ML20205S785

Report to Congress on Abnormal Occurrences. October-December 1985

ML20205S785
Person / Time
Issue date:	05/31/1986
From:	NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:
References
NUREG-0090, NUREG-0090-V08-N04, NUREG-90, NUREG-90-V8-N4, NUDOCS 8606120901
Download: ML20205S785 (47)

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NUREG-0090 Vol. 8, No. 4 Report to Congress on Abnormal Occurrences October - December 1985 1 1

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U.S. Nuclear Regulatory Commission Offico for Analysis and Evaluation of Operational Data p" "' %q, t .

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Available from Superintendent of Documents U.S. Govemment Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication.

Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161 i

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NUREG-0090 Vol. 8, No. 4 Report to Congress on Abnormal Occurrences October - December 1985 Date Published: May 1986 Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission Wcshington, D.C. 20565

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PrOvious R0 ports in Series NUREG 75/090, January-June 1975, NUREG-0090, Vol.3, No.3, July-September 1980, pubitshed October 1975 pub 115hed February 1981 NUREG-0090-1, July-September 1975, NUREG-0090, Vol.3, No.4 October-December 1980, published March 1976 published May 1981 NUREG-0090-2, October-December 1975, NUREG-0090 Vol.4, No.1, January-March 1981, published March 1976 published July 1981 NUREG-0090-3, January-March 1976 NUREG-0090, Vol.4, No.2, April-June 1981, published July 1976 published October 1981 NUREG-0090-4, April-June 1976 NUREG-0090, Vol .4, No.3, July-September 1981, published March 1977 published January 1982 NUREG-0090-5, July-September 1976, NUREG-0090, Vol.4, No.4, October-December 1981, published March 1977 published May 1982 NUpEG-0090-6, October-December 1976, NUREG-0090, Vol.5, No.1, January-March 1982, published June 1977 pub 11shed August 1982 NUREG-0090-7, January-March 1977, NUREG-0090, Vol.5, No.2, April-June 1982, published June 1977 pubitshed December 1982 NUREG-0090-8, April-June 1977, NUREG-0090, Vol.5, No.3, July-September 1982, published September 1977 published January 1983 NUREG-0090-9, July-September 1977 NUREG-0090, Vol .5, No.4, October-December 1982, published November 1977 published May 1983 NUREG-0090-10, October-December 1977 NUREG-0090, Vol.6, No.1, January-March 1983, published March 1978 published September 1983 NUREG-0090, Vol.1, No.1, January-March 1978, NUREG-0090, Vol.6, No.2. April-June 1983, published June 1978 published November 1983 NUREG-0090, Vol .1, No.2. April-June 1978, NUREG-0090, Vol.6, No.3, July-September 1983, published September 1978 published April 1984 NUREG-0090 Vol.1, No.3, July-September 1978, NUREG-0090, Vol .6 No.4, October-Decerrber 1983, published December 1978 published Fay 1984 NUREG-0090 Vol.1, No.4, October-December 1978, NUREG-0090, Vol 7, No.1, Jaruary March 1984, published March 1979 published July 1984 NUREG-0090, Vol.2, No.l. January-March 1979, NUREG-0090, Vol.7, No.2, April-June 1984, published July 1979 published October 1984 NUREG-0090, Vol.2, No.2. April-June 1979, NUREG-0090, Vol.7, No.3, July-September 1984, published November 1979 published April 1985 NUREG-0090, Vol .2, No.3, July-September 1979 NUREG-0090, Vol.7 No.4, October-December 1984, published February 1980 published May 1985 NUREG-0090, Vol.2, No.4, October-December 1979, NUREG-0090, Vol.8, No.1, January-March 1985, published April 1980 published Auoust 1985 NUREG-0090, Vol.3, No.1, January-March 1980, NUREG-0090, Vol.8, No.2, April-June 1985, published September 1980 published November 1985 NUREG-0090, Vol .3, No.2, April-June 1980, NUREG-0090, Vol.8, No.3, July-September 1985, published November 1980 published February 1986

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 i to December 31, 1985.

The report states that for this reporting period, there were two abnormal occur-rences at the nuclear power plants licensed to operate. The first involved inoperable main steam isolation valves and the second involved management defi-ciencies at Fermi Nuclear Power Station. There were three abnormal occurrences at the other NRC licensees. Two involved diagnostic medical misadministrations and the other involved a thera)eutic medical misadministration. There were no abnormal occurrences reported )y the Agreement States.

The report also contains information updating some previously reported abnormal occurrences.

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CONTENTS PjLg

, ABSTRACT .......................................................... iii PREFACE ...........................................................

vii INTRODUCTION ................................................. vii THE REGULATORY SYSTEM ........................................ vii REPORTABLE OCCURRENCES ....................................... viii AGREEMENT STATES ............................................. ix FOREIGN INFORMATION .......................................... x REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, OCTOBER-DECEMBER 1985 . 1 l

i NUCLEAR POWER PLANTS ......................................... 1 85-19 Inoperable Main Steam Isolation Valves ........ 1 85-20 Management Deficiencies at Fermi Nuclear Power Station ................................. 4 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants) ...... 7 OTHER NRC LICENSEES (Industrial Radiographers, Medical Insti tutions , Industrial Users , etc. ) . . . . . . . . . . . . . . . . . . . . . . 7 85-21 Diagnostic Medical Misadministration .......... 7 l 85-22 Therapeutic Medical Misadministration ......... 8 85-23 Diagnostic Medical Misadministration .......... 9 l

AGREEMENT STATE LICENSEES .................................... 9 l REFERENCES ........................................................ 11 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA ......................... 13 APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES ... 15 NUCLEAR POWER PLANTS ......................................... 15 79-3 Nuclear Accident at Three Mile Island ......... 15 i

83-3 Failure of Automatic Reactor Trip System ...... 17 83-6 Uncontrolled Leakage of Reactor Coolant Outside Primary Containment ................... 20 84-2 Through Wall Crack in Vent Header Inside BWR Containment Torus ............................. 21 85-7 Loss of Main and Auxiliary Feedwater Systems 23 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants) ...... 25 t

1 84-19 Buildup of Uranium in Ventilation System ...... 25 v

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CONTENTS (continued)

P_ag OTHER NRC LICENSEES ........................................ 26 85-8 Diagnostic Medical Misadministration ........ 26 85-9 Diagnostic Medical Misadministration ........ 26 85-15 Therapeutic Medical Misadministration ....... 27 85-16 Therapeutic Medical Misadministration ....... 27 APPENDIX C - OTHER EVENTS OF INTEREST ........................... 29 REFERENCES (FOR APPENDICES) ..................................... 35 j

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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 abnor-mal occurrences involving facilities and activities regulated by the NRC. An abnormal occurrence is defined in Section 208 as an unscheduled incident or event which the Commission determines is significant from the standpoint of public health or safety.

Events are currently identified as abnormal occurrences for this report by the NRC using the criteria delineated in Apaendix A. These criteria were promulgated in an NRC policy statement which was pu)11shed in the Federal Register on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952). In order to provide wide dissemination of information to the public, a Federal Register notice is issued on each abnormal occurrence with copies distributed to the NRC Public Document Room and all Local Public Document Rooms. At a minimum, each such notice contains the date and place of the occurrence and describes its nature and probable consequences.

The NRC has reviewed Licensee Event Reports, licensing and enforcement actions (e.g., notices of violations, civil penalties, license modifications, etc.),

generic issues, significant inventory differences involving special nuclear material, and other categories of information available to the NRC. The NRC has determined that only those events, including those submitted by the Agree-ment States, described in this report meet the criteria for abnormal occurrence reporting. This report covers the period from October 1 to December 31, 1985.

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

THE REGULATORY SYSTEM The system of licensing and regulation by which NRC carries out its responsi-bilities is implemented through rules and regulations in Title 10 of the Code of Federal Regulations. Toaccomplishitsobjectives NRC regularly conducts licensingproceedings,inspectionandenforcementactIvities evaluation of operatingexperienceandconfirmatoryresearch,whilemaintalningprogramsfor establishing standards and issuing technical reviews and studies. The NRC's

! role in regulating represents a complete cycle, with the NRC establishing stan-dards and rules; issuing licenses and permits; inspecting for compliance; en-

forcing license requirements; and carrying on continuing evaluations, studies l

and research projects to improve both the regulatory process and the protection of the public health and safety. Public participation is an element of the regulatory process.

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

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be achieved and maintained through regulations which specify requirements which will assure the safe use of nuclear materials. The regulations include design and quality assurance criteria appropriate for the various activities licensed by NRC. An inspection and enforcement program helps assure compliance with the regulations.

Most NRC licensee employees who work with or in the vicinity of radioactive materials are required to utilize personnel monitoring devices such as film badges or TLD (thermoluminescent dosimeter) badges. These badges are processed periodically and the exposure results normally serve as the official and legal record of the extent of personnel exposure to radiation during the period the badge was worn. If an individual's past exposure history is known and has been sufficiently low, NRC regulations permit an individual in a restricted area to receive up to three rems of whole body exposure in a calendar quarter. Higher values are permitted to the extremities or skin of the whole body. For unre-stricted areas, permissible levels of radiation are considerably smaller. Per-missible doses for restricted areas and unrestricted areas are stated in 10 CFR Part 20. In any case, the NRC's policy is to maintain radiation exposures to levels as low as reasonably achievable.

REPORTABLE OCCURRENCES Actual operating experience is an essential input to the regulatory process for assuring that licensed activities are conducted safely. Reporting requirements exist which require that licensees report certain incidents or events to the NRC. This reporting helps to identify deficiencies early and to assure that corrective actions are taken to prevent recurrence.

For nuclear power plants, dedicated groups have been formed both by the NRC and by the nuclear power industry for the detailed review of operating experience to help identify safety concerns early, to improve dissemination of such infor-and mation}ons.tofeedbacktheexperienceintolicensing, operat regulations,and In addition, the NRC and the nuclear power industry have ongoing efforts to improve the operational data system which include not only the type, and quality, of reports required to be submitted, but also the method used to analyze the data. Two primary sources of operational data are reports submitted by the licensees under the Licensee Event Report (LER) system, and under the Nuclear Plant Reliability Data (NPRD) system. The former system is under the control of the NRC while the latter system is a voluntary, industry-supported system operated by the Institute of Nuclear Power Operations (INP0), a nuclear utility organization.

Some form of LER reporting system has been in existence since the first nuclear power plant was licensed. Reporting requirements were delineated in the Code of Federal Regulations (10 CFR), in the licensees' technical specifications, and/or in license provisions. In order to more effectively collect, collate, store, retrieve, and evaluate the information concerning reportable events, the Atomic Energy Commission (the predecessor of the NRC) established in 1973 a computer-based data file, with data extracted from licensee reports dating from 1969. Periodically, changes were made to improve both the offectiveness of data processing and the quality of reports required to be submitted by the licensees.

