ML20140C088

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Report to Congress on Abnormal Occurrences.October - December 1983
ML20140C088
Person / Time
Issue date: 05/31/1984
From:
NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:
References
NUREG-0090, NUREG-0090-V06-N04, NUREG-90, NUREG-90-V6-N4, NUDOCS 8406190041
Download: ML20140C088 (29)
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s NUREG4090 Vol. 6, No. 4 l

Report to Congress on Abnormal Occurrences October - December 1983 l

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

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NUREG M Vol. 6, No. 4 1

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l Report to Congress on

Abnormal Occurrences l

October - December 1983 Dita Published: May 1984 l

Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission Wahington, D.C. 20555 v a *%,,,

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Previous Reports in Series a

i NUREG 75/090, January-June 1975, NUREG-O')q0, Vol .2, No.3, July-September 1979, published October 1975

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published February 1980 NUREG-0090.-1, July-September 1975, NURGi-0090, Vol .2, No.4, October-December 1979, published harch 1976 published April 1980 NUREG.0090-2, October-December 1975, NUREG-0090, Vol.3, No.1, January-March 1980, published March 1976 published September 1980 NUREG-0090-3, January-March 1976, NUREG-0090, Vol .3, No.2, April-June 1980, published July 1976- publishet November 1980 NUREG-0090-4, April-June 1976, NUREG-0090, Vol .3, No.3, July-September 1980, published March 1977 published February 1981 NUREG-0090-5, July-September 1976, NUREG-0090, Vol .3, No.4, October-December 1980, published March 1977 published May 1981 NUREG-0090-6, October-December 1976, NUREG-0090, Vol .4, No.1, January-March 1981, published June 1977 published Jply 1981 WUREG-0090-7[ January-March 1977, NUREG-0090,,Vol.4, No.2, April-June 1981, published June 1977 published October 1981 NUREG-0090-8, Aprff-June 1977, NUREG-0090,Vol.4,No.3, July-September 1981, published September 1977 published January 1982 NUREG-0090-9, July-September 1977, NUREG-0090, Vol.4, No.4, October-December 1981, published Notember 1977 published May 1982 NUREG-0090-10, October-December 1977, NUREG-0090, Vol.5, No.1, January-March 1982 published March 1978 ptblished August 19P(.

NUREG-0090, Vol .1, No.1, January-March 1978, NUREG-0090, Vol.5, No.2, April-June 1982, published June 1978

, publ( ..ed December 1982 NUREG-0090, Vol.1, No.2, April-June 1978, NUREG-009'J, Vol .5, No.3, July-September 1982, published September 1978 published January I?83 NUREG-0090, Vol.1, No.3, July-September 1978, NUREG-0090, Vol.'5, No.4, October-December 1982, published December 1978 ,

published May 1983 NOREG-0090, Vol.1, No.4, October-December 1978, NUREG-0090, Vol.6, No.1, January-March 1983, published Msrch 1979 published September 1983

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NUREG-0090, Vol .6, No.2, April-June 1983, NUREG-0090, Vol.2, No.1, January-March 1979, published Julj 1979 published November 1983 NUREG-0090, Vol . ', Ho.2, . April-June 1979, NUREG-0090, Vol.6, No.3, July-September 1983.

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I ABSTRACT Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety and requires a quarterly report of such events to be made to Congress. This report covers the period from October 1 to December 31, 1983.

The report states that for this report period, there was one abnormal occurrence at the nuclear power plants licensed by the NRC to operate. The item involved generic problems pertaining to a specific manufacturer's emer-gency diesel generators. There was one abnormal occurrence for the other NRC licensees. The item involved an overexposure of a radiographer. There was one abnormal occurrence reported by an Agreement State. The item involved an overexposure to a radiographer.

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

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1 l-i CONTENTS Page ABSTRACT ......................................................... iii PREFACE .......................................................... vii INTRODUCTION ............................................ ... vii THE REGULATORY SYSTEM ......................................., vii REPORTABLE' OCCURRENCES ...................................... viii AGREEMENT STATES ............................................ ix FOREIGN INFORMATION ......................................... x REPORT.T0 CONGRESS ON ABNORMAL OCCURRENCES, OCTOBER-DECEMBER 1983 ............................ ............. 1 NUCLEAR POWER PLANTS ........................................ 1 83-15 Emergency Diesel Generator Problems ............. 1 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants) ..... Es OTHER NRC' LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, Etc.) ..........'.......... 5 x

83-16 OverexposureofaRadiogr4pher........'.......... 5 AGREEMENT STATE LICENSEES ...................J. ............'. 7 s

AS83-10 Overexposure of a Radiographer ....T........... 7 i b REFERENCES '

9 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA ......;.......;,.......... .

11 APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES .. 13 NUCLEAR POWER PLANTS .t........................................ 13 79-3 -Nuclear Accident at:Three Mile Island ............ 13 83-5 Large. diameter Pipe Cracking in Boiling N Water Reactors (BWRs) .......................... 15

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APPENDIX C - OTHER EVENTS OF INTEREST 17-REFERENCES (FOR APPENDICES) .................................. .-.. .

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5 PREFACE INTRODUCTION I

The Nuclear Regulatory Commission reports to the Congress each quarter under

. provisions of Section 208 of the Energy Reorganization Act of 1974 on any e abnormal occurrences involving facilities and activities regulated by the NRC.

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

Events are currently identified as abnormal occurrences for this report by the NRC using the criteria delineated in Appendix A. These criteria were promul-gated in an NRC policy statement which wasl 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 i Document Room and all local public document rooms. At a minimum, each such i 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 t

material, and other categories of information available to the NRC. -The NRC i has determined that only those events, including those submitted by the Agree-i ment States, described in this report meet the criteria forLabnormal occurrence i reporting. This report covers the period between October 1 to December'31, 1983.

Information reported on each event includes: date and place; nature and

. probable consequences; cause or causes; and actions taken to prevent recurrence.

1 THE REGULATORY SYSTEM The system of licensing ~and regulation by which NRC carries out its'responsi-

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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.

L 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 i

standards and rules;: issuing licenses and permits; inspecting for compliance;-

! enforcing license requirements; and carrying on continuing evaluations, studies and research projects to improve both the regulatory process'and the l- protection of the public health and. safety. 'Public participation is an element t- sof 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

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N be achieved and maintained through regulations which specify requirements which will assure the safe use of nuclear materials. The regulations include design and quality assurance criteria appropriate for the various activities licensed by NRC. An inspection and enforcement program helps assure compli-ance with the regulations. Requirements for reporting incidents or events exist which help identify deficiencies early and aid in assuring that correc-tive action is taken to prevent their recurrence.

