ML20210J210

Report to Congress on Abnormal Occurrences.July - September 1985

ML20210J210
Person / Time
Issue date:	02/28/1986
From:	NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:
References
NUREG-0090, NUREG-0090-V08-N03, NUREG-90, NUREG-90-V8-N3, NUDOCS 8604030385
Download: ML20210J210 (48)

Text

.--- _ - _

~

NUREG-0090 Vol. 8, No. 3 i Report to Congress on Abnormal Occurrences July - September 1985 t

U.S. Nuclear Regulatory Commission Offica for Analysis and Evaluation of Operational Data

,p* *'%q, N

!!s84 0090 R a8H8 "' ""PDR

+ .

Available from Superintendent of Documents U.S. Government Printing Office

Post Office Box 37082 Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication.

Single copies of this publication

are available from National Technical informatum Sennce, Springfield, VA 22161 t

4

?

I

NUSEG-0090 Vol. 8, No. 3 Report to Congress on

. Abnormal Occurrences July - September 1985 1

Data Published: February 1986 i Offics for Analysis and Evsluation of Operational Data U.S. Nuclear Regulatory Commission Wcshington, D.C. 20555 y~ 9,

~. . i. . . f '

f

PrGvious Raporta la Sgries NUREG 75/090, Januarv-June 1975, NUREG-0090, Vol.3, No.3, July-September 1980, published October 1975 published 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 NU9EG-0090-3, Janu'ary-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-Seotember 1976, NUREG-0090 Vol .4, No.4, October-December 1981, published March 1977 published May 1982 NUREG-0090-6, October-December 1976 NUREG-0090, Vol .5, No.1, Janbary-March 1982, .>

published June 1977 published August 1982 NUREG-0090-7, January-March 1977, NUREG-0000, Vol .5, No.2, April-June 1982, published June 1977 published December 1982 NUREG-0090-8, April-June 1977 NUDEG-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-0040 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 Sept, ember 1978 published April 1984 NUREG-0090, Vol.1, No.3, July-September 1978 NUREG-0090, Vol .6, No.4, October-December 1983, published December 1978 published May 1984 NUREG-0090, Vol.1, No.4, October-December 1978, NUREG-0090, Vol.7, No.1, January-March 1o84, published March 1979 published July 1984 l

NUREG-0090, Vol.2, No.1, 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 19R4, ~

published November 1979 pubitshed April 1985 l

NUREG-0090, Vol.2, No.3, July-September 1979, NUDEG-0090, Vol .7, No.4, October-December 1984, published February 1980 published May 1985 NUREG-0090, Vol.2, No.4, October-December 1979, NUDEG-0000. Vol .R No.1, January-March 1985, published April 1980 published August 1Q85 NUREG-0090, Vol.3, No.1, January-March 1980, NUREG-0090, Vol.8, No.2 April-June 1985, published September 1980 published November 1985

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 July 1 to September 30, 1985.-

The report states that for this reporting period, there were three abnormal occurrences at the nuclear power plants licensed to operate. These events in-volved, respectively, (1) management control deficiencies, (2) inoperable steam generator low pressure trip, and (3) management deficiencies at Tennessee Valley Authority. There were four abnormal occurrences at the other NRC licensees.

Three of the events involved medical misadministrations - two therapeutic and one diagnostic. The other event involved exposure of radiographic personnel due to management and procedure control deficiencies. There were no. abnormal occur-rences reported by the Agreement States.

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

iii

CONTENTS P_ age 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, JULY-SEPTEMBER 1985 . . . . . . 1 NUCLEAR POWER PLANTS ............................................ 1 85-12 Management Control Deficiencies ..................... 1 85-13 Ir. operable Steam Generator Low Pressure Trip ........ 5 85-14 Management Deficiencies at Tennessee Valley Authority .................................... 9 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants) ......... 17 OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, Etc.) ......................... 17 l 85-15 Therapeutic Medical Misadministration ............... 18 I 85-16 Therapeutic Medical Misadministration ............... 20 85-17 Exposure of Radiographic Personnel Due to Management and Procedure Control Deficiencies........ 21 85-18 Diagnostic Medical Misadministration ................ 23 AGREEMENT STATE LICENSEES................... .................... 24 REFERENCES ........................................................... 25 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA ............................ 27 APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES ...... 29

[ NUCLEAR POWER PLANTS ............................................ 29 i

79-3 Nuclear Accident at Three Mile Island ................ 29 83-15 Emergency Diesel Generator Problems .................. 30 85-7 Loss of Main and Auxiliary Feedwater Systems ......... 31 APPENDIX C - OTHER EVENTS OF INTEREST ................................ 33 REFERENCES (FOR APPENDICES) ............................ ............. 39 v

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 ab-normal occurrences involving facilities and activities regulated by the NRC.

An abnormal occurrence is defined in Section 208 as an unscheduled incident or svent which the Commission determines is significant from the standpoint of public health or safety.

Events are currently identified as abnormal occurrences for this report by the NRC using the criteria delineated in Appendix A. These criteria were promul-gated in an NRC policy statement which was published in the Federal Register 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 dccument rooms. At a minimum, l

l 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 occur-rence reporting. This rmort covers the period from July 1 to September 30, 1985.

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

THE REGULATORY SYSTEM The system of licensing and regulation by which NRC carries out its responsi-bilities is implemented through rules and regulations in Title 10 of the Code of Federal Regulations. To accomplish its objectives, NRC regularly conducts licensing proceedings, inspection and enforcement activities, evaluation of operating experience and confirmatory research, while maintaining programs for establishing standards and issuing technical reviews and studies. The NRC's role'in regulating represents a complete cycle, with the NRC establishing stan-dards and rules; issuing licenses and permits; inspecting for compliance; en-forcing license requirements; and carrying on continuing evaluations, studies and research projects to improve both the regulatory process and the protec-tion of the public health and safety. Public participation is an element of the regulatory process.

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

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 anil enforcement program helps assure compliance with the regulations.

Most NRC licensee employets who work with or in the vicinity of radioactive materials are required to utilize personnel monitoring devices such as film badges or TLD (thermolumi1escent dosimeter) badges. These badges are pro-cessed 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 re-stricted 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 unrestricted areas, permissible levels of radiation are consider-ably smaller. Permissible 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-mation, and to feed back the experience into licensing, regulations, and operations.

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

Some form of LER reporting system has been in existence since the first nuclear port plant was licensed. Reporting requirements were delineated in the Code ot 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 informa' ion concerning reportable events, the Atomic Energy Commission (the predecessor of the NRC) established in 1973 a computer-based data file, with data extracted from licensee' reports dating from 1969. Periodically, changes were made to improve both the effectiveness of data processing and the quality of reports required to be' submitted by the licensees.

viii

1 1

Effective January 1, 1984, major changes were made to the requirements to report j 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 requirements of significant events at operating nuclear power reactors (10 CFR S 50.72). This was published in the Federal Register (48 FR 39039) on August 29, 1983, with corrections (48 FR 40882) published on September 12, 19S3.

Among the changes made were the use of terminology, phrasing, and reporting thresholds that are similar to those of 10 CFR S 50.73. Therefore, most events reported under 10 CFR S 50.72 will also require an in-depth follow-up report under 10 CFR S 50.73.

The NPRD system is a voluntary program for the reporting of reliability data by nuclear power plant licensees. Both engineering and failure data are to be submitted by licensees for specified plant components and systems. In the past, industry participation in the NPRD system was limited and, as a result, the Commission considered it may be necessary to make participation mandatory in order to make the system a viable tool in analyzing operating experience. How-ever, on June 8, 1981, INP0 announced that because of its role as an active user of NPRD system data, it would assume responsibility for management and funding of the NPRD system. INP0 reports that significant improvements in licensee participation are being made. The Commission considers the NPRD system to be a vital adjunct to the LER system for the collection, review, and feedback of operational experience; therefore, the Commission periodically monitors the progress made on improving the NPRD system.

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

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

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

AGREEMENT STATES Section 274 of the Atomic Energy Act, as amended, authorizes the Commission to enter into agreements with States whereby the Commission relinquishes and the States assume regulatory authority over byproduct, source and special nuclear ix

materials (in quantities not capable of sustaining a chain react' ion). Compar-able and compatible programs are the basis for agreements.

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

In early 1977, the Commission determined that abnormal occurrences happening at facilities of Agreement State licensees should be included in the quarterly 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.

\

x

REPORT.T0 CONGRESS ON ABNORMAL OCCURRENCES JULY - SEPTEMBER 1985 NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to operate during the third calendar quarter of 1985. As of the date of this report, the NRC had determined that the following events were abnormal occurrences.

85-12 Management Control Deficiencies The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see Example 11 of "For All Licensees") of this report notes that serious deficiencies in management or procedural controls in major areas can be considered an abnormal occurrence.

Date and Place - Because of continuing problems with the operation of LaSalle Nuclear Power Station, the NRC Region III Office initiated a special Task Force in July 1985 to perform an in-depth review of the facility's operations. Among the problems which triggered the task force study were three instances where errors in installation during equipment modifications affected the operability of Emergency Core Coolant System (ECCS) and the' shutdown cooling systems. The Task Force identified a number of items indicative of poor management ~ performance by the licensee (Commonwealth Edison Company). The LaSalle Nuclear Power Station, a two unit facility utilizing boiling water reactors designed by General Electric, is located in LaSalle County, Illinois.

Nature and Probable Consequences - LaSalle Units 1 and 2 received full power licenses in June 1982 and March 1984, respectively. Subsequently, the licensee has experienced numerous personnel, equipment, and regulatory problems, many of which can be attributed to deficiencies in management controls. These recurring problems have not individually been of major safety significance, but represent

, a trend which is not acceptable over the long run at an operating nuclear station.

To fully assess the scope and nature of the problems, the NRC Region III Office initiated a special Task Force review of the performance of the LaSalle Station in July 1985. The task group identified a number of items indicative of poor management performance. Among the principal findings of the task force, and other NRC inspections:

Three instances in which installation errors occurred during equipment modifications at the facility in 1985. One event led to the inoperability of Unit 2 Emergency Core Cooling System (although the plant was shut down at the time) for several days. In addition, for part of this time secondary containment was not maintained as required by Technical Specifications. The other two events affected the operability of the shutdown cooling systems of Units 1 and 2, re-spectively. These events are discussed further below.

1

A total of 172 violations of NRC requirements have been identified by NRC inspections from 1982 through July 1985. Three fines have been assessed, and a fourth one has been proposed (for the modification errors mentioned above, and described further below).

Twenty-four instances between October 1984 and July 1985 where personnel errors or other actions during maintenance or modification work affected the operations of the plant. Included were five instances where reactor scrams (automatic shutdowns) were triggered by persons performing maintenance or modi-fication work.

Repeated equipment problems--caused either by hardware failures or personnel errors--have occurred in a number of systems. Among the problems have been 25 failures of the control room ventilation system toxic gas-detectors, 56 Licensee Event Reports (submitted to the NRC by the licensee) or Deviation Reports (inter-nal licensee reports) on fire protection system problems, and 10 failures of the vent stack wide range gas monitors.

The plant appears to routinely operate with several Limiting Condition for Operation (LCO) " time clocks" running. -The plant's Technical Specifications contain numerous instances where the plant must reduce power or shut down within a specified time period if certain conditions or equipment problems exist.

During a two week period reviewed by the Task Force, the number of time clocks for these LCOs averaged from three to six per unit at any one time.

The number of outstanding work requests for repairs or maintenance of con-trol room equipment remains high--running about 80 per unit in September 1985.