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Effective January 1, 1984, major changes were made to the requirements to report to the NRC. A revised Licensee Event Report System (10 CFR S 50.73) was estab- -

lished by Commission rulemaking which modified and codified the former LER sys-tem. The purpose was to standardize the reporting requirements for all nuclear power plant licensees and eliminate reporting of events which were of low indi-vidual significance, while requiring more thorough documentation and analyses by the licensees of any events required to be reported. All such reports are to be submitted within 30 days of discovery. The revised system also permits licensees to use the LER procedures for various other reports required under specific sections of 10 CFR Part 20 and Part 50. The amendment to the Commis-sion's regulations was published in the Federal Register (48 FR 33850) on July 26, 1983, and is described in NUREG-1022, " Licensee Event Report System,"

and Supplements 1 and 2 to NUREG-1022.

Also effective January 1, 1984, the NRC amended its immediate notification re-ower reactors (10 CFR guirements 5 50.72). This of was significant events published in the at:ederal o)erating nuclear Register p(48 FR 39039) on August 1983, with corrections (48 FR 40882) published on September 12, 1983. Among the changes made were the use of terminology, phrasing, and reporting thresholds that are similar to those of 10 CFR S 50.73. Therefore, most events reported under 10 CFR S 50.72 will also require an in-depth follow-up report under 10 CFR S 50.73.

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

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order to make the system a viable tool in analyzing operating experience. How-ever, on July 8,1981, INP0 announced that because of its role as an active user of NPRD system data, it would assume responsibility for management and fund-ing of the NPRD system INP0 reports that significant im)rovements in licensee participation are being made. The Commission considers t1e NPRD system to be a vital adjunct to the LER system for the collection, review, and feedback of operational experience; therefore, the Commission periodically monitors the progress made on improving the NPRD system.

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

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

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

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

States assume regulatory authority over byproduct, source and special nuclear materials (in quantities not capable of sustaining a chain reaction). Comparable and compatible programs are the basis for agreements.

Presently, information on reportable occurrences in Agreement State licensed activities is publicly available at the State level. Certain information is also provided to the NRC under exchange of information provisions in the agreements.

In early 1977, the Commission determined that abnormal occurrences happening at facilities of Agreement State licensees should be included in the quarterly report to Congress. The abnormal occurrence criteria included in Appendix A is applied uniformly to events at NRC and Agreement State licensee facilities.

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

FOREIGN INFORMATION .

The NRC participates in an exchange of information with various foreign govern-ments which have nuclear facilities. This foreign information is reviewed and considered in the NRC's assessment of operating experience and in its research and regulatory activities. Reference to foreign information may occasionally be made in these quarterly abnormal occurrence reports to Congress; however, only domestic abnormal occurrences are reported.

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REPORT TO CONGRESS ON ABNORMAL OCCURRENCES OCTOBER-DECEMBER 1985 NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to

, operate during the fourth calendar quarter of 1985. As of the date of this report, the NRC had determined that the following events were abnormal occurrences.

[ Note: There were two other significant events which occurred during the latter hiTf of the fourth calendar quarter of 1985, f.e., (1) loss of power and water hammer event at San Onofre Unit 1 on November 21,1985,and(2)lossofinte-grated control system power and overcooling transient at Rancho Seco on Decem-ber 26, 1985. Both events were being investigated by NRC Incident Investigation Teams. As of the end of the fourth calendar quarter, the Teams had not yet completed their investigations. Therefore, the events will be included in the next quarterly abnormal occurrence report.]

85-19 Inoperable Main Steam Isolation Valves The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see general criterion 2) of this report notes that major degradation of essential safety-related equipment can ,

be considered an abnormal occurrence.

Date and Place - On September 27, 1985, duringmainsteamisolationvalve(MSIV) surveillance testing at Brunswick Unit 2 Carolina Power and Light Company (the licensee)discoveredthatMSIVsF028A,Fd22C,andF028Cwouldnotfastclose.

At the time of the event, the plant was in cold shutdown following a controlled shutdown from power on September 26, 1985, as a precautionary measure in advance of then a)proaching hurricane Gloria. Brunswick Unit 2 is a General Electric-designed ) oiling water reactor located in Brunswick County, North Carolina.

Nature and Probable Consecuences - The main steam line isolation system includes isolation valves that, incividually or collectively, are designed to:

1. Closewithinthetimeestablishedbydesignbasisaccident(DBA) analysis to limit the release of reactor coolant or radioactive materials; i 2. Close when required, despite the single failure of either valve or the associated controls, to provide a high level of reliability for the safety function;

3. Use separate energy sources, as the motive force to independently close theredundantisolationvalvesineachindividualsteamline;

4. Uselocalstoredenergy(compressedairand/orsprings)tocloseatleast  ;

l one isolation valve in each steam line without relying on continuity of any variety of electrical power for the motive force to achieve closure; I 5. Be able to close during or after design basis seismic loadings to assure isolation; and 1

6. Be testable during normal operating conditions, to demonstrate that the valves will function.

Two MSIVs (F022 and F028) are welded in a horizontal run of each of the four main steam lines (A-D), with one valve as close as possible to the primary containmentbarrierinside,andtheotherjustoutsidethebarrier. The valves, when closed, form part of the nuclear system process barrier for openings outside the primary containment, and part of the primary containment barrier for nuclear system breaks inside the containment.

The valve is held open by pneumatic pressure and loss of air allows a spring assist to close the valve. The control unit is attached to an air cylinder and contains the pneumatic, ac, and dc control valves for opening, closing, and slow speed exercising of the main valve. The control power for each valve is supplied at 120 volts ac, 60 Hz, and 125 volts dc. Remote manual switches in the control room enable the reactor operator to operate or close each valve at fast speed (3 to 10 seconds) or at slow speed (45 to 60 seconds) for exercising or testing.

In the event that the main steam line should rupture downstream from the valve, the steam flow quickly increases to 200 percent of rated flow and is limited from further increase by the venturi flow restrictor upstream of the valves.

During approximately the first 75 percent of closing, the valve has little effect in reducing flow because the flow is choked by the venturi restrictor upstream from the valves. After the valve is more than about 75 percent closed, flow is reduced as a function of the valve area versus travel characteristic.

ThesafetyobjectivesoftheMSIVsareto:

1. Prevent damage to the fuel barrier by limiting the loss of reactor coolant in case of a major leak from the steam piping outside the primary containment;

2. Limit release of radioactive materials by closing the nuclear system process barrier in case of gross release of radioactive materials from the reactor fuel to the reactor cooling water and steam; and

3. Limit release of radioactive materials by closing the primary containment barrier in case of a major leak from the nuclear system inside the primary containment.

In the direct cycle boiling water nuclear power plant, the reactor steam goes to the turbine and to other equipment outside the reactor containments. Radio-active materials in the steam are released to the environs through process open-ings in the steam system, or they can escape from accidental openings. A large break in the steam system can void the water from the reactor core faster than it is replaced by feedwater. The analysis of a complete sudden steam line breakoutsidetheprimarycontainmentshowsthatthefuelbarrierisprotected against loss of cooling if MSIV closure takes 10.5 seconds or less (including as much as 0.5 seconds for the instrumentation to initiate valve closure after the break). .

Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only 2

I one valve in each line is required to maintain the integrity of the containment.

The Technical Specification surveillance requirements are based on the operating h 4 tory of this type valve. The maximum closure time of 5 seconds has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

Initial investigation determined a problem existed within the double solenoid valve on the actuator of MSIV F028A. In addition, a subse uent fast closure test of MSIV F028C revealed a closure time of approximate 1 45 seconds. A de-cision was made not to troubleshoot the F022C valve until t could be removed and disassembled. On September 28, 1985, the licensee began to disassemble visuallyinspect,andcollectsamplesofforeignmaterialsinthedoublesolenoid valvesofthethreesubjectMSIVs.

From the visual ins)ection, it was concluded the F028C valve had failed to fast close due to the ex1aust port being blocked. The F022C valve failed to fast close because the solenoid valve disc was adhered to the valve seat and the valve could not move. The F028A valve failed either because the solenoid valve disc was stuck to its seat or because disc material had broken off and plugged the exhaust port.

An evaluation of the solenoid valve from the F028C by the valve vendor, ASCO, concluded the valve elastomer had degraded due to contamination. The vendor did not determine the contaminant; however, it is felt it was not introduced during manufacture or assembly of the solenoid valve assembly.

On September 28, 1985, a systems engineering task force was established by the licensee to determine the cause of the failures and to recommend and implement corrective actions.

The pilot valves that failed are ASCO Model 8323A36E double solenoid valves.

Each valve has one ac and one de coil. Both solenoids must be deenergized for the valve to close the supply port and open the exhaust port. The valves operate with both solenoids normally energized. This model was installed in Unit 1 in June 1983 and in Unit 2 in August 1984. They are environmentally qualified replacements for nearly identical valves which used Buna N as the elastomer.

These valves utilize ethylene propylene (EP) in place of the Buna N. The EP material is rated for higher temperature use and has much higher radiation resis-tance than Buna N.

Because EP is resistant to high levels of radiation, it is the material of choicer for environmentally qualified (EQ) applications. However, EP absorbs hydrocarbons and, like a sponge, softens and swells up. Laboratory analysis of the three subject solenoid valves showed a significant amount of hydrocarbons in the valve body of the F028C solenoid valve. The precise hydrocarbon has not yet been identified.

Cause or causes - The precise failure mechanism of the elastomer material has not yet been identified; however, it is believed that temperature, contamination, and internal geometry played a part in the failure mechanism. Temperature alone can reproduce the failure mechanism but not at a temperature that is reasonably expected to occur in the plant. The physical evidence showed signs of swelling of the EP which is known to occur when it is exposed to hydrocarbons. The geometryIssuchthattheEPcanswellandfilltheexhaustport,thusblocking 3

that path and providing a frictional force in opposition to the valve opening force. In addition, exposure to hydrocarbons may cause the elastomer in EP to vaporize at a lower temperature. Valves that are not subjected to the same conditions of temperature, contamination, and internal geometry are not expected to experience this failure.

Actions Taken to Prevent Recurrence Licensee - The Unit 2 MSIV solenoid valves have been replaced with valves utilizing viton as the elastomer. The valves on Unit I were replaced during the ongoing 1985 Unit 1 refueling / maintenance outage by valves utilizing viton.

Viton is impervious to the hydrocarbon contamination and licensee tests have shown that it can withstand temperatures that will degrade EP. While the geom-etry of the viton valves is the same as the EP, reduction of two of the three effects (i.e. , temperature, contamination, and geometry) provides an increased degree of confidence that the viton valves will not fail in this mode. Viton, however, is less resistant to radiation than EP by a factor of ten; therefore, the material needs replacing more frequently.

During the next scheduled outage of sufficient length following six months of operation for either unit, a sampling of the MSIV double solenoid valves will be replaced. The removed solenoids will be evaluated as ) art of the continuing failure analysis of the ASCO solenoids. It is expected t1e solenoid valves will be replaced at approximately three year intervals due to the radiation susceptibility of viton.