After the accident at Three Mile Island in March 1979, the NRC and other groups-(a Presidential Commission, Congressional and NRC special inquiries, industry, special interests, etc.) spent substantial efforts to analyze the accident and its implications for the safety of operating reactors and to identify the changes needed to improve safety. Some deficiencies in design, operation and regulation were identified that required actions to upgrade the safety of nuclear power plants. These included modifying plant hardware, improving emergency preparedness, and increasing considerably the emphasis on human factors such as expanding the number, training, and qualifications of the reactor operating staff and upgrading plant management and technical support staffs' capabilities. In addition, each plant has installed dedicated telephone lines to the NRC for rapid communication in the event of any inci-dent. Dedicated groups have been formed both by the NRC and by the industry for the detailed review of operating experience to help identify safety concerns early, to improve dissemination of such information, and to feed back the experience into the licensing and regulation process.

Most NRC licensee employees who work with or in the vicinity of radioactive materials are required to utilize personnel monitoring devices such as film badges or TLD.(thermoluminescent dosimeter) badges. These badges are processed periodically and the exposure results normally serve as the official and legal record of the extent of personnel exposure to radiation during the period the badge was worn. If an individual's past exposure history is known and has been sufficiently low, NRC regulations permit an individual in a restricted area to receive up to three rems of whole body exposure in a calendar-quarter.

Higher values are permitted to the extremities or skin of the whole body. For unrestricted areas, permissible levels of radiation are considerably 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 expo-sures to levels as low as reasonably achievable.

REPORTABLE OCCURRENCES Since-the NRC is responsible for-assuring that regulated _ nuclear activities are conducted safely, the nuclear industry _is required to~ report incidents'or events which involve a variance from the regulations, such as personnel over-exposures, radioactive material releases above prescribed limits, and malfunc '

-tions of safety related equipment. Thus, a reportable occurrence-is any incident or event occurring at a licensed facility or related to licensed-activities which.NRC licensees are required to report to the NRC. The NRC-evaluates each reportable occurrence _to determine the. safety implications

. involved.

Because of the broad scope of regulation and the conservative attitude toward safety,.there are'a large number of events reported to the NRC. The information viii

provided in these reports is used by the NRC and the industry in their

~ continuing evaluation and improvement of nuclear safety. Some of the reports describe events that have real or potential safety implications; however, most of the reports received from licensed nuclear power facilities describe events that did not directly involve the nuclear reactor itself, but involved equip-ment and components which are peripheral aspects of the nuclear steam supply system, and are minor in nature with respect to impact on public health and safety. Many are discovered during routine inspection and surveillance testing and are corrected upon discovery. -Typically, they concern single malfunctions of components or parts of systems, with redundant operable components or systems continuing to be available to perform the design function.

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

Dissemination includes deposit of incident reports in the NRC's public docu-ment rooms, special notifications to licensees and other affected or inter-ested groups, and public announcements. In addition, information on report-able events received from NRC licensees 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 at 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 rel nquishes and the States assume regulatory authority over byproduct, source and special nuclear l materials (in quantities not capable of sustaining a chain reaction). Compa-rable 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 agree-ments. NRC prepares a semiannual summary of this and other information in a document entitled, " Licensing Statistics and Other Data," which is publicly

! available.

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.

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

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9 REPORT TO CONGRESS ON ABNORMAL OCCURRENCES OCTOBER-DECEMBER 1983 l NUCLEAR POWER PLANTS

[ The NRC is reviewing events reported at the nuclear power plants licensed to E operate during the fourth calendar quarter of 1983. As of the date of this r

report, the NRC had determined that the following was an abnormal occurrence.

83-15 Emergency Diesal Generator Problems The following information pertaining to this event is also being reported

. concurrently in the Federal Register. Appendix A (see Example 12 of "For All E Licensees") of this report notes that incidents with implications for similar facilities (gencric incidents), which create major safety concern, can be considered an abnormal occurrence. The problem discussed below involving the Transamerica Delaval, Inc. (TDI) emergency diesel generators (EDGs) at the L Shoreham Nuclear Power Plant was previously described in Appendix C of NUREG-0090, Vol. 6, No. 3. It was not reported as an abnormal occurrence at that time

- because the immediate problem involved a plant still under construction.

However, it was mentioned that reliability of the TDI EDGs remained under active review. It has now been determined that the question of reliability of f TDI diesels has generic implications and should be reported as an abnormal y occurrence.

g Date and Place - On August 12, 1983, EDG-102 at the Shoreham Nuclear Power g Plant (99% construction completion) failed due to a fractured crankshaft. The applicant for the plant is Long Island Lighting Company. The plant is a boil-6 ing water reactor and is located in Suffolk County, New York. There are three

, EDG units at Shoreham, all manufactured by TDI. During the following investiga-tions of the failure and needed repairs, several conditions were identified which raised questions about the reliability of all TDI diesels at other nuclear power stations.

B-g Nature and Probable Consequences - The failure at Shoreham occurred after s 1.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> of testing at the two-hour overload rating (3900 kW). At the time g of failure, EDG-102 had accumulated about 718 operating hours and about 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> at the 110% overload rating. The test in progress when the crankshaft fractured 6,

was being performed to demonstrate EDG load carrying ability following replacement I of all eight cylinder heads with a newer design (originally supplied cylinder heads had developed leaks from the cooling water area).

m, The EDG-102 crankshaft fracture occurred on the generator (load) side of the

{ No. 7 cylinder and extended through the load side crank arm into the crank pin. (The No. 8 cylinder is closest to the load.) Examination of the other h

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two EDGs identified cracks similar in location and orientation to the one in which developed into a fracture on EDG-102. In addition, four of 24 connect-f ing rod bearings were found to contain cracks in the bearing shells.

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  • The EDGs are TDI Model DSR-48 diesels. These EDGs are the only DSR-48 diesels manufactured with a crankshaft assembly having an 11" crank pin diameter and 13" crankshaft diameter (11 x 13). On November 3, 1983, the applicant and its technicel consultant repor:.ed that the crankshaft failures were definitely caused by a basic design inadequacy. Independent analysis by the contractor established that the crankshaft was overstressed relative to industry standards, a conclusion supported by various considerations, including: industry-standard torsional analysis methods, detailed stress analyses, and actual torsional test results on EDG-101. Factors contributing to the bearing cracks were found to include unsupported, overhung bearing ends, excessive crack pin journal yawing, and the presence of large pores or voids in the aluminum bearing shells.