These outstanding work requests may not be significant on an individual basis, but they tend to decrease the reactor operators' confidence in control room instruments and indicators.

There is an excessive backlog of equipment modifications. In September 1985 the number totaled 543, not including those in progress and 270 of these have been designated as priority modifications. The priority modifications include 85 resulting from commitments to the NRC. The licensee's ability to complete these modifications in a timely manner is in doubt--only 74 mofifications were completed between January and August 1985.

As mentioned previously, among the problems which resulted in the task force study were three instances where equipment modification errors affected the operability of the ECCS and the shutdown cooling systems. These events are described as follows.

1. Unit 2 was. shut down in February 1985 for an outage that included installa-tion of environmentally. qualified electrical equipment. The Unit has three divisions of ECCS equipment. Division III of the ECCS was removed from service in March 1985 for normal maintenance. Between April and June 1985, due to'in-adequate controls in the design, inspection, and testing areas, the piping to

~

two reactor vessel water level actuation switches in Division I of the Unit 2 Emergency Core Cooling System (ECCS) was installed backwards and, as a result, the Division I ECCS pumps would not have initiated as required on a low-low-low reactor vessel water level trip signal. On June 5, 1985, while unaware that Division I was inoperable, the licensee removed Division II of the ECCS from 2

service for modification. Therefore, all three ECCS divisions were inoperable and automatic initiation capability of the ECCS in response to a low-low-low reactor vessel ~ water level signal was lost until the problem was discovered and corrected on June 10, 1985.

In addition, again because the licensee was unaware that Division I of the ECCS was inoperable, the secondary containment was declared inoperable from June 5 to June 8, 1985, due to maintenance on the reactor building ventilation system. Even though the reactor was in cold shutdown, failing to maintain containment integ-rity when all ECCS capability is lost is a violation of Technical Specifications.

The event was caused by a lack of adequate design documentation, inspection, and testing controls. The NRC considered this violation particularly significant since its causes were almost identical to a violation which was discovered in Unit 1 in April 1985, and for which the licensee was cited for inadequacies in design and test controls. On April 17, 1985, while performing monthly functional tests on Unit 1, the licensee found that two switches for the Unit 1 Automatic Depressurization System (ADS) were miswired, making the trip system "B" for ADS initiation inoperable.

Following the discovery on June 10, 1985, of the loss of automatic actuation of ECCS. capability, NRC Region III sent a Confirmatory Action Letter to the licensee on June 17, 1985, documenting the steps to be taken by the licensee both prior and after startup of Unit 2 (Ref. 1).

2. On July 17, 1985, the licensee discovered that the piping to the Unit 1 Residual Heat Removal (RHR) shutdown cooling pump high suction flow alarm and isolation switches was installed backwards. A verification walkdown failed to identify this improper installation. This installation resulted in these switches being inoperable during power operation, and a Technical Specification Limiting Condition for Operation was exceeded. Although there are several re-dundant signals that may provide this same system isolation function, this vio-lation demonstrates another example of the lack of adequate design document and testing controls in the licensee's program.

The NRC Region III forwarded a Confirmatory Action Letter to the licensee on July 22, 1985, documenting additional actions to be taken by the licensee prior to startup of either Unit 1 or Unit 2 (Ref. 2).

3. During this same period of time, another instance was discovered in which the piping to the two Unit 2 RHR Shutdown Cooling pump suction high flow isola-tion switches was installed backwards. The licensee failed to recognize this im-proper installation during a verification walkdown, but after a review of data associited with an alternate test, identified the problem with the installation of the lines to the switches. Although the Technical Specification does not re-quire these switches to be operable in cold shutdown, this violation demonstrates further design and testing failures in the licensee's modification program.

Cause or Causes - The deficiencies appear to have resulted from a failure to aggressively resolve equipment problems, inadequate planning and control of site activities, and an excessive number of personnel errors--all of which are indicative of significant deficiencies in the licensee's site management struc-tures and systems to control site activities.

3

Actions Taken to Prevent Recurrence Licensee - The immediate actions required prior to startup (as described in Ref. 1 and Ref. 2) were completed and Unit 2 was authorized to start up on July 20, 1985. The licensee is performing the actions required after Unit 2 startup. In addition, the licensee has completed the actions required prior to Unit 1 startup (the licensee had shut down Unit 1 for maintenance on July 12, 1985; the plant was restarted on July 27, 1985.)

Over the past two years the licensee has undertaken a company-wide Regulatory Improvement Program to improve the performance of its management for its nuclear power plants. This program has included issuance of policy directives, organiza-tion modifications, some personnel changes, increased management involvement in the day-to-day operations at the nuclear facilities (by both corporate and sta-tion management), training activities, and efforts to reduce the number of per-sonnel errors and procedural violations.

The licensee also retained a consultant to review its station operations.

The licensee's Regulatory Improvement Plan has been less effective at the LaSalle Station than at the other Commonwealth Edison facilities. Some evidence of improved performance has been observed, but additional steps are needed to obtain effective, continued improvements.

Additional corrective actions may be required in response to the NRC escalated enforcement action, issued Septembar 27, 1985, des.ribed below.

NRC - Considerable effort has been required to monitor the licensee's performance and to review and document violations. As discussed previously, numerous viola-tions have been found and several fines have been assessed.

In regard to the equipment modification errors which occurred in June and July 1985, on September 27, 1985, the NRC issued a notice of violation and pro-posed imposition of civil penalties of $125,000 (Ref. 3). The base penalty for violations would normally be $50,000; however, the amount was increased to

$125,000 because of the number of incidents and because of previous poor per-formance of the licensee in similar areas.

The NRC letter noted that the violations demonstrated a need for the licensee to re-examine its commitments made to the NRC with regard to operability testing.

On October 30, 1984, the licensee failed to perform adequate tests on the Standby Gas Treatment System (SBGT) after maintenance work was performed. As a result, plant personnel were nnt aware that the SBGT was inoperable until the problem was brought to their attention by the NRC Resident Inspector. That event re-suited in a $25,000 civil penalty (Ref. 4). In its response, the licensee stated, "In order to preclude this type of problem in the future, LaSalle Station I will require that a test be conducted to demonstrate operability anytime a safety-related system is returned to service. A Post Maintenance Operational Test Checklist has been developed to ensure that the post maintenance test specified adequately demonstrates system operability in light of work performed."

The violations cited in the September 27, 1985 NRC letter indicate that more effective controls must be implemented to ensure that operability tests will 4

be performed on safety-related systems after maintenance or modification and before these systems are returned to service.

The results of the Task Force formed by NRC Region III in July 1985 have been discussed with the licensee's Chief Executive Officer.

On November 22, 1985, the Regional Administrator of Region III issued a letter to the licensee under 10 CFR S 50.54(f) requesting information on the licensee's plans to improve its performance in managing its maintenance, operation, and modification activities, including those problems identified in the Task Force report (Ref. 5).

The licensee replied to the request on December 23, 1985 (Ref. 6). NRC Re-gion III is currently evaluating the adequacy of the response. Further regula-tory action will be taken, as appropriate.

Future reports will be made as appropriate.

85-13 Inoperable Steam Generator Low Pressure Trip The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see the second general criterion) of this report notes that a major degradation of essential safety-related equip-ment can be considered an abnormal occurrence.

Date and Place - On August 7, 1985, Maine Yankee Atomic Power Company found 9 of the 12 pressure transmitters that monitor pressure of the three steam genera-tors (SGs) inoperable due to closed or partially closed root valves. These transmitters provide low steam pressure inputs to the Reactor Protection System, the Main Steam Isolation System, and the Feedwater I olatic., System. The closed root valves caused three of the four low-SG pressure logic ciannels of these systems to be inoperable. The significance of this is that in tN event of a steam line rupture the subsequent reactor trip, main steam isolation and main feedwater isolation would not have initiated automatically on low steam pressure signals. This condition had existed since June 20, 1984.

Maine Yankee utilizes a three loop pressurized water reactor designed by Combus-tion Engineering, Inc. and is located in Lincoln County, Maine.

Background

Low steam pressure provides several protection signals in the event of a main steam line break. The three main steam lines exit containment into the mechani-cal penetrations room where each line is provided with an isolation non-return valve and an excess flow check valve. Each of these steam lines has four instru-ment taps between the non-return valve and the containment wall. Each instrument tap contains a root valve, instrument isolation valve, a drain valve, and a pressure transmitter. The pressure transmitters provide low pressure signals to four independent measurement channels, designated as channels A, B, C and D.

Each channel provides the following:

5

f Pressure indication on the main control board (MCB) sigma meter for each SG. (Sigma meters are a brand name of a small horizontally mounted meter. All twelve sigma meters are located in the same area of the MCB so that all channels can be easily compared.)

An input to the Reactor Protection System (RPS). The low-SG pressure

trip setpoint is 485 psig. RPS logic is any 2 of 4 low-SG pressure

! signals on any two independent protection channels from one or more SGs.

}

Independent of RPS:

An input to close all excess flow check valves on low pressure (400 psig) from a single SG and provide a pre-trip alarm (535 psig). Signal is any 2/4 from a single SG.

A low-SG pressure signal to provide a feedwater isolation signal.

j Signal is 2/4 from a single SG.

A low-SG pressure signal coincident with a safety injection actu-ation signal to provide a main feedwater pump train trip. Signal is 2/4 from a single SG.

The purposes of the above trips in the event of a main steam line break are:

to scram the reactor, isolate the SGs, and stop feedwater to the SGs preventing continued steam release, reactor overcooling, and a possible reactor restart.

f In addition to the trip signals, Channel A provides an input to the plant compu-ter, indication on the safe shutdown panel, and a second meter on the MCB. ,

I Nature and Probable Consequences - On August 7, 1985, the control room operators l noted that the SG pressure indication for SG #1 Channel D was reading approxi- l mately 520 psig vice actual SG pressure of 630 psig. Calibration of the pressure transmitter by Instrumentation and Control (I&C) personnel was satisfactory and the technicians suspected that the instrument line may have been blocked. Using the drain valve, the technicians blew down the line and found that after an l initial volume of water was drained, the steam pressure in the line was minimal.

An operator was sent to the mechanical penetration room to check the position of the instrument root valve. The operator, suspecting that the valve (MS-46) for SG #1 Channel B was open, attempted to verify it as open by turning the valve in the open direction and discovered the valve was, in fact, not open.

The next valve check was MS-47 which was also checked in the open direction and was found to be similar to MS-46.

All root valves (four for each of the three SGs for-a total of 12) were checked.

'Three valves were found open (MS-45, 65 and 85). These three valves are the Channel A pressure detector root valves. The other nine root valves were found shut or nearly shut. The operator reported moving a number of valves approxi-mately 1/16 to 1/32 of a turn in the closed direction. Full stem travel is approximately 5 and 1/2 turns. Six valves (MS-44, 64, 66, 84, 86 and 87) were found cracked open and three valves (MS-46, 47, and 67) were found shut.

6

The low-SG pressure trip is designed to operate even when one of the four chan-nels is out of service, with a trip occurring if two of the remaining channels sensed low pressure. However, the closed root valves affected three channels (Channels B, C, and D) and thus the low-SG pressure trip would not have operated if a main steam line break had occurred.

Three possible consequences of the closed root valves were of primary concern:

(1) the loss of a trip signal to the Reactor Protection System which controls the insertion of reactor control rods, (2) delayed isolation of the intact steam generators from the ruptured pipe following a main steam line break accident, and, (3) the prevention of automatic main feedwater pump trip. These consequences are discussed as follows.

