transient..............................
F FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)....
10 86-3 Rupture of a Uranium Hexafluoride Cylinder and Release of Gases...................................
10 OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, etc.)....................
16 86-4 Therapeutic Medical Misadministration..............
16 86-5 Overexposure to a Member of the Public from an Industrial Gauge..............
18 86-6 Breakdown of Management Controls at an Irradiator Facility 19 86-7 Tritium Overexposure and Laboratory Contamination..
21 AGREEMENT STATE LICENSEES..................................
23 AS86-1 Radiation Injury of an Industrial Radiographer.....
24 AS86-2 Contamination of a Scrap Steel Facility............
26 AS86-3 Radiation Injury of an Industrial Radiographer
..s..
28 AS86-4 Radiation Injury of an Industrial Assistant Radiographer.......................................
31 REFERENCES..............................................
35 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA 39 v
1 1
CONTENTS (continued)
Pa2e APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES 41 NUCLEAR POWER PLANTS.......................................
41 1
79-3 Nuclear Accident at Three Mile Island...............
41 85-7 Loss of Main and Auxiliary Feedwater Systems........
42 85-12 Management Control Deficiencies.....................
43 85-13 Inoperable Steam Generator Low Pressure Trip........
44 85-14 Management Deficiencies at Tennessee Valley Authority...........................................
45 85-20 Management Deficiencies at Fermi Nuclear Power Station.............................................
47 OTHER NRC LICENSEES........................................
48 85-10 Breakdown in Management Controls....................
48 APPENDIX C - OTHER EVENTS OF INTEREST...........................
51 REFERENCES (FOR APPENDICES).....................................
59 vi
PREFACE INTRODUCTION The Nuclear Regulatory Commission reports to the Congress each quarter under provisions of Section 208 of the Energy Reorganization Act of 1974 on any abnor-mal occurrences involving facilities and activities regulated by the NRC.
An abnormal occurrence is defined in Section 208 as an unscheduled incident or event which the Commission determines is significant from the standpoint of public health or safety.
Events are currently identified as abnormal occurrences for this report by the NRC using the criteria delineated in Appendix A.
These criteria were promulgated in an NRC policy statement which was published in the Federal Register on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952).
In order to provide wide dissemination of information to the public, a Federal Register notice is issued on each abnormal occurrence with copies distributed to the NRC Public Document Room and all Local Puhlic Document Rooms.
At a minimum, each such notice contains the date and place of the occurrence and describes its nature and probable consequences.
The NRC has reviewed Licensee Event Reports, licensing and enforcement actions (e.g., notices of violations, civil penalties, license modifications, etc.),
generic issues, significant inventory differences involving special nuclear material, and other categories of information available to the NRC.
The NRC has determined that only those events, including those submitted by the Agree-ment States, described in this report meet the criteria for abnormal occurrence reporting.
This report covers the period from January 1 to March 31, 1986.
Information reported on each events includes:
date and place; nature and prob-able consequences; cause or causes; and actions taken to prevent recurrence.
THE REGULATORY SYSTEM The system of licensing and regulation by which NRC carries out its responsi-bilities is implemented through rules and regulations in Title 10 of the Code of Federal Regulations.
To accomplish its objectives, NRC regularly conducts licensing proceedings, inspection and enforcement activities, evaluation of operating experience and confirmatory research, while maintaining programs for establishing standards and issuing technical reviews and studies.
The NRC's role in regulating represents a complete cycle, with the NRC establishing stan-dards and rules; issuing licenses and permits; inspecting for compliance; en-forcing license requirements; and carrying o'n continuing evaluations, studies and research projects to improve both the regulatory process and the protection of the public health and safety.
Public participation is an element of the regulatory process.
In the licensing and regulation of nuclear power plants, the NRC follows the philosophy that the health and safety of the public are best assured through the establishment of multiple levels of protection.
These multiple levels can vii
-a
be achieved and maintained through regulations which specify requirements which will assure the safe use of nuclear materials.
The regulations include design and quality assurance criteria appropriate for the various activities licensed by NRC. An inspection and enforcement program helps assure compliance with the regulations.
Most NRC licensee employees who work with or in the vicinity of radioactive materials are required to utilize personnel monitoring devices such as film badges or TLD (thermoluminescent dosimeter) badges.
These badges are processed
)
periodically and the exposure results normally serve as the official and legal record of the extent of personnel exposure to radiation during the period the badge was worn.
If an individual's past exposure history is known and has been sufficiently low, NRC regulations permit an individual in a restricted area to receive up to three rems of whole body exposure in a calendar quarter.
Higher values are permitted to the extremities or skin of the whole body.
For unre-stricted areas, permissible levels of radiation are considerably smaller.
Per-missible doses for restricted areas and unrestricted areas are stated in 10 CFR Part 20.
In any case, the NRC's policy is to maintain radiation exposures to levels as low as reasonably achievable.
REPORTABLE OCCURRENCES Actual operating experience is an essential input to the regulatory process for assuring that licensed activitie's are conducted safely.
Reporting requirements exist which require that licensees report certain incidents or events to the NRC.
This reporting helps to identify deficiencies early and to assure that corrective actions are taken to prevent recurrence.
For nuclear power plants, dedicated groups have been formed both by the NRC and by the nuclear power industry for the detailed review of operating experience to help identify safety concerns early, to improve dissemination of such infor-mation, and to feed back the experience into licensing, regulations, and operations.
In addition, the NRC and the nuclear power industry have ongoing efforts to improve the operational data system which include not only the type, and quality, of reports required to be submitted, but also the method used to analyze the data.
Two primary sources of operational data are reports submitted by the licensees under the Licensee Event Report (LER) system, and under the Nuclear Plant Reliability Data (NPRD) system.
The former system is under the control of the NRC while the latter system is a voluntary, industry-supported system operated by the Institute of Nuclear Power Operations (INPO), a nuclear utility organization.
Some form of LER reporting system has been in existence since the first nuclear power plant was licensed.
Reporting requirements were delineated in the Code of Federal Regulations (10 CFR), in the licensees' technical specifications, and/or in license provisions.
In order to more effectively collect, collate, store, retrieve, and evaluate the information concerning reportable events, the Atomic Energy Commission (the predecessor of the NRC) established in 1973 a computer-based data file, with data extracted from licensee reports dating from 1969.
Periodically, changes were made to improve both the effectiveness of data processing and the quality of reports required to be submitted by the licensees.
viii
Effective January 1, 1984, major changes were made to the requirements to report to the NRC.
A revised Licensee Event Report System (10 CFR S 50.73) was estab-lished by Commission rulemaking which modified and codified the former LER sys-tem.
The purpose was to standardize the reporting requirements for all nuclear power plant licensees and eliminate reporting of events which were of low indi-vidual significance, while requiring more thorough documentation and analyses by the licensees of any events required to be reported.
All such reports are to be submitted within 30 days of discovery.
The revised system also permits licensees to use the LER procedures for various other reports required under specific sections of 10 CFR Part 20 and Part 50.
The amendment to the Commis-sion's regulations was published in the Federal Register (48 FR 33850) on July 26, 1983, and is described in NUREG-1022, " Licensee Event Report System,"
and Supplements 1 and 2 to NUREG-1022.
Also ef fective January 1,1984, the NRC amended its immediate notification re-quirements of significant events at operating nuclear power reactors (10 CFR S 50.72).
This was published in the Federal Register (48 FR 39039) on August 29, 1983, with corrections (48 FR 40882) published on September 12, 1983.
Among the changes made were the use of terminology, phrasing, and reporting thresholds that are similar to those of 10 CFR S 50.73.
Therefore, most events reported under 10 CFR S 50.72 will also require an in-depth follow-up report under 10 CFR S 50.73.
The NPRD system is a voluntary program for the reporting of reliability data by nuclear power plant licensees.
Both engineering and failure data are to be submitted by licensees for specified plant components and systems.
In the past, industry participation in the NPRD system was limited and, as a result, the Commission considered it may be necessary to make participation mandatory in order to make the system a viable tool in analyzing operating experience, How-ever, on July 8, 1981, INP0 announced that because of its role as an active user of NPRD system data, it would assume responsibility for management and fund-ing of the NPRD system.
INP0 reports that significant improvements in licensee participation are being made.
The Commission considers the NPRD system to be a vital adjunct to the LER system for the collection, review, and feedback of operational experience; therefore, the Commission periodically monitors the progress made on improving the NPRD system.
Information concerning reportable occurrences at facilities licensed or other-wise regulated by the NRC is routinely disseminated by the NRC to the nuclear industry, the public, and other interested groups as these events occur.
Dissemination includes special notifications to licensees and other affected or interested groups, and public announcements.
In addition, information on reportable events is routinely sent to the NRC's more than 100 local public document rooms throughout the United States and to the NRC Public Document Room in Washington, D.C.
The Congress is routinely kept informed of reportable events occurring in licensed facilities.
AGREEMENT STATES Section 274 of the Atomic Energy Act, as amended, authorizes the Commission to enter into agreements with States whereby the Commission relinquishes and the ix
States assume regulatory authority over byproduct, source and special nuclear materials (in quantities not capable of sustaining a chain reaction).
Comparable and compatible programs are the basis for agreements.
Presently, information on reportable occurrences in Agreement State licensed activities is publicly available at the State level.
Certain information is also provided to the NRC under exchange of information provisions in the agreements.
In early 1977, the Commission determined that abnormal occurrences happening I
at facilities of Agreement State licensees should be included in the quarterly reports to Congress.
The abnormal occurrence criteria included in Appendix A is applied uniformly to events at NRC and Agreement State licensee facilities.
Procedures have been developed and implemented and abnormal occurrences reported by the Agreement States to the NRC are included in these quarterly reports to Congress.
FOREIGN INFORMATION The NRC participates in an exchange of information with various foreign govern-ments which have nuclear facilities.
This foreign information is reviewed and considered in the NRC's assessment of operating experience and in its research and regulatory activities.
Reference to foreign information may occasionally be made in these quarterly abnormal occurrence reports to Congress; however, only domestic abnormal occurrences are reported.
s
REPORT TO CONGRESS ON ABNORMAL OCCURRENCES JANUARY-MARCH 1986 NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to operate during the first calendar quarter of 1986.
As of the date of this report, the NRC had determined that the following events were abnormal occurrences.
86-1 Loss of Power and' Water Hammer Event i
The following information pertaining to this event is also being reported con-currently in the Federal Register.
Appendix A (see the second general criterion) of this report notes that a major degradation of essential safety-related equip-ment can be considered an abnormal occurrence.
Date and Place - On November 21, 1985, San Onofre Nuclear Generating Station (SONGS) Unit 1 experienced a partial loss of inplant ac electrical power while the plant was operating at 60 percent power.
Following a manual reactor trip, the plant lost all inplant ac power for 4 minutes and experienced a severe wa*.er hammer in the feedwater system which caused a leak, damaged plant equipment, and challenged the integrity of the. plant's heat sink.
The most significant aspect of the event involved the failure of five safety-related check valves in the feed-water system, without detection, and jeopardized the integrity of safety systems.
The event involved a number of equipment malfunctions, operator errors, and proce-dural deficiencies.
SONGS Unit 1 utilizes a Westinghouse-designed pressurized water reactor.
The plant is operated by Southern California Edison Company (the licensee) and is located in San Diego County, California.
Nature and Probable Consequences - At 4:51 a.m., on November 21, 1985, the plant was operating at 60 percent power when a ground fault was detected by protective relays associated with the "C" transformer, which was supplying offsite power to one of the two safety-related 4160V electrical buses.
The resulting isolation of the transformer caused the safety-related bus to de-energize, which tripped all feedwater and condensate pumps on the east side of the plant.
The pumps on the west side of the plant were unaffected since their power was supplied from another bus which was being fed from the main generator.
The east feedwater pump discharge check valve (FWS-436) failed to seat as the de-energized pump coasted down.
This provided a path for the discharge of the still operating high pressure (1300 psig) west feedwater pump to the low pressure (350 psig) east condensate piping and components.
East flash evaporator condenser tubes became overpressured, ruptured and overpressurized the evaporator shell, causing the shell to develop a fishmouth opening approximately 20 feet long and 2 feet wide, which relieved the pressure.
The operators, as required by emergency procedures dealing with electrical sys-tems, tripped the reactor and turbine-generator.
As a result, the plant experi-enced its first complete loss of steam generator feedwater and inplant electrical power since it began operation.
The manual trip of the main generator caused 1
1
loss of ac power to the remaining inplant loads.
The subsequent 4-minute loss of inplant electric power started the emergency diesel generators (which by design did not load), de energized all safety related pumps and motors, significantly reduced the number of control room instrument indications available for operators to diagnose plant conditions, produced spurious indications of safety injection system actuation, and caused the NRC Emergency Notification System (ENS) phone on the operator's desk to ring spuriously.
Restoration of inplant electric power was delayed by improper operation of an automatic sequencer.
The temporary total loss of steam generator feedwater was the direct result of
~the loss of ac power to' the two main feedwrter and one auxiliary feedwater pump motors, and the designed 3-minute warm-up period of the steam powered auxiliary feedwater pump.
The loss of the feedwater pumps, in combination with the fail-ure of five feedwater check valves to close (one at the discharge of each feed-water pump and one in the feedwater line to each of the three steam generators),
allowed loss of inventory from all three steam generators and the partial voiding of the long horizontal runs of feedwater ;,iping within the containment building.
The subsequent automatic start of feedwater injection by the steam powered auxiliary feedwater pump did not result in the recovery of steam generator level because the auxiliary feedwater being injected into the feedwater lines was flowing backwards through the failed check valves to the ruptured feed heater in the condensate system.
Later, operators isolated the feedwater lines upstream of the failed check valves, as required by procedure, unknowingly initiating the process of refilling the feedwater lines in the containment building.
As the auxiliary feedwater pumps refilled the feedwater piping to the steam generators, conditions were being established for a phenomenon that can generate destructive forces greater than 150,000 pounds-force.
Since the feedwater piping to the steam generators had drained because of the failed check valves, the pipes contained water and steam at high temperature and pressure from the steam generators.
As the auxiliary feedwater system filled the piping with relatively cold water, an instability occurred at the steam / water interface, which created a slug of water in the steam space.
The slug accelerated at great speed, as steam was condensed in front of the slug, until it encountered an obstruction or a change of direction in the piping, such as at an elbow or closed valve.
Upon contact, the slug imparted its energy to the piping with the force of a hammer blow, i.e., a condensation-induced water hammer.
Because of the long (203 feet) horizontal layout of the feedwater piping to the B steam generator and other sustaining conditions, this piping experienced the water hammer.
The forces from the water hammer displaced the 10-inch diameter feedwater piping, distorted its original configuration, and damaged pipe hangers and snubbers.
Outside the containment building, the forces associated with the water hammer were enough to stretch 10 one-half-inch diameter bolts holding the bonnet on a 4-inch bypass check valve by about one-half inch.
All of the bolts were stretched into an hour glass shape.
The steam and water from the check valve body to bonnet interface had sufficient force to blow away the insulation from all the piping located 360 degrees around the check valve.
The significant steam and water leak from this check valve constituted the second leak in the event.
2 i
The design of the steam system at Unit 1 has the three steamlines joined into a common pipe (or steam header) inside the containment building without any valves to prevent simultaneous blowdown of all three steam generators should a leak in a steamline or a feedwater line occur.
Hence, the leak from the B feedwater bypass check valve located outside the containment building communicated with all three steam generators, via the steam header and B feedring, and their steam inventories were vented via the leak to the atmosphere.
In addition, the auxiliary feedwater flow to B steam generator escaped from this leak instead of going to the steam generator.
Despite these problems, operators succeeded in recovering water level in the two steam generators not directly associated with the feedwater piping leak.
With the reestablishment of steam generator levels, the operators safely brought the plant to a stable cold shutdown condition, without a significant release of radioactivity to the environment (a preexisting primary to secondary leak was not exacerbated) and without significant additional damage to plant equipment.
Cause or Causes - The most significant aspect of the event was that five safety-related feedwater system check valves degraded to the point of inoperability with-out detection by the licensee, and that their failure jeopardized the integrity of safety-related feedwater piping.
The root causes of the check valve failures were a combination of inadequate maintenance, inadequate inservice testing, inade-quate design, and inadequate consideration of the effects of reduced power operations.
Actions Taken to Prevent Recurrence Licensee - The licensee has undertaken an extensive study (including testing programs) of the multiple failures associated with the event to determine root causes and effective corrective actions to preclude recurrence.
On April 8, 1986, subsequent to several meetings with NRC staff and the Commis-sion, the licensee submitted a comprehensive report (Ref. 1) documenting the results of their investigations to that date and providing some conclusions and corrective actions being implemented.
The licensee provided additional informa-tion on May 1, 1986 (Ref. 2).
The licensee concluded that the most likely cause of the cable failure which initiated the event was temperature-induced degradation due to the presence of local heat sources such as hot pipe flanges.
Additionally, the licensee concluded that the failure of the five check valves was caused by (1) their proximity to turbulent flow, (2) the fact that the valves were not properly sized for design flow conditions and therefore did not remain fully open in normal operation, (3) the design by which the valve disc was fastened to the valve hinge, and (4) extended reduced flow operation at 90% power which exacerbated the effects of the design deficiencies.
The licensee's actions described in the April 8, 1986 report were extensive and included examinations and corrective actions in the areas of testing, procedures development, training, maintenance, quality assurance, emergency preparedness, post-trip review and safety review programs.
3
(
The licensee committed to and is in the process of implementing a number of corrective actions including repairs and design changes which include redesign and replacement of the damaged feedwater lines, replacement of the failed check valve design with another design, and adding an additional check valve in each feedwater line.
Additionally the licensee has committed to substantial initiatives to improve plant performance.
These initiatives will systematically examine the material condition of the unit and identify and correct systems and components which deviate from defined standard conditions.
The licensee has elicited the aid of recognized experts in this area and has committed to implement necessary ac-tions prior to restart.
Additional actions are being defined to maintain the material standard on an ongoing basis.
NRC - The San Onofre Resident Inspectors arrived at the site shortly after being notified of the event.
They observed licensee actions to assure the plant re-mained in a stable condition and began an initial investigation of the circum-stances associated with the event.
