ML20237H210

Report to Congress on Abnormal OCCURRENCES.October-December 1986

ML20237H210
Person / Time
Issue date:	07/31/1987
From:	NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:
References
NUREG-0090, NUREG-0090-V09-N04, NUREG-90, NUREG-90-V9-N4, NUDOCS 8708170105
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NUREG-0090 Vol. 9, No. 4 Report to Congress on Abnormal Occurrences October - December 1986

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Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washingten, D.C. 20013 7082 l A year's subscription consists of 4 issues for this publication.

Single copies of this publication .i are available from National Technical  !

Information Service, Springfield, VA 22161 l

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l-NUREG-0090 Vol. 9, No. 4 Report to Congress on Abnormal Occurrences l October - December 1986 I

Date Published: July 1987 Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission Washington, DC 20555 l

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Previous Reports in Series NUREG 75/090, January-June 1975, NUREG-0090, Vol . 4, No.1. January-March 1981, published October 1975 published July 1981 NUREG-0090-1, July-September 1975 NUREG-0090, Vol.4, No.2. Aprii-June 1981, 4 I

published March 1976 published October 1981 NUREG-0090-2, October-December 1975, NUREG-0090, Vol.4, No.3, July-September 1981, published March 1976 published January 1982 NUREG-0090-3, January-March 1976, NUREG-0090, Vcl.4, No.4, October-December.1981, published July 1976 published May 1982 ,

NUREG-0090-4. April-June 1976, NUREG-0090, Vol . 5, No.1, January-March 1982 published March 1977 published August 1982 NUREG-0090-5, July-September 1976, NUREG-0090, Vol.5, No.2, April-Jene 1982, published March 1977 published December 1982 NUREG-0090-6, October-December 1976, NUREG-0090, Vol.5, No.3, July-September 1982, 3 published June 1977 published January 1983 NUREG-0090-7, January-March 1977 NUREG-0090, Vol.5, No.4, October-December 1982, published June 1977 published May 1983 NUREG-0090-8 April-June 1977 NUREG-0090, Vol.6, No.1, January-March 1983, ,

published September 1983 published September 1977 l NUREG-0090-9, July-September 1977, NUREG-0090, Vol.6, No.2, April-June 1983, published November 1977 published November 1903 NUREG-0090-10, October-December 1977, NUREG-0090, Vol.6, No.3, July-September 1983, published March 1978 published April 1984 NUREG-0090, Vol.1, No.1, January-March 1978, NUREG-0090, Vol.6, No.4, October-December 1983, j published June 1978 rublished May 1984 NUREG-0030, Vol.1, No.2. April-June 1970, NUREG-0090, Vol.7, No.1, January-March 1984, published September 1978 published July 1984 NUREG-0090, Vol.1, No.3, July-September 1978, NUREG-0090, Vol.7, No.2, April-June 1984, I published December 1978 published October 1984 '

NUREG-0090 Vol.1, No.4, October-December 1978, NUREG-0090, Vol.7, No.3, July-September 1984, published March 1979 published April 1985 NUREG-0090, Vol.2, No.1, January-March 1979, NUREG-0090, Vol .7, No.4, October-December 1984, published July 1979 published May 1985 NUPEG-0090, Vol.2, No.2, April-June 1979, NUREG-0090, Vol.8 No.1, January-March 1985, puolished November 1979 published August 1985 NUREG-0090, Vol,2, No.3, July-September 1979, NUREG-0090, Vol .8, No.2. April-June 1985, published February 1980 published November 1985 NUREG-0090, Vol.2, No.4, October-December 1979, NUREG-0090, Vol .8 No.3, July-September 1985, published April 1980 published February 1986 NUREG-0090, Vol.3, No.1, January-March 1980, NUREG-0090, Vol.8 No.4, October-December 1985, published September 1980 published May 1986 NUREG-0090, Vol .3, No.2, April-June 1980, NUREG-0090, Vol.9 No.1, January-March 1986, published hovember 1980 published September 1986 NURfa-0090,Vol.3,No.3, July-September 1980, NUREG-0090, Vol.9, No.2, April-June 1986, published February 1981 published January 1987 NUREG-0090, Vol.3, No.4, October-December 1980, NUREG-0090, Vol.9, No.3,.luly-September 1986, published May 1981 published April 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 October 1 to December 31, 1986.

The report states that for this reporting period, there were three abnormal occurrences at the nuclear power plants licensed to operate. The events were (1) loss of low pressure service water systems at Oconee, (2) degraded safety systems due to incorrect torque switch settings on Rotork motor operators at Catawba and McGuire Nuclear Stations, and (3) a secondary system pipe break resulting in the death of four persons at Surry Unit 2. There were six abnormal occurrences at the other NRC licensees. One involved release of americium-241 inside a waste storage building at Wright-Patterson Air Force Base; three in-volved medical misadministration, one therapeutic and two diagnostic; one involved a suspension of license for servicing teletherapy and radiography units; and one involved an immediately effective order modifying license and order to show cause issued to an industrial radiography company. There were no abnormal occurrences reported by the Agreement States.

l The report also contains information updating some previously reported abnormal occurrences, iii 1

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ABSTRACT .............. ................ ........................ iii PREFACE . ............ ......................................... vii INTRODUCTION ............................................... vii THE REGULATORY SYSTEM ...................................... vii l

REPORTABLE OCCURRENCES ..................................... viii

AGREEMENT STATES ........... .................. ............ ix FOREIGN INFORMATION ............ ......... ................. x l

i REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, OCTOBER-DECEMBER 1986... .............. .. . .......................... 1 i

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! NUCLEAR POWER PLANTS ....................................... 1 86-20 Loss of Low Pressure Service Water l Systems at 0conee................................... 1 86-21 Degraded Safety Systems Due to Incorrect Torque Switch Settings on Rotork Motor Operators at Catawba and McGuire Nuclear Stations... 5 86-22 Secondary System Pipe Break Resulting in Death of Four Persons at Surry Unit 2............... 7 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants) .... 13 OTHER NRC LICENSEES (Industrial Radiographer, Medical Insti tutions , Industrial Users , etc. ) . . . . . . . . . . . . . . . . . . . . 13 86-23 Release of Americium-241 Inside a Waste Storage Building at Wright-Patterson Air Force Base......... 13 86-24 Therapeutic Medical Misadministration ...... ....... 16 86-25 Suspension of License for Servicing Teletherapy and Radiography Units................... 17 86-26 Diagnostic Medical Misadministration ............... 20 86-27 Diagnostic Medical Misadministration ..... ......... 21 86-28 Immediately Effective Order Modifying License and Order to Show Cause Issued to an Industrial Radiography Company................ 22 AGREEMENT STATE LICENSEES .................................. 25 REFERENCES ...................................................... 27 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA ....................... 31 v

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CONTENTS (continued)

Page APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES 33 NUCLEAR POWER PLANTS ................................... ... 33 77-9 Environmental Qualification of Safety-Related Electrical Equipment Inside Containment.............. 33 79-3 Nuclear Accident at Three Mile Island................ 34 86-15 Differential Pressure Switch Problem in Safety Systems at LaSalle Facility................ 36 FUEL CYCLE FACILITIES ...................................... 37 86-3 Rupture of Uranium Hexafluoride cylinder and Release of Gases ............................................ 37 OfHER NRC LICENSEES ........................................ 39 86-10 Willful Failure to Report a Diagnostic Medical Misadministration............................ 39 APPENDIX C - 0THER EVENTS OF INTEREST ........................... 41 REFERENCES (FOR APPENDICES) ..................................... 51 1

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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 abnormal 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 promul-gated in an NRC policy statement which was published in the Federal Register l on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952). In order to provide l wide dissemination of information to the public, a Federal Register notice is

! issued on each abnormal occurrence with copies distributed to the NRC Public Document Room and all Local 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 pe P d from October 1 to December 31, 1986.

Information reported on each event includes: date and place; nature and prob-able consequences; cause or causes; and actions taken to prevent recurrence.

THE REGULATORY SYSTEM The system of licensing and regulation by which NRC carries o>it its responsi-bilities is implemented through rules and regulations in Title 10 of the Code of Federal Regulations. To accomplish its objectives, NRC regularly conducts licensing proceedings, inspection and enforcement activities, evaluation of operating experience and confirmatory research, while maintaining programs for establishing standards and issuing technical reviews and studies. The NRC's role in regulating represen s a complete cycle, with the NRC establishing stan-dards and rules; issuing licenses and permits; inspecting for compliance; en-forcing 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.

In the licensing and regulation of nuclear power plants, the NRC follows the l philosophy that the health and safety of the public are best assured through the establishment of multiple levels of protection. These multiple levels can a

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be achieved and maintained through regulations which 3pecify requirements which will assure the safe use of nuclear materials. The regulations inc!ade design and quality assurance criteria appropriate for the various activiti,es licensed by NRC. An inspection and enforcement program helps Asure complidnce with the regulations. V .

Most NRC licensee employees who work witth or in the vicinity of radioactive materials are required to utilize personnel monitoring devices such as film badges or TLD (thermoluminetcent dosimeter) badges These badges are processed periodically and the exposure results normally serve n. the official and legal record of the extent of personnel exposure to r&dia Mon (uring t,he period the .I l

badge was worn. If an individual's past qxposure history is'known and has been sufficiently low, NRC regulations permit;an individual.in a restricted area to i i receive up to three rems of whole body exposure in a calendar quarsr. Higher .'

values are permitted to the extremities or skin of the whole body! ;f'or unre-1 stricted areas, permissible levels of rddiation are considerably raaller. Per-missible doses for restricted areas and unrestricted areas areTstated in 10 CFR i' Part 20. In any case, the NRC's policy is to maintain radiatish exposures to levels as low as reasonably achievable. ,

REPORTABLE OCCURRENCES i Actual operating experience is an essential input to the regulatory process for i I

assuring that licensed activities are conducted safely. Reporting requirements exist which require that licensees report certain incidents or events to the NRC. This repot ting helps to identify deficiencies early and to atsure that corrective actions are taken to prevent recurrence.

For nuclear power plants, dedicated groups have been formed both by the NRC and by the nuclear power industry for the detailed. review of operating experience to help identify safety concerns early, to improve dissemination of such infor-mation, and to feed back the experience into licensing, regulations. and ,

operations.

In addition, the NRC and the nuclear power industry have ongoing efforts to O improve the operational data system which include not only the type, and qual-ity, of reports required to be submitted, but also the method used to analyze the data. Two primary sources of operr ^.ional data are reports submitted by the licensees under the Licensee Event Report (LER) system, and under the Nuclear Plant Reliability Data (NPRD) system. The former system is under the control of the NRC while the latter system is a voluntary, industry-supported system operated by the Institute of Nuclear Power Operations (INPO), a nuclear utility organization.  ;

Some form of LER reporting system has been in existence sirae the first nuclear power plant was licensed. Reporting require ents were delineated in the Code of Federal Regulations (10 CFR), in the licer s s' technical specification,ns, and/or in license provisions. In order to more effectively col?cct, colfate, store, retrieve, and evaluate the information ancerning reportable events, the Atomic Energy Commission (the predecessor of the NRC) established in 1973 a computer-based data file, with data extracted from licensee reports dating from 1969. Periodically, changes were made to iwove both the effectiveness of data processing and the quality of reports eequired to be submitted by the licensees.

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. Effective January 1, 1984, majcr changes were made to the requirements to re-port to the NRC. A revised Licensee Event Report System (10 CFR S 50.73) was-I established by Commission rulemaking which modified and codified the former LER system. The purpose was to standardi2e 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 anslyses by the licensees of any events required to be reported. All such re-pw ts are to be submitted within 30 days of discovery. The revised system also permits licensees to use the LER procedures for various other reports required under specific sections of 10 CFR Part 20 and Part 50. The amendment to the 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 re-quirements of significant events at operating nuclear power reactors (10 CFR S 50.72). This was published in the Federal Register (48 FR 39039) on August 29, 1983, with corrections (48 FR 40882) published on September 12, 1983. Among the changes made were the use of terminology, phrasing, and reporting thresholds that are similar to those of 10 CFR S 50.73. Therefore, most events reported under 10 CFR S 50.72 will also require an in-depth follocup report under 10 LFR S 50.73.

The NPRD system is a voluntary program for the reporting of reliability data by nuclear power plant licensees. Both engineering and failure data are to be submitted by licensees for specified plant components and systems. In the past, industry participation in the NPRD system was limited and, as a result, the Commission considered it may be necessary to make participation mandatory in order to make the system a viable tool in analyzing operating experience.

However, on July v, 1981, INP0 announced that because of its role as an active user of NPRD system data. it would assume responsibility for management and funding of the NPRD system, INP0 reports that significant improvements in licensee participation are being made. The Commission considers the NPRD sys-tem to be a vital adjunct to the LER system for the collection, review, and feedback of operational experience; therefore, the Commission periodically moni-tors the progress made on improving the NPRD system.

Information concerning reportable occurrences at facilities licensed or other-wise regulated by th'e NRC is routinely disseminated by the NRC to the nuclear industry, the public, and other interested groups as these events occur.

Dissemination includes special notifications to licensees and other affected or interested groups, and public announcements. In addition, information on reportable events is routinely sent to the NRC's more than 100 local public document rooms throughout the United States and to the NRC Public Document Room in Washington, D.C.

The Congress is routinely kept informed of reportable events occurring in licensed facilities.

AGREEMENT STATES Section 274 of the Atomic Energy Act, as amended, authorizes the Commission to enter into agreements with States whereby the Commission relinquishes and the i

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1 States assume regulatory authority over byproduct, source and special nuclear j materials (in quantities not capable of sustaining a chain reaction). Compara-ble 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 is applied uniformly to events at NRC and Agreement' State licensee facilities.

Procedures have been developed and implemented and abnormal occurrences reported by the Agreement States to the NRC are included in these quarterly reports to Congress.

FOREIGN INFORMATION The NRC participates in anl 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.

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i REPORT TO CONGRESS ON ABNORMAL OCCURRENCES-OCTOBER-DECEMBER 1986 NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to operate during the fourth calendar quarter of 1986. As of the date of this report, the NRC had determined that the following events were abnormal occurrences.

86-20 Loss of Low Pressure Service Water Systems at Oconee t

l The following information pertaining to this event is also being reported con-l currently in the Federal Register. Appendix A (see the third general crite-rion) of this report notes that major deficiencies in design, construction, use of, or management controls for licensed facilities or material can be con-sidered an abnormal occurrence. In addition, Example 10 (of "For All Licensees")

of Appendix A' notes that a major deficiency in design, construction, or opera-tion having safety implications requiring immediate remedial action can be considered an abnormal occurrence.

Date and Place - On October 1, 1986, Duke Power Company (the licensee) twice attempted an electrical load shed surveillance test of circuits on Oconee Unit 2, which was shut down for refueling at the time. During both tests, the low pressure service water (LPSW) system was lost. Investigation revealed a ques-tionable design feature which was also applicable to Units 1 and 3. Therefore, the LPSW systems for all three units were considered inoperable and on October 2, 1986, orderly shutdowns of Units 1 and 3 (both operating at 100% power at the time) were commenced.

Oconee Units 1, 2, and 3 each utilize a Babcock & Wilcox-designed pressurized water reactor; the facility is located in Oconee County, South Carolina.