In the unlikely event of a large main steam line rupture, the reactor trip on low-SG pressure would be the first trip received by the RPS resulting in a reac-tor scram within several seconds. The closed root valves would have prevented this trip. However a number of other trips were available to scram the reactor, e.g., delta T power, high nuclear flux, or low primary system pressure. Thus the closed root valves would have caused a delay in reactor scram.

Of greater concern was the possibility of extended steam blowdown outside con-tainment and reactor overcooling. The closed root valves would delay main steam and feedwater isolation and prevent automatic trip of the main feedwater pumps.

Thus, if a main steam line rupture had occurred, the steam blowdown and flow of main feedwater to the affected SGs would have continued beyond the termination point currently assumed in plant safety analyses. This would result in overcool-ing of the reactor vessel and possible reactor recriticality. In addition, for ruptures outside containment, there would be adverse environmental conditions for equipment in the t tine building.

Cause or Causes - Inadequate administrative controls resulted in the SG pressure instrument root valves being left in the closed position. A hydrostatic test required that the root valves be closed by operations personnel and the instru-ment isolation valves be closed by I&C personnel. The test procedure did require I&C personnel to reopen the instrument isolation valves but did not specifically call for the root valves to be opened by operations personnel. The licensee does not manipulate root valves on normal system alignments. The three root valves for Channel A were opened following completion of a plant modification involving installation of the subcooled margin monitor (SMM). The two activities (hydrostatic test and plant modification) were worked concurrently and both required the "A" root valves to be shut. However, maintenance controls for the SMM modification properly directed opening of the "A" root valves following com-pletion of the modification. Consequently, 9 of the 12 root valves remained in a closed or nearly closed position for a fifteen month period.

The licensee found on September 3, 1985 that the installation of the SMM modifi-cation also adversely affected the low-SG pressure reactor trip. Errors in design and post installation testing of the SMM modification resulted in the Channel A portion of low-SG pressure reactor trip being disabled. Therefore, even though the Channel A root valves were open, a proper trip signal would not have been received by the RPS. The feedwater trip for Channel A was unaffected.

It should be noted that since the Channel A root valves were open, and the root valves for Channels B, C and D leaked or were partially open, the sigma meters 7

in the control room accurately displayed steam-generator pressure. However, due to the restricted path for sensing steam pressure, the RPS would not have responded in accordance with design and tripped t.he reactor on low steam pressure.

Actions Taken to Prevent Recurrence Licensee - The licensee has developed a program to correct and prevent recurrence of the mispositioning of instrumentation root valves as well as inadequate design change review as delineated below:

Root Valves

1. Verify as open all root valves associated with safety related instrumenta-tion identified in the Technical Specifications.

2. Incorporate all root valves identified in (1) above into appropriate operat-

.ing procedures to ensure that they are open and verified to be open prior to startup from each refueling outage.

3. Verify that appropriate administrative controls (procedures) exist on all isolation valves associated with instrumentation identified in (1) above.

4. Review all special tests and temporary procedures prior to use to ensure that each valve position is individually specified when realigning systems.

5. Review the generic procedures governing the preparation and review of pro-cedures to ensure valve positions are individually specified when realigning systems.

Design Changes / Post Maintenance Testing

1. Redesign the SMM circuitry to eliminate the common connection.

2. Review all previous design changes which involve or could interact with the safety instrumentation system identified in the Technical Specifications.

3. Provide an independent design review of all the design changes identified in (2) above.

4. Develop functional test requirements that are more comprehensive for all systems identified in (2) above that have been significantly modified since the issuance of the facility operating license (1972), including those undergoing modification during the current outage.

5. Provide documented assistance that the functional test requirements identi-fied in (4) above have been previously satisfied, and no further modifica-tions have been performed, or perform new comprehensive functional tests to satisfy the applicable requirements.

6. Review the engineering design change request (EDCR) procedures to specifi-cally identify instrumentation lead "commoning" as requiring special design review emphasis, l

j

7. Review the quality assurance procedures to require that a comprehensive functional test be performed on modified circuits including any associated circuits that may be affected.

8. Review the EDCR procedures to require a second independent design review of all EDCRs associated with instrumentation identified'in (2) above, with particular emphasis to detecting sneak or interactive circuits.

9. Review the procedures governing design change implementation instructions to provide an independent review to ensure that the functional test require-ments are appropriately comprehensive.

NRC - An enforcement conference was held with the licensee in the Region I Office on September 9,1985. At the conference, the licensee was asked to discuss root causes of the occurrences from the aspect of both the root valve closure and the breakdowns in the design review process and post maintenance testing.

At the enforcement conference, the licensee offered a comprehensive analysis of all aspects of the occurrences and outlined a comprehensive program for cor-rective actions.

The licensee formally transmitted their corrective action program in a letter to the Region I Office on September 13, 1985 and addressed the points outlined above. A special inspection was initiated following the enforcement conference to assess the adequacy of implementation of the proposed licensee corrective actions. The inspection concluded on October 21, 1985, the day before recovery from the then ongoing refueling outage. The inspection found the licensee's corrective program comprehensive, well conceived, and properly implemented.

There were no issues found by the inspection that impacted on timely plant startup.

On October 29, 1985, the NRC Region I Office issued _a Severity Level II violation and civil penalty in the amount of $80,000 (Ref. 7). The Regional Administrator emphasized that corrections were needed in the areas of improved administrative control of valves, control of design changes, preparation and implementation of temporary procedures, and control of the post maintenance or post modification testing process including test design, test procedures, and their review.

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

85-14 Management Deficiencies at Tennessee Valley Authority The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see Example 11 of "For All Licensees") of this report notes that serious deficiencies in management or pro-cedural controls in major areas can be considered an abnormal occurrence.

Date and Place - Because of serious NRC concern regarding significant program-matic and management deficiencies at Tennessee Valley Authority (TVA), on September 17, 1985, the NRC issued a request for information pursuant to 10 CFR S 50.54(f) to enable the NRC to determine whether or not the licenses 9

for the Browns Ferry and Sequoyah facilities should be modified or suspended or the application for the Watts Bar facility should be denied (Ref. 8). The licensee's facilities are described as follows:

Number Reactor Facility of Units Designer Reactor Type Facility Location Browns Ferry 3 General Electric Boiling Water Limestone County, AL Sequoyah 2 Westinghouse Pressurized Water Hamilton County, TN Watts Bar 2 Westinghouse Pressurized Water Rhea County, TN Bellefonte 2 Babcock & Wilcox Pressurized Water Jackson County, AL Operations at all three Browns Ferry units have been suspended by the licensee since March 1985. Operations at both Sequoyah units have been suspended by the r licensee since August 1985. The two units at Watts Bar are under construction and fuel loading for Unit I had been projected for January 1986; however, the licensee issued a stop-work order on all safety related welding activities for both units during August 1985. This order was lifted in September 1985.

e Nature and Probable Consequences - The September 17, 1985, NRC letter forwarded to the licensee the latest Systematic Assessment of Licensee Performance (SALP)

Reports. This assessment was prepared by the described above, as well as for the TVA headoua)taff for each of the facilities rters' functions.

The SALP program is an integrated NRC staff effort to collect available obser-vations and data on a periodic basis and to evaluate licensee performance based upon this information. SALP is supplemental to normal regulatory processes used to ensure compliance to NRC rules and regulations. SALP is intended to be sufficiently diagnostic to provide a rational basis for allocating NRC resources and to provide meaningful guidance to 4he licensee's management to promote qual-ity and safety of plant construction and operation.

The collection of performance observations and data are then reviewed by a SALP Board, composed of NRC senior personnel, to assess licensee performance. Licensee performance is assessed in gelected functiona} areas, depending upon whether the facility is in a constructi30, preoperational, or operating phase. Each func-tional area normally represents areas which are significant to nuclear safety and the environment, and whkh are normal programmatic areas. Some functional areas may not be assessed because of little or no licensee activities or lack of meaningful observations. Special areas may be added to highlight significant observations.

Each functional area is classified into one of three performance categories.

Briefly, these are (1) Category 1: Reduced NRC attention may be appropriate, (2) Category 2: NRC attention should be maintained at normal levels; and (3) Category 3: Both NRC and licensee attention should be increased. The SALP Board also categorizes the performance trend over the course of the SALP assess-ment period, i.e. , the performance isJmproving, remaining constant, or declining.

The SALP reports for TVA covered the pe'riod from March 1,1984 through May 31, 1985 (Watts Bar Unit 1 was for January 1, 1985 through May 31, 1985).

Based on an overall assessment of the licensee's perfot.7ance, the NRC has con-cluded that TVA has demonstrated inef fective management of its nuclear program.

This poor performance is indicated by:

10

Four successive SALP periods with Category 3 performance in Plant Operations for the Browns Ferry facility, Three successive SALP periods with Category 3 performance in Quality Assurance and Administrative Controls Affecting Quality for both the Browns Ferry and Sequoyah facilities, Three successive SALP periods with Category 3 performance in Mainte-nance and Security and Safeguards for the Browns Ferry facility, Multiple escalated enforcement actions, including a Confirmatery Order regarding the Browns Ferry Regulatory Performance Improvement Program, an Order regarding identification, evaluation and reporting of signifi-cant issues, frequent enforcement conferences, and several significant civil penalties since March 7, 1984, Numerous significant events since March 1,1984, at TVA facilities.

Several of these events are described below.

Browns Ferry Facility

1. On August 14, 1984, overpressurization of the Unit 1 core spray system occurred while conducting a core spray system logic surveillance test with the reactor operating at 100 percent power. The air control solenoid for the actu-ator of the core spray system inboard isolation valve was incorrectly rebuilt during maintenance sometime prior to the beginning of the current fuel cycle on December 29, 1983. The isolation valve is a check valve with an air actuator which is used to move the flapper for test purposes. This, in conjunction with operator error and procedural deficiencies caused the opening of the inboard injection valve during the performance of the core spray logic surveillance test. Backflow of reactor coolant at reactor system pressure into the low pressure core spray system resulted. The low pressure section of the core spray system was overpressurized and portions of the core spray system piping were heated to approximately 400 F.

The high pressure / low pressure isolation arrangements provided between the high pressure reactor coolant system and the low pressure core spray system were substantially degraded, reducing primary system containment integrity and pro-viding the potential for structural damage to the core spray system. In addition, as a result of this event, thirteen persons received minor radioactive skin contamination.

The event was caused by a combination of personnel errors, lack of control over maintenance activities, inadequate post-maintenance testing, and procedural deficiencies.

This event, together with other events involving degraded isolation valves in emergency core cooling systems, were reported as abnormal occurrence A0 84-8 in NUREG-0090, Vol. 7, No. 3 (" Report to Congress on Abnormal Occurrences:

July-September 1984"). As described in the report, on January 28, 1985, the licensee we' issued a civil penalty in the amount of $100,000 (Ref. 9).

i

2. While conducting a shutdown margin test on Unit 3 on October 22, 1984 (after an extended shutdown for refueling, plant modifications, and inspections),

11

numerous procedural and equipment deficiencies necessitated a re-evaluation, which extended the outage for another month.

The Technical Specifications require that the correct rod withdrawal sequence be verified prior to reactor startup. However, an incorrect rod withdrawal sequence was programmed into the' Rod Worth Minimizer computer program and due to inadequate verification of the program, the errors in control rod programming were not discovered until 31 rods had been fully withdrawn from the core. No rods were actually withdrawn in the wrong sequence, however.