In 1974, the licensee contracted with TDI to purchase three EDGs for the Shoreham station. This was the first order received by TDI to provide an EDG for a commercial nuclear power station. Pre-operational testing of the engines at Shoreham commenced in late 1981. Each engine has eight cylinders in a straight line (straight-8). One of the Shoreham engines had been used by TDI to qualify the straight-8 series (R48) diesel engine for nuclear service.

Since testing began, the licensee has experienced several problems with the EDGs. Many component parts required reworking, redesign, and/or replacement.

At the present time, only two plants with operating licenses have TDI engines installed. One is San Onofre Unit 1 which has been shut down since February 27, 1982 for seismic modifications. The other is Grand Gulf which is authorized for power only up to 5%. A third operating plant, Rancho Seco, is presently installing TDI engines to supplement the existing non-TDI engines.

Grand Gulf has also experienced several problems with TDI engines. In 1981, preoperational testing of two V-16 engines at Grand Gulf commenced. These engines represent the first V-16 units ordered from TDI; one of the Grand Gulf engines was used to qualify the entire TDI V-16 line of machines for nuclear applications.

There has been a total of 57 TDI engines ordered for 16 nuclear power plant sites in the United States. A list of these sites is shown in Table 1. Only San Onofre Unit 1, Grand Gulf, and Shoreham have any significant equipment run time; therefore, the experience base of TDI units in United States nuclear service is limited.

Cause or Causes - The large number of failures together with the inspection history of TDI described below, indicate that quality assurance problems exist at TDI. l Actions Taken to Prevent Recurrence Long Island Lighting Company - The licensee has replaced the three 11 x 13 crankshaft assemblies with the 12 x 13 crankshaft assemblies like those report-edly installed in all other DSR-48 diesels. In addition, the connecting rod bearings were replaced with bearings designed to accommodate the new 12" crank pin diameter and to address the factors which caused the earlier bearings to develop cracks.

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Table 1 Nuclear Plants with Transamerica Delaval, Inc.

Diesel Generators Engine Site Licensee Location Model No.

Bellefonte Tennessee Valley Authority Jackson County, AL DSRV 16 Catawaba Duke. Power Co. York County, SC DSRV 16 Comanche Peak Texas Utilities Generating Co. Somervelle County, TX DSRV 16 Grand Gulf Mississippi Power & Light Co. Claiborne County, MS DSRV 16 Harris Carolina Power & Light Co. Wake & Chatham DSRV 16 Counties, NC Hartsville* Tennessee Va' ley Authority Trousdale & Smith DSRV 16 Counties, TN Midland Consumers Power Co. Midland County, MI DSRV 12 Perry Cleveland Electric Illuminating Lake County, OH DSRV 16 Co.

Phipps Bend

  • Tennessee Valley Authority Hawkins County, TN DSRV 16 Rancho Seco Sacramento Municipal Sacramento County, CA DSR 48 Utility District River Bend ** Gulf States Utilities West Feliciana DSR 48 Parish, LA San Onofre Southern California San Diego County, CA DSRV 20 Edison Co.

Shoreham Long Island Lighting Co. Suffolk County, NY DSR 48 Vogtle Georgia Power Co. Burke County, GA DSRV 16 WPPSS 1 Washington Public Power Benton County, WA DSRV 16 Supply System WPPSS 4* Washington Public Power Benton County, WA DSRV 16 Supply System

  • Project delayed or cancelled.
    • River Bend Unit 2 has been cancelled.

Note: Of the plants listed above, only San Onofre Unit 1, Rancho Seco, and Grand Gulf have received operating licenses.

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The applicant still intends to apply for a license to operate the Shoreham )

facility with the TDI diesel generators. However, as part of a long-term .

solution for the TDI diesel problems, the applicant has recently placed purchase orders for three diesel generators from Colt Industries. It is understood that

, the applicant intends to ultimately replace the TDI diesels with Colt diesels.

j Delivery of the Colt diesels is scheduled for the fall of-1985 which coincides

! with the completion of a new diesel generator building that is currently under l construction.

Additional actions will also be required in conjunction with the other actions described below.

1 Other Licensees - By letter dated December 23, 1983, the NRC staff was informed j that a TDI diesel engine owners group has been formed to address the EDG reliability issue.

i NRC - The staff continues to gather information regarding problems concerning TDI units, reviewing specifics of the problems, and developing a course of action to assure that the affected plants have reliable EDG capability.

! The NRC Region IV Vendor Inspection Branch performed inspections of the TDI facility in Oakland, California during July, September, and October 1983.

j These inspections were performed at the request of Region I (Region I has j responsibility for inspection activities at the Shoreham facility) and in response to allegations of irregularities in the quality assurance program.

Several potential nonconformances with NRC requirements were found during the

! July 1983 inspections. During the September and October 1983 inspections, the

! staff identified conditions which indicate that portions of the TDI quality l assurance program may not have been carried out in accordance with the pro-

! visions of 10 CFR 50, Appendix B.

i The staff has met with the applicant for Shoreham and the licensee for Grand l Gulf to discuss the failures to date, the results of the Shoreham investigation, and the actions to be taken to recover from the failures. The staff has also developed several lists of questions that it feels need to be addressed as part of the TDI engine evaluations. One list, which has been sent to all TDI diesel owners, requested specific information about each engine. Another was sent to TDI on December 1, 1983, requesting information about the design develop-ment history of various parts of TDI machines. Delaval responded on December 16, 1983.

On January 15, 1984, a special NRC project group was formed to coordinate the overall NRC review of TDI diesel generators.- Their primary responsibility is to evaluate the overall qualification of TDI diesel generators for nuclear service. Pacific Northwest Laboratory has been chosen to assist the staff in assessing and evaluating the corrective action plans being submitted by utilities possessing TDI diesel generators.

I The staff held a meeting on January 26, 1984 with seafor utility executives representing each of the applicants listed in Table 1. The staff informed them of its concerns regarding the breakdown in quality assurance in the TDI manufacturing facility and emphasized the significance of the widespread operating problems to date with TDI engines.

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r The staff believes that before additional licensing action is taken to authorize the operation of a nuclear power plant with TDI engines, these issues, relating to quality assurance, operating experience, and the ability of the machines to reliably perform their intended function, must be addressed.

On August 30, 1983, the NRC issued Inspection and Enforcement Information Notice No. 83-58 to licensees to infctm them of the Shoreham event (Ref. 1).

Previous to the Shoreham event, the NRC issued Information Notice No. 83-51 to licensees to inform them of various diesel generator problems (Ref. 2).