Background - At Oconee, the condenser circulating water (CCW) system takes suc-tion from Lake Keowee and supplies water to the main condensers. Unlike most nuclear power plants, the CCW pumps at Oconee also perform safety-related func-tions which include: supplying a source of water to the LPSW system, the cool-ing water pump for the standby shutdown facility (SSF) emergency diesel genera-tor (EDG), a supply to the SSF auxiliary service water (ASW) pump, and the pri-mary source for cooling the turbine-driven auxiliary feedwater pump for long-term cooling. The LPSW system supplies cooling water for the decay heat removal function of the low pressure injection system and other safety-related equipment.

The LPSW system pumps take suction on the upstream side of the condenser from the CCW system crossover lines between Oconee Units 1, 2, and 3.

Each of the four CCW pump motors for each Oconee unit is capable of being powered from either of two emergency hydro generators. However, the Oconee plant is designed to accommodate a loss (shedding) of the CCW pumps and still provide LPSW pump suction through a siphon arrangement. The siphon is neces-sary because of a high point in the CCW piping just downstream of the CCW pumps and upstream of the LPSW pump suction. This high point may be as much as 25 feet above the level of Lake Keowee (depending upon lake level).

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i Nature and Probable Consequences - On October 1, 1986, while Unit 2 was in a refueling outage, the Unit 2 load shed test was performed. At Oconee, a load i shed of non essential loads is initiated when emergency power is required via j the underground feeder from Keowee Hydro Station. The load shed protects this I power path f rom overload. I When the load shed test was initiated, the condenser circulating water pumps were deenergized. Normally, this causes the gravity flow system to automat-ically align and to allow the flow of water from the Lake Keowee intake struc-ture through the condenser and discharging to the Keowee tailrace into Lake l Hartwell. The elevation difference and a siphon effect are used to cause the condenser circulating water to continue to flow. This mode of operation of the CCW system is referred to as the emergency condenser circulating water (ECCW) system. For this test, the condenser gravity drain to the Keowee tailrace was blocked because this was not part of the test.

After about an hour, the LPSW pumps began to cavitate and stop pumping. One LPSW pump was stopped by the control room operator, and a second LPSW pump was observed to have low discharge pressure and cycling amps. Various high tem-perature alarms for the components cooled by LPSW were received in the control room. CCW flow was restored by restarting a CCW pump and the plant was restored to its normal power condition without any plant damage or system upsets. Prior to the occurrence, two LPSW pumps were operating with approximately 13,000 gpm/

pump. The CCW crossover header, which provided suction for the LPSW pumps, was being supplied by the Unit 2 CCW pumps at the time.

In the evening of October 1, 1986, the test was repeated, but this time the gravity drain feature was also tested. The results were the same as in the first test, i.e., loss of LPSW system function due to loss of LPSW pump suc-tion. NRC Region II was advised of these results late in the evening, and concurred with the licensee that Units 1 and 3 (operating at 100% power) could continue to operate until the test data could be fully evaluated. At 9:00 a.m.

on October 2, 1986, evaluation of the tests revealed that the operation of this design feature (the ECCW system) was questionable for Units 1 and 3, and that I

this resulted in inoperability of all LPSW systems for Oconee. As a result, an orderly shutdown of the two operating units was begun as required by Technical Specification 3.3.7. Both units reached cold shutdown conditions by October 3, 1986.

The licensee's analysis of the event showed that the loss of suction to the LPSW pumps was caused by a loss of the previously described siphon. The CCW pump discharge flange is normally nine.-feet below the surface of Lake Keowee when the lake is at full level. However, because of drought conditions, the lake level was about six feet below the flange at the time of the load shed test. During operation at these reduced lake levels, minor water leakage of the flange had been observed. This leakage was insignificant during plant operation. However, with the CCW pumps off (shedded), air inleakage at this flange caused the high point in the CCW system piping to drain and resulted in a loss of siphon flow.

Siphon flow, if initiated, could not be sustained in the system, as originally designed and built, during low lake level conditions because of air inleakage at the CCW pump discharge flange. It appears that ;.revious surveillance tests 2 j l

were not of sufficient duration to determine that siphon flow was sustained.

Since the large volume of water contained in the CCW lines provided LPSW flow for about an hour before the loss of LPSW suction, it appears that load shed testing personnel, in the past, may have been misled into thinking siphon flow had been sustained.

As mentioned briefly in the " Background" information above, the Oconee CCW syrtem is designed to also provide suction and discharge (heat sink) for the cooling water pump for the SSF EDG and a supply for the SSF ASW pump. The SSF was designed to be a backup means to achieve and maintain the plant in a hot shutdown condition. Analysis performed subsequent to the above load shed test showed that if siphon flow were lost in the CCW system pipe, the CCW system could not provide an adequate heat sink for SSF operation to meet its design basis of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of operation. In addition, when the CCW pumps are not operat-ing, the CCW system should provide emergency gravity-siphon CCW system flow to the main condensers to recover condensate for DHR following certain postulated events until the DHR system is in operation. The gravity flow is possible be-cause the CCW system discharge from the main condenser is shifted to an alter-nate pipe that discharges downstraam of Lake Keowee dam at an elevation well below the CCW system intake. This feature of the CCW system also was disabled by the loss of siphon.

The safety significance of the October 2, 1986 event at Oconee is that the event revealed an unanticipated failure mode which resulted in the loss of the ECCW system and the safety-related functions of the LPSW system. This situation has existed ever since the first Oconee unit went into operation in 1973.

Cause or Causes - The root cause of this incident is the inadequate design and testing of the ECCW system. This led to a failure of the ECCW system to perform its intended function as described in the FSAR under all assumed conditions. Inadequate original design evaluation of the ECCW system and the lower than normal lake level due to extreme drought conditions of Lake Keowee

, are contributing factors to the cause of this incident.

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Actions Taken to Prevent Recurrence Licensee - The licensee modified the discharge flange on all CCW pumps to pre-vent air inleakage when the lake level is below the discharge flange. The LPSW pumps were successfully tested for several hours with the CCW pumps off and the lake level below the discharge flange. The emergency CCW gravity-siphon flow to the main condersers and the SSF EDG cocling water pump also were successfully tested under the above conditions. In addition, the SSF cooling water pump was modified to take a separate and independent suction from Lake Keowee.

The licensee inspected each continuous vacuum priming (CVP) line at the CCW system intake for blockage. Unit 3 lines were clear and vacuum was established on Unit 3 intake high point vents. Unit 1 and Unit 2 lines were found blanked off with blind flanges which prevented the CVP pumps from developing adequate vacuum on the CCW system intake high point vents to overcome air inleakage at the pumps. These flanges had apparently not been removed at the completion of the original system hydrostatic testing. The flanges were removed and a vacuum was established on Units 1 and 2 intake high point vents.

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Successful testing was performed on the condensate steam air ejector, the tur-bine bypass valves (TBV), and the siphon effect. By October 23, 1986, all three units were returned to service.

l' Further corrective actions planned by the licensee include:

(a) review and )

analyze the seismicity of the CCW system, (b) develop a program to include CCW l system piping in a routine inspection, (c) review the validity of the testing program to ensure that systems and components are tested adequately, and (d) review Technical Specifications to determine if any revisions are necessary.

NRC - The NRC Resident' Inspectors for the Oconee site were observing the Unit 2 i load shed test in progress on October 1, 1986, and observed the failures. They' observed licensee actions to assure that the Units remained in a stable condi-tion and notified Region II. Initial investigation of the circumstances ,

associated with this event began while surveillance test data was being analyzed. (

On October 4, 198G, the Deputy Director of vae Division of Reactor Projects, Region II, went to the site to observe the licensee's repairs for a fix of the problem, to review the potential of the proposed fix to solve the problem, and to assess the overall significance of the event.  ;

On October 8, 1986, a meeting between NRC and licensee personnel was held in  ;

the NRC Region II Office to discuss actions being taken and planned by the licensee to repair and demonstrate operability of the Oconee units. Actions to be taken and required conditions prior to restart of the units were also discussed.

The NRC Resident Inspectors witnessed the repairs and the subsequent testing to confirm the overall adequacy of the licensee's corrective action prior to  ;

restart of the units.

On October 14, 1986, an NRC Management Meeting with the licensee was held in Bethesda, Maryland, to review the completed modifications'and test results.

The licensee presented the results of the surveillance which were conducted following the described modifications to the pumps.

NRC Inspection Reports No. 86-26 and No. 86-33 concerning the incident were forwarded to the licensee on October 23, 1986, and December 1, 1986, respec--

tively (Refs. 1 and 2). On December 22, 1986, an-NRC Enforcement Conference was held at the NRC Re;;%, il Office to discuss the event with the licensee (Ref. 3). On February 5,1987, the NRC issued a Notice of Violation to the licensee (Ref. 4), regarding the operability of the ECCW system. The viola-tion was classified as Severity Level IV (on a scale where Levels I and V are the most and least severe, respectively).

On January 30, 1987, the NRC issued Inspection and Enforcement Information Notice No. 87-06 (Ref. 5) to all nuclear power reactor facilities holding an operating license or a construction permit to inform them of the Oconee event.

This item is considered closed for the purposes of this report.

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86-21 Degraded Safety Systems Due to Incorrect Torque Switch Settings on Rotork Motor Operators at Catawba and McGuire Nuclear Stations

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The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see Example 10 of "For All Licensees") of this report notes that a major deficiency in design, construction, or operation having safety implications requiring immediate remedial action can be considered an abnormal occurrence.

Date and Place - On October 23, 1986, Duke Power Company (the licensee) dis-covered that numerous safety-related valves at Catawba Nuclear Station were degraded due to incorrect torque switch settings. On October 28, 1986, a similar situation was also found at the licensee's McGuire Nuclear Station. Catawba Units 1 and 2, and McGuire Units 1 and 2 are Westinghouse-designed pressurized i

water reactors. The Catawba Station is located in York County, South Carolina l and the McGuire Station is located in Mecklenburg County, North Carolina.

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Background - At Catawba and McGuire, and many other plants, Rotork Actuators are used for remote control of plant valves. Many of these valves are in safety related systems. The actuators are driven by electric motors. The size of the motor and the actuator depends on the size of the valve and the force or torque necessary to open and close the valve. Rotork Actuators have five torque switch settings which the licensee had assumed represented 40, 55, 70, 85 and 100 percent of the maximum rated torque output. The required torque switch setting for each actuator is determined based on the maximum differential pres-sure expected on its associated valve during anticipated events.

1 If incorrect torque switch settings are used, the valves may not perform as designed (e.g., the actuator motor may switch off before the valves complete their travel) thereby possibly degrading their function to avoid, or mitigate, the consequences of a transient or a design basis accident. Improper torque switch settings, on motor operators manufactured by various vendors, have been a contributing cause in a number of significant events, one of the most serious of which was the complete loss of main and auxiliary feedwater event at Davis-Besse on June 9, 1985. The Davis-Besse event was reported as abnormal occurrence No. 85-7 in NUhcG-0090, Vol. 8, No. 2 (" Report to Congress on Abnormal Occurrences: April-June 1985").

On November 15, 1985, the NRC issued IE Bulletin No. 85-03 (Ref. 6) to all holders of nuclear power reactor operating licenses or construction permits for action. The Bulletin described various events (including the June 9,1985 event at Davis-Besse) during which motor-operated valves failed on demand, in a common mode, due to improper switch settings. The Bulletin also described numerous previously issued NRC reports, Information Notices, Bulletins, and Circulars (as far back as 1972) involving problems with torque switches.

Bulletin No. 85-03 requested licensees to develop and implement a program to ensure that switch settings on certain safety-related motor-operated valves are selected, set, and maintained correctly to accommodate the maximum-differential pressures expected on these valves during both normal and abnormal events within the design basis. It was during the licensee's followup to the Bulletin that the problem at Catawba and McGuire was discovered.

5

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Nature and Probable Consequences Catawba Nuclear Station (CNS) l 1

On October 23, 1986, with both CNS units in cold shutdown, a valve was being i repaired. The motor and worm gear had been replaced and the valve's actuator 1 was being calibrated per CNS plant procedure. After setting the torque switch i to the specified setting, per procedure, while the valve was on the test bench, i the torque output was checked. The result indicated the torque output was lower {

than required. Subsequently, a performance curve (percent torque output versus torque switch setting) was obtained. As stated above, the licensee had assumed that setting 1 represents 40 percent of rated torque and setting 5 represents 100 percent of rated torque.

A linear or straight curve had been assumed by the licensee for the range from 40 percent to 100 percent. However, results of the performance curve indicated a non-linear relationship between percent torque output versus torque switch setting. Also, the curve obtained was generally lower than the linear rela-tionship previously thought to be correct.

Based on these results, the valve being tested at CNS was determined to be incapable of performing its intended function. Subsequent review indicated that at CNS this situation was not unique to the single valve discussed.

Fifty-three valves were determined to potentially be affected. Systems which contained safety-related valves (many of which were containment isolation valves) which could potentially be affected were:

(a) Chemical and volume control systems (b) Component cooling system (c) Residual heat removal system (d) Ice condenser refrigeration system (e) Safety injection system (f) Nuclear service water system (g) Containment hydrogen purge system (h) Breathing air system (i) Instrument air system (j) Containment air release and addition system McGuire Nuclear Station (MNS)

Based on the above event at CNS, it was determined at MNS on October 28, 1986 ,

that the charging line outside containment isolation valves for both MNS units l were technically inoperable. In fact, analysis of the two valves, which would be required to close during safety injection, indicated that they may not be able to do so under differential pressure conditions which could be encountered following a loss-of-coolant accident.

Unit I was operating at 100 percent and Unit 2 was operating at 47 percent of full power. With both trains of the emergency core cooling system thus inoperable, the licensee commenced shutdown of both Units. A plan was estab-lished to inspect all safety-related Rotork motor operators and perform a detailed engineering evaluation on each affected valve to determine the appro-priate corrective action to ensure its operability.

1 i l

6

Cause or Causes - The cause of the problem appears to be poor communication between the motor operator vendor and the licensee. The vendor subsequently stated that whenever the factory torque setting is changed in the field, an individual calibration curve or bench test 1.s required to accurately determine torque output. In fact, based on verbal communication with the vendor, the licensee utilized a linear or straight line curve for the relationship between percent torque versus torque switch setting.

Actions Taken to Prevent Recurrence Licensee - The licensee took the following corrective steps:

a) Operability requirements for all safety-related Rotork motor operators were evaluated individually.

b) Corrective action for each motor operator evaluated.was performed, if required.

c) Corrective maintenance procedures for Rotork motor operators were revised to require a bench test of the operator when setting torque switches.

d) A detailed inspection of each safety-related Rotork motor operator l installed was performed.

e) All significant discrepancies found during inspection were corrected.

f) A new document which specifies the required torque output (in terms of torque output, and not torque switch setting) for each valve with a Rotork motor operator was completed and approved.

g) The licensee committed to implement a program by November 15, 1987, to ensure that torque switch settings on all safety-related motor operated valves are selected, set, and maintained correctly. This commitment was made in accordance with the previously discussed NRC IE Bulletin No. 85-03.

NRC - The NRC monitored the licensee's corrective actions to assure that they were responsive. Subsequently, NRC met with the licensee and Rotork to follow progress on the issue.

On November 3, 1986, the NRC issued IE Information Notice No. 86-93 (Ref. 7),

which described the McGuire event, to all nuclear power reactor facilities holding an operating license or a construction permit.

This item is considered closed for the purposes of this event.