The Technical Specifications require that jet pumps be demonstrated to be oper-able prior to startup. Two jet pump differential flow instruments were inoper-able due te valve alignment errors and were therefore unavailable for the demon-stration of jet pump operability. The reactor was taken to the startup mode and made critical in violation of Technical Specifications.

In addition to the above, various other procedural steps of several different procedures were not accomplished which were required to be completed prior to criticality. One of the examples involved the failure to perform steps in one procedure which were identified as critical steps that resulted in the low pressure coolant injection mode of the residual heat removal system not being fully operable and could have led to a violation of Technical Specification Limiting Condition for Operation. A subsequent procedure which required verifi-cation that the' critical steps were accomplished was also not completed. Addi-tional examples of procedural' violations as well as two examples of inadequate precedures were also identified.

The event was caused by a lack of discipline in the conduct of operations.

This was exemplified by personnel errors, inattentiveness to procedural details, inadequate control of safety equipment during return-to service following main-tenance, and a fundamental misunderstanding of the definition of " reactor startup".

On February 27, 1985, the NRC issued a Notice of Violation and Proposed Imposition of Civil Penalties in the amount of $112,500 for violations associated with the improper startup (Ref. 10).

3. During a reactor startup on Unit 3 on February 13, 1985, the' licensee failed to satisfy Technical Specification requirements for reactor vessel water level instrumentation operability. Specifically, due to normal error inherent to low reactor pressures and temperatures, two Yarway water level instruments indicated a reactor vessel w~ater level about 39" (25" greater than the actual vessel water level). Two of the three GEMAC water level instruments were indicating a reactor vessel water level of 37", and the third GEMAC water level instrument was indi-cating 10". The operators believed the GEMAC instrument indicatin'g 10" was erroneous because the other two were approximately in numerical agreement with the Yarways. (The 10", 37", and 39" levels are control point levels; even at 10", there was over 17 feet of water above the top of the core.)

Even when a half scram occurred as a result of low reactor water level, the operators failed to determine which instruments were providing correct reactor water level indication. However, the heatup was discontinued until the instru-ments began to converge. The NRC believes reactor operation should have been suspended until the cause of the problem was determined. Such action is neces-sary because the errors in the GEMAC instruments were caused by a malfunctioning 12

reference leg which was common to the Barton water level instruments and which degraded two channels of the one-of-two-taken-twice logic associated with the reactor water level scram in the Reactor Protection System. Instead, operators reset the half scram by raising the reactor vessel water level in manual control, and continued the heatup of the system.

This event was considered serious by the NRC because operators had sufficient information to indicate that important instrumentation was inoperable and, in-stead of identifying and fixing the cause of the problem, they continued with the reactor startup. In addition, the Plant Superintendent for Operations and Operations Supervisor became aware of this event approximately at midnight on February 13, 1985; however, the fact that the water level instruments were in-operable was not recognized and they did not direct discontinuation of startup.

The event resulted from a failure to take corrective action when a similar reac-tor vessel water level instrument problem occurred on November 20, 1984. If effective corrective actions had been taken at that time, the event in February could have been prevented. Contributing causes were a lack of knowledge by the operators due to a deficient training program and communications and coordination problems between operators, maintenance, and management.

On July 22, 1985, the NRC issued a Notice of Violation and Proposed Imposition of Civil Penalties in the amount of $150,000 (Ref. 11).

4. During a routine inspection on August 16, 1985, it was determined that major discrepancies existed in the design of cable tray supports at all three Browns Ferry units. Cable tray supports in the control bay area were not designec' to accommodate seismic loading. Cable tray supports in the Diesel Generator Buildings were improperly designed in that the seis.nic loads used in the design calculations were obtained from the Reactor Building seismic analysis instead of the Diesel Building seismic analysis. In addition, cable tray support calcu-lations in the Reactor Duilding showed a lack of thoroughness, clarity, consis-tency, and accuracy. As a result, many supports may not be able to serve their intended function during a seismic event.

During this same inspection, the licensee's implementation of corrective action to address a known deficiency related to cable trays was found to be inadequate.

In February, 1981, the licensee became aware of many overloaded cable trays in the cable spreading room of the control bay and a corrective action report was initiated. The root cause determination and corrective action associated with the report was delinquent and ineffective until July, 1985. The actions taken between the time period of February, 1981 to July, 1985 consisted of forwarding the information to various design groups within TVA, and attempts at preventing additional cable trays from becoming overloaded. The cable trays which were known to be overloaded were not evaluated until July,1985, when the licensee determined that the cable trays could not be cons.dered qualified for a seismic event.

The inadequate design of safety related cable tray supports was caused by in-adequate design controls during the construction phase of the plant. This con-dition was aggravated during subsequent modifications which resulted in cable trays being overloaded beyond their original design. A lack of aggressive action to correct the deficiencies once they were identified in 1981 further exemplified 13

the inability of management to coordinate resources available in the design, modification, and quality assurance organizations in a timely manner.

5. On September 24, 1985, . the licensee declared all eight emergency diesel generators associated with the standby a.c. power supply and distribution system inoperable. The diesels were considered inoperable for two reasons. Although some of the diesels have been in service for about thirteen years, the manufac-turer's recommended three, six and twelve year inspections and maintenance ac-tivities had not been performed. (The recommended annual maintenance had been performed.) Simultaneously, the diesel battery racks that support the batteries which are required for startup and operation of the diesels were found to be not qualified for the loads result:ng from a postulated seismic event. As a result of the diesel generators being inoperable, the licensee was unable to satisfy three Technical Specification requirements regarding diesel generator operability and emergency core cooling system operability. The resulting un-analyzed condition prompted compensatory measures and several safety evaluations.

This occurrence is considered significant by the NRC in that once again, the degraded condition of the plant could have been prevented had prop'er corrective action been initiated following previous identification of the problem. The failure to perform the manufacturer's recommended maintenance on the diesel generators was identified by the NRC resident s taff and cited as a violation of Technical Spuifications on July 16, 1984. The licensee reported on August 16, 1984, that full compliance with the requirement would be met on October 5, 1984; however, full compliance was never achieved.

Sequoyah Facility

1. On April 19, 1984, a significant event at Unit 1 occurred involving damage to a compression fitting at the incore probe seal table. Unit I was at 30%

power, with maintenance in progress for cleaning of the interior of the D-12 thimble tube (stainless steel tubing about 0.3 inch 0.D.). The cleaning assembly for drybrushing of the thimble tube was inserted about 80 feet into tube D-12 when the high pressure seal fitting, which forms the reactor coolant system (RCS) pressure boundary, failed. At the first indication of leakage, the eight workers in the incore instrument room immediately left the room through the containment airlock without injury or significant exposures. Shortly after the workers left the area, RCS pressure caused ejection of thimble tube 0-12. The RCS pressure caused a 25-35 gallons per minute (gpm) average unisolable reactor coolant leak. After extensive preplanning and mockup training, plant personnel recovered the highly radioactive thimble tube over the period of April 25-28, 1984.

The cause of the event is attributed to failure of the licensee to control mod-ifications made to the cleaning fixture used to support the drybrushing apparatus.

The tool had been repeatedly modified since 1979 by plant personnel without per-forming technical evaluations or tests to determine the effects of the modifica-tions of the tool on the thimble tube seal reactor coolant pressure boundary.

On May 7, 1985, the NRC issued a Notice of Violation and Proposed Imposition of Civil Penalties in the amount of $112,500 (Ref. 12). This event was reported in Appendix C of NUREG-0090, Vol. 7, No. 3 (" Report to Congress on Abnormal Occurrences: July-September 1984").

14

2. During early 1985, the NRC expressed concern to TVA regarding environmental qualification of electrical equipment issues. TVA hired an independent contrac-tor to review the documentation to determine compliance with 10 CFR 6 50.59 (Ref. 13). The contractor found that the documentation appeared to be inadequate at all TVA sites. As a result of this, as well as other certain technical con-cerns, the licensee shut down the operating Sequoyah Units 1 and 2 on August 21, 1984; as discussed previously, all three Browns Ferry units were already shut down.

Watts Bar and Bellefonte Facilities Construction at Watts Bar Unit 1 is essentially complete and the licensee had projected a fuel load date of March 1985; Unit 2 is about 75% complete. Unit 1 fuel load has been delayed pending resolution of various concerns raised by the NRC staff and TVA employees. Bellefonte Units 1 and 2 are about 86% and 56%

complete, respectively.

1. On July 18 and 19, 1985, the licensee issued stop-work orders on installa-tion of Class 1E electrical cable at the Bellefonte and Watts Bar sites, respec-tively. The orders followed a review by TVA of its general construction specifi-cations which sets electrical cable installation requirements for all TVA facili-ties. TVA advised NRC Region II that this review found an inadequacy in cable pull tension requirements with respect to industry practices.

2. On August 23, 1985, TVA issued a stop-work order on all safety-related welding activities at the Watts Bar site as the result of preliminary NRC Region II inspection findings which raised questions on the adequacy and accu-racy of welder recertification. On August 23, 1985, NRC Region II issued a confirmation-of-action letter which provides that TVA will throughly review its welder recertification pr -am, determine if appropriate code welding activities have been conducted by properly certified welders, and determine the safety significance of any welding activities conducted by uncertified welders (Ref. 14).

TVA agreed not to resume safety-related welding activities at the site without NRC concurrence. Welding activities were resumed in September 1985 after NRC Region II reviewed the certification program.

3. On August 29, 1985, the NRC issued a Notice of Violation and Proposed Impo-sition of Civil Penalty in the amount of $100,000 to TVA for violations involving control room design modifications at Watts Bar Unit 1 (Ref. 15). The violations concerned inaccurate status reports submitted from November 8, 1983 to October 3, 1984. The licensee stated that certain items were complete when in fact they were not, despite notice from the NRC inspectors that the reports were not accu-rate. The violations resulted from carelessness and inattention to detail in assuring the accuracy of information submitted to the NRC and was indicative of a breakdown in management controls.

Cause or Causes - Most of the problems encountered by the licensee were caused primarily by breakdowns in management and procedural controls, with personnel errors also a contributing factor. Deficiencies have been noted in procedures, training of personnel, maintenance, required documentation, accuracy of submit-tals to the NRC, and in taking timely, effective action on proolems which are identified. The NRC believes that there is a lack of effective management both at the Corporate and site levels.

15

Actions Taken to Prevent Recurrence Licensee - The licensee, in addition to taking corrective actions in regard to specific problem areas, has made attempts to improve management at their plants.

For example, the licensee initiated a Regulatory Performance Improvement Program at the Browns Ferry Facility in mid-1984 to improve performance. However, this has not apparently been effective as evidenced by the NRC SALP report as enclosed in the previously referenced NRC September 17, 1985 letter (Ref. 8).

TVA is evaluating the apparent ineffectiveness of the Program and has acknowl-edged the management shortcomings, attributing them to the past organization which has lad to a lack of responsibility, accountability and productivity. A reorganization was accomplished with key personnel changes. At Browns Ferry, the plant manager, assistant plant manager for maintenance and the operations supervisor have been replaced. A corporate entity was established with a single chain of command responsible for all nuclear activities. This eliminated the dual organizational structure which has, in the past, separated the engineering /

construction activities from the operating activities. Prior to the. restart of Browns Ferry, the licensee plans an in-depth operational readiness review to verify the integrity of personnel, procedures and equipment. An industry peer review to be conducted by personnel from other utilities and the Institute of Nuclear Power Operations is also planned.

TVA has recently undertaken reorganization at its other sites to effect more timely resolution of potential safety issues. This action is essentially a decentralization of plant-specific engineering staff deemed necessary to support the operating staff.