On November 21, 1985, the NRC Region V Regional Administrator forwarded a Con-firmatory Action Letter to the licensee (Ref. 3) indicating, in part, that the licensee would not perform any additional work on equipment that malfunctioned during the event until the NRC investigation could review the licensee's proposed actions.
The letter also confirmed an understanding that the plant was not to be restarted until authorized by the NRC Region V Regional Administrator or his designee.
On November 22, 1985, responsibility for the incident investigation was assigned to a special NRC Incident Investigation Team (IIT) by the NRC Executive Director for Operations at the request of the Region V Regional Administrator, in confor-mance with an NRC staff proposed Incident Investigation Program.
The Team, com-posed of six technical experts, was to (1) determine pertinent facts related to the event, (2) identify the probable cause, and (3) make appropriate findings and conclusions to form the basis for possiole follow-on actions.
The Team began their investigation at the plant site on November 23, 1985.
The equipment which malfunctioned was quarantined.
The Team collected and evaluated information to determine the sequence of opera-tor, plant, and equipment responses during the event and the causes of equipment malfunctions.
The sequence of these responses was determined primarily by inter-viewing personnel who were at the plant during the event and by reviewing plant data from the period immediately preceding and during the event.
The Team also toured the plant to examine the equipment which malfunctioned.the equipment that was key to mitigating the transient, and the control room instrumentation and controls.
The Team also interviewed plant management personnel and NRC Region V personnel who arrived at the site soon after the plant was stabilized about their knowledge of the plant response and operator actions.
By correlating plant records with personnel statements on their actions and observations, the Team was able to compile a description of the event.
The results of the Team's investigation were issued in NUREG-1190 (Ref. 4).
Problems identified included issues specific to SONGS Unit 1 and several pos-sible generic issues.
In addition, the Team concluded that the most significant aspect of the event was that five safety-related feedwater system check valves 4
)
3 degraded to the point of inoperability during a period of less than a year, without detection, and that their failure jeopardized the integrity of safety related feedwater piping.
The root causes of the check valve failures have been determined by the licensee and are under independent review by the NRC.
Potential contributors to this problem include inadequate inservice testing (IST), inadequate design, and inade-quate consideration of the effects of reduced power operations.
The licensee's IST program (submitted to but not yet approved by NRC) provided for testing a sampling of the check valves each quarter, but permitted deferral of testing when plant conditions were inappropriate (e.g., plant in operation).
The testing was also intendad to identify vhlve failure, not degradation or impending failure.
The IST was therefore not effective in identifying the check valve failures before the event occurred.
Finally, reduced power operations at Unit 1 are now routine because of steam generator tube plugging and sleeving, and the reduced feedwater flow may have increased the susceptibility of check valve components to hydraulically-induced vibration.
s The NRC continues to be involved in the resolution of this event and related matters.
The event provided an opportunity for the NRC to learn from experience and to feed back the per.tinent lessons into NRC and licensee activities.
The NRC Executive Director for Operations has directed NRC program managers to conduct an in-depth reappraisal of the effectiveness of their programs in light of the lessons of the SONGS Unit 1 event with the view of making the NRC programs more effective.
An NRC action plan has been developed through a cooperative effort of the Offices of Nuclear Reactor Regulation, Inspection and Enforcement, and Region V.
This plan resulted in three basic types of actions that the staff is undertaking:
I Evaluation of licensee corrective actions and evaluations that are re-
)
quired for restart in accordance with the NRC's action list.
Most i
notably, these include an assessment of the licensee's review of plant material condition and readiness for operation.
Evaluation of generic implications of the SONGS Unit 1 event through a sampling of industry experience and technical evaluations of root causes, e.g., check valve design implementation.
(Lead iesponsibility for resolving generic issues related to check valve failures has been assumed by industry per a meeting with the NRC Executive Director for-l Operations on April 7, 1986.
The NRC will monitor and review industry actions.)
Evaluation of NRC requirements and positions in light of existing imple-mentation practices, root causes of the event, and samples of industry practices.
These actions outline a program that evaluates the SONGS Unit 1 readiness for restart and assures that generic aspects are considered.
On January 6, 1986, the NRC Office of Inspection and Enforcement issued Informa-tion Notice No. 86-01 (Ref. 5) to all nuclear power reactor facilities holding an operating license or a construction permit to inform them of the San Onofre event.
5 Q
Future reports will be made as appropriate.
A A A A A A A A 86-2 Loss of Integrated Control System Power and Overcooling Transient The following information pertaining to this event is also being reported con-currently in the Federal Register.
Appendix A (see the third general criterion) of this report notes that major deficiencies in design, construction, use of, or management controls for licensed facilities or material can be considered an abnormal occurrence.
Date and Place - On December 26, 1985, Rancho Seco Nuclear Generating Station, located in Sacramento County, California, experienced a loss of dc power within the integrated control system (ICS) while the plant was operating at 76 percent power.
Following the loss of ICS de power, the reactor tripped on high reactor coolant system (RCS) pressure followed by a rapid overcnoling transient and automatic initiation of the safety features actuation system on low RCS pressure.
The overcooling transient continued until ICS dc power was restored 26 minutes after its loss.
The significance of the event is that a nonsafety related system failure initiated a plant transient which could have been more severe under other postulated scenarios.
The Rancho Seco Nuclear Generating Station, operated by the Sacramento Municipal i
Utility District (SMUD), is a Babcock & Wilcox (B&W)-designed pressurized water reactor.
Nature and Probable Consequences - At 4:14 a.m. on December 26, 1985, the plant was operating at 76 percent power, when a loss of ICS de power occurred as a result of a single failure.
The loss of dc power to the ICS (a nonsafety-related system) caused a number of feedwater and steam valves to reposition automatically and also caused the loss of remote control of the affected valves from the con-trol room.
In addition, the main feedwater (MFW) pump turbines slowed to minimum speed and the auxiliary feedwater (AFW) pumps started.
The immediate result was a reactor coolant system (RCS) undercooling condition that resulted in the reac-tor tripping on high pressure.
The reactor trip was followed by an overcooling condition that resulted in safety features actuation and excessive RCS cooldown.
The transient was initiated by the failure of a single monitoring module in the nonsafety-related ICS (i.e., the spurious tripping of the power supply monitoring module that interrupted all +/-24 Vdc power).
The most probable cause of this failure was a bad crimp connection in the wiring between the +24 Vdc bus and the power supply monitor which caused the module to sense undervoltage and interrupt all de power.
The operators did not immediately recognize that the dc switches had tripped and therefore did not reset the switches to promptly restore de power within the ICS.
As a result, nonlicensed operators were sent to isolate the affected steam and feedwater valves locally with handwheels.
During the first 7 minutes of the incident, the excessive steam and feedwater flows resulted in a rapid RCS cool-down of over 100*F.
The pressurizer emptied and a small bubble formed in the reactor vessel head.
The RCS cooldown continued and the RCS depressurized to about 1064 psig and then began to repressurize.
This repressurization resulted in the RCS entering the B&W-designated pressurized thermal shock (PTS) region.
6 1
d The atmospheric dump valves and turbine bypass valves were isolated within 9 minutes after the reactor trip.
However, the operators experienced difficulty closing the ICS-controlled AFW flow control valves.
One of the flow control valves was finally shut; however, the second AFW flow control valve was damaged and failed open.
The associated AFW manual isolation valve was found to be stuck open.
Therefore, both AFW pumps continued to feed and overfill one steam gen-er-ator. Water began to overflow into the main steam lines.
About 26 minutes after the reactor trip, the operators ' restored power within the ICS by reclosing two switches in an ICS cabinet.
The operators were then able to close the open AFW flow control valve from the controi room, which stopped the RCS cooldown, and started stabilizing the plant.
The RCS had cooled down a total of 180 F in this 26-minute period.
While changing a valve lineup in the suction of the pump used to supply RCS makeup l
(makeup pump), the last suction valve to the makeup pump was inadvertently shut.
This resulted in the overheating and destruction of the makeup pump.
About 450 gallons of contaminated water were spilled on the floor.
This failure did
]
not directly affect the incident since a high pressure injection (HFI) pump was available to supply RCS makeup.
In addition, the spilled water did not result in any significant onsite or offsite radioactivity release or personnel dose.
Operators later stabilized the plant and brought it to a cold shutdown without a significant release of radioactivity to the environment and without significant additional damage to plant equipment.
The December 26, 1985 overcooling incident did not seriously threaten the integ-rity of the Rancho Seco reactor vessel.
However, the plant has had a number of I
overcooling incidents in its 12 year operating history.
Each time this occurs the potential exists for additional operator errors and equipment failures that might exacerbate the event and threaten reactor vessel integrity.
Thus, the sig-nificance of this incident lies in the fact that under alternate scenarios more serious consequences could occur.
The incident at Rancho Seco was also significant because a single failure in the I
ICS, which is a nonsafety-related system, subjected the plant to an undesirable overcooling transient.
During the transient, the RCS cooled down 100 F in 26 minutes, the pressurizer emptied, a bubble formed in the reactor vessel head, the plant entered the licensee-defined pressurized thermal shock region, the safety features actuation system (LFAS) actuated, and water overflowed from a steam generator into the main steam lines.
Cause or Causes - The fundamental causes for this transient were design weak-nesses and vulnerabilities in the ICS and in the equipment controlled by that system.
These weaknesses and vulnerabilities were not adequately compensated by other design features, plant procedores or operator training.
These weak-nesses and vulnerabilities were largely known by the licensee and the NRC staff by virtue of a number of precursor events and through related analyses and studies.
Yet, adequate plant modifications were not made so that this event would be improbable, or so that its course or consequences would be signifi-cantly altered.
l 7
I
l Actions Taken to Prevent Recurrence Licensee - The licensee has undertaken extensive study (including controlled disassembly, examination and testing) of the multiple failures associated with the event to determine root causes and to take corrective actions to prevent recurrence.
Some specific improvements have been identified by these efforts and are being implemented prior to plant startup.
These are desc.ibed in the licensee's February 19, 1986 summary report to the NRC (Ref. 6).
Plant Modifications 1.
Replacement of power distribution wiring of the ICS Power Supply Monitor, to reduce the resistance in series with the voltage being monitored.
2.
Provisions to ensure proper isolation and control room manual control capability for turbine bypass valves, atmospheric dump valves, and auxiliary feedwater flow control valves, under circumstances of ICS failure.
Training Classroom and simulator training related to response to ICS power loss conditions, handling of overcooling and potential pressurized thermal shock, recovery from safety system actions and implementation of emergency plan procedures.
Maintenance Program 1.
Repair of damaged equipment that is required for normal and abnormal operating conditions.
2.
Verification of acceptable condition of equipment in the non-nuclear systems of the plant.
3.
Development of a preventive maintenance program for non-nuclear balance-of plant equipment.
Emergency Procedures Development of event-related procedures to complement the symptom-related j
emergency procedures, for ICS power loss and safety feature actuation system recovery.
NRC - Upon being notified of the event, the NRC Resident Inspectors for the plant arrived shortly thereafter.
They observed licensee actions to assure the plant remained in a stable condition and began an initial investigation of the circumstances associated with the event.
On December 26, 1985, the Regional Administrator of the NRC Region V Office for-warded two Confirmatory Action letters to the licensee (Refs. 7 and 8) indicating that the licensee would perform a root cause analysis prior to return to power and would not perform any additional work on equipment that malfunctioned during i
the event until the NRC could evaluate the event.
8
On December 27, 1985, an NRC Augmented Inspection Team (AIT) was sent to the site by the Regional Administrator and started transcribed personnel interviews on December 28.
fue initial results of this investigation effort indicated that the event was complex and had potentially significant generic implications.
On December 31, 1985, the responsibility for the incident investigation was expanded to a special NRC Incident Investigation Team by the NRC Executive Director for Operations (EDO) at the request of the Region V Regional Adminis-trator, in conformance with an NRC staff proposed Incident Investigation Program.
The Team, composed of six technical experts, was to (1) fact-find as to what happened, (2) identify the probable cause as to why it happened, and (3) make appropriate findings and conclusions to form the basis for possible follow-on actions.
The Team consisted of the AIT members supplemented by additional staff.
It continued the investigation started by the AIT at the plant site.
The equip-ment which malfunctioned was quarantined.
The Team collected and evaluated information to determine the sequence of opera-tor, plant, and equipment responses during the event and the causes of equipment malfunctions.
The sequence of these responses was determined primarily by inter-viewing personnel who were at the plant during the event and by reviewing plant data for the period immediately preceding and during the event.
The Team also toured the plant to examine the equipment which malfunctioned, the equipment that was key to mitigating the trans-ient, and the control room instrumentation and controls.
The Team also interviewed plant management personnel and NRC Region V personnel who arrived at the site soon after the plant was stabilized about their knowledge of the plant response and operator actions.
By correlating plant records with personnel statements on their actions and observations, the Team was able to compile a picture of the event.
Duringandsubsequenttotheironsiteactivities,theTSmreviewedandconcurred in specific troubleshooting plans developed by the licensee for equipment dis-assembly, inspection and testing.
Several of these activities were witnessed by NRC inspectors.
The results of the Team's investigation are contained in NUREG-1195 (Ref. 9).
Problems identified included issues specific to Rancho Seco and several possible generic issues.
The NRC continues to be involved in the resolution of this event and related matters.
The NRC EDO has directed NRC program managers to conduct further generic and plant specific follow-up actions.
Development of NRC plant specific action plans commenced while the IIT was on-site in January 1986, and have been expanded subsequently to include a review of the completeness of prior staff and licensee actions associated with the control systems.
The NRC EDO has also addressed this event in a January 24, 1986 (Ref. 10) letter to the B&W Owners Group (B&WOG) which stated that the NRC staff will reassess the overall safety of B&W plants.
Following the TMI accident there has been a grow-ing realization among the NRC staff of.the sensitivity of B&W plants to opera-tional transients.
A number of recent events at B&W-designed reactors have reinforced their concerns regarding these designs and lead them to conclude that there is a need to re-examine the basic design requirements for B&W reactors.
While they believe that this reassessment is needed, they also believe that B&W reactors can safely continue to operate in the interim.
The B&WOG has committed to taking a leadership role in the reassessment.
9
The January 24, 1986 letter to the B&WOG from the NRC EDO communicated plans for the design reassessment and outlined the scope envisioned for the study.
A pro-gram plan for the reassessment was subsequently developed by the NRC Office of Nuclear Reactor Regulation (NRR) Staff and transmitted to the B&WOG in a March 13, 1986, letter to the B&WOG Chairman (Ref. 11).
The plan calls for studies in the area of operating experience, transient analysis, and probabilistic risk assess-ment.
The plan also identified those areas that the Staff expects the B&WOG to take the lead or play a major role in completing.
As outlined in a March 21, 1986 memorandum from the NRC ED0 to the NRC Commis-sioners (Ref. 12), the B&WOG will assume a strong leadership role in accomplish-ing key aspects of the overall effort, where such involvement is appropriate.
In a meeting with the Staff on April 8, 1966, they presented their program plan for reducing the reactor trip frequency and improving the transient response of B&W-designed plants.
This program was formally submitted on May 15, 1986.
The NRR Staff is currently reviewing this plan.
In addition to addressing those issues which have arisen directly as a result of the December 26, 1985 cooldown transient, the NRC Region V Office has re-evaluated the status of prior Rancho Seco open inspection findings to identify matters which should be. resolved prior to restart of the plant.
The licensee and NRR have included these in restart plans.
Also, the NRC Staff has encour-aged the licensee to reexamine the status of all critical plant systems to as-sure readiness for operation and maximum reliability, so that operation of the plant may be continued with a low probability of disruption from internal causes.
Some of these efforts will be observed by NRC inspectors.
To supplement this, the licensee has initiated a performance improvement program which will address management, training, and maintenance issues.
These actions outline a program that evaluates the Rancho Seco restart program, and assures that generic aspects are considered.
On January 31, 1986, the NRC Office of Inspection and Enforcement issued Infor-mation Notice No. 86-04 (Ref. 13) to all nuclear power facilities holding an operating license or a construction permit to inform them of the Rancho Seco event.
Future reports will be made as appropriate.
FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)
The NRC is reviewing events reported by these licensees during the first calendar quarter of 1986.
As of the.date of this report, the NRC had determined that the following event was an abnormal occurrence.
86-3 Rupture of a Uranium Hexafluoride Cylinder and Release of Gases The following information pertaining to this event is also being reported con-currently in the Federal Register.
Appendix A (see the general criterion) of this report notes that a major reduction in the degree of protection of the public 10
health or safety can be considered an abnormal occurrence.
In addition, Example 11 (of "For All Licensees") of Appendix A notes that serious deficiency in management or procedural controls in major areas can be considered an abnormal occurrence.
Date and Place - At 11:30 a.m. on January 4,1986, a cylinder filled with uranium hexafluoride (UF ) ruptured while it was being heated in a steam chest at the s
Sequoyah Fuels Corporation's Sequoyah Facility near Gore, Oklahoma.
One worker died from pulmonary edema caused by inhalation of hydroflucric acid, a reaction product of UFs and airborne moisture.
Much of the facility complex and some off-site areas to the south were contaminated with hydrofluoric acid, and a second reaction product, uranyl fluoride.
The interval of release was approximately 40 minutes.
The licensee experienced another incident involving an overfilled uranium hexa-fluoride cylinder on March 13, 1986; however, in this incident the overfilled cylinder was not heated and no damage to the cylinder occurred.
Some other events involving overfilled uranium hexafluoride cylinders at Allied Chemical Company, Metropolis, Illinois, are discussed in the Annex to this abnor-mal occurrence.
Allied Chemical Company is a division of Allied-Signal Corpora-tion of Morristown, New Jersey.
Nature and Probable Consequences - At approximately 10: 00 a.m. on January 3, 1986, the filling of a 14-ton capacity cylinder with UFs was commenced.
This operation continued during the following work shifts.
During the early morning of January 4, a chemical operator was unable to add further material into the cylinder, even though the targeted load of 27,500 pounds had not been achieved.
The cylinder and its attendant cart had been placed on a scale during the filling process in order to monitor the net weight of the cylinder.
At this time, the scale indi-cated that the cylinder contained 26,400 pounds of product.
The chemical operator inspected the cylinder and observed that the cart on which it sat had not been fully moved onto the scale platform.