86-22 Secondary System Pipe Break Resulting in Death of Four Persons at Surry Unit 2 The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see Example 10 of "For All Licensees") notes that a major deficiency in design, construction, or operation r

7 1

l 1

having safety implications requiring immediate remedial action can be considered  :

an abnormal occurrence.

i Date and Place - On December 9, 1986, with both Surry Units 1 and 2 operating i at 100% reactor power, a Unit 2 reactor trip (scram) followed by a main feedwater (MFW) line rupture occurred. Eight individuals were injured by the escaping steam and water. Four of those injured subsequently died.

l The Surry Power Station consists of two Westinghouse-designed pressurized water i reactors. The station is operated by Virginia Electric and Power Company (the I

licensee) and is located in Surry County, Virginia.

Nature and Probable Consequences - On December 8, 1986, Unit 2 had completed a i

refueling outage and had returned to full power operation. Unit 1 was also operating at 100% power.

On December 9, 1986, at about 2:20 p.m. (EST), a low-low level in the "C" steam generator (S/G) of Unit 2 caused an automatic reactor trip and an automatic start of the two motor-driven auxiliary feedwater pumps. The reactor trip also resulted in a trip of the Unit 2 main turbine generator.

The control room operators noted the S/G code safety valves lifting and regulated S/G pressure through the atmospheric dump valves. Approximately 30 seconds after the trip, the unit's electrical busses auto-transferred to offsite power.

A small steam release noise was heard followed by a very loud noise approxi-mately five seconds later.

A shift supervisor, who was in the turbine building observing construction activity around the MFW pumps, realized that a large break had occurred and ran to the control room to alert the control room operators. He also told them that people had been injured. All secondary pumps were secured and the break iso-lated. Water to the S/Gs was supplied by the auxiliary feedwater system.

The primary system responded normally to the loss of load transient. Reactor coolant temperature, pressure, and pressurizer level were stabilized in the desired band.

A notification of unusual event was declared at 2:27 p.m. At 2:30 p.m., ground and air ambulances were called to take the injured people to the hospital. At 2:40 p.m., the unusual event was upgraded to an Alert in order to ensure accountability of all station personnel.

Unit 2 was stabilized by 2:34 p.m. with two reactor coolant pumps running and primary system pressure and temperature being maintained by relieving steam to the atmosphere. No radioactive releases resulted from the event and a cooldown was initiated.

At 3:48 p.m., accountability of personnel was completed and at 4:25 p.m., the Alert was terminated. At 3:55 a.m. on December 10, 1986, Unit 2 was placed on the residual heat removal system; at 7:04 a.m., the Unit achieved cold shutdown conditions.

8 i

Later investigation showed that an 18-inch suction line to the A train main feedwater pump had ruptured at the elbow where the line connects to the 24-inch condensate header. At the time of the rupture, contractor personnel employed by Daniel Construction Company of Greenville, South Carolina, and by Insulation Specialties, Inc. , of Hopewell, Virginia, were doing instrument j line relocation and pipe insulation work in the general area of the ruptured pipe. As a result of the escaping steam and water, six of these individuals were hospitalized for treatment of severe burns. Three were evacuated directly from the site by helicopter, and three others were taken off site by l

ambulance. The other two, who were less severely injured, were treated at a clinic and rele' sed.

One of those hospitalized died the afternoon of December 10, 1986, and another victim died on December 11, 1986. Two others died several days later. The other two remained in serious to critical condition. One of the two improved and was later released but the other was still in serious condition more than a month after the accident.

l The escaping steam and water also caused various system interactions which l complicated the licensee's handling of the event. For example, one important complication was that water and steam saturated a security card reader which

! shorted out the entire plant card reader system. As 6 result, key-cards would not open plant doors. Security personnel responded to the control room and provided access control while doors into the control room were opened for easy access and to improve control room ventilation. Guards admitted personnel on l the basis of personal recognition and excluded non-essential personnel. The card reader system returned to service approximately 20 minutes after the pipe break and functioned normally thereafter.

An operator reported being delayed in the stairway outside the control room as a result of the card reader failure. Due to the hot water conditions on the turbine building basement floor and the discharge of the Halon fire suppression system in the emergency switchgear rooms below the control room and the carbon dioxide fire suppression system in the cable tray rooms above the control room (see discussion below), the operator had no safe way to exit the stairway other than the control room itself. The operator was admitted to the control room by someone opening the door from inside the control room. The licensee is con-sidering installing electronic override switches which would permit the opening of electronically locked doors in emergency situations.

Another important complication was that within minutes of the feedwater pipe rupture event in the Unit 2 turbine building, portions of the Unit 2 turbine building sprinkler system actuated. Sixty-two sprinkler heads opened in the immediate area of the feedwater pipe rupture due to the high heat levels associated with the event. As they opened, these sprinkler heads immediately began discharging water to cool the turbine building atmosphere.

Water short-circuited control systems in the area for the Halon and carbon dioxide fire suppression systems; as a result, the systems activated. Control room habitability became a concern because doors were blocked open to allow better control room access without recognizing that carbon dioxide had been discharged in the areas above. The carbon dioxide was apparently coming into the control room from the turbine building hallway. Control room personnel in 9 l

the main control room annex and near the main control room door experienced shortness of breath, dizziness and nausea. However, once they recognized that carbon dioxide was present, the control room operators took appropriate correc-j tive actions and initiated control room emergency air supply fans which placed I the main control room at a higher pressure than the turbine building. This action assisted in diluting and exhausting the existing carbon dioxide levels and kept any additional carbon dioxide from infiltrating into the main control room.

i l About 3:30 p.m. on December 9, 1086, a decision was made by NRC Region II management to send an inspection team to the site. This team consisted of -l Regional-based personnel and the senior resident inspectors from North Anna and l Surry. The Region II inspection team arrived on site about 9:30 p.m. on (

December 9, 1986, to assess the operational status of the unit and to inspect {

the damaged area of the Unit 2 turbine building. During the morning of l December 10, 1986, the team's status was upgraded to an Augmented Inspection 1 Team (AIT), and an engineer from the NRC Office of Nuclear Reactor Regulation, -

knowledgeable in water hammer phenomena, was assigned to the team. The AIT was tasked with the job of examining the licensee's response to the incident and performing a separate investigation.

Following the termination of the Alert classification on December 9,1986, licensee management initiated recovery activities. An organization was established and resources identified for evaluating the incident and recommend-  !

ing recovery actions. The licensee's preliminary findings resulting from the Unit 2 main feed pump suction pipe rupture indicated that there may have been significant thinning of the pipe wall due to a corrosion / erosion mechanism not-fully understood at the time. Since the same mechanism could similarly affect Unit 1, at 12:30 p.m. on December 10, 1986, the licensee decided to shut down i Unit 1. The shutdown was initiated at 5:30 p.m. on December 10, 1986, and the 1 unit was off the line at 10:47 p.m. Subsequently, the unit was placed in a cold shutdown condition.

Based on the licensee's investigation and the NRC Region II AIT inspection /

investigation, it has been concluded that the following factors attributed to the main feedwater pipe rupture events.

1. Pipe Wall Thinning on "A" Main Feed Pump Suction Line The 18-inch suction line which supplies the "A" Train Main Feed Pump was fabricated using ASTM A-106, Grade B, Extra Strong, carbon steel seamless pipe and ASTM A-234, Grade B, Extra Strong, WPB carbon steel wrought fit-tings which had a nominal wall thickness of .50 inches 110% at installation.

Since installation, the bulk single phase corrosion / erosion mechanism had subsequently reduced the original wall thickness. Ultrasonic wall thick-ness measurements and micrometer measurements taken on the elbow following the failure showed a gradually sloping wall thickness loss over much of the suction line. At several locations, usually near welds, localized cavities had been formed in the elbow inner surface by the corrosion /

erosion process. The remaining wall thickness of these localized areas has been measured as low as .048 inches while adjacent locations were .090 to .140 inches in thickness. Using the code minimum wall equation and assuming an internal pipe pressure of 600 psig, a temperature of 370 F, 10

and an ultimate strength of 60,000 psi results in a calculated burst thick-( ness of .090 inches and a yield thickness of .173 inches. This difference in the pipe thickness directly contributed to the pipe failure.

2. Main Steam Trip Valve and Instrument Air Pressure Surry Unit 2 tripped as designed when a low-low S/G water level protec-tion signal occurred on the "C" Steam Generator Reactor Protection System.

Instrumentation. This occurrence was a result of the unplanned closure of the "C" main steam trip valve (MSTV). The MSTV closure was initiated by a slight reduction in instrument air pressure combined with valve tr,isassembly.

The "C" MSTV would not be expected to close with the slight reduction in instrument air pressure; however, the investigation revealed that the valve disc had not been in a fully ope 1 position because of misalignment of the valve bonnet that occurred when the valve was overhauled during the past refueling outage. This slight reduction in instrument air pressure allowed enough air pressure decrease for the steam flow to force the disc shut.

This directly contributed to the trip, but not to the pipe rupture.

3. Main Feed Pump Discharge Check Valve The licensee's inspection of the 2A main feed pump check valve revealed two hardware deficiencies, i.e., a missing disc hinge pin and a displaced seat-disc assembly. Further investigation showed that these deficiencies did not contribute to the cause of the break, but may have permitted an additional amount of water to be discharged through the break.

Cause or Causes - The investigations have indicated that the rupture of the suction line elbow resulted from the combination of wall thinning due to bulk single phase corrosion / erosion and a normal feed pump suction pressure tran-

, sient. The root causes appear to be a design deficiency (associated with piping I configuration and flow velocity) and operational circumstances associated with water chemistry.

Actions Taken to Prevent Recurrence Licensee - The licensee developed a comprehensive plan for inspection, evalua-tion, and modification / replacement, as necessary for components-(e.g., piping, uain steam trip valves, main feed pump discharge check valves), systems (e.g. ,

fire protection system) and procedures (e.g. , security plan, emergency plan and communications). The licensee also developed a station startup plan. On January 14, 1987, the licensee forwarded to the NRC a report entitled "Surry l Unit 2, Reactor Trip and Feedwater Pipe Failure Report" (Ref. 8), which pro-vided detailed information on the event, together with the station recovery

, plan and corrective actions for NRC review and concurrence prior to station-l restart. As discussed below under "NRC Actions," the licensee's plans and ac-tions were acceptable to the NRC. Subsequently, Surry Units 1 and 2 were returned to service on February 23, 1987 and March 20, 1987, respectively.

The licensee also operates the North Anna nuclear power station which consists of two units similar to those of Surry. The facility is located in Louisa County, Virginia. Subsequently, when pipe wall thinning was found at Surry 11  !

l

Units 1 and 2, the licensee decided to inspect similar piping at North Anna j Unit 1. Therefore, on December 25, 1986, power was reduced from 100%, reach-ing 20% on December 26, 1986. Approximately 4900 ultrasonic inspections were  ;

made on North Anna Unit 1 piping. No measurements indicated pipe wall thick-ness below the required minimum. The feedwater pump suction piping and header wall thicknesses were within original pipe manufacturing specifications, -and i the high pressure drain pump discharge piping was no more than 15 percent below the original specifications. Since no abnormal conditions were found, Unit l' was returned to full power on December 27, 1986. North Anna Unit 2 and addi-tional Unit 1 piping inspections will be performed during future outages.

NRC - As previously mentioned, an NRC AIT was sent to the Surry facility on Dicember 9, 1986. The AIT conducted inspections during the remainder of the week ending December 12, 1986, to ascertain the circumstances involved in the accident. An executive summary was transmitted to the Region II office on Dec uber 17, 1986. This summary provided the significant facts concerning the event. The AIT did not conclude its inspection at that time due to the ongoing activities by the licensee to develop a root cause analysis, which required subsequent inspection activities.

AIT activities continued during the weeks of December 22 and 29, 1986, and January 12, 1987. An AIT exit meeting with plant management was held on January 14, 1987, after review of the licensee's investigative report (Ref. 8) and proposed corrective actions which were presented to the NRC on January 12, 1987. In addition to the AIT inspection activities, inspectors knowledgeable in security, fire protection systems, water chemistry and check valve design were assigned to review specific concerns in these areas. Where applicable, their inspection findings were incorporated into the AIT report.

The AIT Inspection Report was forwarded to the licensee on February 10, 1987 (Ref. 9). The forwarding letter stated that the AIT concurred in the licensee's planned actions in order to restart the plant. The forwarding letter also included a Notice of Violation regarding maintenance procedures for overhauling a main steam trip valve. The violation was classified as Severity Level IV (on a scale in which Levels I and V are the most and least significant, respectively). q On December 16, 1986, the NRC issued Inspection and Enforcement Information Notice No.86-106 ("Feedwater Line Break") to all nuclear power reactor facil-ities holding an operating license or a construction permit (Ref. 10). The Notice alerted addressees of a potentially generic problem with feedwater pipe thioning and other problems related to the Surry Unit 2 event. Supplement 1 to that Notice, issued February 13, 1987, provided additional information about thinning of piping walls which led to the pipe break (Ref. 11). Supplement 2 to the Notice, issued March 18, 1987, provided information about potentially generic systems interaction problems that were caused by the release of large quantities of feedwater (Ref. 12).

This item is considered closed for the purposes of this report.

• • • • • • • • • • l

)

12  ;

FUEL CYCLE FACILITIES (Other Than Nuclear Power Plants)

The NRC is reviewing events reported by these licensees during the fourth cal-endar quarter of 1986. As of the date of this report, the NRC had not determined that any events were abnormal occurrences.

OTHER NRC LICENSEES (Industrial Radiographer, Medical Institutions, Industrial Users, etc.)

There are currently about 9,000 NRC nuclear material licenses in ef fect in the United States, principally for use of radioisotopes in the medical, indus-trial, and academic fields. Incidents were reported in this category from licensees such as radiographer, medical institutions, and byproduct material users.

The NRC is reviewing events reported by these licensees during the fourth cal-endar quarter of 1986. As of the date of this report, the NRC had determined that the following events were abnormal occurrences.

86-23 Release of Americium 241 Inside a Waste Storage Building at Wrioht-Patterson Air Force Base 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 /

! procedural controls in major areas can be considered an abnormal occurrence.

Date and Place - On September 18 and October 6,1986, a drum containing radio-active waste was opened to inspect its contents at Wright-Patterson Air Force Base, located near Dayton, Ohio. Opening the drum caused a significant release of americium-241 inside the waste storage building, resulting in extensive con-tamination of the facility; significant efforts and costs have been incurred in decontaminating the facility, as well as in the investigations being performed by the NRC and other agencies.

Background - The United States Air Force (USAF) holds a license, issued by the NRC on June 26, 1985, which grants the USAF the authority to issue Permits to Air Force locations where the NRC has regulatory jurisdiction. The management and control of this license are the responsibilities of the USAF Radioisotope Committee. The Executive Secretary of this Committee is located at Brooks Air Force Base near San Antonio, Texas. Wright-Patterson Air Force Base is the holder of a USAF Radioactive Material Permit issued by the Committee on December 18, 1985, as a conversion of a previously issued NRC specific license.

Nature and Probable Consequences - In response to a request by the USAF Radio-isotope Committee, the Wright-Patterson Radiation Safety Officer and another individual began an inventory on September 18, 1986 of radioactive waste drums which were in storage at the Air Base. Five drums were not labeled as to their 13

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contents. When one of the unlabeled drums was opened, the two individuals per-forming the inspection noticed that Beir alpha radiation detectp had gone .'