TVA has initiated some policy changes to help correct weaknesses in nuclear and operating experience of some of the lower levels of management, as well as reac-tor operators. During the past several years, a number of key managers with ex-tensive nuclear and operating experience left TVA. A number of licensed reactor operators and senior reactor operators have also left to work at other utilities.

In addition to the above, TVA is required to address the general and specific concerns noted in the NRC September 17, 1985 letter.

As previously noted, TVA has shut down all of their operating plants (Browns Ferry and Sequoyah facilities) until plant specific problems and general concerns are resolved to the satisfaction of the NRC.

NRC - In order to assure high level attention to the problems at TVA, an NRC Serior Management Team (consisting of the Executive Director for Operations, the Directors of the Offices of Nuclear Reactor Regulation, Inspection and En-forcement, and Investigations, the Regional Administrator of Region II, and senior managers from these four Offices) was formed and meets regularly to dis-cuss and implement corrective actions. These corrective actions consist of an augmented inspection program for Browns Ferry; Commission briefings; a re-evaluation of the Browns Ferry Regulatory Performance Improvement Plan; augmented Systematic Appraisal of Licensee Performance of all TVA facilities; and an in-depth operational readiness inspection program.

On July 3, 1985 (Ref. 16), and August 1, 1985 (Ref. 17), the Executive Director for Operations (EDO) forwarded to TVA concerns with TVA performance. On 16

i September 10, 1985, the NRC Team met to review the latest SALP reports for TVA's four sites and for TVA headquarters functions. Based on continuing concerns with TVA performance, the EDO forwarded the previously referenced 10 CFR S 50.54(f) letter (Ref. 8).

The letter _ addressed a number of concerns, including corporate oversight, qual-ifications of new personnel, commitment control, timely resolution of conditions adverse to quality, adequacy of the operational readiness plan, maintenance improvement program, plant modification control, evaluation of seismic design concerns, environmental qualification of electrical equipment important to safety, onsite independent safety engineering group, fire protection program, and out-standing licensing issues.

The letter requested information'in accordance with "1e following schedules:

1. Information specific to Sequoyah 60 days prior to restart of either' Sequoyah unit.

2. Information specific to Browns Ferry 90 days prior to restart of any Browns Ferry unit.

3. Information specific to Watts Bar 90 days before the licensee antici-pates requesting a fuel-load license for Watts Bar Unit 1.

4. Information specific to TVA corporate 60 days prior to startup of any of the TVA Units or a request for licensing of Watts Bar Unit 1.

On November 1, 1985, TVA submitted the requested information specific to Sequoyah and TVA corporate changes. These submittals are under review. The information pertaining to the Browns Ferr" and Watts Bar facilities is expected in early 1986.

Future reports will be made as a yr ite.

FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)

The NRC is reviewing events reported by these licensees during the third 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.)

Tht:re are currently more than 8,000 NRC nuclear material licenses in effect in the United States, principally for use of radioisotopes in the medical, indus-trial and academic fields. Incidents were reported in this category from li-censees such as radiographers, medical institutions, and byproduct material users.

17

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

During the third calendar quarter of 1985, an overview of 1984 misadministration events disclosed that the first two events below (which occurred during 1984)

.should have been reportable as abnormal occurrences.

85-15 Therapeutic Medical Misadministration The following information pertaining to this event is also being reported concur-rently 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 - From October 17, 1984 to November 1, 1984, a patient treated on the cobalt-60 teletherapy unit at the University Health Center of Pittsburgh's Joint Radiation Oncology Center, Magee - Women's Hospital site, received a radiotherapy administration of 3584 rads rather than the prescribed 2000 rads.

Nature and Probable Consequences - The patient was receiving the second of two courses of therapy to the ninth and tenth ribs using cobalt-60 external beam therapy. This region was being treated palliatively to relieve pain from meta-static disease. There was, in addition, a primary site under treatment with an entirely separate treatment plan which included the lung and mediastinum. Both treatment plans involved supraclavicular regions. The course of treatment to the primary site proceeded normally to its conclusion.

The first course to the metastatic area, 2000 rads in five treatments, prescribed on September 13, 1984, had been completed without any problems. The prescription for the second course of treatment, prescribed on October 16, 1984, was 2000 rads to be delivered in ten treatments. However, when the treatment dose for the second course was calculated, the dosimetrist assumed that the prescription was the same as the earlier one.

Rather than checking the prescription and preparing a new calculation as required by the Joint Radiation. Oncology Center (JROC) procedures, the dosimetrist relied on a verbal communication and only decay-corrected the output from the first treatment. As a result, the patient began to receive treatment fractions that were twice those of the prescribed dose. This error was not discovered until the patient had received 3584 rads in the second course of therapy, when one of the treatment technologists noticed that the delivered dose differed from the prescribed dose by greater than 10%. Further treatment was stopped at that time.

The consequence of this incident was that the patient received an unprescribed

' dose to the ribs of 1584 rads. The licensee reported that although the patient received more radiation than was prescribed, the patient has not, and likely will not suffer any ill-effects other than a modestly aggravated soft tissue reaction. The licensee reported that the actual dosage received is within a clinically acceptable range for the desired effect.

Cause or Causes - The cause was failure to comply with established procedures and oversights by the responsible staff, as evidenced by the following:

18

1. A requisition slip, required by established procedure, was not issued to request the second course calculation by the dosimetrist. The dosimetrist proceeded on verbal instructions which were relayed rather than received directly.

2. The dosimetrist failed to read the prescription in the patient's chart.

3. The dosimetrist failed to initiate a new calculation sheet for the new course of treatment.

4. A second dosimetrist, requested to make a correction for a separation change after the third treatment of the course in question, made that correcti)a on the wrong sheet. Finding no calculation sheet for the second course, due to error 3, she made the corrections on the calculation sheet of the first course, overlooking the fact that the dates did not coincide.

5. The physician reviewing the chart after the fifth treatment of the course in question, failed to observe that the dose had already exceeded the pre-scription, and marked the chart " continue", referring to the primary treat-ment course of greater concern, going on concurrently.

6. The error was not noticed by the treatment technologists in their routine handling of the patient until the ninth treatment.

Actions Taken to Prevent Recurrence Licensee - The policies and procedures which should have prevented this occur-rence are clearly stated in a Physics Procedure Manual in the possession of each of the dosimetry staff. The policies were reviewed in inservice meetings held for the staff on July 2,1984 and July 16, 1984 which dosimetrists and technologists attended.

Enforcement of these policies will receive the continued vigilance of management and supervisory staf f. A general meeting of the entire physics and dosimetry staff was held on November 8, 1984 to discuss this issue and to re-emphasize the importance and necessity of strictly following the procedures. In addition, management and physics personnel have discussed the occurrence, the consequences, and the importance of following the standard practices with each of the persons involved. The dosimetrist who committed the most critical errors (numbers 2 and 3 above) has received a formal admonishment in writing which is incorporated as part of her personnel record. A memorandum has also been delivered to the physicists in charge at each site reminding them of their responsibility for assuring that procedures are followed correctly at their sites.

NRC - No violations of NRC regulations were associated with this incident. An NRC medical consultant is reviewing the case. Upon receipt of the consultant's report, an inspection will be scheduled.

This incident is' closed for the purposes of this report.

19

85-16 Therapeutic Medical Misadministration The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see the general criteria) of this report notes that an event involving a moderate or more severe impact on public health c' safety can be considered an abnormal occurrence.

Date and Place -_On October 25, 1984, NRC was notified that a patient of. the Milton S. Hershey Medical Center in Hershey, Pennsylvania, received 15 milli-curies of iodine-131 rather than the prescribed dose of 10 millicuries.

Nature and Probable Consequences - A 10 millicurie dose of iodine-131 had been ordered for one patient for treatment for hyperthyroidism and a 5 millicurie dose had been ordered for a second patient as a whole body scanning dose. When the first patient arrived, the technologist opened the bottle containing capsules which had the patient's name on it, dumped them in the patient's hand and gave the patient water with which to take them. The technologist neither verified the activity nor the number of capsules. Later in the day when the second patient arrived, the bottle with this patient's name on it was found to be empty.

Several days passed before the licensee was successful in contacting the first patient who remembered taking three capsules, rather than the appropriate two.

The capsules were provided in vials labelled with each patient's name by Nuclear Pharmacy, Inc. It appears that although Nuclear Pharmacy accurately verified the activity of each capsule before dispensing, they dispensed three capsules in one bottle and none in the second. The prescriptive information labels in-dicated two capsules of 5 millicuries each in the first bottle and one capsule of 5 millicories in the second' bottle.

The referring physicians determined that effects should be minimal and involve only an increased probability of the first patient ultimately developing hypo-thyroidism.

Cause or Causes - The cause was failure on the part of a nuclear medicine tech-nologist to verify the activity of the administered dose and failure on the

~

part of Nuclear Pharmacy to properly dispense two patient doses.

Actions Taken to Prevent Recurrence Licensee - The technologist involved, who was new and apparently not thoroughly familiar with procedures, was reprimanded for not following required procedures.

.All technologists were retrained regarding procedures and requirements for receipt of radioisotopes, survey of radioisotopes received, dose calibrator verification of all radiopharmaceutical activities, administration to patients, record keeping and notification of incidents. The written procedures were modi-fied and all technologists who are hired in the future will be required to demon- 1 strate thorough understanding of the procedures before being given approval tc l administer radioisotopes to patients.

Nuclear Pharmacy, Inc. - As a result of this occurrence and a recent enforcement action, Nuclear Pharmacy, Inc. has significantly improved management control of its dispensing operations. Significant improvement was also made to record-keeping and auditing procedures.

20

I NRC - The incident is being reviewed by an NRC medical consultant. The corrective actions taken by the licensee will be reviewed during a future routine inspection.

This incident is closed for the purposes of this report.

85-17 Exposure of Radiographic Personnel Due to Management and Procedural Control Deficiencies The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see Example 11 of "For All Licensees") of this report notes that serious deficiency in management or pro-cedural controls in major areas can be cunsidered an abnormal occurrence.

Date and Place - On August 5, 1985, Western Stress, with offices located in Evanston, Wyoming and Houston, Texas, notified the NRC Region IV Office that radiographic personnel had received whole body radiation exposures in excess of NRC regulatory limits. Subsequent NRC inspections showed that the root causes were due to serious management and procedural control deficiencies.

Nature and Probable Consequences - On August 1, 1985, a radiographer and his helper went to a site at Table Rock, Wyoming to perform radiography with a radio-graphic camera containing a 29 curie Ir-192 source. After radiographing welds, the helper began developing the film while the radiographer disassembled the radio-graphic equipment. After being informed that one of the films did not receive the proper amount of exposure, the radiographer went to examine it. As he did, the helper reconnected the drive cable and radioactive source guide tube to the radiography camera. The source was cranked out, and a second exposure of the weld was made. Next, the radiographer developed the film as the helper discon-nected the equipment. Neither realized that the radioactive source had not been connected to the drive cable and therefore could not be returned to the exposure device. Consequently, the radioactive source remained in an unshielded condition at the end of the guide tube. The radiography camera was placed near the rear of the truck, but the guide tube and cable were left lying on the ground approximately 20 feet from the weld. The two men then prepared to perform the job of stress relief around the weld. The process took about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. During this operation, they were either in the truck cab or watching the instrumentation near the weld.