Further reports will be made as appropriate.

FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)

The NRC is reviewing events reported by these licensees during the fourth calendar quarter of 1983. As of the date of this report, the NRC had not determined that any events were abnormal occurrences.

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

There are currently more than 8,000 NRC nuclear materal 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 cateaory from licensees such as radiographers, medical institutions, and byproduct material users.

The NRC is reviewing events reported by these licensees during the fourth calendar quarter of 1983. As of the date of this report, the NRC had deter-mined that the following were abnormal occurrences.

83-16 Overexposure of a Radiographer The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see General Criterion 3) of this report notes that major deficiencies in design, construction, use of, or management controls for licensed facilities or material can be considered an abnormal occurrence. since the licensee had not installed a radiation detector as required by the NRC, a radiographer received a serious overexposure to his right thumb.

Date and Place - On January 9, 1984, the NRC Region I office was informed that on December 20, 1983, a radiographer working at Pittsburgh Testing Laboratory in Pittsburgh, Pennsylvania, received an estimated 3400 rem to his right thumb and 2.9 rem to the whole body while performing radiography with a state (Penn-sylvania) regulated x ray unit. During the same calendar quarter, the radio-grapher also received exposure from NRC licensed material resulting in a combined whole body exposure of 3.1 rem for the quarter. Both the extremity and whole body doses are in excess of applicable 10 CFR 920 quarterly restrictions.

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Nature and Probable Consequences - During November 1982, the state-regulated x-ray unit was replaced. At that time, the interlock on the room door was disconnected, and never reconnected. On December 20, 1983, the radiographer turned on the x-ray unit to allow it to warm up prior to making his first exposure. He later entered the radiography room to set his film and make final adjustments in the position of the piece to be radiographed. This involved localizing the beam center with a plumb-bob, which had to be held under the beam port with the radiographer's thumb. There were no indicators inside the room which showed when the x-ray beam was on. The radiographer realized that he had been exposed when he returned to the console to start the exposure and found that the beam was already on. It is estimated that the radiographer's thumb was in the beam port for about 5 seconds.

The licensee reported the incident to, and it was investigated by, the Commonwealth of Pennsylvania because the source of the overexposure (the x-ray tube) is regulated by the state. The state's reconstruction of the event, using ionization chambers, indicated that the radiographer's right thumb received an estimated 3400 rem, and that the whole-body dose was about 5-10 rem. The radiographer's film badge showed a dose of 2.9 rem for the month of the incident, and, including the two previous month's exposure, was 3.1 rem for the fourth quarter of 1983.

As discussed later, the NRC Region I also performed an inspection. The room used for x-ray radiography is also used for radiography with NRC licensed materials. As such, the room was supposed to have been equipped with a radiation-sensitive visible and audible alarm. It was found that such an alarm had never been installed.

The exposure to the radiographer's right thumb has resulted in erythema and blistering. At the time of the NRC inspection, the radiographer had not yet been seen by a medical specialist in radiation injuries, and, consequently, additional medical information is unavailable. NRC Region I has urged the licensee to obtain a consultant in radiation injuries.

Cause or Causes - The principal causes of the incident were the failure of the licensee to install the NRC-required radiation alarm and the absence of an interlock on the room door. A contributing cause was the absence of a beam-status indicator inside the room.

Actions Taken to Prevent Recurrence Licensee - The interlock on the room door was immediately reconnected. The radiation alarm required by the NRC has been ordered, and radiography with NRC-licensed materials will not be resumed in the room until the alarm is in place.

State Agency - Ine Commonwealth of Pennsylvania investigated the incident and found that the interlack on the room door was disconnected. The Agency also reconstructed the event to estimate the amount of radiation received by the radiographer. The Agency is taking appropriate actions with the licensee.

NRC - Since NRC licensed material was also involved in the radiographer's whole body overexposure for the calendar quarter, and since the licensee had a 6

history of-several violations of NRC requirements, the NRC also performed an investigation. The NRC confirmed the State Agency's findings and found that the NRC required radiation alarm had not been installed. Had the alarm been installed, the radiographer would have known that the x-ray beam was on before he attempted to adjust it.

The NRC Region I office held an enforcement conference with.the license'e on January 31, 1984. NRC Region I expressed their serious concern with the licensee management's procedures and controls for the use of radiography 4 devices, as evidenced not only with the immediate violations, but also with the numerous violations of requirements found during several previous inspec-tions. On March 2, 1984, the NRC sent to the licensee a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $8000, which the licensee subsequently paid.

This incident is closed for purposes of this report.

AGREEMENT STATE LICENSEES Procedures have been developed for the Agreement States to screen unscheduled incidents or events using the same criteria as the NRC (see Appendix A) and report the events to the NRC for inclusion in this report. During the fourth calendar quarter of 1983, one of the Agreement States reported the following abnormal occurrence to the NRC.

AS83-10 Overexposure of a Radiographer Appendix A (Example 1 of "For All Licensees") of this report notes that an exposure of the feet, ankles, hands or forearms of any individual to 375 rems or more of radiation can be considered an abnormal occurrence.

Date and Place - On July 8, 1983, a radiation safety officer (RS0) for X-Ray Inspection Company, a Louisiana industrial radiography licensee, located in

. Lafayette, Louisiana, received a significant overexposure to a finger while performing work at Conoco Oil Company in Westlake, Louisiana.

Nature and Probable Consequences - On July 8, 1983, the RSO received a call from two of his radiographers, indicating that a 108 curie iridium-192 source was stuck in the exposed position between the camera and the end of the source tube. He went to the radiography jobsite to evaluate the situation. Upon arrival, he visually' inspected the radiographic exposure device and noted that the outlet nipple on the camera was broken and the drive cable could be seen.

He did not'know the exact location of the source but knew that it was in the exposed position because his survey meter indicated a full up-scale reading.

He did not try to locate the source in the source tube but instead, picked up the source tube in both hands and tried to pull the source tube from the camera in an attempt to remove the source from the source tube.- During this effort, he apparently placed the index finger of his left hand very close to the iridium source.

About 13 days later, he experienced pain in his left index finger and eventu-ally, a blister developed. The individual's pocket dosimeter was discharged beyond its range during this retrieval, and since he was not wearing a whole-body TLD badge, it was quite difficult to establish a.whole-body dose.

7

However, from the reenactment, it was estimated that the whole-body dose was approximately 3 rems. From the clinical indications, the dose to the finger was estimated at between 4,000 and 8,000 rads. The individual is receiving medical treatment for this injury.