NRC - A detailed inspection to review the licensee actions with respect to the solenoid failures was conducted by NRC Region II. Inspectors from NRC Region II and NRC headquarters reviewed test results at both the site and the utility's research facility. An Inspection and Enforcement Information Notice regarding this event is under consideration.

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

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! 85-20 Management Deficiencies at Fermi Nuclear Power Station l

The following information pertaining to this event is also being re)orted con-currently"in the Federal Register. Appendix A (see Example 11 of ";or All Licensees ) of this report notes that serious deficiencies in management or proceduralcontrolsinmajorareascanbeconsideredanabnormaloccurrence.

Date and Place - On December 24, 1985, NRC Region III issued a letter under 10 CFR 50.54(f) seeking information from Detroit Edison Company on plans to im-prove the regulatory performance of its Fermi Unit 2 Nuclear Power Station

! (Ref. 1). The letter identified a series of operational and equipment problems occurring at Fermi Unit 2 and attributed them to ineffective management systems, t The plant, which utilizes a General Electric-designed boiling water reactor, is located in Monroe County, Michigan.

l Nature and Probable Consequences - In July 1985 the licensee began to experience I a series of operational problems, some of which were the result of personnel errors.

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On July 1, 1985, during startup of the reactor, a reactor operator failed to follow the procedure for withdrawing control rods from the reactor core. The operator withdrew 11 rods to their fully withdrawn position (position 48) instead of to the next intermediate position (position 04). As a result the reactor went critical (i.e., a sustained nuclear chain reaction was started) earlier in the startup procedures than planned.

When instrumentation indicated unexpected conditions in the reactor, the operator reinserted the control rods to stop the chain reaction. Later, the startup was resumed and criticality occurred at the expected point. Later analysis showed that the reactor had gone critical when the 11 rods were withdrawn.

The NRC's inspection and review of this premature criticality event identified nine apparent violations of NRC requirements associated with it. Because the plant had just achieved its initial criticality on June 21, 1985, and therefore had little residual radioactivity in the reactor core, this event was of minimal safety consequences. No damage to the fuel would be expected under these circumstances.

Subsecuently, additional problems occurred at the plant, some of which were causec by personnel errors while others were strictly equipment failures. These problems included:

1. The South Reactor Feedwater Pum) turbine was damaged during testing on July 22, 1985. The pump turbine had a listory of excessive vibration during preoperational testing. The second feedwater pump was not affected by the vi-bration problem. Operation at power levels up to 50 percent is possible with only one feedwater pump in operation.

2. On July 26, 1985, while a diesel generator was undergoing testing, a low flow of cooling water for the diesel was observed. The diesel was shut off, and the licensee's subsequent investigation determined that a cooling tower bypass valve was closed. This valve is required to be open for the flow of cooling water to the diesel generator and to other safety system equipment in Division 1, including the core spray system and the residual heat removal system.

The licensee determined that the valve had been left in the closed position while running the Reactor Heat Removal Service Water System on July 23, 1985.

The plant is required to have both Divisions of its emergency core cooling system operable, while the plant is in startup or operating status. The plant was in a startup mode when the valve was closed, but shut down later because of the previously reported feedwater aump turbine problem. (Theshutdownfortuitously avoided further violation of t1e plant Technical Specification requirement that the plant be placed in cold shutdown if one ECCS division is inoperable.)

3. On September 2, 1985, the licensee discovered that a Containment Monitoring System Valve was open and uncapped, thus providing a breach of the primary reactor containment. The valve was left 03en following installation of the valve in June 1985. (This event was descri)ed in Appendix C, "Other Events of Interest," in the Abnormal Occurrence Report to Congress for the Third Quarter 1985, i.e., NUREG-0090, Vol. 8, No. 3.)

i i 4. Since receiving an operating license in March 1985, Fermi Unit 2 has l experienced numerous personnel errors. Of 78 Licensee Event Reports submitted by the licensee between March and November 1985, 41 involved personnel errors.

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5. NRC inspectors, in an inspection report covering the events listed above as well as other problems at the Fermi Unit 2 facility, identified 26 items of apparent violation of NRC requirements (Ref 2) Enforcement action on these items is pending.

The unit was shut down on October 10, 1985, for the installation of a remote shutdown control panel to meet an NRC requirement for the capability of shutting the plant down safely from a point outside the main control room.

During the outage, significant problems were encountered with the facility's emergency diesel generators. During testing of the No. 13 diesel generator in November 1985, excessive noise and vibration were detected. The diesel generator was shut down, and an examination showed evidence of bearing damage. Subsequent inspections of two additional diesel generators (the site has a total of four) showed similar damage. The plant had previously encountered bearing problems with two diesel generators in January 1985 and the equipment was repaired.

The licensee and the diesel vendor, Fairbanks Morse, believe the damage is attributable to an insufficient break-in )eriod. Further testing of the die-sels is underway, and the root cause of t1e bearing problems remains under re-view by the licensee, its consultants, and the NRC.

Cause or Causes - The principal cause is attributed to ineffective management systems as reflected in a series of operational and equipment problems, and numerous personnel errors and violations of technical requirements.

Actions Taken to Prevent Recurrence Licensee - The licensee has instituted various measures to improve its regulatory performance, including retraining of personnel and revision of procedures. An independent overview committee of consultants has been formed by the licensee to review the management and operation of the Fermi facility.

NRC - Following the July 1, 1985, premature criticality incident, NRC Region III issued a Confirmatory Action Letter (Ref. 3) to Detroit Edison, confirming the licensee's agreement not to operate the unit above five percent power until the premature criticality incident was fully analyzed and corrective action taken. Operation above five percent power would not occur without authorization from NRC Region III. That restriction was not lifted prior to the October 10, 1985, outage.

The December 24, 1985, request for information under 10 CFR 50.54(f) sought the licensee's response on the adequacy of management and management structures, on changes in controls needed to improve re ulatory performance and on actions plannedtoensurereadinessofthefaciltytoresumeoperatIonsandtesting activities (Ref. 1).

Future reports will be made as appropriate.

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FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)

The NRC is reviewing events reported by these licensees during the fourth calen-dar quarter of 1985. As of the date of this report, the NRC had not determined that any events were abnormal occurrences.

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

There are currently more than 8,000 NRC nuclear material licenses in effect in the United States, principally for use of radioisotopes in the medical, industrial and academic fields. Incidents were reported in this category from licensees such as radiographers, medical institations, and byproduct material users.

The NRC is reviewing events reported by these licensees during the fourth calen-dar quarter of 1985. As of the date of this report, the NRC had determined that the following events were abnormal occurrences.

85-21 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 criteria) of this report notes that an event involving a moderate or more severe impact on public health or safety can be considered an abnormal occurrence.

Date and Place - On August 14, 1985, a patient at the Letterman Army Medical Center, Presidio of San Francisco, California, was inadvertently given the wrong radiopharmaceutical for a scheduled thyroid uptake study and scan. This resulted in an administered dose which exceeded the prescribed dose by approximately a factor of 30. f Nature and Probable Consequences - The attending physician mistakenly prescribed a dose of 150 microcuries of I-131 instead of I-123. The radiopharmacy misin-terpreted the prescribed dose as 5 millicuries of I-131. The 5 millicuries of I-131 were adrinistered to the patient on August 14, 1985. The mistake was identified on August 16, 1985, when the patient returned to the hospital for ,

the uptake stLdy and scan.

The level of radiopharmaceutical administered is commonly given to patients for certain other diagnostic procedures, and, despite the mistake, the diagnostic scan and uptake desired for this patient were able to be accomplished without administration of any additional radiation beyond the initial dose. Also, the patient had previously undergone a partial thyroidectomy and was taking thyroid hormones for thyroid gland suppression. The licensee states that, because of i

these circumstances, no adverse clinical symptoms were expected as a result of the misadministration.

l The patient and the attending physicians were notified of the misadministration

and the licensee began an immediate investigation to determine what factors and

circumstances may have contributed to the incident.

7

Cause or Causes - The misadministration was caused when the attending physician prescribed the wrong radiopharmaceutical which was further misinterpreted by the radiopharmacy as a request for 5 millicuries of I-131.

Actions Taken to Prevent Recurrence Licensee - Effective August 16, 1985, the licensee instituted a new hospital pro-cedure which would require that only the nuclear medicine staff would administer radiopharmaceuticals and the radiopharmacist must authorize the release of radio-pharmaceuticals from the pharmacy. All prescriptions will be in writing.

NRC - The circumstances of the misadministration were discussed with the Ticensee. The licensee's corrective actions were determined to be acceptable.

The NRC does not plan any further actions.

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

• • • AAAAA 85-22 Therapeutic Medical Misadministration The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see the general criteria) of this re) ort notes that an event involving a moderate or more severe impact on public 1ealth or safety can be considered an abnormal occurrence.

Date and Place - On October 9 1985, at the Queen's Medical Center, Honolulu, Hawaii,apatientwastoreceIve1000radstothelaterallimbalareaofthe right eye using a Sr-90 applicator; however, the medial limbal area of the right eye was treated.

Nature and Probable Consequences - The attending physician realized the error when the patient returned for his second treatment on October 16 1985. The patientandreferringabysicianwereinformedoftheerrorimmedlately. Accord-ing to the attending p1ysicians, there are no complications to be anticipated due to the incorrect treatment.

Cause or Causes - The misadministration was caused when the attending physician misinterpreteTan area containing scar tissue in the medial limbal area of the right eye as the area to be treated. There were no written treatment instructions for the attending physician.

Actiors Taken to Prevent Recurrence Licensee - Effective November 20, 1985, the hospital has made it mandatory that a written requisition must be submitted by the referring physician prior to treatment. This requisition must include the patient's name and a description of the intended treatment area clearly identified. The physician must then sign and date the requisition. This requisition shall be kept in the patient's chart along with the treatment summary.

NRC - The circumstances of the misadministration were reviewed during a visit to the hospital on December 19 1985, by members of the NRC Region V management staff. Thelicensee'scorrectIveactionsweredeterminedtobeacc? table.

The NRC does not plan any further actions.

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

85-23 p h iostic Medical Misadministration The fo:h, wing information pertaining to this event is also being reported concur-rently a the Federal Register. Appendix A (see the general criteria) of the report . tes that an event involving a moderate or more severe impact on public health o safety can be considered an abnormal occurrence.

Date and Place - On December 9, 1985, a patient at Hospital Universitario, San Juan Puerto Rico, received 4.98 millicuries of iodine-131 instead of a 10 to 15microcuriesdoseusuallygivenfora24-hourthyroiduptaketest.

Nature and Probable Consequences - The patient arrived at the hospital's Nuclear Medicire Division on December 9, 1985, to receive iodine-131 for a 24-hour thy-roid test. The test was part of the physician's plan to evaluate the patient for hyperthyroidism. The usual dose for such a diagnostic procedure at the Nuclear Medicine Division is 10 to 15 microcuries. Instead, the technologist mistakenly administered a dose of 4.98 millicuries which is the dose usually given for whole body scans with iodine-131.