"off scale" (caused by a radiation level in excess of the measuring scale).

The individuals immediately left the storage building and were assisted by two additional individuals in removing their protective garments. The two.individ-

~

uals who had opened'the drum and one of the adelitional technicians had some minor radioactive contamination remaining on. '

i hands, necks, and shoes.

They subsequently decontaminated themselves u>.,1,1 soap and water.

Subsequent radiation surveys revealed the presence of radioactive contamination inside the building. There were also detectable releases outside the structure, but these were below NRC release criteria for unrestricted areas. The waste .

I storage building is located in a remote controlled-access munitions storage' area.

On October 1,1986, the Air Force Radiological Assistance Team and a private contractor (Chem-Nuclear Services) arrived at the Air Base to assist in decon-tamination work and repackaging of the waste. Because of an unexplained high .

radiation level emanating from one of the drums, the drum was opened on -

October 6, 1986, and the contents transferred to a larger drum. -The drum which was opened on October 6 was the same drum that was opened.on September 18, and.

the reopening resulted in further release of radioactive materials inside the building. Although they were wearing anti-contamination clothing, the workers received contamination on their personal clothing, and one individual received an uptake (inhalation) of airborne americium-241, which may have exceeded the NRC limit of 3.8 nanocuries. The actual uptake .is in the process of being determined, pending further testing and analysis. The preliminary evidence, however, is that the uptake is below the level at which detectable medical effects would be expected.

A technician reentered the building three days later and determined that the  !

removable surface contamination levels had increased to as high as 2,875,000 disintegrations per minute. This level of contamination requires cleanup.to prevent personnel contamination or release to the environment. Workers in the facility would have to wear respiratory protection. For an area that'is'open to the public, the NRC guidelines call for removal of contamination above 20 i disintegrations per minute. In additien to the surface contamination, the measurement of airborne radioactivity in the building was approximately one I million times the NRC limit for a restricted area. 1 Actual decontamination work began October 30, 1986, and continued until November 18, 1986. At that time all radioactive waste drums had been examined and repackaged (with the exception of the drum which was the source of the- l americium-241 contamination). Most radioactive contamination was successfully decontaminated by the Air Force's contractor. Approximately 100 microcuries of contamination remained in the building, awaiting a decision on further decontamination work or dismantling and disposal of the building. Access to the building remains restricted.

The drum containing the americium-241 (estimated by external measurements to be 1.6 to 2.2 curies) will be transferred to a Department of Energy facility for storage.

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The costs to date (as of mid to late February 1986) for decontamination, repack-aging, and disposal are approximately $500,000. Additional costs will be incurred during final disposition of the storage building.

Subsequent investigation by the Air Force and by the NRC determined that the barrel containing the americium-241 had originated at a former NRC licensee's facility and had been accepted by an individual at the Air Base for disposal in the 1970s. (The circumstances of the transfer of the waste remain under investigation.) The waste barrel had apparently remained in storage since that time.

Cause or Causes - The root cause appears to be attributed to deficient management /

procedural controls. However, the event remains under investigation by the NRC Office of Investigations, and a complete understanding of all contributing causes awaits their report.

Actions Taken To Prevent Recurrence Licensee - The licensee's investigation of the incident is continuing. The Radiation Safety Officer and two other individuals associated with the handling of the incident have been removed from any work involving NRC-licensed radio-active materials; this was documented by a Confirmatory Action Letter issued by NRC Region III on February 5, 1987 (Ref. 13).

The licensee has provided information to cther Air Force bases on handling practices for opening waste drums and on the NRC's reporting requirements.

In addition, the USAF will have to decide whether further decontamination attempts should be undertaken, or whether the waste storage building should be dismantled and disposed of as radioactive waste.

! NRC - After the NRC learned of the scope of the contamination incident, NRC inspection personnel were dispatched from Region III to review the circumstances of the incident and to monitor the licensee's decontamination activities. Per-sonnel from Oak Ridge Associated Universities were also retained by the NRC to perform radiation surveys at the Air Base and to assist the NRC in reviewing the decontamination plans and activities.

Region III issued a Confirmatory Action Letter on October 27, 1986, documenting the Air Force's agreement (a) not to enter the building without NRC authori-zation; (b) to provide training for fire protection personnel; (c) to maintain a fire and security watch at the site; (d) to notify the NRC promptly of any problems concerning the contaminated building such as fire, damage due to storms, or detection of contamination outside the building; and (e) to provide an Incident Report to the NRC within seven days (Ref. 14).

The NRC is continuing its investigation of the circumstances surrounding the l contamination incident and its handling by the Air Force, including the report-1 ing of this incident to NRC. On February 19, 1987, the NRC issued a letter

(Ref. 15) to the Air Force requiring it to submit written information on how it will assure (a) that the NRC receives complete, timely, and accurate informa-tion; (b) that appropriate individuals will be fully aware of NRC reporting 15

requirements; and (c) that NRC-licensed activities will be conducted in com-pliance with NRC regulations. This information is necessary for the NRC to determine whether the license should be modified or other enforcement action taken.

Future reports will be made as appropriate.

86-24 Therapeutic Medical Misadministration The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see the general 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 - On October 6-8, 1986, a patient at the Cleveland Clinic Foun-dation, Cleveland, Ohio, received a series of cobalt-60 therapeutic radiation exposures which resulted in a radiation exposure that was about 67 percent greater than the prescribed dosage.

Nature and Probable Consequences - A 58 year-old female patient received two radiation treatments a day for three consecutive days, October 6-8, 1986, for treatment of bone marrow disease. Because of an error in calculating treatment time, these treatments resulted in the patient receiving a radiation dose of approximately 2,000 rads head-to-waist, as opposed to the intended 1,200 rads.

The patient was discharged from the hospital on October 10, 1986, but was re-admitted on October 20, 1986, for symptoms believed to result from the radi-ation exposure (unable to swallow, fever, and chills). She was discharged after treatment, but later admitted to Cleveland Metropolitan Hospital Burn Clinic on November 10, 1986, with skin burns. The patient died on November 18, 1986. l 1

The licensee did not discover the therapeutic treatment error until November 11,  !

1986, when a dosimetrist reviewed the patient's treatment records and checked i the calculations. NRC regulations stipulate that such misadministration be i reported to the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after they are discovered; however, the licensee did not report it to the NRC until November 17, 1986. The delay was apparently due to the licensee's failure to realize that a misadministration of this type requires immediate notification.

A panel of NRC medical consultants reviewed the case and concluded that the radiation treatments had " minimal effect, if any, upon the fatal outcome of her disease." The skin burns were not attributable to the radiation treatment, but rather to a variety of drugs (i.e. , chemotherapy) given to the patient prior' to and in addition to her radiation treatments.

Cause or Causes - The misadministration was caused by an error in the calcula-tions performed to determine the exposure time to deliver the desired radiation dosage. The physicist who performed the calculations used the distance from the cobalt-60 radiation source to the patient, instead of the distance from ,

the exterior of the radiation therapy device to the patient. The physicist 16

entered the measurement into a programmable calculator that already accounted for the internal distance from the radiation source to the exterior of the de-vice. Therefore, the internal distance was added twice with the result that a longer treatment time was scheduled. (The further the source is from the patient, the longer the treatment time required.-)

In 1982 Cleveland Clinic adopted-new procedures as a result of a therapeutic misadministration at that time. These new procedures included a system of l

dual verification of all dose calculations prior to the first day of treatment.

l In this case, however, the procedure was not followed and there was no recheck I

of the physicist's calculations prior to treatment.

Actions Taken to Prevent Recurrence

_ Licensee - The licensee has adopted revisions to its procedures providing that all dose calculations will be independently performed by two qualified individ-uals and that, prior to the first treatment, the technologist will verify that the duplicate calculations have been performed. In addition, the treatment data will be reviewed weekly by the chief technologist. A quality assurance audit by the licensee's Radioisotope Committee is to be performed quarterly for a year and then annually thereafter.

NRC - On November 20, 1986, NRC Region III issued a Confirmatory Action Letter documenting the licensee's agreement to institute the improvements in its procedures listed above (Ref. 16).

A special NRC inspection was conducted beginning November 20, 1986 (Ref. 17).

The inspection identified two violations of NRC requirements, i.e., failure to report the therapeutic misadministration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and failure to obtain approval of the licensee's Radioisotope Committee for physicians to use NRC-licensed materials. (This second violation is not directly related to the misadministration.) On April 15, 1987, the NRC issued a proposed civil penalty of $2,500 which the licensee subsequently paid.

The NRC retained a special medical panel to review the case, consisting of two physicians and a physicist. As previously mentioned, the panel concluded that the patient's deteriorating condition, ending in her death, was not the result of the misadministration.

This item is considered closed for the purposes of this report.

86-25 Suspension of License for Servicing Teletherapy and Radiography Units 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. In addition, Example 11 of "For All Licensees" in Appendix A notes that serious deficiency in management or procedural controls in major areas can be considered an abnor-mal occurrence.

Date and Place - On October 10, 1986, the NRC Office of Inspection and En-forcement issued an order suspending certain NRC-licensed service activities of 17 j I

Advanced Medical Systems, Inc., of Geneva, Ohio'(Ref. 18). This action was taken after the NRC determined that the firm had been using untrained and un-qualified employees to service cobalt-60: teletherapy units.

Nature and Probable Consequences - Advanced Medical Systems (AMS) is licensed l by the NRC to install, service, maintain, and dismantle radiography and tele- i therapy units. (A teletherapy machine contains cobalt-60 and.is used in medi-  ;

cal facilities for the radiation treatment of cancer. A radiography unit con-tains cesium-137 or cobalt-60 and is used to make X-ray like pictures of metal products and welds.) AMS is also licensed to possess cobalt-60 and cesium-137 for manufacturing the sealed sources used in the teletherapy and radiography units. This source manufacturing, which is performed at a separate AMS facility in Cleveland, Ohio, was not affected.by the NRC Order.

By a special safety inspection conducted on September 17 through November 12, 1985 at AMS (Ref. 19), the NRC confirmed allegations that since the Spring of 1985, and as recently as September 1986, licensee employees were directed to perform certain service and maintenance activities on teletherapy. equipment at medical facilities even though these individuals lacked (1) NRC authorization, (2) the required training to perform the directed maintenance, and (3) the-appropriate radiation detection and monitoring equipment or the required service manuals.

Only those AMS technicians who are specifically named on the NRC license, or who were approved by the AMS Safety Committee, may service safety-related components l on a teletherapy unit. Prior to either NRC or AMS Safety Committee approval,

the technician must have had 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of classroom and 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of laboratory training; approximately six months of on-the-job training or- prior related employment; and satisfactorily completed a written examination.

The potential safety consequences of work performed by an unqualified or un-authorized technician is that a faulty repaired or serviced teletherapy unit could expose the AMS repairman, a medical patient, or a hospital tei:hnician to j excessive radiation. 1 Through the course of its inspection efforts, NRC Region III determined that unqualified AMS repairmen had serviced teletherapy units at seven medical insti-tutions in the midwestern and eastern United States. An Order was sent to each institution by the NRC's Office of Inspection and Enforcement requiring the institution to perform full calibration measurements on its teletherapy units prior to the treatment of patients, unless "those calibration requirements have been satisfactorily completed subsequent to maintenance or service by AMS repre- i sentatives." In addition, each institution was required to have its teletherapy 1 unit (s) fully inspected and serviced within 90 days of the Order by technicians other than AMS; until the full inspection and servicing were completed, each licensee was to perform periodic spot-checks of its teletherapy unit (s) every seven days.

The NRC's inspection of AMS also disclosed that the firm has been installing a defective teletherapy unit timer. Operational malfunction of the timer could ,

result in radiation misadministration to patients and possible excessive radiation exposures to teletherapy unit operators. .

18 L

l i1 l

Cause or Causes - The cause of this event was the apparent disregard of NRC's l regulations and requirements by the licensee.

Actions Taken to Prevent Recurrence Licensee - The NRC's October 10, 1986, Suspension Order required AMS to make available to the NRC all employee training records on the servicing of tele-therapy units, all leak test records of sealed cobalt-60 sources, and all invoice and service reports of teletherapy unit maintenance and service work.

AMS was given 20 days from the date of the Order to show-cause why the Order should not have been issued.

On December 23, 1986, the licensee met with NRC Region III officials in Glen Ellyn, Illinois, seeking a rescission of the October 10, 1986 Order. On January 7,1987, the NRC denied the licensee's request, in part because the licensee failed to " provide reasonable assurance of the protection of the public health and safety..."

In a letter dated January 23, 1987, the licensee agreed to perform all work on teletherapy units with only licensed, qualified technicians. AMS also agreed to conduct more frequent field inspections of their employees'. work, including conducting a field audit of an individual's first job, followed by quarterly audits for the next six months, and semiannual audits thereafter. AMS also committed to bringing in an outside consultant semiannually to independently audit their service and repair program.

l NRC - On October 29, 1986, the NRC issued Inspection and Enforcement Bulletin No. 86-04 to all NRC licensees authorized to use cobalt-60 teletherapy units (Ref. 20). The Bulletin directed licensees to instruct their technicians on how to recognize defective timers and the mitigating actions to be taken if a malfunction occurs. It also required the licensees to notify the NRC of the i presence of any defective timer and the corrective actions taken.

On February 2,1987, the Regional Administrator of NRC Region III relaxed the October 10, 1986 Order, following a review of the licensee's stated corrective actions. In addition to the licensee's commitments described above, Region III required AMS to (1) notify a Regional Branch Chief within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a tele-therapy unit service request, (2) provide a description of the work to be done, (3) name the individual assigned to perform the work, and (4) notify the NRC of the date the work is to be performed.

The licensee has asked for a hearing on the Suspension Order. A hearing date before an NRC Administrative Law Judge has not yet been scheduled.

On April 8, 1987, the NRC issued Inspection and Enforcement Information Notice No. 87-18, to caution applicable licensees against using non-licensed mainten-ance personnel to service their teletherapy equipment (Ref. 21).

Future reports will be made as appropriate.

19

86-26 Diagnostic Medical Misadministration The following information pertaining to this event is also being reported con-currently ir 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 abncrmal occurrence.

Date and Place '- On October 21, 1986, a patient at St. Luke's Hospital, Racine, l

Wisconsin, received a whole body iodine-131 diagnostic scan while the intended 4 procedure was to be a thyroid- scan, J Nature and Probable Consequences'- On October 6,1986, a patient received a I diagnostic thyroid scan using iodine-123, an accelerator-produced radioisotope l (accelerator-produced radioisotopes are not subject to NRC regulation, but are underStatejurisdiction). The attending physician then gave oral instructions for an iodine-131 scan because the previous scan was not definitive. The nuclear medicine technologist erroneously arranged for a whole body scan instead of a thyroid scan as intended by the physician. The whole body scan involved 1.53 millicuries of iodine-131, which is approximately 30 times the normal 50 microcurie dosage for a thyroid scan.  ;

i After the scan was performed on October 21, 1986, the attending physician dis- l covered the error. The whole body scan, however, did provide the physiciar. ]

with the diagnostic information desired.

j The radiation exposure, while in excess of that intended, did not result in any immediate medical effects, according to the licensee. Had a typical dosage of iodine-131 for a therapeutic procedure been administered (i.e., 4 to 6 milli-curies), rather than the 1.53 millicuries actually administered, a significant reduction in thyroid activity could have resulted. Thyroid da.nage, however, can be compensated for through the use of medication.

