ML20234D895
| ML20234D895 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/1987 |
| From: | NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-0090, NUREG-0090-V10-N02, NUREG-90, NUREG-90-V10-N2, NUDOCS 8801070275 | |
| Download: ML20234D895 (53) | |
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NUREG-0090 Vol.10, No. 2 Report to Congress on Abnormal Occurrences April - June 1987 U.S. Nuclear Regulatory Commission Office for Analysis and Evaluation of Operational Data pa"%
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"SR2 38#H S722 o 0090 R PDR
F NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, l
Washington, DC 20013-7082
- 3. The National Technical information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and
!!censee documents and correspondence.
The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.
Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other iederal agencies and reports prepared by the Atomic j
Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions, Federal Reg / ster notices, federal and j
state legislation, and congressional reports can usually be 'abtained from these libraries.
l Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.
Single copies of NRC draf t reports are available free, to the extent of supply, upon written request to the Division of Information Support Services, Distribution Section,' U.S, Nuclear Regulatory Commission, Washington, DC 20555.
Copies of industry codes and standards used in a substantive manner in the NRC regulatory process i
j are maintained at the NRC Library,7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards institute,1430 Broadway, New York, NY 10018.
NUREG-0090 Vol.10, No. 2 Report to Congress on Abnormal Occurrences April - June 1987 Date Published: November 1987 Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission Washington, DC 20555 p,....,
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Provious Reports in SQries NUREG 75/090, January-June 1975, NUDEC-0090, Vol.4, No.2 April-June 1981, published October 1975 published October 1981 NUREG-0090-1, July-September 1975 NUDEG-0090, Vol.a. Wo.3, July-September 1981, published March 1976 published January 1982 l
NUDEG-0090-2, October-December 1975, NUDEG-0090 Vol,4,'No.4, October-December 1981, l
published March 1976 published May 1982
)
NUREG-0090-3, January-March 1976, NUREG-0090, Vol.5, No.1, January-March 1982, published July 1976 published August 1982 i
l NUREG-0090 4. April-June 1976, NUREG-0090, Vol.5, No.2, April-June 1982, published March 1977 published December 1982 NUREG-0090-5, July-September 1976, NUREG-0090, Vol.5, No.3, July-September 1982, published March 1977 published January 1983 NUREG-0090-6, October-December 1976, NUREG-0090,Vol.5,No.4,betober-December 1982, publisted June 1977 published May 1901 i
NUREG-0090-7, January-March 1977, NUREG-0090 Vol.6, No.1, January-March 1983, j
publisted June 1977 published September 1983 NUREG-0090-8, April-June 1977, NUREG-0090, Vol.6, No.2, April-June 1983, published September 1977 published November 1983 NUREG-0090-9, July-September 1977, NUREG-0090, Vol.6, No.3, July-September 19R3, published November 1977 published April 1984 NUREG-0090-10, October-December 1977, NUREG-0090, Vol.6, No 4, October-Cecember 1983, publisted March 1978 published May 1984 NUREG-0090, Vol.1, No.1, January-March 1978 NUREG-0000, Vol.7, No.1, January-March 1984, published June 1978 published July 1984-NUREG-0090 Vol.1, No.2, April-June 1978, NUREG-0090, Vol.7, No.2, April-June 1984, published September 1978 published October 1984 NUREG-0090, Vol.1, No.3, July-September 1978, NUREG-0090 Vol.7, No.3, July-September 1984, publisted December 1978 published April 1985 NUREG-0090, Vol.1, No.4, October-recember 1978 NUREG-0090, Vol.7, No.4 October-December 1984, published March 1979 published May 1985 NUREG-0090, Vol.2, No.1, January-March 1979, NUDEG-0090, Vol.8, No.1, January-March 1985, published July 1979 published August 1985 NUREG-0090, Vol.2, No.2, April-June 1979, NUREG-0090, Vol.8, No.2. April-June 1985, publisted November 1979 published November 1985 NUREG-0090, Vol.2, No.3, July-September 1979 NUREG-0090, Vol.8, No.3,.luly-September 1985, publisheu February 1980 published February 1986 NUREG-0090, Vol.2 No.4, October-December 1979, NUREG-0090, Vol.8, No.4, October-December 1985, published April 1980 published May 1986 NUREG-0090, Vol.3, No.1, January-March 1980, NUREG-0090, Vol.9, No.1, January-March 1986, published September 1980 published September 1986 NUREG-0090, Vol.3, No.2, April-June 1980, NUREG-0090 Vol 9 No.2, April-June 1986, published November 1980 published January 1987 NUREG-0090, Vol.3, No.3, July-September 1980, NUREG~0090, Vol.9, No.3, July-September 1986, published February 1981 published April 1987 NUREG-0090, Vol.3, No.4, October-December 1980, NUREG-0090, Vol.9, No.4, October-December 1986, published May 1981 published July 1987 NUREG-0090, Vol.4, No.1, January-March 1981, NUREG-0090, Vol.10, No.1, January-March 1987, published July 1981 published October 1987
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 Congress.
This report covers the period from April 1 to June 30, 1987.
The report states that for this reporting period, there were no abnormal occurrences at the nuclear power plants licensed to operate.
There were five abnormal occurrences at the other NRC licensees.
Three involved medical misadministration (two diagnostic and one therapeutic); one involved the issu-ance of an NRC Order to remove a hospital's radiation safety officer due to falsification of certain records; and one involved a significant breakdown in management and procedural controls at an industrial radiography licensee.
There was one abnormal occurrence reported by an Agreement State (Idaho).
The item involved radiographer overexposure.
The report also contains information updating some previously reported
]
abnormal occurrences.
i i
iii
CONTENTS Page iii ABSTRACT................................................................
vii PREFACE.................................................................
vii INTRODUCTION.......................................................
vii THE REGULATORY SYSTEM..............................................
viii REPORTABLE OCCURRENCES.............................................
x AGREEMENT STATES...................................................
x FOREIGN INFORMATION................................................
REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, APRIL-JUNE 1987.............
1 1
NUCLEAR POWER PLANTS...............................................
l FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)............
1 OTHER NRC LICENSEES (Industrial Radiographer, Medical Institutions, Industrial Users, etc. )............................
1-1 87-9 Diagnostic Medical Misadministration........................
I 87-10 Therapeutic Medical Misadministration.......................
2 3
J 87-11 Diagnostic Medical Misadministration.........................
87-12 NRC Order Issued to Remove a Hospital's Radiation Safety Officer....................................................
5 87-13 Significant Breakdown in Management and Procedural Controls j
at an Industrial Radiography Licensee........................
7 q
l AGREEMENT STATE LICENSEES..........................................
8
/
AS87-3 Radiographer Overexposure..................................
9 j
i 13 i
REFERENCES..............................................................
APPENDIX A - ABNORMAL OCCURRENCE CRITERIA..............................
15 APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES..........
17 i
NUCLEAR POWER PLANTS...............................................
17 77-9 Environmental Qualification of Safety-Related Electrical 17 Equipment Inside Containment................................
79-3 Nuclear Accident at Three Mile Island........................
18 8137 Blockage of Coolant Flow to Safety-Related Systems and 21 Components..................................................
85-14 Management Deficiencies at Tennessee Valley Authority.......
23 87-1 NRC Order Suspends Power Operations at Peach Bottom Facility Due to Inattentiveness of the Control Room Staff...
27 i
i l
v l
4
l 1
CONTENTS (continued)
Page OTHER NRC LICENSEES................................................
28 86-25 Suspension of License for Servicing Teletherapy and Radiography Units............................
28 APPENDIX C - OTHER EVENTS OF INTEREST..................................
31 REFERENCES (FOR APPENDICES).........................................
39 vi l
l
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 is<ued on each abnormal occurrence with copies distributed to the NRC Public D,cument Room and all Local Public 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 April 1 to June 30, 1987.
Information reported on each event includes:
date and place; nature and probable consequences; cause or causes; and actions taken to prevent recurrence.
1 THE REGULATORY SYSTEM The system of licensing and regulation by which NRC carries out its responsibil-itie:; 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; enforcing license requirements; and carrying on 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.
vii i
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 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 l
receive up to three rems of whole body exposure in a calendar quarter.
Higher l
values are permitted to the extremities or skin of the whole body.
For unre-stricted areas, permissible levels of radiation are considerably smaller.
Pe r-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.
l REPORTABLE OCCURRENCES Actual operating experience is an essential input to the regulatory process for assuring that licensed activities 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 l
to help identify safety concerns early, to improve dissemination of such infor-l 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 I
of the NRC while the latter system is a voluntary, industry-supported system operated by the Institute of Nuclear Power Operations (INP0), 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 i
Atomic Energy Commission (the predecessor of the NRC) established in 1973 a viii j
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.
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 esta-blished by Commission rulemaking which modified and codified the former LER system.
The purpose was to standardize the reporting requirements for all nuclear power plant licensees and eliminate reporting of events which were of low individual significance, while requiring more thorough documentation and All such analyses by the licensees of any events required to be reported.
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 The amendment required under specific sections of 10 CFR Part 20 and Part 50.
to the Commission'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 effective January 1,1984, the NRC amended its immediate notification requirements 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 l
Commission considered it may be necessary to make participation manadatory 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 to NPRD system data, it would assume responsibility for management and funding of the NPRD system.
INP0 reports that significant improvements in licensee The Commission considers the NPRD system to be a j
participation are being made.
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.
I I
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 1
l 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.
ix o
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 States assume regulatory authority over byproduct, source and special nuclear materials (in quantities not capable of sustaining a chain reaction).
Compar-i able 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 at facilities of Agreement State licensees should be included in the quarterly reports to Congress. The abnormal occurrence criteria included in Appendix A are 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 i
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.
i x
REPORT TO CONGRESS ON ABNORMAL OCCURRENCES APRIL-JUNE 1987 NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to operate during the second calendar quarter of 1987.
As of the date of this report, the NRC had not determined that any events were abnormal occurrences.
FUEL CYCLE FACILITIES (Other Than Nuclear Power Plants)
The NRC is reviewing events reported by these licensees during the second calen-dar quarter of 1987.
As of the date of this report, the NRC had not determined l
that any events were abnormal occurrences.
j OTHER NRC LICENSEES (Industrial Radiographer, Medical Institutions, Industrial Users, etc.)
There are currently about 9,000 NRC nuclear material licenses in effect in the l
l United States, principally for use of radioisotopes in the medical, industrial, l
and academic fields.
Incidents were reported in this category from licensees l
I such as radiographer, medical institutions, and byproduct material users.
The NRC is reviewing events reported by these licensees during the second calen-dar quarter of 1987.
As of the date of this report, the NRC had determined that the following events were abnormal occurrences.
87-9 Diagnostic Medical Misadministration 1
The following information pertaining to this event is also being reported con-currently in the Federal Register.
Appendix A (see the general criterion) of l
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 January 21, 1987, a 66 year-old female at Halifax-South Boston Community Hospital, South Boston, Virginia, received 782 microcuries of I-131 instead of a 100-microcurie dose usually given for a thyroid scan.
Nature and Probable Consequences - The purpose of the scan was to rule out the presence of a substernal thyroid, following removal of the normal thyroid many years ago.
The thyroid scan and confirming computerized axial tomography (CAT) l scan demonstrated the presence of a nonfunctional substernal thyroid.