The patient's referring abysician was notified of the misadministration. Based on statements from the p1ysician, the patient was a likely candidate for iodine-131 therapy for treatment of the hyperthyroid condition; therefore, the probable consequences for the patient would be consistent with the projected medical treatment.

Cause or Causes - As discussed above, the reason for the misadministration was due to an error by the technologist.

Action Taken to Prevent Recurrence Licensee - Review with the nuclear medicine staff the protocol used for hyper-thyroid patients dosed with radiciodine.

NRC - The incident and the licensee's protocol will be reviewed during the next NRC routine inspection.

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

AGREEMENT STATE LICENSEES Procedures have been develo)ed for the Agreement States to screen unscheduled incidents or events using t1e 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 1985, the Agreement States reported no abnormal occurrences to the NRC.

9 I

REFERENCES

1. 10 CFR S 50.54(f) letter from James G. Keppler, Regional Administrator, NRC Region III, to Wayne H. Jens, Vice President-Nuclear Operations, Detroit Edison Company, Docket No. 50-341, December 24, 1985.*

2. Letter from Charles E. Norelius, Director, Division of Reactor Projects, NRC Region III, to Wayne H. Jens, Vice President-Nuclear Operations, Detroit Edison Company, forwarding Inspection Report No. 50-341-84-040, Docket No. 50-341, January 7, 1986.*

3. Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to Wayne H. Jens, Vice President-Nuclear Operations, Detroit Edison Company, Docket No. 50-341, July 16,1985.*

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

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APPENDIX A ABNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormal occurrence determinations were set forth in an NRC policy statement published in the Federal Register on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952).

Aneventwillbeconsideredanabnormaloccurrenceifitinvolvesamajor reduction 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. Majordegradationofessentialsafety-relatedequipment;or

3. Majordeficienciesindesign, construction,useof,ormanagementcontrols for licensed facilities or material.

Examples of the types of events that are evaluated in detail using these criteria are:

For All Licensees

1. Exposure of the whole body of any individual to 25 rems or more of radia-tion; exposure of the skin of the whole body of any individual to 150 rems or more of radiation; or exposure of the feet, ankles, hands or forearms of any individual to 375 rems or more of radiation (10 CFR S20.403(a)(1)),

or equivalent exposures from internal sources.

2. An exposure to an individual in an unrestricted area such that the whole-body dose received exceeds 0.5 rem in one calendar year (10 CFR S20.105(a)).

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

4. Radiation or contamination levels in excess of design values on packages, or loss of confinement of radioactive material such as (a) a radiation dose rate of 1,000 mrem per hour three feet from the surface of a package containing the radioactive material, or (b) release of radioactive material from a package in amounts greater than the regulatory limit.

5. Any loss of licensed material in such quantities and under such circum-stances that substantial hazard may result to persons in unrestricted areas.

6. A substantiated case of actual or attempted theft or diversion of licensed material or sabotage of a facility.

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7. Any substantiated loss of special nuclear material or any substantiated inventorydiscrepancywhichisjudgedtobesignificantrelativetonormally expected performance and which is ;udged 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. Amajordeficiencyindesign,constructionoroperationhavingsafety implications requiring immediate remedial action.

11. Serious deficiency in management or procedural controls in major areas.

12.

Seriesofevents(whereindividualeventsarenotofmajorimportance)Ities recurring incidents, and incidents with implications for similar faci 1 (genericincidents),whichcreatemajorsafetyconcern.

For Commercial Nuclear Power Plants

1. Exceeding a safety limit of license technical specifications (10 CFR 550.36(c)).

2. Majordegradationoffuelintegrity,primarycoolantpressureboundary,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. Discoveryofamajorconditionnotspecificallyconsideredinthesafety 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 550.36(c)).

2. A major condition not specifically considered in the safety analysis report or technical specifications that requires immediate remedial action.

3. An event which seriously compromised the ability of a confinement system to perform its designated function.

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APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the October through December, 1985 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 necessary to prevent recurrence of previously reported abnormal occur-rences. The referenced Congressional abnormal occurrence reports below provide the initial and any updating information on the abnormal occurrences discussed.

Those occurrences not now considered closed will be discussed in subsequent reports in the series.

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

" Report to Congress on Abnormal Occurrences: January-March 1979," and undated in each subsequent report in this series, i.e., NUREG-0090, Vol. 2, No. 2 through Vol. 8, No. 3. It is further updated as follows.

Reactor Building Entries During the fourth calendar quarter of 1985, 81 entries were made into the reactor building. There have been a total of 783 entries since the March 1979 accident. ,

Major reactor building activities during this period consisted of preparations for and the commencement of defueling operations. Installation and preopera-tional testing of defueling systems and equipment were completed, including the Defueling Water Cleanup Syrtem, the Canister Positioning System, the Canister Transfer System, and the Reactor Building Canister Handling Bridge. Defueling began during this period, as discussed below. Video mapping of the debris bed was performed during a brief defueling delay. Also, during this period, concrete samples of the highly radioactive reactor building basement were collected using a robotic device. Surveys conducted using video cameras and thermoluminescent dosimeters indicated little fuel deposition in the pressurizer.

Reactor Vessel Defueling Operations On Thursday, October 31, 1985, GPU Nuclear began the first phase of defueling the damaged TMI-2 reactor core, when specially trained operators used long-handled manual and hydraulic tools to move damaged fuel and core debris within the reactor vessel. Initial debris movements provided sufficient space in the reactor vessel for the installation and rotation of the Canister Positioning System (CPS), a carousel device capable of holding five defueling canisters.

Defueling operators then used various tools to separate fused end fittings and to cut up partial fuel assemblies to prepare the debris for loading into fuel canisters. Operations were briefly delayed when boron precipitation problems occurred in the hydraulic fluid used to power the long-handled tools. A new i type of borated hydraulic fluid was put into the system, which alleviated the precipitation problem.

In December of 1985, several fuel canisters were filled with core debris consist-ing primarily of fuel assembly end fittings, control rod spiders, and small 15 1 *

• pieces of partial fuel assemblies. Packing efficiency was enhanced by loading debris into special buckets which were in turn loaded into the fuel canisters.

Following the filling of the first group of fuel canisters, the licensee success-fully tested and operated the Vacuum Defueling System, filling a knockout can-ister with loose granular debris. By January 10, 1986, there were four loaded fuel canisters and one loaded knockout canister in the CPS. On January 12, 1986, three of the fuel canisters were sealed, dewatered, and successfully transferred to the storage racks in the "A" spent fuel pool in the fuel handling building. Radiation levels during these initial defueling operations remained low. Dose rates on the defueling work platform have been in the range of 10 mR/hr. Measured dose rates near the shielded canisters during transfer were below 40 mR/hr. On the basis of air sample data from the first month of defuel-ing, the licensee decided to discontinue the use of respirators during defueling activities. This action increased worker efficiency, and coupled with the addi-tion of a third defueling shift, will increase the pace of the defueling program.

EPICOR-II/ Submerged Demineralizer System (SDS) Processing Approximately 119,600 gallons of water were processed through the EPICOR-II sys-tem during the re)orting period. The SDS processed approximately 635,022 of water during tie quarter, primarily in support of defueling activities. gallons Liner Shipments Seven EPICOR-II dewatered resin liners were shipped from the site to Hanford, Washington, during the reporting period. Nine EPICOR-II spent resin liners were shipped to U.S. Ecology at Richland, Washington during the quarter.

Strontium-90 Inspection A special inspection was conducted from September 30 through October 2 1985 toreviewthelicensee'sreportthatduetomisinterpretationofacallbration procedure, the strontium-90 (Sr-90) activities reported for some solid waste shipments were one-half the actual activities. The inspectors also reviewed the licensee's efforts to determine the corrective actions needed. The licensee agreed to perform an independent assessment of the chemistry Quality Assurance /

Quality Control (QA/QC) program; implement a formalized QA/QC program; and to assure that all procedures used at TMI-2 for Sr-90 analyses are clear and accurate.

TMI-2 Advisory Panel Meetings The Advisory Panel for the Decontamination of Three Mile Island Unit 2 (Panel) met on October 16, 1985, in Lancaster, Pennsylvania. At this meeting, the Panel received a presentation from Dr. H. A. Muller, State Health Secretary, Pennsyl-vania Department of Health, regarding his department's recent study entitled

" Cancer Mortality and Morbidity (Incidence) Around TMI." The results of the study do not indicate an increased risk of developing cancer for local residents near TMI. The Panel also received a status report on preparations for early defueling from GPU Nuclear. Members of the NRC staff briefed the Panel on the impending reorganization of the TMI Program Office.

On November 19, 1985, the Panel met with the five Commissioners of the NRC in Washington, DC. Topics of discussion included the liennsee's defueling plans, 16

measures to prevent criticality during defueling, the defueling schedule, Panel activities concerning health effects issues, and the status of licensee plans for the disposition of processed accident water.

On December 12, 1985, the Panel met in Harrisburg, Pennsylvania. At that meeting, Mr. and Mrs. Aamodt discussed the findings of their survey on health effects in the vicinity of TMI and rebutted the findings of the Pennsylvania Department of Health. The Panel also was briefed by GPU Nuclear on the status of defueling operation.

Future reports will be made as appropriate.

83-3 Failure of Automatic Reactor Trip System i This abnormal occurrence was originally reported in NUREG-0090, Vol. 6, No. 1, >

" Report to Congress on Abnormal Occurrences: January-March 1983." It was up-dated and closed out in NUREG-0090, Vol. 6, No. 3. It is being reopened to de-l scribe several recent problems with Westinghouse and General Electric type reac-tor trip breakers (RTBs).

Westinghouse Type D8-50 RTBs On October 29, 1985, D. C. Cook Nuclear Station Unit No. 2* tripped from approximately 80% of full power on a spurious indicated low flow condition.

Immediately following the reactor trip, it was discovered that the Train "A" RTB did not open, and that the trip function was carried out by only the "B" train RTB. Subsequent action by the licensee included quarantine of the failed train "A" reactor trip breaker. The NRC sent an Augmented Incident Response Team to the site for further investigation.

All of the reactor trip breakers at D.C. Cook Unit 2 were Westinghouse type 08-50 and had been refurbished at Westinghouse in July 1985. On November 3, 1985, the "B" train RTB (which had successfully tripped the reactor on Octo-ber29,1985) failed to pass the undervoltage trip attachment (UVTA) force margin test.

trip installed. UnlikeD.C.CookUnit1}*neitherRTBhadtheautomaticshunt The Westinghouse spec fication for the D8-50 RTB specifies a minimum trip force margin of 20 ounces above the maximum force required to trip the breaker. Testing of the "A" train RTB indicated that the UVTA force margin had dropped to five ounces or less. Testing of the "B" train RTB indicated that ttie UVTA force margin was no longer as great as 20 ounces. The root cause of the lost force margin is being determined. Preliminary investigation indi-cates that the loss may be due to lack of proper lubrication combined with as-

! sembly deficiencies in the UVTA itself.