This condition occurred because the cylinder, being the largest design filled at the facility, was not properly positioned on the cart so as to allow clearance at the front end of the cylinder when the cart was moved onto the scale platform.
When the cart and cylinder were repositioned onto the scale platform, the scale dial indicator registered its maximum possible reading of approximately 29,500 pounds.
The cylinder had been filled with a quantity of UFs in excess of the amount measur-able witn the scale and in excess of the maximum shipping weight specification of the cylinder which is 27,560 pounds.
At approximately 6:15 a.m.,
the chemical operator began to evacuate UFs from the cylinder back into plant process vessels.
He was relieved by the day shift chemical operator at 8:00 a.m.,
and the evacuation process continued until the material began to solidify in the cylinder.
The operator consulted with the assistant shift supervisor, who is the ranking production manager on site, and who instructed the operator to move the cylinder to a steam chest located out-side the process building.
The steam chest was to be used to heat the cylinder to approximately 210 F, thus liquifying the contained UFs.
Although some mate-rial had been removed from the cylinder, the scale indicator still registered approximately 29,500 pounds before the cylinder was removeo.
Heating an over-filled cylinder was later noted to be contrary to company procedures.
11
I At approximately 11:30 a.m.,
the cylinder ruptured in the steam chest.
The cylinder ruptured while it was being heated because of the expansion of uranium hexafluoride as it changed from the solid to the liquid phase.
Liquid UFs flowed from the 4-foot lengthwise rupture and rapidly reacted with moisture in the air to form uranyl fluoride and hydrofluoric acid.
The resulting vapor cloud was car.*ied south by southeast by a wind gusting to 25 mph.
The cloud enveloped the process building, and the acidic vapor fatally injured the chemical operator located within a structure approximately 70 feet southwest of the cylinder.
Most of the approximately 40 workers at the site were in the plant lunch room and qu'ickly evacuated the building.
The airborne release con-tinued for about 40 minutes crossing an interstate highway one mile to the south and private residences beyond.
The licensee immediately notified various local, state, and federal officials.
Four injured workers were transported to a local hospital.
A private physician arrived at the site within one hour of the accident and examined plant workers.
During the afternoon, downwind residents were personally notified to go to nearby hospitals and clinics for examinations.
The NRC Region IV Duty 0.fficer was notified of the incident by the NRC:HQ Opera-tions Officer by pager at approximately 12:25 p.m.
The Region IV Incident Response Center was staffed and communication links with NRC Headquarters and the licensee began at 12:55 p.m.
Six NRC personnel were immediately dispatched and began arriving at the site at 6:00 p.m.
Additional NRC personnel were dispatched to the site during the following days to oversee bioassay of workers and residents, evaluation of offsite effluents, and decontamination of the plant complex.
An NRC Augmented Investigation Team was formed to investigate the incident.
Their findings were reported in NUREG-1179, Vol. 1, published during February 1986 (Ref. 14).
An assessment of the public health impact of the accident was pub-lished during March 1986 as NUREG-1189 (Ref. 15).
After the January accident, t'he licensee planned to drain UFs remaining in plant vessels into 10-ton shipping cylinders in order to enable modification of facili-ties and equipment at the plant.
A procedure for the work was reviewed by NRC and the work commenced on March 12, 1986.
The procedure limited the filling of the cylinders to 20,000 pounds each.
The maximum shipping weight specification of the cylinders was 21,030 pounds.
During the draining process on March 13, 1986, a scale malfunctioned which caused UFs to be drained into a cylinder in excess of both of the above limits.
The final net weight of the cylinder was 26,017 pounds.
Most of the excess material was immediately evacuated from tte cylinder before the UFs solidified.
The final net weight of the cylinder was 21,203 pounds.
l The root cause of this second incident was identified as inadvertent damage to a scale apparently when it was decontaminated after the first incident.
Results l
of the NRC investigation of this overfilling event, together with a report of a detailed metallurgical examination performed on the cylinder damaged on January 4, 1986, were reported in NUREG-1179, Vol. 2, published during June 1986 (Ref. 16).
I Cause or Causes - The NRC Augmented Investigation Team (AIT) which investigated both incidents reported the following causes in NUREG-1179, Vol. 1 and Vol. 2, respectively.
12 l
January 4, 1986 Incident 1.
The cylinder was overfilled because it was not placed fully on the scales.
Plant facilities were not designed to accommodate 14-ton cylinders, and associated equipment were not designed to prevent improper positioning of cylinders on the scales.
2.
The time required for filling the cylinder was long enough to allow par-tial solidification of the UFe, which inhibited product removal from the cylinder.
3.
The precise weight of the cylinder was not readily determinable after it was overfilled.
4.
There was no secondary or alternative way to measure the quantity of mate-rial in a cylinder being filled.
5.
Employees violated company procedures when they heated an overfilled cylin-der.
Workers, including line management personnel, had not been suffi-ciently trained in regard to company procedures.
Procedural controls such as checklists or approval points were not esed.
6.
Equipment for monitoring or automatically venting cylirders that are being l
heated was not used.
In summary, the factors can be aggregated into the following ca ases of the accident:
The physical equipment and facilities used for filling and weighing UFs cylinders were inappropriate for safe use with 14-ton cylinders.
The training of workers in operating procedures and ensuring the implemen-tation of the procedures were not carried out effectively.
March 13, 1986 Incident 1.
The scale used for weighing the cylinder being filled malfunctioned.
2.
The procedures for draining did not include any provisions for ensuring proper scale function.
3.
The supervisor in charge of the operation did not recognize early indica-tions of malfunction.
(An operator advised his management of peculiar scale behavior during the filling of the cylinder.)
Actions Taken to Prevent Recurrence - Both NRC and licensee actions to prevent recurrence are currently in progress.
The fo.llowing summarizes actions as of mid-April 1986.
Licensee - The licensee has committed to keep the plant shut down until equipment modifications are made, plant personnel are retrained, plant procedures are rewritten, organization changes have been implemented, and NRC approves plant restart.
13
NRC - A Lessons Learned Task Group reviewed regulatory practices in regard to such fuel facilities in general.
The Group interviewed appropriate members of the NRC staff, licensee, State, and local authorities.
A Lessons Learned Report was completed in May 1986.
A request to restart the facility was received by NRC in May 1986 and is under review.
NRC is monitoring licensee plant modifi-cation work.
Enforcement actions are pending.
The staff is also compiling a list of followup items that need to be considered and addressed.
Additional items are anticipated from the Lessons Learned lask l
Group and other sources.
Upon completion of the list, action items will be grouped into categories and priorities assigned.
Tasks will be undertaken based on priorities and resource requirements.
In the meantime, the staff is moving ahead on a number of near-term follow-on actions, such as:
(1) verification by NRC of existing emergency phone numbers; (2) requiring licensees to verify quarterly emergency numbers and availability of emergency response assistance; (3) informing DOE and other licensees, who are conducting operations involving UFe, of the accident and providing relevant reports; and (4) conducting an independent review of the material licensing and inspection programs by a study group.
To assure that licensees have an updated list of telephone numbers to the NRC Operations Center and Regional Offices, the NRC Office of Inspection and Enforcement issued Information Notice No. 86-28 on April 24, 1986 (Ref. 17).
The event remains under review by the NRC, and future reports will be made as appropriate.
Annex During the publicity associated with the Sequoyah Fuels Accident, NRC Region III (Chicago) received an inquiry from a newspaper reporter about an incident on December 7, 1984 at Allied Chemical Company, Metropolis, Illinois, involving overfilling and subsequent damage to a uranium hexafluoride cylinder.
The li-censee was asked about the incident and provided the following information.
(The incident had not been previously reported to the NRC.
The licensee stated that it had considered reporting it, but concluded that it did not meet any NRC reporting requirements).
On December 7, 1984, an overfill incident occurred in which a cylinder was overfilled and the cylinder subsequently damaged during heating of the cylinder to remove the excess uranium hexafluoride.
There was no release of any uranium hexafluoride to the environment associated with the incident, and there were no injuries.
In the incident, a 48-inch diameter cylinder, with a maximum capacity of 26,560 pounds, was filled with 33,000 pounds of uranium hexafluoride.
The weight recording device was apparently faulty and showed an incorrect weight during the filling operation.
Based on the length of time the filling had been underway, licensee personnel suspected that it had been overfilled and moved it to another scale to be weighed.
The second scale showed it to contain 33,000 pounds.
The cylinder was returned to the filling position and about 500 pounds of UFe was drawn off before the cylinder cooled and the UFs solidified.
The cylinder was then moved to another fill location where a steam chest was placed over it to heat 14
the cylinder.
A line was attached to the cylinder to draw off the UFs as the cylinder was heated, but the line was blocked.
Plant personnel were unable to clear the line, and so the cylinder was heated for about 2-1/2 hours with the cylinder valve closed.
The steam chest was then removed, and the cylinder was moved to the weighing location to draw off the UFs.
At that time, plant personnel noted that the three stiffening rings which surround the cylinder were cracked at a welded joint.
A portion of the UFs was then drawn 1
off at the scale location, and then the cylinder was moved to another fill loca-tion where the remainder of the UFs was drawn off, while applying heat in a steam chest.
It was later observed that the cylinder was slightly deformed-placing a j
straight edge along the cylinder wall showed a deformation of approximately 1/2 inch.
j The licensee later provided information to the NRC on overfill incidents at the Metropolis facility for the time period 1981 through 1985.
During the five year period, there were 41 overfills--of which three were greater than 1,000 pounds.
The three were 1,183 pounds in 1981, 5,448 in 1984 (described above), and 2,140 in 1985. With the exception of the December 7, 1984 event, none of the other overfill incidents involved damage to the cylinders.
No releases of UFs occurred in any of the incidents.
Another overfill incident occurred on March 23, 1986, when a cylinder was filled with 28,207 pounds (an overfill of 1,367 pounds).
The excess was successfully removed without applying additional heat.
This incident was attributed to the failure of an operator to "zero out" the scale to account for the empty weight of the cylinder combined with the erroneous calculation by another operator of the time required to fill the cylinder.
Subsequent to the December 1984 incident, the licensee installed new load cells (scales) at each fill location to provide clearer, more reliable weight measure-ments in the control room.
A scale was also added to the overhead crane used to lift the cylinders to allow weighing of the cylinders without transporting them more than 50 feet to another weighing location.
After the January 1986 Sequoyah Fuels accident, the licensee installed a flow totalizer which measures the flow rate of the liquid UFs and has an alarm and automatic shutdown function based on total flow and data from the load scale.
The licensee has also initiated improvements in its training and retraining programs, procedures, and level of supervision for cylinder filling activities.
In response to the January 1986 accident at Sequoyah Fuels, NRC Region III con-ducted a special inspection at the Metropolis facility on January 14-15, 1986 to observe the Allied Chemical Company cylinder handling procedures.
Additional inspections were conducted to examine the circumstances of the December 7, 1984, incident, and the licensee's actions to preclude the occurrence,of significant overfills (Ref. 18).
Region III issued a Confirmatory Action Letter to the licensee on January 10, 1986 (Ref. 19), documenting the licensee's agreement that no overfilled cylinders would be heated without the review and concurrence of Region III.
A second Confirmatory Action Letter was issued on March 24, 1986 (Ref. 20), documenting the licensee's planned actions in response to the March 23, 1986 overfill incident.
These cor-rective measures include increased supervision of filling activities, prohibiting 15 h
cylinder filling unless two independent methods are available to determine the amount of UFs in a cylinder, and completion of the installation of the new UFc flow readout and alarm functions by April 15, 1986.
A special NRC inspection, by a seven member team, was conducted in mid-April 1986 to extensively review the licensee's activities.
The team identified two viola-tions of NRC requirements:
radiation survey instruments did not have the required sensitivity; and a procedure concerning the handling of overfilled cylinders did not have all the proper internal approvals (Ref. 21).
An emergency planning inspection was conducted March 31-April 14, 1986, and the inspectors determined that while the licensee had an on-site emergency contin-gency plan, off-site emergency response capablity for the area surrounding the plant was poorly coordinated.
The inspectors also found that training of off-site emergency response personnel to respond to emergencies at the plant was inadequate (Ref. 22).
These items will be reviewed during a future inspection.
In addition the NRC is currently reviewing its requirements for emergency plan-ning at fuel facilities.
Although no regulations exist for off-site emergency response, the licensee has taken the initiative to work with the appropriate off-site groups to establish a coordinated capability.
On June 27, 1986, the NRC forwarded to the Allied-Signal Corporation (parent company of Allied Chemical Company) a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $25,000 (Ref. 23).
The violations included the failing to report the December 7, 1984 incident to the NRC and for three instances of failing to follow procedures during the March 23, 1986 overfill incident.
- aa a*
OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, etc.)
There are currently more than 8,000 NRC nuclear material licenses in effect in the United States, principally for use of radioisotopes in the medical, industrial and academic fields.
Incidents were reported in this category from licensees such as radiographers, medical institutions, and byproduct material users.
The NRC is reviewing events reported by these licensees during the first calen-dar quarter of 1986.
As of the date of this report, the NRC had determined that the following events were abnormal occurrences.
86-4 Therapeutic Medical Misadministration The following information pertaining to this event is also being reported con-currently in the Federal Register.
Appendix A (see the general criteria) of this report notes that an event involving a moderate or more severe impact on public health or safety can be considered an abnormal occurrence.
Date and Place - On February 7, 1986, a patient at Washington Hospital Center, Washington, D.C. received a cobalt-60 teletherapy treatment of 150 rads to the abdomen which was intended for another patient.
16
.A
Nature and Probable Consequences - On February 6, 1986, an attending surgeon of the Renal Transplant Unit ordered radiation therapy as follows far one of his patients:
150 rads per day to be repeated every other day for a total of 600 rads.
The treatment was intended to forestall rejection of the kidney implanted on the previous day.
The Unit clerk, in entering the order for the treatment into the computer for transmission to the Radiation Therapy Department for scheduling pur-poses, ordered the treatment for the wrong patient through careless use of the computer light pen.
The wrong patient, who was also a kidney transplant recipient, was brought to the radiation therapy department on the morning of February 7.
A radiation therapy physician checked her chart, noted that there was no order in the chart for radia-tion therapy, but, contrary to hospital policy, directed the technologist to administer the treatment, since the computer schedule showed this patient's name.
The mistake was discovered that afternoon and the correct patient was subsequently treated.
The consequence of this incident was that the patient received 150 rads to the abdomen contrary to the wishes of her physician.
It should be noted, however, that her physician stated later that if in the future she showed signs of rejec-tion of the kidney that had just been implanted, he would prescribe a similar course of radiation therapy.
It should also be noted that some physicians who perform renal implants routinely prescribe radiation therapy without waiting for evidence of rejection.
The licensee's medical staff has concluded that the patient should experience no clinical complications.
Cause or Causes - The cause of the event was the failure of the radiation therapy physician to follow proper procedure.
The physician should have investigated why a patient presented for radiation therapy did not have an order for such therapy written in her chart.
Actions Taken to Prevent Recurrence Licensee - The licensee voluntarily suspended patient treatment pending the results of an internal investigation, and discussion of these results with NRC Region I.
Subsequently, the licensee committed to assure that an authorized physician reviews every patient chart prior to initiation of treatment and confirms that treatment has been requested and is appropriate, and to require consultation between an authorized user and the referring physician prior to the initiation of treatment of any patient.
NRC - The licensee was inspected by an NRC Region I inspector on February 10-11, T986.
The subject event was reviewed in detail.
On February 11, 1986, Region I issued a Confirmatory Action Letter documenting the licensee's commitment described above.
The incident was reviewed by an NRC medical consultant.
A Confirmatory Order Modifying License was issued on May 29, 1986 (Ref. 24).
The Order required that an authorized physician user review every teletherapy patient chart to confirm that cobalt-60 teletherapy treatment has been requested and that 17 I
the authorized physician user consult with the referring physician or the Chief Resident prior to the initial treatment of each teletherapy patient.
In their response to the Order, Washington Hospital Center confirmed that the required procedures had been in place since February 18, 1986.
The May 29, 1986 NRC letter (Ref. 24) also forwarded a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $5,000.
Unless new significant information becomes available, this item is considered closed for the purposes of this report.
- A 86-5 Overexposure to a Member of the Public from an Industrial Gauge The following information pertaining to this event is also being reported con-currently in the Federal Register.
Appendix A (see Example 2 of "For All Licensees") of this report notes that an exposure to an individual in an unre-stricted area such that the whole-body dose received exceeds 0.5 rem in one calendar year can be considered an abnormal occurrence.
Date and Place - On February 19, 1986, while checking a licensee which had ap-parently ceased operations, an NRC Region III inspector determined that an in-dustrial gauge, containing a sealed source of cobalt-60, was in an unrestricted area of the former factory site.
Subsequent inspection determined that at least two members of the public received exposures to radiation as a result of the improper disposal of the gauge.
The gauge had been licensed to C-E Glass, Inc.,
a Division of Combustion Engineering, Inc.
The company operated a facility in St. Louis, Missouri, until October 1981.
Nature and Probable Consequences - C-E Glass, Inc., was licensed in 1971 for the use of a level measurement gauge containing 2.5 curies of cobalt-60.
The source was replaced in June 1978.
In October 1981, the facility and equipment at C-E Glass's site was transferred to Hordis Brothers, Inc., which continued operations until May 1982.
C-E Glass violated two NRC regulations - (1) transferring the gauge to an unauthorized organization (Hordis Brothers did not have an NRC license) and (2) failing to notify the NRC that it had ceased all operations at the St. Louis facility.
The facility and equipment were later sold by Hordis Brothers to a salvage com-pany.
The gauge was placed near a scrap pile at the site, and a salvage company employee removed the gauge's shutter control in early December 1984.
For the next two months, two employees of the salvage company handled the gauge and worked near it.
It was later moved to a scrap pile where access by other individuals was limited.
The gauge was then located by the NRC inspector, assisted by salvage company employees, on February 19, 1986, and later removed from the site by representa-tives of Combustion Engineering Company, who took it to another Combustion Engineering facility for storage and eventual disposal.
18
\\
Interviews with the two salvage company employees determined that they frequently worked or took breaks in the vicinity of the gauge.
Calculations based on the radiation level--with the shutter of the gauge open--concluded that one individual would have received a radiation exposure to his buttocks of 0.6 to 1.7 rem and to his leg of 69 to 208 rem.