Cause or Causes - The misadministration was caused by the nuclear medicine tech-nologist's misinterpreting the attending physician's oral instruction. The physician requested an " iodine-131 scan," which the technologist incorrectly assunied to be whole body scan. Typically, the licensee uses iodir.e-123 for thyroid scans and iodine-131 for either thyroid scans or whole body scans.

Actions Taken To Prevent Recurrence Licensee - The licensee has revised its procedures for prescribing radiciodine for medical procedures and provided training on the revised procedures. All prescriptions are now to be in written form and will be reviewed by a nuclear medicine physician and verified by the technologist prior to administration of the radiopha'rmaceutical to the patient.

NRC - The NRC ccoducted a special inspection on December 15, 1986, to review the circumstances of the misadministration (Ref. 22). The inspection did not identify any violations of NRC requirements, but determined that improvements were needed in the patient prescription process to preclude similar misadmini-strations in the future.

NRC Region III issued a Confirmatory Action Letter on January 9,1987 (Ref. 23),

20

documenting the licensee's agreement to change its procedures. The changes will be incorporated into the facility's NRC license.

The NRC also retained a medical consultant to evaluate the misadministration and its possible medical effects. The consultant's report is pending.

On April 15, 1987, the NRC issued Information Notice No. 85-61, Supplement 1

(" Misadministration to Patients Undergoing Thyroid Scans") to licensees which described various misadministration and corrective actions taken by some li-censees which have been found effective to prevent such misadministration (Ref. 24). Information Notice No. 85-61, previously issued on July 22, 1985, discussed several other similar misadministration (Ref. 25).

Unless new, significant information becomes available, this item is considered closed for the purposes of this report.

86-27 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 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 November 18, 1986, a patient at Toledo Hospital, Toledo, Dhio, received a misadministration of a radiopharmaceutical when the wrong radio-active material was administered. It is estimated that the patient's thyroid received a dose of about 6,760 rads.

Nature and Probable Consequences - The physician of a 62 year old female patient planned a bone scan for the patient as an outpatient at the Diagnostic Center at Toledo Hospital. The bone scan normally involves a 20 millicurie dose of l

technetium-99m MDP. The hospital's procedures provide that the referring physician's office notify the Diagnostic Center by telephone of the scheduled procedure. The procedure is then scheduled, and the hospital's nuclear medicine department is notified to order the radiopharmaceutical.

In this instance, the physician's office notified the Diagnostic Center, but kept no record of the telephone conversation. The intended procedure was a bone scan, but the Center's receptionist recorded a " total body scan, rule out metastases, carcinoma." This was interpreted by the nuclear medicine depart-ment as an order for a thyroid metastatic disease scan, which is also known as a " total body scan." Toledo Hospital normally uses a 20 millicurie dose of iodine-131 for such a procedure, which is usually performed on patients who have had their thyroid removed. (The organ principally affected by an iodine dose is the thyroid.) The nuclear medicine department confirmed with the Cen-ter's receptionist that the thyroid metastatic disease scan was the prescribed procedure. The receptionist, however, did not verify the procedure with the referring physician's office.

On November 18, 1986, the patient was administered the iodine-131. She returned to the Diagnostic Center the following day and said she was scheduled for a bone 21

)

scan. Since the Center had no bone scan scheduled, the error was consequently discovered. )

i The patient had previously been diagnosed as having mild hypothyroidism (under-active thyroid) and was taking medication to make up for the decreased thyroid function. The iodine-131 dosage was estimated to cause a 6,760 rad dose to the thyroid, while other organs received a rolatively small dose. (A rad is a standard measure of absorbed dose.) Thi; dose to the thyroid is less than would normally be expected for 20 millicuries of iodine-131, because of the i patient's reduced thyroid function. If the patient had had a normally func-I tioning thyroid, the expected dose would have been three to seven times what this patient is estimated to have actually received.

Nevertheless, the 6,760 rad thyroid dose is expected to significantly decrease the patient's thyroid function, necessitating an increase in the medication (thyroxin) the patient was already receiving. The prescribed thyroxin dosage was increased to three times the original prescribed dose. Both the hospital and the patient's physician plan to continue to monitor the patient.

Cause or Causes - The apparent cause of the misadministration was a failure to accurately communicate the prescribed procedure to the hospital's Diagnostic Center. The precise method of failure could not be determined since the pa-tient's physician did not have a record of the telephone conversation in which the procedure was scheduled.

Actions Taken to Prevent Recurrence Licensee - The hospital has instituted a change in its procedures for scheduling outpatient diagnostic doses. All prescriptions for nuclear medicine procedures are to be in written form and reviewed by a nuclear medicine physician and veri-fied by a technologist prior to the administration of the radiopharmaceutical to the patient. i NRC - NRC Region III conducted a special inspection at Toledo Hospital on Novem-  ;

ber 25, 1986, to review the circumstances of the misadministration (Ref. 26). j No violations of NRC requirements were found during the inspection. NRC Re- '

gion III issued a Confirmatory Action Letter to the hospital on November 21, 1986, documenting the hospital's agreement to change its procedures for sched-  ;

uling procedures involving radiopharmaceuticals (Ref. 27). The NRC also re- j tained a medical consultant to review the possible health effects of the  ;

misadministration. 1 Unless new, significant information becomes available, this item is considered closed for the purposes of this report.

86-28 Immediately Effective Order Modifying License and Order to Show Cause Issued to an Industrial Radiography Company The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see the third general subcriter-ion) of this report notes that major deficiencies in management controls for licensed facilities or material can be considered an abnormal occurrence.

22 l

Date and Place - On December 30, 1986, the NRC issued an Order to Met-Chem Test-ing Laboratories of Utah, Inc. of Salt Lake City that in effect prohibits the company from involving a senior management employee in the performance or supervision of any NRC licensed activities (Ref. 28).

I Background - The licensee is the holder of both a general license pursuant to 10 CFR 9150.20 and a specific license (License No. 43-26821-01) pursuant to  ;

10 CFR Part 30, issued by the NRC. The general license authorizes the licensee to conduct the same activity in non-Agreement States pursuant to the provisions of 10 CFR 9150.20 as the licensee is authorized to conduct by its specific license from the State of Utah, an Agreement State. The NRC specific license authorizes the licensee to use the licensed materials in performing industrial 1 radiography and replacing sources, and to use an EON Model 64-764 calibrator (which contains a radioactive source) for calibration of survey instruments at locations where NRC maintains jurisdiction. The NRC license for industrial radiography was issued in July 1986.

Nature and Probable Consequences - The NRC Order was issued to remove the senior vice president from any assignment or position influencing or involving the l performance or supervision of any licensed activities. This action was taken following an NRC investigation initiated in 1985 as a result of inspector obser- i vations made during a routine inspection. The NRC decided to issue the Order i after an NRC investigator obtained a sworn statement on August 21, 1986, from the senior vice president in which he admitted that while employed as the office i manager for the predecessor radiography company (Met-Chem Engineering Labora-tories, Inc.), he had typed a letter and forged on it the signature of a radio-grapher for the purpose of explaining away an overexposure indicated on the radiographer's film badge.

The overexposure, while not clinically significant, was reportable according to NRC regulations. The letter f alsely stated that the radiographer's dosimeter and film badge were left in a shirt pocket and the shirt was placed in an area near a radiation source resulting in an overexposure reading, but not an overexposure to the radiographer himself.

i Had the NRC been provided with correct information, inspection actions regarding the overexposure would have been taken against Met-Chem Engineering Laboratory, Inc., the now defunct former company. Further, had the NRC known that a senior management employee of the licensee had withheld reportable information concern-ing radiation exposures, the specific license for the present company would not have been issued. The false statements made by the senior vice president call into question his candor in dealing with the NRC, and demonstrate that there was no longer reasonable assurance that the licensee would comply with NRC requirements while the individual was involved in licensed activities.

1 Cause or Causes - The employee willfully made false statements to, and withheld I information from, the NRC. On August 13, 1986, the employee denied to an NRC

inspector and an NRC investigator any knowledge of how the forged letter was generated. However, on August 21, 1986, he admitted that he had indeed l generated, and signed, the letter.  ;

l The employee has stated that the reasons he wrote the forged letter were (1) he i

23 a-_- _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _

did not want anything to stop the sale of certain Met-Chem Engineering Labora-tories, Inc. properties to a third party, and (2) he did not want the NRC to know about the overexposure because he believed it would not have been desir-able to have the NRC looking into the matter during the sale negotiation period.

1 Actions Taken to Prevent Recurrence Licensee - The licensee responded to the NRC Order on January 15, 1987. The licensee stated that the employee terminated employment at Met-Chem Testing Laboratories during November 1986, to accept employment with a company which j neither has a radioactive materials license nor handles any radioactive materials.

The licensee held meetings with all authorized users of radioactive materials to restate the instructions they are givet, during training, which includes total I compliance to NRC requirements and to be honest and cooperate totally with NRC personnel. On January 5,1986, the authorized users of radioactive materials signed a statement that they have read and understand the December 31, 1986 NRC Order.

NRC - The NRC Order contained the following provisions, effective immediately-  ;

i (1) License No. 43-26821-01 is amendec by adding the following condition:

The employee shall be removed from any assignment or position influencing or involving the performance or supervision of any licensed activities I (e.g., as an authorized user), including the supervision of any Radiation l Safety Officer (RS0). 1 q

(2) The licensee shall show cause in the manner hereinafter provided why the 1 license amendment set out in paragraph (1) above should not become permanent. )

(3) The employee shall be removed from any assignment or position influencing or involving the performance or supervision of any licensed activities permitted under the general license issued pursuant to 10 CFR K150.20.

1 (4) The licensee shall show cause in the manner hereinafter provided why the provisions in paragraph (3) above should not become permanent.

(5)

Prior to conducting the licensee (any shall a) licensed notify activities in writing after receipt all personnel of thisinOrder, involved the performance and supervision of licensed activities at Met-Chem Testing Laboratories of Utah, Inc. of this Order and the importance of strict adherence to NRC requirements and complete candor with NRC personnel, and (b) certify to the NRC that each Authorized User and RSO has read the notification and Order and understands its contents.

(6) The NRC Region IV Regional Administrator may relax or rescind any of the above provisions for good cause shown by the licensee.

The NRC is evaluating the licensee's response to the Order, to determine whether it is satisfactory, and/or whether further enforcement action is required.

24

i 3 s.

a ..

\

Future reports will be made as appropriate.

• x**.****** 1 i

AGREEMENT STATE LICENSEES-1 Procedures have been developed for the Agreelnent States to screen unscheduled.

incidents or events using the same criteria as the NRC (See Appendix A) and. ,

report the' events _to the NRC for inclusion in this report. During the fourth  !

calendar quarter of 1986, the Agreement States reported no abnormal occurrences to the NRC.  ;

t 1

25

REFERENCES

1. Letter from Virgil L. Brownlee, Chief, Reactor Projects Branch 3, Division of Reactor Projects, NRC Region II, to H. B. Tucker, Vice President, Nuclear Production Department, Duke Power Company, forwarding Inspection Report Nos. 50-269/86-26, 50-270/86-26, and 50-28'/86-26, Docket Nos. 50-269, 270, and 287, October 23, 1986.*

2. Letter from J. Nelson Grace, Regional Administrator, NRC Region II, to H. B. Tucker, Vice President, Nuclear Production Department, Duke Power Company, forwarding Inspection Report Nos. 50-269/86-33, 50-270/86-33, and 50-287/86-33, Docket Nos. 50-269, 270, and 287, December 1, 1986.*

3. Letter from J. Nelson Grace, Regional Administrator, NRC Region II, to H. B. Tucker, Vice President, Nuclear Production Department, Duke Power Company, forwarding " Meeting Summary - Report Nos. 50-269/87-07, 50-270/87-07, and 50-287/87-07," Docket Nos. 50-267, 270, and 287, February 11, 1987.*

4. Letter from J. Nelson Grace, Regional Administrator, NRC Regicn II, to H. B. Tucker, Vice President, Nuclear Production Department, Deke Power Compan;, forwarding a Notice of Violation, Docket Nos. 50-269, 50-270, and 50 287, February 5,1987.*

5. U.S. Nuclear Pegulatory Commission, Inspection and Enforcement Information Notice No. 87-06, " Loss of Suction to Low Pressure Service Water System Pumps Resulting from Loss of Siphon," January 30, 1987.*

6. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 85-03, " Motor-0perated Valve Common Mode Failures During Plant Tran-sients Due to Improper Switch Settings," November 15, 1985

7. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-93, " Inspection and Enforcement Bulletin No. 85-03 Evaluation of Motor Operators Identifies Improper Torque Switch Settings," November 3, 1986.*

8. Letter from R. F. Saunders, Station Manager, Virginia Electric and Power Company, to U.S. Nuclear Regulatory Commission, Document Control Desk, forwarding (1) Licensee Event Report No. 86-20-01; and (2) "Surry Unit 2, Reactor Trip and Feedwater Pipe Failure Report," Revision No. 0, dated January 14, 1987; Docket No. 50-281, January 14, 1987.*

• Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).

27

9. Letter from J. Nelson Grace, Regional Administrator, NRC Region II, tn W. L. Stewart, Vice President, Nuclear Operations, Virginia Electric and Power Company, forwarding (1) Notice of Violation; and (2) NRC Augmented Inspection Team Report Nos. 50-280/86-42 and 50-281/86-42; Docket Nos. 50-280 and 50-281, February 10, 1987.*

10. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No.86-106, "Feedwater Line Break," December 16, 1986.*

11. Supplement 1 to Information Notice No.86-106 issued February 13, 1987.*

12. Supplement 2 to In'ormation Notice No.86-106 issued March 18, 1987.*

13. Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to Colonel Frederick Bode, Commanding Officer, Wright Patterson Air Force Base, License No. 34-00472-02, February 5, 1987.*

14. Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to Colonel Charles Fox, Jr. , Base Commander, Wright Patterson Air Force Base, License No. 34-00472-02 October 27, 1986.*

15. Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to Colonel Bryant D. Mauk, Chairman, USAF Radioisotope Com-mittee, forwarding a Demand for Information, Docket No. 030-28641, License No. 42-23539-01AF, February 19, 1987.*

16. Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to Frank Thomas, M.D., Cleveland Clinic Foundation, License No. 34-00466-02, Docket No. 30-00394, November 20, 1986.*

17. Letter from Jack A. Hind, Director, Division of Radiation Safety and Safeguards, NRC Region III, to William S. Kiser, Chairman, Board of Governors, Cleveland Clinic Foundation, forwarding Inspection Report No. 86-01, Docket No. 30-00394, February 12, 1987.*

18. Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to Seymour S. Stein, President, Advanced Medical Systems, Inc., forwarding an Order Suspending License and To Show Cause (Effective Immediately), License No. 34-19089-01, Docket No. 30-16055, October 10, 1986.*

19. Letter from Jack A. Hind, Director, DSision of Radiation Safety and Safeguards Section, NRC Region III, to S. S. Stein, President, Advanced MedicalSystems,Inc.,forwardingInspectionReportNo.86-01, No. 30-16055, November 25, 1986 Docket

20. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 86-04, " Defective Teletherapy Timer That May Not Terminate Treatment Dose," October 29, 1986.*

• Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).