On November 5, 1985, the NRC staff issued Inspection and Enforcement Bulletin

, No. 85-02 (Ref. B-1) to all power reactor licensees and applicants describing i the two D.C. Cook events. The bulletin required that nine Westinghouse PWR l facilities which had not yet installed the automatic shunt trip coil must:

a) perform the UVTA force margin test within seven days, b) incorporate the UVTA force margin test into the normal RTB testing performed each month,

• D. C. Cook Units 1 and 2 are pressurized water reactors operated by Indiana &

, Michigan Electric Company and located in Berrien County, Michigan.

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F' $

c) declare a RTB inoperable if it fails to pass the force margin test and notify the NRC within four hours of such a failure, and d) provide a written report to the NRC discussing specific actions taken pursuant to the bulletin.

Of the nine Westinghouse PWR facilities required to be tested, only the Kewaunee Nuclear Power Plant

• has reported RTBs failing the UVTA force margin test. On November 7, 1985, Kewaunee reported that both a "B" train RTB bypass breaker and an "A" train RTB failed to trip vith a 20-ounce weight attached to the trip bar. All of the nine Westinghous-e PWR facilities without the automatic shunt trip coil c,re scheduled to install the device prior to startup or by May 1986.

General Electric Type AK-2-25 RTBs On June 5,1985, one of the dc RTBs at the Rancho Seco Nuclear Power Generating Station ** failed to open when its UVTA was actuated during a test. The RTBs at Rancho Seco are General Electric (GE) type AK-2-25 and had iust been refurbished by GE-Atlanta and certified for service as safety-related PfBs by Babcock and Wilcox (B&W) Lynchburg. Although the UVTA had de-energized, its armature had not moved out of the energized position. Investigatico revealed that the trip paddle had jammed against the armature, and as a result, the RTB would not trip.

Subsequently, when the shunt trip coil was actuated, the trip paddle rotated to a position in which the armature could not engage the trip paddle then the UVTA is de energized. Further investigation revealed that the clearance between the roller rivet and armature within the UVTA was significantly greater thao the specified allowable range. This increased downward displacement was sufficient to allow the trip paddle to interfere with the armature. On July 17, 1985, Inspection and Enforcement Information Notice No. 85-58 (Ref. B-2) describing this problem was issued to all nuclear power facilities designed by B&W and Combustion Engineering (CE) and holding an operating license or construction permit. ,

Recently, there have been two failures of the UVTAs in the GE AK-2-25 type RTBs at the Calvert Cliffs Nuclear Power Plant:*** one in February 1985 and one in July 1985. In the first failure, analysis revealed that several laminated sec-tions that are part of the armature had slipped down and eliminated the air gap between the armature and the pole face. The physical contact between the lami-nations and pole face allowed the armature to be held down by residual magnetism after the dc power was removed, resulting in a slow response time, well above the licensee's acceptance criteria. In the second failure, laminations had moved down only slightly to make the air gap below tolerance, but physical contact with the pole face was not established. While this configuration did not affect the response time, it did affect the pickup and dropout voltages of the UVTA, causing them to be low.

• Kewaunee is a pressurized water reactor operated by Wisconsin Public Service Corporation and located in Kewaunee County, Wisconsin.

• • Rancho Seco is a pressurized water reactor operated by Sacremento Municipal Utility District and located in Sacramento County, California.

• • • Calvert Cliffs Units 1 and 2 are pressurized water reactors operated by Baltimore Gas & Electric Company and located in Calvert County, Maryland.

18

On July 22, 1985, one of the GE AK-2-25 dc RTBs at the Oconee Nuclear Station Unit 1* failed to meet its required trip response time during on-line testing of an RPS channel while the unit was operating at 100% power. The failed UVTA was a new device installed on the RTB. A detailed inspection of the breaker showed a metal burr on the head of one of the mounting studs for the UVTA. The licensee concluded that the probable cause of the failure was that the armature of the UVTA brushed against the mounting stud as the armature moved toward the tripped position. It also was shown that the mounting stud heads of the new devices had square edges rather than round ones like on the older devices.

According to the licensee, the possible reduced clearance between the armature and the mounting studs could have caused the contact, resulting in a slow res-ponse time. The licensee's review of previous RTB failures has indicated that another new UVTA installed in the same breaker had failed once before on April 29, 1985, because of mechanical binding. The cause of the previous failure, which also resulted in a slow trip response time, had been attributed to some parti-cles, possibly paint chips or metal shavings, stuck in the pivot point of the UVTA.

On December 17, 1985, with Arkansas Unit 2** operating at 30% power, a Plant Protective System (PPS) monthly surveillance revealed that the armatures on undervoltage devices for RTBs 4 and 8 were in a mid position. The technician responsible for performance of the Channel 'B' PPS test suspended the test and notified the shift supervisor of his finding. A licensed control room operator and electrical maintenance personnel were dispatched to the RTBs to verify and evaluate the finding. The electrical maintenance personnel tripped the RTBs by perturbing the magnetic field of the undervoltage device causing it to actuate the RTB trip shaft demonstrating actuation capability. The RTBs were removed for testing and assessment of the undervoltage devices. These tests showed that from the approximated "as found" condition, the undervoltage devices could successfully trip the RTBs open when power was removed from the device simulat-ing a PPS actuation. The cause of the event is unknown, but proper verification of the position of the undervoltage devices when the RTBs were closed for plant startup on December 14, 1985 could possibly have prevented this incident. As a result, procedures are being changed to require a visual verification of the undervoltage armature position when closing the RTBs. Since the undervoltage devices were capable of performing their design function, there was no degrada-tion of the level of safety afforded by this system nor was there a violation of the unit's Technical Specification.

In addition to the RTB failures discussed above, GE notified the NRC and affected facilities on September 13, 1985, of certain defects in the UV trip devices supplied for use on AK and AKR type low-voltage power circuit breakers. Subse-quently, GE issued Service Advice Letter No. 300 on September 26, 1985, that outlines the actions to be taken with respect to those defects. One of the defects addressed by GE involved insufficient clearance between the armature and mounting stud, similar to the Oconee problem. The other defect involved improper painting of the mating surfaces of the armature and pole pieces in ac powered UV devices.

• 0conee Units 1, 2, and 3 are pressurized water reactors operated by Duke Power Company and located in Oconee County, South Carolina.

• • Arkansas Unit 2 is a pressurized water reactor operated by Arkansas Power

& Light Company and is located in Pope County, Arkansas.

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Information regarding the events at Calvert Cliffs, Oconee, and the GE notifica-tion letter was previded to all nuclear power facilities designed by B&W and GE and holding an operating license or construction permit in Inspection and En-forcement Information Notice No. 85-58, Supplement 1, on November 19, 1985 (Ref. B-3).

The specific and general issues associated with these RTB UVTA failures continue to be under active review by the nuclear industry and the NRC. However, unless significantly new issues are identified, this item is considered closed for the purposes of this report.

83-6 Uncontrolled Leakage of Reactor Coolant Outside Primary Containment This abnormal occurrence, which described a sustained and uncontrolled loss of hot pressurized reactor cooling water outside the primary containment at Hatch Unit 2 on August 25, 1982, was reported and closed out in NUREG-0090, Vol. 6, No. 3, " Report to Congrass on Abnormal Occurrences: July-Se tember 1983." It was reopened, and closed out, in NUREG-0090, Vol. 8, No. 2, p' Report to Congress on Abnormal Occurrences: April-June 1985," to report an event with certain similarities which occurred on June 12, 1985, at Oyster Creek Unit 1.

The following event also has certain similarities to the two previously reported events. The event occurred on September 19, 1985, at Dresden Unit 3 while the plant was at 83% power. Dresden Unit 3 is a General Electric-designed boiling water reactor operated by Commonwealth Edison Company. The plant is located in Grundy County, Illinois.

The reactor tripped when a technician accidentally moved a circuit card while removing a temporary lead used in testing. The movement of the card led to the closing of the turbine control valves, triggering the reactor shutdown.

During the recovery from the reactor trip, difficulty was encountered in reset-ting the reactor protection system (RPS) channel B. Channel A was reset without incident. In this configuration (i.e., channel A RPS reset, channel B RPS still in the scram condition) the air pressure system, which operates valves in the scram discharge system, was not at full operating pressure. As a result the scram discharge volume (SDV) vent and drain valves opened, but the air pres-sure was insufficient to close the scram inlet and outlet valves.

The SDV is a system of pipes which contain the water displaced when the control rods are driven into the reactor core during a trip (scram). After reactor scram, normal procedure is to reset the RPS circuits, which then allows the ac-cumulated water to drain from the SDV. At the same time, the scram inlet and outlet valves are designed to close automatically to cut off any flow of reactor cooling water into the SDV.

In tue Dresden Unit 3 event, the valve alignment (SDV vent and drain valves open and the scram inlet and outlet valves remain open) allowed reactor cooling water to flow through the seals on the control rod drives and into the reactor building through the SDV vents and drains. The vents are open to the reactor building atmosphere, and the drain valves are routed to the reactor building equipment drain tank.

20 l

Once the reactor operators discovered the leakage path, the SDV vent and drain valves were manually closed, terminating the reactor cooling water leakage after about 23 minutes. Steam released through the leakage pathway caused some con-tamination of the lower three levels of the reactor building.

About 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 16 minutes after the scram, licensee personnel were able to fully reset the RPS. Apparently, the electrical contacts on the reactor mode switch were in an intermediate position between the " shutdown" and " refuel" positions, instead of being fully in the " refuel" position necessary to reset the RPS.

The licensee's corrective action included extensive testing, equipment overhaul, and procedure changes.

Testing of the air pressure system--which operates the valves in question--

showed that in all cases the air pressure was degraded when the A channel of the RPS was reset, but the B channel remained tripped. In a reverse situation, with A channel tripped and B channel reset, the air pressure promptly returned to normal levels. The degraded air pressure was attributed to air leakage in the scram pilot air solenoid valves. Twenty of the scram 3ilot air solenoid valves were rebuilt to determine if the air leakage could )e eliminated, but subsequent testing showed that the system air pressure still remained low (with i

the A channel reset and the B channel remaining tripped).

The licensee's engineering department is studying the air supply system to determine if any improvements can be made.

The licensee has revised its scram recovery procedures to make certain the SDV vent and drain valves are closed prior to the resetting of the RPS. This would be expected to prevent a recurrence of the September 19, 1985, event because the closed vent and drain valves would prevent any reactor cooling water releases if the air system pressure were degraded. Inaddition,thescramrecoverypro-cedure now provides that the reactor mode switch be placed in the " shutdown' position to help prevent any further mispositioning of the mode switch.