Upon completion of work, all equipment was placed in the truck including the guide tube containing the source and the men returned the truck and equipment to the Western Stress facility in Evanston, Wyoming. The following day, two radiographers took the truck to a job site at Black Canyon, Wyoming. Radiography was performed using the guide tube containing the 29 curie Ir-192 source attached to a radiography camera different from the camera used the previous day. Several exposures were made, and the developed film was found to have double images.

The radiographer was then aware of the problem and placed the exposure device and guide tube in a transport container and covered it with bricks. They con-tacted the company Radiation Safety Officer, and the truck and equipment were returned to the Western Stress facility in Evanston. The source was secured by the Radiation Safety Officer the following day.

21

The personnel dosimeters of the employees involved in the incident were evalu-ated. They indicated whole body radiation doses of 22.1, 7.4, and 0.6 rem to the original radiographer, his helper, and another employee, respectively.

On August 6, 1985, NRC Region IV inspectors met with representatives of Western Stress in Evanston, Wyoming and discussed the incident and evaluated information relative to the event.

Subsequently, an anonymous caller contacted NRC Headquartefs on August 12, 1985, concerning Western Stress and stated that there had been other work involving the truck containing the exposed source by radiographic personnel who were un-authorized and who did not wear personnel monitoring devices. The job site, at which work was performed after the Table Rock work and before the Black Canyon .

work, was at Green River, Wyoming.

On August 13, 1985, NRC Region IV was notified by Western Stress maragement that additional use of the truck containing the radioactive source while in the unshielded condition had not been reported to the NRC during the week of August 6.

An extensive inspection was initiated by NRC Region IV personnel at Evanston, Wyoming on August 14, 1985. The inspection confirmed the information reported by the anonymous caller and later reported by company management. Interviews with radiographic personnel and reenactment of the events indicated that as many as six members of the general public may have received some exposures.

However, best estimates are that the exposures were very low.

Oak Ridge Associated Universities Medical and Health Science Division performed cytogenetic dosimetry evaluations on blood samples taken from the radiographic personnel involved. Results of these studies showed at the 80 percent confi-dence level that the original radiographer's dose was not smaller than 8 rad nor larger that 31 rad and his helper's dose was not greater than 15 rad.

Cause or Causes - The root cause was due to a serious breakdown in management controls and oversight of the licensed program.

Actions Taken to Prevent Recurrence 4

Licensee - On August 9, 1085, Western Stress voluntarily agreed to suspend op-erations until manage e+ had made the necessary changes in the program to satisfy the NRC that they could meet the NRC's regulatory requirements. Per-mission to resume operation was given on October 3,1985, af ter an additional NRC Region IV inspection confirmed that program improvements had been made. A license amendment was subsequently issued on October 4, 1985, to Western Stress,

.which included procedural and management changes.

1 NRC - On August 21, 1985, an enforcement conference was held in the NRC Region IV Office with members of Western Stress management. Items discussed were: the use of unauthorized radiographers to perform radiography, failure to wear personnel monitoring equipment, multiple failures to make radiation surveys

. required by regulations or company procedures, and the general breakdown in management controls and oversight of the licensed program.

The event remains under review by the NRC.

22 I

Future reports will be made as appropriate.

85-18 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 19, 1985, Riverside Methodist Hospital of Columbus, Ohio, reported to the NRC that a 78 year-old patient had received a radiation exposure from a diagnostic test that was 10 times greater than that which had been intended.

Nature and Probable Consequences - On August 17, 1985, the patient underwent a blood pool imaging study. The diagnostic test involves injecting a radioactive riterial (sodium pertechnetate-99m) into the patient and then recording the move-ment and location of the radioactive material with a scanning device. The dia-gnostic test called for use of 20 millicuries ~of the sodium pertechnetate-99m, but the patient received 200 millicuries of the material.

A technologist prepared the material for the test, using a dose calibrating device to measure the amount of radioactivity. Measurements were made of the bulk supply of the sodium pertechnetate and of the single-dose syringe prepared by the technologist. The dose calibrator malfunctioned in both measurements, showing a measurement which was 1/10th of the actual amount. Therefore, the dose, measured as 20 millicuries in the calibrator, was actually 200 millicuries.

The error was discovered when the scanning test was performed. The licensee calculated that the patient received a whole body radiation dose of 3.28 to 3.5 rads. (A rad is a standard measure of radiation exposure.) This level.is far below the point where any detectable medical effects would be anticipated.

Cause or Causes - The misadministration was caused by the malfunction of the dose calibrator. The digital display on the calibrator misplaced the decimal point, thereby leading to the use of 200 millicuries of the material instead of the intended 20 millicuries.

. - a calibrator had previously malfunctioned in June 1935 and was returned to th- r:nufacturer'for service. The malftnction - by a factor of 10 resurfaced again in August, but could have been corrected by removing and reinserting the container being measured. The Chief Technologist was not informed of the problem, and no action was taken at.that time.

Actions Taken To Pre' vent Recurrence Licensee - After the misadministration occurred, the licensee attempted to duplicate the instrument malfunction, but was unable to do so. It was placed back in service until August 22, when the malfunction reoccurred. The device was then returned to the manufacturer for repair.

23

NRC - A special inspection (Ref. 18) was conducted on September 3, 1985, to review the circumstar.::= of the misadministration. The licensee's handling of

~

the incident and the corrective measures taken were found to be acceptable. No violations of NRC regulations were identified.

This incident is closed for the purposes of this report.

AGREEMF.NT STATE LICENSEES Procedures have been developed for the Agreement States to screen unscheduled incidents or events using the same criteria as the NRC (See Appendix A) and report the events to the NRC for inclusion in this report. During the third calendar quarter of 1985, the Agreement States reported no abnormal occurrences to the NRC.

24

REFERENCES

1. Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to Cordell Reed, Vice President, Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, June 17, 1985.*

2. Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to Cordell Reed, Vice President, Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, July 22, 1985.*

3. Letter from James G. Keppler, Regional Administrator, NRC Region III, to James J. O'Connor, President, Commonwealth Edison Company, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, Docket Nos. 50-373 and 50-374, September 27, 1985.*

4. Letter from James G. Keppler, Regional Administrator, NRC Region III, to James J. O'Connor, President, Commonwealth Edison Company, forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, Docket Nos. 50-373 and 50-374, March 21, 19.85.* .

5. 10 CFR 50.54(f) letter from James G. Keppler, Regional Administrator, NRC Region III, to Cordell Reed, Vice President, Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, November 22, 1985.*

6. Letter from Cordell Reed, Vice President, Commonwealth Edison Company, to James G. Keppler, Regional Administrator, NRC Region III, Docket Nos. 50-373 and 50-374, December 23, 1985.*

7. Letter from Thomas E. Murley, Regional Administrator, NRC Region I to J. B. Randazza, Vice President - Nuclear Operations, Maine Yankee Atomic Power Company, forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, Docket No. 50-309, October 29, 1985.*

8. 10 CFR 50.54(f) letter from William J. Dircks, NRC Executive Director for, Operations, to Charles Dean, Chairman, Board of Directors, Tennessee Valley Authority, Docket Nos. 50-259, 50-260, 50-296, 50-327, 50-328, 50-390, 50-391, 50-438, and 50-439, September 17, 1985.*

9. Letter from James P. O'Reilly, Regional Administrator, NRC Region II, to H. G. Parris, Manager of Power and Engineering, Tennessee Valley Authority, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties,

' Docket Nos. 50-25!, 50-260, and 50-296, January 28, 1985.*

10. Letter from J. Neh m Grace, P!gional Administrator, NRC Region II, to H. G. Parris, Manage' of Pove, e'd En1inering, Tennessee Valley Authority, forwarding a Notice o: Vial- tion anu Pn ,osed Imposition of Civil Penalties, Docket Nos. 50-259, 50-260, and 50-296, % ruary 27, 1985.*

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

25

11. Letter from J. Nelson Grace, Regional Administrator, NRC Region II, to H. G. Parris, Manager of Power and Engineering, Tennessee Valley Authority, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, Docket Nos. 50-259, 50-260, and 50-296, July 22, 1985.*

12. Letter from J. Nelson Grace, Regional Administrator, NRC Region II, to H. G. Parris, Manager of Power and Engineering , Tennessee Valley Authority, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, Docket No. 50-327, May 7, 1985.*

13. U.S. Nuclear Regulatory Commission, " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," Federal Register Vol. 48, N' 15, January 21, 1983, 2729-2734.

14.' Confirmation of Action Letter from J. Nelson Grace, Regional Administrator, NRC Region II, to H. G. Parris, Manager of Power and Engineering, Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, August 23, 1985.*

15. Letter from J. Nelson Grace, Regional Administrator, NRC Region II, to H. G. Parris, Manager of Power and Engineering, Tennessee Valley Authority, forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, Docket No. 50-390, August 29, 1985.*

16. Letter from William J. Dircks, NRC Executive Director for Operations, to Charles Dean, Chairman, Board of Directors, Tennessee Valley Authority, Docket Nos. 50-259, 50-260, 50-296, 50-327, 50-328, 50-390, 50-391, 50-438, and 50-439, July 3,' 1985.*

17. Letter from William J. Dircks, NRC Executive Director for Operations, to C. H. Dean, Jr. , Chairman, Board of Directors, Tennessee Valley Authority, August 1, 1985.*

~

18. Letter froa D. J. Sreniawski, Chief, Nuclear Materials Safety Section'2, NRC Region III, to M. Vergales, Vice President of Professional Services, Riverside Methodist Hospital, transmitting Inspection Report No. 030-02669/85001(NMSS), October 7,1985.*

26 l

l

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

An event will be considered an abnormal occurrence if it involves a major 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. Major degradation of essential safety-related equipment; or

3. Major deficiencies in design, construction, use of, or management controls for licensed facilities or material.

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

l For All Licensees

1. Exposure of the whole body of any indivicual 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 620.403(a)(1)),

or equivalent exposures from internal sources.

2. An exposu,w 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.

27

7. Any substantiated loss of special nuclear material or any substantiated inventory discrepancy which is judged to be significant relative.to normally expected performance and which is judged to be caused by theft or diversion or by substantial breakdown of the accountability system.

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

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

9. An accidental criticality (10 CFR S70.52(a)).

10. A major deficiency in design, construction or operation having safety implications requiring immediate remedial action.

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

12. Series of events (where individual events are not of major importance),

recurring incidents, and incidents with implications for similar facilities (generic incidents), which create major safety concern.

For Commercial Nuclear Power Plants

1. 2xceeding a safety limit of license technical specifications (10 CFR S50.36(c)).

2. Major degradation of fuel integrity, primary coolant pressure boundary, or primary containment boundary.

3. Loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g.,

loss of emergency core cooling system, loss of control rod system).

4. Discovery of a major condition not specifically considered in Uie safety.

analysis report (SAR) or technical specifications that requires immediate remedial action.

5. Personnel error or procedural deficiencies which result in loss of plant  !

capability to perform essential safety functions such that a potential  !

release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency.

core cooling system, loss of control rod system).

For Fuel Cycle Licensees

1. A safety limit of license technical specifications is exceeded and a plant shutdown is required (10 CFR S50.36(c)).

2. A major condition not specifically considered in the safety analysis 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.

28

APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the July through September, 1985 period, the NRC, NRC licensees, Agree-ment States, Agreement States 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 updated in each subsequent report in this series, i.e., NUREG-0090, Vol. 2, No. 2 through Vol. 8. No. 2. It is forther updated as follows:

Reactor Building Entries 4

During the third calendar quarter of 1985, 59 entries were made into containment.