(A rem is a standard measure of radiation exposure.)
NRC regulations do not permit radiation exposures to members of the public from licensed activities to exceed 0.5 rem.
The second individual would have received a significantly lower radiation dose.
The first individual has been examined by a physician and his blood count, bone marrow, and physical condition were reported to be normal.
Cause or Causes - The uncontrolled use of the gauge and radiation exposure of at least two individuals were caused by the transfer of the gauge by the licensee to an unauthorized organization.
There was therefore no control over access to the gauge, and a salvage company employee removed the shutter control, allowing the shutter of the gauge to open.
(Had the shutter of the gauge remained closed, the radiation dose to persons in the area would be substantially less than with an open shutter.)
Actions Taken to Prevent Recurrence Licensee - The licensee is no longer in business and has no other gauges in its possession.
NRC - An NRC inspector located the gauge and locked the shutter in its closed position.
He then arranged for the licensee's corporate organization to remove the gauge to another site for storage and eventual disposal.
NRC inspectors surveyed the former C-E Glass site to make certain there were no other gauges there.
A medical consultant was retained to review the circumstances of the case and to provide assistance to the exposed individuals' physicians.
On June 30, 1986, the NRC forwarded to the licensee a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $15,000, for violations associated with the handling of the gauge (Ref. 25).
On May 5, 1986, the NRC issued Inspection and Enforcement Information Notice No. 86-31 to all NRC licensees authorized to possess and use industrial nuclear gauges to inform them of this event (Ref. 26).
Further information was provided to these licensees on July 14, 1986 by Supplement 1 to the Information Notice (Ref. 27).
This incident is considered closed for the purposes of this report.
86-6 Breakdown of Management Controls at an Irradiator Facility The following information pertaining to this event is also being reported con-currently in the Federal Register.
Appendix A (see Example 11 of "For All Licensees") of this report notes that serious deficiency in management or pro-cedural controls in major areas can be considered an abnormal occurrence.
19
i Date and Place - On March 3, 1986, the NRC issued an Order Suspending License (Effective Immediately) to Radiation Technology, Incorporated (RTI) of Rockaway, New Jersey (Ref. 28).
The Order was based on NRC inspections which identified a number of instances of bypassing safety interlock systems; these indicated a significant breakdown in the licensee's management control system.
5 Nature and Probable Consequences - RTI has been licensed to operate a large irradiator near Rockaway, New Jersey, since November 1970.
The licensee's irradiator uses sealed cobalt-60 sources to produce high intensity gamma ray fields for the sterilization of medical equipment and supplies and for various other industrial and scientific applications.
In addition, the licensee has long sought FDA approval to irradiate food products for routine consumption.
At the time of the March 3, 1986 NRC Order, the President of the company was also j
the Chairman of the Board of Directors, and the Radiation Safety Officer.
RTI also owns and operates irradiators through wholly owned subsidiaries in North l
Carolina and Arkansas, both Agreement States.
Another wholly owned subsidiary,
{
South Jersey Process Technology, recently built in Salem, New Jersey, was licensed by Region I on March 14, 1986.
RTI has been the subject of several escalated enforcement actions in the past; the most noteworthy was in 1977 when a plant worker at the Rockaway facility 1
I was able to walk into the irradiation room while the cobalt-60 was exposed be-l cause safety interlocks on the personnel access door, designed to prevent such entry, had been made inoperable.
The employee received a radiation dose of i
j 150-300 rem,.far in excess of regulatory limits.
The license was temporarily suspended following this incident until the licensee took necessary corrective a
actions.
(This incident was reported as abnormal occurrence No. 77-10 in NUREG-0090-10, " Report to Congress on Abnormal Occurrences:
October-December 1977.")
The events giving rise to the most recent Suspension Order first came to light i
during a routine NRC inspection in September 1984.
The inspector discovered that the licensee had been operating the irradiator since April 1984 with an inoperable safety interlock on one of the two conveyor openings used to transfer product into the irradiation room.
On September 26, 1984, Region I issued a Confirmatory l
Action Letter that documented the licensee's commitment to operate the facility only if all safety interlocks were operable and to cease operations if any safety interlock failed to function as required.
Review of relevant documentation by the inspector indicated that this bypassing of interlocks was implemented by the operators under the supervision of the Operations Manager.
In November 1985, the i
interlock was replaced with a new design without required NRC approval.
4 I
During a recent inspection on February 26, 1986, the staff determined that the j
licensee had been operating the facility for several days prior to the inspection j
in spite of the malfunction of a radiation monitor which actuates the lock that
=
assures that the personnel door to the irradiation room cannot be opened while i
the sources are exposed.
Rather than fix the monitor prior to continued opera-tion, as is required by the license, the licensee chose to operate the irradiator i
and, when necessary, opened the door by improperly tripping the door lock when i
the cobalt-60 appeared to have returned to its shielded position.
Following this i
discovery, the staff requested that the licensee cease all operations until the monitor was repaired; conducted daily inspections to assure the facility was being operated safely and that all interlocks were functioning; and prepared the 20 i
4
previously mentioned Order Suspending the License which was issued on March 3, 1986 (Ref. 28).
Subsequently, the licensee requested lifting of the suspension by letters to the NRC dated March 4 and 5, 1986.
After the Region I staff met with the li-censee on March 6, a more complete submission was provided by the licensee on March 10.
This latter submission proposed interim plant operations under the surveillance of an independent Third Party, reporting directly to a member of the RTI Board of Directors, who, along with the licensee, would be responsible for assuring that the facility would be operated safely and in compliance with all NRC requirements.
Further, an independent Fourth Party would monitor the activities of the Third Party on a weekly basis.
Both parties would provide uncensored reports directly to the NRC.
Following consideration of the proposal and agreement of the licensee to additional items, the staff concluded that tem-porary resumption of facility operations under these conditions would not endanger the health and safety of the public.
Accordingly, a Conditional Rescinding of the Order Suspending License was issued on March 13, 1986 (Ref. 29).
The licensee agreed to the terms of this Order in a letter dated March 13, 1986.
Cause or Causes - The root cause can be attributed to a serious breakdown in the licensee's management controls.
Actions Taken to Prevent Recurrence Licensee - The actions taken by the licensee are described above.
NRC - The NRC is continuing to inspect the performance of this licensee at frequent intervals.
A recent license amendment appointed an individual, who joined the company in March 1986, as the new Radiation Safety Officer.
The individual who formerly held this position no longer has direct contact with, or responsibility for, this function.
At a recent meeting of the Board of Directors, this same individual resigned as President, but remains Chairman of the Board.
The responsibilities of President are being shared among three Vice Presidents while a new President is sought.
In addition to the previously described actions taken by the NRC, on June 23, 1986 the NRC suspended the license again based on investigative findings indicating repeated and intentional violations of NRC requirements and impeding NRC inspec-tion and investigations (Ref. 30).
The license is presently suspended pending further action by the NRC.
South Jersey Process Technology, a subsidiary, has recently begun commercial operation of a more modern in-air irradiator in Salem, New Jersey.
The licensee is attempting to build another irradiator in the Port of Elizabeth, New Jersey.
No formal application has been received by the NRC for this facility.
Future reports will be made as appropri' ate.
86-7 Tritium Overexposure and Laboratory Contamination The following information pertaining to this event is also being reported con-currently in the Federal Register.
Appendix A (see Example 11 of "For All 21
Licensees") of this report notes that serious deficiency in management or procedural controls in major areas can be considered an abnormal occurrence.
Date and Place - During a routine inspection on March 12, 1986 at Ferris State College, Big Rapids, Michigan, an NRC inspector determined that, based on a review of bioassay test results, a licensee researcher had received an over-exposure to tritium (hydrogen 3) during experiments on August 3, 1985, equivalent to a whole body exposure of about 21 rems.
Continuing NRC inspections showed that two laboratories were contaminated.
In addition, numerous deficiencies in the licensee's use and control of byproduct radioactive material were identified.
Nature and Probable Consequences - Ferris State College had a broad scope license from the NRC for the possession and use of byproduct radioactive material for training and research purposes.
The NRC inspector found that on August 3, 1985 a researcher was performing work in a ventilated glove box using 5 curies of tritium in a laboratory at the licensee's facility.
(A glove box is a sealed box with viewing windows and gloves affixed to the box, allowing hazardous materials to be normally handled safely.) After completing the work, the researcher performed a urine bioassay test, which showed a tritium level of 520,000 counts per minute per 0.2 milli-liters of urine.
This level of tritium would indicate an internal uptake of 10,000 MPC hours, compared to the NRC quarterly limit of 520 MPC hours for occupational radiation exposures.
(An MPC hour is equivalent to one hour of exposure to the maximum permissible concentration of a specific radioactive material.) This intake would be equivalent to a whole body exposure of 21 rem.
A second bioassay test, 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> later, showed an internal intake of 1,210 MPC hours.
A radiation exposure of 21 rems (a rem is a standard measure of radiation exposure) would not normally be expected to produce any medically observable effects.
Bioassay test results following another 5-curie experiment on December 1, 1985, showed a level of 239 MPC hours.
This exposure was within the NRC limit, but was required to be reported to the NRC because of the urine concentration.
Surveys by the NRC--and subsequently by the licensee--showed the laboratory to be contaminated.
A second laboratory on a different level of the same building was also found to be contaminated.
Surveys by the licensee and the NRC did not iden-tify any contamination in the hallways or other public areas of the building.
Cause or Caust- - The tritium overexposure appeared to result from the failure of the researcher to properly seal off the glove box in which the tritium was being used.
The glove box was pressurized with nitrogen gas which apparently forced the tritium gas through a blower fan into the laboratory rather than through the glove box vent system.
The discharge of the tritium into the laboratory caused both the exposure to the researcher and the contamination of the laboratory.
A research assistant also received some exposure as a result of the experiment or the laboratory contamination, but this exposure was within the NRC limits, according to bioassay test results.
22 i
NRC inspections identified numerous violations of NRC requirements (as discussed below), some of which may have contributed to the overexposure and laboratory contaminations.
Actions Taken to Prevent Recurrence Licensee - After being notified of the initial NRC inspection findings, the licensee removed the researcher from any work involving radioactive material, restricted access to the laboratory areas, and began decontamination of the laboratory facility.
Decontamination was subsequently completed, and the facility was released for normal use.
The licensee also provided information on the contamination to students or other persons who may have used the building where the laboratories are located and offered to provide bioassay testing for any concerned individuals.
No one requested the testing.
Additional actions may be necessary in response to the numerous violations identified during the NRC inspections.
NRC - The NRC issued Confirmatory Action Letters to the licensee on March 19 and 21, 1986, documenting the licensee's agreement to remove the researcher from work involving radioactive materials, to restrict access to the laboratory areas, to undertake decontamination of the facility, and to stop all licensed activities except those associated with the nuclear medicine school.
NRC inspectors inspected the facility on several occasions to gather additional information on the licensee's handling of radioactive materials and to monitor the decontamination efforts.
Confirmatory radiation surveys were also performed.
NRC inspections, which began March 12, 1986 and continued through April 17, 1986, also identified a total of 20 violations of NRC requirements.
These violations included failure to perform required surveys for radioactive contamination, failure to check the glove box ventilation system for proper operation, failure to perform required bioassay tests in some instances, failure to take required follow up actions when certain bioassay results are obtained, failure to report the overexposure, and failure to restrict access to the laboratory.
On April 28, 1986, the licensee's NRC license was amended, significantly re-strict'ng the scope of the authorized activities and providing that any new activitiee must be reviewed and approved by the NRC.
Enforcement action is pending on the violations identified during the NRC inspections.
Further reports will be made as appropriate.
A A A A A A A A AGREEMENT STATE LICENSEES Procedures have been developed for the Agreement States to screen unscheduled incidents or events using the same criteria as the NRC (See Appendix A) and 23
report the events to the NRC for inclusion in this report.
During the first calendar quarter of 1986, the Agreement States reported the following abnormal occurrences to the NRC.
AS86-1 Radiation Injury of an Industrial Radiographer Appendix A (see Example 1 of "For All Licensees") of this report notes that exposure of the whole body of any individual to 25 rems or more of radiation, or exposure of the extremities of any individual to 375 rems or more of radia-tion, can be considered an abnormal occurrence.
Date and Place - On April 20, 1984, an individual employed by BF Inspection Services in Midland, Texas, received an exposure that resulted in a radiation burn while performing radiography in Seminole, Texas.
The licensee failed to notify the Texas Bureau of Radiation Control (Agency) of the incident after it knew the radiographer had received a radiation burn.
The burn was reported by Permian Industrial X-ray, present employer of the individual, to the Agency on November 8, 1985, when the radiographer had an apparent recurrence of the wound.
Nature and Probable Consequences - The radiographer and assistant arrived at the job site at 9:00 a.m. and used a 90 curie iridium-192 source to x-ray pipe in a pipe rack until 11:00 a.m..
The job was being performed by Conam Inspection, and BF Inspection had been subcontracted to assist on the job.
The radiographer worked on the pipe rack removing exposed film and placing new film on the weld to be x-rayed.
The assistant worked the crank out.
The radiographer did not have a survey instrument with him.
He depended on the assistant to tell him when the radiation levels decreased on the survey meter kept at the crank out.
At approximately 10:00 a.m., the radiographer took the exposed film to the cen-tral processing facility. When this film was developed, it was darker than it should have been and appeared to be exposed longer than necessary.
At 11:00 a.m., the radiographer decided to stop work and go to lunch.
At this time he locked the radiographic device, removed the guide tube and put the dust cover on the front of the device.
No survey was performed on the radiographic device or guide tube at this time.
The equipment was placed in a bucket used to transport equipment on the pipe rack and left there during lunch.
Prior to leaving the job site, the radiographer went to the job supervisor.
The radiographer was informed that it appeared that he needed to decrease his exposure time and was asked for any other exposed film.
The radiographer told the supervisor that he would bring the remaining exposed film for developing after lunch.
When the crew returned from lunch, the radiographer decided to quit for the day, since he could not determine the cause of the film being overexposed.
The equip-ment was moved from the bucket to the back of the truck.
The radiographer carried the guide tube and crankout in his left hand and the radiographic device in his right hand.
When the radiographer disconnected the crankout from the radiographic device, he discovered that the source was not connected to the drive cable.
The radiographer then looked at his pocket dosimeter and found that it was discharged beyond its range.
He then asked the assistant to look at his dosimeter and was 24
informed that it was not discharged beyond its range.
The assistant radiographer was instructed to check the radiation level and he told the radiographer that he had a reading of 90.
(The radiographer and the assistant did not remember what scale the survey meter was set on.)
The radiographer shook the guide tube and heard something rattling.
He carried the guide tube, still rolled up, and the radiographic device to a large concrete slab approximately 60 feet from the truck.
The guide tube was unrolled and when he shook it, the source fell out.
The radiographic device was placed on top of the source and the radiographer went back to the truck.
He then approached the I
device with the crankout and passed the drive cable through the radiographic device.
The device was then moved behind the connector end of the source pigtail and the pigtail and the drive cable were connected.
The source was returned to its shielded position and locked in place.
When the crew left the job site, the radiographer notified the job supervisor of the disconnect.
The supervisor instructed the radiographer that he should notify his radiation safety officer (RS0) of the disconnect and to leave his film for processing.
The supervisor stated that when this film was developed, it appeared that it may have been fogged.
When the radiographer returned to the office, he found that the RSO was gone for the weekend and did not attempt to notify him of the disconnect until Monday, April 22.
On Monday, the RSO and the radiographer inspected the equipment.
According to the radiographer, nothing was found to be wrong with the equipment.
The radiographer stated that he turned in his and his assistant's film badges for immediate processing.
At this time, the radiographer did not demonstrate any symptoms of a radiation injury.
Approximately five to seven days after the disconne t,
the thumb, index and middle fingers of both the radiographer's hands bet.sme red and swollen.
The radiographer was seen by a doctor and the three blw d tests performed were within normal limits.
The medical expenses were paid for by BF Inspections.
After a period of approximately two months, the radiographer's hands appeared to heal.
During the first week of November 1985, the radiograpner was working for another company and the middle and index fingers of his left hand became red and swollen.
He again went to see a doctor.
He notified his employer of the injury.
The company RSO then notified the Agency of the injury.
Based on statements by the radiographer, Agency investigators calculated his exposure from carrying the equipment to the truck and recovering the source.
The radiographer may have received up to about 29,000 rems to his left hand and about 47 rems whole body exposure.
During the Agency's investigation, several items of disagreement arose.
While the radiographer stated that he turned in his and his assistant's film badges, the R50 for BF Inspection Service stated that he asked for the badges and was informed that they were in the truck used at the job site.
The RSO instructed the radiographer to give him the film badges but did not follow up when he did not receive the badges.
The question also arose as to whether there was an 25
equipment malfunction.
According to the radiographer, there was no equipment malfunction.
The RSO stated during the investigation that the connector on the j
drive cable was worn and caused the disconnect.
When questioned concerning the cause of the disconnect, the radiographer stated that the only conclusion he could reach is that he did not connect the source pigtail to the drive cable when he assembled the device.
Cause or Causes - The apparent cause of the disconnect is that the source pigtail was not correctly connected to the drive cable when the equipment was set up.
The exposure and subsequent burn resulted when the radiographer did not follow the licensee's Operating Procedures or the Texas Regulations for Control of Radiation, and failed to perform a survey of the radiographic device or guide tube between radiographs, when the equipment was secured for lunch, or at the end of the day.
The radiographer also failed to follow the licensee's Emergency Procedures for a source disconnect.
Actions Taken to Prevent Recurrence Licensee - At this time, the licensee's response to the Agency's compliance letter was not satisfactory as to what actions it has taken to prevent occurrence of this type of accident.
The l'icensce's initial report of the incident did not address calculations of the radiographer's exposure, nor measures taken to prevent a recurrence.
Agency - The Agency has cited the licensee for 14 items of non-compliance with the Texas Regulations for Control of Radiation and is undertaking escalated enforcement.
The investigation of this incident is continuing in an attempt to obtain additional information.
Unless new significant information becomes available, this incident is closed for the purposes of this report.
AS86-2 Contamination of a Scrap Steel Facility Appendix A (see the general criterion) of this report notes that a moderate or more severe impact on the public health or safety can be considered an abnormal occurrence.