The scram recovery procedure revision also contains an instruction to trip the l

l RPS fully if one channel fails to reset.

l The event was reviewed by the NRC Resident Inspector and by Office of Inspection and Enforcement personnel. In addition, on December 23, 1985, the NRC issued Inspection and Enforcement Information Notice No. 85-95 (" Leak of Reactor Water l to Reactor Building Caused by Scram Solenoid Valve Problem") to all boiling l

water reactor facilities to inform them of the event (Ref. B-4).

This item is considered closed for the purposes of thi's report.

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84-2 Through Wall Crack in Vent Header Inside BWR Containment Torus This abnormal occurrence, discovered at Hatch Unit 2 on February 3, 1984, was originally reported in NUREG-0090, Vol. 7, No. 1, " Report to Congress on Abnormal Occurrences: January-March 1984." It was updated and closed out in NUREG-0090, 21 i

i Vol. 7, No. 4. It is being reopened to report a recent similar event at Hatch Unit 1. Hatch Units 1 and 2 are boiling water reactors operated by Georgia Power Company and located in Appling County, Georgia.

As discussed in NUREG-0090, Vol. 7, No. 1, on February 3, 1985, a through wall crack was discovered in the vent header within the containment torus which de-graded the containment pressure suppression capability of Hatch Unit 2. The cause of this crack was determined to be brittle fracture and was attributed to impingement of low temperature nitrogen onto the vent header. It was learned that the licensee had discovered yet another crack on December 15, 1984,.during inservice inspection of selected welds. This time the crack was in the Unit 1 i drywell inerting and purge line. In a recent meeting with the NRC staff, the licensee stated that the likely cause of the crack was the growth of a pre-existing weld defect although cold nitrogen injection from the makeup system could not be ruled out as the primary cause.

! The crack was discovered during inservice inspection (ISI) tests which used a magnetic particle inspection method. During this testing a linear through wall crack approximately 2-3/4 inches long was discovered in a weld located in the

! 18-inch nitrogen inerting and purge line between the drywell penetration and i

inboard containment isolation valve. The crack was located approximately two feet i downstream from the point where a 2-inch nitrogen makeup line enters the 18-inch l purge line. Cold nitrogen entering the purge line from the makeup line makes contact with the purge line in the area of the weld, the contour of which makes it a likely spot for a thermally induced failure to occur.

The crack has been ground out and the weld has been repaired and satisfactorily tested. The licensee's corrective actions included raising the nitrogen makeup low temperature alarm set point to 10 F from 0 F and revising the annunciator response procedures to terminate nitrogen makeup when the annunciator is actuated.

In meetings with the licensee, the NRC staff has learned that the licensee per-formed only a visual inspection of the Unit 1 nitrogen inerting and purge line welds rather than the ultrasonic testing (UT) recommended by the reactor vendor (General Electric) as follow-up to the original Hatch Unit 2 vent header crack event. (Most utilities responding to a staff questionnaire indicated they did not consider the NRC Bulletin applicable to the nitrogen purge lines. They also indicated that they did not use UT.) The visual inspections would not have discovered the crack found during the ISI. Therefore, the crack could have existed at the time vendor-recommended inspections were implemented. In addition, the nitrogen makeup line was not considered by the licensee to be covered within the scope of the vendor's Service Information Letter (SIL) as a source of cold nitrogen (less than 40 F) into containment systems, because of:

the use of relatively low flow rates; and the long run of bare piping inside t the warm reactor building. The licensee did confirm, however, the makeup portion of the inerting system was in use for an 18 1/2-hour period in February 1984 I. when the ambient temperature did not get above 22 F. Makeup line nitrogen I temperatures during this period were not known.

On December 31, 1985, the NRC issued Inspection and Enforcement Information Notice No. 85-99 (Ref. B-5) which describes the Unit 1 crack experience and the

, licensee's corrective actions. In addition, on December 23, 1985, the NRC issued l a letter to the BWR Owners Group requesting a meeting to review the effectiveness 22

of voluntary industry actions taken in response to the original vent header crack at Hatch Unit 2 in view of the crack discovered at Unit 1.

Unless new, significant information becomes available, this item is considered closed for the purposes of this report. The NRC staff will continue to review the industry actions taken in response to the SIL and plans to meet with the BWR Owners Group for a regulatory closecut of the issue.

85-7 Loss of Main and Auxiliary Feedwater Systems This abnormal occurrence, which occurred at Davis-Besse on June 9, 1985, was originally reported in NUREG-0090, Vol. 8, No. 2, " Report to Congress on Abnormal Occurrences: April-June 1985," and updated in NUREG-0090, Vol. 8, No. 3. It is further updated as follows.

As mentioned in the previous update, based upon the findings of the NRC Incident Investigation Team reported in NUREG-1154 (Ref. B-6), the NRC identified the concerns Toledo Edison Company (the licensee) should address for NRC review before resumption of operation of the plant can be approved. These concerns were identified to the licensee in a letter dated August 14, 1985 (Ref. B-7).

The licensee has responded in a document submitted to the NRC on September 10, 1985, entitled, " Davis-Besse Course of Action" (Ref. B-8). The NRC staff has essentially completed its review of this document and is currently preparing a Safety Evaluation Report to address plant re-start. The NRC Commissioners were briefed by the staff on December 18, 1985, on the status of the staff's review regarding plant re-start.

NRC Region III continues to monitor the licensee's activities at the plant site using two Resident Inspectors, supplemented as required by Regional Office-based personnel. Major activities have included:

1. The installation of the new startup feedwater pump, which is powered by an electric motor, has been completed and testing is complete except for tests at pressure. NRC inspectors observed the installation activities and continue to monitor the testing.

2. The licensee has completed its troubleshooting and root cause evaluation of the equipment that failed during the June 9, 1985, event. The reports of its findings, which has been submitted to the NRC, includes an action plan for each item to resolve the problems which occurred. NRC Region III is monitoring the actions being taken.

3. The inspection of 2,425 pipe hangers and supports for safety-related systems was completed in late December 1985 and the results are undergoing evaluation by the licensee. The inspections led to preparation of Nonconformance Reports on 2,152 of the hangers and supports. These Nonconformance Reports are currently being evaluated by the licensee.

4. Major work on the System Test and Review Program was initiated in late l

1985. This program involves a thorough review of 34 systems important to safety. The intent of the program is to verify that the systems would operate 23 I

when needed under all credible operating and accident conditions. An NRC test review team is providing weekly coverage for this program.

5. On December 13, 1985, the Director, NRC Office of Inspection and En-forcement, issued a Notice of Violation and Proposed Imposition of Civil Penalties to Toledo Edison Company for 10 violations associated with the June 9, 1985, loss of feedwater event (Ref. B-9). The proposed civil penalties included

$100,000 for each of eight violations, and $50,000 for each of the remaining two, for a W al of $900,000.

The eight violations, each carrying a $100,000 fine, were:

(a) Failure to identify that the torque switch bypass settings for certain safety-related motor operated valves were improper. Proper testing would have shown that the design settings on the switches involved with the closed valves were improper, causing the valves to remain shut when called upon to open.

(b) Failure to determine that the auxiliary feedwater pump turbines would not operate properly with steam supplied through the crossover steam )iping.

(c) Failure of its preoperational testing program to identify t1at the two auxiliary feedwater isolation valves would not function under certain conditions.

(d) Failure to assure that maintenance procedures for the performance of some safety-related activities were adequate.

(e) Failure to test certain valves quarterly as required.

(f) Failure to take prompt and adequate corrective action when a problem was identified previously with one of the two closed valves.

(g) Failure to satisfy the NRC requirements that no single failure would prevent safety systems from functioning (the two valves remained closed even after the operator pushed the correct buttons).

(h) Failure to adequately train auxiliary operators on resetting the over-speed trip mechanism on the auxiliary feedwater pump turbines. .

The two violations, each carrying a $50,000 fine, were:

(i) Failure by the licensee to identify maintenance and troubleshooting deficiencies at the plant.

(j) Failure to verify that the installations of pipe hangers were carried out in conformance with design drawings.

Inspection and Enforcement Bulletin No. 85-03, " Motor-0perated Valve Common Mode Failure during Plant Transients Due to Improper Switch Settings," was issued on November 15, 1985 (Ref. B-10), in part because of experience gained during the event at the Davis Besse plant on June 9, 1985. The bulletin describes a problem resulting from improper setting of torque bypass switches on auxiliary feedwater valves equipped with Limitorque motor operators. Torque switches are incorporated in Limitorque motor operators to prevent damage to the valve or operator when motor torque is excessive. The torque bypass switches are neces-sary so that larger than normal torques can be applied when unseating valve disks or gates. At Davis Besse, the torque bypass switches were set so that they would not accommodate the increased force caused by the pressure differ-entials across the valves when the valves were in the closed position. Thus the valves could not be unseated before the torque switches were enabled and automatically disconnected power to the motor operators. This caused delay in i reestablishing adequate cooling for the reactor.

t i

24

The bulletin also described other events at various plants which involved motor operated valves failing on demand, in a common mode, due to improper switch settings. The bulletin requested all holders of nuclear power operating licenses or construction permits 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.

l Future reports will be made as appropriate.

FUEL CYCLE FACILITIES (Other than Nuclear Power Plants) 84-19 Buildup of Uranium in a Ventilation System This abnormal occurrence was originally reported and closed out in NUREG-0090, Vol. 7, No. 4, " Report to Congress on Abnormal Occurrences: October-December 1984." It is being reopened to report updating information.

As previously reported, on October 5, 1984, Nuclear Fuel Services, Inc. notified the NRC that an excessive buildup of uranium had been discovered in the new ventilation system (including a scrubber) of the scrap recovery facility at their plant located near Erwin, Tennessee. The previously reported actions taken to prevent recurrence are updated as follows.

Licensee - The licensee implemented a program for routinely monitoring material accumulation in the ventilation systems throughout the plant. The NRC approved a license amendment incorporating this monitoring program. The licensee con-ducted a design review to identify engineering improvements which will reduce the amount of uranium entering the ventilation system. The licensee installed a cyclone separator in the ventilation line upstream of the venturi scrubber which is removing solid material before it reaches the scrubber. Also, a portion of the duct and bottom elbow where the air enters the scrubber is being continu-ously flushed with liquid to prevent buildup. Status reports on these engineer-ing improvements are provided to the NRC on a routine basis. In addition, the licensee will respond to the NRC enforcement action described below.

NRC - A special inspection was conducted at the licensee's Erwin, Tennessee facility by the NRC Region II Office during the period of October 5-18, 1984.

Significant failures to comply with NRC regulatory requirements were identified, i.e., failure to perform adequate investigations and take appropriate corrective actions, as required by the license, for violations of criticality safety action limits placed on the accumulation of uranium in the ventilation system. The conditions of degraded safety and safeguards had existed for a significant period of time.