There have been a total of 702 entries since the March 1979 accident. Reactor building activities during this period consisted primarily of preparations for early defueling, including assembly of the rotating defueling work platform and installation of the service work platform, jib cranes, defueling tool racks, and cable management system. Installation of the canister positioning system, canister handling bridge and vacuum defueling system commenced during the quarter, and preoperational testing of the defueling water cleanup system was conducted.

Also during this quarter, additional video inspections of the reactor vessel lower head were performed, as discussed below.

Lower Reactor Vessel Head Inspection Additional characterization of the reactor vessel lower head region was performed in July 1985. Video inspections were conducted in areas not previously examined and water and debris samples were taken. The inspections indicated that the debris bed was more shallow and that individual pieces were smaller than in those regions inspected in February 1985. Radiation measurements of debris bed samples indicated a much smaller concentration of radioactive material than expected. An analysis of the water sample indicated that the chemical and phys-ical properties of the water in the lower head are consistent with those else-where in the vessel.

EPICOR-II/ Submerged Demineralizer System (SDS) Processing The EPICOR-II system processed approximately 182,208 gallons of water during the reporting period. The SDS processed approximately 64,539 gallons of water during that time.

29

Temporary Suspension of Waste Burial Permit On August 6, 1985, the State of Washingt:n suspended GPU Nuclear's permit to ship radioactive waste from Three Mile Island to Richland, Washington. This action was taken after the discovery of three misclassified barrels of Class A waste out of a total shipment of 104. The three barrels exceeded the limits specified in 10 CFR S 61.55 and should.have been classified and handled as Class B waste. On August 16, 1985, GPU Nuclear's burial privileges were restored after Washington State officials approved the corrective actions taken by GPU Nuclear.

TMI-2 Advisory Panel Meetings The Advisory Panel for the Decontamination of Three Mile Island Unit 2 (Panel) met on July 18 1985, in Lancaster, Pennsylvania. The Panel received a status report on cleanup activities from GPU Nuclear and a renort from the NRC on activities requiring completion prior to the start of defueling. An Environ-mental Protection Agency (EPA) representative described the upgrad'd e aquatic monitoring program in the vicinity of TMI and the NRC staff presented a summary of the health effects studies conducted in the area following the accident.'

On September 11, 1985, the Panel met in Annapolis, Maryland. At this meeting the Panel received presentations from GPU Nuciear on the status of the cleanup, cleanup funding, and disposition of the processed accident generated water.

Representatives of the Department of Energy, EPA, and the State of Maryland Power Plant Siting Program described their respective agency's responsibilities I

l and activities with respect to TMI. The NRC staff briefed the Panel on the planned reorganization of the Office of Nuclear Reactor Regulation and the re-sulting changes to the TMI Program Office.

Further reports will be made as appropriate.

l * * * * * *

• 1 83-15 Emergency Diesel Generator Problems This abnormal occurrence was originally reported in NUREG-0090, Vol. 6, No. 4,

" Report to Congress on Abnormal Occurrences: October - December 1983," and updated in a subsequent report in the series, i.e., NUREG-0090, Vol 7, No. 3.

It is further updated as follows.

As mentioned in the previous reports, during a load test on August 12, 1983, the main crackshaft on one of the three emergency diesel generators (EDGs) at the Shoreham Nuclear Power Station (N.Y.) broke in two. The EDGs at Shoreham were manufactured by Transamerica Delaval, Inc.-(TDI), which has also supplied 54 other EDGs to 14 other nuclear power plant sites in the United States.

Also previously mentioned, during the evaluation of the Shoreham EDG failure, a broad pattern of deficiencies emerged with regard to critical engine components in EDGs manufactured by TDI. These deficiencies appeared to stem from inadequacies in design, manufacture, and quality assurance / quality control 30

by TDI. In response to these problems,13 nuclear utilities formed an Owners Group to establish a program for upgrading and confirming the adequacy of the TDI diesels for nuclear service.

During Fiscal Year 1985, the Owners Group Program proceeded beyond the existing problem areas to systematically consider all EDG components important to the operability and reliability of the engines. This step beyond the limited num-ber of components with known design and manufacturing problems is intended pri-marily to ensure that significant new problem areas do not develop due to de-ficiencies in design or quality of manufacture. The Owners Group has performed design reviews and recommended needed component upgrades, modifications, and component inspections to validate the quality of manufacture and assembly. In addition, a comprehensive engine mai.'tenance and surveillance program to be implemented by the individual owners was prepared.

The NRC staff has concluded that issues warranting priority attention have been adequately resolved at several plants such that the TDI EDGs will provide reliable service through at least the first refueling outage. As a result, Supplemental Safety Evaluation Reports concerning TDI diesel generators were issued to support issuance of Operating Licenses for Shoreham Nuclear Power Station Unit 1, Catawba Nuclear Station Unit 1, and River Bend Station Unit 1, and to support operation of San Onofre Nuclear Generating Station Unit 1, until its next refueling outage. The Staff expects to complete its. final evaluation of the Owners Group findings and recommendations from the program t later this year.

i Finally, hearings before the Atomic Safety and Licensing Board (ASLB) on the subject of the TDI engines at Shoreham Nuclear Power Station Unit 1, were completed on March 12, 1985. The Board found there is reasonable assurance that the TDI diesels can perform their required safety functions for the first refueling cycle. In addition, an ASLB hearing conducted on April 10 and 11, 1985, dismissed all contentions regarding the TDI diesels at Perry Nuclear Power Plant Unit 1.

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

85-7 Loss of Main and Auxiliary Feedwater Systems This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 2,

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

The NRC Incident Investigation Team, which was established by the Executive Director for Operations and dispatched to the site several days after the June 9, 1985, Davis Besse event, completed its fact finding efforts in early July. The Team's findings are reported in NUREG 1154, " Loss of Main and Auxiliary Feedwater at the Davis-Besse Plant on June 9, 1085" (Ref. B-1). The report was released to the public at the end of July. The specific sequence of events, operator actions and equipment failures and malfunctions are documented in the report.

31

On-July 24, 1985, the Commissioners were briefed by ti.e Incident Investigation Team on their findings. With the investigation complete,~ Toledo Edison company embarked on a program of troubleshooting and analysis to identify the fundamental cause for each of the equipment malfunctions or failures and to formulate correc-tive actions to assure equipment reliability. These efforts continued throughout the remainder of the quarter.

Based upon the findings of the Incident Investigation Team reported in NUREG-1154, the NRC identified the concerns Toledo Edison Company should address.for NRC review before resumption of operation of the plant can be approved. These con-cerns were identified to the licensee in a letter dated August 14, 1985 (Ref. B-2). The licensee responded in a document submitted to the NRC on Sep-tember 10, 1985, entitled " Davis-Besse Course of Action" (Ref. B-3), which is now under review by the NRC staff.

. Shortly after the event of June 9, 1985, the licensee undertook steps to accel-erate the installation of an electrically driven startup feedwater pump. This pump had previously been scheduled to be installed in the next refueling outage which was to have started in the spring of 1986. By early September the new pump had been delivered to the site and installation was started.

During the period, Toledo Edison Company continued in plant examinations of the as-built condition of various safety related pipe hangers and supports to determine any deviations from_ design drawings. These examinations have revealed numerous discrepancier, and analysis is being performed to determine the safety significance, if any, of these deviations.

Further reports will be made as appropriate. j 32

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. Two Stuck Control Rods During Testing On May 30, 1985, while performing a cold, full-flow control rod drop test at Point Beach Unit 1, control rod F-12 stuck at about 60 inches above the bottom of the core. On May 31, 1985, during a hot, full-flow drop test on all the other rods, rod J-4 (which dropped successfully during the cold test) dropped only 45 inches, sticking at about 99 inches from the bottom of the core.

Point Beach Unit 1, operated by Wisconsin Electric Power Company, utilizes a pressurized water reactor designed by Westinghouse. The plant is located in Manitowoc County, Wisconsin.

The unit was shut down on April 5, 1985, for refueling, maintenance, and modi-fications. Part of the modification program included work on the control rod drive system to replace control rod guide tube (CRGT) flow inserts. The previ-

' ous flow inserts were replaced since they were susceptible to stress induced cracking.

In the previous design, each flow insert was held down by four flexure pins.

(The flow insert and flexure pins are located at the top of each guide tube.)

During the modifications, these flexure pins and flow inserts were removed and the inserts replaced by ,;ew " flexure pinless" flow inserts (referred to as "flexureless inserts"). In addition, the guide tube alignment pins (split pins), located at the bottom of each guide tube, were replaced since they were also susceptible to stress induced cracking. The flexure pins were removed by a device which would cut the pin and hold it to keep it from falling into the control rod drive mechanism.

On May 30, 1985, as part of the preparations for plant startup, a cold, full-flow control rod drop test was performed. Rod F-12 did not drop to the bottom of the core, but stuck about 60 inches from the bottom. Attempts to move the rod by stepping were unsuccessful.

.The licensee decided to proceed with reactor coolant system heatup in order to perform an inservice leak test on the system as well as hot rod testing of the remaining rods. During the hot (370 degrees F), full-flow rod drop testing, rod J-4 (which had successfully dropped in the cold test) became stuck about 99 inches from the bottom of the core. The licensee then returned the plant to cold shutdown.

In order to determine the cause of the problem, it was necessary to lift the vessel head and remove rods F-12 and J-4, as well as the flexureless inserts, from their guide tubes and perform a visual inspection of the inserts and the guide tube internals. Inspection of the flexureless insert removed from the 33

J-4 guide tube showed that a piece of the skirt section was missing. Inspection of the J-4 and F-12 guide tubes disclosed the causes of the rods failing to drop completely into the core: (1) the missing piece of flexureless insert skirt was found in the J-4 guide tube on a guide tube card corresponding to the elevation at which rod J-4 had stuck, and (2) one of the previously used flexure pins was found in the F-12 guide tube on a guide tube card corresponding to the elevation at which rod F-12 had stuck.

The flexure pin and piece of flexureless insert were removed and no other debris was found. All flexureless inserts were removed and inspected. Three were found to have bent tabs. These three and the broken one from location J-4 were replaced from spares. It is believed that the cause of the damaged flexureless inserts was due to an inadequate installation procedure. This procedure was revised and all flexureless inserts were reinstalled with no problems. The con-trol rods from positions F-12 and J-4 were replaced and the control rod drive shaft from position F-12 was replaced due to damage noted during visual inspections.

The reason that the loose flexure pin had been left in the F-12 guide tube was due to inadequate inventory control during the modification process.

After replacing the rods and vessel head, cold and hot rod drop tests were satisfactorily performed on June 16 and June 18, 1985, respectively. Later on June 18, 1985, the reactor went critical, ending the refueling outage.

The event had no impact on public health or safety since it occurred during f testing which was routinely performed to determine whether there were any prob-lems prior to taking the reactor critical.

2. Diesel Generator Load Sequencing Inoperable On August 22, 1985, Iowa Electric Light and Power Company reported that the emergency-diesel generators (EDGs) at Duane Arnold had been declared inoperable after determining that EDG load sequencing would not occur properly under cer-tain conditions. Duane Arnold, located in Linn County, Iowa, utilizes a boiling water reactor designed by General Electric.

On November 4, 1984, there was an explosion and fire in the auxiliary trans-former. Since that time, the auxiliary transformer had been out of service.

The plant non-vital loads (normally supplied by the auxiliary transformer) were being supplied by the startup transformer. The plant vital loads (normally supplied by the startup transformer) were being supplied by the standby trans-former.