No adverse effects to the patient are expected from the reported misadminis-tration.
The dose to the whole body was estimated as 0.37 rem (assuming a 15%
1
thyroid tissue uptake) and a thyroid tissue dose of 625 rem.
Patients are often administered radiciodine following surgical or radioactive thyroid removal to check for hidden thyroid tissue.
Cause or Causes - The misadministration was caused by the nuclear medicine technician's misinterpretation of the dose calibrator value.
Actions Taken to Prevent Recurrence Licensee - The nuclear medicine technician was instructed to verify that the dose was within the proper range for a given procedure and to check with the radiologist prior to administration.
i NRC - A telephonic contact was made to the radiologist reporting this misadmin-istration for additional information and assurance that corrective action had been taken.
The incident will be reviewed during the next NRC routine inspec-tion at the hospital.
This item is considered closed for the purposes of this report.
l 1
87-10 Therapeutic Medical Misadministration I
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 an event involving a moderate or more severe impact on public health or safety can be considered an abnormal occurrence.
Date and Place - From April 20-22, 1987, a patient treated on the cobalt-60 teletherapy unit at St. Peter's Medical Center, New Brunswick, New Jersey, re-l ceived a radiotherapy administration of 600 rads to the lumbar spine area, which was not the prescribed treatment site.
1 Nature and Probable Consequences - The putient, diagnosed as having breast cancer with metastasis to the bone, was undergoing treatment to the thoracic l
spine of 3000 rads in 15 fractionated doses of 200 rads each.
She had pre-l viously undergone palliative treatment to the lumbar spine and sacral hip areas l
and still retained the tattoo marks for those treatment fields.
The technol-ogist mistakenly used these tattoos _to position the patient for treatment, j
rather than the tattoos defining the thoracic spine treatment area.
During the course of treatment, the patient was treated as both an in patient i
and out patient.
The misadministration occurred while the patient was an in-patient.
During treatment set up on April 20, 21 and 22, 1987, the patient's gown was only raised far enough to expose the tattoos in the previously treated lumbar spine and sacral hip areas and the technologist involved mistakenly assumed that the lumbar spine tattoos defined the currently prescribed treat-ment field.
Had the technologist raised the gown to expose the entire back, the tattoos in the thoracic spine area would have been seen.
The technologists involved with the patient's treatment noted that the light field was larger than the tattooed field, but assumed the discrepancy was due 2
l
L to skin shif ting and did not notify the supervising technologist, radiation oncologist, or medical physicist. When the patient returned for treatment on April 23, 1987 as an out patient, the gown she wore opened'in the back and the entire back was exposed during treatment set-ups..The technologists ~then l
realized that the patient had been erroneously treated in the lumbar spine area, rather than the prescribed thoracic spine area.
They immediately noti-fied the supervising technologist and radiation oncology physician.
The consequence of this incident was that a patient received an unprescribed dose to the lumbar spine of 600' rads.
The patient's referring physician.and radiotherapist concluded that the dose'would have no detrimental clinical, effect due to the patient's current disease state.
Cause or Causes - The causes are attributed to human errors, including failures to comply with established procedures, i.e.,
1.
The technologist did not expose the patient's entire back during treat-ment set-up; l
2.
The two technologists did: not perform all simulation and set-up proce-dures together; 3.
The technologist who originally simulated, tatto'oed and set up the patient for the initial treatreents did not realize-the error in subsequent set-ups; and 4.
The technologists did not-follow established procedures in the event the light field does not match the patient tattoo marks, which require notify-ing the supervising technologist, the radiation oncologist, or the medical physicist.
Actions Taken to Prevent Recurrence Licensee - The licensee's immediate and planned corrective actions included:
a review of internal policies to evaluate possible changes to prevent further misadministration; a training session with all technologists to review the i
)
incident and internal policies; special training sessions for the technolo-gists involved and review of all their work; and immediate probation of the two technologists.
NRC - A senior Region I NRC inspector conducted'a~ routine inspection of the teletherapy program and review of the misadministration on April 28, 1987.
No violations of NRC regulations were associated with this incident. =An NRC l
medical consultant is reviewing the case.
This item is considered closed for the purposes of-this report.
87-11 Diagnostic Medical Misadministration The following information pertaining to this event is also being reported con.
currently in the Federal Register.
Appendix A (see the general criterion) of 3
}
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 June 3, 1987, NRC received written notification that on May 20, 1987, a patient at the National Institutes of Health, Bethesda, Mary-land, received 120 millicuries of technetium-99m pertechnetate rather than the prescribed radiopharmaceutical, 10 millicuries of gallium-67 citrate.
Nature and Probable Consequences - A patient, scheduled to be injected with 10 millicuries of gallium-67 on May 20, 1987, was administered a radiopharmaceutical on that day and asked to return on May 22 for a scan.
The patient study did not show the typical gallium-67 citrate uptake and an energy spectrum obtained by the gamma camera indicated that technetium-99m had been injected, and not the prescribed gallium-67.
The radiopharmacist reviewed the usage records for May 20, 1987 and discovered s
- r. 3.3 milliliter excess of gallium-67.
The only technetium-99m radiopharmaceu-Lical which could not be accountrd for was technetium-99m pertechnetate.
The radiopharmacist concluded that approximately 120 millicuries of technetium-99m in 3.3 milliliters were withdrawn by mistake by the radiopharmacist and was not assayed for activity in a dose ;alibrator.
This radiopharmaceutical was then dispensed to a physician who administered it to the patient.
The licensee informed the NRC. hat the Chief of the Nuclear Medicine Depart-i ment, the Chief of the Radiopharmacy and the Chief of the Radiation Safety Branch were notified as soon as the misadministration was discovered.
The re-ferring physician was notified by written memorandum.
The patient experienced no adverse effect from this misadministration but received the following un-warranted approximate organ doses:
Tissue Rads Bladder Wall 10.2 Gastrointestinal Tract Stomach Wall 6.1 Upper Large Intestinal Wall 14.4 Lower Large Intestinal Wall 13.2 Red Marrow 2.0 Testes 1.1 Thyroid 15.6 Brain 1.4 Whole Body 1.3 Cause or Causes - The causes are attributed to failure on the part of the radiopharmacist to read labels on stock solutions and the failure to assay for activity before administration to the patient.
Actions Taken to Prevent Recurrence Licensee - All radiopharmacy personnel have been retrained in the existing policies requiring that all labels be checked and all radiopharmaceuticals assayed in a dose calibrator before being dispensed.
4
1 NRC - Region I reviewed this incident during a routine inspection of the licensee on June 8-12, 1987.
One apparent violation, failure to assay the dose before administration to the patient, was associated with this incident.
This item is considered closed for the purposes of this report.
87-12 NRC Order Issued to Remove a Hospital's Radiation Safety Officer 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.
Jate and Place - On June 15, 1987, an Order Modifying License, Effective Imme-diately, was issued to Milford Memorial Hospital, Milford, Delaware (Ref. 1).
The action was based on (1) the falsification of daily constancy checks of the dose calibrator by the licensee's two technologists, and (2) the falsification of records of Radiation Safety Committee (RSC) meetings by the Radiation Safety Officer (R50) for about 15 years.
Nature and Probable Consequences - As part of an NRC inspection at Milford Hospital on December 17, 1986, an NRC inspector reviewed the records of daily constancy checks performed on the dose calibrator.
The inspector observed that during a period of time in 1986, the recorded results of the constancy checks were almost always the same value.
In the presence of the licensee's RSO at the time, the inspector asked one of the two licensee technologists responsible for performing the constancy checks if these tests had been per-formed.
She initially stated that the constancy checks had been performed daily.
However, when the technologist performed the constancy check procedure a short time later in the presence of the inspector and obtained a sigaf ficantly dif-I ferent value than previously recorded, she admitted that she had recorded data j
in the past without actually performing the check.
The other technologist also admitted that she had documented the results of daily constancy checks without having performed the checks.
Subsequent to the inspection, the inves-tigation determined that these records were falsified for the period May 6, 1986 through December 17, 1986.
Although the RSO at that time stated that he had performed an audit of these specific records of constancy checks on November 16, 1986, he did not recog-nize that the records had been falsified.
Apparently, the RSO verified-that records of constancy checks existed', but he did not assess'the accuracy of the records.
During an interview with investigators from the NRC Office of Investigations (01) on May 18, 1987, the Assistant Administrator of the hospital stated that during a review of previous RSC meeting minutes, he noticed that there were minutes for a January 20, 1987 meeting that he neither attended nor was given J
notice of despite his previous instructions to the RSO that he or the Hospital 5
Administrator be present at those meetings.
As a result of his inquiries he had found that these RSC meetings, which were required by the license to be conducted quarterly, had not been conducted for at least the past year, but that the RSO had created a record each quarter to represent that the meetings had occurred.
j The R50 subsequently admitted to 01 investigators that no RSC meetings had been held since approximately 1970, but that false records had been prepared to indicate that the meetings had occurred.
These false records had been pre-1 sented to NRC inspectors during various NRC inmections as evidence that the RSC meetings had occurred, as required.
Specific meeting minutes of the RSC also had been provided to the NRC, in letters dated April 7, and May 14, 1982, to resolve NRC concerns regarding the licensee's application for license renewal dated February 23, 1982.
The consequence of these occurrences was a reduction in the level of safety associated with the use of licensed material by this licensee.
No specific hazard was identified.
Cause or Causes - The causes of these occurrences appear to be a lack of ade-quate management control by the licensee and a lack of integrity on the part of individual members of the licensee's staff.
1 Actions Taken to Prevent Recurrence Licensee - The licensee suspended the RSO (a physician) from his duties as RSO shortly after determining that he had. falsified the records.
Subsequent to the NRC Order, the licensee suspended him from all duties but later permitted him to function in accord with the restriction specified by the NRC Order.
The licensee is conforming to the various provisions of the NRC Order described below.
NRC - The June 15, 1987 Order required:
(1) the removal of the RS0; (2) the suspension of the R50's authorization to independently use or supervise the use of licensed material as currently permitted by the license; (3) the per-formance of monthly independent audits of the licensee's radiation safety pro-gram by an independent party; and (4) a review of the Radiation Safety Program oy the new RSO, correction of deficiencies identified, and certification by the licensee to the NRC that the nuclear medicine program is being operated safely and in accordance with NRC requirements.
A subsequent NRC inspection has shown that the licensee is in compliance with the Order.
This item is considered closed for the purposes of this report.
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l
87-13 Significant Breakdown in Management and Procedural Controls at an Industrial Radiography Licensee 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 a major deficiency in management or pro-cedural controls in major areas can be considered an abnormal occurrence.
Date and Place - On June 17, 1987, the NRC issued an Order Modifying License (Effective Immediately) to United States Testing Company, Inc., Unitech Services Group (USTU), San Leandro, California, which required the licensee to temporarily cease all operations until certain specified corrective actions were taken (Ref. 2).
Nature and Probable Consequences - During an indepth special safety inspection (Ref, 3) on February-10 through June 1, 1987 of USTU in San Leandro, California, it was i etermined that the large radiography firm had committed numerous vio-lations of NRC and Agreement State requirements.
Based on initial findings, a Confirmatory Action Letter (CAL) (Ref. 4) was sent to the licensee regarding --
radiation safety certification of radiographer and radiographer assistants on February 13,198'/.