An Enforcement Conference to discuss these matters was held with the licensee at the NRC Region II Office on October 29, 1984. On February 21, 1985, the NRC l forwarded to the licensee a Notice of Violation and Proposed Imposition of Civil

! Penalty in the amount of $20,000 (Ref. B-11). In addition to the civil penalty, ,

, the NRC believed that further remedial action was needed to ensure that the licensee improved management oversight of operations and initiated appropriate investigations when action limits were exceeded. Therefore, the February 21, l 1985, NRC letter also enclosed an Order Modifying License. The Order amended l

l 25 l

l

the license to require the licensee to expand the duties and responsibilities of its Internally Authorized Change Council. The licensee denied the violation, requested mitigation of the civil penalty, and requested modification of cer-tain language in the Order Modifying License. On October 18, 1985, the NRC issued a revised Order Modifying License to incorporate changes (Ref. B-12).

As a result of staff review of the licensee's response to the Proposed Imposi-tion of Civil Penalty, the NRC issued an Order Imposing Civil Monetary Penalty on November 27, 1985 (Ref. B-13). This order reduced the civil penalty by 25%

l to $15,000 because of the licensee's extensive and comprehensive corrective actions. The civil penalty was paid on December 30, 1985.

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

OTHER NRC LICENSEES 85-8 Diagnostic Medical Misadministration This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 2,

" Report to Congress on Abnormal Occurrences: April-June 1985," and is updated as follows.

As described in the previous report, on August 7, 1984, a patient was admin-istered a 10 millicurie dose of iodine-131 instead of an intended 400 micro-curie dose of iodine-123. The misadministration occurred at the Hospital of St. Raphael located in New Haven, Connecticut. NRC Region I requested an NRC medical consultant to assess the clinical aspects of the occurrence and the appropriateness of the actions taken by the licensee.

The NRC medical consultant agrees with the licensee's treatment plan and believes

( that the prognosis may be slightly more favorable than that originally expected t

by the licensee. The consultant concluded that the dose to the thyroid was approximately 6000 rads. The licensee was issued a Notice of Violation by NRC Region I for failing to report the misadministration to the NRC within the required time and for failing to require a written request from the physician performing the procedure.  ;

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

85-9 Diagnostic Medical Misadministration

! This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 2,

" Report to Congress on Abnormal Occurrences: April-June 1985," and is updated as follows.

As described in the previous report, on March 19, 1985, a patient was admin-istered 5 millicuries of iodine-131 instead of an intended 10 millicuries of technetium-99m for a routine thyroid scan. The misadministration occurred at Mercy Hospital located in Pittsburgh, Pennsylvania. NRC Region I requested an NRC medical consultant to review the event.

The NRC redical consultant agreed with the licensee's prompt action in admin-istering potassium perchlorat'a. He estimated a dose of 45 rads to the thyroid 26 1

which is considerably less than the 1000 rads estimated by the licensee. The l consultant concluded that the increase in the probability of developing cancer from this exposure was inconsequential especially in light of the fact that the patient has subsequently undergone a partial thyroidectomy.

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

l ********

i 85-15 Therapeutic Medical Misadministration This abnormal occurrence was originally reported, and closed out, in NUREG-0090, Vol. 8, No. 3, " Report to Congress on Abnormal Occurrences: July-September l

1985." It is being reopened to report updated information.

As described in the previous report, from October 17, 1987 to November 1, 1984, a patient received a radiotherapy administration of 3584 rads to a portion of

, the body instead of an intended 2000 rads. The misadministration occurred at I the University Health Center of Pittsburgh's Joint Radiation Oncology Center.

l Magee-Women's Hospital site, located in Pittsburgh, Pennsylvania. NRC Region I l

requested an NRC medical consultant to review the case.

The medical consultant confirmed that the second course of treatment to meta-stases involving the 9th and 10th ribs was about 40% greater than that which had been planned. There has been a moderate amount of fibrosis in this area but the tumor control has been excellent with no recurrence of pain or other complairit. He concluded that the misadministration did not result in significant harm to this patient and in fact might be suspected to have produced an excellent palliative therapeutic result.

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

85-16 Therapeutic Medical Misadministration This abnormal occurrence was originally reported, and closed out, in NUREG-0090, Vol. 8, No. 3, " Report to Congress of Abnormal Occurrences: July-September 1985." It is being reopened to report updated information.

As described in the previous repo t, on October 25, 1984, the NRC was notified that a patient received 15 millicuries of iodine-131 rather than the prescribed dose of 10 millicuries for treatment of hyperthyroidism. The misadministration occurred at the Milton S. Hershey Medical Center located in Hershey, Pennsyl-vania. NRC Region I requested an NRC medical consultant to review the event.

The medical consultant estimated the delivered dose to the thyroid to be 12,000 rads rather than the intended 8,000 rads. The patient is clinically hypothyroid and is doing well on medication. The consultant concluded that hypothyroidism also would have resulted from the intended dose and that many clinicians routinely choose the higher dose for initial treatment.

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

27

APPENDIX C OTHER EVENTS OF INTEREST The following items are described below because they may possibly be perceived by the public to be of public health significance. The items did not involve a major reduction in the level of protection provided for public health or safety; therefore, they are not reportable as abnormal occurrences.

1. Systems Interaction Event On May 15, 1985, Hatch Unit 1 experienced a systems interaction event due to leakage of fire protection water through the ventilation system into instrument cabinets, shorting one of the two redundant power supplies and/or possible inter-mittent shorting of logic system contacts in the Analog Transmitter Trip System (ATTS) panel. Hatch Unit 1 is a General Electric-designed boiling water reactor operated by Georgia Power Company. The facility is located in Appling County, Georgia.

The event began about 8:35 p.m. when an instrument water line was damaged, appar-ently by a crane hook that was dragged below a crane that was passing above the control room ventilation system equipment along the line. The instrument water supply line depressurized causing the fire suppression water deluge system for the charcoal filters in the control room ventilation system to actuate. The i deluge system is located in the control room ventilation system charcoal filter housing which is located above the control room.

Following actuation of the deluge system, approximately 15 to 25 gallons of water backed up into the ventilation header before the system could be secured.

The backup was caused by plugged drains in the charcoal filter housing. Water eventually leaked through a hole in the ventilation ductwork that was located above the ATTS panel in the control room. When the water sprayed onto the panel, one of two redundant panel power supplies ap?arently shorted because of water intrusion inW the p nel. As a result, a low low-set safety relief valve

(LLS-SRV) began to cycle open and closed. The SRV cycled three times and then opened and remained open. The operators attempted to deenergize the open SRV in accordance with procedure HNP-1-1907, Failure of Safety / Relief Valves to Operate, but could not locate the correct fuse within the panel. The procedure had not been revised to incorporate the new fuse locations for the SRVs. The l operator manually scrammed the reactor from 75% power. A false turbine high

! exhaust pressure trip signal also was generated, temporarily disabling the high j pressure core injection (HPCI) system. The reactor core isolation cooling (RCIC) i system was inoperable at the time, so neither HPCI nor RCIC was immediately available for use. Fortunately, neither system was needed during the event.

This is because the water level was restored and maintained by the reactor feed-

water system until the MSIVs were shut. Subsequent to MSIV closure, water level

! was maintained by the control rod drive (CRD) system with the excess water being dumped to the condenser via the reactor water cleanup system. The LLS-SRV closed without operator action at 9:52 p.m.

The event is of considerable concern because of the potential for multiple safety system failures through unanalyzed systems interactions. In this event, the water from the fire suppression deluge system in the control room caused opening l of a safety relief valve and loss of primary system inventory. The event could 1

! 29

have been seriously aggravated by the spurious HPCI turbine high watec intrusion.

Because the RCIC system was inoperable at the time of the event, no sifety-related high pressure injection system would have been immediately available to restore water level should that have been necessary. The HPCI turbine trip signal was reset shortly after it occurred, however, and the system was returned to operability.

Perhaps more serious is the potential effect the water could have had on numerous other safety systems. The ATTS panels have permissive and arming logic and trip logic for various safety systems, as well as water level inputs to the HPCI, RCIC, core spray (CS), automatic depressurization system (ADS), residual heat removal (RHR) system, and diesel activation logic. It is difficult to predict the anomalous behavior that could occur if both power supplies had been lost, or if other portions of the logic had been shorted; but quite possibly, several safety systems could have malfunctioned, seriously handicapping the operators during their efforts to stabilize the unit.

Prior to this event, no procedures were in place at Hatch Unit 1 for adequately cleaning the ventilation plenums or drains in the charcoal filter units. Had these procedures been prepared and implemented, the drains would have functioned as designed with no serious adverse effects. In response to this event, the licensee cleaned and inspected drains in the remaining filter units and prepared cleanout and inspection procedures to be added to the maintenance schedules.

Since the event, procedure HNP-1-1907 has been replaced by procedure 34AB-0PS-007-1 which reflects the correct SRV fuse locations. Also, the design change request procedure, HNP-0-0809, which was then used when the work was completed for the SRV fuses in November 1984, has since been replaced by procedure 42EN-ENG01-0. The failure to incorporate the design change for the SRV fuse locations in the operating procedure was charged as a violation by the NRC.

All other plant equipment functioned as required during this event, and plant personnel responded promptly and correctly. These actions limited the conse-quences such that the only effect was a stuck open relief valve, which did not adversely affect reactor safety during this event. Therefore, the event had no adverse affect on plant safety, and the health and safety cf the public were not affected.

On October 31, 1985, the NRC issued Inspection and Enforcement Information Notice No. 85-85 (" Systems Interaction Event Resulting in Reactor System Safety Relief Valve Opening Following a Fire-Protection Deluge System Malfunction") to all nuclear power reactor facilities holding an operating license or a construction permit to inform them of the event (Ref. C-1).

2. Suspension of License of a Medical Research Institute During an NRC inspection on July 12, 1984, at the Institute for Medical Research of Bennington (Vermont), the licensee was found to be in possession of quantities of radioactive material in excess of those authorized by the license and to have permitted use of licensed materials although all individuals previously approved to supervise use of materials had left the Institute. On July 18, 1984, the Institute committed to several actions which were documented in a Confirmatory Action Letter from Region I. These actions included the cessation l

30 1

of all uses of licensed material until a request for amendment of the Institute's license was approved to include qualified individuals to supervise the use of licensed material.

An unannounced re-inspection was conducted on September 19, 1985, since the Institute had not yet requested an amendment to the license to add qualified users. It was discovered that small quantities of licensed material had been used by two investigators after the licensee's commitment to cease all such use. In addition, during November 1984, the Institute's possession of licensed material had again exceeded the limits specified in the license. Because of the licensee's failure and apparent inability to comply with the conditions of its license or its commitments to NRC, an Order Suspending License (Effective Immediately) was issued on October 25, 1985 by the Director of the NRC Office of Inspection and Enforcement (Ref. C-2).

On November 6, 1985, an enforcement conference was held with the management of the Institute. The Institute's management admitted that it had not exercised the appropriate control over the program and had not understood the requirements of the license or the NRC's regulations. As a result of this meeting, the li-censee agreed to revise the radiation safety program including procedures for ordering material, training and approval of personnel, inventories, and proce-dures for safe use of material.