Cause or Causes - The RSO was well aware of the mistakes which had been made during the source retrieval but could not explain why he had taken such action.

The only reason offered was that the customer was rushing him to get the job completed so that production would not be hindered.

Actions Taken to Prevent Recurrence Licensee - The licensee reinstructed the RSO regarding the proper procedures to be used.

Louisiana Nuclear Energy Division - Appropriate violations have been cited for allowing excessive exposure to the individual and for the individual's failure to use the appropriate personnel monitoring devices.

This incident is closed for purposes of this report.

8

REFERENCES

1. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 83-58, "Transamerica Delaval Diesel Generator Crankshaft Failure," August 30, 1983.*
2. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 83-51, " Diesel Generator Events," August 5, 1983.*
  • Available in NRC Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).

9 i iii gl

~

3

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APPENDIX A ABNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormal occurrence determinations were set forth in an NRC policy statement published in the FEDERAL REGISTER on February 24, 1977 (Vol. 43, No. 37, pages 10950-10952).

Events involving 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:

For All Licensees

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

or equivalent exposures from internal sources.

2. An exposure to an individual in an unrestricted area such that the whole-body dose received exceeds 0.5 rem in one calendar year (10 CFR S20.105(a)).
3. The release of radioactive material to an unrestricted area in concentra-tions which, if averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 500 times the regulatory limit of Appendix B, Table II, 10 CFR S20 (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 regulatory limit (10 CFR 671.36(a)).
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.

11

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 facili-ties (generic incidents), which create major safety concern.

For Commercial Nuclear Power Plants

1. Exceeding a safety limit of license Technical Specifications (10 CFR

$50.36(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 S100 guidelines could result from a postulated transient or accident (e.g., loss of emer-gency core cooling system, loss of control rod systen).
4. Discovery of a major condition not specifically considered in the Safety Analysis Report (SAR) or Technical Specifications that requires immediate remedial action.
5. Personnel error or procedural deficiencies which result in loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR S100 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 Licenses l

1. A safety limit of license Technical Specifications is exceeded and a 1 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.

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i APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the October through December 1983, period, the NRC, NRC licensees, Agreement States, Agreement State licensees, and other involved parties, such as reactor vendors and architects and engineers, continued with the implementa-

tion of actions necessary to prevent recurrence of previously reported abnormal l occurrences. The referenced Congressional abnormal occurrence reports below provide the initial and any updating information on the abnormal occurrences 3 discussed. These occurrences not now considered closed will be discussed in l subsequent reports in the series.

NUCLEAR POWER PLANTS i 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 subsequent reports in this series, i.e., NUREG-0090, Vol. 2, No. 2; Vol. 2, i No. 3; Vol. 2, No. 4; Vol. 3, No. 1; Vol. 3, No. 2; Vol 3., No. 3; Vol. 3,
No. 4; Vol. 4, No. 1; Vol. 4, No. 2; Vol. 4, No. 3; Vol. 4, No. 4; Vol. 5,

! No. 1; Vol. 5, No. 2; Vol. 5, No. 3; Vol. 5, No. 4; Vol. 6, No. 1; Vol. 6, No. 2; and Vol. 6, No. 3. It is further updated as follows.  ;

i Reactor Building Entries  ;

1

During the fourth calendar quarter of 1983, 15 entries were made into contain-ment. There have been a total of 312 entries since the March 28, 1979 accident.

j Major activities included sampling the reactor coolant drain tank and the l

retrieval of three core debris samples.

EPICOR-II/ Submerged Domineralizer System (SDS) Proces h The EPICOR-II system processed approximately 71,000 gallons of water during the fourth quarter of 1983. The SDS processed approximately 66,000 gallons of water during the same time period.

EPICOR-II/Prefilter and SOS Liner Shipments A total of 30 EPICOR-II domineralizers were shipped from the TMI site to Hanford, Washington. As of December 16, 1983, all EPICOR dominaralizers

! meeting present disposal criteria and that would exceed Class A Criteria (10 CFR 6, effective December 27,1983) have been shipped for disposal. One SDS liner was also shipped to the Hanford facility.

Spent Fuel Pool "A" Refurbishment Work continued on the refurbishment of spent fuel pool "A". This involves the

! removal of concrete shield blocks, tanks from the upper and lower tank farm, and support steel. The use of the "A" spent fuel pool will be required for the transfer and temporary storage of fuel and 6:.bris from the damaged reactor core.

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s Auxiliary and Fuel Handling Building Activities Work on the expansion of the 328 ft, elevation decontamination facility continued during this quarter. Partial operation of the facility has begun.

Full operation should occur after the receipt of additional equipment com-ponents. Decontamination activities in the auxiliary and fuel handling building consisted of some surface scabbling and cubicle decontamination by hydrolazing.

Reactor Building Polar Crane On November 18, 1983, the staff approved the licensee's safety evaluation for the refurbishment and use of the Reactor Building Polar Crane. The crane had been the subject of allegations made by several GPUNC and contractor (Bechtel) employees relating to mismanagement, NRC/ licensee collusion, unsafe modifica-tions, and harassmant.

Sonic Core Topographical Model A computer generated map of the core void was completed from sonic ...easurements which were obtained inside the reactor vessel in August and September 1983. A scale, plastic model of the damaged core was also constructed from the sonic data. Based on the sonic measurements, the cavity volume in the damaged area of the core is 330 cubic feet or 26\ percent of the original core volume. The irregular cavity bottom is generally 5 feet below the top of the core region, with the deepest point, a narrow channel, being 6\ feet deep. Laterally, the cavity extends to the core forming walls in several areas.

Of the 177 fuel assemblies in the reactor, 42 assemblies around the core perimeter exhibit some continuous vertical development through the void region.

The cross sections of 23 of these standing assemblies were less than 50% of their fuel pins, and 2 assemblies appear to be relatively intact. The sonic plot showed that fuel assembly segments, typically 2 to 10 inches long are randomly attached to the underside of the plenum. The top 2 to 4 feet of several assemblies on the west side of the core overhang the void. In several areas where the core forming wall was exposed, the sonic device mapped the 3/4 inch thick stainless steel plates which form the perimeter of the core.

On the east side of the core, one area of the core forming wall appears to be bowed outward by 2 inches.

The-sonic topographical data is being evaluated and will be useful in planning for plenum and fuel removal. The data supplements the previously obtained closed circuit television tapes of the void and at the present stage of dis-assembly and defueling planning does not alter the existing concepts for future work.