28

21. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 87-18, " Unauthorized Service on Teletherapy Units by Non-Licensed Maintenance Personnel," April 8, 1987.*

22. Letter from W. L. Axelson, Chief, Nuclear Materials Safety Section 2, NRC Region III, to Ray L. Dilulio, President, St. Luke's Hospital, forwarding Inspection Report No. 86-01, Docket Nc. 030-03425, January 15, 1987.*

23. Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to Ray L. DiIulio, President, St. Luke's Hospital, License No. 48-02096-01, Docket No. 030-03425, January 9, 1987.*

24. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 85-61, Supplement 1, " Misadministration to Patients Undergoing Thyroid Scans," April 15, 1987.*

j 25. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 85-61, " Misadministration to Patients Undergoing Thyroid Scans,"

July 22, 1985.*

26. Letter from W. L. Axelson, Chief, Nuclear Materials Safety and Safeguards Branch, NRC Region III, to William Jeffries, Radiation Safety Officer, Toledo Hospital, forwarding NRC Inspection Report No. 86-01, Docket No. 030-02685, January 9, 1987.*

27. Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to William F. Jeffries, Radiation Safety Officer, Toledo Hospital, License No. 34-01710-05, Docket No. 030-02685, November 21, 1986.*

28. Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to J. W. Lewis, Chairman of the Board, Met-Chem Testing l Laboratories of Utah, Inc., forwarding an Order Modifying Licen and Order to Show Cause (Effective Immediately), Licensee No. 43-26821-01, Docket No. 30-29056, December 30, 1986.*.

• Available in NRC Public Document Room, 1717 H Street, NW, Washir.gton, DC 20555, foo inspection and copying (for a fee).

29

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-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 cancentra-tions which, if averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 500 times the regulatory limit of Appendix B, Table II, 10 CFR Part 20 (10 CFR S20.403(b)).

4. Radiation or contamination levels in excess of design values on packages, or loss of confinement of radioactive mahrial 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 sucn 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.

31

7. Any substantiated loss of special nuclear material er 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 accountability system.

8. Any substantial breakdown of physical security or material control (i.e. ,

access control, containment, or accountability systems) that significantly weakened the protection against theft, diversion, or sabotage.

9. An accidental criticality (10 CFR 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 S50.36(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.

32

APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the October through December 1986 period, the NRC, NRC licensees, Agree-ment States, Agreement State Licensees, and other involved parties, such as reactor vendors and architects and engineers, continued with the implementation of actions necessery to prevent recurrence of previously reported abnormal occur-rences. The referenced Congressional abnormal occurrence reports below provide the initial and any vodating information on the abnormal occurrences discussed.

The updating provided generally covers' events which took place during the re-port 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 Qu lification of Safety-Related Electrical Equipment Inside Containmen.

l 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; /ol. 3, No. 2; Vol. 4, No. 2; Vol. 5, No. 2. Vol. 6, No.1; and closed out in Vol. 8, No. 2. It is being reopened to report the following new information; the information is current as of the end of 1986 l

The NRC is currently conducting an Environmental Qualification (EQ) Inspection Program which consists of inspections at utility engineering offices and nuclear plant sites to evaluate implementation of EQ programs in compliance with 10 CFR K50.49. The program also includes inspections at EQ test laboratories and quaiified equipment manufacturers' test facilities to evaluate test methods and practices, and verify compliance with quality assurance, EQ technical, and other regulatory requirements.

During the course of this inspection series, the NRC discovered a number of instances in which heat shrinkable electrical insulation tubing manufactured by Raychem, the principal supplier of this type of nuclear environmentally qualified insulation system, was improperly installed. The tubing is primarily used to frsulate and environmentally seal electrical connections which may be subjected to a harsh environment during postulated plant accidents such as loss of coolant accidents (LOCA) or high energy line breaks (HELB). Raychem tubing was found ,

installed in configurations which deviated from the configurations which had been tested and qualified. The improper installations were found in use with various qualified components in potentially harsh environmental zones both in-side and outside containment. In most instances, the deficiencies would not be expected to cause equipment failure under accident conditions. However, the full extent of safety system degradation could not be accurately assessed.

As a result of these discoveries, the NRC issued Inspection and Enforcement Information Notice No. 86-53 on June 26,1986, (Ref. B-1), which alerted remain-ing plants of the problem. Corrective actions are in progress or have been completed at plants where Raychem deficiencies have been identified and the EQ inspection program routinely examines this issue to verify that the utilities 33

are effectively conducting their own inspections and taking appropriate correc-tive actions. Additionally, the NRC has issued a Temporary Instruction for use by NRC regional inspectors and resident inspectors to expedite verification that all plants are finding and correcting these deficiencies.

In an effort to reduce the number of required repairs, and to show that most installed splice configurations are qualifiable as they are, several utilities have sponsored additional EQ testing. This testing has qualified a much wider range of splice configurations than was originally qualified by Raychem and has significantly reduced the scope of the problem. Pursuit of this issue will continue until resolved at all plants and effective measures to prevent recurrence are implemented.

During the EQ inspection at Commonwealth Edison Company's (CECO) Dresden Nuclear Power Station, NRC inspectors found that General Electric type F-01 Electrical Penetration Assemblies contained wire butt splices using insulated crimp connec-tors made by the Amp Company. The splice insulation sleeves were made of nylon material with open ends which did not seal tightly to the wire insulation.

Prompted by HRC questioning of the applicability of CECO's EQ test documentation to the installed configurations of the splices, CECO sponsored additional EQ testing to qualify the splices for their applications. In these tests, the Amp connectors failed by causing short circuits due to degradation of the nylon insulation. The tests indicated that these open end nylon insulated Amp butt splice connectors were unqualified for their identified harsh environmental applications.

Connectors of this type were found in similar applications at CEC 0's Quad Cities Nuclear Power Station and Iowa Electric's Duane Arnold Energy Center. In all cases, licensees have taken prompt action in effecting repairs using qualified tape or heat shrinkable insulation tubing and sealing methods. To alert other licensees to this problem, NRC issued Inspection and Enforcement Information Notice No.86-104 on December 16, 1986 (Ref. B-2). The actions of licensees in response to the Information Notice will be monitored by regional and resi-dent inspectors and will be addressed during the future inspections in the ongoing EQ Inspection Program.

Unless new, significant information becomes available, this item is considered closed for the purposes of this report.

79-3 Nuclear Accident at Three Mile Island This abnormal occurrence was originally reported in NUREG-0090, Vol. 2, No.1,

" Report to Congress on Abnormal Occurrences: January-March 1979," and updated in each subsequent report in this series, i.e., NUREG-0090, Vol. 2, No. 2 through Vol. 9, No. 3. It is further updated for this report period as follows.

Reactor Building Entries During the fourth calendar quarter of 1986, 88 entries were made into the TMI-2 reactor building, bringing the total number of entries since the March 1979 accident to 1135. Entries are currently made by several crews each day to per-form defueling and supporting tasks. Specific reactor building activi'cies 34

conducted during this period included core drilling using the core boring machine as a defueling tool, mapping and video examinations of the core debris bed, removal of end fittings and broken drill strings, robotic decontamination of basement walls, and efforts to improve water clarity in the reactor coolant system.

Reactor Vessel Defueling Operations In October 1986, the drilling equipment that had been used earlier for core sampling activities was modified and reinstalled on the defueling platform.

Full-scale drilling operations were conducted between late October and late November. A total of 425 holes were drilled into the hard crust region of the core to condition this material for subsequent removal. The entire cross sec-tional area of the core, except for a 2 foot ring at the periphery, was drilled to a depth of between 1\ and 4 feet. While this effort was successful in break-ing up the hard crust, the subsequent removal of the resulting debris proved difficult, due to the number of relatively large pieces of core material remain-ing in the vessel. Consequently, defueling progress during the quarter was limited. Through December 1986,.approximately 61,000 pounds (20%) of the core debris had been removed from the reactor vessel.

In late November 1986, the licensee, General Public Utilities Nuclear Corporation (GPUNC), performed topographic mapping and video examinations of the core debris bed to develop more effective methods of removing the core debris generated by the drilling operation. Concurrent with these activities, GPUNC conducted numerous tests on coagulant additives intended to improve filter performance in the defueling water cleanup system (DWCS). By early January 1987, an effective coagulant compound was identified and was added to the reactor vessel water to be used in conjunction with a diatomaceous earth filter aid in the DWCS. This system has greatly improved the DWCS filter performance and is expected to main-tain good water clarity for the duration of defueling activities. Additionally, new tooling and techniques appear to be effective for removal of the larger pieces of core debris. Through the end of January 1987, 72,000 pounds (24%) of core debris had been removed from the reactor vessel, an indication that the pace of defueling has increased.

Decontamination and Waste Disposal Activities In early December 1986, GPUNC conducted a decontamination experiment in the reactor building basement. A high pressure hydrolazer mounted on a robotic vehicle was used to scarify the concrete basement walls. The licensee is continuing to examine the use of robotics technology in decontamination applications.

During the fourth quarter of 1986, the licensee continued to apply proven decon-tamination techniques in the auxiliary and fuel handling building (AFHB). By year's end, decontamination of approximately 66% of the contaminated surface area in the AFHB had been completed. The licensee expects to complete AFHB decontamination during CY 1987.

During the fourth quarter of 1986, 10 EPICOR-II spent resin liners were shipped offsite to Hanford, WA for disposal as low-level waste.

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In December 1986 and January 1987, two additional shipments of core debris were transported by rail to the Idaho National Engineering Laboratory (INEL) for examination and interim storage. A total of 48,000 pounds of core debris has been shipped to INEL to date.

In December 1986, the NRC staff issued, for public comment, draft Supplement No. 2 to NUREG-0683, " Programmatic Environmental Impact Statement (PEIS)" for the decontamination of TMI-2 (Ref. B-3), which deals with the disposal of slightly contaminated accident generated water (AGW). Ten alternative methods of disposal were evaluated in some detail in the draft supplement, including the licensee's proposal for forced evaporation of the water with disposal of the solidified residue at a low-level waste burial site. The potential environ-mental impacts for all ten alternatives evaluated were determined to be very small. Following receipt of comments on the draft supplement, the NRC staff will issue a final supplement to the PEIS and provide a recommendation to the Commission. The Commission will ultimately approve the method of disposal for the AGW.

TMI-2 Advisory Panel Meetings The Advisory Panel for the Decontamination of TMI-2 met on October 8 and December 10, 1986, in Harrisburg, Pennsylvania. At the October meeting the Panel was briefed by the licensee on the progress of defueling and the status l of funding for the remainder of the cleanup. The Environmental Protection i Agency presented the results of its review of the TMI offsite radiation monitor-ing program. At the December meeting, the Panel was briefed by the licensee on their proposal for the Post-Defueling Monitored Storage of TMI-2.

Future reports will be made as appropriate.

86-15 Differential Pressure Switch Problem in Safety Systems at LaSalle Facility This abnormal occurrence was originally reported in NUREG-0090, Vol. 9, No. 3,

" Report to Congress on Abnormal Occurrences: July-September 1986." It is updated through December 1986 as follows.

On December 18, 1986, Commonwealth Edison Company (the licensee) reported that five S0R, Incorporated (formerly the Static "0" Ring Pressure Switch Company) differential pressure switches utilized in safety-related applications had failed in the prevfous 14-month period at LaSalle. Prior to this report, the licensee had considered the failures to be " random failures". This new report indicates that all five failures were related to failures of the diaphragm in the switches (i.e., pin hole leaks or tears) which raises reliability and design concerns.

Failures that occurred prior to the June 1, 1986 event were found during normal surveillance testing. Failures af ter June 1986 have been found during the aug-mented test program initiated following the June 1, 1986 event. Systems and components in which failures have occurred include: Unit 2 ADS (automatic de-pressurization system) permissive (Level 3); Unit 2 RCIC (reactor core isolation cooling) Line Break Protector; and Unit 1 RHR (residual heat removal) mini-flow 36

alarm. The root cause of the switch diaphragm failures is still under investigation by the licensee and the switch manufacturer.

On April 1,1987, the NRC staff issued a safety evaluation report to the licen-see approving its evaluation of test results to justify continued operation of both Units 1 and 2 (Ref. 8-4). Additionally, the long-term action plan for com-pensating for erratic differential pressure switch operation was also approved.

The NRC staff is continuing to monitor the performance of the SOR differential pressure switches in accordance with the approved lor.g-term action plan.

Future reports will be made os appropriate.

• • • • • ++***

FUEL CYCLE FACILITIES 86-3 Rupture of Uranium Hexafluoride Cylinder and Release of Gases This abnormal occurrence, involving Sequoyah fuels Corporation, Gore, Oklahoma, was originally reported in NUREG-0090, Vol. 9, No.1. " Report to Congress on Abnormal Occurrences: January-March 1986," and updated in subsequent reports in this series, i.e., NUREG-0090, Vol. 9, No. 2 and Vol. 9, No. 3. Some other events involving overfilled uranium hexafluoride (UF ) cylinders at Allied-Signal Cor-paration'sfacilitiesinMetropolis,IllinoiswereaIsodiscussedasanAnnexto the abnormal occurrence in NUREG-0090, Vol. 9, No. 1. The abnormal occurrence and the Annex are updated through mid-March 1987, as follows.

Sequoyah Fuels Corporation (SFC)

As discussed in the previous report (i.e., NUREG-0090, Vol. 9, No. 3), on November 14, 1986 the NRC authorized SFC to restart production activities.

The NRC provided 24-hour oversight of restart and production activities, and continued the requirement for an independent oversight organization to provide a third-party examination of operations.

On December 15, 1986, the licensee began filling the first UF 6 cylinder (14 ton) which was completed on December 17. The remodeled steam chest was successfully used on December 22 to heat a cylinder containing out-of-specification UF in order to remove the product for reprocessing. BylateDecember,SFCwashper-ating at a production rate of 300 metric tons of uranium (mtu) per month. The independent oversight organization released two reports to SFC staff describing its activities, findings, and recommendations.

Although minor equipment problems were encountered as each piece of equipment was started up, SFC believes that they have not had as many problems as they originally anticipated.

Continuing into early 1987, no significant safety concerns have

• een noted, and SFC continued to operate without incident. On February 28, 1986, NRC Region IV reduced their oversight activities from a 24-hour-per-day basis to one shift per day, seven days a week. On February 24, 1987, SFC requested a reduction in the coverage by the independent oversight organization. The NRC responded to SFC's request on March 18, 1987, denying a reduction in independent oversight 37

coverage until the licensee submitted additional infordtion regarding oversight reduction criteria and the status of quality assurance coverage at the facility.

Further information has been submitted by the licensee in response to this re-quirement and is under review by the NRC.

As discussed in NUREG-0090, Vol. 9, No. 3, on October 14, 1986 the NRC issued to SFC a Notice of Violation and Proposed Imposition of Civil Penalties in the amount of $310,000 (Ref. B-5). The violations which were directly associated with the January 4, 1986 accident were categorized as a Severity Level I problem (on r scale in which Levels I and V are the most and the least significant, respectively) and accounted * * $300,000 of the proposed civil penalty.

On November 13, 1986, the licensee responded by presenting a number of arguments protesting the civil penalties and in support of remission or mitigation of the civil penalties. The NRC has reviewed the licensee's response and concluded that the alleged violations did occur as stated in the Notice of Violation and no mitigation of the civil penalties is warranted. Therefore, on February S, 1987, the NRC issued to SFC an Order Imposing Civil Monetary Penalties in the amount of $310,000 (Ref. B-6). In early March 1987, the licensee paid the civil penalty in full.