Upon completion of the full inspection, which covered the licensee's activities from January 1, 1985 to March 1, 1987, NRC issued the previously mentioned Order Modifying License on June 17, 1987 (Ref. 2).
At the time of the inspection, USTU was licensed by the NRC and several Agree-ment States to perform industrial radiography.
The licensee employed approxi-mately 200-300 radiographer, assistant radiographer and trainees, and con--
ducted radiographic operations at 11 locations under NRC jurisdiction and 35 locations under Agreement State jurisdiction.
As the result of the inspection, it was determined that the licensee was (1) allowing individuals to perform' radiography after failing one or more certification examinations, (2) allowing individuals to perform radiography before all training and examinations were
~
r.ompleted, and (3) allowing individuals with expired certifications to perform radiography.
Also, three radiation overexposure and associated evaluations were not reported.
In addition, numerous other radiation safety violations associated with field audits, radiation surveys, inoperable survey instruments, surveillance over high radiation areas, and proper maintenance and equipment inspections were identified at the NRC and Agreement State locations.
Deficient implementation of radiation safety requirements by -this licensee re-suited in the use of radioactive materials by inadequately trained personnel, thereby endangering themselves and co-workers.
In fact, the.NRC inspection-j was initiated by an incident on February 5, 1987, involving the overexposure
^
of inadequately trained personnel (.a radiographer and an assistant radiog-rapher) at a USTU job site in Arizona, an Agreement State.
This event was reported as Agreement State abnormal occurrence AS87-1 (" Breakdown.in Manage-ment and Procedural Controls at an Industrial Radiography-Licensee") in NUREG-0090, Vol'. 10, No. 1 (" Report to Congress on Abnormal Occurrences:
-January-March 1987"),
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Cause or Causes - The root cause appears to be attributed to widespread'disre-l gard for compliance with regulatory requirements.. However, the event remains under investigation by the NRC Office of Investigations, and a complete under-standing of all contributing.causes awaits their report.
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Actions Taken to Prevent Recurrence Licensee - As discussed further below, the licensee has taken, or is taking, I
appropriate actions in response to the february 13, 1987 CAL, and the June 17, l
1987 NRC Order.
N_RC - Initial findings of the NRC indepth special safety inspection indicated that the licensee was using radiographer that had not received required radia-tion safety training.
The CAL issued on February 13, 1987, required a licensee official to verify in writing that assigned radiographer,, by name, have received appropriate training.
Subsequent inspections by the NRC and Agreement States have verified licensee conformance with the CAL.
The Order Modifying License incorporated a two phase action plan.
The licensee is required to enlist a consultant to assist in performing an assessment of program deficiencies and necessary corrective actions.
In the interim, the licensee may continue operations only if very stringent on-site management controls are in place as prescribed by the Order.
This includes assignment of a qualified Radiation Safety Officer (R50) at each major project site or centralized facilities for temporary job sites, with responsibility for radia-tion safety program implementation and the authority to shut down any opera-tions not in regulatory compliance.
The NRC Region V staff has reviewed the training and certification documenta-tion of the new Radiation Safety Officers submitted in compliance with the Order.
All documentation was acceptable.
A reinspection schedule has been established which will examine the actions taken by the licensee, pursuant to the Order, at selected job sites under NRC and Agreement State jurisdiction.
j The consultant's action plan has been evaluated and approved with minor revi-sions by Region V as stipulated in the Order.
On September 25, 1987, NRC Information Notice No. 87-45 ("Recent Safety-Related Violations of NRC Requirements by industrial Radiography Licensees") was issued to all NRC licensees authorized to possess and use sealed sources for industrial radiography to inform them of the event (Ref. 5).
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l Future 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 ircidents or events using the same criterie as the NRC (see Appendix A) and report the events to the NRC for inclusion in this report.
During the second calendar quarter of 1987, an Agreement State reported the following abnormal occurrence to the NRC.
8
l AS87-3 Radiographer Overexposure Appendix A (see Example 1 of "For All Licensees") of this report notes that exposure of the whole body of any individual to 25 rem or more of radiation; j
exposure of the skin of the whole body of any individual to 150 rem or more of radiation; or exposure of the feet, ankles, hands or forearms of any indi-vidual to 375 rem or more of radiation can be considered an abnormal occurrence.
Date and Place - On December 9, 1986, an industrial radiographer and a radiog-rapher's assistant, employed by Northwest X-ray, Idaho Falls, Idaho, received whole body overexposure while performing radiography in a multi-level hot cell at the Chemical Processing Plar' : ; the Idaho Nctional Engineering Labora-tory (INEL) near Idaho Falls.
In addition, the radiographer received an over-l exposure to both hands.
Nature and Probable Consequences - The individuals were using a 50 curie Industrial Nuclear Company (INC) Model 7 iridium-192 source in a Tech / Ops Model 660 exposure device.
The radiographer (who had 5 years of experience) had difficulty in climbing with the survey meter in hand.
It was left with the exposure device and was not used to survey the guide tube between expos-During attempts to lock the exposure device, the assistant noticed the ures.
survey meter was off scale.
The assistant took the survey meter behind a concrete wall to check it out.
The meter was still off scale.
He switched it to the X100 position twc it remained off scale.
He then switched the meter to the battery test position, the off position and finally the X1 position where it read zero. The radiographer and the assistant ansumed that the meter was acting up so they returned to work (the reading remained zero thus confirming the assumption).
The exposure device was moved to another level of the scaffolding (despite the fact that the camera could not be locked) and the guide tube was coiled and un-coiled and otherwise handled by the radiographer.
The radiographer cranked out the source, the assistant picked up the survey meter (it still read zero I
on the X1 scale) and went behind a concrete wall.
When the exposure was com-pleted, they started to enter the room but the survey meter immediately went off scale.
At this point, they decided that a problem existed, and left the area and roped off the entrance to the area.
They removed their anti-contamination coveralls (which had precluded ready access to their pocket do-simeters) and verified that their dosimeters were off scale.
Emergency proce-dures were then implemented.
The source was later verified to be disconnected.
The assistant received a documented exposure of 3.4 rem whole body, and INEL-estimated exposures of 6 rem to the lens of the eye, 5 rem to the left hand, i
and 20 rem to the right hand.
The radiographer received documented exposure of 7.8 rem whole body and INEL-1 estimated exposures of 50 rem to the lens of the left eye, 70 rem to the lens of the right eye, and entrance doses of 2000 and 1700 rem to the left and 1
right hands, respectively.
The licensee's consultant estimated the radiog-rapher's hand entrance doses to be 560 rem and 380 rem for the left and right hand, respectively.
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Both individuals were examined by INEL's Medical Director.
No signs of injury were found.
The assistant was released and the radiographer will be followed medically for several months.
Due to the question of State jurisdiction on a Department of Energy (DOE) facility, DOE was asked to conduct the investigation and forward the results to the State.
The licensee retained a consultant to conduct an independent investigation of the incident.
The findings are as follows:
The radiographer did not follow proper procedures as detailed below.
1.
Proper surveys of the guide tube were not made after each exposure (due to the difficulty in climbing with a survey meter in one hand);
2.
Radiography was performed with a camera that was thought to be malfunc-tioning (it would not lock)-
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Radiography was performed with a survey meter that was thought to be mal-l functioning (it was thought to be " acting up"); and 4.
Pocket dosimeters were not readily available for viewing during the job (they were covered by anti-contamination clothing).
The Tech / Ops representative determined (on site) that the camera and a " dummy" Tech / Ops source worked properly and a disconnect situation could not be produced.
The INC representative determined (on site) that the pigtail connection (drive cable to pigtail) appeared to have greater end play than normal.
It was also determined that, when bending the pigtail and drive cable at a 90-degree angle, the connection (due to end play) could become disconnected.
When a good pigtail (from INC's stock) and drive cable were tested, the disconnect did not occur.
For one of the exposures, the guide tube had an extremely tight radius.
- This, combined with the excessive end play and a sudden jerk of the drive cable upon retraction at the end of the exposure, are likely to have caused the disconnect.
Cause or Causes - The causes of the overexposure were failure to perform proper surveys after each exposure and continued use of a survey meter which was suspecteci to be malfunctioning.
The causes of the disconnect wE.re use of a source with e)cessive end play and use of too tight a radius for the guide tube.
Actions Taken to Prevent Recurrence Licensee - The licensee immediately provided preinstruction to all radiographic personnel on radiation safety and operating procedures with emphasis on sur-veys and emergency procedures.
3 Manufacturer - The sealed source with the excessive end play has been returned to the manufacturer for disposal.
The manufacturer will set up an in-house 10
quality assurance procedure to test all source pigtails for proper connection I
and end play.
State Agency - The State Agency has reviewed the licensee's procedures and l
training programs to assure that they are adequate.
This item is considered closed for the purposes of this report.
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REFERENCES 1.
Letter from James M. Taylor, NRC Deputy Executive Director for Regional Operations, to Glenn Davis, Administrator, Milford Memorial Hospital, for-warding an Order Modifying License, Effective Immediately, License No. 07-14900-01, Docket No. 30-08228, June 15, 1987.*
2.
Letter from James M. Taylor, NRC Deputy Executive Director for Regional Operations, to Gene Basile, President, United States Testing Company, Inc., Unitech Services Group, forwarding an Order Modifying License,' Effec-tive Immediately, Docket No. 30-20402, June 17, 1987.*
3.
Letter from Ross A. Scarano, Director, Division of Radiation Safety and l
Safeguards, NRC Region V, to Gene Basile, President, United States Test-ing Company, Inc., Unitech Services Group, forwarding Inspection Report No. 30-20402/87-01, Docket No. 30-20402, June 16, 1987.*
4.
Confirmatory Action Letter from B. H. Faulkenberry, Deputy Regional Administrator, NRC Region V, to Gene Basile, President, United States Test-ing Company, Inc., Unitech Services Group, Docket No. 30-20402, i
February 13, 1987.*
5.
U.S. Nuclear Regulatory Commission, NRC Information Notice No. 87-45, "Recent Safety-Related Violations of NRC Requirements by Industrial I
l Radiography Licensees," September 25, 1987.*
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- Available in NRC Public Document Room, 1717 H Street, NW., Washington, DC 20555, for public inspection and/or copying.
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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 re-duction 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 detail using these crite-ria are:
For All Licensees 1.
Exposure of the whole body of any individual to 25 rems or more of radia-l 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 S20.403(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 S20.105(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 620.403(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 containing the radioactive material, or (b) release of radioactive mate-rial 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.
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7.
Any substantiated loss of special nuclear material or any substantiated inventory discrepancy which is judged to be significant relative to nor-mally expected performance and which is judged to be caused by theft or diversion or by substantial breakdown of the act. countability 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 670.52(a)).
10.
A major deficiency in design, construction, or operation having safety implications 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),
recurring incidents, and incidents with implications for similar facili-ties (generic incidents), which create major safety concern.
For Commercial Nuclear Power Plants 1.
Exceeding a safety limit of license technical specifications (10 CFR S50.36(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 release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cooling 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 650.36(c)).
2.
A major condition not specifically considered in the safety analysis re-port 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.