1 Date and Place - On May 24, 1985 it was discovered that some facilities of Tamco Steel Company of Ontario, California, were contaminated with radioactive i
material (later determined to be cesium).
Nature aad Probable Consequences - On May 23, 1985 two 20 cubic yard roll boxes being transported to the hazardous waste site at Kettlemen Hills, California, i
set off the radiation alarms at the weigh station at Newhall, California.
The trucks were returned to the originator of the shipment at the direction of the i
California Highway Patrol (CHP) on advice of the California Radiologic Health l
Branch (RHB).
The licensee voluntarily ceased operations upon notice by the CHP of the incident at the weigh station.
The furnace was turned off and 26
operations began to go to a cold shut down.
Shipments of product steel were suspended.
On May 24, 1985 inspectors from the California Department of Occupational Health and Safety conducted an initial survey of the roll boxes and the facilities of Tamco Steel Company.
The survey indicated contamination was limited to flue dust, slag piles, bag house (containing flue dust) and associated ducting, in addition to surfaces of the furnace itself.
Samples were collected and sent for analysis.
Direct radiation levels ranged from 0.03 mr/hr to 15 mr/hr.
Analysis of the samples indicated cesium contamination ranging from 2.0 pci/gr to 4 uci/gr.
On May 25, 1985 Tamco had a contractor on site to begin a thorough survey and develop a clean up plan.
The initial plan for decontamination was developed with the RHB.
A planned incremental decontamination program began on May 30, 1985.
Priority was given to operational equipment and facilities.
Clearance inspections conducted by the State followed the progress of the decontamination effort.
Because of its chemical form the cesium was removed from work areas in the mill through flue dust ducting and the waste slag. Workers in the furnace area and bag house most likely to receive internal contamination by inhalation of airborne dust containing cesium were sent to the University of Californ*t at Los Angeles where detailed examinations were conducted.
The examinations did not detect any contamination of the workers.
On August 1, 1985 the State Compliance Inspection Team completed its final survey.
The RHB issued a departmental letter dated October 8, 1985, which released the facilities and equipment for unrestricted use.
Cause or Causes - The Tamco Steel Company processes scrap steel purchased from various suppliers throughout California, Nevada, and Arizona, into construction rebar.
The scrap is segregated by metal type and sent directly to the melting furnace without inspection.
The device or source of the approximately 1.5 curies of cesium was brought into the scrapyard undetected and sent to the furnace as part of a routine melt.
Scrap metal dealers as a normal practice do not screen for radioactive materill.
Actions Taken to Prevent Recurrence Tamco Steel - Tamco Steel installed low-level radiation monitors at the gate to check scrap steel coming into the facilities and product shipment leaving. They also now physically survey all scrap steel before it is placed in the furnace.
State Agencies - As discussed above, the cognizant State Agencies monitored the decontamination of the facility, the actions taken to prevent recurrence by Tamco Steel, and after a final survey of the facility, released the facilities and equipment for unrestricted use.
This incident is considered closed for the purposes of this report.
27
AS86-3 Radiation Injury of an Industrial Radiographer Appendix A (see Example 1 of "For All Licensees") of this report notes that exposure of the whole body of any individual to 25 rems or more of radiation, or exposure of the extremities of any individual to 375 rems or more of radia-tion, can be considered an abnormal occurrence.
Date and Place - On August 25, 1985 an industrial radiographer received a radia-tion injury of his left hand and a whole body overexposure.
At the time of the incident, the employee (employed by Boothe-Twining, Inc.) was performing radi-ography at the company's field site in the Kern River oil field in Bakersfield, California.
He was using a 46 curie iridium-192 source contained in a radio-graphic projector.
Nature and Probable Consequences - The radiographer (who had four years of radiography experience with Boothe-Twining) encountered great resistance with the source " crank out."
He then approached and manually adjusted the camera to reduce the kink in the guide tube.
During this action his hands grasped the lock box and guide tube connector.
At the completion of this readjustment, he moved away from the camera and observed that his 200 mr pocket dosimeter read off scale.
However, he did not report his dosimeter was off-scale but reported a pocket dosimeter reading of 119 mr to his supervisor.
His film badge was sent in for reading approximately seven days after the accident after symptoms of high dose to the left hand were manifested and reported to management.
He was seen by and continues to be under the care of a physician.
In addition, another doctor is participating as the State Agency's medical consultant.
Based on time and motion studies preliminary estimates indicated a left hand dose of about 2,000 rads and a whole body dose of about 6 rads.
The radiographer was accompanied at the work site by a helper who was acting as an assistant radiographer.
She was only indirectly involved in actual performance of radiography, and registered no excessive dose.
She kept back from the shooting area but had observed most of the radiographer's movements during the accident.
A State Investigative Panel was convened pursuant to an Order dated October 16, 1985, to determine the causes of the radiation accident, establish the nature and extent of radiation exposure and any injury, and to recommend corrective ac-tion to prevent future recurrence of such accidents.
The findings of the Inves-tigative Panel are discussed below.
The radiographer had failed to adhere to established radiation safety and operat-ing procedures.
He did not assure that the radiography source was returned to the safe shielded position with the crank and did not perform.a radiation survey on his approach to the camera; therefore, he had no warning that the source was out.
Management had failed to communicate forcefully its intolerance of deviation i
from established safety procedures, particularly the failure to survey while approaching the radiographic projector.
The Investigative Panel found that such deviation was common practice with the overexposed radiographer, and that management knew of it.
Had the radiographer used his survey instrument as l
28 i
required, he would have detected early on that the source was out, in an unsafe position.
Instruction of radiographers, and specifically the overexposed radiographer, was found to be unacceptable in that, (a) there was failure to convey the crucial safety problem to the employee, i.e.,
that the performance of field radiography constitutes a serious hazard and therefore requires strict adherence to safety procedures, (b) the licensee's attempted requirement that employees attend re-fresher training on their own time is unenforceable and contrary to State labor law -- the license assigns the responsibility for providing refresher training to the licensre, not the employee, (c) semi-annual refresher training consisted of infrequent safety " bull sessions" with no structure and no record other than the fact of the bull session, and (d) radiographers were not checked out on equipment prior to use.
General operating procedures were available to radiographers, how-ever, no step-by-step operating, radiation safety and emergency proccdures were provided that were specific to the make and model of projector used.
Responsibility for the radiation safety program, although vested in the Radia-tion Safety Officer (R50) of the licensee, was in fact abrogated by the president of the company.
In his own testimony to the Investigative Panel, the president accepted this responsibility and authority, but did very little to implement the program and clearly would not delegate to others.
In the current case, he pre-vented the RSO from conducting the investigation.
To compound the situation, the compliance history and the president's testimony clearly illustrated his failure to comprehend the company's direct responsibilities for the employee hand burn and for the numerous overexposures.
He maintained that the accident was the employee's fault.
The company's RSO asserted that the conduct of training and management audits were his safety assignment, but his heavy responsibilities in sales, customer relations and quality assurance often took precedence.
Management audits of the overexposed employee's work as a radiographer were not conducted as required by license condition and records were not maintained.
Records for August 23, 1985, of the inspection and maintenance of the Gamma Century projector involved in the accident were reviewed.
These suggest that the equipment was in good functioning condition, yet the Investigative Panel discovered that the lock could be actuated over the drive cable, thus locking the source outside the shield.
According to the Investigative Panel, the compliance history of Boothe-Twining is unacceptably poor.
The company was found to be at fault and cited, in the serious overexposure / injury of an employee, in a 1981 radiography accident.
Since then repeat and serious violations have continued, necessitating an office compli-ance conference on July 5, 1985.
In spite of agreements arising out of that con-ference, the company continued to fail to provide adequate training and audits of employees so that there is a clear and unambiguous understanding of the serious-ness of the performance of field radiographers and the need therefore to follow required safety procedures.
For example, in July 1985 the radiographer involved in the present incident received a whole body dose of 2 rem.
The report of this did not stimulate an appropriate response by management.
29
Failures of the RSO to conduct job site management audits at frequencies promised and to provide comprehensive refresher training for radiographers were contrary to license conditions.
Refresher training for radiographers is required to include review of, (1) radiation safety and operating procedures, while (2) stressing changes in such procedures and (3) measures to be taken to avoid excessive j
exposures.
Cause or Causes - The immediate cause of the overexposure was the failure of the radiographer to adhere to established radiation safety and operating procedures.
As discussed above, contributing causes are the serious breakdowns in management and procedural controls in the licensee's conduct of radiographic operations.
Actions Taken to Prevent Recurrence Licensee - A Notice of Violation was issued to the licensee by the California Division of Occupational Safety and Health (Agency) on December 11, 1985.
The testimony of company employees including management affirmed that the violations did in fact occur.
The response also outlined corrective action to prevent recur-rence of these violations.
The licensee's response to the matter of management audits was judged to be in-adequate.
The licensee was provided additional opportunity to develop an internal audit program to assure that radiographers and radiographers' assistants comply with the State Department of Health Services' regulations and license conditions, and the company's operating and emergency procedures.
Agency - The State held an enforcement conference with the licensee. A consent agreement will be signed between the Director of the State Department of Health Services and the licensee.
The licensee will be placed on a 3 year probation with provisions for suspension if serious noncompliance occurs within that period.
The license will be amended to require a full time RSO and will detail the R50's duties.
The State Investigative Panel concluded that if the radiographer had been wear-ing a functional pocket radiation alarm, the radiographer would have had ample warning that the source was not in its proper shielded position.
The Panel further agreed that the introduction of pocket radiation alarms into the practice of industrial radiographer is now imperative.
Reliability had improved as a result of demand by the nuclear power industry so that the pocket alarm can reasonably be expected to withstand field service, provided radiographers are given appropriate instruction in the use of these devices.
Instruction will also be necessary to prevent use of the radiation alarm as a substitute for quantita-tive assessment by radiation survey meters of radiation fields in conducting radi-ation surveys.
The introduction of pocket radiation alarms is expected to reduce the frequency of excessive exposures and minimize the incident of injuries by giving radiographers timely warning of exposed sources.
California will consider adopting regulations which would require use of appropri-ate pocket radiation alarms for all radiographers and radiographers' assistants.
This requirement would supplement.and not in any way displace the present require-ment for use of a survey meter in conducting required radiation protection surveys for industrial radiography.
30 t
=
California will consider promulgating regulatory requirements and otherwise encouraging the development of a radiographic projector with an integral warning system built into the device to indicate in unambiguous fashion the safe, inter-mediate or unsafe position of the source.
This may be done by announcing proposed legislative requirements to authorize only devices with this feature, starting in 1990.
Unless new significant information becomes available, this incident is considered closed for the purposes of this report.
AS86-4 Radiation Injury of an Industrial Assistant Radiographer Appendix A (see Example 1 of "For All Licensees") of this report notes that exposure of the whole body of any individual to 25 rems or more of radiation, or exposure of the extremities of any individual to 375 rems or more of radiation, can be considered an abnormal occurrence.
Date and Place - On November 9, 1985 an individual employed as an assistant radiographer by Basin Industrial X-Ray in Odessa, Texas, received a radiation burn of his left hand an'd an estimated 129 rems whole body exposure.
The licensee f ailed to notify the Texas Bureau of Radiation Control (Agency) of the incident.
Another licensee informed the Agency on November 26, 1985 that an incident had occurred involving Basin Industrial X-Ray.
Nature and Possible Consequences - The radiography crew was performing work at Fabricators Contractors, Inc., during the evening of November 9, 1985.
The assis-tant radiographer was shooting the welds and the radiographer was developing the exposed film.
At approximately 11:30 p.m., the assistant radiographer noticed that his survey meter, placed approximately two feet in front and to the right of the radiographic device (a 76 curie iridium-192 radiography camera), was off scale after the source was supposed to have been returned to its shielded position.
He then checked his pocket dosimeter and found it was discharged beyond its range.
He notified the radiographer, who was unsuccessful in his attempt to return the source to its shield using the crankout.
The radiographer checked his pocket dosimeter and found that it was not discharged beyond its limit.
He then notified the local supervisor, who was acting as the local radiation safety officer.
The radiographer was instructed to isolate the area and wait for the supervisor.
The supervisor did not retrieve the source but instructed the radiographer in the procedures.
The first action was to remove the crankout from the radio-graphic device.
At this time it was found that the drive cable had broken just behind the connection with the source pigtail.
The radiographer then removed the guide tube fr om the device with a pair of 12 inch channel lock pliers.
Using the pliers and holding the guide tube at arms length he carried it to an open area of the shop and shook the source out of the guide tub'e.
Using the pliers, the pigtail was placed in the device source end first.
The pigtail was then reversed using the pliers and pushed in with the dust cap.
When the con-nector end of the pigtail exited the lock box it was locked in place.
After the source retrieval, it was found that the radiographer's pocket dosimete - had been discharged beyond its range of 200 millirems.
31 1
The equipment was loaded in a truck and the three employees proceeded to the licensee's facility.
After arriving at the facility, the supervisor notified the company's radiation consultant of the disconnect and that the employees' pocket dosimeters had been discharged beyond their limits.
The supervisor was instructed to return both badges for immediate processing and to send the assistant radiographer for a blood test.
The blood sample taken at this time was within normal limits.
On November 29, 1985 the assistant radiographer met with an Agency represent-tive.
At this time, the individual's left hand had redness from the wrist to the base of the little finger.
On December 2, 1985 the individual had a blister from the wrist to the base of the little finger on his left hand.
On December 6, 1985 the Agency contacted the radiation safety officer (RS0) and requested the film badge results.
The Agency was informed that the film badge had been sent in for routine processing and the results were not available.
The licensee was instructed to contact its film badge supplier and to have the radiographer's and assistant's film badges immediately processed.
When the assistant's film badge was processed, it indicated an exposure of 129 rems.
When asked about the radiographer's exposure, the RSO stated that he did not have his badge processed.with the assistant's.
The Agency again instructed the RSO to have the radiographer's film badge immediately processed.
The radiog-rapher's exposure was determined to be 28 rems.
Based on statements made by the assistant radiographer and a re-enactment of the incident, Agency investigators calculated the exposure to the assistant to be about 129 rems whole body.
The estimated exposure to his left hand is subject to considerable uncertainty.
The value may have been as high as 30,000 rems, or even considerably higher.
The Agency's investigation found that the :ndividual had not received radiation safety training or formal training in industrial radicgraphy from the licensee.
It also appeared that the individual had falsified his application stating that he had previous experience.
When asked why the licensee did not report the incident to the Agency, the RSO stated that he did not realize the severity of the incident, since he had not been provided the full details by the radiography crew.
The RSO knew that the assistant radiographer's pocket dosimeter was discharged beyond its range but did not return his film badge for immediate processing.
The licensee failed to perform a detailed investigation of the incident when it appeared that there could have been a serious radiation exposure.
The RSO also informed the Agency that he did not know that the drive cable had been broken. When asked by the Agency investigators, the RSO stated that he could not locate the broken crank-out cable.
Cause or Causes - The apparent cause of the exposure and burn appear to be that the licensee permitted an individual to perform the functions of a radiographer without providing the proper safety training, and that the individual failed to perform surveys between radiographs.
32 I
r
Actions Taken to Prevent Recurrence Licensee - The licensee has started tighter controls on its initial training program and hiring procedures.
t Agency - The Agency has cited the Licensee for items of non-compliance with the f
Texas Regulations for Control of Radiation.
In addition, a complaint has been issued to the licensee, notifying him that the Agency intends to revoke the license. The investigation of this incident is continuing in an attempt to obtain additional information.
Unless new significant information becomes available, this incident is considered closed for the purposes of this report.
I l
33 n-R
REFERENCES 1.
Letter from M. O. Medford, Manager of Nuclear Licensing, Southern California Edison Company, to A. E. Chaffee, Chief, Reactor Project Branch, NRC Region V, Docket No. 50-206, April 8, 1986.*
2.
Letter from M. O. Medford, Manager of Nuclear Licensing, Southern California Edison Company, to G. E. Lear, Director, PWR Project Directorate #1, NRC Office of Nuclear Reactor Regulation, Docket No. 50-206, May 1, 1986.*
3.
Confirmatory Action Letter from J. B. Martin, Regional Administrator, NRC Region V, to H. B. Ray, Vice President, Southern California Edison Company, Docket No. 50-206, November 21, 1985.*
4.
U.S. Nuclear Regulatory Commission, " Loss of Power and Water Hammer Event at San Onofre, Unit 1, on November 21, 1985," USNRC Report NUREG-1190, pub-lished January 1986.**
5.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-01, " Failure of Main Feedwater Check Valves Causes Loss of Feedwater System In.tegrity and Water Hammer Damage," January 6, 1986.*
6.
Letter from R. J. Rodriguez, Assistant General Manager, Sacramento Municipal Utility District, to J. B. Martin, Regional Administrator, NRC Region V and F. J. Miraglia Jr., Director, PWR-B Division, NRC Office of Nuclear Reactor Regulation, forwarding a summary report
" Description and Resolution of issues Regarding the December 26, 1985 Reactor Trip," Docket No. 50-312, February 19, 1986.*
7.
Letter from J. B. Martin, Regional Administrator, NRC Region V, to R. J. Rodriguez, Assistant General Manager, Sacramento Municipal Utility District, requesting that a root cause analysis be completed before return to power, Docket No. 50-312, December 26, 1985.*
8.
Letter from J. B. Martin, Regional Administrator, NRC Region V, to R. J. Rodriguez, Assistant General Manager, Sacramento Municipal Utility District, requesting the licensee to hold in abeyance any repair work planned on equipment that malfunctioned, Docket No. 50-312, December 26, 1985.*
9.
U.S. Nuclear Regulatory Commission, " Loss of Integrated Control System Power and Overcooling Transient at Rancho Seco on December 26, 1985,"
USNRC Report NUREG-1195, published February 1986.**
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, fo. inspection and copying (for a fee).
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection.
Available for purchase from the GP0 Sales Program, Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7982.
35 n
REFERENCES (Continued) 10.
Letter from Victor Stello, Jr., NRC Acting Executive Director for Opera-tions, to Hal Tucker, Chairman, Babcock & Wilcox Owners Group, requesting an evaluation of design of B&W plants for reduction of plant trips and miti-gating transient response, January 24, 1986.*
11.