Annex Update: Allied-Signal Corporation On November 3-7, 1986, representatives of the NRC Office of Nuclear Material Safety and Safeguards, NRC Region II, and the Illinois Department of Nuclear Safety conducted an announced team inspection at the Allied-Signal facility in Metropolis, Illinois. The inspection focused on completion of open items iden-tified in previous inspections; comparison of actions taken at Allied with those taken at SFC; and Allied responses to the Lessons-Learned Recommendations. [As discussed in previous reports, the Lessons-Learned Recommendations were developed by a Lessons-Learned Group, formed by the NRC Acting Executive Director for Operations on February 20, 1986, with a goal to identify actions to prevent incidents similar to the January 4, 1986 SFC accident and to improve response /

follow-on activities by licensees and regulatory agencies. The Group's report was issued during June 1986 as NUREG-1198 (Ref. B-7).] The NRC is continuing to follow the licensee's implementation of the Lessons-Learned Recommendations.

The results of the meeting on November1986. 7,sk.ction +ere discussed The inspection report with was Allied management formally sent to the in an exit licensee on November 18, 1986 (Ref. B-8). The inspection identified one apparent violation (regarding the frequency of monthly training of the fire brigade emergency response team) and various open items to be resolved. The NRC forward-ing letter noted that Allied's Hazard Review Committee, established during the first half of 1986, has made considerable progress in identifying and resolving safety issues.

As discussed previously in NUREG-0090, Vol. 9, No. 1, on June 27, 1986, the NRC forwarded to the licensee a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $25,000 (Ref. B-9). The violations included failing to report a December 7, 1984 incident (involving overfilling of, and subsequent damage to an uranium hexafluoride cylinder; however, there were no releases of gases or personnel injuries involved) to the NRC, and three instances of failing to follow procedures during the March 23, 1986 overfill incident.

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On August 21, 1986, the licensee forwarded a check for $25,000; however, the licensee questioned the assignment of Severity Level III to the collective vio-lations, as well as the magnitude of the proposed civil penalty. The NRC re-viewed the licensee's response and in a December 19, 1986 letter (Ref. B-10) concluded that a category of Severity Level III was proper and that no mitiga-tion of the civil penalty was warranted.

Unless new, significant information becomes available, the SFC abnormal occur-rence, as well as the Allied-Signal Corporation incident, are considered closed for the purposes of this report.

OTHER NRC LICENSEES 86-10 Wijlful Failure to Report a Diagnostic Medical Misadministration.

This abnormal occurrence was originally reported in NUREG-0090, Vol. 9, No. 2,

" Report to Congress on Abnormal Occurrences: April-June 1986." It is updated as follows.

As discussed in the previous report, on June 17, 1986, the NRC forwarded to Mercy Hospital of Wilkes-Barre, Pennsylvania (1) an Order requiring the licensee to show cause why the Chief Nuclear Medicine Technician and the Radiation Safety Officer (RS0) should not be prohibited from the performance or supervision of any licensed activities, and (2) a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $5,000 (Ref. B-11). The reason was that the

individuals willfully failed to report a diagnostic medical misadministration l

to the NRC as required by 10 CFR 35.43. In addition, the RS0 at Mercy Hospital

is also listed as an authorized user of NRC licensed material on the license of Valley Radiology Associates, Inc., Kingston, Pennsylvania. Therefore, on June 17, 1986, the NRC issued a similar Order to this licensee (Ref. B-12).

The technologist involved at Mercy Hospital no longer works in the field of Nuclear Medicine. On July 15, 1986, NRC Region I, with the consent of the licensee, issued an amendment to the Mercy Hospital license which did not in-clude the RS0 involved. While the individual will continue to work at the hos-pital, his activities will be under the supervision of an authorized user. He will not manage the Nuclear Medicine program and is no longer the RSO. Region I has agreed to reconsider adding the individual as an authorized user, if the hospital formally requests it, after a period of one year. On October 17, 1986, the licensee sent a check in full payment of the $5,000 civil penalty. On December 24, 1986, the NRC found that the licensee's corrective actions pro-vided adequate cause why the license should not be modified (Ref. B-13).

On December 15, 1986, the license of Valley Radiology Associates, Inc. was amended to delete the same individual as an authorized user, with the consent of the licensee. On December 24, 1986, the NRC found that the licensee's corrective actions provided adequate cause why the license should not be modified (Ref. B-14).

This item is considered closed for the purposes of this report.

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APPENDIX C OTiiER 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 abncrmal occurrences.

1. Diesel Generator Problems During this report period, several significant events involving problems with emergency diesel generator (EDG) units occurred at various nuclear power plant sites. EDG units are provided as a source of electric power for safety-related equipment in the event of an accident situation in which a total loss of offsite power also occurs. The events are described below. These events are not iso-lated incidents, but are indicative of continuing problems due to such causes as EDG design, maintenance, etc.

Gas Leakage into the Jacket Water Cooling System During September 1986 it was reported by Baltimore Gas and Electric Company that diesel engine No. 12 at the Calvert Cliffs Nuclear Power Plant (NPP)* was expe-riencing leakage of gases into the jacket water cooling system (JWCS). Excess gas in the JWCS can result in cavitation of the jacket water pump, low JWCS pressure, and subsequent diesel engine trip. The problem was initially observc4 at a very low level of leakage in September 1985.

After extensive investigation and analysis of the potential sources of leakage, including the installation of additional instrumentation and provisions for i venting, the licensee in conjunction with vendor representatives identified l adapter penetration seals and a turbocharger air cooler (one of two) as the sources of gas inleakage. There are four adapter penetrations in each cylinder that provide for fuel injection, starting air input, and a spare. A cracked cylinder liner was also found. All cylinder liners, adapter seals, and the faulty air cooler were replaced and the engine was operating satisfactorily as of mid-December 1986.

The diesel engine manufacturer, Colt /Fairbanks Morse Engine Division, has issued a special manual, " Maintenance and Surveillance Testing Program," for nuclear applications (Rev. 1, 7/18/86). The manual calls for leak testing adapter seals at intervals no greater than 18 months. The vendor has also recommended that the cylinder liners be replaced every 10 years.

Failure of a Connecting Rod On October 24, 1986, Commonwealth Edison Company reported the failure of a con-necting rod in diesel engine No. 1B at the Zion NPP.** The connecting rod had

• Calvert Cliffs NPP is a two Unit facility located in Calvert County, Maryland.

Both Units utilize Combustion Engineering-designed pressurized water reactors.

• • Zion NPP is a two Unit facility located in Lake County, Illinois. Both Units utilize Westinghouse-designed pressurized water reactors.

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been thrown through a large hole it made in the engine block, along with oil, water, and various other parts of the engine.

Prior to the failure, the diesel engine had been undergoing preventive mainte-nance testing. The testing showed slightly low compression in three cylinders (cylinders 4L, SR, and 7L). To correct this problem, the heads on these three cylinders were removed, and the pistons, rings, cylinder liners, and gaskets were replaced. The engine was reassembled and post-maintenance testing was initiated. After the engine had operated for approximately one and one-half hours at 1000 kilowatts (kW) load (25 percent of full capacity), loud knocking noises were heard that escalated rapidly to massive engine failure.

Review of maintenance and reassembly activities revealed that the bolts that fasten the articulated connecting rod to its rod pin were torqued to 690 foot-pounds (ft-lb) instead of the required 1140 ft-lb. Inadequate torquing of these bolts resulted in a gradual loosening and failure of the bolts.

The engine was repaired in place with the assistance of the engine manufacturer and consultants. The engine completed its 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test run satisfactorily on February 18, 1987, and is available for service. NRC issued two notices of violation to the licensee as a result of this engine failure.

Failure of Connecting Rods On December 23, 1986, Arizona Public Service Company reported a catastrophic failure of two connecting rods (rods 9R and 9L) in diesel engine No. 3B at the Palo Verde Nuclear Generating Station, Unit No. 3.*

i At the time of the failure, the No. 3B diesel engine had operated about l eight minutes of a scheduled two hour run at 110 percent of full power. The l failure resulted in the connecting rods, pistons, counterweight and other parts being ejected from the engine. Even though the fuel oil supply to the engine was terminated when the engine was tripped, the combustion process was sustained in the remaining 18 cylinders by the hot lubricating oil mist in the crankcase mixing with air and being drawn into the intake manifold through the now empty 9R and 9L cylinders.

The engine continued to run at approximately 250 rpm (compared to a normal ro-tational speed of about 600 rpm) for about one hour. Operation er finally stopped by the plant fire brigade by spraying fire suppression foam into the lubricating oil sump through the hole in the side of the engine.

Investigation and analysis of the event resulted in the conclusion that the cause of the connecting rod failure was stress concentration at the oil hole between the main and articulating connecting rod pin bores aggravated by the composition and thickness of the iron plating. The rod pin bores had been over-machined during manufacture and the correct diameters were restored by the use of iron plating of 50 to 60 mils in thickness. Unfortunately, the plating

• Palo Verde Nuclear Generating Station is a three Unit facility (as of late March 1987, Unit 3 was not yet operational) located in Maricopa County, Arizona.

All three Units utilize Combustion Engineering-designed pressurized water reactors.

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extended into the oil holes joining the rod pin bores and was not removed to restore the required radius. This, in combination with the iron plating's columnar grain structure which has a tendency for cracking, and the area of stress concentration caused by the oil holes, resulted in cracking of the iron plating and the parent metal and ultimate catastrophic failure of the engine.

The engine manufacturer, Cooper Energy Services, reviewed its manufacturing records and determined that a total of four other connecting rod pin bores were overbored and iron plated in this manner. Two (including the failed rod) were installed in Palo Verde engine No. 3B, one was installed in Palo Verde engine No. 2A, and one was installed at Commonwealth Edison Company's Byron Nuclear Power Station.*

All of the affected connecting rods were removed and replaced with correctly machined connecting rods. In addition, the manufacturer has notified the licensees for two other plants that their diesel engines contain five connecting rods with iron plating but not in areas that are subject to high stress loads.

Four of the five connecting rods are located at Corr.monwealth Edison Company's Braidwood NPP** and one at Niagara Mohawk Power Company's Nine Mile Point Unit 2t. The connecting rod has been replaced at Nine Mile Point Unit 2. The licensee for Braidwood is determining the corrective action to be taken.

Cracked Heads on Diesel Engines On October 7, 1986, during a station refueling outage, Nebraska Public Power District discovered while conducting an annual inspection of station EDG No. 2 at Cooper Nuclear Stationtt that 12 of 16 cylinder heads were cracked.

! The cracked heads were the result of the use of heads originally designed to accommodate fuel consisting of 94 percent natural gas and 6 percent diesel fuel l oil. The heads contain extra passageways for higher volume gas flow which resulted in a smaller stress margin to withstand engine upset conditions (e.g. ,

primarily engine cooldown and incorrect fuel timing). The heads were of the gas / diesel configuration considered in 1970 when the engines were manufactured.

As of mid-February 1987 all 16 cylinder heads had been magnaflux tested and hydrostatically tested with the result that 10 heads were replaced with heads of a new design, 4 heads were found crack-free and were returned to the engine, and 2 heads were replaced with crack-free heads of the old design.

• Byron Nuclear Power Station is a two Unit facility located in Ogle County, Illinois. Both Units utilize Westinghouse-designed pressurized water reactors.

• • Braidwood NPP is a two Unit facility located in Will County, Illinois. As of late March 1986, Unit I had a zero power license and Unit 2 was still under construction. Both Units utilize Westinghouse-designed pressurized water reactors.

TNine Mile Point is a two Unit facility located in Oswego County, New York.

Both Units utilize General Electric-designed boiling water reactors.

ftCooper Nuclear Station is a single Unit facility located in Nemaha County, Nebraska. The Unit utilizes a General Electric-designed boiling water reactor.

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Station EDG No. I was also inspected. All of the cylinder heads were crack-free and were returned to the engine. The licensee will replace all heads of the old design for both engines with heads of the new design at the earliest prac-tical opportunity. The manufacturer of these diesel engines, Cooper Energy Services, has reviewed its records and has determined that the only other engines subject to this potential problem are the five engines at the Zion NPP. However, all of the cylinder heads for those engines have already been replaced several years ago with heads of the new design.

2. NRC Augmented Inspection Team Sent to Hope Creek On September 24, 1986, NRC Region I sent an Augmented Inspection Team (AIT) to the Hope Creek Nuclear Power Plant to investigate numerous equipment problems and test anomalies experienced during loss of offsite power (LOP) tests performed on September 11 and 19, 1986. Hope Creek, operated by Public Service Electric and Gas Company (the licensee), utilizes a General Electric-designed boiling water reactor. The plant is located in Salem County, New Jersey. The plant first achieved criticality on June 28, 1986 and first generated electricity on August 1, 1986.

During Se pember 1986, the plant was in the power ascension tes_ing program.

An important part of the program is the LOP test. Its purpose is to demonstrate whether the plant response is satisfactory and in accordance with the plant design for concurrent loss of the turbine generator and all offsite power sources.

The LOP test was initiated on September 11, 1986 from about 21.5% power and l with the main turbine generator loaded to 165 MWe. The first indication of an

! unsatisfactory plant response was the failure of the "C" emergency diesel gen-I erator (EDG) output breaker to close automatically. Soon after, an observed failure of the reactor auxiliary cooling system coincident with increasing dry-well pressure resulted in the test being aborted by the licensee. The reactor was scrammed. Normal offsite power was then manually restored to the station.

Twenty-four observations were made by the licensee during this test. These observations occurred during the time from initiation of the test until the reactor vessel water level and pressure were controlled and the reactor scram was reset.

The most significant observations on September 11, 1986 were: (1) EDG "C" out-put breaker failed to close; (2) main steam relief valve position indication was lost; (3) power supplies for the source and intermediate range neutron de-tector drives and main steam line acoustic monitors were lost; (4) 17 control rods did not provide a normal full-in position indication; (5) reactor' auxiliary cooling system flow was lost; (6) EDGs "A" and "B" governors transferred frcm isochronous (frequency control) to speed droop (load control) mode without op-erator action; and (7) the "B" safety auxiliary cooling system pump failed to auto-start.

On September 19, 1986 the licensee performed a cold LOP test. The purpose of this test was to demonstrate that the plant response was in accordance with plant design for loss of all offsite power sources after the licensee had assured that the previous test observations had been investigated and resolved. This LOP test was initiated with the reactor at cold (T = 200 F) shutdown temperature and pressure conditions with the reactor mode switch in shutdown.

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During the September 19, 1986 test, the licensee identified a total of 17 ob-servations. The most significant observations were: (1) the "B" safety auxi-liary cooling system loop head tank level indicator failed; (2) one control room emergency ventilation (air recirculation) system fan failed to start; and (3) one drywell fan also failed to start.

As a result of the unsatisfactory test results, an NRC AIT was formed and sent to the site to: (1) independently assess the root cause of each observation; (2) review the effectiveness of the corrective actions planned or taken; and (3) assess the overall implications of the test results. The AIT inspection began on September 25 and ended October 3, 1986.

A second cold LOP test was conducted on October 2, 1986. The AIT witnessed this test and assessed the results. One test observation was a repeat of a previous observation and involved a Bailey logic module.