By application dated December 5, 1985, the Institute requested a license amend-ment to include these new procedures and to add qualified individuals to super-vise the use of materials. The licensee submitted documentation to the Region I Office that demonstrated fulfillment of the commitments made during the Novem-ber 6, 1985 enforcement conference. Subsequently, in a letter to the Institute dated March 20, 1986, the Region I Administrator concluded that the terms of the Order suspending the license had been met and approved resumption of licensed activities.

The event did not result in a significant threat to licensee personnel or to public health or safety. However, action to suspend the license was necessary to assure corrective actions were taken in response to the NRC inspection find-ings that the licensee exhibited a complete failure to understand NRC rules and regulations and that licensee management was not exercising proper control over the use of licensed materials.

3. HPCI Turbine Exhaust Multiple Check Valve Failures On November 4, 1985, while the plant was shut down, Long Island Lighting Company found that all three valves (two check valves and one gate valve) in the High Pressure Coolant Injection (HPCI) System turbine steam discharge line were de-graded at Shoreham Nuclear Power Station Unit 1. Shoreham is a General Electric-designed boiling water reactor located in Suffolk County, Long Island, New York.

At Shoreham, the HPCI pumps are driven by a single steam turbine. In the un-likely event of a loss of coolant accident (LOCA), nuclear steam from a main steam line drives the turbine. The steam leaves the turbine via the steam dis-charge line, flowing through two Anchor-Darling swing check valves (18V-21 and 18V-22), a motor operated gate valve (M0V-44), into the primary containment, and discharging at low pressure into the suppression pool water volume where 31

the steam is condensed. (The turbine can also be driven by auxiliary steam when, for example, the system is tested during a plant shutdown.)

On October 30, 1985, prior to finding the two check valves inoperable, mainten-ance personnel were working on gate valve MOV-44 to reduce valve leakage when unidentified components were found in the valve. Upon further investigation, it was determined that the unidentified components were the disc, swing arm and bolt from the check valve immediately upstream, 18V-22 (HPCI Turbine Exhaust Downstream Check Valve). The disc was wedged at the inlet of the gate valve.

On November 2, 1985, maintenance was initiated to repair 18V-22 in order to restore the valve to operational status. Maintenance personnel disassembled the valve and discovered the disc and swing arm from the second check valve further upstream in the line, 18V-21 (HPCI Turbine Exhaust Upstream Check Valve),

wecged in the inlet of 18V-22. On November 4, inspection and repair of 18V-21 began in order to restore the valve to operational status. The valve was disassembled and absence of the disc and mechanism was confirmed.

HPCI was last run on September 25, 1985 without incident. MOV-44 was last stroked satisfactorily on July 30, 1985 per Technical Specification requirements.

A local leak rate test was last performed satisfactorily on June 14, 1984 for penetration X-13, which is composed of MOV-44, 18V-22 and 18V-21.

Further investigation of the malfunctions indicate that, in both cases, the hinge support piece separated from the bonnet, allowing the disc of each check valve to move downstream and become wedged in the inlet of the downstream valve.

The hinge support piece came loose from the valve bonner, because the cap screws that fasten the support piece to the bonnet were not lock welded per vendor drawings, and backed out during operation.

The internals of the damaged valves are being repaired by the vendor, after which the hinge support piece will be reinstalled with lock wired cap screws, and the valves returned to operational status. The NRC Resident and Regional Inspectors will continue to monitor the progress of the licensee's corrective actions.

The licensee has committed to replacing the existing swing check valves with a new design lift check at the first refueling outage. The lift type check valve is an improved design for applications in steam systems. This requires extensive stress analysis and pipe support analysis due to the increased weight of the lift check valve.

Even though the event had minimal effect on public health or safety, the con-dition did degrade the HPCI system. The valves are considered primary contain-ment isolation valves.

The event is of interest since it represented multiple failures of components due to a single root cause, with possible generic implications.

Damage to, and potential failure of, HPCI and Reactor Core Isolation Cooling (RCIC) System turbine exhaust swing check valves in boiling water reactors has been a subject of concern for some time. Inspection and Enforcement Information Notice No. 82-26 (Ref. C-3) and General Electric (GE) Application Information Document No. 56 both addressed this subject, and recommended that certain actions 32

s be taken to minimize the possibility of valve damage and/or failure. These recommendations included: 1) certain starting and o)erating precautions, 2) that the check valve, exhaust line vacuum breaker, and ex1aust line sparger be de-signed in accordance with the GE system design specifications, 3) that the check valve be located as close to the containment as possible, and 4) that the turbine exhaust check valve internals should be visually inspected on a routine schedule such as at every refueling outage.

33

1 REFERENCES FOR APPENDICES B-1 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No.85-02,"UndervoltageTripAttachmentsofWestinghouseDB-50 Type Reactor Trip Breakers, November 5, 1985.*

B-2 U.S. Nuclear Regulatory" Commission, Inspection and Enforcement Informa-I tion Notice No. 85-58, Failure of a General Electric Type AK-2-25 Reactor

! Trip Breaker," July 17, 1985.*

B-3 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Informa-tion Notice No. 85-58, Suaplement 1, " Failure of a General Electric Type l AK-2-25 Reactor Trip Breaker," November 19, 1985.*

l B-4 U.S. Nuclear Regulatory" Commission, Inspection and Enforcement Informa-l tion Notice No. 85-95, Leak of Reactor Water to Reactor Building Caused by Scram Solenoid Valve Problem," December 23, 1985.*

B-5 U.S. Nuclear Regulatory" Commission, Inspection and Enforcement Informa-tion Notice No. 85-99, Cracking in Boiling-Water-Reactor Mark I and

Mark II Containments Caused by Failure of the Inerting System,"

l December 31, 1985.*

B-6 U.S. Nuclear Regulatory Commission, " Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9, 1985," USNRC Report NUREG-1154, published July 1985.**

B-7 10 CFR S 10.54(f) letter from Harold R. Denton, Director, NRC Office of l Nuclear Reactor Regulation, to Joe Williams, Jr., Senior Vice President-Nuclear, Toledo Edison Company, Docket No. 50-346, August 14, 1985.*

B-8 Letter from John P. Williamson, Chairman and Chief Executive Officer, Toledo Edison Company, to Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, Docket No. 50-346, September 10, 1985.*

! B-9 letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to Joe Williams, Jr. , Senior Vice President-Nuclear, Toledo Edison Company, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, Docket No. 50-346, December 13, 1985.*

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

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

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s B-10 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 85-03, " Motor-0perated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings," November 15, 1985.*

B-11 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to C.W. Taylor, President, Nuclear Fuel Services, Inc., for-warding (a) Notice of Violation and Proposed Imposition of Civil Penalty, and (b) Order Modifying License, Docket No.70-143, February 21, 1985.*

B-12 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to F.K. Guinn, Plant Manager, Nuclear Fuel Services, Inc.,

forwarding a Revised Order Modifying License, Docket No.70-143, October 18, 1985.*

B-13 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to C.W. Taylor, President, Nuclear Fuel Services, Inc., for-warding an Order Imposing Civil Monetary Penalty with Appendix, Docket No.70-143, November 27, 1985.*

C-1 U.S. Nuclear Regulatory Commission, Ins)ection and Enforcement Information Notice No. 85-85, " Systems Interaction Event Resulting in Reactor System Safety Relief Valve Opening Following a Fire-Protection Deluge System Malfunction," October 31, 1985.*

C-2 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to Peter A. Lalley, Director, Institute for Medical Research of Bennington, forwarding an Order Suspending License (Effective Immedi-ately), License No. 44-18388-01, Docket No. 30-14984, October 25, 1985.*

C-3 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 82-26 "RCIC and HPCI Turbine Exhaust Check Valve Failures,"

July 22, 1982.* ,

9 36

u S. EUCLENA R EGULATFJV COuutssiON i atPOR T NuMSE A #dsti.a er FioC. saa vor No.

• ears SIAC Pomes 33s S','- BIBLIOGRAPHIC DATA SHEET Gg G0g nE N.TavCT,0N.ON T E .m sg 3 LE AVE SLANK f

2 TITLE AND SUSTITLE R: port to Congr s on Abnormal Occurrences October - Decemb r 1985 4 DjlIE REPORT COMPLET ED MoNT YEAR

.. .uT O.. . Mayr 1986

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May l 1986 y PE. FOmweNG ORGAN 82Af SON NAuf AND Malte ADDRESS fiac%mle Cool e Ps4DJECT 4T A$n veOmn UN#T NUMBER Office for Analysis and Evaluation of Operational Data /

U.S. Nuclear Regulatory onnission p "N oa oa AN' NuMsEa Washington, DC 20555

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i. s,oNso .No Oao AN,1 AT ,0N Navi ANo . Ait,No Acoat Same as 7, above. [ Quarterly D PERroD COk E RED fiacev s,ve mesi October - December 1985 12 SuPFLEMENT ARY NOf t1

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13 A85f xd.CT f200 words or 'enes Section 208 of the Energy Reorganiz , tion Act of 1974 identifies an abnormal occurrence as an unscheduled incid 'nh or event which the Nuclear Regulatory Commission determines to be signi cah from the standpoint of public health and safety and requires a quarte ly re rt of such events to be made to Congress. Thisreportcoverstfeperio October 1 to December 31, 1985.

During the report period, ther/ were two bnormal occurrences at the nuclear power plants licensed to opegate. The firgt involved inoperable main steam isolation valves and the segond involved ma agement deficiencies at Fermi Nuclear Power Station. The're were three abn rmal occurrences at the other NRC licensees. Two invol/ed diagiostic medic misadrr.inistrations and the other involved a therapetitic medical misadminis{tration. There were no abnormal occurrences re60rted by the Agreement States. The report also contains information 6 dating some previously repbr d abnormal occurrences.

19 DOCU.ENT AN ALvus - e at twomas Esca Ptoses it Av AiLAsitiT v ETATEVENT Unlimited

'6 SECLR TV CLASS'81 CATION Main Steam Isolation Valves; Inoperable Pilot Valves; "

uoENT.ntas.OPEN ENoto f t...

Degraded Elastomer Material; Management Deficiencies at Nuclear Power f,"(assified Plant; Unexpected Criticality; Personnel Errors; Therapeutic and Unclassified Diagnostic Medical Misadministrations; Systems Interaction Event-Multiple Check Valve Failures; License Suspension; Reactor Trip System Failures; Leakage of Reactor Coolant Outside Primary Containment; Pipe ,, ,a ,c t Cracks.

UNITED STATES ,,, cat ,oexTKCLAss RATE ' -

NUCLEAR RESULATORY COMMISSION Posiaos e nts ruo o WASHINGTON, D.C. 20666 wfsT"8c.

PtmMrT % G 47 0FFICIAL BUSINESS .,

PENALTY FOR PRIVATE USE,4300 ~;

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