Letter from Dennis M. Crutchfield, Assistant Director for Technical Support, Division of PWR Licensing-8, NRC Office of Nuclear Reactor Regulation, to Hal Tucker, Chairman, Babcock & Wilcox Owners Group, March 13, 1986.*
12.
Memorandum from Victor Stello, Jr., NRC Acting Executive Director for Orera-tions, to the Commissioners, entitled "B&W Design Reassessment," March 21, 1986.*
13.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-04, " Transient Due to Loss of Power to Integrated Control System at a Pressurized Water Reactor Designed oy Babcock & Wilcox,"
January 31, 1986.*
14.
U.S Nuclear Regulatory Commission, " Rupture of Model 48Y UFe Cylinder and Release of Uranium Hexafluoride," USNRC Report NUREG-1179, Vol.1, published February 1986.**
15.
U.S. Nuclear Regulatory Commission, " Assessment of the Public Health Impact from the Accidental Release of UFs at the Sequoyah Fuels Corporation at Gore, Oklahoma," USNRC Report NUREG-1189, published March 1986.**
l 16.
U.S. Nuclear Regulatory Commission, " Rupture of Model 48Y UFs Cylinder and Release of Uranium Hexafluoride," USNRC Report NUREG-1179, Vol. 2, published June 1986.**
17.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-28, " Telephone Numbers to the NRC Operations Center and Regional Offices," April 24, 1986.*
18.
Letter from Jack A. Hind, Director, Division of Radiation Safety and Safeguards, NRC Region III, to A. D. Riley, Plant Manager, Allied Chemical Company, forwarding Inspection Report No. 40-3392/86001 (DRSS), Docket No. 40-3392, April 4, 1986.*
19.
Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to A. J. Cipolla, Plant Manager, Allied Chemical Company, I
Docket No. 40-3392, January 10, 1986.*
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection.
Available for purchase from the GPO Sales Program, Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7982.
36 i
__fL_________
l
.___A--
REFERENCES (Continued) 20.
Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to A. D. Riley, Plant Manager, Allied Chemical Company, Docket No. 40-3392, March 24, 1986.*
r 21.
Letter from Jack A. Hind, Director, Division of Radation Safety and Safe-guards, NRC Region III, to J. C. Bishop, Acting Plant Manager and General Manager UFs/ Fluoride Programs, Allied Chemical Company, forwarding a Notice of Violation and Inspection Report No. 40-3392/86003 (DRSS), Docket No. 40-3392, April 25, 1986.*
22.
Letter from W. D. Shafer, Chief, Emergency Preparedness and Radiological Protection Branch, NRC Region III, to A. D. Riley, Plant Manager, Allied Chemical Company, forwarding Inspection Report No. 40-3392/86002 (DRSS),
Docket No. 40-3392, April 21, 1986.*
23.
Letter from James G. Keppler, Regional Administrator, NRC Region III, to L. L. Taunton, Vice President, Operations, Engineered Materials Sector, Allied-Signal Corporation, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, Docket No. 40-3392, June 27, 1986.*
24.
Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to Jay Shriver, Associate Administrator for Clinical Affairs, Washington Hospital Center, forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, also forwarding a Confirmatory Order Modifying License, Docket Nos. 30-09588 and 30-01325, May 29, 1586.*
25.
Letter from James G. Keppler, Regional Administrator, NRC Region III, to Reynold L. Hoover, Corporate Director, Health, Safety, and Environmental C.atrol, Combustion Engineering Company, forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, Docket No. 030-05165, June 30, 1966.*
26.
'J.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-31, " Unauthorized Transfer and Loss of Control of Industrial Nuclear Gauges," May 5, 1986.*
27.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-31, Supplement 1, " Unauthorized Transfer and Loss of Control of Industrial Nuclear Gauges," July 14, 1986.*
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection.
Available for purchase from the GP0 Sales Program, Superintendent of Documents, U.S. Government Printing Of fice, Post Of fice Box 37082, Washington, DC 20013-7982.
37 i
a
4 REFERENCES (Continued) 28.
Letter from James M. Taylor, Director, NRC Office of Inspection and Enforce-ment, to Martin A. Welt, Ph.D., President, Radiation Technology, Inc., for-warding an Order Suspending the License (Effective Immediately), Docket No. 30-07022, March 3, 1986.*
29.
Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement to Martin A. Welt, Ph.D., President, Radiation Technology, Inc.,
forwarding a Conditional Rescission of Order Suspending License, Docket No. 30-07022, March 13, 1986.*
30.
Letter from James M. Taylor, Director, NRC Office of Inspection and Enforce-ment, to Radiation Technology, Incorporated executives (Dr. Robert Cockrell, Vice President, Operations and Engineering; Mr. George Sadek, Vice Presi-dent, Finance and Administration; and Dr. Arnold Orlander, Vice President, Sales and Marketing), forwarding an Order Suspending Licenses (Effective Immediately), Docket Nos. 30-07022 and 30-19146, June 23, 1986.*
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection.
Available for purchase from the GP0 Sales Program, Superin-dendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7982.
j 38 l
t D
yu___
APPENDIX A ABNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormal occurrence determinations were set forth in an NRC policy statement published in the Federal Register on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952).
An event will be considered an abnormal occurrence if it involves a major reduc-tion in the degree of protection of the public health or safety.
Such an event would involve a moderate or more severe impact on the public health or safety and could include but need not be limited to:
1.
Moderate exposure to, or release of, radioactive material licensed by or otherwise regulated by the Commission; 2.
Major degradation of essential safety-related equipment; or 3.
Major deficiencies in design, construction, use of, or management controls for licensed facilities or material.
Examples of the types of events that are evaluated in aetail using these criteria are:
For All Licensees 1.
Exposure of the whole body of any individual to 25 rems or more of radia-tion; exposure of the skin of the whole body of any individual to 150 rems or more of radiation; or exposure of the feet, ankles, hands or forearms of any individual to 375 rems or more of radiation (10 CFR $20.403(a)(1)), or equivalent exposures from internal sources.
2.
An exposure to an individual in an unrestricted area such that the whole-body dose received exceeds 0.5 rem in one calendar year (10 CFR 620.105(a)).
3.
The release of radioactive material to an unrestricted area in concentra-tions which, if averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 500 times the regulatory limit of Appendix B, Table II, 10 CFR Part 20 (10 CFR S20.403(b)).
4.
Radiation or contamination levels in excess of design values on packages, or loss of confinement of radioactive material such as (a) a radiation dose rate of 1,000 mrem per hour three feet from the surface of a package con-taining the radioactive material, or (b) release of radioactive material from a package in amounts greater than the regulatory limit.
5.
Any loss of licensed material in such quantities and under such circum-stances that substantial hazard may result to persons in unrestricted areas.
6.
A substantiated case of actual or attempted theft or diversion of licensed material or sabotage of a facility.
39-
7.
Any substantiated loss of special nuclear material or any substantiated in-ventory discrepancy which is judged to be significant relative to normally expected performance and which is judged to be caused by theft or diversion or by substantial breakdown of the accountability system.
8.
Any substantial breakdown of physical security or material control (i.e.,
access control, containment, or accountability systems) that significantly weakened the protection against theft, diversion or sabotage.
9.
An accidental criticality (10 CFR S70.52(a)).
10.
A major deficiency in design, construction or operation having safety impli-cations requiring immediate remedial action.
11.
Serious deficiency in management or procedural controls in major areas.
12.
Series of events (where individual events are not of major importance), re-curring incidents, and incidents with implications for similar facilities (generic incidents), which create major safety concern.
For Commercial Nuclear Power Plants 1.
Exceeding a safety limit of license technical specifications (10 CFR 650.36(c)).
2.
Major degradation of fuel integrity, primary coolant pressure boundary, or primary containment boundary.
3.
Loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emer-gency core cooling system, loss of control rod system).
4.
Discovery of a major condition not specifically considered in the safety analysis report (SAR) or technical specifications that requires immediate remedial action.
5.
Personnel error or procedural deficiencies which result in loss of plant capability to perform essential safety functions such that a potential re-lease of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cool-ing system, loss of control rod system).
For Fuel Cycle Licensees 1.
A safety limit of license technical specifications is exceeded and a plant shutdown is required (10 CFR 550.36(c)).
2.
A major condition not specifically considered in the safety analysis report or technical specifications that requires immediate remedial action.
3.
An event which seriously compromised the ability of a confinement system to perform its designated function.
40
APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the January through March 1986 period, the NRC, NRC licensees, Agreement States, Agreement State Licensees, and other involved parties, such as reactor vendors and architects and engineers, continued with the implementation of ac-tions necessary to prevent recurrence of previously reported abnormal occur-rences. The referenced Congressional abnormal occurrence reports below provide the initial and any updating information on the abnormal occurrences discussed.
Those occurrences not now considered closed will be discussed in subsequdnt re-ports in the series.
NUCLEAR POWER PLANTS 79-3 Nuclear Accident at Three Mile Island This abnormal occurrence was originally reported in NUREG-0090, Vol. 2, No. 1,
" Report to Congress on Abnormal Occurrences:
January-March 1979," and updated in each subsequent report in this series, i.e., NUREG-0090, Vol. 2, No. 2 through Vol. 8, No. 4.
It is further updated as follows.
Reactor Building Entries During the first calendar quarter of 1986, 82 entries were made into the TMI-2 reactor building, bringing the total number of entries since the March 1979 acci-dent to 865.
Reactor building activities during this period centered around the ongoing defueling operations, as discussed below.
Additional reactor building activities included collecting pressurizer sludge samples and performing examina-tions of the steam generator upper and lower head spaces.
Reactor Vessel Defueling Operations Substantial progress was made during the quarter in removing fuel and structural debris from the damaged TMI-2 core.
As of April 20, 1986, a total of 44 defueling canisters had been loaded with fuel debris and transferred to shielded storage locations in the "A" Spent Fuel Pool in the Fuel Handling Building.
The total weight of the debris removed from the reactor vessel thus far is 50,170 lbs, rep-resenting 16% of the 308,000 lbs of debris estimated to be in the vessel. Visi-t bility in the reactor vessel has been progressively decreasing due to the rapid growth of microorganisms in the reactor coolant.
These biological growths, which include algae, fungi, bacteria, and aerobic and anaerobic oroganisms, were ini-tially identified as the cause of filter plugging in the Defueling Water Cleanup System, which has been ineffective in maintaining water clarity.
The licensee has reached a point at which further defueling operations require improved visibility; therefore the concentration of these microorganisms in the reactor vessel water must be substantially reduced.
Numerous biocides were tested or considered as a means to kill the growths but were found to be inef-fective or unacceptable for use.
Raising the water temperature to 180 F also proved unsuccessful in tests.
Based on test results, the licensee plans to kill the organisms by exposing them to high pressure.
Manual brushing and hydrolaz-ing will initially be performed to remove growths from reactor vessel surfaces.
41
Reactor coolant will then be circulated through a high pressure positive dis-placement pump, where the rapid pressurization and depressurization should de-stroy most of the organisms.
The pump effluent will be circulated through a diatomaceous earth filter to remove the resulting debris.
The operation of this system is expected to take one week and, if successful, may be used periodically to maintain water clarity at an acceptable level for the remainder of defueling.
Dose rates associated with defueling activities have remained low.
Dose rates on the defueling work platform average approximately 8 mrem /hr, and the highest mea-sured dose rates during canister transfer from the reactor vessel to the storage racks have been less than 40 mrem /hr.
The licensee has been using four defueling crews a day to increase the actual defueling time to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> a day, a 200% im-provement since the beginning of defueling activities in October 1985.
EPICOR-II/ Submerged Demineralizer System (SDS) Processing Approximately 362,462 gallons of water were processed through the SDS during the reporting period.
Approximately 77,835 gallons were processed through the EPICOR-II system during the quarter.
Liner Shipment One depleted, dewatered EPICOR-II resin liner was sent to Richland, WA during the quarter.
TMI-2 Advisory Panel Meeting The Advisory Panel for the Decontamination of Three Mile Island Unit 2 met on February 12, 1986 in Harrisburg, Pennsylvania.
The Panel was briefed by General Public Utilities Nuclear Corporation (GPUN) on the progress of the TMI-2 cleanup and on the licensee's actions to correct deficiencies in the radiochemistry pro-gram.
The Panel was also briefed on the status of current NRC activities.
Rep-resentatives from the U.S. Department of Energy (00E) described plans for off-site shipment and storage of the damaged fuel and debris removed from the reactor vessel.
The material will be transported by rail from the TMI site to a DOE interim storage facility in Idaho.
Ms. Jane Lee, a local citizen, addressed the Panel on health effects issues in the vicinity of TMI.
She criticized the recent health effects study issued by the Pennsylvania State Department of Health.
The Commission has recently appointed two new members to the Advisory Panel, to fill the vacancies created by the resignations of Mayor Robert Reid of Middletown, PA and Dr. Thomas Cochran.
The new Panel members are:
Mr. Frederick S. Rice, Chairman of the Dauphin County Board of Commissioners, and Dr. John Luetzelschwab, Professor of Physics at Dickinson College in Carlisle, PA.
Future reports will be made as appropriate.
85-7 Loss of Main and Auxiliary Feedwater Systems This abnormal occurrence, which occurred at Davis-Besse on June 9, 1985, was originally reported in NUREG-0090, Vol. 8, No. 2, " Report to Congress on Abnormal 42 l
Occurrences:
April-June 1985," and updated in NUREG-0090, Vol. 8, No. 3 and Vol. 8, No. 4.
It is further updated as follows.
As mentioned in the previous update, based upon the findings of the NRC Incident Investigation Team reported in NUREG-1154 (Ref. B-1), the NRC identified the con-cerns Toledo Edison Company (the licensee) should address for NRC review before resumption of operation of the plant can be approved.
These concerns were iden-tified to the licensee in a letter-dated August 14, 1985 (Ref. B-2).
The licensee has responded in a document submitted to the NRC on September 10, 1985, entitled, " Davis-Besse Course of Action" (Ref. B-3).
The NRC staff has essentially completed its review of this document and is currently preparing a Safety Evaluation Report to address plant re-start.
Several open items still exist for which the Staff has requested additional information from Toledo Edison.
The current schedule calls for completion of the Safety Evaluation Report by May 31, 1986, and briefing of the ACRS in early July.
The licensee is conducting its System Review and Test Program in which 34 safety systems are being extensively reviewed, including an evaluation of the system design requirements.
Previous surveillance tests of the systems are being ana-lyzed and additional tes, ting is being performed to demonstrate the operability of the systems.
The testing program includes 172 existing or modified surveillance tests and 106 new test procedures.
On March 16, 1986 the licensee identified possible cracks in the shafts of the reactor coolant pumps.
There are four pumps, two associated with each of the two steam generators.
One shaft was replaced with a spare shaft and the original shaft was sent to the Babcock and Wilcox Research Center for examination.
Pre-liminary test results were unable to duplicate the crack indications previously observed through ultrasonic testing.
Other non-destructive tests also failed to identify any evidence of cracking and destructive examination of the shaft con-firmed that no cracking was present.
(Also note Item 2, Appendix C, of this report.)
The licensee intends to continue its replacement program for the shafts, and the current outage will be extended until October 1986 to allow time for fabrication and installation of the new shafts.
The licensee will also be replacing the four bolts which connect the shafts to the pump impellers.
Examination of the bolts in one pump showed one bolt to be sheared, two with cracks, and one with a possible crack.
Future reports will be made as appropriate 85'-12 Management Control Deficiencies This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 3,
" Report to Congress on Abnormal Occurrences:
July-September 1985."
It is up-dated as follows.
As previously reported, the NRC Region III Office initiated a specific Task Force in July 1985 to perform an in-depth review of the LaSalle Nuclear Power Station, 43
because of the licensee's (Commonwealth Edison Company) continuing problems with the operation of the facility.
Subsequently, on November 22, 1985 the Regional Administrator of Region III issued a letter to the licensee (Ref. B-4) under 10 CFR S 50.54(f) requesting information on the licensee's plans to improve its performance in managing its maintenance, operation, and modification activities, including the problems identified in the Task Force's report.
The licensee replied to the request on December 23, 1985 (Ref. B-5).
The re-sponse described the licensee's program to reduce the backlog of outstanding work requests, minimize reactor scrams and unneeded actuations of the plant's emer-gency safety systems, improve the control of maintenance and plant modification activities, and upgrade the management at both the corporate and station level.
The licensee has met periodically with the NRC staff to review the status of its improvement program.
Data collected by the licensee through April 1986 generally shows evidence that the improvement program is effective in reducing such parame-ters as the number of automatic shutdowns (scrams) and unneeded actuations of emergency safety systems, the backlog of work requests, and the number of Licensee Event Reports attributed to personnel errors.
Region III will continue,to monitor the licensee's program in implementing its measures to improve regulatory performance.
LaSalle Unit 1 was shut down for a refueling and maintenance outage in October 1985 and remained shut down throughout the First Quarter of 1986.
During the outage, the licensee conducted inservice testing of the operability of pipe snub-bers (shock absoroers to protect the piping in the event of an earthquake or other pipe movement).
This testing determined that 51 or 375 small bore pipe snubbers and 10 of 275 large bore pipe snubbers failed to meet the testing re-quirements.
Further testing and review of the snubbers are underway.
On March 19, 1986, the NRC proposed a $50,000 fine against Commonwealth Edison for a violation occurring in October 1985 at LaSalle Unit 2 (Ref B-6).
The al-leged violation involved two of the three divisions of the unit's Emergency Core Cooling System being inoperable for a 13-hour period on October 5, 1985. One di-vision was out of service and a second division was rendered inoperable when a valve was closed for maintenance.
NRC requirements stipulate that the plant must restore one of the divisions within an hour or begin shutting the plant down.
The closed valve was ultimately reopened, restoring the division to operability, but the two divisions had been out of service for a 13-hour period.
Future reports will be made as appropriate.
- aaa*
85-13 Inoperable Steam Generator Low Pressure Trip This abnormal occurrence was originally reported, and closed out, in NUREG-0090, Vol. 8, No. 3, " Report to Congress on Abnormal Occurrences:
July-September 1985."
It is being reopened to report new information.