Advisory Panel On November 29, 1983, Arthur E. Morris, Mayor of Lancaster, Pennsylvania, was appointed Chairman of the Advisory Panel for the Decontamination of the Three Mile Island Nuclear Station, Unit 2, by NRC Chairman Nunzio J. Palladino. The Advisory Panel obtains local citizen views and provides the Commission with valuable counsel on the actions to be proposed and taken by the NRC regarding cleanup of the damaged reactor.

14

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On December 8, 1983, the Advisory Panel held a meeting in Harrisburg, Pennsyl-vania. Representatives from the NRC, EPA and DOE provided an update of their respective agency's activities relative to the cleanup effort.

The Panel was given a presentation by GPUN personnel which provided both an overview of the licensee's safety evaluation and the sequence of activities associated with the planned reactor pressure vessel head lift.

Dr. Bernard J. Snyder, Director, NRC TMI Program Office (TMIPO), presented the Panel with copies of a TMIP0 fact sheet and resumes of the TMIPO staff.

Mr. B. K. Kanga, Director TMI-2, GPUN, answered questions posed by the Panel on the issue of funding. Currently the licensee estimates approximately

$75 million in funds for TMI-2 activities will be available during calendar year 1984. However, there is still a fair degree of uncertainty associated with the 1984 funding levels. The Advisory Panel passed a resolution that states that the Panel is against the consideration of the restart of TMI Unit I until committed funding for the cleanup of the damaged Unit 2 reactor is in place. The vote of this resolution was five in favor, one opposed, and one abstention.

Further reports will be made as appropriate.

A A A A A A

] 83-5 Large Diameter Pipe Cracking in Boiling Water Reactors (BWRs),

This abnormal occurrence was originally reported in NUREG-0090, Vol. 6, No. 3,

" Report to Congress on Abnormal Occurrences: July-September 1983." It is further updated as follows.

As stated in the previous report, NRC Orders were issued on August 26, 1983 to the four licensees of five plants which had not yet begun inspections of piping. The Orders confirmed accelerated inspection schedules which had been developed by the licensees in a meeting with the NRC. The four licensees and their respective plants are as follows: (1) Tennessee Valley Authority (Browns Ferry Unit 3, located in Limestone County, AL), (2) Carolina Power &

Light Company (Brunswick Unit 2, located in Brunswick County, NC), (3) Boston Edison Company (Pilgrim Unit 1, located in Plymouth County, MA), and (4) Common-wealth Edison Company (Dresden Unit 3, located in Grundy County, IL; and Quad Cities Unit 2, located in Rock Island County, IL).

The inspections have now either been completed, or, in the case of the Pilgrim plant, the licensee decided to undertake a pipe replacement program rather than repair. The results for the five plants are shown in Table B-1.

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L.

j: Table B-1 Inspection Results and Repairs for Plants Subject to

August 26, 1983 NRC Order

! No. of L Welds in No. Cracks No.

l Plant Program Inspected Detected Repaired *

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Browns Ferry Unit 3 191 191 1 1 l Brunswick Unit 2 131 131 24 9 i Dresden Unit 3 337 240 64 61 Pilgrim Unit 1 REPLACING PIPE

Quad Cities Unit 2 225 225 23 10 4.

As discussed in the previous report, Georgia Power Company, licensee for Hatch Unit 2, located in Appling County, Georgia, has also decided to initiate a l pipe replacement program. -This plant, even though it has only a relatively

brief operating history, showed extensive cracking. The licensee shut down l- the plant on January 13, 1984 to begin the replacement program. Similarly, i Northern States Power, licensee for Monticello, shut down the plant on

! February 3, 1984, to replace recirculation system piping. The licensee i anticipates an' approximate 30. week outage, i

j Further reports'will be made as appropriate.

i 1

! "Not all-cracks detected will necessarily require repairs. Cracks' identified i by ultrasonic' testing are evaluated to determine the size and depth. Some

cracks'are determined to be sufficiently minor that they do not require

! repair. These minor cracks will then be tested again in the future to deter-l mine if there have been any changes-in size or configuration.

t I

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f APPENDIX C OTHER EVENTS OF INTEREST I

The following events are described below because they may possibly be perceived by the public to be of public health significance. The events 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. Contamination Due to Failed Fuel By letter dated November 7, 1983, the Sierra Club requested that the NRC halt dry cask shipments of spent fuel, including shipments from West Valley, New York and the Cooper Nuclear Station in Nebraska, until appropriate analyses of a May 1980 incident involving possible oxidation of spent fuel at Battelle Columbus Laboratories (Battelle) of West Jefferson, Ohio, concerning a Model No. NFS-4 shipping cask are performed and factored into licensing requirements (Ref. C-1).

The Sierra Club letter was treated as a request for action under 10 CFR S2.206.

The details associated with the Battelle incident are as follows.

On May 1, 1980, an irradiated fuel assembly having known, severe fuel cladding failures was shipped from the Haddam Neck facility (operated by Connecticut Yankee Atomic Power Company and located in Middlesex County, Connecticut) to Battelle for postirradiation examination. The fuel assembly was shipped dry (normal atmosphere air) inside a Model No. NFS-4 shipping cask. The cask arrived at Battelle on May 2, 1980. Before immersing the cask in the fuel pool, the fuel assembly was cooled by slowly filling the cask cavity with water while venting the cask to the hot cell through a connected hose. Steam was initially discharged from the hose indicating that the assembly was thermally hotter than fuel previously handled. A high radiation level alarm was also activated within the hot cell.

I

, Following cavity flooding, the cask was lowered into the pool and the cask head removed. A dark cloud emanated from the cask, spread through the pool water, rose to the surface, and spread contamination throughout the high-bay cask handling area. The event caused " chirpers" (radiation detectors) worn by the operators to respond and caused a radiation level of about 200 mr/hr three feet above the water level, as measured by a portable instrument. Floor smears showed that contamination had deposited on the room surfaces. Five personnel were working in the area at the time. Respirators were not being worn since normally they would not be needed. The personnel continued work until the fuel assembly was removed from the cask and placed in a pool storage rack (about one hour after cask lid removal). Subsequent entries into the pool area were made by the personnel wearing respirators.

Nasal swabs, film badge measurements, urinalyses, fecal samples, and in vivo counts were obtained from the five individuals involved in the incident; none sindicated significant doses to any of the individuals. The highest film badge measurement was 220 mr gamma.

l Continuous air monitors were in operation during and after the cask opening.