As a result of analysis by training department employees, the licensee deter-mined on August 22, 1985, that in the present electrical bus configuration, if the standby transformer stopped supplying power, the startup transformer con-tinued to supply power, and a subsequent loss of coolant accident (LOCA) signal (high drywell pressure or triple low reactor water level) occurred, the four residual heat removal (RHR) pumps and two core spray pumps would load onto their respective diesel generators simultaneously instead of sequencing on as designed.

The licensee could not show that the EDGs could supply the starting power neces-sary to start the pumps simultaneously. The licensee, therefore, declared an unusual event and declared both EDGs technically inoperable.

34

The licensee took interim corrective action within twenty-four hours by placing four test switches in the test position. This forces the affected pumps to sequence onto their power supply regardless of whether the power supply is a transforner or a EDG. The licensee is in the process of developing a permanent modification to correct the problem.

The WRC Senior Resident Inspector at the plant monitored the licensee's actions tnroughout the event. The temporary corrective action was thoroughly reviewed prior to implementation to ensure that the desired effect was achieved without any inadvertent adverse effects. The NRC reviewed the sequencing systems for similar plants to ensure that other plants did not have a similar design defi-ciency. No other plants were found to have this design.

The root cause was attributed to the original des'ign which did not provide for the plant systems to recognize the effect on load sequencing if the standby transformer is deenergized, the startup transformer remains energized and the vital buses are being supplied by the standby transformer. If a total loss of off-site power occurred and a subsequent LOCA signal was received, the loads would have sequenced properly since the startup transformer would also have deenergized. If the standby transformer had stopped supplying power, the startup transformer had continued to supply power and a subsequent LOCA signal had occurred, the LOCA loads would not have sequenced onto their respective EDGs. The consequences of this would most likely have been a failure of the EDGs to supply the emergency loads. The failure of the EDGs to supply the vital buses is indicated in the control roem by a variety of indications. The fact that the startup transformer is st'.11 energized is also indicated in the control room and the operator in the control room can manually transfer.the vital buses to the startup transformer. Although it would take time for the operator to recognize, diagnose, and correct the problem, the startup transformer would '

still be supplying power to tha two electric reactor feedwater pumps and the two electric condensate pumps which would be supplying water to the reactor vessel. Therefore, the probability of significant impact on public health or safety as a result of this event was low.

3. Truck-Train Wreck Involving Spill of Uranium Concentrates On August 27, 1985, at 4:33 p.m. CDT, a truck loaded with 53 drums of natural uranium concentrates collided with a train crossing Highway 52, about 12 miles south of Fessenden, North Dakota, an NRC Agreement State. The shipment was enroute from Saskatchewan Mining Development Corporation, Saskatchewan, Canada to the Sequoyah Fuels (formerly called Kerr-McGee Nuclear Corporation) facility at Gore, Oklahoma. During the accident, the truck driver was killed, the train locomotives derailed without injury to anyone, and most of the drums were rup-tured, spilling concentrates about 200 feet along the track and immediate sur-rounding areas. Approximately.10 tons of concentrate spilled covering the train i engine, tractor-trailer, railroad track, and immediate vicinity.

Local fire and police departments responded to the accident within 30 minutes and in turn notified the State of North Dakota emergency response team who arrived at the site within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Access to the area was closed and arrange-ments were made to check for internal and external contamination of personnel in the area. The NRC Headquarter's Duty Officer was notified of the accident at 6:15 p.m. CDT on August 27, 1985, who in turn notified NRC Region IV. FEMA, as well as DOT and DOE were also notified; however, they decided not to respond 35 l

to the accident. NRC Region IV dispatched two health physicists.to the site on August 28, 1985, who provid'ed technical. assistance to the State until the cleanup was completed. A response team from the shipper arrived at the site on August 30, 1985, and a contractor crew of 20 decontamination personnel was subsequently hired and utilized by the shipper.

-The State of North Dakota remained in control of the accident and established a control center at the site.

Uranium concentrates present no significant external radiation hazard to per-sonnel; however, internal uptake of the concentrates in sufficient amounts could cause heavy metal damage to the kidneys. Persons involved in the acci-dent were given bioassays to assess any internal uptake and all results were negative.

Following the accident and until decontamination of the area was complete, traffic was detoured to an alternate route. During the decontamination phase, airborne concentrations of the uranium concentrates were well controlled. All

~

uranium concentrates were recovered using vacuum cleaners, and the ground was decontaminated to background levels by scraping the top soil with a front loader scraper. Recovered concentrates were placed in drums and packaged for return to the shipper in Canada. Parts of the train engine were disassembled for ship-ment to Canada. The remainder of the train was decontaminated to levels speci-fied by the State decontamination guidelines (which follow NRC guidelines).

The tractor-trailer'was loaded on another trailer and shipped back to Canada along with 37 trailer loads of the above decontamination material. The body of the truck driver was decontaminated at a Minot, North Dakota hospital with water which drained to the sanitary sewer at levels permitted by State regulations.

The decontamination of the site was completed on September 12, 1985, with no significant external or internal exposure to personnel involved in the accident or to workers involved in subsequent cleanup activities at the site.

There was considerable media interest in the accident, and subsequent to the accident the Governor of North Dakota formally complimented the NRC Region IV staff for their support in the matter.

4. Degraded Containment. Integrity On September 2, 1985, Detroit Edison Company discovered an open valve in a small-diameter containment monitoring system line at the Fermi Unit 2 Nuclear Power Plant. The valve had been left open since June 21, 1985. The open valve repre-sented a violation of containment integrity. Fermi Unit 2, located in Monroe County, Michigan, utilizes a boiling water reactor designed by General Electric.

The open valve represented a significant breach of containment integrity since t' it went undetected for such a long period of time. However, the effects on public health and safety were minimal, even if a design basis accident had occurred. The licensee estimated that the leak rate through.the unisolated line connecting the torus air space with secondary containment would be approxi-mately 35,000 SCFH under limiting accident conditions. This represents an in-crease in primary containment leak rate over what is currently assumed in the design basis loss of coolant accident (LOCA) analysis by at least two orders of magnitude. Further evaluation by NRC of the radiological consequences associated 36

with this leakage indicates that should a design basis LOCA have occurred during the period the containment breach existed, the existing low levels of radioac-tivity in the core would offset the increased leak rate and serve to limit cal-culated doses at the site boundary to levels below 10 CFR Part 100 limits.

On March 25, 1985, an engineering design package was issued for the installation of four test connection cutoff valves in the primary containment monitoring system. The valves were installed--and the work orders for the installation were signed off as completed on June 20, 1985. Although required by the work orders, there were no local leak rate tests performed on the monitoring system modifications. The licensee also failed to add the containment monitoring sys-tem valves to the checklists used for routinely verifying containment integrity.

On September 2, 1985, the licensee discovered that one valve was open. The plant had operated in this condition since June 21, 1985, at power levels g:n-erally up to 5%. The licensee closed the valve. The sample system valves were added to the appropriate check lists used in verifying containment integrity.

The local leak rate tests, omitted following the valves' installation, were performed in September 1985. Three of the installations met the test criteria; the fourth installation required rework before it met the test criteria.

The NRC's Resident Inspector at the plant reviewed the circumstances surrounding the open valve. Three apparent violations of NRC requirements were identified:

operating from June 21, 1985, until September 2, 1985, without the required con-tainment integrity; failure to perform the Local Leak Rate Tests; and failure e

to place the valves on the checklists used for verifying containment integrity.

Enforcement action on these apparent violations is pending.

i 37

REFERENCES FOR APPENDICES B"1 U.S. Nuclear Regulatory Comraission. " Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant or. June 9, 1985," USNRC Report NUREG-1154, published July 1985.*

B-2 10 CFR S 10.54(f) letter from Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, to Joe Williams, Jr. , Senior Vice President, Nuclear, Toledo Edison Company, Docket No.50-34G, August 14, 1985.5*

B-3 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 IC, 1985.**

• Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection. Available for purchase from the GP0 Sales Program, Superin-tendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7982.

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

39

l

=a,c ,omu sas ' u s aiucu.a nnoutaronv cou iss>O= i airo r ~uvet- l AmM Dy TfDC Me M 4e of M A

%"3s'- BIBLIOGRAPHIC DATA SHEET NUREG-0090 un srauctio%s o~ v- svias. Vol. 8 No. 3 uira . o su.ma 3 u . v . . t..

Report to ongress on Abnormal Occurrences July-Septem er 1985 4 omp Oar co ano _

.O ,-

7 l

,s .

. .ur Oa si February / 1986

/ 6 0.f t MtPORT 54uED

7. l Febru6ry 1986

, . .. o. ,~o 0 a. iz.i iO~ ,.... .,<,so .ao .ss ,, ode i, c e, . , . 1 3. .o.. us,,~u....

U.S. Nuclear Regulat (y Commission ~ ,7,,0,0,,,,,,,,,,,

Office for Analysis an Evaluation of Operational Data /

Washington, D.C. 20555

10. 5PONSO*i NG ORG.Ntz.T60N N.wt .8.O M.*Ls4G .OOPE ss tsace wee le Coces lla T YPE OF ALPGRT Quarterly b PE RaOD CGvf REO isar% ewe dsfes/

' ~

July-September 1985 12 SUPPLEMENT.Rv NOTES 13 A857R.CT (200 mores er <esst

( Section 208 of the Energy Reorganization et of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission datermines to be significant from the standpoint pubTic health and safety and requires a quarterly report of such events to be ade to Cqngress. This report covers the period July 1 to September 30, 1985./ During the\ report period, there were three abnormal occurrences at the nuclear power plants 1jcensed to operate. These events j

involved, respectively, (1) manager $nt control def fciencies, (2) inoperable steam generator low pressure trip, and (3) management defitiencies at Tennessee Valley Authority. There were four abnorpal occurrences at the other hRC licensees. Three of the events involved medical misadministrations - two stherapeutic and one diagnostic.

The other ever.t involved expos e of radiographic personh,el due to management and procedure control deficiencies There were no abnormal occurrences reported by the Agreement States. Thereportdiso.containsinformationupdatingsomepreviously reportedabnormaloccurrencesj \

\u k+

I i j " ;=="

.} l

\

s Unlimited 16 SECualT V CL AS5ttit.TaON ITern ee,el

. oo.1,... s,o,i~e.ononaos Management Deficiencies; Personnel Errors; Equipment U Modification Deficiencies;. Regulatory Problems; Equipment Failure; Pressure <r nclassified

.r ,

Transmitters Inoperable; Overpressurization; Cable Tray Supports; Diesel IIn Generators; Seal Table; Welder Certifications; Environmental Qualification "~ classified a a m .ois of Electrical Equipment; Therapeutic and Diagnostic Medical Misadministra-tions; Radiographic Personnel Exposure; Stuck Control Rods; Spill of 4 -ci Uranium Concentrates; Degraded Containment

UNITED STATES NUCLEAR RE~lutAT2:.Y CZMMISSl!N sncut rouarw ctass man =

WASHINGTON, D.C. 20555 SST^ts;"S *D gas,*g g ,, ,

OFFICIAL BUstNESS  ;

PENALTY FOR PRIVATE USE. 4300 120555078877 US NRC 1 1ANIC01CVICY1 -

ADM-DIV 0F TIOC ,

POLICY & PUB MGT W-501 ER-POR NUREG WASHINGTON i OC 20555 G

b 1

C F,

l 4

. 3 en t I

-43

• )

E t CD 3 A

C 2:

'm

$ )3 c1 r

(

c C

5 r

C r

C r

C D

L 3

J C

C

_