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l APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the April through June 1987 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 actions necessary to prevent recurrence of previously reported abnormal occur-The referenced Congressional abnormal occurrence reports below provide rences.
the initial and any updating information on the abnormal occurrences discussed.
The updating provided generally covers events which took place during the report period, thus some information is not current.
Some updating, however, is more current as indicated by the associated event dates.
Open items will be discussed in subsequent reports in the series.
NUCLEAR POWER PLANTS 77-9 Environmental Qualification of Safety-Related Electrical Equipment Inside Containment This abnormal occurrence was originally reported in NUREG-0090-10, " Report to Congress on Abnormal Occurrences:
October - December, 1977" and updated in subsequent reports in this series, i.e., NUREG-0090, Vol. 1, No. 1; Vol. 1, No. 2; Vol. 2, No. 2; Vol. 3, No. 2; Vol. 4, No. 2; Vol. 5, No. 2; Vol. 6, No.
1; Vol. 8, No. 2; and closed out in Vol. 9, No. 4.
It is being reopened to report the following new information; the information is current as of the end of July 1987.
On March 23-27, 1987, an NRC inspection was conducted at Calvert Cliffs Units 1 and 2 to review the status of the licensee's environmental qualifica-tion (EQ) program.
Calvert Cliffs Units 1 and 2 are Combustion Engineering-designed pressurized water reactors, operated by Baltimore Gas and Electric Company (the licensee), and located in Calvert County, Maryland.
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l During the inspection, the NRC found that tape splices identified on electri-cal leads of certain safety-related solenoid valves were not included on the list of electric equipment important to safety required to be maintained in accordance with 10 CFR S50.49, and there was no documentation in a qualifica-tion file to indicate that the splices were qualified to perform their intended function under postulated environmental conditions.
)
In response to the NRC findings, the licensee conducted an immediate followup inspection of their EQ program at Unit 2 which was shut down at the time.
Based on their initial finding of additional unqualified tape splices at I
Unit 2, the licensee voluntarily commenced a shutdown of Unit 1 on April 1, l
1987 to perform an evaluation of the EQ program at both units so as to deter-l l
mine the magnitude of the problem.
During this' evaluation, the licensee iden-j j
tified several other tape splices, as well as several other items of electrical l
{
equipment important to safety, which were either:
(1) not included on the l
list requirer to be maintained in accordance with 10 CFR 650.49; or (2) not environment ly qualified in that there was no documentation available in a i
file to supp rt qualification, or both.
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Licensee corrective actions included:
Immediate replacement of unqualified taped splices, Conducted a comprehensive program to inspect all splices on 10 CFR S50.49 equipment, l
Evaluate the EQ program using both in-house reviews and an independent consultant, Improve EQ program control to prevent recurrence, and Correct identified equipment and record deficiencies.
The NRC conducted a team inspection on May 11-15, 1987, to review the weas of concern and assess readiness for restart; no significant deficiencies were identified.
A more detailed EQ inspection is scheduled following the licensee's funded independent audit of the EQ program.
An enforcement conference was held on May 13, 1987.
Enforcement action is under consideration.
Unless new, significant information becomes available, this item is considered closed for the purposes of this report.
79-3 N_uclear 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 udpated in each subsequent report in this series, i.e., NUREG-0090, Vol. 2, No. 2 through Vol. 10, No. 1.
It is further updated for this report period as follows.
Reactor Building Entries During the second calendar quarter of 1987, 90 entries were made into the TMI-2 reactor building; this brings the total number of entries since the March 1979 accident to 1321.
Defueling operations continued to be the focus of reactor building activities.
Good reactor water clarity was maintained throughout the reporting period.
Significant progress has been made in the removal of sedi-ment deposited during the accident on the reactor building basement floor.
(
Reactor Vessel Defueling Operations j
In April 1987, defueling operations focused on clearing the top of the debris bed in preparation for removing damaged fuel assemblies.
Loose fuel rods were sized and loaded into canisters.
Smaller, loose debris was then loaded into partially filled canisters by a specially designed air lift system in order to j
fill the void spaces and achieve the maximum packing density.
Three damaged assemblies were removed during the month of April.
At the end of the month, defueling operations shifted to the breakup of large pieces of resolidified materials remaining on the debris bed after the core 18
boring operations.
A heavy duty impact funnel was suspended in the reactor vessel above a canister.
The large pieces were then placed in the funnel and an air-operated chisel was used to strike them.
The chisel is essentially a jack hammer with a 12-foot-long bit having the motor located above the water.
Several pieces were easily broken and placed into fuel canisters.
A cavitat-ing water jet apparatus was installed in the reactor building and is available for use.
The jet will be available to collapse the resolidified material on the periphery of the core area (the " donut") and break up pieces of core mate-rial if needed.
A duplicate jet will be installed in the defueling test assembly (DTA) for training purposes.
Defueling operations during May and June 1987 were focused on removing addi-tional damaged fuel assemblies.
Nine assemblies were removed and loaded into canisters in May, and 22 in June.
Through the end of the second calendar quarter of 1987 a total of 34 of the 177 original fuel assemblies have been extracted and loaded into canisters for shipment to the Idaho National Engi-neering Laboratory (INEL).
The primary method of removal was to use a specially designed fuel assembly puller tool to grapple an assembly near its base, loosen it, and slightly raise it to where it could be grappled by its top with a core debris digger tool.
The core debris digger was used to lift the assembly and place it in a canister.
Loose fuel pins were cleared from the grid which supports fuel assemblies from below.
Eight exposed grid loca-tions were plugged to prevent additional debris from moving into the lower core support assembly (CSA) and the lower vessel head area.
The grid plugs are flat metal plates with lifting attachments welded on top and a pipe stub weldeo onto the bottom that extends down into the grid flow hole.
A video inspection of the cleared core region was performed in June.
This inspection confi rmed the existence of several small holes indicating damage to the core former wall.
The source of the damage is not yet known but it appears that molten core material burned or melted through the core former wall in at least this location.
The core former wall is an internal structure.
No evi-dence of damage to the reactor vessel itself has been identified.
The water clarity within the reactor vessel continued to be good.
The "B" train of the defueling water cleanup system (DWCS) was put into operation with a reclaimed filter canister installed.
This second train of the system will provide increased performance and flexibility in processing reactor coolant I
water.
The temporary reactor vessel filtration system, previously used to help gain water clarity, was disconnected and removed from its processing location to a high radiation storage area in the reactor building.
The licensee is evaluating methods for disassembling and defueling the regions beneath the grid plate (i.e., core support assembly).
EG&G has successfully i
completed testing of a core boring machine and various drilling bits to simu-i late drilling holes through the lower CSA.
Additional tests will be performed to determine if the drilling machine can be used in dismantling the lower CSA.
A plasma arc torch was received onsite and pre-operational testing in air is I
underway.
The plasma arc torch may be used to cut up the lower CSA for defuel-ing.
A manipulating arm which will operate with the plasma arc torch was set up and pre-operational tests were performed in air.
The DTA in the turbine building has been modified to include a partial mockup of the lower CSA.
This will allow the plasma arc torch and the manipulating arm to be installed for l
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19
training and checkout.
Other alternatives being considered are electron dis-charge machining (EDM), and metal disintegration machining (MDM) processes.
These devices are under evaluation.
Cask and Liner Shipments During the second calendar quarter of 1987, there was one offsite shipment of TMI-2 core debris to INEL.
Two loaded shipping casks, each holding seven defueling canisters, were transferred by rail during the reporting period.
Through June 1987, approximately 88,100 lbs. of core debris (30% of the total estimated quantity) had been shipped.
Seven EPICOR liners and three 4' x 4' liners containing dry active waste were also shipped offsite during the report-ing period.
In May 1987, the first of four cuno filters was shipped to INEL for disposal as abnormal waste.
The cuno filters were originally used in the submerged de-mineralized system (SDS) to filter reactor building basement water before the use of sand filters.
Through an agreement with the DOE, the filters have been shipped to INEL for research and storage.
EPICOR II/ Submerged Demineralized System (SDS) Processing Through June 1987, a total of 4,505,022 gallons of water have been processed through the SDS and a total of 3,490,376 gallons have been processed through the EPICOR II system.
For the reporting period, approximately 60,000 gallons and 175,000 gallons were processed by the SDS and EPICOR 11 systems, respectively.
Decontamination / Dose Reduction Activities Decontamination activities continued in the TMI-2 auxiliary and fuel handling building (AFHB) during the second calendar quarter of 1987.
These activities centered around steam vacuum cleaning, scabbling, and hands-on decontamination of AFHB cubicles.
As of the end of June 1987, 94 cubicles out of 143 have been decontaminated to the licensee's established end point levels.
Scabbling of the highly contaminated seal injection valve room continued during the reporting period.
Sediment removal from the auxiliary building sump was com-pleted during the second calendar quarter.
The successful operation reduced general area radiation dose rates from over 2 R/hr before sediment removal to approximately 350 mR/hr.
Remote reconnaissance vehicle No. 1 (RRV-1) continued to remove sediment from accessible areas of the reactor building basement.
The sediment was transferred to the spent resin storage tanks in the auxiliary building basement and will later be solidified.
The top flight of the open stairwell between El. 305' and El. 282' was removed to permit robotic access to previously unexamined areas.
RRV-3 was lowered into this region to remove sediment and debris.
Proposal to Dispose of Accident-Generated Water In June, the NRC staff issued the final supplement to the " Programmatic Environ-mental Impact Statement" (PEIS) related to the disposal of accioent generated water (Ref. B-1).
A draft supplement had been circulated to allow public 20
3 comment. After considering comments on the droft supplement, the NRC staff concluded that the licensee's proposal to evaporate accident-generated water is an acceptaP disposal plan. The staff found that evaporation of the water at the TMI siu, followed by solidification and disposal of the remaining low-level radioactive solids will not significantly affect the cuality of the human environment. The staff has also concluded that any adverse impacts from the disposal program are outweighed by its benefits.
TMI-2 Advisory Panel The Advisory Panel for the Decontamination of Three Milt Island Unit 2 (Panel) met with the NRC Commissioners on April 16, 1987, in Washington, DC.
They discussed the issue of the TMI-2 accident-generated water disposal. The Panel was asked to provide the Commission with its view of when the water disposal issue should be ultimately resolved.
On June 10, 1987, the Panel met in Lancaster, Pa. At the meeting, Mr. J. Roth provioed a short summary of the Panel's meeting with the NRC Commissioners.
GPU Nuclear (the licensee) also provided a status report on the cleanup.
Dur-ing the remainder of the meeting the Panel received public comment on the issue of the NRC's timing of a decision on the disposal of the accident-generated water. As a result of the discussion, Chairman Morris indicated he would inform the Commissicr. that the Panel deemed it inappropriate to provide spe-cific advice on the timing issue. The Panel felt that the staff and Commission should take as much time as necessary for a careful evaluation of the issue.
The Environtrental Protection Agency (EPA) also presented information on its l
plans for rcducing environmental monitoring activities in the TMI area.
Future reports will be made as appropriate.
81-7 Blockage of Coolant Flow to Safety-Related Systems and Components This abnormal occurrence was criginally reported in NUREG-0090, Vol. 4, No. 4,
" Report to Congress on Abnormal Occurrences: October-December 1981."
It was updated, and closed out, in NUREG-0090, Vol. 7, No.1, " Report to Congress on Abnormal Occurrences: January-March 1984."