[ The highest air activity detected was for a period of about 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> shortly 17

i i

i after the cask opening, in which both alpha and beta concentrations were about 20 times maximum permissible concentration levels. The concentration of radioactivity in the fuel pool water reached a peak of about 400 and 50 times the concentration limits imposed by a license condition for beta and alpha activities respectively. These concentrations were reduced to the permitted levels over a period of weeks by circulating the water through the installed ion exchange resin beds. The decontamination of the surface areas and equipment in the pool areas required significant labor and supplies. Before cleanup began, contamination levels were up to 150 times and 1000 times the licensee's control limits for alpha and beta gamma activities, respectively. There was no release of radioactive material from the building. Normal work activities in the laboratory were not interrupted by the decontamination efforts.

Analysis showed that the contaminating material was fuel and fission products in the form of very fine particles. The radioactive particulate material, which was capable of becoming airborne, represented a radiological source term which had not been anticipated.

The specific mechanisms by which these particles were produced and by which they were transported are not precisely known. However, it is believed that the particulates were primarily caused by the oxidation of UO2 fuel. A sig-nificant amount of fuel was exposed to the air environment of the cask since the fuel assembly was known to contain several fuel rods with severe cladding splits (thereby exposing the fuel pellets inside). Some irradiated fuel material in the form of UO 2and higher oxides could have been released from fafled fuel rods into the confines of the shipping cask during transportation and handling, subsequently being released tc the fuel pool when the cask lid was removed.

Two important parameters affecting oxidation rate are temperature and time.

The amount of oxidation increases with higher fuel temperatures and the length of time the fuel is exposed to an oxygen-rich environment (such as normal air).

The residual heat content of the fuel assembly, and consequently the fuel temperatures reached, were considerably higher than expected. Original calcula-tions performed by Connecticut Yankee indicated a decay heat content of 2.09 kW.

After the incident, Battelle made their own calculations (based on a later industry standard) which indicated a decay heat rating of greater than 2.50 kW.

Since the NRC had imposed a restriction to 2.5 kW on the shipping cask, Battelle notified Connecticut Yankee. The latter recalculated and found that the decay heat rating was actually as much as 3.50 kW; these calculations were verified <

by the cask licensee.

! In addition, the fuel was exposed to air longer than anticipated. The dry fuel assembly had been in the cask for six days prior to its removal at i

Battelle. The fuel assembly was loaded into the cask on April 26, 1980 at the Haddam Neck facility. However, departure was delayed until May 1, 1980 because of problems in decontaminating the cask walls to below permissible l levels, and excessive radiation levels emanating from the cask which' required placing external shielding on the container transport vehicle.

18

. - - . _ . -. _ ~ _ __ _.__- . _ . _ _ - . _ . _ _ _ _ .

V l . Corrective actions taken by Battelle included (1) performing a comprehensive

+'

. review of hot cell laboratory receiving procedures and modifying or supplement-

~

{ ing them as necessary, (2) at least until the review was completed, requiring the use of respiratory protection during cask unloading procedures involving failed fuel assemblies,~(3) requiring approval of the hot cell laboratory

i. operations manager of case reviews submitted to the Radiological Safety Com-j Cittee, and (4) improving communications with shippers regarding potential i

hazards. associated with shipments.

i The circumstances associated with the incident, and subsequent activities to decontaminate the cask, were included in a routine inspection at Battelle per-formed by NRC Region III on September 22-26 and November 12, 1980. The inspec-

' tion report was forwarded to Battelle on December 8, 1980 together with a Notice
of Violation (Ref. C-2). The violations involved (1) an employee receiving an j - cverexposure to the hand during Batte11e's preparation of the NFS-4 cask for i

reuse, and (2) radioactivity in the fuel storage pool exceeding license i conditions.

The NRC amended the Certificate of Compliance for the Model No. NFS-4 cask to

preclude shipment of failed fuel assemblies (U0, pellets) which are oxidized.
Shipment of other failed fuel was authorized in a dry inert atmosphere; in the

( absence of oxygen, oxidation of any exposed 002would be suppressed.

1 L The radioactivity releases into the air of the high-bay area and resultant i laboratory personnel exposures, associated with the May 2, 1980 incident at
Battelle, were small. In addition, there was no release of radioactive material i from the building. Therefore, there was no impact on public health or safety and the event is not considered reportable as an abnormal occurence.

{

In view of the 10 CFR $2.206 action request mentioned above, the NRC staff [

reevaluated the Battelle incident. The staff concluded that while fuel oxida-i tion does not'significantly alter the risks of transport, it could increase' l f the risks of personnel exposure during receiving and handling operations, i Therefore, all Certificates of Compliance were revised to require spent fuel j casks to be inerted for shipment to prevent handling problems from oxidized-  ;

i- fuel at facilities receiving spent fuel. In addition, shipments of known or '

l suspected failed fuel assemblies (fuel rods) may not be made unless each fuel j assembly is appropriately canned for shipment. Based on the staff reevaluation, i and the revisions made to the Certificates of Compliance, the Sierra Club's request was denied on April 13, 1984, by the Director, NRC Office of Nuclear Material Safety and Safeguards (Ref. C-3). The decision will constitute the

! final' action of the Commission 25 days after the date of issuance of the decision-  :

unless the Commission, on its own motion, institutes a review of the decision l within that time.

I i ~ 2.. Failed Fuel Assemblies i During Cycle 5 of-the Millstone Unit 2 reload core, the licensee (Northeast I i Nuclear Energy. Company)'noted elevated levels of radioactive iodine and other

fission products in_the reactor coolant. Millstone Unit 2 is a Combustion 4 Engineering (CE)~ designed plant, utilizing a pressurized water reactor, and is located in New London County,-Connecticut, t

I- 19

, L m ., ._ _. . _--_-,-. _.- - , . - _ _ - , . ~ - , _ , _ . - - - - , _ , . . ,-_._..m_,r

-By the end of Cycle 5 operation, the primary system activity was about two

. percent of the plant's Technical Specification limit; this was indicative of about 10 to 30 fuel pin failures. The plant was shut down on May 28, 1983, for refueling and maintenance. The licensee established a fuel pin failure investigation program. Fuel sipping (analysis of the fuel assembly for leak-age of fission products) was conducted on the entire core and 26 fuel assem-blies were identified which had one or more failed fuel pins. Five of the assemblies were supplied by one vendor and were scheduled for discharge at the end of Cycle 5; these assemblies had been irradiated for several cycles.

Twenty-one of the assemblies were supplied by another vendor and had been scheduled for reinsertion for Cycle 6; these assemblies had seen no more than a few cycles of operation.