As previously reported, on October 29, 1985 the NRC Region I Office issued a Se-verity Level II violation and civil penalty in the amount of $80,000 (Ref. B-7) to the licensee.
The Regional Administrator emphasized that corrections were 44 l
i needed in the areas of improved administrative control of valves, control of de-sign changes, preparation and implementation of temporary procedures, and control of the post maintenance or post modification testing process including test de-sign, test procedures, and their review.
Subsequently, the licensee submitted corrective actions and requested a reduction in the severity level of the violation.
After consideration of the licensee's d
l request, the NRC decided that no reduction was warranted.
Therefore, on Janu-ary 22,1986 the NRC imposed a civil penalty of $80,000 (Ref. B-8).
On Janu-4 l
ary 28, 1986 the licensee paid the civil penalty.
l This item is considered closed for the purposes of this report.
i 85-14 Management Deficiencies at Tennessee Valley Authority This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 3,
" Report to Congress on Abnormal Occurrences:
July-September 1985."
It is up-dated as fcilows.
The NRC Staff, led by a senior management team, has identified a number of major Tennessee Valley Authority (TVA) issues requiring resolution prior to the restart of any of the TVA reactors.
The Staff briefed the Commission on these issues on January 7,1986, and stated that Sequoyah Unit 2 was expected to be the first reactor ready to resume operation. TVA briefed the Commission on a number of TVA activities on January 9, 1986, and noted the appointment of Steven White as the new Manager of Nuclear Power.
The Staff briefed the Commission on TVA status on February 7, 1986 and March 11, 1986.
By letter dated March 10, 1986, TVA submitted the.r revised response to the Sep-tember 17, 1985, 10 CFR S 50.54(f) letter (Ref. B-9) regarding corporate con-cerns.
This submittal is under review by the Staff.
TVA intends to revise the response submitted concerning Sequoyah, but a revised response has not been re-ceived.
TVA has not responded to the Browns Ferry and Watts Bar concerns.
There are six areas where there have been considerable TVA activity and which have also received a significant level of Staff attention.
These areas are equipment qualification, operational readiness, employee concerns, welding, elec-trical design calculations, and simulator evaluations of Sequoyan licensed per-sonnel.
The following paragraphs summarize activity during the first quarter CY 1986 in these areas, focusing primarily on the Sequoyah facility.
Equipment Qualification As TVA has completed portions of the equipment qualification (EQ) effort, the Staff conducted on-site reviews at Sequoyah, including an audit in November 1985, a two-week inspection in January 1986, and an inspection during the week ending i
February 14, 1986.
Additional on-site reviews will be conducted.
Discrepancies identified by both the Staff and TVA remain to be corrected; how-ever the TVA EQ Program appears to be sound.
The Staff currently estimates that all EQ discrepancies will be resolved by June 1986 and TVA will be able to cer-tify that Sequoyah is in compliance with the EQ rule.
45
Cables and unqualified electrical equipment are currently being replaced or up-graded in the Browns Ferry Unit 2 drywell (first Browns Ferry unit scheduled to be restarted).
TVA and the Staff will face additional effort to assure EQ compliance at Browns Ferry and Watts Bar.
Operational Readiness Operational readiness inspections at Sequoyah by NRC Region II continued during the first quarter CY 1986.
Some deficiencies have been identified which will require resolution prior to restart.
These deficiencies are in the areas of sur-veillance activities, maintenance, operating experience, pipe supports, and radia-tion protection.
Employee Concerns Nearly 5,000 TVA employee concerns have been raised; some of these involve safety-related and intimidation, harassment, or wrongdoing issues.
About 400 of these concerns apply to the Sequoyah facility.
TVA has established a program for evaluating employee concerns and is making the transition from a program adminis-tered by Quality Technolbgy Company (QTC) to a TVA-administered program.
The l
Staff has conducted several inspections of the QTC-administered employee concerns p rogram.
Resolution of employee concerns may be a pacing item leading to Sequoyah re-start or Watts Bar licensing.
On February 11, 1986, TVA submitted a summary of the TVA-Administered Employee Concerns Program and the methodology TVA will use for resolution of concerns gen-erated by the Watts Bar Program.
The Staff responded to TVA regarding this sub-mittal on February 28, 1986, and requested a TVA staff briefing and response to a number of Staff concerns.
TVA and QTC terminated their contract in April 1986.
The Staff has obtained copies of all of the QTC generated employee concerns records and is reviewing and evaluating many of these issues, as well as providing technical issues to TVA for resolution while maintaining employee confidentiality.
Th' aff is currently interviewing individuals and following up on allegations b'
a t directly to the NRC by concerned individuals.
In addition, the Staff is n
initiating an effort to review, on an expedited basis, all intimidation and ha-rassment issues and request identification on TVA actions taken on those items determined to be safety significant.
Welding Activities completed by the Staff and TVA in the first quarter CY 1986 are as follows:
The Staff reviewed the TVA Weld Reinspection Plan for Sequoyah and found it acceptable, subject to comments.
TVA completed reinspection of 800 welds at Sequoyah during the week ending February 14, 1986.
46
Members of the Staff were at the Sequoyah site February 24-28, 1986, con-ducting independent inspection of about 300 welds using the NRC's mobile nondestructive examination van.
Electrical Design Calculations Activities completed in the first quarter CY 1986 in this area are as follows:
The Staff conducted an on-site inspection of the Sequoyah facility in mid-January to evaluate the TVA program and on-site progress.
TVA has completed the Sequoyah electrical design review.
TVA is examining the entire design and configuration control program for Sequoyah, including areas outside of the electrical area.
This may result in considerable TVA and Staff effort in the next few months, and could become a pacing item for Sequoyah re-start.
Sequoyah Simulator Evaluations In response to high failure rates for operator requalification exams at Browns Ferry, the Staff has conducted simulator evaluations of Sequoyah operating per-sonnel.
These evaluations were completed during the week ending February 28, 1986 and the results were that 21 out of 24 personnel performed adequately. The one crew (3 persons) that was weak is receiving additional training and will be re-evaluated by TVA.
NRC will audit the licensee's retraining and re-evaluation.
Future reports will be made as appropriate.
85-20 Management Deficiencies at Fermi Nuclear Power Station This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 4,
" Report to Congress on Abnormal Occurrences:
October-December 1985."
It is up-dated as follows.
As part of its response to the NRC's request for information under 10 CFR S 50.54(f) (Ref. B-10), Detroit Edison Company established an Independent Over-view Committee to assess the management of the Fermi Unit 2 plant and to provide the licensee with recommendations for improvements.
The Committee issued its preliminary evaluation report on January 30, 1986.
Its recommendations included the hiring of individuals with previous experience at commercial nuclear power plants to fill key management positions at Fermi Unit 2.
The committee also recommended strengthening of the Nuclear Engineering organiza-tion and assuring full support of the Nuclear Production organization by other organizational components at the site.
Further reports are expected from the Overview Committee.
The licensee has also prepared a Performance Improvement Program for its security program as a result of numerous violations of NRC requirements and inadequate 47 D
l
implementation of the licensee's security plan.
On May 21, 1986, the NRC for-warded to the licensee a Notice of Violation and Proposed Imposition of Civil Penalty ($50,000), for 13 violations of security requirements (Ref. B-11).
The licensee has paid the fine.
The licensee has extended its outage which began in October 1985.
During the outage extension, the licensee will perform maintenance and testing previously scheduled for an October 1985 outage and replace the neutron sources used for monitoring reactor startup.
The sources had become depleted during the current extended outage.
During the outage the emergency diesel generators, previously reported to have incurred bearing damage, were repaired and subjected to an ex-tensive break-in and test program.
The type of lubricating oil for the diesels was changed and repairs were made to the fuel oil systems to preclude leakage into the lubricating oil, causing reduced viscosity.
The repair and test program was reviewed and accepted by the NRC.
Future reports will be made as appropriate.
OTHER NRC LICENSEES 85-10 Breakdown in Management Controls e
This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 2,
" Report to Congress on Abnormal Occurrences:
April-June 1985."
It is updated as follows.
As mentioned in the previous report, during August of 1984 inspections vero con-ducted by NRC Region I inspectors at the Cleveland, Gnio facility of Pittsburgh Testing Laboratory.
Several violations of NRC requirements were identified, in-cluding the performance of radiography by unqualified individuals.
The NRC Of-fice of Investigations performed an investigation and concluded that the Radia-tion Safety Officer at the Cleveland facility willfully violated NRC requirements and gave false information to an inspector.
As a result, on May 24, 1985, the Director of the NRC Office of Inspection and Enforcement issued an Immediately Effective Order removing the Radiation Safety Officer in Cleveland and requiring ceratin other corrective actions.
Subsequently, the case was referred to the U.S. Department of Justice (D0J) for prosecution.
The NRC held additional enforcement action in abeyance pending ac-tion by the D0J.
On February 19, 1986, the former Radiation Safety Officer of the Cleveland facility, and the President of Pittsburgh Testing Laboratory, act-ing for the Corporation, appeared in the U.S. Federal District Court for +he Northern District of Ohio and pled guilty to violations of 18 USC 1001 and the Atomic Energy Act.
In accordance with a plea bargaining agreement, the former Radiation Safety Officer was fined $2,500 and the Corporation was fined $15,000.
In addition, the judge strongly admonished the defendants regarding the serious-ness of their actions.
l On April 7, 1986, the NRC issued to the licensee a Notice of Violation and Pro-posed Imposition of Civil Penalties in the amount of $58,000 (Ref. B-12).
This enforcement action was taken because of the deliberate violations of NRC require-l ments', the falsification of records by site management, and the lack of condor demonstrated by both site and corporate management in their dealings with the NRC.
48 1
t
The licensee responded on May 6, 1986.
The NRC decided that no mitigation of the civil penalty was warranted.
Accordingly, on June 2, 1986, the NRC forwarded an Order Imposing Civil Monetary Penalties in the amount of $58,000 (Ref. B-13).
Unless new s'ignificant information becomes available, this item is considered closed for the purposes of this report.
49 o
a
APPENDIX C OTHER EVENTS OF INTEREST The following items are described below because they may possibly be perceived by the public to be of public health significance.
The items did not involve a ma-jor reduction in the level of protection provided for public health or safety; therefore, they are not reportable as abnormal occurrences.
1.
Failure of Lifting Rig Attachment While Lifting an Upper Guide Structure On November 6, 1985, Florida Power and Light Company (the licensee) experienced a potentially significant problem at their St. Lucie' Unit 1 facility while moving a heavy load over the reactor core in preparation for refueling.
St. Lucie Unit 1 is a Combustion Engineering-designed pressurized water reactor located in St. Lucie County, Florida.
While lifting the upper guide structure (UGS) from the reactor vessel, one of three lifting rig attachments gave way.
This placed the UGS lifting rig in a position canted upward approximately six inches and the guide structure canted downward approximately s,ix inches at one of the three attachment points.
The licensee attempted to lower the load back to its installed position, but the load cells indicated binding.
Therefore, the attempt was terminated after lowering the load a few inches.
The 50 ton load was left suspended about eight feet above the irradiated fuel.
The licensee declared an unusual event in accordance with plant emergency proce-dures.
All core alterations were suspended and containment integrity was en-hanced by resuming full use of the airlocks.
Temporary, primary manway covers were installed on both hot and cold legs to enhance the nozzle dams.
The UGS is supported in the reactor vessel by its upper flange.
It is aligned by eight alignment keys, four at the top and four at the bottom.
The structure fits down inside the core support barrel, just'above the fuel, assemblies.
The fuel assembly alignment plate is the bottom component of the UGS.
The lifting rig is attached to the upper guide structure by three vertically oriented bolts.
These bolts are attached from above the water line by torque tools that run down the hollow columns of the rig.
Combustion Engineering's procedure for attaching the rig calls for checking for thread engagement and torquing each bolt to 50 foot pounds.
The licensee's procedure had omitted the step concerning the check for thread engagement.
Subsequent inspection of the bolt that had pulled loose indicated that part of the last thread was stripped.
The failure assumed was that this bolt cross-threaded or bound due to rig-to guide-structure misalignment during attachment and reached the 50 foot pound torque requirement with only part of one thread engaged.
During the lift, the few inghes of unengaged bolt shaft were pulled through the lifting rig until the bolt head rested on the rig's surface at the bottom of the column, resulting in an imperceptible tilt.
The resulting lat-eral load was initially supported by the guide pins.
Wheg the rig and guide structure were lifted about eight feet, it is surmised that sufficient lateral motion was permitted to allow the thread of the improperly engaged bolt to strip.
51 o
A
To prevent a recurrence of the UGS failure, the licensee has revised the appro-priate procedure to incorporate sign-off steps for full thread engagement of the lifting bolts prior to lifting the UGS.
The safety implication of dropping heavy loads into the open reactor vessel is that irradiated fuel assemblies might be sufficiently damaged to release the radio-active gases and iodines held within the fuel clad gap.
The event, however, did not result in any actual adverse effects on health and safety of the public or licensee personnel.
NRC Inspection and Enforcement Information Notice No. 86-06 was issued on February 3,1986 to notify licensees of this potentially significant event (Ref. C-1).
l naa*****
i 2.
Degraded Reactor Coolant Pump Shafts j
On January 1,1986, Crystal River Unit 3* was shutdown because of a problem with i
reactor coolant pump (RCP) "A".
Examination showed that the RCP "A" shaft had failed completely within the hydrostatic bearing due to fatigue propagation of small cracks.
The cause is still uncertain but may involve material properties, design inadequacies, manufacturing cracks, residual stresses, bending moments, or i
thermal cycling, j
The failure occurred rapidly with essentially no warning to operators.
It has j
been determined that the reactor coolant flow rate decreased from the four pump i
value to the three pump value within three sec6nds following failure of the 1
shaft.
The post-trip review indicates that the reactor tripped on a power / flow mismatch signal approximately five seconds after the failure.
The RCP seals were damaged but did not leak at the time of shaft failure.
The RCP "B" shaft was removed and extensive cracks were indicated at a slightly dif-i ferent location.
RCP "C" shaft ultrasonic testing (UT) indicated a crack less severe than "B" pump.
RCP"D"shaftUTindicatedacrackofthesamemagnitude l
as "B" pump.
All eight cap bolts securing the impeller to the shaft on A" and i
"B" pumps were cracked in multiple places (some broken).
Five of eight pins which take the torque between the impeller and shaft on "A" and "B". pumps were i
also cracked.
Their appearance is similar to the defective cap bolts.
The reac-i tor coolant pumps at Crystal River Unit 3 were manufactured by Byron-Jackson (BJ).
i i
The Toledo Edison Company, which utilizes RCPs similar to those at Crystal River Unit 3, took advantage of being shutdown to examine the RCP shafts at Davis-Besse Unit 1.**
They used the same team from Babcock & Wilcox (B&W) that examined the Crystal River Unit 3 RCP shafts.
UT examination showed crack indications in all four RCP shafts.
These RCPs were also manufactured by BJ, and are of the same horsepower and speed (RPM) as the Crystal River Unit 3 RCPs.
Additionally, the pump shafts are made of the same material (ASTM A461 GR 660) and have seal injectioncooling.
- Crystal River Unit 3 is a pressurized water reactor, design by Babcock &
Wilcox, operated by Florida Power Corporation and located in Citris County, Florida.
- Davis-Besse Unit 1 is a pressurized water reactor, designed by Babcock &
Wilcox, operated by Toledo Edison Company and located in Ottawa County, Ohio.
52 A
o
On March 21, 1986, the NRC Office of Inspection.and Enforcement transmitted In-formation Notice No. 86-19 on the Crystal River Unit 3 and Davis-Besse Unit 1 RCP shaft inspections to all nuclear power reactor facilities holding an operating license or a construction permit (Ref. C-2).
Meanwhile, one of the shafts from Davis-Besse was sent to the B&W Research Center for further testing.
As discussed further in Appendix B, A0 85-7, of this,re-port, preliminary test results at B&W failed to duplicate the crack indications.
Subsequent destructive examination of the shaft confirmed that there were no cracks in the shaft.
The licensee intends to replace the RCP shafts, even though the crack indications could not be confirmed.
The bolts which connect the shafts to the impeller of one pump were found to be damaged.
The need to replace the bolts in all four pumps contributed to the decision to continue the shaft re-placement program.
The NRC Office of Nuclear Reactor Regulation is continuing to investigate the generic implications of multiple RCP shaft cracking in similarly designed pumps.
This is because a number of nuclear power plants utilize BJ RCPs, and the failure of an RCP shaft can be similar to a seized shaft accident in which specified ac-ceptable fuel design limits (SAFDL) may be exceeded.
A A A A A A A A 3.
Earthquake in Vicinity of a Nuclear Power Plant At 11:46 a.m., January 31, 1986, an earthquake measuring 5.0 on the Richter Scale struck northeastern Ohio. The epicenter of the quake was about 10 miles south of the Perry Nuclear Plant, located in Lake County, Ohio.
The plant is a General Electric-designed boiling water reactor and is operated by Cleveland Electric Illuminating Company.
At the time of the event, the plant was in preoperational testing and had not yet received an operating license; therefore, there was no fuel in the reactor (new fuel, however, was being stored on site).
There were no effects on public health or safety; however as noted below, the event received considerable media and Con-gressional interest.
At 12:06 p.m., following a brief assessment of this situation, the control room chief supervisor implemented emergency procedures.
Non-essential employees were directed to leave the site and trained emergency personnel reported to their as-signed work stations.
A walkdown of the plant by the licensee (and later by an NRC inspection team from the NRC Region III office and from NRC headquarters in Bethesda, Maryland) revealed no structural damage.
(Earthquake instrumentation at the Fermi nuclear plant in Monroe, Michigan, and the Palisades Nuclear Plant near South Haven, Michigan, also registered the quake movement; no damage was reported at either plant.)
The in plant seismic monitoring instruments were triggered at Perry, some of which indicated high frequency exceedances (above 20 Hz) which raised questions on the plant's seismic design adequacy.
This information was duly noted in the media, and resulted in widespread controversy and concern.
l l
53 h - - -
A
On February 11, 1986, the licensee held a special briefing on the results of its investigation of the earthquake and its effect on the Perry plant.
It concluded that although some instrumentation showed a high reading, the ground motion was of very short duration and of low energy, and therefore was of no safety significance.