It is being reopened to report the following information.
j Service water systems (SWSs) of nuclear power pl6nts are typically open cycle systems. An "open-cycle SWS" implies that water is pumped directly from a river, cooling pcnd, or ocean body, into the service water intake structure.
- However, along with water with open cycle systcms, mud, silt, sand, algae, bacteria, fungi and aquatic organisms may also be pumped into the SWSs. Although gratings, screens, and filtcrs block out many of the impuritics, fouling of SWSs is an existing problem that must be satisfactorily resolveo.
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Safety-related SWSs, which already have separate and redundant piping systems, share the same intake structure and ultimate heat sink.
Thus, they share a potential for common mode failure due to service water impurities.
To deal with this concern, the NRC staff developed Generic Issue (GI)-51, " Improving Reliability of Open Cycle Service Water Systems."
Originally, GI-51 was concerned with biofouling (principally by fresh water Asiatic clams, and by marine mussels and oysters).
In recent years, biofouling-i related SWS problems have become less prevalent, but fouling by silt / mud and I
corrosion products persist in the industry.
Recent problems at McGuire l
(October 1985), Farley 1 (August 1986), Grand Gulf (August 1986), Robinson (October-December,1986), Not th Anna 1/2 (February 1987), Oconee 1/2/3 (March-April 1987), and Rancho Seco (June 1987) illustrate silt-related SWS clogging problems.
Information Notice No. 86-96, " Heat Exchanger Fouling Can Cause Inadequate Operability of Service Water Systems," dated November 20, 1986, was issued because of the problems at Farley and McGuire (Ref. B-2).
In 1985, GI-51 was modified to include fouling by silt, mud, and corrosion and the potential for their interaction with biofouling.
The NRC Office of
(
Nuclear Regulatory Research (RES) is coordinating the effort and has produced 1
two recent reports on the problem, NUREG/CR-4626, Volumes 1 and 2 (Ref. B-3).
Volume 2 is specific to silt and corrosion.
In addition, RES has modified the l
contract work so that the remaining technical analyses will be produced in a more timely manner.
This will allow work to begin on the regulatory analysis that will lead to recommendations for improving SWS reliability.
It is antic-ipated that a draft regulatory analysis will be available during the second quarter of FY 1988, l
The following discussion of the problem that affected all three Oconee units illustrates the significance and impact of silt deposition.
In early April 1987, Duke Power Company (the licensee,. informed the NRC staff that an analysis of test data for Oconee Units 1, 2 and 3 indicated that there was recent foul-ing in the low pressure service water (LPSW) system (lake water) side of the reactor building cooling units (RBCU) and low pressure injection (LPI) coolers.
The Oconee units are Babcock & Wilcox-designed pressurized water reactors located in Oconee County, South Carolina.
This fouling had resulted in an inability to transfer total design-basis accident (the loss-of-coolant accident, LOCA) heat loads.
Consequently, the licensee was required to reduce power levels in Oconee Units 1 and 2 to a maximum of 91.5% and 81.7%, respectively, to match accident heat transfer requirements with the capability of the de-graded heat exchangers.
Oconee Unit 3 was shut down at the time; its affected heat exchangers were cleaned and performance tested oefore the unit resumed full power operation.
The RBCUs provide the design heat removal capacity following a loss-of-coolant accident.
All three coolers operate continuously and circulate the steam-air mixture past the cooling tubes to transfer heat from the containment atmosphere to the LPSW, which is passed through the cooler tubes.
The LPI system in the recirculation mode provides for long-term cooling by injecting water from the reactor building sump to the core.
Heat is transferred through LPI coolers to the LPSW systems.
In addition to the LPI and RBCU coolers, the LPSW systems cool the high pressure injection pump motor bearing coolers, the motor-driven 22
emergency feedwater pump motor air coolers, and the turbine-driven emergency feedwater pump turbine bearing oil coolers.
All of these could have been ad-versely impacted by the fouling.
Further progress on the resolution of this problem will be reported biannually in NUREG-0933, "A Prioritization of Generic Safety Issues" (Ref. B-4).
This item is considered closed for the purposes of this report.
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," and up-dated in Vol. 9, No. 1; Vol. 9, No. 2; and Vol. 9, No. 3.
It is further up-dated through July 1987 (except as noted otherwise), as follows.
On February 9, 1987, the Commission established the Office of Special Projects (OSP) to, among other things, efficiently and aggressively (1) resolve the problems causing the shutdown of Tennessee Valley Authority's (TVA's) Sequoyah and Browns Ferry facilities and prevent their recurrence and (2) evaluate the problems that prevented the licensing of TVA's Watts Bar facility to ensure that TVA is taking appropriate measures to bring the facility into compliance with the Commission's regulatory requirements.
The staff has identified a number of major issues requiring resolution prior to the restart of any of the TVA reactors and has provided periodic status reports to the Commission, the most recent of which was issued July 16, 1987.
4 TVA submitted Revision 3 of the Corporate Nuclear Performance Plan on Decem-ber 4, 1986 and Revision 4 on March 26, 1987.
Revision 2 of the Sequoyah Plan was submitted on July 2, 1987.
Revision 1 of the Browns Ferry Performance Plan was submitted July 1, 1987.
Corporate Activities The staff review of the Corporate Nuclear Performance Plan is complete.
The staff issued its conclusions by letter dated July 28, 1987, to the TVA Board of Directors and in NUREG-1232, Volume 1, " Safety Evaluation Report (SER) on Tennessee Valley Authority Revised Corporate Nuclear Performance Plan,"
published July 1987 (Ref. B-5).
The staff concluded that TVA has acceptably addressed the corporate-level concerns raised by the staff in its 10 CFR S50.54(f) letter to TVA dated September 17, 1985 (Ref. B-6).
The staff included an evaluation of the ACRS concerns, provided in the August 12, 1986 report, and TVA's September 26, 1986 response to these concerns, in its SER.
The staff will continue to monitor TVA's corporate-level nuclear activities to determine if past problems recur, and to ensure that TVA continues to seek out root causes of problems and continues to conduct critical independent reviews or self-reviews of the TVA nuclear organization.
23
The staff requested that TVA provide notification at least 30 days before making any permanent change in this organization or permanent replacement of senior managers, including the Site Directors.
The staff is continuing its evaluation of TVA's handling of the harassment and intimidation (H&I) issues.
The staff is currently planning an in-depth review of H&I concerns.
This will include a review of (1) the results of the Inspector General's (IG's) investigation reports and actions taken by TVA in those cases in which the allegations were substantiated, and (2) the TVA actions to program-matically eliminate those factors that contributed to the H&I environment.
Sequoyah TVA submitted Revision 2 of the Sequoyah Nuclear Performance Plan on July 2, 1987.
TVA has scheduled Sequoyah Unit 2 (SQN 2) to be restarted in late 1987 and Sequoyah Unit 1 (SQN 1) restarted approximately six months after Unit 2.
Since January 1987, the staff and TVA have made substantial progress on the resolution of key issues affecting SQN 2.
In the discussion which follows, the staf f presents four areas:
(1) the schedule for SQN 2 restart, (2) the Inte-grated Design Inspection (IDI) effort, (3) environmental qualification of elec-trical equipment (EQ), and (4) other significant issues.
The staff discusses the significant issues which appear at this time to have the greatest potential to affect SQN 2 restart:
(a) the adequacy of cable installation, (b) TVA regen-eration of civil engineering design calculations, (c) the restart test program, (d) the fuse replacement program, and (e) the readiness of the plant staff to resume operations.
(1) Schedule for SQN 2 Restart The staff approved, in a June 9, 1987 letter to TVA, the restart criteria used by TVA to identify issues that must be resolved prior to Sequoyah and Browns ferry restarts.
Having approved the restart criteria, OSP identified the staff activities necessary before the restart of SQN 2.
These were identified in the OSP Sequoyah Activities Schedule sent to the Commission on June 23, 1987.
Revision 1 of the schedule estimated commencement of non-nuclear heatup of SQN 2 in late Summer 1987, and criticality no earlier than November 6, 1987.
However, the IDI review has identified a number of technical issues which must be resolved by TVA prior to restart of SQN 2.
Pending resolution of these and other issues, as of the end of October 1987 the NRC staff currently estimates criticality no earlier than March 1988.
A Commission decision on criticality will not be requested until after the satisfactory completion of the IDI issues, as well as the imple-mentation of corrective actions identified by other technical programs.
(2) The Integrated Design Inspection The staff has been closely monitoring TVA's efforts to identify and resolve the many issues that need to be addressed prior to Sequoyah restart.
Although the efforts have been quite extensive, the identification of issues has been an evolving process and somewhat fragmented.
In examining these issues, the staff 24
l 1
i observed that TVA's engineering and design review efforts to date have concen-trated on separate programs looking horizontally across the safety systems at Sequoyah and focused on the specific concerns raised.
TVA's efforts have not included a vertical review through one or more safety systems that would pro-i vide atsurance that all major design and construction problems have been iden-l tified and resolved prior to SQN 2 restart.
Accordingly, the staff decided that an independent design / construction verifica-tion including significant aspects of the interactions and interfaces through-out design, engineering and construction, of at least one safety-related system should be undertaken.
The staff gave TVA the choice of hiring an independent contractor or having the NRC undertake the task.
1VA stated that an independent contractor would take at least nine months to complete the effort, unreasonably delaying Sequoyah restart, and it could not commit to such a program at that time.
Therefore, the NRC staff elected to perform the review itself.
The staff chose the Essential Raw Cooling Water System for the " vertical slice" review.
The entrance meeting for this effort was held on July 8, 1987 and the report documenting the inspection review is expected to be completed in November 1987.
The review has identified a number of technical issues that must be resolved by TVA prior to restart.
Additional information will be provided in a subsequent abnormal occurrence quarterly report after the IDI inspection report is issued.
(3) Environmental Qualification of Electrical Equipment The main issue which precipitated the shutdown of the Sequoyah units in 1985 was the environmental qualification of electrical equipment (EQ).
The staff has completed a comprehensive review of the Sequoyah EQ program including a number of inspections of its implementation and concluded that, after satisfactory completion of identified plant modifications. the program will be in compliance with the requirements of 10 CFR 550.49.
These modifications will be completed and completion will have been certified by TVA prior to restart.
The staff now considers Sequoyah's EQ program to be one of the better programs in the nuclear power industry.
(4) Other Significant Issues 1
Additional significant technical and operational issues which have the greatest potential to affect SQN 2 restart are discussed below:
(a) Adequacy of Cable Installation Concerns were raised by TVA employees regarding cable installation.
The q
concerns were that integrity of the insulation of the cables had been I
degraded in three installation situations:
cable pullbys, cable jamming, j
and long vertical cable runs supported at the top.
To resolve these concerns, a testing program was developed by TVA and submitted on April 8, 1987.
Based on test results and concerns regarding the test conditions, TVA undertook an effort to reevaluate this proposed test program and has submitted a revised program dated July 31, 1987.
The principal differences between the two test programs include revisions to test parameters, change in acceptance criteria and test conditions.
The staff is reviewing this revised program.
If replacement of cables is required, this could become the pacing item for SQN 2 restart.
1 25
"> m (b) Desiqn Calculations Review TVA instituted a review of essential design calculations to verify that the calculations existed and were adequate to support the design of Sequoyah.