Of the total of 41 observations reported from the September 11 and 19,1986 LOP tests, the overall safety significance was concluded to be relatively minor except for the Bailey solid state logic module failures. These modules, manu-factured by the Bailey Meter Co., are multipurpose electronic devices used extensively throughout the plant for control and safety functions. Of eight hardware failures identified during this review, six were attributable to various malfunctions with Bailey logic modules.

Three weaknesses with the Bailey logic modules were found: (1) the dependency on common equipment for accomplishment of automatic and manual safety actions for the actuated safety system equipment; (2) limited test provisions to assure the online operability of the Bailey logic modules after their installation into the equipment cabinets, and (3) the usefulness of the bench test equipment in assuring that the Bailey logic modules are operable. The AIT was also con-cerned that the failure rate of the Bailey logic modules appeared high. These weaknesses are especially significant since all of the balance of plant safety-related systems (and a part of one nuclear steam supply system) use Bailey modules to develop the safety system logic and actuation functions.

A number of minor plant design, construction, and manufacturing problems were also identified. Several specific weaknesses in the scope of various system preoperational tests were revealed since the LOP tests were the first integrated demonstration of the plant response to this type of event. Several subtle interactions involving the dependency of various systems on cooling and instru-ment air supporting systems were revealed.

A number of observations resulted because instruments or other equipment lost power during the test. A number of these instances involved the apparent failure to meet Final Safety Analysis Report (FSAR) commitments to provide reliable power to specific instruments or equipment.

In summary, the AIT determined that the results of the LOP tests indicated cer-tain weaknesses in the design, construction, and testing programs for Hope Creek.

With the exception of the Bailey modules, the AIT found the weaknesses to be minor in nature. For the Bailey modules, however, the AIT identified concerns with the adequacy of bench and surveillance testing and a failure rate which was higher than expected.

45

Following the AIT inspection, the licensee presented a program to resolve con-cerns related to the Bailey modules. The program will include:

An in-house data assessment program which will include a review of each in plant module failure and a determination by the manufacturer of the individual component which failed and, to the degree possible, the cause of the failure.

An assessment, by the manufacturer, of module failures at installations of other users.

An accelerated aging and cycling test program, with a final reliability analysis report by the end of the second quarter of CY 1987.

A monthly trending program that will provide a report bi-monthly indicating:

(1) the number of module failures having an adverse affect on system func-tion; (2) resulting time in a Limiting Condition for Operation; and (3) the number of failures determined by surveillance. This program will apply to both IE and non-1E systems.

A report of Bailey's recommendations to improve module reliability based upon their observations at Hope Creek of site environment, handling, and testing techniques.

The modification of existing module test equipment and procedures to permit module testing without staple jumper removal.

The development and procurement of a test rig capable of bench testing I modules for all utilized functions prior to November 1987. Testing would i be conducted without removing staple jumpers or the field programmable

logic array chips.

The determination of the feasibility and implications of modifying the existing Bailey system to permit in-situ testing.

This program is underway and the results are under continued NRC review.

The findings of the AIT are contained in NRC Inspection Report No. 50-354/86-50 which was forwarded formally to the licensee on October 31, 1986 (Ref. C-1).

The NRC forwarding letter also listed several issues which the AIT identified as potentially warranting enforcement action in terms of violations or devia-tions. The issues will be reviewed during future NRC inspections, and if warranted, enforcement actions will be taken as appropriate.

3. Conviction of International Nutronics, Inc. , and One Employee in Federal District Court International Nutronics, Inc., (INI) a California corporation, and Eugene O'Sullivan, a Corporate Vice President and Corporate Radiation Safety Officer of INI, were convicted on October 29, 1986, in Federal District Court in Newark, New Jersey. They had been charged with two counts of willful viola-tion of the incident notification requironents of the Atomic Energy Act, one 46

count of willfully furnishing false information to a government agency, one count of conspiracy to conceal the incident from the NRC, and five counts of mail and wire fraud. Bruce Thomas, the Plant Manager and Radiation Safety Officer of INI's Dover, New Jersey, facility was acquitted on all nine counts.

INI was fined $35,000, the maximum fine. Mr. O'Sullivan was given a suspended sentence and two years probation. Both convictions have been appealed.

The charges resulted from a December 1982 incident involving a spill of radio-actively contaminated water at the INI irradiation facility in Dover, New Jersey.

The spill resulted in widespread contamination of the facility, including the ground immediately under and adjacent to it. Decontamination was begun in 1983 and completed in early 1986. The facility has been released for unrestricted use, and the INI license has been terminated at their request.

4. NRC Augmented Inspection Team Sent to Hatch Facility On December 4, 1986, NRC Region II sent an Augmented Inspection Team to the E.I. Hatch Nuclear Plant to investigate a release of radioactive water from the plant refueling floor into the environment discovered on December 3, 1986.

Hatch is a two-unit facility, both of which utilize a General Electric-designed boiling water reactor. The facility is operated by the Georgia Power Company (the licensee) and is located in Appling County, Georgia.

Prior to December 2,1986, the air regulator (100/20 psi) supplying the inflat-able seals on the transfer canal seal assembly, which seals the gap between the Unit 1 and Unit 2 reactor buildings, failed and the lever valve upstream of the regulator was placed in a throttle position to compensate. This throttle posi-tion was almost shut. On December 2, 1986, at approximately 10:00 p.m. (EST),

this throttled valve was closed when a Plant Equipment Operator (PE0), who was restoring equipment maintenance clearances on the air system, noticed that the valve was almost shut; assumed it to be slightly out of position in the open direction; and placed the valve in the closed position. This action secured the pressurization air to all six inflatable seals on the transfer canal seal assembly. The inflatable seals, not being entirely leak tight, slowly deflated creating a leak path from the spent fuel pool into the gap between the two reac-tor buildings and from there to the environment and to areas of Unit 1 and 2 reactor buildings, Unit 1 and 2 turbine buildings, the control building, the hot machine shop, and the nitrogen storage area.

On December 3,1986 the licensee discovered unidentified water leakage at several locations in the reactor and turbine buildings. Searches were initiated to locate the source of the leakage. Several low fuel pool level alarms were re-ceived during this period but the operating personnel did not correlate the various water leaks in the reactor and turbine buildings with the low fuel pool level alarms. (Normal fuel pool makeup had been about once per shift prior to the leak.) When at 9:37 a.m., December 3, the fuel pool cooling pumps tripped on low-low surge tank level, along with the low fuel pool level alarm, operations personnel went to the refueling floor to %vestigate.

47

The flow past the transfer canal seals was seen by the operators. The air sup-ply was restored to the seals which stopped the leak. Fuel pool level was ob-served to be about five and one-half feet below normal. No significance change in radiation levels on the refueling floor was observed. Calculations indicate that approximately 140,000 gallons had leaked from the fuel pool. Of this, about 17,500 gallons were collected in the buildings and processed through the radwaste system. About 80,000 gallons were released to the swamp located on the licensee's property east of the plant. The rest of the water was absorbed in the three and one-half inch gap between the Unit 1 and Unit 2 buildings.

During the night of December 3 and the morning of December 4,1986, the licensee took immediate corrective action to determine the extent of, and to limit the spread of, contamination. A series of water samples were collected to determine which outfall areas of the swamp contained contaminated water. Additionally, a series of dikes (dams) were constructed to minimize the spread of the contami-nation. The licensee promptly performed radiation surveys of the affected buildings and the outfall area of the swamp to determine the extent of the release. The licensee estimated that the maximum amount of radioactive material that could have been released to the outfall area of the swamp was approximately 0.4 curies.

During the afternoon of December 4, 1986, direct radiation readings along the outfall which led to the swamp ranged from 1-2 mrem /hr. Direct radiation read-ings along the shoreline of the swamp ranged from 150-250 uR/hr (normal readings were typically 8-10 uR/hr). Water sample results at the location identifiei as a potential pathway to the river indicated no detectable activity. The water from behind the dikes and at the entrance to the swamp pool was pumped back to tanker trucks, filtered through demineralizers, and then discharged through the plant radwaste system with all normal discharge precautions in place.

Fuel pool to transfer canal gates were installed on with December 4, 1986, and by December 5, 1986, the air supplies were configured with the transfer canal gate seals fed from both units. The inner seals were supplied from one unit and the outer seals were supplied from the other unit, thus achieving redundancy.

Also, by December 6, 1986, the transfer canal seals were pressurized from each unit such that a single loss of air supply would not cause the seals to deflate.

In addition to the cleanup efforts required (for the inplant areas and the swamp area located near the cooling towers on the east side of the plant) and establishment of sampling stations at various locations, the licensee planned an augmented environmental sampling program to detect any significant future migration of the radioactive material in excess of permissible levels.

The NRC AIT concluded that the organization, staffing, controls and coordination of the recovery efforts were rapid and effective. The licensee's Technical Support Center functioned well as the recovery center. The staffing of the recovery center was adequate, with all needed disciplines represented. The personnel at the spill area in the swamp were well managed and effective in containing the extent of the contamination within the site boundaries and in minimizing the area of the swamp which was contaminated.

The AIT inspection was conducted from December 4 through December 7, 1986. The inspection findings are contained in NRC Inspection Report Nos. 50-321/86-41 and 50-366/86-41. The report was forwarded formally to the license on 48

January 8,1937 (Ref. C-2). A violation was identified during the inspection involving failure to follow procedures that resulted in the event. On April 8, 1987, NRC Reg 4on III forwarded to the licensee a Notice of Violation and Pro-posed Imposit"on of Civil Penalty in the amount of $50,000 (Ref. C-3). Subse-quently, the licensee paid the civil penalty.

On February 2c,1987, the NRC issued Inspection and Enforcement Notice No. 87-13 (Ref. C-4). The Notice described the Hatch event and alerted addressees to the potential for high radiation fields following loss of water from the fuel pool.

As described in the Notice, analysis by the licensee after the event has shown that if water had been completely lost from the fuel transfer canal, radiation fields would ae high enough such that remedial measures may have been difficult to take. Some irradiated control rod blades, stored on short hanger rods clipped over the side of the spent fuel pool, would become completely uncovered, result-ing in general area radiation levels of about 100 R/hr at the edge of the spent fuel pool and about 1 R/hr six feet from the pool edge. (To avoid this potential hazard, the licensee is shipping the control rod blades off the site.) However, no fuel damage of the stored spent fuel components in the pool would be expected because even if the water level dropped to the bottom of the fuel transfer canal, about two feet of water would remain over the top of the spent fuel.

The Notice also mentioned that concern has been raised about the design of the seal systern and the leak detection system for the seals. The NRC is currently evaluating these designs to determine whether they are adequate.

I 49

REFERENCES FOR APPENDICES B-1 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-53, " Improper Installation of Heat Shrinkable Tubing,"

June 26, 1986.*

B-2 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No.86-104, " Unqualified Butt Splice Connectors Identified in Qualified Penetrations," December 16, 1986.*

8-3 U.S. Nuclear Reculatory 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," NUREG-0683, Docket No. 50-320, Draft Supplement No. 2 (dealing with disposal of accident-generated water) issued for public comment during December 1986.*

B-4 Letter from Robert Bernero, Director, Division of BWR Licensing, NRC Office of Nuclear Reactor Regulation, to Dennis L. Farrar, Director of Nuclear Licensing, Commonwealth Edison Company, forwardino " Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Continued Use of SOR Inc. Differential Pressure Switches at La Salle County Stasion, Units 1 and 2," Docket Nos. 50-373 and 50-374, April 1,1987.*

B-5 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to J. G. Randolph, President, Kerr-McGee Center, Sequoyah Fuels Corporation, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, Docket No. 40-08027, October 14, 1986.*

B-6 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to J. G. Randolph, President, Kerr-McGee Center, Sequoyah Fuels Corporation, forwarding an Order Imposing Civil Monetary Penalties, Docket No. 40-08027, February 5, 1987.*

l B-7 U.S. Nuclear Regulatory Commission, " Release of UFg from a Ruptured Model 48Y Cylinder at Sequoyah Fuels Corporation Facility: Lessons - Learned Report," USNRC Report NUREG-1198, published June 1986.**

B-8 Letter from Jack A. Hind, Director, Division of Radiation Safety and Safe-guards, NRC Region III, to J. C. Bishop, Plant Manager, Metropolis Works, Allied Chemical Company, forwarding (1) a Notice of Violation, and (2) Inspection Report No. 86-06, Docket No. 40-3392, November 18, 1986.*

• Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).

Single copies of NPC draft reports are available free, to the extent of supply, upon written request to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

• • Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection. Available for purchase from the GPO Sales Program, Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7982.

51

B-9 Letter from James G. Keppler, Regional Administrator, NRC Region III, to L. L. Taunton, Vice President, Operations, Engineered Materials Sector, Allied-Signal Corporation, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, Docket No. 40-3392, June 27, 1986.*

B-10 Letter from James M. h ylor, Director, NRC Office of Inspection and En-forcement, to L. R. Taunto., Vice President, Operations, Engineered Mate-rials Sector, Allied-Signal Corporation, Docket No. 40-3392, December 19, 1986.*

B-11 Letter from James M. Taylor, Director, NRC Office of Inspection and Er forcement, to W. David Keating, Vice President, Ancillary Services, Mercy Hospital, forwarding (1) an Order to Show Cause Why the License Should Not Be Modified and (2) a Notice of Violation and Proposed Imposition of Civil Penalty, Docket No. 30-02971, June 17, 1986.*

B-12 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to Salvatore M. Imperiale, M.D., Director of Nuclear Medicine, Valley Radiology Associates, Inc., forwarding an Order to Show Cause Why the License Should Not Be Modified, Docket No. 30-15110, June 17, 1986.*

B-13 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to Robert J. Moylan, Executive Vice President, Mercy Hospital, Docket No. 30-02971, December 24, 1986.*

B-14 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to Salvatore M. Imperiale, M.D. , Director of Nuclear Medicine, Valley Radiology Associates, Inc., Docket No. 30-15110, December 24, 1986.*

C-1 Letter from William F. Kane, Director, Division of Reactor Projects, NRC l Region I, to C. A. McNeill, Jr. , Vice President-Nuclear, Public Service Electric and Gas Company, forwarding (1) potential enforcement issues, and (2) Region I Augmented Inspection Team Report No. 50-354/86-50, Docket No. 50-354, October 31, 1986.*

C-2 Letter from J. Nelson Grace, Regional Administration, NRC Region II, to J. H. Miller, Jr. , President, Georgia Power Company, forwarding Inspection Report Nos. 50-321/86-41 and 50-366/86-41, Docket Nos. 50-321 and 50-366, January 8, 1987.*

C-3 Letter from J. Nelson Grace, Regional Administrator, NRC Region II, to James P. O'Reilly, Senior Vice President, Georgia Power Company, forwarding a Notica of Violation and Proposed Imposition of Civil Penalty, Docket Nos. 50-321 and 50-366, April 8, 1987.*

C-4 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Notice No. 87-13, " Potential for High Radiation Fields Following Loss of Water from Fuel Pool," February 24, 1987.*

• Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).

52

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d (3)There a seco .ere si x-a bnormal occurrences at the other NRC licencees. olved release of building at Wright-PatterOneilAirForceBase;threricium-241 involved medical insidemisadministration, a waste storege one therapeutic and two agnostic; one involved. suspension of license for servicing teletherapy and radiogr. Shy units; and one involve an immediately effective order modifying license and. der to show cause issued to . industrial radiography company. There were.'o abnormal occurrences reporte y the Agreement States. The report also contair . information updating some previouf y reported abnormal occur-rences.

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