The NRC Staff and the NRC's Advisory Committee on Reactor Safeguards (ACRS) both calculated the effect of the earthquake on the Perry plant.
The Staff, in a Sup-plement to the Perry Safety Evaluation Report, concluded that Perry's seismic design was adequate, and that the plant could be licensed without undue risk to the health and safety of the public.
The Staff said that the earthquake stresses at Perry were substantially lower than the corresponding design stresses, and, therefore, were of no safety significance.
In a report to NRC Chairman Nunzio Palladino dated March 17, 1986, the ACRS agreed with the Staff's conclusion, while adding that there currently exists some possi-bility that the earthquake was related to deep well injection activities that took place near Perry in years past.
It also noted that the earthquake could have been the result of solution mining in the area.
On April 8, 1986, the House of Representatives Subcommittee on Energy and the Environment, Committee on Interior and Insular Affairs, held a hearing on the earthquake.
The U.S. Geological Survey and the licensee have completed their investigations of the possibility of the agricultural liquid waste injection wells as a cause of the January 31, 1986 earthquake.
These investigations have concluded that the possibility of the earthquake being caused by the injection wells is low, and that the preponderance of evidence obtained from the investigations performed support that the cause was tectonic in origin.
The licensee has completed the seismological / geological confirmatory work re-quired by the Staff for the full power licensing of Perry Unit 1.
It is the in-tent of this work to reaffirm the plant's seismic design.
The licensee expects to have made all submittals on this work to the Staff by the end of June 1986.
Current schedules project completion of the Staff's evaluations in late July 1986 and the issuance of the full power license in early August 1986.
maaaa aaa 4.
Inoperable Standby Liquid Control System During surveillance tests at Vermont Yankee on February 8, and 11,1986, the Standby Liquid Control (SLC) system was found to be inoperable when the redundant explosive squib valves failed to actuate.
Although the SLC system was not re-quired to be operable at the time of the test (since the plant was shut down and defueled to replace the recirculation system piping), investigation showed that the system was not functional as required by the technical specifications during l
the plant operating cycle that started in July 1984 and ended in September 1985.
l Vermont Yankee is a boiling water reactor (8WR) designed by the General Electric Company, and is operated by the Vermont Yankee Nuclear Power Corporation (the licensee).
The plant is located in Windham County, Vermont.
54 A
d
l The SLC system requires manual initiation and explosive firing of two parallel i
redundant squib valves to inject borated water into the reactor vessel in the event that soluble boron is required as a backup to the control rod system for reactivity control.
During surveillance tests on February 8 and 11, 1986 with the plant shut down and defueled, both valves failed to fire when called for by the system initiation switch.
Permanently installed plant wiring is connected to the explosive squib valves to provide firing circuits and firing circuit continuity monitoring.
The squib valve is actuated by shearing a cap on the valve inlet fitting, which opens a flow path through the valve.
The cap is sheared off by a trigger assembly that is actuated by an explosive charge in the primer subassembly.
The primer charge for a squib valve is ignited by the firing circuit by concurrently applying 120 Vac to two bridgewires embedded in the charge.
Either of the bridgewires can ignite the charge.
When the SLC system is in the standby mode, the electrical continuity of the squib valve firing circuit is monitored and indicators are lit on the control room panel when a firing path for each valve exists.
During surveillance tests on February 8 and 11, 1986, both the A and B squib valves failed to fire when the control switch was placed in the System 1 and System 2 positions, respectively.
It was notable that the continuity monitoring circuits for both valves' indicated the system was " ready to fire."
However, the firing circuits could not work as wired.
Investigation of the firing circuits found that the primer chambers within each squib valve were mis-wired by having both 120 Vac high side leads wired to one ignition bridgewire, and both neutral leads wired to the other bridgewire.
Both squib valves were last tested satisfactorily during a July 11, 1984 surveillance.
The replacement squib valve explosive primers, insta41ed following the July 1984 test and presumed to be electrically the same as the primers successfully fired, were electrically different due to a changed connector pin to bridgewire config-uration.
The primer wiring change was a manufacturing error unknown to the vendor, Conax Corporation of Buffalo, New York, that had not been detected by the licensee during receipt inspection or preservice testing of the parts prior to installation in the SLC system.
The design of the continuity monitoring circuit was not capable of detecting the problem.
Therefore, during plant operating cycle #11, which ran from July 1984 until September 1985, the SLC system was not available as a backup to the control rod system because the squib discharge valves would not have opened upon demand.
However, the control rods and reactor protection system were operable during the entire period when the SLC system was not functional.
Additionally, the Recircu-lation Pump Trip (RPT) and Alternate Rod Insertion (ARI) systems were operable during the last operating cycle.
In the event that an anticipated transient had occurred concurrent with the failure of the control rod scram function (an ATWS event), the RPT system could have been used to mitigate the event by reducing reactor power, and the ARI system could have been used to provide for slow rod insertion.
If a common mode failure had occurred, such that both the control rod scram fur ction and the ARI system were inoperable, then no additional reac-tivity control system would have been available to mitigate the consequences of the event.
However, the probability of such a common mode failure, occurring simultaneously with an anticipated plant transient, is considered to be small.
55
The principal cause for the squib valve failures and the loss of the SLC system function was the incorrect wiring in the primer chamber supplied by the vendor in 1983 and initially installed in July, 1984.
A total of six defective primer chambers were identified either installed or in stores at the Vermont Yankee site.
Following notification of the problem at Vermont Yankee, the vendor identified 51 other possible defective primer chambers that had been supplied to other facilities.
Actions regarding this item are discussed further below.
A secondary cause for the event was the failure by the licensee to detect the manufacturing defect prior to using the parts in the plant.
The error was iden-tified by the licensee's staff in February 1986 while testing the installed valves at the end of the operating cycle.
If a test had been conducted for a representa-tive sample of the chambers installed in the SLC system in 1984, the loss of SLC system function could have been prevented.
Although a squib valve from the vendor's same manufacturing lot was " bench" tested in the plant maintenance shop prior to installation, the " bench" test was deficient in that it only verified the adequacy of the explosive material, but did not test the electrical wiring configuration.
NRC inspection also found irregularities in the plant wiring in that the as-found firing circuit wiring differed from the design drawings.
The difference occurred following a des'ign change to the firing circuits in 1977 when diffi-culties encountered during the installation resulted in a needed field modifi-cation of the firing circuit.
The adequacy of the modified circuit was demon-strated by the successful completion of the annual surveillance tests from 1977 to 1984 with primer chambers of the type supplied in 1977.
However, the field modification did not go through the normal review process and therefore was not reflected in a change to the as-built drawing of the firing circuit.
Since the field modification would have been approved in 1977 if submitted, the primary cause for the SLC system failure remains the primer manufacturing error and the failure of the licensee to detect the error.
Licensee actions to prevent recurrence of the event were in progress as of the date of this report and were under review by the NRC Region I staff.
All de-fective primer chambers were separated from good ones and plans were in progress to return the defective ones to the vendor.
Changes were made to the procedures used to test the primer chambers to correct the deficiencies in the " bench" test, and to provide for an in-situ test of a representative primer chamber from the vendor's same manufactured lot prior to the use of new chambers in the SLC sys-tem.
The surveillance testing of the SLC system will be repeated using the updated procedures prior to declaring the SLC system operable and installing fuel in the reactor.
The licensee plans to evaluate the feasibility and cost benefit of making the continuity monitoring circuit capable of testing not only continuity, but also readiness to fire.
Plant design drawings will be updated to reflect the as-built conditions.
Plant startup from the recirculation pipe i
replacement outage is scheduled for June 1986.
The Conax Corporation submitted a report to the NRC on March 5, 1986 per the requirements of 10 CFR Part 21 to describe the nature and the probable cause of the manufacturing error.
The error occurred during a temporary change in the manufacturing location for the parts and was caused by a combination of assembly and inspection by inexperienced personnel, the lack of connector pin identifica-tions on the connector, and by incorrect assembly operation sheets for the primers.
1 I
l 56
Conax identified a corrective action plan in the Part 21 report that, when com-pleted, should prevent recurrence of the manufacturing error.
The vendor also reported that, based on telephone contacts with other users of the potentially defective parts as of February 28, 1986, no other faulty units had been located.
The NRC Region I staff is reviewing the licensee's corrective actions and will verify that actions are completed to restore the SLC system to an operable status per the technical specifications prior to subsequent plant operations.
Following notification of the SLC system test failure, the NRC worked with the
' licensee and vendor staffs to determine whether the problem applied to other operating plants.
The vendor identified six plants aside from Vermont Yankee that could possibly have defective parts.
The action taken by the vendor is discussed above.
The NRC Office of Inspection and Enforcement issued Information Notice No. 86-13 on February 21, 1986 (Ref. C-3) to all BWR facilities to describe the problems at Vermont Yankee and to recommend actions be taken at the other facilities to check the SLC system firing circuits and the primer chambers supplied by Conax.
The Information Notice also identified the plants that had potentially suspect primer chambers.
Part numbers and serial numbers of the suspect chambers were also provided.
Based on the suggestions from the Vermont Yankee staff, the NRC staff is planning to issue a supplement to Information Notice No. 86-13 that will clarify the cause of the problem and emphasize to power plant utilities the importance of adequately testing vendor supplied parts prior to use at their facilities.
The results of the special inspection performed by NRC Region I of the circum-stances associated with the event were forwarded to the licensee on March 11, 1986 (Ref. C-4).
On April 22, 1986, NRC Region I forwarded a Notice of Violation to the licensee (Ref. C-5).
No civil penaltv was imposed.
The event was of concern because for a period of about 15 months, both redundant components of the SLC system were inoperable.
Had the system been called upon to operate, it would have taken the operators a considerable amount of time to analyze the problem and to correct it.
The ATWS rule (i.e., 10 CFR 50.62) for BWRs requires a reliable ARI system, a reliable SLC system, and a reliable system for tripping the recirculating pumps (for Vermont Yankee, this system is designated as the RPT system).
The latter system is to quickly reduce the reactor power.
As previously discussed, both the ARI and RPT systems were operable during the nearly 15 month period to miti-gate the consequences of an ATWS event.
For the loss of the SLC system to be consequential, one would have to postulate a second, or concurrent, failure, such as loss of the ARI system.
Such conditions are possible because the normal control rod scram subsystem and the ARI system share certain components (e.g., scram discharge volume).
However, the probability of such a condition, occurring simultaneously with an anticipated reactor transient, is considered small; therefore, the Vermont Yankee event is considered below the threshold for abnormal occurrence reporting.
57
REFERENCES FOR APPENDICES B-1 U.S. Nuclear Regulatory Commission, " Loss of Main and Auxiliary Feed-water Event at the Davis-Besse Plant on June 9, 1985," USNRC Report NUREG-1154, published July 1985.**
B-2 10 CFR S 50.54(f) letter from Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, to Joe Williams, Jr., Senior Vice President, Nuclear, Toledo Edison Company, Docket No. 50-346, August 14, 1985.*
B-3 Letter from John P. Williamson, Chairman and Chief Executive Officer, Toledo Edison Company, to Harold R. Denton, Director, NRC Office of Nu-clear Reactor Regulation, Docket No. 50-346, September 10, 1985.*
B-4 10 CFR S 50.54(f) letter from James G. Keppler, Regional Administrator, NRC Region III, to Cordell Reed, Vice President, Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, November 22, 1985.*
B-5 Letter from Corde.11 Reed, Vice President, Commonwealth Edison Company, to James G. Keppler, Regional Administrator, NRC Region III, Docket Nos. 50-373 and 50-374, December 23, 1985.*
B-6 Letter from James G. Keppler, Regional Administrator, NRC Region III, to James J. O'Connor, President, Commonwealth Edison Company, forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, also for-warding Inspection Reports No. 50-373/85033 (DRP) and No. 50-374/85034 (DRP), Docket Nos. 50-373 and 50-374, March 19, 1986.*
B-7 Letter from Thomas E. Murley, Regional Administrator, NRC Region I to J. B. Randazza, Vice President - Nuclear Operations, Maine Yankee Atomic Power Company, forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, Docket No. 50-309, October 29, 1985.*
B-8 Letter from James M. Taylor,' Director, NRC Office of Inspection and En-forcement, to J. B. Randazza, Vice President - Nuclear Operations, Maine Yankee Atomic Power Company, forwarding an Order Imposing a Civil Mone-tary Penalty, Docket No. 50-309, January 22, 1986.*
B-9 10 CFR S 50.54(f) letter from William J. Dircks, NRC Executive Director for Operations, to Charles Dean, Chairman, Board of Directors, Tennessee Valley Authority, Docket Nos. 50-259, 50-260, 50-296, 50-327, 50-328, 50-390, 50-391, 50-438, and 50-439, September 17, 1985.*
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection.
Available for purchase from the GP0 Sales Program, Superindendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7982.
59
B-10 10 CFR 9 50.54(f) letter from James G. Keppler, Regional Administrator, NRC Region III, to-Wayne H. Jens, Vice President-Nuclear Operations, Detroit Edison Company, Docket No. 50-341, December 24, 1985.*
B-11 Letter from James G. Keppler, Regional Administrator, NRC Region III, to Frank E. Agosti, Vice President, Nuclear Operations, Detroit Edison Com-pany, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, also forwarding Inspection Report No. 50-341/85047 (DRSS),
Doc ht No. 50-341, May 20, 1986.*
B-12 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to M. Ruyan, President, Pittsburgh Testing Laboratory, for-warding a Notice of Violation and Proposed Imposition of Civil Penalties, Docket No. 030-05985, April 7, 1986.*
B-13 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to M. Ruyan, President, Pittsburgh Testing Laboratory, for-warding an Order Imposing Civil Monetary Penalties, Docket No. 030-05985, June 2, 1986.*
C-1 U.S. Nuclear Regulatory" Commission, Inspection and Enforcement Informa-tion Notice No. 86-06, Failure of Lifting Rig Attachment While Lifting the Upper Guide Structure at St. Lucie Unit 1," February 3, 1986.*
C-2 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Informa-tion Report No. 86-19, " Reactor Coolant Pump Shaft Failure at Crystal River," March 21, 1986.*
C-3 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-13, " Standby Liquid Control System Squib Valves Failure to Fire," February 21, 1986.*
C-4 Letter from Richard W. Starostecki, Director, Division of Reactor Projects, NRC Region I, to Warren P. Murphy, Vice President and Manager of Operations, Vermont Yankee Nuclear Power Corporation, forwarding Inspection Report No. 50-271/86-05, Docket No. 50-271, March 11, 1986.*
C-5 Letter from Thomas E. Murley, Regional Administrator, NRC Region I, to Warren P. Murphy, Vice President and Manager of Operations, Vermont Yankee Power Corporation, forwarding a Notice of Violation, Docket No. 50-271,
[
April 22, 1986*
l l
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).
60
N#C POAM 35 W S. NuCLli.4 kEruLATOR7 COesnelSSIOas i atrOnT Nuveta #Aargase ey TsOC, see yet ife,,asays
'.'Rc's iio:'
NUREG-0090 BIBLIO2RAPHIC DATA SHEET Vol. 9, No. 1 3 oi.non SEE tNs7 RUCTIONS ON YME REVERSE 2 YsTLE AN0 50Gf sTLE J LE AVE SLANE Report to Congress on Abnormal Occurrences January-March 1986
. oArE RE.OR, coo,arto l
, EAR oONr.
September 1986
- aur-ORisi 6 D AT E REPOR Y ISSUED MONTM vtAR September 1986
- PERFORMING ORGANil AT 60N NAVE AND W A8 LING ADDRESS asses or te ceses S PROJEcfiT A5m WORK UNif NuweER w
Office for Analysis and Evaluation of Operational Data U.S. Nuclear Reculatory Corm ssion Washington DC 20555 10 SPONSORING ORGANi2Af EON NAYE AND MAitsNG ADDRESS fracmerte Coorr its TYPE OF REPORY Same as 7 above.
Quarterly b PEReOD cOv tREO rince s,ve sseesi s
January-March 1986 12 SUPPL EME N T AR v NOT E S 13 Aesf R ACT (200 weres or sessJ Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to be significant from the standpoint of public health dnd safety and requires a quarterly report of such events to be made to Congress. This report covers the period January 1 to March 31, 1986. Duri ng the report period, there were two abnormal occurrences at the nuclear power plants licensed to operate. The events were (1) a loss of power and water hammer event and (2) a loss of integrated control system power and overcooling transient. There were five abnormal occurrences at the other NRC licensees.
The events were (1) a rupture of an uranium hexafluoride cylinder and release of gases, (2) a therapeutic medical misadministration, (3) an overexposure to a member of the public from an industrial gauge, (4) a breakdown of management controls at an irradiator facility, and (5) a tritium overexposure and laboratory contamination. There were four abnormal occurrences reported by the Agreement States. Three of the events involved radiation injuries to people working either as radiographers or assistant radiographers; the other event involved contamination of a scrap steel facility. The report also contains information updating some previously reported abnormal occurrences.
,. Av,A g e,
- i. oOcuuf N, AN AL,s.
...a v.ORos oE scm, Proms Water Hammer; Loss of Feedwater; Ruptured Components; Loss of Integrated Unlimited Control System Power; Overcooling Transient; Rupture of Uranium e stcuR.r classiPicariON Hexafluoride Cylinder; Release of Toxic Gas and Death of Employee; t ra-Therapeutic Medical Misadministration; Overexposure of Unclassified
. iDENei.,E Ri OPEN ENo'o ""5 Member of Public; Breakdown of Management Controls; Overexposure of Employees; Laboratory Contamination; Radiation Injury of Radiographers; Unclassified Scrap Steci Facility Contamination; Failure of Lifting Ri'g; Failure of Reactor Coolant Pump Shafts; Earthquake; Inoperable Standby Liquid Control
a
System
' UNITED STATES speciAt sounts.ctAss aArt NUCLEAR CEZULATCEY COMMISSION cosrAce e sses ruo WASHINGTON, D.C. 20666 wfA7"Ec.
PERMIT No G47 OFFICIAL BUSINESS PENALTY FOR PR!VATE USE, $300 1ANIC01CV1CY1 I
120555078877 us NRC gjfV puk 1 BR-PDR NUREG 0
g W-501 DC 20555 j
yASHINGTON
,G Q
ZCD:xi /
!o E>
a._
n O
s O
o a
e
,