TVA completed an initial review of a sample of engineering calculations in the electrical, nuclear, mechanical and civil engineering areas.
Although no significant problems were identified in the nuclear and mechanical engineering calculations, significant problems were iden-tified in the electrical and civil areas.
TVA regenerated all electrical calculations.
In the civil area, about 5000 pipe support calculations were determined to be missing.
The staff has required TVA to regenerate the missing civil calculations, to the extent practicable, prior to re-start.
The staff will inspect the regeneration of the calculations and assumptions used in the electrical calculations.
This may become a pacing item for restart.
(c) Restart Test Program In response to all of the programs under way since the shutdown of the Sequoyah plant, the staff is requiring a comprehensive restart test pro-gram to ensure that the plant safety systems are functional before restart of Sequoyah.
A summary description of the program and a listing and schedule of required testing identified by this program were submitted on May 26 and July 6, 1987, respectively.
Inspections have been conducted on this program and the staff is in general agreement with the TVA program.
The staff will closely follow the restart test program.
(d) Fuse Replacement
]
There was widespread use of Bussman fuses as primary protection for safety-related circuits at Sequoyah.
The certification for these fuses could not be verified.
Problems with replacement Bussman MIS-5 fuses led to use of l
Littelfuse FLAS-5 fuses.
Recently, it was discovered that 69 (Ref. B-7) l of the new Littelfuse fuses had failed at Sequoyah.
This matter is under i
review by TVA and the staff.
(e) Operational Readiness In addition to completion of other ongoing technical programs, actions are 4
required to assure that the Sequoyah plant and its management are ready for plant restart.
The TVA staff is required to 1) establish and implement final verification programs to assure all actions required for restart are implemented, and 2) verify that all commitments and regulations are satis-fied.
TVA has established general guidance for operational readiness veri-fication and has conducted an interim operational readiness review.
Actions to establish site implementing documents for operational readiness reviews must still be finalized.
The NRC will independently assess TVA's readiness in these areas.
For the past several months, TVA management and NRC attention has been concentrated on the review and root cause evaluation of several personnel 26 i
errors associated with recent operational and testing events. This eval-uation identified several examples of failure to follow proceoures,~in-adequate procedures, and an overall lack of control over testing evolutions including system and equipment status.
These problems are currently being considered for escalated enforcement, and TVA's corrective actions for this issue are turrently being evaluated by the staff.
Browns Ferry TVA submitted Revision 1 to the Browns Ferry Nuclear Performance Plan on-July 1, 1987. TVA has scheduled restart for Browne Ferry Unit 2 for mid-1988.
Units 1 and 3 would require additional time prior to restart.
Since January 1987, the staff and TVA have made some progress toward the reso-lution of key issues affecting Browns Ferry Unit 2, and a detailed discussion of these issues will be provided in future reports.
Watts Bar On June 30, 1987, the staff issued an order extending the construction comple-tion date of Watts Bar Unit 1 to September 1, 1988 and fo'r Watts'Bar Unit 2 to January 1, 1990.
TVA requested these dates in its January 29, 1987 letter Pacing items are expected to be TVA completion of the Design Baseline and.
.i Licensing Verification Program, reanalysis of piping and supports, and resolu-'
tion of employee concerns. - Detailed schedules for resolving these issues have not been received from TVA and'the submittal date for the Watts Bar portion of the Nuclear Performance Plan remains uncertain.
Detailed discussion of these issues will be provided in future reports.
Future reports will be made as appropriate.
87-1 NRC Order Suspends Power Operations of Peach Bottom Facility Due to Inattentiveness of the Control Room Staff This abnormal occurrence was originally reported in NUREG-0090, Vol.10, No.1,
" Report to Congress on Abnormal Occurrence:
January-March 1987." It is up-dated through mid-August 1987 as follows.
Since the original report the licensee, Philadelphia Electric Company,' has taken several actions.
Their internal security department has conducted an, investigation of the inattentiveness to duty issue and also covered related I
concerns such as the reading of non-technical material and playing of video games in the control room.
The results of this investigation, along with the Management Analysis Company' investigation, has been utilized to develop a restart plan.
There have been several meetings during which the licensee has briefed NRC man-agement on the status of the investigations and on the development of the recovery plan.
The recovery plan, entitled the " Commitment' to Excellence Action Plan," was sent to the NRC on August 7, 1987 for review.
j 27 i
'l
The licensee has also commenced a training program for the licensed operators which is intended to address the issues which led to the March 31, 1987 shut-down order.
The adequacy of the restart plan must be determined by the NRC and properly implemented by the licensee before the restart of either unit will be permitted.
Future reports will be made as appropriate.
OTHER NRC LICENSEES 86-25 Suspension of License for Servicing Teletherapy and Radiography Units This abnormal occurrence, involving Advanced Medical Systems (AMS), Inc. of Geneva, Ohio, was originally reported in NUREG-0090, Vol. 9, No. 4, " Report to Congress on Abnormal Occurrences:
October-December 1986."
It is updated, through July 1987, as follows.
The previous report mentioned that the NRC issued an order to the licensee on October 10, 1986 (Ref. B-8), suspending certain NRC-licensed service activi-ties.
The licensee had been using untrained and unqualified employees to service cobalt-60 teletherapy units.
On February 2, 1987, the NRC relaxed the order based (1) on the licensee's stated corrective actions, and (2) some addi-tional requirements imposed by the NRC.
The circumstances associated with this event remain under investigation.
Meanwhile, a separate issue pertaining to the licensee's teletherapy source fabrication facility (located at 1020 London Road in Cleveland, Ohio) has re-suited in the NRC issuing to the licensee on July 23, 1987 an Order Modifying License, Effective Immediately, and a Demand for Inforraation (Ref. B-9).
The l
issue involves excessive radioactive contamination and radiation levels at the licensee's London Road facility, which results in a significant potential for unnecessary radiation exposures for workers at this facility.
Since the first quarter of 1986, the NRC has made repeated efforts (including license amendments) to get the licensee to take steps to initiate meaningful decontamination efforts at the facility and modify the facility to minimize contamination.
AMS has failed to take such steps and has indicated that it will not begin such steps until March 1988 at the earliest, citing lack of available profits from its business due to the NRC suspension of its service license from October 10, 1986 to February 2, 1987.
Meanwhile, corrective efforts have been minimal and contamination and radiation levels remain ex-cessive and are increasing.
Therefore, the July 23, 1987 NRC Order Modifying License, Effective Immediately, requires the licensee to commence decontamination of the London Road facility by August 31, 1987, and to commence the required redesign, reconstruction, and upgrading of the facility, also by August 31, 1987.
The Demand for Infor-mation requires AMS to submit certain financial information in order for the i
NRC to determine whether it can have reasonable assurance that in the future l
28
1 l
the licensee will conduct its activities in accordance with the Commission's requirements and expeditiously conduct required decontamination, redesign, reconstruction, and upgrading of its facilities and programs.
Under the terms of the Order, the licensee or any other person who has an interest adversely affected by this Order may, within 20 days of the date of receipt of the Order, file a written answer under oath or affirmation and may also request a hearing.
Future reports will be made as appropriate.
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29
1 APPENDIX C 1
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 major reduction in the level of protection provided for public health or safety; therefore, they are not reportable as abnormal occurrences, t
1 Occasionally, this Appendix will include events involving exposures to very small areas of the skin (one square centimeter or less) which technically exceed the exposures shown in Appendix A (see Example 1 of "For All Licensees") of this report.
The radiobiological literature indicates that an overexposure to a small area of skin (less than one square centimeter) would have much less health significance than a similar dose to larger areas of the body; conse-quently, such exposures would generally not be considered a major reduction in public health or safety (the general abnormal occurrence criterion) and there-fore not reportable as abnormal occurrences.
However, all such events, together with the circumstances associated with the events, are re' viewed individually to determine their relative significance, and if warranted, will be reported as abnormal occurrences.
1.
Therapeutic Medical Misadministration On January 13, 1987, a radiation therapy technician working under the license issued to Valley View Regional Hospital, Ada, Oklahoma, telephoned the head-l quarters duty officer to report a possible cobalt-60 therapeutic misadministra-tion.
The incident occurred on December 29, 1986, when a patient undergoing radiation therapy was administered a treatment dose with the source head rotat-ing through part of the treatment dose instead of the prescribed fixed head position.
There was also some damage reported to the teletherapy unit when the source head collided with the patient table.
The incident occurred as a result of failure of the technician to switch the 1
therapy unit out of the rotational mode following a previous exposure.
The rotational exposure to the patient was terminated by the technician after 12 seconds had elapsed, delivering a 25 rad dose to the patient.
Although the delivered dose was less than a 10 percent error in the 5000 rad total treat-ment dose, the incident was a misadministration because it involved a route of administration other than that intended by the prescribing physician.
Since the licensee did not recognize the incident as a misadministration, it was not initially reported within the required time frame.
No adverse effects were suffered by the patient although had the patient been immobile or sedated, physical injury could have occurred from the rotating source head.
The incident was caused by training deficiencies which contributed to inatten-tiveness on the part of the technician.
Failure to properly recognize and report the misadministration was the result cf licensee confusion over the definitions of therapy misadministration as specified in 10 CFR Part 35.
31 l
1 The licensee has implemented retraining in areas related to therapy procedures, emergencies, and reporting requirements.
A procedure manual has been developed and implemented.
In addition, the licensee is considering hiring an additional therapy technician and is investigating a collision circuit for the therapy unit.
An NRC inspector was sent to the hospital to investigate this matter and to perform a full initial inspection of the teletherapy. license.
Also, NRC con-ducted an enforcement conference with the licensee on May 21, 1987.
At the conference, NRC explained and clarified for the licensee the basis for this 1
incident being characterized as a misadministration.
In addition, other acti-vities are underway or being considered by NRC to inform other hospitals of the potential for similar incidents.
l 2.
Overexposure of a Technologist's Thumb l
l
\\
I l
On April 8, 1987, a technologist employed by E.I. duPont de Nemours & Company, Inc., Medical Products Department, Billerica, Massachusetts, received an esti-l mated 658 rem dose to approximately one square centimeter of skin on his left I
thumb.
l The technologist was performing a series of quality control tests on the molybdenum-99 stock solution used to prepare technetium-99m generators.
These generators are utilized in hospitals for diagnostic nuclear medicine studies.
The tests require that several 1.3 milliliter samples, each containing approx-imately 450 millicuries of molybdenum-99, be removed from a shielded hot cell and transported into an adjoining room where the tests are performed.
l While removing a screw cap vial containing one of the molybdenum-99 samples from l
the hot cell, the vial was accidentally dropped on the floor.
A small amount of solution was apparently ejected from the vial when it struck the floor, result-ing in the cap and the floor becoming contaminated with molybdenum-99.
The technologist responsible for performing the tests picked the vial up with for-ceps, placed it into a lead shielded container, and transported it to the test area.
In the test area, contrary to company procedures, the technologist used his gloved left hand to unscrew the vial cap to perform the test, rather than using the required remote handling tools.
When the test was completed, he again used his gloved left hand to screw the cap back on the vial.
During this process the skin on his left thumb became contaminated with approximately 126 micro-curies of molybdenum-99, due to a small hole in the glove.