Ultrasonic inspections showed that there were 32 failed fuel pins in the 26 fuel assemblies. The licensee evaluated a number of possible failure mechanisms and concluded that the failures apparently resulted from multiple sources, none of which were indicative of a situation that may lead to con-tinued serious degradation of the fuel cladding. For example, there was evidence of debris induced wear of the cladding and one case of confirmed grid spring / fuel rod fretting. The probable cause of the latter was a damaged cell, most likely related to either fuel manufacturing or handling. To reduce the possibility of debris induced wear, the licensee performed an extensive cleanup of the primary system.

Visual inspections revealed 15 fuel assemblies to have broken holddown springs.

The probable cause was attributed to system flow induced vibration, near the core periphery, leading to fatigue failure of the springs. The licensee's analysis concluded that although broken, the springs remained functional and the assemblies could continue in operation. Future new fuel assemblies will have redesigned springs.

Further inspections revealed two fuel assemblies with struc.tural damage; one of the two also had a broken holddown spring. However, the requirements for structural integrity, such as strength and loading capability, were still met for normal as well as for accident conditions by these fuel assemblies. The cause was attributed to insufficient gap clearances between the assembly structurals and the fuel alignment plate. Additional clearances will be incorporated into new fuel assemblies.

The problems described above necessitated a revision to the licensee's origi-nally planned reload core for Cycle 6. A combination of new and previously discharged fuel assemblies were used to replace the leaking and two damaged i fuel assemblies. Nine fuel assemblies, each with a sin ie w broken holddown spring, were also used for Cycle 6 (the licensee decided that repair of these springs on the irradiated fuel assemblies would involve'a high risk of damaging fuel pins).

As part of the scheduled shutdown activities, the reactor vessel internals were inspected. Damage was noted to both the thermal shield and the core barrel. The damage appeared similar to that experienced at St. Lucie Unit 1, another CE designed facility. The damage incurred at St. Lucie Unit 1 was described in Appendix C of a previous issue of these quarterly abnormal occur-rence reports (NUREG-0090, Vol. 6, No. 2). As described in that report, it'is 20

believed that the damage to the thermal shield was not a single event, but rather occurred over a period of time and was related to mechanical stress caused by flow induced vibrations. Also as described in the report, the use of a thermal shield is a design option in CE plants. The licensee for Mill-stone Unit 2 decided to operate the plant in the future without the thermal shield. The thermal shield was therefore removed and the minor damage to the core support barrel was repaired.

The Cycle 6 core reload changes, together with operation without the thermal shield, were submitted to the NRC for approval. NRC approval was granted on December 30, 1983 and the plant achieved criticality on January 5,1984.

The number of fuel pins which failed constituted less than 0.1% of the total number of fuel pins in the core. The resultant primary activity was only a small fraction of the Technical Specifications limitations. Therefore, there was no impact on public health or safety and the event is not considered report-able as an abnormal occurrence.

1 l

I 21

REFERENCES (FOR APPENDICES)

C-1 -Letter from M. Resnikoff, Sierra Club, to C. MacDonald, Chief, Transporta-tion Certification Branch, Division of Fuel Cycle and Material Safety, NRC Office of Nuclear Material Safety and Safeguards, November 7,1983.*

C-2 Letter from J. G. Keppler, Director, NRC Region III, to Dr. E. W. Unger, Director, Battelle Columbus Laboratories, forwarding an inspection report and Notice of Violation, Docket Nos.70-008, 30-5728, and 50-006, December 8, 1980.*

C-3 Letter from John G. Davis, Director, NRC Office of Nuclear Material Safety and Safeguards, forwarding " Director's Decision (DD-84-9) Under 10 CFR S2.206," to Dr. Marvin Resnikoff, Sierra Club, April 13, 1984.*

i i

l

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

23 L

- a U 5j NUCJ kJ RE1ULnTORY COMMISSION' BIBLIOGRAPHIC DATA SHEET NUREG-0090, V . 6. No.4 4 1gTLE AN D SUBT8TLE IAdd Vorume No., ot apprmvoser) 2. (Leme blerk)

R port o Congress on Abnormal Occurrences -

Octobe ecember 1983 3. RECIPIENT'S A SSION NO.

7. AUTHORISI S. DATE REl'Old COMPLETED j M ON TH l YEAR M[ay 1984
3. PERFORMING OHG JilAllON N AME AND MAILING ADDFiESS I/actum Isa Co*/ OATE flEPORT ISSUED U.S. Nuclear ~ egulatory Commission Office for Ant sis and Evaluation of Operational - " [" MaY 1984 Data *f**"*"*'

W;chington, DC 555 gft,,,y,,,,

12. St'ONSORING ORGANilATION ME AND MAILING ADDRESS (lactue lea Co*/
10. PROJECT / TASK / WORK UNIT NO.

U.'S. Nuclear Regulat Commission Office for Analysis an Evaluation of Operation it. CONTRACT NO Data W2dhington, DC 20555 13 1YPE OF REPQHT / PE 00 COVE RE D (tectuove deres/

Ouarterly ctober-December 1983

15. SUPPLFMEN TAHY NOTES 14 Iteme n/st4)
16. A~S1H ACT CJ0 wonts or less}

Section 208 of the Energy Reorganizat' Act of 1974 identifies an abnormal occurrence as an unscheduled inciden o vent which the Nuclear Regulatory Commission determines to be signifi nt f the standpoint of public health or safety and requires a quarterly. eport o such events to be made to Congress. This report covers the criod Octo r 1 to December 31, 1983.

During the report per.tod, there s one abnormal ecurrence at the nuclear power plants licensed by the NR to operate. The tem involved generic problems portaining to a speci c manufacturer's am gency diesel generators.

There was one abnormal occurro co for the other NRC. censees. The item involved an overexposure of a adiographe r.' There was ne abnormal occurrence reported by an Agreement Sta . The item involved an o rexposure to a rcdiographer. ,/ ,

come previou reported The cbnormsl report also contait)s in ormation updating /"-

occurrences. f

17. KEY WORDS ANO f OCUME NT AN ALYSIS 1 74 DE SC rip TO RS EDG; Diesel Generator; S orcham; Radiographer overexposure; Trnnsamerica,Delaval; Re iability; Quality Assurance; Radiography; Failed FuelP Radioactiv Contamination f

.)

titi IDENilF fEHS OPEN ENDE D TE RMS 18 AV AILABILITY ST ATEMENT 19 SECURITY CLASS (Thss report / 21 NO OF PAGE3 J nclammifIod

20. iE CURI TY CL ASS (Thes osvel 22 PRICE Unlimited -

Uncina.4r4.2--'^ s

.pcc sonu us o m ,

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