Subsequent decon-tamination efforts reduced the fixed molybdenum-99 contaminating his skin to 24 microcuries.
The company's final dose assessment of 658 rem to one square centimeter of the basal layer of skin on the individual's thumb has been reviewed by the NRC staff and is considered acceptably accurate.
There are probably no consequences of the exposure to the technologist's left thumb, according to the treating physician.
An NRC medical consultant is reviewing the exposure and the treating physician's assessment.
32
While there are other contributing factors, the principal causes of the inci-dent were the failure of the technologist to use required remote handling tools when handling the unshielded vial and the failure to monitor a potentially contaminated vial before handling.
The Radiation Protection Officer of the facility performed an investigation of the incident.
As a result of the investigation, the company revised and clari-fied the applicable standard operating procedures relative to precautions and protocol for handling the sampling vials.
The Operations Manager of the facil-ity issued a memorandum to all Supervisors requiring them to review the inves-tigation report of the incident and the radiation protection procedures for handling radioactive materials with all employees.
Further, licensee management personnel have initiated an extensive review of all radioactive material handl-ing procedures to ensure adequacy and to assure they are distributed to all operating personnel.
An inspection was conducted by Region I staff on April 9, 1987.
Two violations were identified pertaining to exposure in excess of regulatory limits and fail-ure to follow procedures.
An NRC medical consultant was assigned to assess the clinical aspects of the exposure.
An enforcement conference was held with the licensee on May 8, 1987.
On May 28, 1987 a $12,500 fine was proposed (Ref. C-1),
which was subsequently paid by the licensee.
3.
NRC Augmented Inspection Team Sent to Diablo Canyon Nuclear Power Plant On April 13, 1987, an NRC Augmented Inspection Team (AIT) was sent to the Diablo Canyon Nuclear Power Plant after an event at Unit 2 on April 10, 1987 raised some concern regarding safe operation of the plant.
Diablo Canyon, a two-unit site operated by the Pacific Gas and Electric Company, utilizes Westinghouse-designed pressurized water reactors.
The plant is located in San Luis Obispo County, California.
The AIT conducted a special safety inspection at the Diablo Canyon site from April 15 through April 21, 1987, to examine the circumstances of the event in detail.
The report of the inspection findings, issued as NUREG-1269 (Ref.
C-2), was forwarded to the licensee on June 19, 1987 (Ref. C-3).
On April 10, 1987, seven days after being shut down for its first refueling, Unit 2 experienced a loss of Residual Heat Removal (RHR) system flow for a period of approximately one and one-half hours.
The RHR system is designed to remove fission product decay heat from the reactor core.
]
The event occurred while the Reactor Coolant System (RCS) was depressurized and the reactor vessel water level was drained to approximate mid-level of the hot leg piping.
The reactor containment building equipment hatch (a large, approx-imately 20-foot diameter bolted steel closure) had been removed prior to the event, and plant personnel were in the process of removing the primary (RCS) side access covers (manways) on the steam generators to gain access to the steam generator channel head areas.
Bolts securing one manway cover had been loosened; however, the cover had not been removed.
Therefore, only a small pathway for kakage of RCS water existed via the manway covers.
A one-inch 33
diameter vent valve on the reactor vessel head was open, with tygon (plastic) tubing connected to it.
The RCS water temperature increased from 87 F to bulk boiling conditions and the RCS was subsequently pressurized to approximately 7 to 10 psig during the event.
Although airborne radioactivity increased inside the containment build-ing during the event and approximately 30 to 50 gallons of RCS water leaked l
from the loosened steam generator access cover into the containment building, no significant quantity of radioactivity was released to the environment.
Radiation-exposures to plant personnel were determined to be well within NRC limits.
The AIT concluded that the mode of operation at the time of the event presented unusual challenges to the plant operators.
During this mode of operation, reactor water level must be maintained within a narrow range to allow draining of the steam generator tubes and subsequent access of personnel into the steam generator channel head areas, while maintaining the water level high enough to prevent air entrainment (due to vortexing) at the inlet to the RHR pumps.
The loss of RHR flow which initiated the event on April 10, 1987 resulted from air entrainment at the inlet to the RHR pumps when water level was reduced, due to unexpected leakage from the'RCS, to the point where vortexing occurred and plant operators turned off the pumps.
An RHR pump was restarted after water was added to the RCS from the refueling water storage tank to raise the RCS water level above that where vortexing
)
would occur.
The plant operators were delayed in their decision to add water to the RCS due to uncertainty as to the status of work in progress to remove the steam generator manway covers.
Approximately one and one-half hours after the RHR pumps had been shut off, it was determined that personnel had not entered the steam generator channel head area, water was added to the RCS, and an RHR pump was restarted - thus, terminating the loss of RHR flow event.
Unexpected leakage from the RCS occurred when a plant engineer opene'd a valve to commence a test on a containment building penetration without notifying operators in the control room.
Valves intended to isolate the penetration l
line from the RCS during the test had not been closed securely.
Thus, water from the volume control tank drained unexpectedly through the valve which was opened by the test engineer into the reactor coolant drain tank.
The AIT identified concerns regarding the adequacy of RCS water temperature and level instrumentccion during the event.
Upon loss of RHR flow, the temperature of the RCS was unknown to the plant operators since core outlet thermocouple within the reactor vessel had been disconnected'in preparation for the removal of the reactor vessel head.
The plant operators expected the RCS water temperature to increase at approximately one degree per minute upon loss of RHR cooling, whereas the temperature actually increased at a rate of approximately 2.7 degrees per minute.
The temporary reactor vessel water level instrumentation appeared to indicate a higher than actual level due to air entrainment and hydraulic. effects not understood by the plant operators.
Concerns were also identified regarding the adequacy of normal and emergency operating procedures for the mode of operation at the time of the event.
34
The AIT concluded that many of the concerns identified during the special inspection of this event are potentially generic to other pressurized water reactors licensed by the NRC.
The AIT also determined that the mode of operation of the plant at the time of the event was not described in the Final Safety Analysis Report, and therefore, had not been formally evaluated by the NRC staff.
Pacific Gas and Electric Company committed to several facility and procedural improvements prior to a return to the mode of operation at the time of this event.
The containment building equipment hatch was replaced and other major pathways to the environment, such as the personnel air lock, were closed.
Two core exit thermocouple were reconnected.
A second reactor vessel level instrumentation channel, redundant to that installed on the cold leg piping, was installed on the hot leg.
Revisions were made to normal and emergency operating procedures to provide added precautions and recovery actions to be taken in the event RCS level decreases below acceptable values.
In addition, the scope of work to be performed on systems connected to the RCS was restricted to those items which did not communicate between the containment building and the environment or have the potential to reduce RCS inventory.
These commitments were confirmed in a letter to the NRC Region V office on May 4, 1987 (Ref. C-4).
On May 27, 1987, the NRC issued NRC Information Notice No. 87-23, "Luss of Decay Heat Removal During Low Reactor Coolant Level Operation," to all holders of an operating license or construction permit for pressurized water reactor facilities (Ref. C-5).
This Information Notice discussed the circumstances of the event at Diablo Canyon, Unit 2, on April 10, 1987, as well as several other similar events at other pressurized water reactors.
The actions taken by the Diablo Canyon plant staff as well as actions recommended previously to prevent loss of decay heat removal capability and to improve recovery, were discussed.
This Notice also stated that the NRC was considering additional generic action on this issue.
On July 9,1987, the NRC Of fice of Nuclear Reactor Regulation sent letters to all NRC licensees of pressurized water reactors addressing NRC concerns and lessons learned from the loss of RHR event at Diablo Canyon on April 10, 1987 (Ref. C-6).
This letter requires licensees to provide to the NRC, within 60 days, a descrip-tion of the operation of their plants during the approach to a partially filled RCS condition and during operation with a partially filled RCS.
Licensees are w-to include the following in their response to this letter.
(1) Descriptions of the circumstances and conditions under which their plant (s) s would enter into and be operated with the RCS partially filled, the in-strumentation and alarms provided to the operators during such operation, the containment closure condition required, the procedures in the control room (including their analytic basis), and the training provided to plant operators and other affected personnel during RCS partially filled conditions; s
s 35 i
y
~
(2) The identification of all pumps that can be used to control RCS inventory; (3) The identification of personnel available to plant operators with spe-cialized knowledge of phenomena and instrumentation associated with RCS partially filled operations; and (4) A description of any changes made to the facility (ies) or procedures as a result of the consideration of the above items; including when such changes were, or are scheduled to be, made.
The information provided by licensees in response to this letter will be used by the NRC staff to assess conformance of the plants with their licensing basis and determine whether additional NRC action is necessary.
4.
Patient Killed by a Falling Shield Head of a Teletherapy Machine On May 19, 1987, NRC Region III was notified by Bartholomew County Hospital in Columbus, Indiana, that a patient had been killed when the approximately 3,000
{
pound shield head of a teletherapy machine fell on her.
The patient, a 59 year-old woman, was being treated for a cancer in the neck A teletherapy device uses a sealed cylirder containing radioactive area.
cobalt 60 as a radiation source for treatment of cancer and other diseases.
The cobalt 60 is contained in a lead and steel shielded " head" which is posi-tioned over the treatment area on the patient.
A shutter is then opened to expose that portion of the patient's body to the radiation for a prescribed period of time.
As the teletherapy head was being lowered into position, it became disengaged from its drive gear and fell onto the patient.
The Columbus fire Department was summoned to remove the patient from beneath the teletherapy head, and she was pronounced dead.
NRC Region III dispatched an inspection team to the hospital.
The team deter-mined that the radioactive cobalt 60 source was in its shielded position and that the accident did not affect the integrity of the shielding.
Therefore, there was no radiation hazard to persons investigating the accident.
The accident was investigated by the device manufacturer, Atomic Energy of Canada, Ltd. (AECL), by ue Food and Drug Administration, which'has regulatory jurisdiction over medical treatment devices, by the Indiana State Board of l
Health, and by the hospital's insurance company.
l The FDA determined that the cause of the accident was the loosening of a l
coupling which connected the screw drive shaft, which raised and lowered the shielded head, to the motor drive unit which turned the shaft.
Set screws I
holding the coupling were found to be loose, and the coupling had apparently worked up the shaft until the shaft suddenly was able to rotate freely, allow-l ing the shielded head to drop.
There were no violations of NRC regulations identified during the investigation (Ref. C-7).
36 l
The manufacturer subsequently issued a User Bulletin to its customers on May 29, 1987, informing them of the accident and specifying inspection and maintenance proceuures for AECL teletherapy units of a design similar to that at Bartholomew County Hospital.
The incident resulted in extensive area news media interest.
Since the device failure did not involve the NRC-regulated radioactive source nor were there any abnormal radiation levels, this event is not considered to be an abnormal occurrence.
5.
Earthquake in Southeastern Illinois Felt at Several Nuclear Power Plants On June 10, 1987, an earthquake occurred in Southeastern Illinois resulting in Unusual Events being declared at six nuclear power stations in the Midwest.
No damage was observed at any of the plants.
An Unusual Event is the least severe of the four emergency classifications of events at nuclear power sta-tions. When an Unusual Event is declared, the utility must notify appropriate state and local agencies as well as the NRC.
The earthquake was centered near Lawrenceville in Southeastern Illinois'and was felt in at least 11 states.
The Clinton Nuclear Power Stat:on, located about 100 miles from the earthquake's center, was the closest nuclear power station.
Unusual events were declared at the Clinton facility and Quad Cities 1 and 2 in Illinois, D. C. Cook 1 and 2 in Michigan, and Prairie Island 1 and 2 in Minnesota, based on the response of seismic instrumentation.
Two additional plants, Dresden 2 and 3 in Illinois and Palisades in Michigan, declared Unusual Events based on the observations of plant personnel.
Ground motion was also felt by persons at several other nuclear plants in the Midwest.
All six plants which declared Unusual Events performed surveys of plant equip-ment and structures.
No damage was observed.
The Clinton facility reported that a motor in the electro-hydraulic system started automatically at the time of the earthquake.
The Palisades plant reported that a level alarm was re-ceived from its Safety Injection Tank.
The tank level was normal, but the alarm was determined to be the result of a momentary level change caused by the earthquake ground motion.
The earthquake received wide news media attention, including reports that it had resulted in Unusual Event declarations at six nuclear plants.
There was, however, no effect on public health or safety.
6.
NRC Augmented Inspection Team Sent to Perry Nuclear Power Station On June 18, 1987, an NRC Augmented Inspection Team (AIT) was sent to the Perry Nuclear Power Station to review the circumstances-associated with a June 17, 1987 unplanned reactor scram at Perry Unit I due~to. closure of the outboard Main Steam Isolation Valves (MSIVs).
Perry Unit 1 is a General Electric-designed boiling water reactor, operated by Cleveland Electric Illuminating Company, and located in Lake County, Ohio.
37
The scram occurred while the plant was at approximately 29 percent power and a reactor natural recirculation test was underway.
The "A" electrical bus in the Reactor Protection System (RPS) tripped as a result of the failure of S circidt card in a circuit protection device.
The tripping of this bus cut oft < <.:ctric power to the outboard MSIVs, causing them to close automatically.
With chese MSIVs closed, the steam flow from the reactor was stopped, initiating the reac-tor scram.
Each steam line has two MSIVs - an inboard valve located inside the reactor containment and an outboard valve located outside the containment.
Upon reenergization of the RPS bus, the outboard MSIVs opened.
The actuation of the MSIVs is controlled by two solenoids " each valve.
According to the design described in the plant's Final Safety Analysis Report (FSAR), each of the two solenoids receive power from a different electrical bus.,
" A" and "B," in the RPS.
The power to both solenoids would have to be lost for them to close.
Therefore, according to the FSAR design, the tripping of the "A" bus should not have caused the valves to close or to reopen when the bus was reenergized.
Subsequent investigation determined there wns a design error which had not been identified in the plant's preoperational testing.
Both solenoids on each of the outboard MSIVs had been connected to the "A" bus.
(The original design configu-ration by General Electric had provided that both solenoids on the outboard MSIVs be powered by the "A" bus and both solenoids on the inboard MSIVs be powered by tne "B" bus.
This design was changed in 1977, but the' change was not made in the installation drawings for the Perry plant.)
The preoperational testing did not identify the error because of an inadequate test procedure and confusing labelling of components being observed during the tests which led to inadequate performance and witnessing of the test procedure.
Further reviews of previous events where an RPS bus had been deenergized showed that for each of these events, the reactor was in cold shutdown with the MSIVs l
closed.
Thus, the previous events provided no indication of the miswiring l
problem.
Based on the unexpected closure of the MSIVs on loss of a single RPS bus trip, the NRC AIT was formed and sent to the site.
The charter of the team was to perform a fact-finding review of the events, to communicate these facts to NRC regional and headquarters personnel, to identify any potential generic safety concerns, and to document the results of the onsite review.
The team also moni-tored the licensee's extensive reviews of the events, which determined that the design error and preoperational testing error were isolated occurrences.
The licensee changed the MSIV circuitry to conform to that described in the FSAR.
The AIT report was forc rded to the licensee on July 30, 1987 (Ref. C-8).
The forwarding letter stated that the NRC was concerned about the series of errors that were made that allowed the plant to be designed, constructed, and tested without identifying that the plant was not built in accordance with the FSAR.
Enforcement action with respect to these errors will be forthcoming in a subse-quent report.
While the miswiring led to an unnecessary plant scram, the safety of the plant was not significantly affected.
The MSIVs could have been opened, if necessary, by transferring to an alternate power supply.
In addition, the circuits con-trolling the closing of the MSIVs were not affected by the miswiring and would have closed the valves should the closure be required by plant conditions.
38
i i
1 REFERENCES I
FOR APPENDICES i
t B-1 U.S. Nuclear Regulatory Commission, " Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979 Accident, Three Mile Island Nuclear Station, Unit 2, Docket No. 50-320," USNRC Report NUREG-0683, Supplement No. 2 (Final Supplement Dealing with Disposal of Accident-Generated Water),
published June 1987.**
B-2 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Notice No. 86-96, " Heat Exchanger Fouling Can Cause Inadequate Operability of Service Water Systems," November 20, 1986.*
B-3 U.S. Nuclear Regulatory Commission, " Improving the Reliability of Open-Cycle Water Systems," USNRC Report NUREG/CR-4626:
Volume 1: "An Evaluation of Biofouling Surveillance and Control Tech-niques for Use at Nuclear Power Plants, prepared by Battelle Pacific Northwest Laboratories, published September 1986.**
I Volume 2: " Application of Biofouling Surveillance and Control Techniques to Sediment and Corrosion Fouling at Nuclear Power Plants,"
prepared by Battelle Pacific Northwest Laboratories, published March 1987.**
B-4 U.S. Nuclear Regulatory Commission, "A Prioritization of Generic Safety Issues," USNRC Report NUREG-0933, published December 1983,** with Supple-i ments issued periodically.**
l i
B-5 U.S. Nuclear Regulatory Commission, " Safety Evaluation Report (SER) on Tennessee Valley Authority Revised Corporate Nuclear Performance Plan,"
i USNRC Report NUREG-1232, Volume 1, published July 1987.**
j B-6 10 CFR S50.54(f) letter from William J. Dircks, NRC Executive Director for Operations, to Charles Dean, Chairman, Board of Directors, Tennessee Valley Authori' - Docket Nos. 50-259, 50-260, 50-296, 50-327, 50-328, 50-390, 50-391,90-438, and 50-439, September 17, 1985.*
B-7 Letter from L. M. Nobles, Plant Manager, Tennessee Valley Authority, Sequoyah Nuclear Plant, to NRC Document Control Desk, forwarding Licensee Event Report (LER)-50-324/87-30, Docket No. 50-324, July 21, 1987.*
B-8 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforce-ment, to Seymour S. Stein, President, Advanced Medical Systems, Inc.,
l
- Available in NRC Public Document Room, 1717 H Street, NW., Washington, DC 20555, for public inspection and/or copying.
- Avail able for purchase from the Superintendent of Documents, U.S. Government -
Printing Office, P.O. Box 37082, Washington, DC 20013-7082.
Also available from the National Technical Information Service, 5285 Port Royal Road, Spririgfield, VA 22161.
A copy is also available for public inspection.and/
or copying at the NRC Public Document Room, 1717 H Street, NW., Washington, DC.
l, 39
i forwarding an Order Suspending License and To Show Cause (Effective Imme-diately), License No. 34-19089-01, Docket No. 30-16055, October 10, 1986.*
{
B-9 Letter from James M. Taylor, NRC Deputy Executive Director for Regional Operations, to Seymour S. Stein, President, Advanced Medical Systems, Inc.,
forwarding an Order Modifying License, Effective Immediately, and Demand
'1 for Information, License No. 34-19089-01, Docket No. 30-16055, July 23, 1987.*
C-1 Letter from William T. Russell, Regional Administrator, NRC Region I, to Roger Heiser, Operations Manager, E.I. duPont de Nemours & Company, Inc.,
forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, License No. 20-00320-21, Docket No. 30-28902, May 28, 1987.*
C-2 U.S. Nuclear Regulatory Commission, " Loss of Residual Heat Removal System, Diablo Canyon, Unit 2, April 10, 1987," USNRC Report NUi!EG-1269, published June 1987.**
C-3 Letter from John B. Martin, Regional Administrator, NRC Region V, to J. D. shiffer, Vice President, Pacific Gas and Electric Company, forward-ing Inspection Report No. 50-323/87-18 (NUREG-1269), Docket No. 50-323, June 19, 1987.*
C-4 Letter from James D. Shiffer, Vice President, Nuclear Power Generation, Pacific Gas and Electric Company, to John B. Martin, Regional Adminis-trator, NRC Region V, Docket No. 50-323, May 4, 1987.*
C-5 U.S. Nuclear Regulatory Commission, " Loss of Decay Heat Removal During Low Reactor Coolant Level Operation," NRC Information Notice No. 87-23, May 27, 1987.*
C-6 Generic Letter No. 87-12, " Loss of Residual Heat Removal (RHR) While the Reactor Coolant System (RCS) is Partially Filled," from Frank J..Miraglia, Associate Director for Projects, Office of Nuclear Reactor Regulation, to all Pressurized Water Reactor Licensees, July 9, 1987.*
C Letter to John C. McGinty, Jr., Bartholomew County Hospital, License No. 13-01631-06, from W. L. Axelson, Chief, Nuclear Material Safety and Safeguards Branch, NRC Region III, forwarding. Inspection Report No. 30-00193/87-01, License No. 13-01631-06, August 5, 1987.*
C-8 Letter to Murray R. Edelman, Vice President Nuclear Group, The Cleveland Electric Illuminating Company, from Charles E. Norelius, Director, Division Of Reactor Projects, NRC Region III, forwarding Augmented Inspec-tion Team Raport No. 50-440/87-14, Docket No. 50-440, July 30, 1987.
- Available in NRC Public Document Room, 1717 H Street, NW., Washington, DC 20555, for public inspection and/or copying.
- Available for purchase from the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082.
Also available from the National Technical Information Service, 5285 Port Royal Road,-
Springfield, VA 22161.
A copy is also available for public inspection and/
or copying at the NRC Public Document Room, 1717 H Street, NW., Washington, DC.
40
N:tc POXM 336 U S NUCLE AR REGUL ATORY COMMisseON i mteoHI NvMeta sou pera er isDC sad ve, No.,,earJ 12 84)
E7o'*EE BIBLIOGRAPHIC DATA SHEET SE E INSTRUCTIONS,0N THE HE W EH5E g
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Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event whic t e Nuclear Regulatory Commission determines I
to be significant from the standpoint of lic health and safety and requires a I
quarterly report of such events to be mac o Congress. This report covers the period April 1 to June 30, 1987. During the r or' period, there were no ' abnormal occurrences at the nuclear power plants licensed t oper e.
There were five abnormal occurrences at the other NRC licensees. Three in lved m ical misadministration (two diagnostic and one therapeutic); one involved t issuanc of an NRC Order to remove a hospital's radiation safety officer due to fal fication o certain records; and one involved a significant breakdown in managemen and procedur 1 controls at an industrial radio-graphy licensee. There was one a ormal occurren reported by an Agreement State-(Idaho). The item involved radi grapher overexpos res. The report also contains information updating some prev ' usly reported abno al occurrences.
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Residual Heat emoval System; Reactor Core Bulk Boiling; Patient Killed i
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